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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Spent Nuclear Fuel and Reprocessing Waste Inventory

This report provides information on the inventory of spent nuclear fuel (SNF) in the United States located at Nuclear Power Reactor (NPR) and Independent Spent Fuel Storage Installation (ISFSI) sites, as well as SNF and reprocessing waste located at U.S. Department of Energy (DOE) sites and other research and development (R&D) centers as of the end of calendar year 2024. Actual quantitative values for current inventories are provided along with inventory forecasts derived from examining different future nuclear power generation scenarios, based on information available and assumptions made at the time the scenarios were developed. The report also includes select information on the characteristics associated with the wastes examined (e.g., type, packaging, heat generation rate, decay curves).

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Temperature-Time Modeling of Spent Fuel Cladding in Dry Storage Casks

This report uses thermal and decay heat modeling to investigate spent fuel performance and the potential for cladding to anneal in dry storage conditions. Annealing is an important feature to investigate in SNF cladding because it has a direct relationship to cladding response in storage, transportation and disposal conditions. Annealed cladding may have lower strength than unannealed cladding, however its increased ductility would provide better protection against rupture in high strain rate situations such as severe accidents. These consequences are the driver for this work however are outside of the scope of this report. The modeling focused on three representative storage systems, the TN-32B, MAGNASTOR with TSC-37 canister, and NUHOMS AHSM with a 32PTH2 canister. This covers the vertical dual-purpose, vertical ventilated and horizontal ventilated casks respectively. Decay heat modeling using high and low enrichment assemblies was used to bound the decay heat curves that might be expected in dry storage. To bound the temperature relationship, the storage casks were modeled starting at the design basis heat loads with heat decaying through time. Although the results are bounding there is not an attempt to maximize conservatism, rather the intent to form a reasonable upper limit on temperature that will be broadly applicable to the U.S. cask fleet. This will allow materials testing to focus on relevant conditions for annealing that may affect the U.S. spent fuel inventory. The results show a clear dependence on heat load pattern in time in Figure S-1 and Figure S-2. This dependence is due to the different assembly decay heat curves for different assemblies in preferentially loaded casks. It shows the need for careful decay heat modeling when examining in service fuel temperatures that are less than the cask design basis heat load. The body of the report shows percent cladding cutoffs of 300°C and 350 °C as well. These results can be used to inform testing and conclusions about cladding performance through time and the potential for cladding annealing during dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Biohydrogen Generation from Un-Pretreated Spent Coffee Grounds with Clostridium Thermocellum

Coffee is among the world's most widely consumed beverages, generating 18 million wet tonnes of waste spent coffee grounds annually. Disposal of SCGs is complicated by their high moisture content, their recalcitrant chemical composition, and the presence of caffeine and other bioactive compounds. Dark fermentation is a promising technology for conversion of waste cellulosic biomass into renewable hydrogen, carbon dioxide, and volatile fatty acids. Typical SCGs have high moisture content, are pre-milled to fine particle sizes, and contain 8%-19% cellulose and up to 40% hemicellulose, making them an attractive feedstock for dark fermentation. In this study, we investigate biohydrogen production from unpretreated SCGs using Clostridium thermocellum strain KJC19-9, a cellulolytic bacterium engineered to co-utilize xylose, specifically examining the inhibitory effects of residual caffeine and high solids loadings on microbial growth and hydrogen production efficiency. While SCGs were initially resistant to both cell growth and biohydrogen production, a strategy to co-ferment with low concentrations of cellobiose dramatically enhanced process performance, reducing lag phase duration and enabling cellulosome production for efficient cellulose hydrolysis. This co-substrate approach generated up to 944 mL hydrogen per L reaction from 50 g/L of spent coffee grounds over 120 h (1.34 mol H2/mol carbohydrate), validating SCGs as a promising dark fermentation feedstock and underscoring the value of mixed substrate approaches in consolidated bioprocessing.

08 HYDROGEN↗
DOE OSTI2015

Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated k eff margin varies from 0.05 to almost 0.3 Δk eff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Spent Nuclear Fuel Vibration Integrity Study

The objective of this research is to collect dynamic experimental data on spent nuclear fuel (SNF) under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT), the hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The collected CIRFT data will be utilized to support ongoing spent fuel modeling activities, and support SNF transportation related licensing issues. Recent testing to understand the effects of hydride reorientation on SNF vibration integrity is also being evaluated. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance, Fuel structure contributes to the SNF system stiffness, There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response, and that data will also be presented.

bending fatigue↗
DOE OSTI2016

Spent Nuclear Fuel Transport Reliability Study

This conference paper was orignated and shorten from the following publisehd PTS documents: 1. Jy-An Wang, Hao Jiang, and Hong Wang, Dynamic Deformation Simulation of Spent Nuclear Fuel Assembly and CIRFT Deformation Sensor Stability Investigation, ORNL/SPR-2015/662, November 2015. 2. Jy-An Wang, Hong Wang, Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications, NUREG/CR-7198, ORNL/TM-2014/214, May 2015. 3. Jy-An Wang, Hong Wang, Hao Jiang, Yong Yan, Bruce Bevard, Spent Nuclear Fuel Vibration Integrity Study 16332, WM2016 Conference, March 6 10, 2016, Phoenix, Arizona.

bending fatigue tester↗
DOE OSTI2022

Considerations For The Management Of Spent Nuclear Fuel from Advanced Reactors

The intent of this presentation is to give some key considerations for the management of spent nuclear fuel from Advanced Reactors. It will be presented at the Management of Spent Fuel, Radioactive Waste and Decommissioning in SMRs or Advanced Reactor Technologies conference organized by OECD – NEA in Ottawa (Canada) on November 2022 in the session: Operational feedback on managing and disposal of existing waste streams and how some of these concepts can be applied in future endeavours, such as SMRs/Advanced Reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

LWR fuel assembly designs for the transmutation of LWR Spent Fuel TRU with FCM and UO{sub 2}-ThO{sub 2} Fuels

In this paper, transmutation of transuranic (TRU) nuclides from LWR spent fuels is studied by using LWR fuel assemblies which consist of UO{sub 2}-ThO{sub 2} fuel pins and FCM (Fully Ceramic Microencapsulated) fuel pins. TRU from LWR spent fuel is loaded in the kernels of the TRISO particle fuels of FCM fuel pins. In the FCM fuel pins, the TRISO particle fuels are distributed in SiC matrix having high thermal conductivity. The loading patterns of fuel pins and the fuel compositions are searched to have high transmutation rate and feasible neutronic parameters including pin power peaking, temperature reactivity coefficients, and cycle length. All studies are done only in fuel assembly calculation level. The results show that our fuel assembly designs have good transmutation performances without multi-recycling and without degradation of the safety-related neutronic parameters. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Recovery of minor actinides from spent fuel using TPEN-immobilized gels

A series of separation experiments was performed in order to study the recovery process for minor actinides (MAs), such as americium (Am) and curium (Cm), from the actual spent fuel by using an extraction chromatographic technique. N,N,N',N'-tetrakis-(4-propenyloxy-2-pyridylmethyl) ethylenediamine (TPPEN) is an N,N,N',N'-tetrakis (2-pyridylmethyl) ethylenediamine (TPEN) analogue consisting of an incorporated pyridine ring that acts as not only a ligand but also as a site for polymerization and crosslinking of the gel. The TPPEN and N-isopropylacrylamide (NIPA) were dissolved into dimethylformamide (DMF, Wako Co., Ltd.) and a silica beads polymer, and then TTPEN was immobilized chemically in a polymer gel (so called TPEN-gel). Mixed oxide (MOX) fuel, which was highly irradiated up to 119 GWD/MTM in the experimental fast reactor Joyo, was used as a reference spent fuel. First, uranium (U) and plutonium (Pu) were separated from the irradiated fuel using an ion-exchange method, and then, the platinum group elements were removed by CMPO to leave a mixed solution of MAs and lanthanides. The 3 mol% TPPEN-gel was packed with as an extraction column (CV: 1 ml) and then rinsed by 0.1 M NaNO{sub 3}(pH 4.0) for pH adjustment. After washing the column by 0.01 M NaNO{sub 3} (pH 4.0), Eu was detected and the recovery rate reached 93%. The MAs were then recovered by changing the eluent to 0.01 M NaNO{sub 3} (pH 2.0), and the recovery rate of Am was 48 %. The 10 mol% TPPEN-gel was used to improve adsorption coefficient of Am and a condition of eluent temperature was changed in order to confirm the temperature swing effect on TPEN-gel for MA. More than 90% Eu was detected in the eluent after washing with 0.01 M NaNO{sub 3} (pH 3.5) at 5 Celsius degrees. Americium was backwardly detected and eluted continuously during the same condition. After removal of Eu, the eluent temperature was changed to 32 Celsius degrees, then Am was detected (pH 3.0). Finally remained Am could be stripped from TPPEN-gel by changing the pH of the eluent to 2.0. These results These results prove that the proposed recovery process for MAs is a potential candidate for future reprocessing methods based on the extraction chromatographic technique. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Quantification of reactivity margin in BWR spent fuel pools

Boiling water reactor (BWR) spent fuel pool (SFP) criticality safety analyses typically use the in-core k{sub inf} criterion methodology to demonstrate margin to safety. This method models all the fuel in the SFP with a limiting lattice design at a peak reactivity just above a specific in-core k{sub inf} limit. To bound all the fuel in the pool and future fuel with reasonable margin, these limits are often set conservatively high compared to the peak reactivities of actual BWR bundle designs. When current regulatory expectations are met for the quantification of bias and uncertainty adders to the criticality analyses, some results are quantified that are close to the 0.95 in-rack reactivity limit required by 10CFR50.68. The small margins being calculated in the criticality safety analyses have resulted in concerns that SFPs have less margin to safety than was present in the past; however, inherent in the peak reactivity in-core k{sub inf} criterion methodology used to assess these margins is a very conservative model of the fuel. While it is known that all the fuel in a BWR SFP is typically burned well past its peak reactivity exposure, it has been unknown how much reactivity margin could be quantified if bundles were modeled more realistically by accounting for their actual discharge exposure profiles, operational void fractions, control histories, and depletion fuel temperatures. In this study, fuel bundles were depleted to their explicit nodal discharge exposure while accounting for actual operational conditions defined by GEH/GNF's core simulator code, PANACEA. The use of these depleted isotopics in the criticality safety analysis is commonly referred to as taking 'burn-up credit' in the analysis. This paper provides a quantification of the reactivity margin available from burn-up credit in two currently operating BWR spent fuel pools. The evaluation demonstrates that many of the bundles in these storage configurations are significantly below the in-rack reactivity limit, even with the necessary bias and uncertainty adders included. The utilization of BWR burnup credit appears to be an attractive solution to consider when evaluating options for reduced neutron absorber credit or for quantifying additional margin to the regulatory in-rack k{sub eff} limit. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

A study on the assessment criteria of DHC in PWR spent fuel

One of the major degradation mechanisms of PWR spent fuel cladding is delayed hydride cracking (DHC), which can activate within a relatively low temperature range. Its maximum crack growth rate is estimated to be within the range of 10{sup -7} m/s, at which the cladding can rupture in 1 to 2 years during dry storage. For this reason, it is essential to prevent DHC crack growth itself, but most of the existing DHC studies have been focused on the crack growth rate. In this regard, the values and characteristics of the threshold stress intensity factor (K{sub IH}) for PWR Zircaloy-4 cladding were described. Based on the limited existing test results and analysis, the experimentally measured K{sub IH} were compared with the predicted K{sub IH}. The comparison shows that the predicted values are much lower than the measured ones and both theoretical estimation and measured values of K{sub IH} increase as the temperature increases. The possibilities of degradation by DHC were also discussed during long-term dry storage. For a more reliable analysis, additional information on a pre-existing crack of spent fuel is needed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Preliminary Concept of Operations for the Spent Fuel Management System+ - 17134

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the US Department of Energy's Office of Nuclear Energy is tasked with identifying, planning, and conducting activities for developing interim storage and transportation capabilities in support of an integrated waste management system (IWMS). The IWMS is expected to transport Spent Nuclear Fuel (SNF) from a reactor and deliver it to a repository in a suitable container, with the possibility of interim storage. The system will be composed of multiple subsystems including possible interim storage facilities (ISFs), one or more repositories, facilities to package and/or repackage SNF, and transportation. The project team is analyzing options for an IWMS to identify viable options. To support analysis, the project team has developed a concept of operations framework for SNF related components of the potential integrated system and the interdependencies between those components. A goal of this work is to aid system analysts in developing consistent models across the project. The concept of operations framework will be updated periodically as new developments emerge. At a high level, SNF is expected to be transported from reactor sites to a repository. Initially, SNF is unloaded from reactors and placed in spent fuel pools (SFPs) for wet storage at utility sites. To maintain space in SFPs, SNF is placed in containers at reactor sites for storage and/or transportation only after the SNF has cooled enough to satisfy loading limits. Once transportation requirements are met, the SNF may be transported to an ISF, where it is stored until a repository is developed, or it is transported directly to a repository if one is available.While the high level operation of the system is straightforward, analysts must evaluate numerous alternatives, such as the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if needed), repackaging technology, types of containers used, repository design, component sizing, and timing of events. These alternatives arise due to technological, economic, or policy considerations. This paper provides an overview of the potential alternatives identified in the concept of operations framework at a conceptual level. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

The Pretreatment and Immobilization Test on The Greater Than Class C Spent Ion Exchange Resins from Taiwan Research Reactor - 17177

In this work, the experience on the removal of greater than Class C bead ion exchange resins from Taiwan Research Reactor spent fuel pool was reported. All of the spent ion exchange resins had been flushed, dewatered, and packed into lead-shield containers for 60 drums and pending for stabilization. The meta-kaolin type of geo-polymer was selected to immobilize the ion exchange resin. According to the 'Regulations on Final Disposal of Low Level Radioactive Waste and Safety Management of the Facilities' Act in Taiwan, the immobilization waste form shall meet seven testing items. Herein, several immobilize samples were prepared with the following amounts of ion exchange resin, 6, 12 and 18 wt.%. The compressive strength of the samples ranged from 2.9 to 18.2 MPa. The Cs and Sr ions were effectively trapped by ion exchange resins. After the resins incorporated with the geo-polymer matrix, the captured ions were expected difficult to leach out. However, it is found that at the higher contents of ion exchange resins in the immobilize samples exhibited poor water resistance due to the swelling effect. Several parameters such as the ratios of SiO{sub 2}/Na{sub 2}O and powder/solution also the blast furnace slag dosage were investigated to improve processibility and improve performance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

The new CASTOR{sup R} geo - A Comprehensive Solution for Transport and Storage of Spent Nuclear Fuel, MOX and Damaged Fuel - 17202

Dry interim storage has become a common solution for the disposal of spent fuel in recent years. However, in particular the complete defueling of NPP prior to decommissioning and dismantling will dramatically increase the demand worldwide especially for non-standard fuel. Here we present the new dry storage system by GNS for international markets with its capability to also store MOX and damaged spent fuel. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

A Classification of Commercial Nuclear Reactor Sites and the Subsequent Impacts on Preparing Spent Nuclear Fuel for Shipment - 17355

The US Department of Energy is conducting analyses to inform future decisions regarding an integrated waste management system for the acceptance of spent nuclear fuel (SNF) from commercial nuclear reactor sites. The ability to prepare SNF for shipment from reactor sites is likely to be restricted by ongoing operations at those sites. This report examines the extent to which those operations will restrict the reactor sites' availability to prepare SNF for shipment. Specifically, the analysis focuses on the impact and frequency of refueling operations within the commercial reactor sites in the US. In this report, a characterization of reactor sites into families is described. This characterization is based on the number of operating reactors on site, the length of refueling cycles for those reactors, and the configuration of spent fuel pools on site. Each reactor family is analyzed to estimate the amount of time for which each reactor site in the family would be available to prepare SNF shipments. This analysis is a work in progress, and the results are used to inform the development of new models of an integrated waste management system for commercial SNF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

The transportation factor in site evaluations for spent nuclear fuel storage and disposal facilities - 15460

New facility siting studies will be necessary to fulfill the Blue Ribbon Commission (BRC) on America's Nuclear Future 2012 recommendation to develop consolidated interim storage (CIS) for spent nuclear fuel in the next decade. Future site evaluations for used nuclear fuel (UNF) storage and disposal facilities must address transportation requirements, especially access to mainline railroads, early in the site selection process. The U.S. Department of Energy (DOE) in 2013 proposed a strategy for managing spent nuclear fuel in accordance with the BRC recommendations by establishing a pilot interim storage facility by 2021 and a consolidated interim storage facility by 2025. The authors demonstrate that currently available geographic information systems and already adopted siting criteria (defining favorable and unfavorable conditions) allow the identification of the transportation portion of a storage facility siting process to begin without delay. Ready access to the mainline rail system is absolutely critical since existing rail access, or easily constructed access, is a highly favorable condition for successful siting. Despite widespread rail abandonments over the past three decades, vast areas of the United States retain mainline rail service or could connect to existing mainlines by building new rail spurs of less than 25-50 miles. This precondition for transportation planning is critical. Planners need to start with the beginning - availability of rail transportation to and from the interim storage facility/facilities. A second step in transportation screening would be to evaluate those potential site locations within 50 miles of existing mainline railroads using other appropriate variables. This effort could begin with the application of the 1984 repository siting guidelines for transportation (10 CFR 960.5-2-7)[1]. Repository surface facilities have essentially the same transportation requirements as storage facilities. The guidelines in 10 CFR 960 provide a readily available template for evaluating sites based on transportation risks, impacts and costs. Initial site screening, after considering proximity to mainline railroads as the precursory step, should evaluate: 1) proximity to highly populated areas; 2) proximity to interstate highways; 3) proximity to environmentally sensitive areas and lands protected by federal, state, or tribal governments; and 4) proximity to military installations, hazardous materials facilities, or other facilities that might impact transportation safety and security.Several challenges exist which prevent effective transportation planning. The diversity of canister and cask types and SNF characteristics and conditions, the increasingly limited rail access at the shipping sites and the ease with which public attention can be marshalled against a waste management program, all suggest that future efforts need to identify potential storage sites that reduce transportation impacts and costs by reducing near-site transportation construction and reduce total system transportation impacts. This can most easily be accomplished by reducing shipping distances in a process that is easily defined and implemented. The paper concludes that the DOE should develop a site screening and evaluation program with transportation as a central consideration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Development of disposal container designs for disposal of high level waste and legacy spent fuel in the UK - 15471

Radioactive Waste Management Limited (RWM), a fully owned subsidiary of the Nuclear Decommissioning Authority (NDA), is responsible for the implementation of geological disposal in the UK. Three illustrative concepts of a geological disposal facility (GDF), corresponding to three generic geological environments - higher strength rock, lower strength sedimentary rock and evaporite rock - have been developed to demonstrate the viability of geological disposal of intermediate level waste, high level waste and spent fuel. It is envisaged that HLW which has been encapsulated in Waste Vitrification Plant (WVP) canisters, advanced gas-cooled reactor (AGR) SF which has been stripped of their spacers and packaged into slotted fuel cans, and PWR SF will be packaged into disposal containers for disposal. A set of schematic designs of these disposal containers was developed in the late 2000's. The design concept for the higher strength rock environment was based on the joint Posiva/SKB disposal container design, and the design concept for the lower strength sedimentary rock and evaporate rock illustrative environments was based on the disposal container concept from Switzerland's National Cooperative for the Disposal of Radioactive Waste (Nagra). The size, mass and dimensions of the disposal containers in the scheme designs vary significantly. The construction method and the lid closing method also vary between the disposal containers for the different waste contents. A design study has indicated that optimization of the number of HLW WVP canisters and the number of AGR slotted fuel cans in the disposal containers, standardizing construction method and closure arrangement across the containers for each geology type, providing handling features at both ends of the disposal container, and balancing of shielding around the containers, could bring significant advantage to the overall GDF programme by reducing the number of disposal containers, reducing GDF footprint, and reducing the extent of excavations, spoil generated and materials that are required for construction. A project to develop the scheme designs to conceptual level was carried in 2011 to 2013. In this project, the above changes were taken into account, technical underpinning of the designs was carried out and manufacturing method was considered. Two disposal container variants for different host rock environments have been developed: - Variant 1, for compatibility with the higher strength rock illustrative disposal system design. This is constructed from copper, with a cast iron insert and is based on the joint Posiva/SKB disposal container design. - Variant 2, for compatibility with the lower strength sedimentary rock and evaporite rock illustrative disposal system designs. This is a steel container based on the NAGRA disposal container design. Each variant has three internal configurations to accommodate 3 HLW vitrified canisters, 16 consolidated cans of AGR SF and 4 PWR spent fuel assemblies. These designs have been shown to bring significant benefit to the GDF designs, bringing a significant reduction to the operating programme and GDF programme cost, including reducing the number of disposal containers by 33% for HLW, 50% for AGR SF and reducing the size of the GDF footprint by 25%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

A Preliminary Analysis on the Applicability of Deep Borehole Disposal for Spent Fuels in Korea - 15496

A deep geological disposal system, the disposal depth is about 500 m below ground, is considered as the safest method to isolate the spent fuels (SF) or high-level radioactive waste from the human environment with the best available technology at present time. The disposal safety of this system has been demonstrated with underground research laboratory and some advanced countries such as Finland and Sweden are implementing their disposal project on commercial stage. However, if these high-level radioactive wastes can be disposed of in deeper and more stable rock formation than deep geological disposal depth, it has several advantages. Therefore, as an alternative disposal concept, i.e., deep borehole disposal technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, the general concept for deep borehole disposal of spent fuels or high level radioactive wastes was reviewed. And the key technologies and challenges in development of this disposal method with the nuclear environment were analyzed. Also, very preliminary deep borehole disposal concept including disposal canister concept was developed according to the nuclear environment. These results will be used as an input for the analyses of applicability for DBD in Korea. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗