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DOE OSTI2020

Generic FEPs Catalogue and Salt Knowledge Archive

This report describes the development of a comprehensive catalogue of generic features, events, and processes (FEPs) that are potentially important for the post-closure performance of a repository for high-level radioactive waste (HLW) and spent nuclear fuel (SNF) in salt (halite) host rock. The FEPs and other supporting information have been entered into a “SaltFEP” Database. The generic salt repository FEPs include consideration of relevant FEPs from a number of U.S., Dutch, German, and international FEP lists and should be a suitable starting point for any repository program in salt host rock. The salt FEP catalogue and database employ a FEP classification matrix approach that is based on the concept that a FEP is typically a process or event acting upon or within a feature. The FEP matrix provides a two-dimensional structure consisting of a Features/Components axis that defines the “rows” and a Processes/Events axis that defines the “columns” of the matrix. The design of the FEP classification matrix is consistent with repository performance assessment – the Features/Components axis is organized vertically to generally correspond to the direction of potential radionuclide migration (from the waste to the biosphere) and the Processes/Events axis is designed to represent the common two-way couplings between thermal processes and other processes (such as thermal-mechanical or thermal-hydrologic processes). Related FEPs can be easily identified – related FEPs will typically be grouped in a single matrix cell or aligned along a common row (Feature/Component) or column (Process/Event). The online SaltFEP database can be downloaded from www.saltfep.org. It contains the FEP matrix, the FEPs, and the associated processes for each FEP. It provides a starting point to create and document site-specific individual FEPs. Furthermore, the FEP matrix is connected to the Salt Knowledge Archive (SKA), a database of about 20,000 references and documents representing the historical knowledge on radioactive disposal in salt. This work is the result of an ongoing collaboration between researchers in the U.S., the Netherlands, and Germany, and supports the NEA Salt Club Mandate. It builds upon prior work which is documented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

A proposed solution to the challenges of disposal of greater than class C low-level radioactive waste - 15172

In 1982, the US Nuclear Regulatory Commission (NRC) adopted rules for the disposal of low-level radioactive waste. These rules established new standards for the siting, construction, operation and closure of low-level radioactive waste disposal facilities. The new rules also set out a classification system for low-level radioactive waste based on half-life and concentration of the radionuclides in the waste stream. The defined classes - Class A, Class B and Class C - were based on increasing toxicity with Class A being the least hazardous and Class C being relatively more hazardous. Radioactive waste that could not fit the classification scheme due to its long half-life or its high concentration of radioactive material was deemed to be greater than Class C low-level radioactive waste, with the ascribed acronym 'GTCC' low-level radioactive waste. Such waste was deemed not suitable for shallow land disposal based on the design standards for low-level radioactive waste landfills in place at that time. This resulted in a radioactive waste stream that had no disposal option. To remedy this problem, the US Congress in 1985 stipulated that the US Department of Energy would be responsible for the disposal of GTCC waste. Subsequently, the NRC amended its regulations to require GTCC low-level waste to be disposed in a geologic repository unless the Commission approved proposals for disposal in a low-level radioactive waste disposal facility. The Commission did not concurrently establish any specific technical requirements for GTCC disposal in a low-level radioactive waste disposal facility, but did infer that a site-specific performance analysis could be conducted in support of any proposal for land disposal of such waste. In light of this, Waste Control Specialists LLC (WCS) has conducted such a site-specific performance analysis that has indicated the disposal facility at Andrews Texas will perform satisfactorily as a disposal facility for certain types and quantities of GTCC waste. Concurrently with this analysis, WCS has submitted to the Texas Commission on Environmental Quality a petition for rulemaking that will permit disposal of GTCC waste in their disposal facility. This paper summarizes the technical and legal analyses that were conducted in support of the proposed rulemaking. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Brine Availability Test in Salt: Thermal-Hydrological-Mechanical-Chemical Simulations of a Heated Borehole in Salt - 20239

The Brine Availability Test in Salt (BATS) is a collaboration between Los Alamos, Sandia, and Lawrence Berkeley National Laboratories that is being conducted to reduce the uncertainties associated with disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in a hypothetical bedded salt repository. Bedded salt formations may be an acceptable host rock for disposal of nuclear waste due to salt's high thermal conductivity, extremely low permeability, and self-healing capability. In this paper, we report on the first round of the BATS project, a 'shakedown test' known as Phase 1s, with an emphasis on the supporting numerical modeling. The experiment was conducted at the Waste Isolation Pilot Plant (WIPP) using previously drilled horizontal boreholes. The test included a heater borehole in which temperature, pressure, and moisture was monitored, and two nearby temperature monitoring boreholes. Numerical simulations that reflect the 36-year-long history of activity at the WIPP site are necessary to develop the appropriate initial conditions. Testing with different heater designs reveals that a 750 W radiative heater within a borehole can achieve the 120 deg. C borehole wall temperature goal. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Leaching of Degraded Preservative Matrices, Based on Sodium Aluminophosphate Glasses, for High-Level Wastes

The chemical stability of vitrified high-level nuclear industry wastes in contact with water was studied experimentally in application to underground disposal of the wastes. The experiments were performed on samples of sodium aluminophosphate glass used in industrial solidification of wastes. Stable isotopes of strontium, cesium, uranium, and rare-earth elements as geochemical analogs of actinides were introduced into the glass to simulate radionuclides. It was observed that water vapor affects the degradation of the glass when vitrified wastes are placed in temporary repositories. It is shown that as a result of the degradation the rate of leaching of the glass in water increases significantly. In addition, a substantial portion of radionuclide simulators in colloidal form goes into the water medium.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Realization of the National Disposal Facility for Radioactive Waste in Bulgaria - 16611

Radioactive waste is generated in the Republic of Bulgaria from the operation and decommissioning of the country's nuclear power plants and from other sources in research, medical applications, and the industry. The need for building a National Disposal Facility (NDF) is recognized by the Bulgarian government and described and justified in its Environment Impact Assessment Decision. The Bulgarian State Enterprise for Radioactive Waste Management (SERAW) is endeavoring to build a near-surface repository for short-lived low- and intermediate-level radioactive waste (SL-LILW) to discharge its statutory responsibilities in waste management is. The European Union (EU) finances the establishment of the NDF through the Kozloduy International Decommissioning Support Fund (KIDSF). After an international tendering procedure, SERAW placed a contract for the development of the design of the NDF to the Consortium of Westinghouse Electric Spain S.A.R and DBE Technology GmbH of Germany, with ENRESA, the Spanish National Waste Management Agency providing technical review and support. Two Bulgarian companies participate also in the project as subcontractors - EQE Bulgaria and to a lesser extent KK-Project. The repository project official name is 'Technical Design and ISAR Preparation for the National Disposal Facility at Bulgaria', usually abbreviated as 'R-Project 5'. After obtaining SERAW's approval for the preferred repository conceptual design variant, the R Project 5 focused on developing the Technical Design on the one hand, and on preparing the Intermediate Safety Analysis Report (ISAR) to demonstrate the safety and suitability of the proposed NDF design on the other. These documents were handed over to the Bulgarian Nuclear Regulatory Authority (BNRA) as well as to the different authorities involved in the comprehensive licensing process required for nuclear facilities in the country. The corresponding permits have mostly been received or are expected for early 2016. In the last quarter of 2015 the procurement procedure for repository construction as turn-key project was started with the pre-qualification of possible constructors for the Phase I repository realization on a lump sum contract basis. Repository construction is now expected to start in 2016, with the start of waste disposal now scheduled for around 2020. With the realization of the Bulgarian National Disposal Facility an important milestone for waste disposal in the European Union is about to be achieved. This paper is a sequel of previously presented papers on the NDF construction, now including details of the complex licensing process that this disposal facility underwent, now in its final stages. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Operational Safety and Post-Closure Safety: Investigation of their Interactions by a FEP Analysis - 18040

The consideration of operational safety for a repository for high-level waste is a challenge, since it has to integrate the dimension of long-term safety. The operator has to demonstrate on the one hand the safety of the workers in the mine and of men and environment during operation. On the other hand a proper operation of the mine has to guarantee that the state of the facility at the time of closure will be compatible with the requirements for long-term safety. To achieve this, a pre-closure safety analysis should include (a) a general description of the structures, systems, components, equipment, and process activities during the operational phase, (b) identification and a systematic analysis of naturally occurring and human-induced hazards and (c) identification and documentation of the strong relationships between operational safety and post-closure safety. In a German R and D project, a FEP catalogue for the operational phase was developed based on the national disposal concepts for clay and salt. The FEP catalogue contains analyses of the processes and events that affect the components (features) of the repository system in its entirety. The FEP catalogue was applied to identify hazards for the operational phase and to document the impacts of these FEP (including the identified hazards) on post-closure safety. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Preliminary Analysis of Postclosure DPC Criticality Consequences

One of the objectives of the United States (U.S.) Department of Energy's (DOE) Office of Nuclear Energy's Spent Fuel and Waste Science and Technology Campaign is to better understand the technical basis, risks, and uncertainty associated with the safe and secure disposition of spent nuclear fuel (SNF) and high-level radioactive waste. Commercial nuclear power generation in the U.S. has resulted in thousands of metric tons of SNF, the disposal of which is the responsibility of the DOE (Nuclear Waste Policy Act 1982). Any repository licensed to dispose the SNF must meet requirements regarding the longterm performance of that repository. For an evaluation of the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration. Of particular interest is the potential behavior of SNF in dual-purpose canisters (DPCs), which are currently being used to store and transport SNF but were not designed for permanent geologic disposal. A two-phase study has been initiated to begin examining the potential consequences, with respect to longterm repository performance, of criticality events that might occur during the postclosure period in a hypothetical repository containing DPCs. Phase I, a scoping phase, consisted of developing an approach intended to be a starting point for the development of the modeling tools and techniques that may eventually be required either to exclude criticality from or to include criticality in a performance assessment (PA) as appropriate; Phase I is documented in Price et al. (2019). The Phase I approach guided the analyses and simulations done in Phase II to further the development of these modeling tools and techniques as well as the overall knowledge base. The purpose of this report is to document the results of the analyses conducted during Phase II. The remainder of Section 1 presents the background, objective, and scope of this report, as well as the relevant key assumptions used in the Phase II analyses and simulations. Subsequent sections discuss the analyses that were conducted (Section 2), the results of those analyses (Section 3), and the summary and conclusions (Section 4). This report fulfills the Spent Fuel and Waste Science and Technology Campaign deliverable M2SF-20SN010305061.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

GDSA Repository Systems Analysis Investigations in FY2021

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Disposition (SFWD), has been conducting research and development on generic deep geologic disposal systems (i.e., geologic repositories). This report describes specific activities in the Fiscal Year (FY) 2021 associated with the Geologic Disposal Safety Assessment (GDSA) Repository Systems Analysis (RSA) work package within the SFWST Campaign. The overall objective of the GDSA RSA work package is to develop generic deep geologic repository concepts and system performance assessment (PA) models in several host-rock environments, and to simulate and analyze these generic repository concepts and models using the GDSA Framework toolkit, and other tools as needed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

GDSA Repository Systems Analysis Investigations in FY2022

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Disposition (SFWD), has been conducting research and development on generic deep geologic disposal systems (i.e., geologic repositories). This report describes specific activities in the Fiscal Year (FY) 2022 associated with the Geologic Disposal Safety Assessment (GDSA) Repository Systems Analysis (RSA) work package within the SFWST Campaign. The overall objective of the GDSA RSA work package is to develop generic deep geologic repository concepts and system performance assessment (PA) models in several host-rock environments, and to simulate and analyze these generic repository concepts and models using the GDSA Framework toolkit, and other tools as needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Advective gas flow in bentonite: Development and comparison of enhanced multi-phase numerical approaches

Understanding the impact of repository gas, generated from degradation of waste and its interaction with the host rock, is essential when assessing the performance and safety function of long-term disposal systems for radioactive waste. Numerical models based on conventional multi-phase flow theory have historically been applied to predict the outcome and impact of gas flow on different repository components. However, they remain unable to describe the full complexity of the physical processes observed in water-saturated experiments (e.g., creation of dilatant pathways) and thus, the development of novel representations for their description is required when assessing fully saturated clay-based systems. This was the primary focus of Task A within the international cooperative project DECOVALEX-2019 (D-2019) and refinement of these approaches is the primary focus of this study (Task B in the current phase of DECOVALEX-2023). This paper summarises development of enhanced numerical representations of key processes and compares the performance of each model against high-quality laboratory test data. Experimental data reveals that gas percolation in water-saturated compacted bentonite is characterised by four key features: (i) a quiescence phase, followed by (ii) the gas breakthrough, which leads to a (iii) peak value, which is then followed by (iv) a negative decay. Three models based on the multiphase flow theory have been developed. These models can provide good initial values and reasonable responses for gas breakthrough (although some of them still predict a too-smooth response). Peak gas pressure values are in general reasonably well captured, although maximum radial stress differences are observed at 48 mm from the base of the sample. Here, numerical peak values of 12.8 MPa are predicted, whereas experimental values are about 11 MPa. These models are also capable of providing a reasonable representation of the negative pressure decay following peak pressure. However, other key specific features (such as the timing of gas breakthrough) still require a better representation. The model simulations and their comparison with experimental data show that these models need to be further improved with respect to model parameter calibration, the numerical representation of spatial heterogeneities in material properties and flow localisation, and the upscaling of the related physical processes and parameters. To further understand gas flow localisation, a new conceptual model has been developed, which shows that discrete channels can possibly be induced through the instability of gas-bentonite interface during gas injection, thus providing a new perspective for modeling gas percolation in low-permeability deformable media.

58 GEOSCIENCES↗
DOE OSTI2016

Evaluation of used fuel disposition in clay-bearing rock.

The R&D program from the DOE Used Fuel Disposition Campaign (UFDC) has documented key advances in coupled Thermal-Hydrological-Mechanical-Chemical (THMC) modeling of clay to simulate its complex dynamic behavior in response to thermal and hydrochemical feedbacks. These efforts have been harnessed to assess the isolation performance of heatgenerating nuclear waste in a deep geological repository in clay/shale/argillaceous rock formations. This report describes the ongoing disposal R&D efforts on the advancement and refinement of coupled THMC process models, hydrothermal experiments on barrier clay interactions, used fuel and canister material degradation, thermodynamic database development, and reactive transport modeling of the near-field under non-isothermal conditions. These play an important role to the evaluation of sacrificial zones as part of the EBS exposure to thermallydriven chemical and transport processes. Thermal inducement of chemical interactions at EBS domains enhances mineral dissolution/precipitation but also generates mineralogical changes that result in mineral H2O uptake/removal (hydration/dehydration reactions). These processes can result in volume changes that can affect the interface / bulk phase porosities and the mechanical (stress) state of the bentonite barrier. Characterization studies on bentonite barrier samples from the FEBEX-DP international activity have provided important insight on clay barrier microstructures (e.g., microcracks) and interactions at EBS interfaces. Enhancements to the used fuel degradation model outlines the need to include the effects of canister corrosion due the strong influence of H 2 generation on the source term.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Evaluation of used fuel disposition in clay-bearing rock

The R&D program from the DOE Used Fuel Disposition Campaign (UFDC) has documented key advances in coupled Thermal-Hydrological-Mechanical-Chemical (THMC) modeling of clay to simulate its complex dynamic behavior in response to thermal and hydrochemical feedbacks. These efforts have been harnessed to assess the isolation performance of heat-generating nuclear waste in a deep geological repository in clay/shale/argillaceous rock formations. This report describes the ongoing disposal R&D efforts on the advancement and refinement of coupled THMC process models, hydrothermal experiments on barrier clay interactions, used fuel and canister material degradation, thermodynamic database development, and reactive transport modeling of the near-field under non-isothermal conditions. These play an important role to the evaluation of sacrificial zones as part of the EBS exposure to thermally-driven chemical and transport processes. Thermal inducement of chemical interactions at EBS domains enhances mineral dissolution/precipitation but also generates mineralogical changes that result in mineral H2O uptake/removal (hydration/dehydration reactions). These processes can result in volume changes that can affect the interface / bulk phase porosities and the mechanical (stress) state of the bentonite barrier. Characterization studies on bentonite barrier samples from the FEBEX-DP international activity have provided important insight on clay barrier microstructures (e.g., microcracks) and interactions at EBS interfaces. Enhancements to the used fuel degradation model outlines the need to include the effects of canister corrosion due the strong influence of H2 generation on the source term.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Lessons Learned from the Development of the DGR Konrad - 19619

The Federal Republic Germany builds the deep geological repository Konrad. The repository is suitable for low- and intermediate-level radioactive waste (LAW and MAW) and it is the first German repository site, which is not located in a salt geological formation. Konrad is a former iron ore mine in Salzgitter (in southeast Lower Saxony) with an active production time between 1961 and 1976. In 2007, the BGE started with the conversion of the mine infrastructure above and underground to build the repository while using the shafts and some roadways of the former mine. The experiences gained from ten years project progress with its complexities and difficulties can be used for similar repository projects in the future. The Konrad repository is designed for 303,000 m{sup 3} of LAW and MAW, which will be emplaced in chambers. The emplacement chambers are located outside of the existing mine, so an impairment by the old mine workings can be excluded. So far, the first area with emplacement chambers has been excavated. In order to be able to start repository operations in 2027, a wide range of work in building construction, civil engineering, shaft sinking and infrastructure development above and below ground has been necessary and will be completed in the next years. The main activities are focused on the Konrad 1 and 2 sites and the part of the infrastructure located in the future control area, which is used for repository operations. There are still many challenges to overcome in the Konrad repository project, although many processes have been put on the right track in recent years. Looking back on past activities and in the cautious preview, this paper gives some indications from which lessons can be drawn and should be applied to similar projects in the future. - Prefer small dimensions for the infrastructure, such as insets and roadway cross-section, and the appropriate machinery and handling equipment; - Detailed process planning for construction in existing infrastructure in the run-up to implementation; - Intensive communication with contracted partners on the preparation of planning documents for nuclear examination procedures and ensuring an excellent contract management; - Early start of the implementation planning for projects under nuclear law, taking into account the relevant boundary conditions; - Ensuring the support of the project by the local community and other stakeholders. Overall, the use of parts of the existing above and underground infrastructure for the construction of a new repository is more difficult than for a complete new construction. But using the shafts of an existing mine delivers the advantage of very good knowledge about the underground environment. In addition, the areas necessary for the construction and operation of the buildings and facilities are already available and secured and do not compete with other projects.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Magnetite Dissolution using Hydrazine-Acid Solution for Chemical Decontamination

Chemical decontamination is committed to giving safe decommissioning of nuclear facilities as well as maintenance during the normal operation under sufficiently reduced radiation field. The removal of radioisotopes from the primary coolant system of nuclear power plant takes place with the dissolution of metal oxide layer. Dissolution of metal oxides in aqueous acidic solutions is primarily controlled by the protonation of metal-oxygen bonding to release the metal ions from the oxide lattices but in the presence of reducing or chelating agents dissolution is often facilitated by the enhanced mass transfer of metal ions together with chelating reagents as well as metal reduction to soluble state such as insoluble ferric ions to soluble ferrous ions of hematite. The mechanisms and kinetics related to the metal oxides dissolution in varied acidic reagents have been reported for many decades even though a general picture of dissolution is not available due to the peculiarity of each oxide in structure, stability of metal ions, solubility, and reactivity. In a number of experimental results and practical reports, organic acids such as oxalic acid, citric acid, ascorbic acid, malonic acid, thioglycolic acid, etc. were used as solvent reagents and many other literatures reported the effectiveness of the chelating agents in dissolution rate such as iminodiacetic acid (IDA), diethylene triamine pentaacetic acid (DTPA), nitrilo triacetic acid (NTA), and ethylenediaminetetraacetic acid (EDTA) by the sorption of dissolved metal ions forming stable metal chelates. Also reductive dissolution by the addition of reducing reagents such as hydrazine, vanadium as well as organic acids was reported with effective experimental results. However, organic acids and chelating ligands are often problematic on the safety of radioactive wastes. Organic compounds in a repository potentially form a complex with radionuclides, which increases the mobility of radionuclides and consequently increases the probability of the radioactive contamination of groundwater. Furthermore oxalic acid can cause a severe intergranular attack (IGA) to the type 304 stainless steel. Therefore, the use of organic acids or chelating reagents even though those effects to the dissolution performance is much greater than other chemical components must be minimized or carefully determined with appropriate treatment processes. Alternatively dissolution accommodated by inorganic reducing reagents or catalysts in inorganic acid solutions have been extensively studied for the feasible dissolution outcomes. The mixture of sulfuric acid and sulfurous acids as a dissolution media was reported to increase dissolution rate due to the effect of protonation by H{sub 2}SO{sub 4} and reducing effect of H{sub 2}SO{sub 3}. The increased dissolution of magnetite was also found in sulfur dioxide solution containing copper ions because SO{sub 2} plays as a reducing catalyst to form FeHSO{sub 3}{sup +} with Fe(III) from iron oxides and accordingly the concentration of Fe(III) ions decreases to accelerate the iron oxide dissolution. The present study is concerned with the feasibility of not using the organic compounds for the dissolution of oxide layers. We propose an acidic solution containing sulfuric acid, hydrazine, and copper ions for the effective dissolution of magnetite powders. The effects of both hydrazine and copper ions as a reducing agent and a catalyst respectively to promote the reducing power, as well as temperature will be investigated on the dissolution of magnetite powder. (authors)

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

Roadmap for disposal of Electrorefiner Salt as Transuranic Waste.

The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Mobility of Radionuclides in Fractured Carbonate Rocks: Lessons from a Field-Scale Transport Experiment

Current research on radionuclide disposal is mostly conducted in granite, clay, saltstone, or volcanic tuff formations. These rock types are not always available to host a geological repository in every nuclear waste-generating country, but carbonate rocks may serve as a potential alternative. To assess their feasibility, a forced gradient cross-borehole tracer experiment was conducted in a saturated fractured chalk formation. The mobility of stable Sr and Cs (as analogs for their radioactive counterparts), Ce (an actinide analog), Re (a Tc analog), bentonite particles, and fluorescent dye tracers through the flow path was analyzed. The migration of each of these radionuclide analogs (RAs) was shown to be dependent upon their chemical speciation in solution, their interactions with bentonite, and their sorption potential to the chalk rock matrix. The brackish groundwater resulted in flocculation and immobilization of most particulate RAs. Nevertheless, the high permeability of the fracture system allowed for fast overall transport times of all aqueous RAs investigated. This study suggests that the geochemical properties of carbonate rocks may provide suitable conditions for certain types of radionuclide storage (in particular, brackish, high-porosity, and low-permeability chalks). Nevertheless, careful consideration should be given to high-permeability fracture networks that may result in high radionuclide mobility.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗