Publications

Abstract not provided.

Commercial generation of energy by nuclear power plants in the United States (U.S.) has produced thousands of metric tons of spent nuclear fuel (SNF), the disposal of which is the responsibility of the U.S. Department of Energy (DOE). Utilities typically utilize the practice of storing this SNF in dual-purpose canisters (DPCs). DPCs were designed, licensed, and loaded to meet Nuclear Regulatory Commission (NRC) requirements that preclude the possibility of a criticality event during SNF storage and transport, but were not designed or loaded to preclude the possibility of a criticality event during the regulated post-closure period following disposal, which could be up to 1,000,000 years (Price, 2019). There are several options being investigated that could facilitate the disposal of SNF stored in DPCs in a geologic repository (Hardin et al., 2015; SNL 2020b; SNL 2021b). These include: (1) repackage the SNF into canisters that are designed to prevent criticality during the regulated post-closure period following disposal, but with an increased disposal cost estimated at approximately $\$$20B in United States dollars (USD) (Freeze et al., 2019); (2) analysis of the probability and consequences of criticality from the direct disposal of DPCs during a 1,000,000-year post-closure period in several geologic disposal media (Price, 2019); and (3) filling the void space of a DPC with a material before its disposal that significantly limits the potential for criticality over the post-closure regulatory period. This report further investigates the third option, filling DPC already containing SNF with a material to limit the potential for criticality over the post-closure regulatory period.

The US Department of Energy (DOE) is investigating the use of different materials that could be used to fill the void space inside a dual-purpose canister (DPC) loaded with spent nuclear fuel (SNF) just before it is emplaced in a deep geologic repository. The purpose of adding filler material is to maintain subcritical conditions in the repository during the postclosure period, which can span up to 1,000,000 years. Several types of materials have been proposed, including metals, cements, particulates, and glass. Part of this investigation addresses how the presence of filler material inside a DPC will affect the performance of the repository with respect to the repository features; the consequences of events that may occur; and the multiple thermal, hydrologic, chemical, and mechanical processes that may occur in a deep geologic repository over long timescales. This report describes some of the filler materials that have been proposed and studied; identifies 11 features, 6 events, and 25 processes that may be affected by the presence of filler materials; and discusses the effects that may require consideration for each feature, event, or process. The results of this study can be used to direct appropriate research and to develop suitable models if the DOE decides to use fillers to maintain subcritical conditions in DPCs used to dispose of SNF.

Commercial generation of energy by nuclear power plants in the United States (U.S.) has produced thousands of metric tons of spent nuclear fuel (SNF), the disposal of which is the responsibility of the U.S. Department of Energy (DOE). Utilities typically utilize the practice of storing this SNF in dual-purpose canisters (DPCs). DPCs were designed, licensed, and loaded to meet Nuclear Regulatory Commission (NRC) requirements that preclude the possibility of a criticality event during SNF storage and transport, but were not designed or loaded to preclude the possibility of a criticality event during the regulated post-closure period following disposal, which could be up to 1,000,000 years (Price, 2019).

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The performance of APCs with relatively low compressive strength and poor stability under hydrothermal conditions make them less than desirable for the DPC use case. Meanwhile, grossite as a primary filler material or as a modifier has resulted in marked improvements in the properties of several DPC cement filler candidates. Grossite CAPCs have substantial mechanical strength even after irradiation. However, the significant decrease in strength observed post-irradiation requires further investigation before it is advanced as a material for the use case. As a modifier, grossite improves strength and set times of the APC and WAPC cements. Hibonite CAPCs also show considerable promise although their degradation under hydrothermal conditions is a potentially significant liability. Finally, with recent improvements in working time and compressive strength, the WAPCs remain in contention as viable candidates for the DPC use case.

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The fillers research and development (R&D) program, mostly experimental, is part of a broader R&D program that includes new process modeling and performance assessment of criticality effects and the overall importance of criticality to repository performance (consequence screening). A literature research and consultation effort with experts by Hardin and Brady (2018) identified several potentially effective and workable filler materials including cements (primarily phosphate based), moltenmetal alloys, and low-temperature glasses. Filler attributes were defined, and the preliminary lists were compared qualitatively. Further comparative analysis will be done (e.g., cost estimates) after experimental screening has narrowed the list of alternatives. The following cement filler compositions were selected for experimental development work and accelerated testing in FY20: Aluminum phosphate cements (APCs); more specifically aluminum oxide / aluminum phosphate (Al₂O₃ / AlPO₄) cements in which Al₂O₃ serves as the filler material bound by an AlPO₄ binder formed by the reaction of Al₂O₃ with H₃PO₄; Calcium phosphate cements (CPCs); more specifically composed of pure or nearly pure hydroxyapatite or HAP (Ca₅(PO₄)₃(OH)); Wollastonite phosphate cements (WPC), specifically wollastonite and aluminum or calcium aluminum phosphates in which CaSiO₃ serves as the filler material and the phosphate serves as the binder. The FY20 effort focused on the optimization of compositions and subsequent processing of these three materials to achieve dense and well-consolidated monolithic samples with 30 to 40% porosity and permeabilities of 1 millidarcy. At the close of this progress report the aluminum phosphate cements (APCs) and the wollastonite phosphate cements (WPCs) appear to show the most promise for continued development. Less progress has been made with the calcium phosphate cements (CPCs); their slurry viscosities are high (and difficult to measure) and they exhibit relatively short cure times of 2 to 3 hours with concomitant and excessive volatile (e.g. CO₂) generation.

The fillers research and development (R&D) program, mostly experimental, is part of a broader R&D program that includes new process modeling and performance assessment of criticality effects and the overall importance of criticality to repository performance (consequence screening). A literature research and consultation effort with experts by Hardin and Brady (2018) identified several potentially effective and workable filler materials including cements (primarily phosphate based), moltenmetal alloys, and low-temperature glasses. Filler attributes were defined, and the preliminary lists were compared qualitatively. Further comparative analysis will be done (e.g., cost estimates) after experimental screening has narrowed the list of alternatives. The research presented here is focused Sandia’s efforts for the development of phosphate-based cement fillers. Molten metal filler research is an ongoing activity at Oak Ridge National Laboratories and is not discussed herein. After the completion of the FY20 research effort the following cement filler compositions were selected for further experimental development work and advanced testing in FY21: 1. Aluminum phosphate cements (APCs); more specifically aluminum oxide / aluminum phosphate (Al₂O₃ / AlPO₄) cements in which Al₂O₃ serves as the filler material bound by an AlPO₄ binder formed by the reaction of Al₂O₃ with various phosphate sources; 2. Wollastonite phosphate cements (WPCs), specifically wollastonite and aluminum or calcium aluminum phosphates in which CaSiO₃ serves as the filler material bound by a calcium phosphate that serves as the binder; and 3. Calcium aluminate phosphate cements (CAPCs) specifically grossite (CaAl₄O₇) and hibonite (CaAl₁₁O₁₈) fillers bound by an aluminum phosphate that serves as the binder. This effort focused on the optimization and subsequent processing of these three cements to achieve dense and well-consolidated monolithic samples. Upon completion of the FY21 effort the aluminum phosphate cements (APCs) and the calcium aluminate phosphate cements (CAPCs) show the most promise for advanced testing and scale up. We will begin the work in FY22 focused on testing the performance of these two cements in small scale DPCs as well as advanced materials testing to evaluate cement performance under expected radiation doses and representative post-closure geochemical environments.

The US generates approximately 2,000 Metric Tons of Heavy Metal (MTHM) of commercial spent fuel (SNF) each year and currently stores ~ 85,000 MTHM of commercial SNF at 70+ reactor sites, the disposal of which is the responsibility of the US Department of Energy (DOE). SNF is initially stored in spent nuclear fueld pools (SFPs). SFPs were initially constructed by US utilities for temporary fuel storage, but with no final disposal pathway available, SFPs are reaching capacity. To allow continued operation of the nation's commercial nuclear reactor fleet, utilities started transferring SNF from SFPs (wet storage) to dry cask storage systems, typically using dual-purpose (storage and transportation) canisters (DPCs). And while DPCs were designed, licensed and loaded to meet Nuclear Regulatory Commission (NRC) requirements that preclude the possibility of a critical event during SNF storage and transport, they were not designed or loaded to preclude the possibility of a criticality event during the regulated post-closure period following disposal, which could be up to 1,000,000 years (Price, 2019). DPC filler option criteria are detailed and materials that exhibit these attributes are explored. This document is an update of the SNL Joint Workplan on Filler Investigations for DPCs.

This report represents completion of milestone deliverable M2SF-21SN010309012 “Annual Status Update for OWL and Waste Form Characteristics” that provides an annual update on status of fiscal year (FY 2020) activities for the work package SF-20SN01030901 and is due on January 29, 2021. The Online Waste Library (OWL) has been designed to contain information regarding United States (U.S.) Department of Energy (DOE)-managed (as) high-level waste (DHLW), spent nuclear fuel (SNF), and other wastes that are likely candidates for deep geologic disposal, with links to the current supporting documents for the data (when possible; note that no classified or official-use-only (OUO) data are planned to be included in OWL). There may be up to several hundred different DOE-managed wastes that are likely to require deep geologic disposal. This draft report contains versions of the OWL model architecture for vessel information (Appendix A) and an excerpt from the OWL User’s Guide (Appendix B and SNL 2020), which are for the current OWL Version 2.0 on the Sandia External Collaboration Network (ECN).

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Radioiodine (129I) poses a risk to the environment due to its long half-life, toxicity, and mobility. It is found at the U.S. Department of Energy Hanford Site due to legacy releases of nuclear wastes to the subsurface where 129I is predominantly present as iodate (IO3-). To date, a cost-effective and scalable cleanup technology for 129I has not been identified, with hydraulic containment implemented as the remedial approach. Here, novel high-performing sorbents for 129I remediation with the capacity to reduce 129I concentrations to or below the US Environmental Protection Agency (EPA) drinking water standard and procedures to deploy them in an ex-situ pump and treat (P&T) system are introduced. This includes implementation of hybridized polyacrylonitrile (PAN) beads for ex-situ remediation of IO3-contaminated groundwater for the first time. Iron (Fe) oxyhydroxide and bismuth (Bi) oxyhydroxide sorbents were deployed on silica substrates or encapsulated in porous PAN beads. In addition, Fe-, cerium (Ce)-, and Bi-oxyhydroxides were encapsulated with anion-exchange resins. The PAN-bismuth oxyhydroxide and PAN-ferrihydrite composites along with Fe- and Ce-based hybrid anion-exchange resins performed well in batch sorption experiments with distribution coefficients for IO3- of >1000 mL/g and rapid removal kinetics. Of the tested materials, the Ce-based hybrid anion-exchange resin was the most efficient for removal of IO3- from Hanford groundwater in a column system, with 50% breakthrough occurring at 324 pore volumes. The functional amine groups on the parent resin and amount of active sorbent in the resin can be customized to improve the iodine loading capacity. These results highlight the potential for IO3- remediation by hybrid sorbents and represent a benchmark for the implementation of commercially available materials to meet EPA standards for cleanup of 129I in a large-scale P&T system.

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Commercial spent nuclear fuel (SNF) is accumulating at 72 sites across the U.S., at the rate of about 2,000 metric tons of uranium (MTU) per year. There are currently more than 2,700 dualpurpose canisters (DPCs) loaded with SNF, which are designed for storage and transportation but not disposal. If current storage practices continue, about half the eventual total U.S. SNF inventory will be in about 5,500 dry storage systems by 2035, with the entire inventory stored in 10,000 or more by 2060. The quantity of SNF in DPCs is now much greater than that anticipated in the past, leading the DOE to investigate the technical feasibility of direct disposal of SNF in DPCs. Studies in 2013-2015 concluded that the main technical challenges for disposal of SNF in DPCs are thermal management, handling and emplacement of large, heavy waste packages, and postclosure criticality control (Hardin et al. 2015). Of these, postclosure criticality control is the most challenging, and the R&D needed for this aspect of DPC direct disposal is the primary focus of this report.

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The fillers R&D program, mostly experimental, is part of a broader R&D program that includes new process modeling and performance assessment of criticality effects and the overall importance of criticality to repository performance (consequence screening). A literature research and consultation effort with experts by Hardin and Brady (2018) identified several potentially effective and workable filler materials including cements (primarily phosphate based), molten-metal alloys, and low-temperature glasses. Filler attributes were defined and the preliminary lists were compared qualitatively. Further comparative analysis will be done (e.g., cost estimates) after experimental screening has narrowed the list of alternatives. The following cement filler compositions were selected for experimental development work and accelerated testing in FY19: Aluminum phosphate cements (APCs); more specifically aluminum oxide / aluminum phosphate (Al₂O₃/ AlPO₄) cements in which Al₂O₃ serves as the filler material bound by an AlPO₄ binder formed by the reaction of Al₂O₃ with H₃PO₄; Calcium phosphate cements (CPCs); more specifically composed of pure or nearly pure hydroxyapatite (Ca₅(PO₄)₃(OH)); Magnesium potassium phosphate cements (MKPs) composed of magnesium oxide / magnesium potassium phosphate (MgO / MgKPO₄) cements in which MgO serves as the filler and MgKPO₄ serves as the binder formed by the reaction of MgO with monopotassium phosphate (KH₂PO₄) and tricalcium phosphate ((Ca₃(PO₄)₂); Two additional potential cement materials were explored preliminarily as the result of: (1) continued literature investigations into other filler candidates (wollastonite-based phosphate ceramic) and (2) the experimental discovery of a well-consolidated fly ash phosphate cement during the evaluation of fly ash as a potential filler material with Al₂O₃in APCs. Fly ash phosphate cements, more specifically in which a fly ash material composed primarily of mullite and quartz serves as the filler and is reacted with H₃PO₄ to form amorphous phosphate phase(s) as the binder; Wollastonite aluminum phosphate cements (WAPC), specifically wollastonite / aluminum phosphate (CaSiO₃/ AlPO₄) in which CaSiO₃ serves as the filler material and AlPO₄ serves as the binder formed by Al(OH)₃ or metakaolin as Al sources and H₃PO₄ or ammonium dihydrogen phosphate (ADP) (NH₄H₂PO₄) as phosphate sources. The FY19 effort focused on the optimization of compositions and subsequent processing of these five materials to achieve dense and well-consolidated monolithic samples with relatively low porosity. Once these goals were met basic material properties screening evaluations were performed including an assessment of dissolution resistance in water at elevated temperature (200 °C) and mechanical testing including unconfined compressive strength (UCS) testing. To date, the aluminum phosphate cements (APCs) appear to show the most promise for continued development. They are easily prepared and form smooth pourable slurries that remain stable for days with relatively low viscosities of several thousand centipoise (cP). They are then set at elevated temperatures (e.g., 170 °C) under ambient (0.1 MPa) or elevated pressure (~1MPa). Overall, they demonstrate the best dissolution resistance in water at elevated temperature (200 °C) and good compressive strengths. However, additional effort is required to optimize the APC slurry formulations and the process used for thermal curing these materials. The calcium phosphate cements (CPCs) can be formed at room temperature to produce a well-consolidated body. However, their slurry viscosities are very high (and difficult to measure) and they exhibit relatively short cure times of 2 to 3 hours. Also, dissolution resistance is very poor, the poorest of all the cements examined The same is the case for the small number of MKP cements fabricated; they cure very quickly (10 minutes or less) and disintegrate within a few hours upon immersion in distilled water. Surprisingly, fly ash reacts with phosphoric acid to form dense and well-consolidated cements but the mixture rapidly sets at room temperature (less than 30 minutes) and the subsequent conversion of the binder to an amorphous phosphate phase(s) as a function of temperature is complicated. Finally, the wollastonite aluminum phosphate cements (WAPC) are easily prepared and form smooth pourable slurries that remain stable for several hours. They are then set at 130 °C. A WAPC sample exhibited the highest compressive strengths of all the materials we evaluated but in general their dissolution resistance to water is poor.

This report represents completion of milestone deliverable M2SF-19SNO10309013 "Online Waste Library (OWL) and Waste Forms Characteristics Annual Report" that reports annual status on fiscal year (FY) 2019 activities for the work package SF-19SN01030901 and is due on August 2, 2019. The online waste library (OWL) has been designed to contain information regarding United States (U.S.) Department of Energy (DOE)-managed (as) high-level waste (DHLW), spent nuclear fuel (SNF), and other wastes that are likely candidates for deep geologic disposal, with links to the current supporting documents for the data (when possible; note that no classified or official-use-only (OUO) data are planned to be included in OWL). There may be up to several hundred different DOE-managed wastes that are likely to require deep geologic disposal. This annual report on FY2019 activities includes evaluations of waste form characteristics and waste form performance models, updates to the OWL development, and descriptions of the management processes for the OWL. Updates to the OWL include an updated user's guide, additions to the OWL database content for wastes and waste forms, results of the beta testing and changes implemented from it. Also added are descriptions of the management/control processes for the OWL development, version control, and archiving. These processes have been implemented as part of the full production release of OWL (i.e., OWL Version 1.0), which has been developed on, and will be hosted and managed on, Sandia National Laboratories (SNL) systems. The version control/update processes will be implemented for updates to the OWL in the future. Additionally, another process covering methods for interfacing with the DOE SNF Database (DOE 2007) at Idaho National Laboratory on the numerous entries for DOE-managed SNF (DSNF) has been pushed forward by defining data exchanges and is planned to be implemented sometime in FY2020. The INL database is also sometimes referred to as the Spent Fuel Database or the SFDB, which is the acronym that will be used in this report. Once fully implemented, this integration effort will serve as a template for interfacing with additional databases throughout the DOE complex.

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This report presents a generic (i.e., site-independent) preliminary plan for drilling, testing, sampling, and analyzing data for a deep characterization borehole drilled into crystalline basement for the purposes of assessing the suitability of a site for deep borehole disposal (DBD). This research was performed as part of the deep borehole field test (DBFT). Based on revised U.S. Department of Energy (DOE) priorities in mid-2017, the DBFT and other research related to a DBD option was discontinued; ongoing work and documentation were closed out by the end of fiscal year (FY) 2017. This report was initiated as part of the DBFT and documented as an incomplete draft at the end of FY 2017. The report was finalized by Sandia National Laboratories in FY2018 without DOE funding, subsequent to the termination of the DBFT, and published in FY2019. This report presents a possible sampling, testing, and analysis campaign that could be carried out as part of a future project to quantify geochemical, geomechanical, geothermal, and geohydrologic conditions encountered at depths up to 5 km in crystalline basement.

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This report represents completion of milestone deliverable M2SF-18SNO10309013 "Inventory and Waste Characterization Status Report and OWL Update that reports on FY2018 activities for the work package (WP) SF-18SNO1030901. This report provides the detailed final information for completed FY2018 work activities for WP SF-18SN01030901, and a summary of priorities for FY2019. This status report on FY2018 activities includes evaluations of waste form characteristics and waste form performance models, updates to the OWL development, and descriptions of the two planned management processes for the OWL. Updates to the OWL include an updated user's guide, additions to the OWL database content for wastes and waste forms, results of the Beta testing and changes implemented from it. There are two processes being planned in FY2018, which will be implemented in FY2019. One process covers methods for interfacing with the DOE SNF DB (DOE 2007) at INL on the numerous entries for DOE managed SNF, and the other process covers the management of updates to, and version control/archiving of, the OWL database. In FY2018, we have pursued three studies to evaluate/redefine waste form characteristics and/or performance models. First characteristic isotopic ratios for various waste forms included in postclosure performance studies are being evaluated to delineate isotope ratio tags that quantitatively identify each particular waste form. This evaluation arose due to questions regarding the relative contributions of radionuclides from disparate waste forms in GDSA results, particularly, radionuclide contributions of DOE-managed SNF vs HLW glass. In our second study we are evaluating the bases of glass waste degradation rate models to the HIP calcine waste form. The HIP calcine may likely be a ceramic matrix material, with multiple ceramic phases with/without a glass phase. The ceramic phases are likely to have different degradation performance from the glass portion. The distribution of radionuclides among those various phases may also be a factor in the radionuclide release rates. Additionally, we have an ongoing investigation of the performance behavior of TRISO particle fuels and are developing a stochastic model for the degradation of those fuels that accounts for simultaneous corrosion of the silicon carbide (SiC) layer and radionuclide diffusion through it. The detailed model of the TRISO particles themselves, will be merged with models of the degradation behavior(s) of the graphite matrix (either prismatic compacts or spherical "pebbles") containing the particles and the hexagonal graphite elements holding the compacts.

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This report provides an update to Sassani et al. (2016) and includes: (1) an updated set of inputs (Sections 2.3) on various additional waste forms (WF) covering both DOE-managed spent nuclear fuel (SNF) and DOE-managed (as) high-level waste (HLW) for use in the inventory represented in the geologic disposal safety analyses (GDSA); (2) summaries of evaluations initiated to refine specific characteristics of particular WF for future use (Section 2.4); (3) updated development status of the Online Waste Library (OWL) database (Section 3.1.2) and an updated user guide to OWL (Section 3.1.3); and (4) status updates (Section 3.2) for the OWL inventory content, data entry checking process, and external OWL BETA testing initiated in fiscal year 2017.

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A growing body of research supports widespread future reliance on apatite for radioactive waste cleanup. Apatite is a multi-functional radionuclide sorbent that lowers dissolved radionuclide concentrations by surface sorption, ion exchange, surface precipitation, and by providing phosphate to precipitate low-solubility radionuclide-containing minerals. Natural apatites are rich in trace elements, and apatite's stability in the geologic record suggest that radionuclides incorporated into apatite, whether in a permeable reactive barrier or a waste form, are likely to remain isolated from the biosphere for long periods of time. Here we outline the mineralogic and surface origins of apatite-radionuclide reactivity and show how apatites might be used to environmental advantage in the future.

The sorption of selenite, SeO32−, by carbonate substituted hydroxylapatite was investigated using batch kinetic and equilibrium experiments. The carbonate substituted hydroxylapatite was prepared by a precipitation method and characterized by SEM, XRD, FT-IR, TGA, BET and solubility measurements. The material is poorly crystalline, contains approximately 9.4% carbonate by weight and has a surface area of 210.2 m2/g. Uptake of selenite by the carbonated hydroxylapatite was approximately an order of magnitude higher than the uptake by uncarbonated hydroxylapatite reported in the literature. Distribution coefficients, Kd, determined for the carbonated apatite in this work ranged from approximately 4200 to over 14,000 L/kg. A comparison of the results from kinetic experiments performed in this work and literature kinetic data indicates the carbonated apatite synthesized in this study sorbed selenite 23 times faster than uncarbonated hydroxylapatite based on values normalized to the surface area of each material. The results indicate carbonated apatite is a potential candidate for use as a sorbent for pump-and-treat technologies, soil amendments or for use in permeable reactive barriers for the remediation of selenium contaminated sediments and groundwaters.

Water is the backbone of our economy - safe and adequate supplies of water are vital for agriculture, industry, recreation, and human consumption. While our supply of water today is largely safe and adequate, we as a nation face increasing water supply challenges in the form of extended droughts, demand growth due to population increase, more stringent health-based regulation, and competing demands from a variety of users. To meet these challenges in the coming decades, water treatment technologies, including desalination, will contribute substantially to ensuring a safe, sustainable, affordable, and adequate water supply for the United States. This overview documents Sandia National Laboratories' (SNL, or Sandia) Water Treatment Program which focused on the development and demonstration of advanced water purification technologies as part of the larger Sandia Water Initiative. Projects under the Water Treatment Program include: (1) the development of desalination research roadmaps (2) our efforts to accelerate the commercialization of new desalination and water treatment technologies (known as the 'Jump-Start Program),' (3) long range (high risk, early stage) desalination research (known as the 'Long Range Research Program'), (4) treatment research projects under the Joint Water Reuse & Desalination Task Force, (5) the Arsenic Water Technology Partnership Program, (6) water treatment projects funded under the New Mexico Small Business Administration, (7) water treatment projects for the National Energy Technology Laboratory (NETL) and the National Renewable Energy Laboratory (NREL), (8) Sandia- developed contaminant-selective treatment technologies, and finally (9) current Laboratory Directed Research and Development (LDRD) funded desalination projects.

As the title suggests, this report provides a summary of the status and progress for the Preliminary Design Concepts Work Package. Described herein are design concepts and thermal analysis for crystalline and salt host media. The report concludes that thermal management of defense waste, including the relatively small subset of high thermal output waste packages, is readily achievable. Another important conclusion pertains to engineering feasibility, and design concepts presented herein are based upon established and existing elements and/or designs. The multipack configuration options for the crystalline host media pose the greatest engineering challenges, as these designs involve large, heavy waste packages that pose specific challenges with respect to handling and emplacement. Defense-related Spent Nuclear Fuel (DSNF) presents issues for post-closure criticality control, and a key recommendation made herein relates to the need for special packaging design that includes neutron-absorbing material for the DSNF. Lastly, this report finds that the preliminary design options discussed are tenable for operational and post-closure safety, owing to the fact that these concepts have been derived from other published and well-studied repository designs.

The Waste Form Disposal Options Evaluation Report (SNL 2014) evaluated disposal of both Commercial Spent Nuclear Fuel (CSNF) and DOE-managed HLW and Spent Nuclear Fuel (DHLW and DSNF) in the variety of disposal concepts being evaluated within the Used Fuel Disposition Campaign. That work covered a comprehensive inventory and a wide range of disposal concepts. The primary goal of this work is to evaluate the information needs for analyzing disposal solely of a subset of those wastes in a Defense Repository (DRep; i.e., those wastes that are either defense related, or managed by DOE but are not commercial in origin). A potential DRep also appears to be safe in the range of geologic mined repository concepts, but may have different concepts and features because of the very different inventory of waste that would be included. The focus of this status report is to cover the progress made in FY16 toward: (1) developing a preliminary DRep included inventory for engineering/design analyses; (2) assessing the major differences of this included inventory relative to that in other analyzed repository systems and the potential impacts to disposal concepts; (3) designing and developing an on-line waste library (OWL) to manage the information of all those wastes and their waste forms (including CSNF if needed); and (4) constraining post-closure waste form degradation performance for safety assessments of a DRep. In addition, some continuing work is reported on identifying potential candidate waste types/forms to be added to the full list from SNL (2014 – see Table C-1) which also may be added to the OWL in the future. The status for each of these aspects is reported herein.

This report documents key elements of the conceptual design for deep borehole disposal of radioactive waste to support the development of a universal canister concept of operations. A universal canister is a canister that is designed to be able to store, transport, and dispose of radioactive waste without the canister having to be reopened to treat or repackage the waste. This report focuses on the conceptual design for disposal of radioactive waste contained in a universal canister in a deep borehole. The general deep borehole disposal concept consists of drilling a borehole into crystalline basement rock to a depth of about 5 km, emplacing WPs in the lower 2 km of the borehole, and sealing and plugging the upper 3 km. Research and development programs for deep borehole disposal have been ongoing for several years in the United States and the United Kingdom; these studies have shown that deep borehole disposal of radioactive waste could be safe, cost effective, and technically feasible. The design concepts described in this report are workable solutions based on expert judgment, and are intended to guide follow-on design activities. Both preclosure and postclosure safety were considered in the development of the reference design concept. The requirements and assumptions that form the basis for the deep borehole disposal concept include WP performance requirements, radiological protection requirements, surface handling and transport requirements, and emplacement requirements. The key features of the reference disposal concept include borehole drilling and construction concepts, WP designs, and waste handling and emplacement concepts. These features are supported by engineering analyses.

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The Old Rifle Site is a former vanadium and uranium ore-processing facility located adjacent to the Colorado River and approximately 0.3 miles east of the city of Rifle, CO. The former processing facilities have been removed and the site uranium mill tailings are interned at a disposal cell north of the city of Rifle. However, some low level remnant uranium contamination still exists at the Old Rifle site. In 2002, the United States Nuclear Regulatory Commission (US NRC) concurred with United States Department of Energy (US DOE) on a groundwater compliance strategy of natural flushing with institutional controls to decrease contaminant concentrations in the aquifer. In addition to active monitoring of contaminant concentrations, the site is also used for DOE Legacy Management (LM) and other DOE-funded small scale field tests of remediation technologies. The purpose of this laboratory scale study was to evaluate the effectiveness of a hydroxyapatite (Ca₁₀(PO₄)₆(OH)₂) permeable reactive barrier and source area treatment in Old Rifle sediments. Phosphate treatment impact was evaluated by comparing uranium leaching and surface phase changes in untreated to PO4-treated sediments. The impact of the amount of phosphate precipitation in the sediment on uranium mobility was evaluated with three different phosphate loadings. A range of flow velocity and uranium concentration conditions (i.e., uranium flux through the phosphate-treated sediment) was also evaluated to quantify the uranium uptake mass and rate by the phosphate precipitate.

This study considers the feasibility of large diameter deep boreholes for waste disposal. The conceptual approach considers examples of deep large diameter boreholes that have been successfully drilled, and also other deep borehole designs proposed in the literature. The objective for large diameter boreholes would be disposal of waste packages with diameters of 22 to 29 inches, which could enable disposal of waste forms such as existing vitrified high level waste. A large-diameter deep borehole design option would also be amenable to other waste forms including calcine waste, treated Na-bonded and Na-bearing waste, and Cs and Sr capsules.

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We have performed an initial evaluation and testing program to assess the effectiveness of a hydroxyapatite (Ca10(PO4)6(OH)2) permeable reactive barrier and source area treatment to decrease uranium mobility at the Department of Energy (DOE) former Old Rifle uranium mill processing site in Rifle, western Colorado. Uranium ore was processed at the site from the 1940s to the 1970s. The mill facilities at the site as well as the uranium mill tailings previously stored there have all been removed. Groundwater in the alluvial aquifer beneath the site still contains elevated concentrations of uranium, and is currently used for field tests to study uranium behavior in groundwater and investigate potential uranium remediation technologies. The technology investigated in this work is based on in situ formation of apatite in sediment to create a subsurface apatite PRB and also for source area treatment. The process is based on injecting a solution containing calcium citrate and sodium into the subsurface for constructing the PRB within the uranium plume. As the indigenous sediment micro-organisms biodegrade the injected citrate, the calcium is released and reacts with the phosphate to form hydroxyapatite (precipitate). This paper reports on proof-of-principle column tests with Old Rifle sediment and synthetic groundwater.

Sandia National Laboratories has collaborated with Pleasanton Ridge Research Company (PRRC) to determine whether Sandia Octahedral Molecular Sieves (SOMS) and modified SOMs materials can be synthesized in large batches and produced in granular form. Sandia National Laboratories tested these SOMS and its variants based in aqueous chemical environments for an application-based evaluation of material performance as a sorbent. Testing focused primarily on determining the distribution coefficients (K _d ) and chemical selectivity SOMs for alkali earth (Sr) ions in aqueous and dilute seawater solutions. In general the well-crystallized SOMS materials tested exhibited very high K d values (>10 ⁶ ) in distilled water but K _d values dropped substantially (%7E10 ² -10 ³ ) in the dilute seawater (3%). However, one set of SOMS samples (1.4.2 and 1.4.6) provided by PRRC yielded relatively high K d (approaching 10 ⁴ ) in dilute seawater. Further examination of these samples by scanning electron microscopy (SEM) revealed the presence of at least two phases at least one of which may be accounting for the improved K d values in dilute seawater. Evaluation of Strontium Selectivity by Sandia Octahedral Molecular Sieves (SOMS) January 20, 2016

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Deep Borehole Disposal (DBD) of high-level radioactive wastes has been considered an option for geological isolation for many years (Hess et al. 1957). Recent advances in drilling technology have decreased costs and increased reliability for large-diameter (i.e., ≥50 cm [19.7”]) boreholes to depths of several kilometers (Beswick 2008; Beswick et al. 2014). These advances have therefore also increased the feasibility of the DBD concept (Brady et al. 2009; Cornwall 2015), and the current field test, introduced herein, is a demonstration of the DBD concept and these advances.

The mission of the United States Department of Energy's Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and go vernment - sponsored nuclear energy re search. S ome of the waste s that that must be managed have be en identified as good candidates for disposal in a deep borehole in crystalline rock (SNL 2014 a). In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister - based system that can be used for handling these wastes during the disposition process (i.e., storage, transfers, transportation, and disposal) could facilitate the eventual disposal of these wastes. This report provides information for a program plan for developing specifications regarding a canister - based system that facilitates small waste form packaging and disposal and that is integrated with the overall efforts of the DOE's Office of Nuclear Energy Used Fuel Dis position Camp aign's Deep Borehole Field Test . Groundwork for Universal Ca nister System Development September 2015 ii W astes to be considered as candidates for the universal canister system include capsules containing cesium and strontium currently stored in pools at the Hanford Site, cesium to be processed using elutable or nonelutable resins at the Hanford Site, and calcine waste from Idaho National Laboratory. The initial emphasis will be on disposal of the cesium and strontium capsules in a deep borehole that has been drilled into crystalline rock. Specifications for a universal canister system are derived from operational, performance, and regulatory requirements for storage, transfers, transportation, and disposal of radioactive waste. Agreements between the Department of Energy and the States of Washington and Idaho, as well as the Deep Borehole Field Test plan provide schedule requirements for development of the universal canister system . Future work includes collaboration with the Hanford Site to move the cesium and strontium capsules into dry storage, collaboration with the Deep Borehole Field Tes t to develop surface handling and emplacement techniques and to develop the waste package design requirements, developing universal canister system design options and concepts of operations, and developing system analysis tools. Areas in which f urther research and development are needed include material properties and structural integrity, in - package sorbents and fillers, waste form tolerance to heat and postweld stress relief, waste package impact limiters, sensors, cesium mobility under downhol e conditions, and the impact of high pressure and high temperature environment on seals design.

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This report summarizes the assistance provided to Shafer Ranches, Inc., Hightower Ranch, and Western Environmental by Sandia National Laboratories under a Leveraged New Mexico Small Business Assistance grant. The work was conducted between April to November, 2014. Therefore, Sandia National Laboratories has been asked to investigate and develop a water treatment system that would result in reduced cost associated with infrastructure, maintenance, elimination of importing water, and improved cattle health.

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Nanomaterials and nanotechnology methods have been an integral part of international research over the past decade. Because many traditional water treatment technologies (e.g. membrane filtration, biofouling, scale inhibition, etc.) depend on nanoscale processes, it is reasonable to expect one outcome of nanotechnology research to be better, nano-engineered water treatment approaches. The most immediate, and possibly greatest, impact of nanotechnology on desalination methods will likely be the development of membranes engineered at the near-molecular level. Aquaporin proteins that channel water across cell membranes with very low energy inputs point to the potential for dramatically improved performance. Aquaporin-laced polymer membranes and aquaporin-mimicking carbon nanotubes and metal oxide membranes developed in the lab support this. A critical limitation to widespread use of nanoengineered desalination membranes will be their scalability to industrial fabrication processes. Subsequent, long-term improvements in nanoengineered membranes may result in self-healing membranes that ideally are (1) more resistant to biofouling, (2) have biocidal properties, and/or (3) selectively target trace contaminants.

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