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DOE OSTI2019

Dechlorination of molten chloride waste salt from electrorefining via ion-exchange using pelletized ultra-stable H-Y zeolite in a fluidized particle reactor

Dechlorination of eutectic LiCl–KCl based electrorefner (ER) salt is reported via ion-exchange reaction with protonated ultrastable Y-type (USHY) zeolite bound into mechanically fuidized 45–250 μm diameter particles. Evidence of exchange of cations from the salt (Li + , K + , and fission product cations) into the zeolite lattice replacing H + ions was found based on a change in unit cell size, ICP-MS, XRD and TEM–EDS in addition to detection of HCl of gas. Ion exchange reaction was carried out at 625 and 650 °C, temperatures above the melting point of eutectic LiCl–KCl. Experiments were carried out to optimize zeolite drying temperature, estimate maximum ion-exchange capacity, and determine the thermal stability of USHY zeolite. The results indicate over 90% dechlorination can be achieved without zeolite structure collapse at 625 °C. Finally, this provides a promising route to stabilizing waste from radioactive chloride salts into dechlorinated waste forms for permanent geologic disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

K West Basin Sand Filter Backwash Sample Analysis

A sand filter is used to help maintain water clarity at the K West Basin where highly radioactive sludge is stored. Eventually that sand filter will require disposal. The radionuclide content of the solids trapped in the sand filter will affect the selection of the sand filter disposal pathway. The Pacific Northwest National Laboratory (PNNL) was contracted by the K Basin Operations & Plateau Remediation Project (operations contractor CH2M Hill) to analyze the radionuclide content of the solids collected from the backwash of the K West Basin sand filter. The radionuclide composition in the sand filter backwash solids will be used by CH2M Hill to determine if the sand filter media and retained sludge solids will be designated as transuranic waste for disposal purposes or can be processed through less expensive means. On October 19, 2015, K Basin Operations & Plateau Remediation Project staff backwashed the sand filter into the North Load-Out Pit (NLOP) and immediately collected sample slurry from a sampling tube positioned 24 in. above the NLOP floor. The 764 g sand filter backwash slurry sample, KW-105 SFBW-001, was submitted to PNNL for analysis on October 20, 2015. Solids from the slurry sample were consolidated into two samples (i.e., a primary and a duplicate sample) by centrifuging and measured for mass (0.82 g combined – wet centrifuged solids basis) and volume (0.80 mL combined). The solids were a dark brown/orange color, consistent with iron oxide/hydroxide. The solids were dried; the combined dry solids mass was 0.1113 g, corresponding to 0.0146 weight percent (wt%) solids in the original submitted sample slurry. The solids were acid-digested using nitric and hydrochloric acids. Insoluble solids developed upon dilution with 0.5 M HNO 3 , corresponding to an average 6.5 wt% of the initial dry solids content. The acid digestate and insoluble solids were analyzed separately by gamma spectrometry. Nominally, 7.7% of the 60 Co was present in the insoluble solids; less than 1% of other gamma-emitters (i.e., 137 Cs, 154/155 Eu, and 241 Am) were present in the insoluble solids. Aliquots of the acid digestate were analyzed directly using gamma energy analysis (GEA) and after separations for 238 Pu, 239+240 Pu, 237 Np, and 241 Am radioisotopes using alpha energy analysis (AEA). The 90 Sr was measured by liquid scintillation counting (LSC) on the Sr-separated fraction. The plutonium isotopic distribution of the acid digestate was analyzed following Pu separations by thermal ionization mass spectrometry (TIMS). A table summarizes the results for the primary and duplicate samples. The 239+240 Pu concentration (µCi/g dry) relative to 90 Sr and to 137 Cs concentrations (µCi/g dry) was examined. The K West Basin sludge has a 239+240 Pu/ 90 Sr ranging from 0.1 to 1.2 and the 239+240 Pu/ 137 Cs ratio ranging from 0.10 to 0.47. In contrast, the sand filter backwash solids 239+240 Pu/ 90 Sr ratio was 10.6 and the 239+240 Pu/ 137 Cs ratio was 2.0. The ratio differences indicate a relative enhancement of the Pu concentration in the sand filter solids relative to the 137 Cs and 90 Sr sludge concentrations currently in the K West Basin. A dose-to-curie radioisotope evaluation of the sand filter waste form may need to consider this dissimilarity.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Follow-on Report of Analysis of Approaches to Supplemental Treatment of Low–Activity Waste at the Hanford Nuclear Reservation (Volumes I & II)

The Hanford Site, in southeast Washington State, is preparing to disposition approximately 56,000,000 gallons (56 Mgal) of radioactive and chemically hazardous wastes currently stored in underground tanks at the site. Tank wastes will be divided into a high-activity fraction and a low-activity fraction for subsequent treatment and disposition. A waste processing and treatment facility, the Waste Treatment and Immobilization Plant (WTP), will include the high-level waste (HLW) vitrification facility (WTP HLW Vitrification Facility) for immobilizing the high-activity fraction and a low-activity waste (LAW) vitrification facility (WTP LAW Vitrification Facility) for immobilizing the low-activity fraction. Both facilities will use vitrification technology to immobilize the Hanford tank wastes in a glass waste form.

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DOE OSTI2021

Applying laboratory methods for durability assessment of vitrified material to archaeological samples

Abstract Laboratory testing used to assess the long-term chemical durability of nuclear waste forms may not be applicable to disposal because the accelerated conditions may not represent disposal conditions. To address this, we examine the corrosion of vitrified archeological materials excavated from the near surface of a ~1500-year old Iron Age Swedish hillfort, Broborg, as an analog for the disposal of vitrified nuclear waste. We compare characterized site samples with corrosion characteristics generated by standard laboratory durability test methods including the product consistency test (PCT), the vapor hydration test (VHT), and the EPA Method 1313 test. Results show that the surficial layer of the Broborg samples resulting from VHT displays some similarities to the morphology of the surficial layer formed over longer timescales in the environment. This work provides improved understanding of long-term glass corrosion behavior in terms of the thickness, morphology, and chemistry of the surficial features that are formed.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Supported liquid inorganic membranes for nuclear waste separation

A system and method for the extraction of americium from radioactive waste solutions. The method includes the transfer of highly oxidized americium from an acidic aqueous feed solution through an immobilized liquid membrane to an organic receiving solvent, for example tributyl phosphate. The immobilized liquid membrane includes porous support and separating layers loaded with tributyl phosphate. The extracted solution is subsequently stripped of americium and recycled at the immobilized liquid membrane as neat tributyl phosphate for the continuous extraction of americium. The sequestered americium can be used as a nuclear fuel, a nuclear fuel component or a radiation source, and the remaining constituent elements in the aqueous feed solution can be stored in glassified waste forms substantially free of americium.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Evaluation of Nuclear Spent Fuel Disposal in Clay-Bearing Rock - Process Model Development and Experimental Studies

This report represents the milestone deliverable M2SF-23SN010301072 “Evaluation of Nuclear Spent Fuel Disposal in Clay-Bearing Rock - Process Model Development and Experimental Studies” The report provides a status update of FY23 activities for the work package Argillite Disposal work packages for the DOE-NE Spent Fuel Waste Form Science and Technology (SFWST) Program. Clay-rich geological media (often referred as shale or argillite) are among the most abundant type of sedimentary rock near the Earth’s surface. Argillaceous rock formations have the following advantageous attributes for deep geological nuclear waste disposal: widespread geologic occurrence, found in stable geologic settings, low permeability, self-sealing properties, low effective diffusion coefficient, high sorption capacity, and have the appropriate depth and thickness to host nuclear waste repository concepts. The DOE R&D program under the Spent Fuel Waste Science Technology (SFWST) campaign has made key progress (through experiment, modeling, and testing) in the study of chemical and physical phenomena that could impact the long-term safety assessment of heat-generating nuclear waste disposition in clay/shale/argillaceous rock. International collaboration activities comprising field-scale heater tests, field data monitoring, and laboratory-scale experiments provide key information on changes to the engineered barrier system (EBS) material exposed high thermal loads. Moreover, consideration of direct disposal of large capacity dual-purpose canisters (DPCs) as part of the back-end SNF waste disposition strategy has generated interest in improving our understanding of the effects of elevated temperatures on the engineered barrier system (EBS) design concepts. Chemical and structural analyses of sampled bentonite material from laboratory tests at elevated temperatures are key to the characterization of thermal effects affecting bentonite clay barrier performance. The knowledge provided by these experiments is crucial to constrain the extent of sacrificial zones in the EBS design during the thermal period. Thermal, hydrologic, mechanical, and chemical (THMC) data collected from heater tests and laboratory experiments have been used in the development, validation, and calibration of THMC simulators to model near-field coupled processes. This information leads to the development of simulation approaches to assess issues on coupled processes involving porous media flow, transport, geomechanical phenomena, chemical interactions with barrier/geologic materials, and the development of EBS concepts. These lines of knowledge are central to the design of deep geological backfilled repository concepts where temperature plays a key role in the EBS behavior, potential interactions with host rock, and long-term performance in the safety assessment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Solid-liquid Separation Testing for the Remediation of Hanford Waste Treatment Plant Low Activity Waste Melter Off-Gas Condensate

The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate stream (LAW Off-Gas Condensate) from the off-gas system. The plan for disposition of this stream during Direct Feed LAW operations is to transfer it to the Effluent Management Facility (EMF), where it will be concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. However, this stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the quantity of glass waste produced by allowing further increases in waste loading, resulting in a shortened LAW vitrification mission duration. The concept being tested here involves selectively removing the 99 Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed as a different waste form and avoid reprocessing through vitrification.

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DOE OSTI2021

A Focused Ion Beam-Scanning Transmission Electron Microscopy with Energy-Dispersive X-ray Spectroscopy Study on Technetium Incorporation within Iron Oxides through Fe(OH) 2 (s) Mineral Transformation

Incorporation of Tc(IV) into iron oxide/hydroxide minerals has been explored and proposed as a promising pathway to preventing Tc(IV) reoxidation to environmentally mobile pertechnetate (TcO 4 - ) and improving long-term immobilization of radioactive technetium-99 (Tc). However, visual evidence evaluating the distribution of Tc(IV) incorporated within iron oxide/hydroxide phases has not been available, until now, despite potential implications on Tc(IV) stability within the host phase. For the purpose of this study Tc(IV) incorporation into iron oxide/hydroxide phases was facilitated via Fe(OH) 2 (s) oxidation and mineral transformation to magnetite (Fe 3 O 4 ). Focused ion beam - scanning transmission electron microscopy equipped with energy-dispersive X-ray spectroscopy (FIB/STEM-EDS) methods were then combined with X-ray diffraction and absorption spectroscopy techniques to characterize and visually demonstrate that, for the first time, Tc(IV) is heterogeneously incorporated into different iron oxide/hydroxide phases as Tc(IV)-incorporated magnetite and/or TcO 2 ·2H 2 O(s) via different incorporation mechanisms. Heterogeneous distribution of Tc(IV) in magnetite suggests either (i) TcO4- is reduced quickly at the magnetite surface and then encapsulated into magnetite during continued octahedral crystal growth, or (ii) Tc(IV) alternatively partitioned into multiple layers of a blocky, plate-like morphological magnetite structure showing stratified Tc. With limited Tc-hematite (Fe 2 O 3 ) incorporation, the results suggest that TcO 2 ·2H 2 O(s) is formed and mainly associated/embedded in fibrous nanometer-sized polycrystalline hematite. This work highlights the power of modern state-of-the-art FIB/STEM-EDS approach to provide essential visual insights of the Tc-iron oxide/hydroxide incorporation and generate reliable mechanism-informed designs for waste forms relying on Tc mineral incorporation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2021

Annual Status Update for OWL

This report represents completion of milestone deliverable M2SF-22SN010309082 Annual Status Update for OWL, which is due on November 30, 2021 as part of the fiscal year 2022 (FY2022) work package SF-22SN01030908. This report provides an annual update on status of FY2021 activities for the work package “OWL - Inventory – SNL”. 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), DOE-managed spent nuclear fuel (DSNF), and other wastes that are likely candidates for deep geologic disposal. Links to the current supporting documents for the data are provided when possible; however, 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 report contains new information on sodium-bonded spent fuel waste types and wastes forms, which are included in the next release of OWL, Version 3.0, on the Sandia National Laboratories (SNL) External Collaboration Network (ECN). The report also provides an update on the effort to include information regarding the types of vessels capable of disposing of DOE-managed waste.

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DOE OSTI2017

Enhanced LAW Glass Correlation - Phase 2 (Final Report, Rev. 0)

The Office of River Protection is examining options to optimize the LAW facility and LAW glass waste form. VSL and Atkins have evaluated several potential incremental improvements for ORP in support of its evaluation of WTP LAW facility optimization. The objective of the present work was to continue the development of an Enhanced LAW Glass Correlation that can be used to determine the appropriate waste loading, glass composition, and amounts and types of glass former additives for LAW processing at the WTP for any given LAW composition that is delivered to the LAW vitrification facility. The intent is that this correlation, along with the enhanced LAW glass property-composition models, can be used for LAW processing at the WTP with the higher waste loading LAW glass formulations that have been developed for ORP at the VSL.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Cradle to grave: the importance of the fuel cycle to molten salt reactor sustainability

Advanced reactor technologies are being considered for the next-generation of nuclear power plants. These plants are designed to have a smaller footprint, run more efficiently at higher temperatures, have the flexibility to meet specific power or heating needs, and have lower construction costs. This paper offers a perspective on molten salt reactors, promoted as having a flexible fuel cycle and close-to-ambient pressure operation. A complexity introduced by reducing the reactor footprint is that it may require low-enriched fuel for efficient operation, available from enrichment of the feed salt or by reusing actinides from existing used nuclear fuel (UNF). Recycling UNF has the potential to reduce high-level waste, if done correctly. Release limits from UNF processing are stringent, and processes for waste reduction, fission gas trapping, and stable waste-form generation are not yet ready for commercial deployment. These complex processes are expensive to develop and troubleshoot because the feed is highly radioactive. Thus, fuel production and supply chain development must keep abreast of reactor technology development. Another aspect of reactor sustainability is the non-fuel waste streams that will be generated during operation and decommissioning. Some molten salt reactor designs are projected to have much shorter operational lifetimes than light-water reactors: less than a decade. A goal of the reactor sustainability effort is to divert these materials from a high-level waste repository. However, processing of reactor components should only be undertaken if it reduces waste. Economic and environmental aspects of sustainability are also important, but are not included in this perspective.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Iodine Sorbent Performance in FY 2012 Deep Bed Tests

Nuclear fission results in the production of fission products and activation products, some of which tend to be volatile during used fuel reprocessing and evolve in gaseous species into the reprocessing facility off-gas systems. Analyses have shown that I-129, due to its radioactivity, high potential mobility in the environment, and high longevity (half life of 15.7 million years), can require control efficiencies of up to 1,000x or higher to meet regulatory emission limits. Iodine capture is an important aspect of the Separations and Waste Forms Campaign Off-gas Sigma Team (Jubin 2011, Pantano 2011). Deep-bed iodine sorption tests for both silver-functionalized Aerogel and silver zeolite sorbents were performed during Fiscal Year 2012. These tests showed that: • Decontamination factors were achieved that exceed reasonably conservative estimates for DFs needed for used fuel reprocessing facilities in the U.S. to meet regulatory requirements for I-129 capture. • Silver utilizations approached or exceeded 100% for high inlet gas iodine concentrations, but test durations were not long enough to approach 100% silver utilization for lower iodine concentrations. • The depth of the mass transfer zone was determined for both low iodine concentrations (under 10 ppmv) and for higher iodine concentrations (between 10-50 ppmv); the depth increases over time as iodine is sorbed. • These sorbents capture iodine by chemisorption, where the sorbed iodine reacts with the silver to form very non-volatile AgI. Any sorbed iodine that is physisorbed but not chemically reacted with silver to form AgI might not be tightly held by the sorbent. The portion of sorbed iodine that tends to desorb because it is not chemisorbed (reacted to form AgI) is small, under 1%, for the AgZ tests, and even smaller, under 0.01%, for the silver-functionalized Aerogel.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Analysis Results for Sub-Samples of SRS Tanks 30, 32, and 39 to Support Evaluations of the I-129 Inventory

In order to appropriately model and predict the chemical integrity and performance of cementitious materials used for waste immobilization at the Savannah River Site (SRS), it is critical to understand the I-129 solubility and distribution within the tank farm. Iodine in radioactive waste and in environmental media is typically highly mobile and long lived. Iodine is ubiquitous in SRS tank waste and waste forms. The iodine is assumed to be soluble and present at low levels in Performance Assessments (PAs) for SRS Tank Farms, and is one of the dose drivers in the PAs for both the SRS Salt Disposal Facility (SDF) and the H-Area Tank Farm (HTF). Analysis of tank waste samples is critical to understanding the Tank Farm iodine inventory and reducing disposal uncertainty. Higher than expected iodine levels have recently been observed in residual solids isolated from some SRS tanks prior to closure, indicating uncertainty regarding the chemical species involved. If the iodine inventory uncertainty is larger than anticipated, future work may be necessary to reduce the uncertainty. This memorandum satisfies a portion of the work scope identified in Task Plan SRNL-RP-2016-00651. A separate memorandum issued previously, reported historical unpublished I-129 data, a significant portion of which was below detectable analytical limits. This memorandum includes iodine and general chemical analysis results for six archived SRNL samples which were previously reported to have I-129 concentrations below detectable limits. Lower sample dilution factors were used for the current analyses in order to obtain concentrations above detection. The samples analyzed included surface and depth samples from SRS tanks 30, 32, and 39.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2026

A Brief Review of Technologies for Separating Tritium from the Aqueous Effluent of Reprocessing Plants

The scope of this letter report encompasses a review of technologies applicable to separating tritiated water from light and heavy water with a focus on wastewater arising from the reprocessing of UNF. Reports considering flowsheet options for tritium management are also reviewed. The main objective of this letter report is to support consideration of programmatic opportunities for the Materials Recovery and Waste Form Development (MRWFD) Campaign within the Department of Energy’s Office of Nuclear Energy.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Using Cosmic Ray Muons to Assess Geological Characteristics in the Subsurface

This is a technical presentation that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

System Analysis Modeling and Intermodal Transportation for Commercial Spent Nuclear Fuel

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Deep Borehole Disposal Safety Case

The safety case for deep borehole disposal of nuclear wastes contains a safety strategy, an assessment basis, and a safety assessment. The safety strategy includes strategies for management, siting and design, and assessment. The assessment basis considers site selection, pre-closure, and post-closure, which includes waste and engineered barriers, the geosphere/natural barriers, and the biosphere and surface environment. The safety assessment entails a pre-closure safety analysis, a post-closure performance assessment, and confidence enhancement analyses. This work outlines the assessment basis and safety assessment aspects of a deep borehole disposal safety case. The safety case presented here is specific to deep borehole disposal of Cs and Sr capsules, but is generally applicable to other waste forms, such as spent nuclear fuel. The safety assessments for pre-closure and post-closure are briefly summarized from other sources; key issues for confidence enhancement are described in greater detail. These confidence enhancement analyses require building the technical basis for geologically old, reducing, highly saline brines at the depth of waste emplacement, and using reactive-transport codes to predict their movement in post-closure. The development and emplacement of borehole seals above the waste emplacement zone is also important to confidence enhancement.

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