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Oshiro, Jared M.

Publications and source records attributed to Oshiro, Jared M..

Low Activity Waste Glass Optimization with Property Models from Machine Learning, Part 2: Experimental Validation and Active Learning

The United States Department of Energy is responsible for managing legacy nuclear waste stored in underground tanks at the Hanford Site. To treat the waste, it is planned as the current baseline to separately vitrify low-activity waste (LAW) and high-level waste fractions. Previously, machine learning (ML) based glass property models (e.g., chemical durability, viscosity, electrical conductivity and SO3 solubility) were developed with prediction uncertainties. A waste glass optimization approach was then established to enable the capability of using these ML models in LAW glass formulation. In this study, the previous ML models were first experimentally validated, and the results were incorporated back into the database to update the ML models. The updated models and formulations showed increased waste loading while reducing the failure rate, demonstrating improved predictive accuracy, reduced uncertainties, and the effectiveness of active learning in guiding high-dimensional, nonlinear LAW glass design. This represents the first experimental validation of ML based LAW glass formulation, with practical benefits such as higher waste loading, shorter mission duration, and lower operational risk.

Lu, Xiaonan (ORCID:0000000179708148)↗

Pelletization with Spark Plasma Sintering and Characterization of Metal Iodides: An Assessment of Long-Term Radioiodine Immobilization Options

Four promising iodine “getter” materials (Ag, Cu, Bi, and Sn) for radioiodine capture were assessed in their pure metal-iodide (MI x ) pelletized forms to compare relative chemical durabilities. To study chemical durability, commercial MI x compounds of AgI, BiI 3 , BiOI, CuI, and SnI 4 were converted to dense monolithic pellets using spark plasma sintering. Semidynamic leach testing in the form of modified ASTM C1308 tests was then performed on the pellets in two different forms including unmounted (as-pressed) specimens (i.e., “U”) and epoxy-mounted specimens (i.e., “M”) with polished surfaces. The chemical durability results and sample characterizations showed that three of the five MI x compounds tested (i.e., AgI, CuI, and BiOI) displayed moderate to high leach resistances. Further, the remaining two MI x compounds (i.e., BiI 3 and SnI 4 ), which are both desirable iodine waste forms due to their high iodine loading capacities, readily decomposed during leach testing, indicated by crystallographic changes in the specimens as well as large amounts of iodine detected in the leachate solutions. The instabilities of BiI 3 and SnI 4 raise uncertainties for using the base metals/cations (i.e., Bi 0 /Bi 3+ and Sn 0 /Sn 4+ , respectively) as viable getters for radioiodine capture due to likely poor waste form chemical durabilities after capture and consolidation into waste forms.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Direct Feed High-Level Waste APPS Model Glass Testing (DFHLW APPS) Matrix

This report summarizes the data collected during the batching and melting of the Direct Feed High-Level Waste APPS Model Glass Matrix (DFHLW APPS) to serve as a quality-assured validation of the Aspen Process Performance Simulation (APPS) formulation method. Of 15 glasses tested, 12 satisfied all target property constraints. Two glasses, APPS-05 and -06, formed nepheline on canister centerline cooling heat-treatment and failed the Product Consistency Test response limits. Glass APPS-07-2 formed unacceptably high concentrations of crystals (primarily Na3Nd(PO4)2) when heat treated at 950 °C. All other glasses were found to be satisfactory. The measured property values were compared to predicted values from a set of current models. In many cases the current models were found to be inadequate for design of DFHLW glasses. These models are being adjusted to correct for mispredictions. Other models, e.g., density, toxicity characteristic leaching procedure, and sulfur solubility, are adequate for formulation of DFHLW glasses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Evaluating Iodine Immobilization Technologies: Cermets, Polycermets, and Polyhalmets

The work in this report documents the efforts conducted to assess the feasibility of some of the ideas documented in Pacific Northwest National Laboratory invention disclosure reports (IDRs) including: 1) Iodine capture in polyacrylonitrile (PAN)-containing composite sorbents (32451-E). In this work, the composites evaluated included Ag0, Bi0, Cu0, Bi2S3, and Cu2S embedded in PAN. 2) Metal iodide removal from these sorbents through dissolution in dimethyl sulfoxide (DMSO) (32729-E). In this work, PAN dissolution was evaluated for multiple types of sorbents including Ag-Pan, Bi-PAN, Cu-PAN, Bi2S3-PAN, and Cu2S-PAN. 3) Using metal-sulfide sorbents for iodine capture (32647-E). In this work, the composites evaluated under this IDR included Ag2S, Bi2S3, and Cu2S embedded in PAN. 4) Using low-melting metals to immobilize (encapsulate) iodine-loaded and polymer-containing sorbents into polymer-ceramic-metal (called polycermet) or polymer-halide-metal (called polyhalmet) composite waste forms (32625-E). In this work, the iodine-loaded PAN composites included AgI-PAN, BiI-PAN, and CuI-PAN. 5) Ceramic-metal composite waste form synthesis of polymer-containing materials using low-melting metals like bismuth, tin, or bismuth-tin alloys (32537-E). In this work, the metals evaluated included Bi, 58Bi-42Sn eutectic. 6) Cermets for immobilizing commercial sorbents loaded with radioiodine (32806-E). In this work, AgIX (iodine-loaded silver faujasite zeolite) was evaluated in cermet form.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Glass formulation and lab-scale testing of glasses designed for in-can Melter and in-container Vitrification of high-assay low-enriched uranium aqueous polishing Raffinate waste

Glasses were designed for processing a nuclear waste from aqueous polishing of high-assay low-enriched uranium using either In-Can Melter (ICM) or GeoMelt® In-Container Vitrification™ (ICV) technologies, which operate at temperatures of T p ≤ 1100 °C and T p ≤ 1450 °C, respectively. Due to the different operating conditions, the melt and glass properties were optimized differently for each technology. Each glass was designed to optimize for maximum waste loading while simultaneously satisfying processing (e.g., crystallization, viscosity, and conductivity) and product quality (e.g., durability, hazard characteristic, and crystal content) constraints. Here, the raffinate waste contains high nitric acid (4 M) and low total solids (9 g·L -1 ) concentrations. Feed preparation processes were tested to facilitate concentration and nitrate destruction/removal while controlling redox of the melter feed. A successful feed process including sugar addition and spray-drying was performed to generate an adequate melter feed for ICV processing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Ceramic–Metal (Cermet) Composites: A Review of Key Properties and Synthesis Methods Focused on Nuclear Waste Immobilization

Here this paper reviews key properties, applications, and examples of ceramic-metal composites (cermets), and metal matrix composites (MMC) with emphasis on their applicability as waste forms for immobilizing nuclear waste. While the literature is mature for vitrified and cementitious radioactive waste forms, cermet materials have not received adequate attention as potential candidates for immobilizing nuclear waste. To promulgate this effort, this review connects cermet and MMC design, such as hardened tools, with the chemistry of radioactive waste streams. A discussion on certification and qualification standards for cermet waste forms, literature gaps, and “how-to” sections on cermet processing techniques. Key parameters discussed include thermal conductivity, chemical durability, and waste loading, as well as examples for metal-containing waste forms. As cermet waste forms gain momentum within the community, this review paper aims ensure the end-of-life nuclear fuel cycle is addressed from a materials and waste disposition perspective.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗