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At least 217 records · Page 12

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DOE OSTI2020

Rheology and Flow Evaluation of Neutralized Sodium Reactor Experiment Fuel with Manganous Nitrate

H-Canyon is preparing the Sodium Reactor Experiment (SRE) solutions in Tanks 16.3 and 16.4 for discard to the Savannah River Site (SRS) High Level Waste (HLW) Tanks into Sludge Batch (SB) 10. To meet HLW's criticality requirements, manganese will be added to the SRE solutions. The addition of manganese to the existing thorium and uranium in the SRE solution raised concerns with the flow of this neutralized material as it is discharged from H-Canyon through the gravity drain system to the H-Area Pump Pit (HPP). A neutralized Mn adjusted SRE stream will produce primarily Mn, Th, and U solids that can deter transfer. A 2012 rheology study was completed as part of the flowsheet development.1 However, the initial rheology study did not include the addition of manganous nitrate as a poison. Therefore, H-Canyon Engineering requested Savannah River National Laboratory (SRNL) to determine if the neutralized, Mn-adjusted SRE solutions will flow through the waste header to the HPP. The H-Canyon Technical Task Request (TTR) specified a target of 80 to 1 to bound the uncertainty in Mn target. Parallel studies were being performed to ensure that freshly precipitated Mn did not have a solubility that would result in challenging the DWPF WAC requirements of 70 to 1. This task was requested via a TTR and is governed by a Task Technical and Quality Assurance Plan (TTQAP).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Investigation of Optical, Physical and Corrosion Properties of Gamma-Irradiated International Simple Glass

A simulated high-level waste borosilicate glass was irradiated with gamma dose to simulate in-situ radiation exposure and damage in the glass. After irradiation, several measurements of optical, physical and corrosion properties were determined to compare with the unirradiated glass.

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DOE OSTI2020

Vitrification of High-Cr Glass in Research-Scale Melter

This test report describes the experimental results from a small-scale test using the research-scale melter (RSM) at Pacific Northwest National Laboratory to demonstrate processing of a high-Cr simulated feed stream, designated HLW-HCr-16. The RSM is a small, joule-heated melter capable of processing melter feed continuously. The melter is equipped with Inconel ® 693 electrodes, Monofrax ® K-3 refractory, and an Inconel 690 pour spout. An electric kiln surrounds the melter body and minimizes heat loss from the melter body during operation. The RSM is equipped with an offgas treatment system that employs quenching, wet scrubbing, and high-efficiency mist elimination. The glass-discharge section is heated to facilitate pouring of the glass. The melter is fitted with a melt cavity that is ~25 cm (10 in.) in diameter with a nominal glass depth of 8.9 cm (3.5 in.). The melter was operated with a target glass temperature of 1150°C and target plenum temperature between 550°C and 700°C for this test. The air bubbling rate was 4.2 L/min. Overall, during the continuous operation of the melter for ~ 103 hours, ~ 141 kg of glass was produced. At the conclusion of the test, the melter and exhaust lines were visually inspected for particulate deposition and corrosion. Entrained material had adhered to the underside of the melter lid and to the exhaust piping. Enrichments in elements such as Cl, F, B, K, Li, P, Na, and S were measured in these deposits through inductively coupled plasma–optical emission spectroscopy and X-ray fluorescence analysis. When the melter electrodes and air bubbler tube were removed from the glass in the RSM, the electrodes appeared discolored, but no significant loss of metal was observed. The processing of a high-Cr simulant, HLW-HCr-16, in the RSM produced glass at an average rate of 1.36 kg/h, equaling a melter-surface-area normalized glass generation rate of 654 kg/day/m 2 . The resulting glass met the toxicity characteristic leaching procedure requirement. Test results of crystallinity, electrical conductivity, and viscosity showed good processing properties of this high-Cr high-level waste glass. RSM offgas was also sampled and analyzed at periodic intervals during steady-state operating conditions. The total decontamination factor averaged by four sampling periods was 134. The concentrations of CO and NO x in emissions were 237 to 422 and 69 to 94 parts per million by volume, respectively.

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DOE OSTI2021

Alternative Liquid DSA Containers for Sludge Slurry Storage in the SRNL Shielded Cells

A pending revision of the Documented Safety Analysis (DSA) for the Savannah River National Laboratory (SRNL) defines “DSA containers'’ for liquid sludge slurries as containers not susceptible to a flashing spray release in a fire and requires that a list of DSA containers be maintained. High Level Waste (HLW) samples are typically received from the Savannah River Site (SRS) tank farm after retrieval from the storage tanks. Glass bottles are preferred for long-term storage of radioactive liquid sludge slurry samples but are not on the DSA container list. An evaluation has been conducted to identify vessel lids for the glass bottles that would vent prior to pressurization and liquid superheating in a fire event. Low melting Field's Metal eutectic alloy with a melting point of 62°C has been identified as a preferred material of construction for vessel lids for this application because the material is resistant to puncture and is expected to have good chemical compatibility and radiation stability in the shielded cells environment. A test vessel was designed to evaluate the alloy and confirm that it would melt before the water in the vessel reaches the boiling point. The tests confirmed: 1) that the reported melting point of the alloy was accurate, 2) that a top fashioned from this material will melt creating an open atmosphere in the vessel headspace prior to sample boiling under moderate and fast heating rates, and 3) that nonuniform heating of the bottle from the bottom also results in melting of the alloy prior to boiling. Based on the results, neither an engulfing fire with standing or toppled bottles or a fire producing localized heat at the vessel bottom would result in flashing spray conditions. It is recommended that glass vessels of various volumes ranging from 125 mL to 1 gallon with modified tops containing the specified Field's Metal eutectic alloy which are appropriately designed to avoid flashing spray material releases be added to the DSA container list. These vessels will be utilized in the Shielded Cells for the storage of radioactive liquid samples of sludge slurries. Use of these vessels will also be advantageous for other liquid samples containing accountable amounts of nuclear material throughout other SRNL laboratories. Similarly constructed paraffin vessel lids could also potentially be added to the approved list, though additional testing would be required.

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DOE OSTI2021

Alternative Liquid DSA Containers for Sludge Slurry Storage in the SRNL Shielded Cells

A pending revision of the Documented Safety Analysis (DSA) for the Savannah River National Laboratory (SRNL) defines “DSA containers” for liquid sludge slurries as containers not susceptible to a flashing spray release in a fire and requires that a list of DSA containers be maintained. High Level Waste (HLW) samples are typically received from the Savannah River Site (SRS) tank farm after retrieval from the storage tanks. Glass bottles are preferred for long-term storage of radioactive liquid sludge slurry samples but are not on the DSA container list. An evaluation has been conducted to identify vessel lids for the glass bottles that would vent prior to pressurization and liquid superheating in a fire event. Low melting Field’s Metal eutectic alloy with a melting point of 62 °C has been identified as a preferred material of construction for vessel lids for this application because the material is resistant to puncture and is expected to have good chemical compatibility and radiation stability in the shielded cells environment. A test vessel was designed to evaluate the alloy and confirm that it would melt before the water in the vessel reaches the boiling point. The tests confirmed: 1) that the reported melting point of the alloy was accurate, 2) that a top fashioned from this material will melt creating an open atmosphere in the vessel headspace prior to sample boiling under moderate and fast heating rates, and 3) that nonuniform heating of the bottle from the bottom also results in melting of the alloy prior to boiling. Based on the results, neither an engulfing fire with standing or toppled bottles or a fire producing localized heat at the vessel bottom would result in flashing spray conditions.

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DOE OSTI2021

Appropriateness and Readiness of Cold Crucible Vitrification of Calcine Solids at the Idaho Site and H-Canyon Effluent at the Savannah River Site

This report summarizes the results from an assessment of the technology readiness and deployment of cold crucible induction melter (CCIM) technology to vitrify high level waste (HLW) calcine solids at the Idaho Site and H-Canyon liquid effluent at the Savannah River Site (SRS). The assessment was requested by the Department of Energy’s Office of Environmental Management (DOE-EM), but is not a DOE 413-3-4a assessment for technology deployment. A joint Savannah River National Laboratory (SRNL) and Fluor Idaho Cleanup Project (ICP) team with subject matter experts evaluated the existent literature and experience both within the DOE as well as relevant external experience. The technology assessment focused on the site-specific technology readiness and appropriateness based on a number of system factors including feasibility and appropriateness of the technology for the specific site application, the potential extent of the regulatory and design challenges, and evaluation of the ability to deploy within the treatment needs of each site. This initial assessment provides information to support DOE-EM in making a go/no-go decision to carry out additional work on the technology maturation for critical decisions and planning.

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DOE OSTI2021

Characterization of Properties for a Gamma–Irradiated ‘Revised’ International Simple Glass

Simplified surrogate high-level waste glasses from recent national and international testing were irradiated to gamma doses of 100 Mrad and 200 Mrad. Properties of the irradiated glass were measured for comparison to the pristine unirradiated glass. Glass transition temperatures as well as infrared and Raman spectroscopy measurements show no change in the irradiated glass. Short-term aqueous corrosion tests on the glass powders also indicate no measurable differences comparing the irradiated to pristine samples. Some corrosion differences were observed in the pristine glasses due to slight compositional changes. As expected from the observed radiation darkening of the glass, absorbance spectra of the irradiated glass with the pristine glass used as LDRD–2021–00341 LDRD External Report Summary reference indicates a broad absorbance over the range of 400 to 700 nm that is centered around 550-600 nm.

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DOE OSTI2021

VERSE-LC TCCR Model Calibration using Batch 1A, Batch 2, and Batch 3 Column Data

The objective of this work is to utilize on-line gamma monitoring data collected during Tank Closure Cesium Removal (TCCR) Column Batch 3 operations to calibrate the parameters (essentially, the correction/dilution factors (CF or DF)) used in the VERSE-LC models for the prediction of cesium breakthrough while processing Savannah River Site (SRS) Tank 10H dissolved saltcake High-Level Waste through a packed cylindrical bed of Crystalline Silicotitanate (CST). Calibrated VERSE-LC models could be applied to aid in evaluating future TCCR operations.

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DOE OSTI2023

Evaluation Of Glass Density to Support the Estimation of Fissile Mass Loadings in Sludge Batch 10 Glasses

Per a directive from the Department of Energy Savannah River Operations Office (DOE-SR) in 2008, the fissile mass loading concentration must remain below 897 g/m 3 in each high-level waste (HLW) glass canister produced by the Defense Waste Processing Facility (DWPF). To support Sludge Batch 5 (SB5) processing, the Savannah River National Laboratory (SRNL) developed a technical basis that facilitates the evaluation of fissile mass loading of the glass product. The calculation is based on the iron (Fe) concentration in the glass as determined by measurements from the Slurry Mix Evaporator acceptability analysis as well as the glass density. In April 2022, a subsequent DOE-SR directive increased the fissile mass loading limit to 2500 g/m 3 beginning with Sludge Batch 11. Thus, the 897 g/m 3 limit still applies to Sludge Batch 10 (SB10) processing. For SB5 through initial Sludge Batch 9 (SB9) processing prior to coupled operation with the Salt Waste Processing Facility (SWPF), SRNL provided DWPF a bounding glass density value that was based on a statistical evaluation of density measurements. To eliminate the need for experimental work, a composition-based density model for HLW glasses was developed at SRNL in 2019. The objective of this report is to present the bounding glass density determined with the composition-based density model for SB10 sludge-only (SO) and coupled processing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Evaluation of Glass Density to Support the Estimation of Fissile Mass Loadings in Sludge Batch 10 Glasses

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DOE OSTI2024

Enabling in-situ LIBS measurements of liquids and slurries

The aim of this work was to explore the use of, and further develop, a in-situ and near-real time LIBS (Laser Induced Breakdown Spectroscopy) instrument to measure liquids relevant to the Hanford site. Specifically, we investigated the feasibility of using LIBS to measure the cation composition of HLW (High Level Waste) streams. In addition, we focused on enabling continuous interrogation of those liquids using the LIBS system. For the first part, we investigated simple (sodium nitrate) solutions as well as various Hanford tank waste surrogate slurries. We showed that relevant cations (such as sodium, aluminum, magnesium, potassium) can be measured via LIBS under good conditions but the results – both quantitative and qualitative (aka reproducibility) – strongly depend on the kind and concentration of the material used. The dependence of signal intensity on concentration was also confirmed by other measurements conducted (sodium chloride and sodium nitrate in water). For the second part, we investigated the influence of varying depth and concentration of a liquid sample. We observed that for the concentrations and fill depths investigated, the liquid fill depth did not impact the emission intensity of the analyte, while concentration is confirmed to have significant effects. The results of the sample depth experiments influenced testing of waste surrogates. In addition, we focused on further developing and adapting data fitting tools in order to efficiently, accurately, and reproducibly identify the elements and materials present in a LIBS spectrum.

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DOE OSTI2024

Glass Formulation and Testing with RPP-WTP HLW Simulants (Final Report)

This report presents the results from HLW glass formulation work that was conducted at the Vitreous State Laboratory of The Catholic University of America during Part B1 of the RPP-WTP Project (formally known as TWRS-P) in support of the Privatization contractor, BNFL, Inc. Glass formulation development is required to provide data that are essential for design of the vitrification facility in the near term and for safe and reliable operation of that facility in the longer-term. These data also form the bases from which contractually and technically acceptable immobilized high-level waste (IHLW) products will be devised. Glass formulation therefore impacts many aspects of the vitrification facility design, overall process economics, and contractual compliance. As an integral part of the overall system design, development of glass formulations requires input from many sources including tank waste characterization, pretreatment process definition, flow sheet development, and processing schedules. Accordingly, the glass formulation effort must be responsive to new information as the facility design basis evolves.

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DOE OSTI2024

RANGERS: Modeling Report on Integrity and Performance Assessment of Engineered Barrier Systems in a Salt Repository for HLW/SNF

The Engineered Barrier System (EBS) plays an important role in ensuring the long-term safety and containment of high-level waste (HLW) and spent nuclear fuel (SNF) in deep geological repositories in salt formation. As part of a multi-barrier system, the EBS works alongside the natural barrier, which is the salt formation itself and the technical barrier comprising the disposal casks. The primary function of the EBS is to maintain containment during a defined period until the backfill used in the repository made of crushed salt, develops its sealing capacity through compaction. Over the time, the backfill eventually compacts to a state of low porosity and permeability, acting as a long-term seal. However, until this process is complete, the EBS must retain its structural and functional integrity. Regulatory guidelines in Germany currently require the EBS to remain effective for up to next ice age, that is expected in 50,000 years. The significant hydro-geological and topographic changes expected during an ice age could make it impossible to accurately predict the hydro-chemical conditions within the repository system at that time. In response to these challenges, BGE TECHNOLOGY GmbH (BGE TEC) and Sandia National Laboratories (SNL) have jointly developed a comprehensive methodology for the design and safety assessment of engineered barrier systems within the scope of the RANGERS project. This methodology is tailored for repositories in salt formations. The developed methodology provides a structured approach for designing and assessing the performance of the EBS in salt-based repositories. It begins with defining a sealing concept based on the geological characteristics of the selected site and the overall repository design. The entire repository system, comprising the geological site, repository infrastructure, and EBS, is then subjected to a Features, Events, and Processes (FEP) analysis, focusing solely on those FEPs that affect the EBS. The derived FEPs help identify the loads and stresses acting on the EBS, which serve as the foundation for conducting an integrity assessment. This analysis helps predict the EBS’s evolution and performance over the regulatory time frame, feeding into integrated performance assessment simulations.

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DOE OSTI2025

Vitrification Testing of HLW with High Phosphate

Projections of the number of high level waste (HLW) canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) (e.g., [1]) are based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier Department of Energy (DOE) melter designs, incorporates a glass bubbler system. The bubblers create active glass pool mixing and thereby improve heat and mass transfer and glass melting rate. The WTP HLW melters each have a glass surface area of 3.75 m 2 and depth of ~1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by optimization of the feed and glass formulations, increasing the melter operating temperature above 1150⁰C, and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage.

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DOE OSTI2025

L-SCIE and SUPCRTNE model and database development

This progress report (Level 3 Milestone Number M3SF-25LL010301052) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Host Rock Properties & Processes Work Package SF-25LL01030105. The focus of this milestone is to create an initial SUPCRTNE database and extensively test the SUPCRTNE code. We prepared a draft manuscript describing our workflow and approach for developing a next-generation thermodynamic database for the Spent Fuel and High Level Waste Disposition campaign. The goal is to ease any future formal software qualification effort as was done on the Yucca Mountain Project for codes including SUPCRT92 and EQ3/6. We are attempting to extend the SUPCRTNE database to include more species, mainly from the NEA volumes. We can also take advantage of thermodynamic database activities that are ongoing at Thermochimie and as part of the EURADII program. However, our effort is limited by the funds available and the changes to program scope that was initiated in mid-FY25.

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DOE OSTI2025

Engineered Barrier System Research Activities at SNL: FY25 Report

This report describes research and development (R&D) activities conducted at Sandia National Laboratories (SNL) during Fiscal Year 2025 (FY25) specifically related to the Engineered Barrier System (EBS) Properties & Processes Work Package in the Spent Fuel and High-Level Waste Disposition (SFHLWD) Campaign supported by the United States (U.S.) Department of Energy (DOE). SNL activities focus on understanding EBS component evolution and interactions within the EBS, as well as interactions between the host media and the EBS. The team represented in this report consists of individuals from SNL, Argonne National Laboratory (Advanced Photon Source), and Vanderbilt University. The activities represented in this report include development of anion sorbent materials for buffer formulations, properties of pure bentonite and bentonite mixtures, and characterization of cement interfaces relevant to the EBS. The SNL EBS activities also leverage interactions with EBS teams from other national laboratories, and international collaborations to ensure that the DOE program is active and abreast of the latest advances in nuclear waste disposal.

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DOE OSTI2026

FY24 Status of Engineered Barrier Systems Research

This report describes research and development (R&D) activities conducted during Fiscal Year 2024 (FY24) specifically related to the Engineered Barrier System (EBS) R&D Work Package in the Spent Fuel and High-Level Waste Disposition Campaign supported by the United States (U.S.) Department of Energy (DOE). The R&D activities focus on understanding EBS component evolution and interactions within the EBS, as well as interactions between the host media and the EBS. The R&D team represented in this report consists of individuals from Sandia National Laboratories, Lawrence Berkeley National Laboratory (LBNL), Los Alamos National Laboratory (LANL), and Vanderbilt University. EBS R&D work also leverages international collaborations to ensure that the DOE program is active and abreast of the latest advances in nuclear waste disposal.

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DOE OSTI2022

System Analysis Modeling and Intermodal Transportation for Commercial Spent Nuclear Fuel

The United States Department of Energy (DOE) has long term goals to develop solutions for managing the nation’s spent nuclear fuel (SNF) and high-level waste (HLW) inventory. The Integrated Waste Management (IWM) program under the DOE office of Nuclear Energy (DOE-NE) is employing system-level engineering and analysis principles to inform potential future waste management system architectures. Managing the spent nuclear waste requires the use of system-level analysis software that takes various aspects of the fuel cycle into account like waste generation, on-site/centralized storage, transportation infrastructure, and long-term disposal. The Next Generation System Analysis Model (NGSAM) is an agent-based model that was developed to simulate the transportation and storage of SNF and HLW. As an agent-based model, NGSAM has the capability to detail the interaction and movement of individual components and groups, such as rail cars and casks. The SNF inventory from commercial nuclear reactors is currently in temporary storage at multiple locations spread across the United States. Shipping of SNF from these locations relies on one of three transportation modes: rail, heavy-haul truck, or barge. Out of the three modes identified, rail is generally the most preferred due to the size of the canisters and casks the SNF would be shipped in. However, under some scenarios, a direct rail route might not be readily available to a reactor site or improving the rail infrastructure at shutdown sites might be too cost-prohibitive for utilities to opt for a direct rail transfer. Under such scenarios, using a barge or heavy haul truck to de-inventory the site and transfer the SNF to a nearby intermodal transfer site with adequate rail infrastructure where the payload could be transferred to a rail car might prove to be an attractive option. This work initially presents the various intermodal transportation options that could be used to transfer SNF from reactor sites to rail cars. This is followed by exploring the operational steps in each of these modes to move the SNF from a reactor site and transfer it to a rail car. This work also presents the procedure of implementing the intermodal transfer methodology in NGSAM using various Java methods. Finally, the process times for accomplishing each of the individual steps are furnished. The implementation ideology, assumptions, and future steps are presented in this work.

Gadey, Harish Reddy↗