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At least 163 records · Page 9

Packaging Capacity Calculation: Pu Oxide Packaging Options with the 2 Quart SAVY Container

This calculation supports programmatic efforts to Dilute and Dispose Pu. Metal is to be oxidized at LANL, then transported to SRS for dilution at SRS. Ultimately, the diluted oxide will be disposed of at the Waste Isolation Pilot Plant (WIPP). Currently 3013 containers are used to package the oxide, which are placed in to 9975 Type B shipping containers. The Oxide Packaging Technology Maturation Plan determined that a transition to the used of 2-qt SAVY containers placed in 9977 would increase efficiency in shipping and handling and increase programmatic flexibility. The 2-quart SAVY container meets DOE M 441.1 and TA-55 Documented Safety Analysis requirements for handling and storage of Pu oxide at LANL.

36 MATERIALS SCIENCE↗

Cosmogenic 137 Xe Analysis of Alternatives for nEXO

Neutrinoless double-beta decay is a hypothetical nuclear decay in which two matter particles are produced without corresponding antimatter particles. Observation of neutrinoless double-beta decay would shed light on a potential explanation of matter-antimatter observed in the universe and also indicate the Majorana nature of neutrinos, meaning that the neutrino can act as its own antiparticle. The nEXO experiment is a proposed 5-ton liquid xenon time projection chamber detector with a projected 1.35 × 10 28 yr half-life sensitivity to neutrinoless double-beta decay in 136 Xe [J. Phys. G 49, 015104 (2022)]. While nEXO has a baseline location of SNOLAB, there are several other underground facilities in the world capable of hosting the experiment. The primary change at each site will be cosmogenic 137 Xe backgrounds, caused by varied cosmogenic muon activity at each site. This study finds that while there is a negligible increase in nEXO’s half-life sensitivity at 90% C.L. from improving the cosmogenic 137 Xe veto rejection procedure for SNOLAB, much higher daily muon rates at three other sites will have an adversely negative effect on nEXO’s sensitivity and would likely require further modifications to nEXO’s baseline design to reach nEXO’s scientific goals. Monte Carlo simulations of muons inducing 137 Xe backgrounds were conducted at four underground sites, SNOLAB, SURF, LNGS, and WIPP, each with roughly a factor of 10 times higher daily muon activity than the previous site. There were two primary inputs to these simulations. The first was sample data of cosmogenic muon flight through nEXO’s outer detector, a 12.8m tall and 12.3m wide cylindrical water tank lined with photomultiplier tubes to detect Cerenkov radiation, and all of the secondary interactions produced in the detector. The second input was a parametrization of 137 Xe* production tagging when 137 Xe de-excites and release prompt gammas, as well as accidental tagging caused by neutron capture on 63 Cu, 65 Cu, 19 F, and 1 H throughout the detector. Together, the number of backgrounds relative to the previous 2021 Sensitivity Paper [J. Phys. G 49, 015104 (2022)] was found at each of these four sites across three energy windows, from 400, 1400, and 2600 keV all up to 4600 keV, just above the end point of the prompt gamma spectrum for 137 Xe*. Using level curves for the sensitivity as 137 Xe backgrounds scale relative to baseline calculations, the half-life sensitivity at 90% C.L. for each of trial was found. Even with reduced 137 Xe backgrounds, the results for SNOLAB yielded neglible improvements in nEXO’s predicted sensitivity. On the contrary, the increased backgrounds for the three other considered sites yielded sensitivity estimates below nEXO’s previous projections. This study demonstrates that the shielding from cosmogenic muons at SNOLAB makes it highly preferred even amongst other world leading underground laboratories.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Brine Availability Test in Salt (BATS) FY23 Update

This report summarizes the fiscal year 2023 (FY23) status of the second phase of a series of borehole heater tests in salt at the Waste Isolation Pilot Plant (WIPP) funded by the Disposal Research and Development (R&D) program of the Spent Fuel & Waste Science and Technology (SFWST) office at the US Department of Energy’s Office of Nuclear Energy’s (DOE-NE) Office in the Spent Fuel and Waste Disposition (SFWD) program.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of Radioactive 65 Zn Contamination using 3D Position Sensitive CdZnTe Detectors

SRNL NMG Goal: In collaboration with ORNL, the aim is to identify and explore innovative technology opportunities that accelerate disposition, reduce life cycle costs, minimize worker radiation exposure, and improve worker safety and minimize program risks. • This process will include 3 glovebox systems that require personnel to dilute and package the plutonium for safe storage at WIPP. • For holdup monitoring and verification, gamma ray spectroscopy and imaging are the methods employed to achieve real time analysis of the plutonium waste. 2

Nwadeyi, Valerie E.↗

Improved Compaction Experiments and Modeling of Waste Isolation Pilot Plant Standard, Non-degraded, Waste Containers

A credible simulation of disposal room porosity at the Waste Isolation Pilot Plant (WIPP) requires a tenable compaction model for the 55-gallon waste containers within the room. A review of the legacy waste material model, however, revealed several out-of-date and untested assumptions that could affect the model’s compaction behavior. For example, the legacy model predicted non-physical tensile out-of-plane stresses under plane strain compression. (Plane strain compression is similar to waste compaction in the middle of a long drift.) Consequently, a suite of new compaction experiments were performed on containers filled with surrogate, non-degraded, waste. The new experiments involved uniaxial, triaxial, and hydrostatic compaction tests on quarter-scale and full-scale containers. Special effort was made to measure the volume strain during uniaxial and triaxial tests, so that the lateral strain could be inferred from the axial and volume strain. These experimental measurements were then used to calibrate a pressure dependent, viscoplastic, constitutive model for the homogenized compaction behavior of the waste containers. This new waste material model’s predictions agreed far better with the experimental measurements than the legacy model’s predictions, especially under triaxial and hydrostatic conditions. Under plane strain compression, the new model predicted reasonable compressive out-of-plane stresses, instead of tensile stresses. Moreover, the new model’s plane strain behavior was substantially weaker for the same strain, yet substantially stronger for the same porosity, than the legacy model’s behavior. Although room for improvement exists, the new model appears ready for prudent engineering use.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Brine Availability Test in Salt (BATS) FY24 Update

This report summarizes fiscal year 2024 (FY24) activities centered around a series of field tests in bedded salt at the Waste Isolation Pilot Plant (WIPP) funded by the Office of Spent Fuel and Waste Science and Technology in the Spent Fuel and Waste Disposition (SFWD) program of the US Department of Energy’s Office of Nuclear Energy (DOE-NE). High-level Purpose of Experiments: The Brine Availability Test in Salt (BATS) field tests are revealing both brine occurrence (i.e., where, and how much) and brine migration (i.e., how easily it moves) in the excavation damaged zone (EDZ). This understanding is foundational to develop a safety case for a future heat-generating waste repository in salt, and to starting up a generic repository program in salt to buy down risk. BATS seeks to predict how much brine can flow into both ambient and heated excavations (e.g., boreholes or rooms) in salt. This work is educating and empowering new repository scientists on two fronts: “design and execution of field tests” and “prediction and modeling of coupled processes.” DOE-NE capabilities in salt have grown and been tested through international modeling and benchmarking exercises (e.g., DECOVALEX, RANGERS, KOMPASS, and MEASURES; see Mills et al., 2024). The hands-on expertise we are building is a necessary step towards large-scale disposal demonstrations and eventual implementation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

LANL ACRSP Parameter Recommendations for CRA-2026 Performance Assessment

This report provides recommendations made by the Los Alamos National Laboratory—Carlsbad Operations (LANL-CO) Actinide Chemistry & Repository Science Program (ACRSP) for the following parameters used in the 2026 Compliance Recertification Application (CRA) Performance Assessment (PA) calculations: OXCUTOFF, CONCMIN, CONCINT, CAPMIC, PROPMIC, and SOLMOD (SOLCOH and SOLSOH). A summary of the values recommended for the WIPP parameters are included in the Tables below. The shaded cells indicate a change from the CRA-2019 recommended values. A short summary of the basis and justification for each parameter and recommendation is given in this document.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Custom Equipment Development for Processing of Surplus Plutonium

Facilities and Material • Location: Savannah River Site K-Area, South Carolina • Plutonium designated as surplus is blended with adulterant material, re-packaged and shipped to WIPP in New Mexico • Currently one operational glovebox for downblending, facility modifications and construction of three additional glovebox lines in progress • Current operations are highly manual requiring full PPE and reader/worker steps

Krementz, Daniel [Savannah River National Laborato↗

Foam and Unreacted Material Reduction in Magnesium Oxysulfate for SRPPF Waste Solidification

A magnesium oxysulfate (MOS) grout formulation has been proposed to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). This formulation uses a combination of light-burnt MgO, anhydrous MgSO 4 , and dead-burnt MgO resulting in a grout with good mixability and acceptable density. The setting time for the grout is within the operational limits of the SRPPF facility and the leachate pH is ~9.4, which is within the assumed pH range of the brine from the Waste Isolation Pilot Plant (WIPP). When first tested at the 1-gallon small-scale and 55-gallon full-scale, there was a significant foam layer that raised concerns of a nonhomogeneous final form. A nonhomogeneous mixture could have variable density throughout the waste form and therefore allow for unequal shielding. To reduce the potential for unequal shielding in the final form, a series of foam reduction tests were carried out and a final formulation of 28 wt% light burnt MgO, 10 wt% MgSO 4 , and 62 wt% dead burnt MgO was used to remove the foam in under two hours while still maintaining a mix that has an acceptable density, mix time, peak temperature, and set time.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Effects of SRPPF ARS Effluent Variations on Magnesium Oxysulfate Solidification

A magnesium oxysulfate (MOS) grout formulation has been identified to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). The formulation uses reactive, light burnt MgO (28 wt%), anhydrous MgSO 4 (10 wt%), and dead burnt MgO (62 wt%), resulting in a grout with good mixability, acceptable density, acceptable setting times, and a leachate pH around 9.4, within the assumed Waste Isolation Pilot Plant (WIPP) brine pH range. A series of tests were then designed to observe how this formula reacted to variations in the liquid effluent. Liquid effluent temperature, the caustic pH level, concentration of NaNO 3 , and the presence of neutralization products, trace metals, and/or sodium sulfate (all potential residuals found in the effluent from upstream processing), were evaluated to see how each impacts the grout mixing time, peak temperature, setting time, presence of bleed water, leachate pH, and density. The standard mix has an average mixing time of 18 minutes, peak temperature of 100.5 °C, and set time of 60 minutes. Increasing the temperature of the liquid effluent increased the reaction rate of the mix and reduced mixing and setting times. Between 30 and 40°C the change was relatively small, raising no more than 6 °C compared to the standard mix maximum temperature. Increasing the pH of the liquid effluent had no significant impact on the mix. The addition of neutralization products and trace metals had an overall impact of decreasing the reactivity of the mix. The addition of more ions in the liquid effluent, such as an increase in the concentration of NaNO 3 and the addition of Na 2 SO 4 , generally decreased the reactivity of the mix. None of the variations to the liquid effluent hindered solidification as indicated by the lack of bleed water found on all the samples. The leachate pH values for all mixes tested did not significantly vary from the expected pH of 9.4 and none of the density values fell below 1.8 g/cm 3 . Overall, changes to the liquid effluent were found to have a minimal impact on the formulation suggesting the grout's ability to properly form despite increased temperatures and the presence of residuals typically found in the liquid effluent from the ARS.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Microbial Influence on Radionuclide Transport through the Culebra

The potential release of radionuclides into the Culebra dolomite formation is modeled extensively in the Waste Isolation Pilot Plant’s Performance Assessment (WIPP PA) calculations. Uranium is of chief concern in these modeled scenarios, as it shows the greatest release. Because of this, a series of experiments was undertaken to determine if microbial activity in the Culebra formation could transform uranium, either directly or indirectly, to a state that could be immobilized.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Sulfate Conversion of Reillex HPQ Anion Exchange Resin for Disposal (Interim Report)

This report describes preliminary data to validate the Savannah River Plutonium Processing Facility’s (SRPPF) flowsheet for conversion of used Reillex HPQ anion exchange resin from the nitrate form to the sulfate form. The nitrate form is an oxidizer and therefore does not meet acceptance criteria for disposal at the Waste Isolation Pilot Plant (WIPP). The purpose of this study is to develop data to support acceptance for this disposition pathway. Due to the challenges characterizing the nitrate concentration on solid resin, the data developed to date are based upon indirect analysis of the ion exchange column effluent by ion chromatography. These challenges are discussed and two methods for quantification of nitrate directly on the resin are recommended for further development: TGA-MS and permanganate digestion followed by IC. The resin used for this work was provided in the chloride form; this is the form in which resin is supplied by the manufacturer. However, it had to be converted to the nitrate form, which is the form that will be used in SRPPF’s ion exchange process, prior to use in the sulfate conversion experiments. The chloride-form resin was characterized. A lab-scale procedure for the conversion of Reillex HPQ resin from the chloride to nitrate form was validated. The nitrate-form resin was assessed for particle size and chloride concentration to ensure it met SRPPF’s facility specifications. The baseline sulfate conversion flowsheet was tested. However, nitrate was still detectable in the effluent after approximately 10 bed volumes of 1 M sodium sulfate had been passed through the resin bed. Additional experiments were performed to assess the effect of increasing the feed volume, reducing the flowrate, the use of 2 M sulfuric acid instead of sodium sulfate, and the use of irradiated resin. The sulfuric acid test was the only one which provided a nondetectable nitrate concentration (<0.002 M) in the effluent. Detectable nitrate in the column effluent suggests that nitrate is still present on the resin itself.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Computer vision-based rock bolt detection in orthomosaic imagery obtained in the Waste Isolation Pilot Plant underground facility

Assessing structural integrity of large underground tunnel facilities is often a time consuming and human-labor intensive task. Thus, research using various modes of sensing and automated detection of key structural components in mines is posed to aid in safety assessments and establishing overall structural health. We propose an approach utilizing off-the-shelf camera and lidar technology fixed to a custom sensing platform, image stitching techniques, fine-tuned object detection models, and specialized model-inference methods to automatically detect, count, and map roof bolts for assessment of structural safety in man-made underground tunnels. Results show a novel workflow for effective object counting in orthomosaic tunnel ceiling images generated from collections in GPS-denied mining environments. Additionally, we demonstrate effective fine-tuning of EfficientDet object detectors utilizing state-of-the-art image augmentation techniques known as the mosaic and mixup transformations. Our work is demonstrated on sensed data and imagery collected from the Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) where miles of tunnel ceiling must be assessed for structural integrity.

42 ENGINEERING↗

Deinococcus radiodurans UWO298 Dependence on Background Radiation for Optimal Growth

Ionizing radiation is a major environmental variable for cells on Earth, and so organisms have adapted to either prevent or to repair damages caused by it, primarily from the appearance and accumulation of reactive oxygen species (ROS). In this study, we measured the differential gene expression in Deinococcus radiodurans UWO298 cultures deprived of background ionizing radiation (IR) while growing 605 m underground at the Waste Isolation Pilot Plant (WIPP), reducing the dose rate from 72.1 to 0.9 nGy h –1 from control to treatment, respectively. This reduction in IR dose rate delayed the entry into the exponential phase of the IR-shielded cultures, resulting in a lower biomass accumulation for the duration of the experiment. The RNASeq-based transcriptome analysis showed the differential expression of 0.2 and 2.7% of the D. radiodurans genome after 24 and 34 h of growth in liquid culture, respectively. Gene expression regulation after 34 h was characterized by the downregulation of genes involved in folding newly synthesized and denatured/misfolded proteins, in the assimilation of nitrogen for amino acid synthesis and in the control of copper transport and homeostasis to prevent oxidative stress. We also observed the upregulation of genes coding for proteins with transport and cell wall assembly roles. These results show that D. radiodurans is sensitive to the absence of background levels of ionizing radiation and suggest that its transcriptional response is insufficient to maintain optimal growth.

59 BASIC BIOLOGICAL SCIENCES↗

Influence of ethylenediaminetetraacetic acid on the long-term oxidation state distribution of plutonium

Here, spectrophotometry was used to study the effect of EDTA on plutonium oxidation state distribution as a function of time, pH, and ligand-to-metal ratio (L/M) under anoxic conditions. Novel Pu(V)-EDTA absorption bands were identified at 571, 993, 1105, and 1150 nm with molar absorption coefficients of 15 ± 1, 6 ± 1, 10 ± 1, and 10 ± 1 cm –1 M –1 , respectively. Pu(V)-EDTA spectral changes occurred at L/M < 1, indicating only Pu V O 2 (EDTA) 3- formed with logK = 3.6 ± 0.3. Time-resolved experiments showed EDTA drastically increased the Pu(V/VI) reduction rate, which we propose is driven by amine lone-pair electron donation and the oxidative decarboxylation of EDTA. Oxidation of Pu(III)-EDTA to Pu(IV)-EDTA occurred on a slower time scale (110–237 days) than previously reported (<15 min) and is hypothesized to be radiolysis driven. Pu(V/VI)-EDTA and Pu(III)-EDTA both approached Pu(IV)-EDTA stabilization over time, yet Pu(V/VI)-EDTA solubility data was ≥ 1.0 log 10 units higher than predicted by Pu(IV)-EDTA solubility models, indicating that current thermodynamic models are incomplete. Ultimately, the data show EDTA preferentially stabilizes Pu(IV) over time regardless of initial oxidation state, but Pu(V)-EDTA can persist under environmentally-relevant conditions, emphasizing the need to continue investigating redox reactions, speciation, and behavior of these complexes to support the transuranic waste disposal and surface remediation/containment efforts.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Plutonium Speciation and Oxidation State Distributions in the Presence of Citrate

We explored the speciation and kinetics of the Pu(VI)-citrate and Pu(III)-citrate systems (pH m = 2.5–11.0, I = 0.1 M NaCl, T = 23 °C, O 2 (g) < 2 ppm) using ultraviolet–visible–near-infrared (UV–vis–NIR) spectrophotometry, solvent extraction, and PHREEQC modeling. Formation constants were determined for PuO 2 (HcitH)(aq) (log K° 1,1 = 1.09 ± 0.05) and PuO 2 (HcitH)(citH) 3– (log K° 1,2 = -0.20 ± 0.07), and evidence for (PuO 2 ) m (citH –k ) n (OH) x 2m(3+k)n–x was identified under alkaline conditions. Pu(VI) species were found to be less stable in the presence of citrate than in the absence of citrate (t ≤ 168 days); the rate of reduction increased with increasing pH. Further, the direct reduction of Pu(VI) to Pu(IV) was required to fit experimental data in the presence of citrate but did not improve the fit for Pu in the absence of citrate. We also observed increased Pu(III) stability in the presence of citrate (t ≤ 293 days), with higher concentrations of Pu(III) favored at lower pH. Finally, we provide evidence of a radiolysis-driven mechanism for the citrate-mediated reduction of plutonium that involves electron transfer from the oxidative breakdown of citrate. Our work highlights the need to investigate the redox effect of organic ligands on plutonium oxidation states under repository-relevant conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Architecture Design for Remote Operation of Microreactors: Poster

The nuclear sector is pursuing a number of advanced reactor concepts, one of which is the microreactor, an advanced reactor characterized by a power output of less than 20 MWth. These reactors are intended for use in applications where traditional power solutions are economically or logistically impractical. One key feature required for the successful deployment of microreactors is a remote operation capability, which can dramatically cut staffing expenses by eliminating the need for licensed operators to be present at the site of each reactor and instead concentrate in a centrally located operations center. In moving to a remote operations framework for nuclear reactors, the number of potential attack surfaces for a cyber adversary looking to cause harm or disruption increases. Therefore, a robust cybersecurity architecture is required to mitigate these potential vulnerabilities. This paper explores the functional infrastructure, security, and communication requirements to adapt remote operations to nuclear applications. It then presents a reference architecture for remote operations of microreactors that applies best practices in cybersecurity and remote communications in the context of a digital twin remote operation system.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗