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1,076 records · Page 60
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Assessment of the Corrosion Risks to a Double Shell Tank During the Selective Removal of a Cesium-rich Supernate Layer - 20227

The Direct Feed Low Activity Waste (DF-LAW) activity will provide the initial feed to the Waste Treatment and Immobilization Plant at the Hanford Tank Farm facility. Feed preparation for DF-LAW will be carried out in multiple double shell waste tanks (DST) utilizing the Tank Side Cesium Removal (TSCR) process. Tank AP-106 will be an integral part of the TSCR process and hence key to the DFLAW activity. Prior to being utilized for this purpose, the Cs-137 concentration levels in the AP-106 supernate must be reduced significantly. The AP-106 re-purposing process manages normal waste tank addition procedures such that waste acceptance criteria objectives for Cs-137 removal are achieved. As this process was being developed, several concerns regarding tank integrity during and after repurposing of Tank AP-106, and the associated tanks (i.e., AP-102, AN-101, and AP-108), were addressed. The potential corrosion degradation mechanisms identified based on process knowledge included: general and pitting corrosion, caustic stress corrosion cracking, concentration cell corrosion, and microbiologically induced corrosion (MIC). A systematic approach was undertaken to assess the risk of corrosion during the AP-106 re-purposing process. The assessment considered the initial and present condition of the tank, the service history of the tank, the environmental conditions created in each tank during each stage of the process, process timing, and corrosion testing and modeling that approximated tank conditions. Based on this information, recommendations were provided to minimize the exposure of the tank to corrosive conditions. These recommendations included time limits for each stage of the process, staging of corrosion inhibitor solutions should process delays occur, and waste sampling with potential chemistry adjustments at defined steps in the process. Additionally, recommendations for post-process supernate chemistry adjustments and tank inspections were put forth. The greatest risk for this process occurs during the manipulation of large quantities of Columbia River water (i.e., greater than 450,000 gallons) within AP-106, the associated tanks and the supporting infrastructure (i.e., transfer line piping). This risk becomes more pronounced if there is a process upset and the tank walls remain exposed to a corrosive condition longer than intended. As long as the river water is blended into the supernate in a timely fashion, and the waste is adjusted appropriately with corrosion inhibitors, corrosion may be minimized and the final condition of the supernate in AP-106 and the associated tanks is expected to be in compliance with corrosion chemistry requirements. Contingency plans will be made to minimize the impact of a process upset. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Progress of the enhanced Hanford single shell tank (SST) integrity project - 15497

To improve understanding of the single-shell tanks (SSTs) integrity, Washington River Protection Solutions, LLC (WRPS), the US DOE Hanford Site tank contractor, developed an enhanced Single-Shell Tank Integrity Project (SSTIP) in 2009. An expert panel on SST integrity, consisting of various subject matter experts in industry and academia, was created to provide recommendations supporting the development of the project. This panel developed 33 recommendations in four main areas of interest: structural integrity, liner degradation, leak integrity and prevention, and mitigation of contamination migration. In late 2010, seventeen of these recommendations were used to develop the basis for the M-45-10-1 Change Package for the Hanford Federal Agreement and Compliance Order, which is also known as the Tri-Party Agreement. The change package identified two phases of work for SST integrity. The initial phase was focused on efforts to envelope the integrity of the tanks. The initial phase was divided into two primary areas of investigation: structural integrity and leak integrity. If necessary, based on the outcome from the initial work, second phase would be focused on further definition of the integrity of the tanks and liners. Combined these two phases are designed to support the formal integrity assessment of the Hanford SSTs in 2018 by an Independent Qualified Registered Engineer. As the initial phase of this work completes in early 2015, this paper summarizes the project formation and status the work of the past few years associated the implementation of the Phase one recommendations. Work in the initial phase to further define the DOE's understanding of the structural integrity SSTs involved preparing a modern Analysis of Record (AOR) using finite element analysis. Structural analyses of the SSTs have been conducted since 1957, but these analyses used analog calculations, less rigorous models, or focused on individual structures. As such, an integrated understanding of all of the SSTs has not been developed to modern expectations. In support of this effort, other activities addressed the visual inspection of tank internal conditions and the collection of concrete core samples from the tanks, including a full height tank sidewall core, for analysis of current mechanics properties. The work on the liner leak integrity has examined leaks from 25 tanks with known liner failures. Individual leak assessments were developed for each tank to identify the leak cause and location and estimate historic leak rates. A common cause/failure analysis study was performed to take data from individual tanks to look for trends in the causes of failure. A separate activity is being conducted to examine the propensity for corrosion in select SSTs with aggressive waste layers. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Development of a Conditioned System-Level Groundwater Model to Evaluate Long-Term Groundwater Impacts within a Performance Assessment - 20115

A preliminary performance assessment (PA) of single-shell tank Waste Management Area (WMA) A-AX located at the U.S. Department of Energy (DOE)'s Hanford Site in southeastern Washington is being conducted to satisfy the requirements of DOE Order 435.1 [1] as it relates to closure of the single-shell radioactive waste tanks in the WMA A-AX tank farms. A PA assesses the fate, transport, and impacts of radionuclides within a low-level radioactive waste disposal facility in its assumed closure configuration and the subsequent potential doses to humans over a 1,000-year compliance period and a 10,000-year performance evaluation period. The WMA A-AX preliminary PA evaluation is structured around the complementary use of process-level and system-level models to calculate the facility performance against established DOE Order 435.1 [1] performance objectives. Process-level models are those that represent a detailed phenomenological representation of processes of concern in the PA. Process models typically only represent one or a few of the components of the PA, such as groundwater flow and transport, and must be integrated with other modeling elements to perform PA calculations. System-level models are those that are abstracted from the process models, retaining the essential features of the process model, while allowing integration of all aspects of the PA in a single modeling framework. System-level models are often characterized by coarser numerical discretization, lower dimensionality, or other similar simplifications compared to the process-level model. Traditionally, a three-dimensional (3-D) process-level model is utilized primarily to evaluate the long-term impact on groundwater and the potential doses to individuals who consume contaminated groundwater. System-level models are also utilized to evaluate the groundwater pathway in PAs. These models typically have reduced dimensionality (1-D) and are conditioned utilizing flow fields (Darcy fluxes) and moisture content distributions that are abstracted from the process level models. The abstraction approach assures that the flow field in both models is consistent for a specific set of input parameters for flow, differing only in the discretization and dimensionality of the two models. The preliminary WMA A-AX PA incorporates a detailed representation of the geological system and hydraulic properties within the 3-D model STOMP{sup C} numerical code so that the effects of relevant features and processes on water flow and radionuclide transport in the subsurface can be evaluated. The complementary system-level model is developed utilizing the GoldSim{sup C} code to implement a simplified, 1-D equivalent model to represent the groundwater pathway. Contaminant transport through the vadose zone and unconfined aquifer for Tc-99 and I-129 were evaluated in each of the models. Adjustments to the saturated portion of the GoldSim{sup C} 1-D model were required to mimic the dispersion effect captured with the 3-D STOMP{sup C} model. Once this conditioning was conducted, highly similar results for the transport of Tc-99 and I-129 were achieved at the point of calculation, located 100 m downgradient from the WMA A-AX fenceline. These radionuclides represent elements that are regarded as primary dose drivers in the PA groundwater pathway analysis. The high degree of conformance between the two models suggests that the use of the equivalent 1-D system model is suitable for evaluating the full suite of radionuclides that will be released from the tank sources within WMA A-AX over the 1,000-year compliance period and over the 10,000-year evaluation period. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Developing and Implementing New Waste Chemistry Controls for Hanford's Double-Shell Tanks - 20041

Waste chemistry controls for the Hanford double-shell tanks (DSTs) were established in the 1980's in response to tank failures caused by stress corrosion cracking (SCC) at the Savannah River Site (SRS). The controls were established based on a combination of SRS chemistry limits and corrosion testing results. Conditions in the Hanford DSTs have changed over time, with the temperature in the majority of the DSTs dropping below 50 deg. C. Testing indicated that the original chemistry control specification and SCC testing performed in 2010 did not protect against pitting corrosion mechanisms - what is believed to have caused the failure of tank 241-AY-102. New waste chemistry controls are necessary to protect the DSTs from both SCC and pitting corrosion. Savannah River National Laboratory (SRNL), with input from the Tank Integrity Expert Panel (TIEP) Corrosion Subgroup (CSG), performed a statistically based investigation of nitrate and halide ion induced pitting corrosion. The investigation was intended to develop a comprehensive waste chemistry envelope that minimizes the risk of both SCC and pitting caused by halide and nitrate ions. The waste chemistry envelope needed to be robust enough to address future waste retrieval and process changes which could significantly change the waste composition of the DSTs. The experimental design allowed the statistically significant ions and their effect on pitting corrosion to be determined. SRNL developed the 'pitting factor' which is an empirical relationship between the statistically significant inhibitive species (hydroxide, and nitrite), and pitting inducing species (nitrate, chloride, and fluoride). The coefficients are weighted factors that show the degree of influence the species have on the propensity for pitting. New waste chemistry control limits were recommended by SRNL, and concurred with by the TIEP CSG, that incorporated the pitting factor along with other limits. The controls developed to reduce the risk of pitting corrosion were evaluated to determine if they also protected against SCC. Hundreds of test results were reviewed and compared against the proposed limits. With the exception of one test result, instances where cracking was observed corresponded to chemistry that was outside of the proposed limits - indicating that the chemistry controls developed for pitting also adequately addressed SCC concerns. Existing tank composition estimates were evaluated against the new chemistry controls - three current tank chemistries are suspect, and may not meet the new chemistry control requirements. These included the interstitial liquid (liquid associated with solids) of DSTs 241-AN-102, 241-AN-106, 241-AN-107, and 241-AY-101. Measures are ongoing to evaluate the tanks with the suspect chemistry. These include varying combinations of: performing additional evaluations (history of use, modeling); DST specific laboratory testing; and, pursuing core samples of the DSTs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

AGR-5/6/7 Thermal Model with Non-uniform Gas Gaps

Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tri-structural isotropic fuel particles. In the absence of direct measurement, fuel compact temperatures were calculated using a three-dimensional finite element thermal model, which is subject to simulation uncertainty. The most dominant factor in the uncertainty of calculated fuel temperatures is the gas gap uncertainty due to the nub-to-shell clearance caused by a design error of AGR-5/6/7 capsules. The thermal model was revised to examine the most probable graphite offset position for six different days during the irradiation for Capsules 1 and 2. The analysis varied the offset distance and azimuthal direction at both the top and bottom of the holder. The best-fit offset was estimated based on the minimum root mean square error of the residuals (measured minus calculated) for the operational thermocouples (TCs). From these results, the following conclusions were made: (1) The holder offsets led to slightly lower average temperatures but wider temperature variations (lower minimum and higher peak fuel temperatures) for both Capsule 1 and Capsule 2. (2) During earlier cycles (162A–164B), when numerous TCs were still operational, the best-fit offset distance varied over a specific range for both the top and bottom ([0.002–0.0035 in.] for Capsule 1 and [0.003-0.004 in] for Capsule 2). In contrast, the offset azimuthal direction varied widely, especially for the offset at the bottom of the Capsule 1 holder. This is because holder movement was somewhat constrained at the top of Capsule 1 by the TC leads running through the capsule head and into the holder and by the through tubes in Capsule 2, but the Capsule 1 bottom did not have this type of constraint. (3) During later cycles, when all TCs failed, applying the maximum possible offset of 0.006 in. to the northwest direction for both the top and bottom resulted in a calculated peak fuel temperature of 1557? in Capsule 1 (i.e., a 135? increase from 1422? with zero offset on September 20, 2019 (166A)); the maximum offset of 0.0068 in. to the south for both top and bottom resulted in a calculated peak fuel temperature of 1110°C in Capsule 2 on April 20, 2020 (i.e., a 116? increase from 994? with zero offset (168A)). High peak fuel temperatures in Capsule 1 during Cycle 166A could be the cause of massive particle failure near the end of this cycle. (4) Even though the highest temperature at the tip of Type-N TCs, such as TC-1-7, slightly exceeded 1000?, the temperature along the TC wire reached as high as 1335? assuming an offset of 0.006 in. in the northwest of Capsule 1 holder near the end of Cycle 166A. This temperature significantly exceeds the temperature threshold at which TC degradation is expected to occur, ultimately contributing to considerable particle failures in Capsule 1. For eight Type-N TCs in Capsule 2, the peak TC line temperature was much lower (i.e., 1029°C for TC-2-5), assuming maximum offset of 00068 in. to the south during cycle 168A.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Testing Results from the Low Activity Waste Pretreatment System (LAWPS) Full-Scale Ion Exchange Column Test and Engineering Scale Integrated System Test - 18634

The U.S. Department of Energy (DOE), Office of River Protection's (ORP) primary mission is to retrieve and treat Hanford's tank waste and close the Tank Farms to protect the Columbia River. Mixed radioactive waste is stored in 177 underground tanks at the Hanford Site as reported in DOE/ORP 2003- 02, Environmental Impact Statement for Retrieval, Treatment, and Disposal of Tank Waste and Closure of the Single Shell Tanks at the Hanford Site, Richland WA, Inventory and Source Term Data Package. The 177 underground tanks are estimated to contain about 56 million gallons of waste. The Low Activity Waste Pretreatment System (LAWPS) Project provides for the early production of immobilized low activity waste (ILAW) by feeding LAW directly from Tank Farms to the Hanford Waste Treatment Plant's (WTP's) LAW Facility, bypassing the Pretreatment Facility. Prior to the transfer of feed to the WTP LAW Vitrification Facility, tank supernatant waste will be pretreated in the LAWPS to meet the WTP LAW waste acceptance criteria. The LAWPS will also facilitate the return of secondary liquid wastes from the WTP LAW Vitrification Facility to the Tank Farms. In November of 2015 DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS), awarded a contract to the AECOM team to design, fabricate and operate two test platforms to generate performance data at two different scales and demonstrate technology maturation. The first platform models the ion exchange (IX) columns at full scale, and the second platform models the integrated flowsheet (cross flow filter (CFF) and IX) at 1/9. scale. Overarching technical objectives of performing the full-scale ion exchange test and the engineering-scale integrated test are to: 1. Test interactions between unit process operations; 2. Test removal of non-radioactive cesium from process stream; 3. Test removal of undissolved solids from process stream; 4. Confirm operational flow rate range and stability; 5. Confirm volumetric throughput of waste stream over flow operating range; 6. Confirm range of process temperatures over planned operating range; 7. Establish system hydraulics and process criteria; 8. Coordinate key control and monitoring components; 9. Establish treated product chemistry meets requirements (cesium, solids, and pH); and 10. Test the hazard control strategy for a no flow condition through the IX column. Data collected during testing will be used by WRPS to validate the preliminary design, inform the final design of the permanent plant equipment, and demonstrate that the critical technology elements have advanced to a technology readiness level of 6 (prototypical system validation in a relevant environment). Testing on both platforms started in fall of 2016 generating NQA-1 data. This paper covers test results from the first five phases of testing on the 1/9. scale integrated test platform. A further three phases of testing are currently being executed, these including a 5.6 M sodium simulant containing foulant, a high viscosity simulant and finally a run performed with a lower Cs concentration (all earlier tests were spiked to 4 x Cs to reduce volume and cost). Results from these tests were not available at the time of writing available. Results are also presented from the full scale IX testing. Eight test were performed covering column loading, elution and regeneration; column loading, elution and regeneration with upflow displacement; IX column loading and gas migration; resin removal; solids accumulation; and IX column emergency recirculation and vent. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

AGR-5/6/7 Thermal Model with Non-uniform Gas Gaps

Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tri-structural isotropic fuel particles. In the absence of direct measurement, fuel compact temperatures were calculated using a three-dimensional finite element thermal model, which is subject to simulation uncertainty. The most dominant factor in the uncertainty of calculated fuel temperatures is the gas gap uncertainty due to the nub-to-shell clearance caused by a design error of AGR-5/6/7 capsules. The thermal model was revised to examine the most probable graphite offset position for six different days during the irradiation for Capsules 1 and 2. The analysis varied the offset distance and azimuthal direction at both the top and bottom of the holder. The best-fit offset was estimated based on the minimum root mean square error of the residuals (measured minus calculated) for the operational thermocouples (TCs). From these results, the following conclusions were made: (1) The holder offsets led to slightly lower average temperatures but wider temperature variations (lower minimum and higher peak fuel temperatures) for both Capsule 1 and Capsule 2. (2) During earlier cycles (162A–164B), when numerous TCs were still operational, the best-fit offset distance varied over a specific range for both the top and bottom ([0.002–0.0035 in.] for Capsule 1 and [0.003-0.004 in] for Capsule 2). In contrast, the offset azimuthal direction varied widely, especially for the offset at the bottom of the Capsule 1 holder. This is because holder movement was somewhat constrained at the top of Capsule 1 by the TC leads running through the capsule head and into the holder and by the through tubes in Capsule 2, but the Capsule 1 bottom did not have this type of constraint. (3) During later cycles, when all TCs failed, applying the maximum possible offset of 0.006 in. to the northwest direction for both the top and bottom resulted in a calculated peak fuel temperature of 1557? in Capsule 1 (i.e., a 135? increase from 1422? with zero offset on September 20, 2019 (166A)); the maximum offset of 0.0068 in. to the south for both top and bottom resulted in a calculated peak fuel temperature of 1110°C in Capsule 2 on April 20, 2020 (i.e., a 116? increase from 994? with zero offset (168A)). High peak fuel temperatures in Capsule 1 during Cycle 166A could be the cause of massive particle failure near the end of this cycle. (4) Even though the highest temperature at the tip of Type-N TCs, such as TC-1-7, slightly exceeded 1000?, the temperature along the TC wire reached as high as 1335? assuming an offset of 0.006 in. in the northwest of Capsule 1 holder near the end of Cycle 166A. This temperature significantly exceeds the temperature threshold at which TC degradation is expected to occur, ultimately contributing to considerable particle failures in Capsule 1. For eight Type-N TCs in Capsule 2, the peak TC line temperature was much lower (i.e., 1029°C for TC-2-5), assuming maximum offset of 00068 in. to the south during cycle 168A.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

AGR-5/6/7 Final Release-to-Birth Ratio Data Analysis

AGR-5/6/7 is the last of a series of Advanced Gas Reactor (AGR) experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low-enriched fuel for use in the high-temperature gas cooled reactor (HTGR). AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through the use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO-coated particle fuel. The AGR tests provide valuable irradiation-performance data to support fuel process development, qualify fuel for normal operation and accident conditions, and support development and validation of fuel performance and fission-product (FP) transport models and codes. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. Release-to-birth ratios (R/B) for fission-gas isotopes released from each capsule are calculated from release rates, measured by germanium detectors in the Fission Product Monitoring System (FPMS) installed downstream from each capsule, and birth rates calculated using numerical models of FG generation. The R/Bs are a critical measure of the ability of the fuel kernel, the particle coating layers, and the compact matrix to retain fission-gas atoms, preventing their release into the sweep-gas flow, and the impact of initially defective particles and/or particle-coating failures that occur during irradiation. For fission-gas isotopes, particle failure is defined as failure of all coating layers, allowing gaseous fission atoms to escape from a particle. During the first five cycles (162B ? 165A), R/Bs were stable in the 10-8?10-6 range, and no in-pile particle failures were observed, based on the gross gamma counts. The maximum R/B value of around 2 ? 10-6 for Kr-85m resulted from the presence of as-fabricated exposed kernels (based on the high exposed kernel fraction), the dispersed uranium, and high fuel particle temperatures in Capsule 1. Comparison of capsule-measured R/Bs from these early cycles to predictions using the previously developed AGR R/B model demonstrated FG release from the AGR-5/6/7 TRISO fuel was comparable to that of previous experiments. In addition, the Kr-85m R/B per-exposed-kernel values are comparable to R/B values obtained in AGR-3/4 irradiation experiment and four irradiation experiments performed during 1980s: (1) HRB-17/18, (2) COMEDIE-BD1, (3) HFR B1, and (4) HRB-21. In contrast, all measured R/B values are lower than predictions by the commonly used Richards and German models, which are intentionally conservative. A large number of in-pile particle failures occurred in Capsule 1 by the end of Cycle 166A. During the final four cycles (166A ? 168A), apparent damage to the Capsule 1 gas line appeared to cause FG leakage from that capsule into the other four capsules, resulting in an increase in fission gas (FG) detected in the effluent for all capsules. Isolation of the Capsule 1 gas line during the last three cycles also prevented measurement of its FG release. Thus, R/Bs in all capsules after Cycle 166A are highly uncertain because of undefined amount of leakage from Capsule 1, especially for long-lived isotopes. A few hundred in-pile particle failures were estimated for Capsule 1 before the end of Cycle 166A, but the total number of failures is unknown due to the lack of FG release data in the later cycles. Based primarily on evidence from the gross gamma counts during Cycle 168A, approximately 15 particles failed in Capsule 3 and four particles failed in Capsule 2. In-pile failures in Capsule 3 were anticipated because this capsule was designed to operate beyond the HTGR normal operating temperature range. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on the absence of the typical spikes in gross gamma counts and low failure estimates using the AGR model, developed in INL/EXT-14-32970, for R/B of the short-lived isotopes (Kr-89 and Xe-137) with minimal leakage from Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Chalcogenide Glass Radiation Sensor; Materials Development, Design and Device Testing

For many decades, various radiation detecting material have been extensively researched, to find a better material or mechanism for radiation sensing. Recently, there is a growing need for a smaller and effective material or device that can perform similar functions of bulkier Geiger counters and other measurement options, which fail the requirement for easy, cheap and accurate radiation dose measurement. Here arises the use of thin film chalcogenide glass, which has unique properties of high thermal stability along with high sensitivity towards short wavelength radiation. The unique properties of chalcogenide glasses are attributed to the lone pair p-shell electrons, which provide some distinctive optical properties when compared to crystalline material. These qualities are derived from the energy band diagram and the presence of localized states in the band gap. Chalcogenide glasses have band tail states and localized states, along with the two band states. These extra states are primarily due to the lone pair electrons as well as the amorphous structure of the glasses. The localized states between the conductance band (CB) and valence band (VB) are primarily due to the presence of the lone pair electrons, while the band tail states are attributed to the Van der Waal's forces between layers of atoms. Localized states are trap locations within the band gap where electrons from the valence band can hop into, in their path towards the conduction band. Tail states on the other hand are locations near the band gap edges and are known as Urbach tail states (E u ). These states are occupied with many electrons that can participate in the various transformations due to interaction with photons. According to Y. Utsugi et. al., the electron-phonon interactions are responsible for the generation of the Urbach tails. These states are responsible for setting the absorption edge for these glasses and photons with energy near the band gap affect these states. We have studied the effect of x-rays and γ-rays, on thin film chalcogenide glasses and applied them in conjunction with film incorporating a silver source in a new type of radiation sensor for which we have an US patent application. In this report, we give data about our studies regarding our designed radiation sensor along with the testing and performance at various radiation doses. These studies have been preceded by materials characterization research related to the compositional and structural characteristics of the active materials used in the radiation sensor design. During the work on the project, we collected a large volume of material since every experiment was repeated many times to verify the results. We conducted a comprehensive material research, analysis and discussion with the aim to understand the nature of the occurring effects, design different structures to harness these effects, generated models to aid in the understanding the effects, built different device structures and collected data to quantify device performance. These various aspects of our investigation have been detailed in previous quarterly reports. In this report, we present our main results and emphasize on the results pertaining to the core project goals materials development, sensor design and testing and with an emphasis on classifying the appropriate material and design for the optimal application. The report has three main parts: (i) Presentation of the main data; (ii) Bulleted summary of the most important results; (iii) List of the patent, journal publications, conference proceedings and conferences participation, occurring as a result of working on the project.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Performance Assessment for the Waste Management Area C at the Hanford Site in Southeast Washington - 17515

An initial performance assessment (PA) of Single-Shell Tank (SST) Waste Management Area C (WMA C) located at the U.S. Department of Energy's (DOE) Hanford Site in southeastern Washington has been completed to satisfy the requirements of the Hanford Federal Facility Agreement and Consent Order (HFFACO), as well as other Federal requirements and State-approved closure plans and permits. The WMA C PA assesses the fate, transport, and impacts of radionuclides within residual wastes left in tanks and ancillary equipment and facilities in their assumed closed configuration, and the subsequent risks to humans into the far future. This PA focused on radiological impacts has been developed to meet the requirements for a closure authorization under DOE O 435.1 that includes a waste incidental to reprocessing determination for residual wastes remaining in tanks, ancillary equipment, and facilities. PA results indicate that the performance objectives and measures for the all-pathways dose, the air pathway dose, the radon flux, and groundwater protection are met for both the 1,000-year compliance time period (2020 to 3020) and the post-compliance period (3020 to 12020). The peak dose for the all-pathways analysis in the 1,000-year compliance period is associated with the air pathway, with the peak dose of 4 x 10{sup -5} mSv/yr dominated by tritium resulting from upward gaseous diffusive flux from the residual waste. The peak dose within the sensitivity/uncertainty analysis time period (1,000 to 10,000 years after closure) occurs at about 1,500 years after closure, and results primarily from a peak in {sup 99}Tc groundwater concentration at 100 m down-gradient of the facility. The peak total dose within the sensitivity/uncertainty analysis time period is 0.001 mSv/yr. For all of the sensitivity analyses and uncertainty analyses evaluated, the disposal system also met the performance objectives. In the uncertainty analysis performed with the system level model based on GoldSim{sup C}, the highest calculated groundwater dose in the compliance period was about 7 x 10{sup -4} mSv/yr, and the highest calculated peak dose in the sensitivity/uncertainty analysis period was 0.025 mSv/yr.Doses associated with hypothetical inadvertent human intrusion were calculated for all sources in WMA C and compared to the acute and chronic performance measures in DOE O 435.1. The calculated doses for acute and chronic exposure scenario from potential intrusion into a waste transfer pipeline remain below the DOE O 435.1 performance measure for the time period evaluated beyond 100 years after closure. Acute exposure from drilling into the waste yielded a dose of about 0.36 mSv after the assumed 100-year institutional control period, well below the 5 mSv performance measure for acute exposure. Chronic exposure to waste from drill cuttings in a rural pasture scenario resulted in a dose of about 0.08 mSv/yr after the assumed 100-year institutional control period, well below the 1 mSv/yr performance measure for chronic exposure. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Evidence for triaxial shape coexistence in 74 Ge

The deformation properties of the low-lying states in 74 Ge have been investigated using multistep Coulomb excitation. The measurements were carried out with the advanced 𝛾-ray tracking array, GRETINA, and the CHICO2 particle detector. A comprehensive set of 𝐸⁢2 transition and diagonal matrix elements was deduced following an analysis with the semiclassical coupled-channels code GOSIA. The data were compared with results of calculations carried out within the framework of the generalized triaxial rotor model as well as with the configuration interaction shell model and the symmetric rotor model. Results from calculations with covariant density functional theory were used to construct a five-dimensional collective Hamiltonian for further comparisons with the data. Collectively, the calculations provide an accurate reproduction of the experimental matrix elements and further support an understanding in terms of the coexistence of two axially asymmetric shapes. In conclusion, this leads to an overall interpretation of the underlying structure of 74 Ge requiring triaxiality, as is also the case in the neighboring even-mass Ge isotopes.

59 ≤ A ≤ 89↗
DOE OSTI2021

Evaluation of Electrical Resistivity Tomography to Monitor the Transport of Past Releases Beneath Tank Farms

Underground storage tanks at the Hanford Site, in southeastern Washington State, hold radioactive waste generated from four decades of plutonium production. The 149 single-shell tanks and the 28 double-shell tanks have all exceeded their initial design life of approximately 25 years. At least 67 tanks are assumed to have leaked in the past, resulting in radioactive releases into the vadose zone. Gamma ray logging within dry monitoring wells is currently the primary method for tracking the migration of leaked tank waste through the vadose zone. While this approach provides an accurate assessment of radioactive contamination, that information is only provided near (within ~1m) the borehole, leaving most of the vadose zone unmonitored, particularly the important region directly beneath the tank. This report describes a numerical study that investigates the feasibility and performance of time-lapse 3D electrical resistivity tomography (ERT) for long-term monitoring of a hypothetical tank waste location and migration through the vadose zone. ERT is a method of remotely imaging the bulk electrical conductivity (EC) of the subsurface, which is significantly impacted by the presence of conductive solid and liquid tank waste. The release of liquid tank wastes increases subsurface fluid conductivity and saturation over time, creating a target to use time-lapse ERT for long-term monitoring. Although the presence of metallic infrastructure can cause ERT interference, recent advancements in ERT data processing enable the deleterious effects of buried metallic infrastructure (e.g. pipes, wellbore casings, tanks) to be removed to better determine the liquid tank waste migration over time. Three hypothetical realistic scenarios were simulated in the ERT evaluation. The first two scenarios assume the same leak amount and rate (i.e., between 1/1/1951 and 12/31/1951 at the rate of 347 m 3 per year) but different leaky tanks. Scenario 1 assumes leaks under tank B-102, which is located on the edge of the B-tank farm and surrounded by a few metallic infrastructure including cased pipes/wells/tanks. Scenario 2 assumes leaks under tank B-108, which is located near the center of the B-tank farm and surrounded by larger amount of metallic infrastructure than B-102. Scenario 3 assumes the same metallic infrastructure as B-102, with a more recent contaminant leak that was simulated to have occurred between 1/1/2018 and 12/31/2023 at a rate of 1.89 m 3 per year. The leak time in Scenarios 1 and 2 corresponds to a historical overfill event in 1951 and Scenario 3 corresponds to a recent found tank leak in 2019. In each scenario, a “true” bulk EC model vs. time reflecting contaminant migration was generated. ERT data was simulated from these “true” bulk EC models and a time-lapse ERT inversion produced “imaged” bulk EC vs. time. Three electrode configurations in two, four and eight boreholes surrounding the leak tank were used in the ERT simulations in each scenario. These borehole configurations were considered logistically feasible and cost-effective for monitoring. The hypothetical ERT boreholes are assumed to have non-metallic casing. By comparing the “imaged” bulk EC with the “true” bulk EC, it was demonstrated that the three configurations of wells used (two, four, and eight wells) were able to successfully monitor the migration of tank leaks through the vadose zone, with bulk EC resolution increasing with the number of down borehole ERT arrays for all scenarios. Therefore, the use of eight boreholes to perform ERT monitoring beneath the tanks provided the best spatiotemporal information.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Measurements and models of enhanced recombination following inner-shell vacancies in liquid xenon

Electron-capture decays of 125 Xe and 127 Xe , and double-electron-capture decays of 124 Xe , are backgrounds in searches for weakly interacting massive particles (WIMPs) conducted by dual-phase xenon time projection chambers such as LUX-ZEPLIN (LZ). These decays produce signals with more light and less charge than equivalent-energy 𝛽 decays and correspondingly overlap more with WIMP signals. We measure three electron-capture charge yields in LZ: the 1.1 keV M-shell, 5.2 keV L-shell, and 33.2 keV K-shell at drift fields of 193 and 96.5 V/cm. The LL double-electron-capture decay of 124 Xe exhibits even more pronounced shifts in charge and light. We provide a first model of double-electron-capture charge yields using the link between ionization density and electron-ion recombination, and identify a need for more accurate calculations. Finally, we discuss the implications of the reduced charge yield of these decays and other interactions creating inner-shell vacancies for future dark matter searches.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2026

Operation of a Modular 3D-Pixelated Liquid Argon Time-Projection Chamber in a Neutrino Beam

The 2x2 Demonstrator, a prototype for the Deep Underground Neutrino Experiment (DUNE) liquid argon (LAr) Near Detector, was exposed to the Neutrinos from the Main Injector (NuMI) neutrino beam at Fermi National Accelerator Laboratory (Fermilab). This detector is a prototype of a new modular design for a liquid argon time-projection chamber (LArTPC), comprising a two-by-two array of four modules, each further segmented into two optically isolated LArTPCs. The 2x2 Demonstrator features a number of pioneering technologies, including a low-profile resistive field shell to establish drift fields, native 3D ionization pixelated imaging, and a high-coverage dielectric light readout system. The 2.4-tonne active mass detector is flanked upstream and downstream by supplemental solid-scintillator tracking planes, repurposed from the MINERvA experiment, which track ionizing particles exiting the argon volume. The antineutrino beam data collected by the detector over a 4.5 day period in 2024 include over 30,000 neutrino interactions in the LAr active volume—the first neutrino interactions reported by a DUNE detector prototype. During its physics-quality run, the 2x2 Demonstrator operated at a nominal drift field of 500 V/cm and maintained good LAr purity, with a stable electron lifetime of approximately 1.25 ms. This paper describes the detector and supporting systems, summarizes the installation and commissioning, and presents the initial validation of collected NuMI beam and off-beam self-triggers. In addition, it highlights observed interactions in the detector volume, including candidate muon antineutrino events.

47 OTHER INSTRUMENTATION↗