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At least 883 records · Page 49

Crystallization of a Neptunyl Oxalate Hydrate from Solutions Containing Np V and the Uranyl Peroxide Nanocluster U 60 Ox 30

Uranyl peroxide nanoclusters are an evolving family of materials with potential applications throughout the nuclear fuel cycle. While several studies have investigated their interactions with alkali and alkaline earth metals, no studies have probed their interactions with the actinide elements. This work describes a system containing U 60 Ox 30 , [((UO 2 )(O 2 )) 60 (C 2 O 4 ) 30 ] 60- , and neptunium(V) as a function of neptunium concentration. Ultra-small and small angle X-ray scattering were used to observe these interactions in the aqueous phase, and X-ray diffraction was used to observe solid products. The results show that neptunium induces aggregation of U 60 Ox 30 when the neptunium concentration is ≤ 10 mM, whereas (NpO 2 ) 2 C 2 O 4 ·6H 2 O(cr) and studtite ultimately form at 15–25 mM neptunium. Finally, the latter result suggests that neptunium coordinates with the bridging oxalate ligands in U 60 Ox 30 , leaving metastable uranyl peroxide species in solution, and is an important finding given the potential application of uranyl peroxide nanoclusters in the recycling of used nuclear fuel.

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Modeling gas breakthrough and flow phenomena through engineered barrier systems using a discrete fracture approach

Compacted clays are being considered to build engineered barrier systems (EBS) intended for the safe isolation of high-level nuclear waste (HLW) and spent nuclear fuel (SNF). The corrosion of the metallic canister containing the HLW/SNF will lead to the generation and buildup of the gas pressure in the more internal part of the clay buffer. This phenomenon would eventually trigger the formation and propagation of fractures in the clay barrier, jeopardizing its safety functions. Here, in this work, we propose to use the fragmentation technique (MFT) to model evolving fractures in clays triggered by gas pressurization. The MFT has been successfully used to model the formation of fractures in concrete, drying cracks in soil, hydraulic and thermo-fractures in rocks. In this work, we extend the MFT to deal with multiphase fluid flow in deformable porous media, and we upgraded a fully coupled computer finite element code using the extended technique. The proposed approach is first verified against analytical solutions and is then applied to model gas breakthrough experiments in clays. A very satisfactory performance of the method is observed in all the analyses, showing the potential of the MFT to tackle multiphase flow problems in deformable porous media with evolving discontinuities.

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Is there a Perfect D and D Plan - 20100

Problems with many Deactivation and Demolition (D and D) projects arise not necessarily from the technical difficulty of the actions required in a given facility, but rather from lack of integration of the stove piped planning efforts; engineering, operations, environmental, radiological, human resources, etc. This paper will cover Polestar Technical Service, Inc. (Polestar) systems engineering approach to D and D planning based on the company's experience over the past 27 years. This systematic, four-pronged approach to planning is applicable to any D and D project and addresses several diverse topics in a cohesive and integrated manner. The four areas to be discussed are project scope, requirements, facility end state, work force and base operations. This approach progresses a facility through the typical DOE life-cycle from operations, through deactivation, into surveillance and maintenance (S and M) and to eventual demolition and site remediation. To achieve a final end state the approach brings together the efforts associated with identifying and managing the right set of requirements, detailing the scope of work, establishing the baseline, redefining the work processes, aligning the workforce resources with the work processes, and maintaining the base operations activities as the project progresses through the various life-cycle stages. Getting these building blocks right provides a solid framework for executing any D and D project. (authors)

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Characterization of CST Post-Processing AP-105 Hanford Tank Waste

The primary goal of the Tank Side Cesium Removal (TSCR) system, under development by Washington River Protection Solutions, LLC (WRPS), is to remove entrained solids and 137 Cs from the Hanford tank waste supernate to expedite production of low-activity waste. Ion exchange (IX) testing of 10.9 L of waste from Hanford tank 241-AP-105 (AP-105), performed by Pacific Northwest National Laboratory, used a lead-lag-polish column format, with a bed volume of 10 mL per column, to decontaminate tank waste supernate using crystalline silicotitanate (CST) as the IX media. The AP-105 Cs IX processing test, discussed elsewhere, resulted in a shorter transition zone (i.e., steeper load curve) than those defined by wastes from tanks 241-AP-107 and 241-AW-102.1 The shorter transition zone was indicative of a matrix effect retarding Cs capacity. Therefore, aliquots of spent CST from the lead, lag, and polish columns were subjected to a digestion protocol to quantify analytes retained by the CST and extrapolate the impact on Cs capacity. The spent CST was digested using a combination of 5 M HNO 3 and H 2 O 2 with vigorous heating and stirring. Due to the radiation dose accompanying the 137 Cs on the CST columns, a secondary Cs separation by ammonium molybdophosphate embedded in polyacrylonitrile (AMP-PAN) was performed to separate the 137 Cs from the CST so the samples could be contact-handled for analysis outside of a shielded facility.

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Nevada Higher Education Benefits from US Department of Energy (DOE) Environmental Management (EM) Nevada Program's Transfer of Geologic Samples - 20442

The Nevada National Security Site (NNSS), formerly the Nevada Test Site, was the location of 100 historic atmospheric and 828 historic underground nuclear tests from 1951 to 1992. Related to this historic nuclear testing, geologic and hydrologic studies of the site were conducted utilizing the skills and the expertise of the U.S. Geological Survey (USGS) and national laboratories (Los Alamos National Laboratory, Lawrence Livermore National Laboratory), who were principal in leading the development of weapons and were responsible for specific underground testing programs. The subsurface samples and technical/scientific data associated with NNSS geologic studies (past and present) are preserved and stored at the USGS Mercury Core Library and Data Center located at the NNSS. Currently, the facility stores over 2,000,000 linear feet of cores and cuttings from more than 2,600 drill holes that can be accessed for study. Most of the samples and historic geologic work was focused directly on subsurface geologic settings that relate directly to historic underground nuclear testing. Since 1992, the United States has observed a unilateral moratorium on full-scale nuclear testing, and the U.S. Department of Energy (DOE) Environmental Management (EM) Nevada Program is now responsible for hydrogeologic characterization of the potential impacts to the natural groundwater systems that may have resulted from historic underground nuclear tests. As part of this effort, the EM Nevada Program Underground Test Area (UGTA) Activity has drilled and completed over 58 deep (2,000 - 7,000 ft.) characterization wells, totaling in excess of 170,000 linear feet of cuttings and core samples. Cutting samples were collected as triplicate samples for each respective depth interval, to account for potential later nondestructive/destructive analysis and to preserve samples for regulatory purposes. Recently, it was recognized by the EM Nevada Program that opportunities may exist to reduce the cost and floor space required for the storage of geologic samples at the Mercury Core Library without impacting the integrity and representative nature of the samples necessary for project execution. An initiative was sponsored by EM Nevada Program to evaluate several options: 1) disposal of a portion of cuttings and cores in a land fill setting; 2) reduction in sample volume through skeletonizing core and cuttings from wells to a representative but much smaller number of samples; and 3) solicit potential interest in the academic community where other geoscientists could freely access the samples for studies. The Nevada state university system through the Nevada Bureau of Mines and Geology responded positively to the opportunity to receive these NNSS samples from the EM Nevada Program. In June 2019, approximately 17,000 geologic samples, representing greater than 170,000 linear feet of drilling and weighing over 20,000 pounds, were shipped from the NNSS to the Great Basin Science Sample and Records Library located in Reno, Nevada. The benefits resulting from this transfer were realized by the EM Nevada Program in terms of cost and space savings for geologic sample storage at the NNSS. The transfer did not impact the EM Nevada Program mission as access to representative geologic samples for regulatory and scientific purposes was preserved. Moreover, the Nevada Bureau of Mines and Geology acquired a significant resource of subsurface geologic samples and supporting technical data to support academic and scientific studies in a complex volcanic setting in southwest Nevada. (authors)

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Verification of Special Nuclear Material (SNM) using Active Well Coincidence Counter (AWCC) and a High Purity Germanium (HPGe) Detector

SNM verification can be accomplished via two methods i.e. destructive analysis (DA) or nondestructive analysis (NDA). The DA modes are much more accurate compared to NDA modes; nevertheless, the sample's physical integrity is destroyed in the process. Techniques such as ICP-MS, ID-ICP-MS, TI-MS are few examples of DA, where a bulk material is ground and mixed with a solution (some form of acid) to acquire a homogenous sample. In contrast, NDA modes utilize signals emitted by the sample such as gamma, neutrons, light, or heat to analyze samples under investigation. HPGe detectors are some of the most frequently used gamma detectors to assay SNM. However, in a high activity scenario it provides a large uncertainty and sometimes becomes unfeasible to measure low energy gamma rays. Meanwhile, neutron detectors are not affected by the gamma background, therefore they can be used in addition to the gamma detectors to verify total fissile mass ({sup 235}U + {sup 239}Pu) of SNM. The AWCC is one of the thermal neutron NDA systems developed at the Los Alamos National Laboratory (LANL). It consists of 42 {sup 3}He detectors to measure gross neutrons (singles), coincidence neutrons (doubles), higher order multiplicity counting (triples, quadruples, etc.). It can effectively operate in both active and passive neutron interrogation modes. It can utilize AmLi, AmBe, or {sup 252}Cf active interrogation neutron sources. It is used to verify declared fissile mass ({sup 235}U or {sup 239}Pu) present in SNM. Objectives: Calibrate HPGe and AWCC for lightly irradiated (with enough {sup 137}Cs gamma background) ∼93% enriched HEU fuels; Verify fissile mass ({sup 235}U +{sup 239}Pu) content using AWCC and HPGe; Model and benchmark AWCC in MCNP6. Neutron coincidence and multiplicity counting: Two or more neutrons are coincident if they are detected by the system within the specified gate window; In AWCC: Gate window is 128 μsec; Induced fission in fissile material such as {sup 235}U and {sup 239}Pu - emission of time correlated neutrons - Singles, Doubles, Triples, or Quadruples; No induced fission in non-fissile materials- no correlated neutrons - only singles. SNM can be verified with high accuracy and precision using gamma and neutron NDA instruments such as HPGe and AWCC. Interaction with thermal neutrons. Non-fissile atoms - do not emit correlated neutrons. Fissile atoms - emit correlated neutrons during induced fission events. The time correlated neutrons can be detected by using coincidence or multiplicity counters such as AWCC. Likewise, {sup 235}U produces 186 keV gamma peak that can be measured by a HPGe gamma detector. Counts under 186 keV gamma peak provide information about {sup 235}U content. Combining AWCC and HPGe results can provide better understanding of special nuclear material under investigation.

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Corrosion-Resistant Coatings on Spent Nuclear Fuel Canisters to Mitigate and Repair Potential Stress Corrosion Cracking (FY22 Status)

This report summarizes the activities performed by Sandia National Laboratories in FY22 to identify and test coating materials for the prevention, mitigation, and/or repair of potential chloride-induced stress corrosion cracking in spent nuclear fuel dry storage canisters. This work continues efforts by Sandia National Laboratories that are summarized in previous reports in FY20 and FY21 on the same topic. The previous work detailed the specific coating properties desired for application and implementation to spent nuclear fuel canisters (FY20) and identified several potential coatings for evaluation (FY21). In FY22, Sandia National Laboratories, in collaboration with four industry partners through a Memorandum of Understanding, started evaluating the physical, mechanical, and corrosion-resistance properties of 6 different coating systems (11 total coating variants) to develop a baseline understanding of the viability of each coating type for use to prevent, mitigate, and/or repair potential stress corrosion on cracking on spent nuclear fuel canisters. This collaborative R&D program leverages the analytical and laboratory capabilities at Sandia National Laboratories and the material design and synthesis capabilities of the industry collaborators. The coating systems include organic (polyetherketoneketone, modified polyimide/polyurea, modified phenolic resin), organic/inorganic ceramic hybrids (silane-based polyurethane hybrid and a quasi-ceramic sol-gel polyurethane hybrid), and hybrid systems in conjuncture with a Zn-rich primer. These coatings were applied to stainless steel coupons (the same coupons were supplied to all vendors by SNL for direct comparison) and have undergone several physical, mechanical, and electrochemical tests. The results and implications of these tests are summarized in this report. These analyses will be used to identify the most effective coatings for potential use on spent nuclear fuel dry storage canisters, and also to identify specific needs for further optimization of coating technologies for their application on spent nuclear fuel canisters. In FY22, Sandia National Laboratories performed baseline testing and atmospheric exposure tests of the coating samples supplied by the vendors in accordance with the scope of work defined in the Memorandum of Understanding. In FY23, Sandia National Laboratories will continue evaluating coating performance with a focus on thermal and radiolytic stability.

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A Critical Review of the Circular Economy for Lithium-Ion Batteries and Photovoltaic Modules: Status, Challenges, and Opportunities

To meet net-zero emissions and cost targets for power production, recent analysis indicates that photovoltaic (PV) capacity in the United States could exceed 1 TW by 2050 alongside comparable levels of energy storage capacity, mostly from batteries. For comparison, the total U.S. utility-scale power capacity from all energy sources in 2020 was 1.2 TW (EIA 2022), of which solar satisfied approximately 3% (DOE 2021). With such massive scales of deployment, questions have arisen regarding issues of material supply for manufacturing, end-of-life management of technologies, environmental impacts across the life cycle, and economic costs to both individual consumers and society at large. A set of solutions to address these issues center on the development of a circular economy - shifting from a take-make-waste linear economic model to one that retains the value of materials and products as long as possible, recovering materials at end of life to recirculate back into the economy. With limited global experience, scholars and practitioners have begun to investigate circular economy pathways, focusing on applying novel technologies and analytical methods to fast-growing sectors like renewable energy. This critical review aims to synthesize the growing literature to identify key insights, gaps, and opportunities for research and implementation of a circular economy for two of the leading technologies that enable the transition to a renewable energy economy: solar PV and lithium-ion batteries (LIBs). We apply state-of-the-science systematic literature review procedures to critically analyze over 3,000 publications on the circular economy of solar PV and LIBs, categorizing those that pass a series of objective screens in ways that can illuminate the current state of the art, highlight existing impediments to a circular economy, and recommend future technological and analytical research. We conclude that while neither PV nor LIB industries have reached a circular economy, they are both on a path towards increased circularity. Based on our assessment of the state of current literature and scientific understanding, we recommend research move beyond its prior emphasis on recycling technology development to more comprehensively investigate other CE strategies, more holistically consider economic, environmental and policy aspects of CE strategies, increase leveraging of digital information systems that can support acceleration towards a CE, and to continue to study CE-related aspects of LIB and PV markets.

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Shaker Table Test Plan

Currently, spent nuclear fuel (SNF) is stored in onsite independent spent fuel storage facilities (ISFSIs), which is a dry storage facility, at 55 nuclear power plant sites. The majority of SNF in dry storage is in welded metal canisters (2,917 canisters at the end of 2019). The canisters are loaded for storage in storage overpacks (vertical casks or horizontal storage modules) and placed on outdoor concrete pads. Because the SNF will be stored at ISFSIs for an extended period of time, there is growing concern with regards to the behavior of the SNF within these dry storage systems during earthquakes. To address these concerns, the SFWST program is considering conducting an earthquake shaker table test. The goal of this test is to determine the strains and accelerations on fuel assembly hardware and cladding during earthquakes of different magnitudes to better quantify the potential damage an earthquake could inflict on spent nuclear fuel rods. The seismic integrity of the storage system has been addressed in the past by the US Nuclear Regulatory Commission and is not the focus of this potential test. Instead the DOE would benefit from knowing the condition of the fuel cladding from storage, transportation, to disposal so that it can ascertain repository performance for the fuel and packaging in its final state. A seismic event is part of the possible loading events that the fuel could experience in its lifetime. This report proposes several earthquake shaker table tests with different degrees of complexity. Alternative 1 was defined in the FY20 work scope. Alternatives 2 and 3 were recently developed to take advantage of the NUHOMS 32PTH dry storage canister that may be available in FY21 for this test at a minimum cost to the project. The selection of the alternative(s) will depend on the available budget and the SFWST program priorities for the near future.

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SRPPF Aqueous Recovery System Sorbent Testing

This study details preliminary testing of 7 inorganic sorbents for the solidification of the surrogate SRPPF aqueous recovery system liquid excess material of 4M sodium nitrate. Sorbents were subject to various mixing methods to incorporate sodium nitrate solution including static, rotating, and in situ paddle mixing. Paddle mixing was found to be the most effective mixing technique, while simple static tests were also successful, though only for some of the sorbents. Setting tests were performed to ensure that the sorbent sequestered the liquid without release through a conical paint filter. Recommended set times ranged from near immediate to 7 days. The set mixtures were investigated further using ultra-centrifugal liquid release testing, thermogravimetric analysis, and visual microscopy to understand the mechanism by which the sorbent sequestered the liquid. Most sorbents seem to incorporate liquid by sorbing it onto and into the sorbent particles, while Drierite ® , anhydrous calcium sulfate, created a crystal structure that bound the surrogate liquid. Overall, four sorbents are recommended for continued testing: Aquaset II, Aquaset IIG, Drierite ® , and Spill-X-A.

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Ion-Exchange Modeling of Crystalline Silicotitanate for Cesium Removal - 20283

The Tank Closure Cesium Removal (TCCR) system is a Savannah River Site (SRS) demonstration 'at-tank' process designed to remove {sup 137}Cs from the high-level aqueous tank waste so that the decontaminated solution can be disposed as low-level waste. Cesium is removed by ion exchange (IX) columns using engineered IONSIV{sup TM} R9120-B form of the Crystalline Silicotitanate (CST) media. The TCCR system is deployed at Tanks 10 and Tank 11 in the SRS H Tank Farm. Water is added to the salt-cake in Tank 10 H to dissolve it. The dissolved salt solution waste is pumped out of Tank 10H (feed tank), through filters and IX columns. The decontaminated salt solution is transferred to Tank 11 (receipt tank), and on to Tank 50H for final disposal in the Saltstone Production Facility. The current TCCR can accommodate lead-lag (two-column) or lead-lag-guard (three-column) configurations to optimize media utilization and achieve the target decontamination. To assist the TCCR operations, a parametric study was conducted to evaluate the impact of different parameters (e.g., column configurations (single column, two columns or three columns in series), waste characteristics, operating temperature, process flow rate, CST average particle size) on the IX column performance including CST bed utilization. The initial results indicate that the IX column performance is improved at slower process flow rate, at lower operating temperature, and with smaller CST average particle size. Multi-column configurations are recommended, because the single-column configuration does not utilize CST bed effectively. This paper demonstrates the versatility of the ion exchange modeling to evaluate the effects of CST characteristics and operational parameters on IX column performances. (authors)

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Status of the Down-Blending of Irradiated IGR Graphite Fuel in Kazakhstan

Dry processing remains the primary solution for managing the down-blending of irradiated IGR HEU fuel. Several tasks have been initiated and completed to achieve this, including lab modernization, lab-scale and full-scale testing, designing and fabricating an in-paddle mixing drum, and commissioning both a down-blending system and a crushing and milling system. Additionally, a testing and training center was modernized, the equipment was successfully commissioned, and the IGR HEU fuel blocks were repackaged into daily containers. The training of operators and the construction of a new down-blending facility are tasks that will need to be completed in the near future. The progress in these areas continues to validate that the proposed method for down-blending irradiated HEU graphite fuel, followed by cementation of the down-blended material for permanent disposition, appears to be achievable.

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Tank Side Cesium Removal IXC-150 Cask Detonation Report

A set of experiments were conducted at Los Alamos National Laboratory’s firing point 88 to determine the blast pressure and effects of a stoichiometric gas mixture of Hydrogen (H 2 ) and Nitrous Oxide (N 2 O) on the IXC-150 Storage Pad Vent Stack assemblies. The Lab was asked to design, build, and test the vent stack assemblies using a reaction gas volume of 340 in 3 that mimics the Tank Side Cesium Removal (TSCR) Ion Exchange Column (IXC) assembly. The experimental system with vent stack assembly was designed with input from the Washington River Protection Services (WRPS) to ensure an accurate test setup. The tests were conducted at LANL on the newly built filter/gas assembly at the end of April 2021 timeframe, using WRPS supplied vent stack assemblies and internal HEPA filters. The objective of these tests was to observe that the structural integrity of the vent stack assemblies was maintained and determine the resulting blast overpressures at a distance of 16 and 32 inches from the assembly. All work was performed under the LANL Quality Assurance Program (SD330), using a graded application of ASME NQA-1- 2008/NQA-1a-2009. To ensure that all client quality assurance (QA) expectations were addressed, evaluation of the end data needs was performed and the appropriate controls applied for this work.

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High Solids Performance Testing in a Scaled TSCR System

The Tank Side Cesium Removal (TSCR) project is a technology demonstration that will pretreat Hanford tank waste supernatant in support of the Direct Feed Low-Activity Waste (DFLAW) mission. The TSCR system employs two key separation technologies: dead-end filtration (DEF) and ion exchange (IX) using crystalline silicotitanate (CST) media. DEF will be used to remove undissolved solids from tank waste to protect the functionality of the IX columns and the IX system will remove Cs-137 from tank waste. The separation technologies (DEF and IX) used in TSCR are technically mature and have also been successfully deployed at the Savannah River site in a similar facility known as the Tank Closure Cesium Removal (TCCR) system. While testing with simulants and real waste has been successfully performed under conditions expected during the initial operation of TSCR, test data is absent for assessing off normal high solids loading that may be in the TSCR waste feed. Normal TSCR treatment operations are expected to handle wastes with solids content on the order of 200 ppm, and off normal solids loading could be much larger than the nominal level. The testing program described in this report was conducted to understand the consequence of operating the TSCR system at elevated solids loadings up to the high-solids limit of 15,000 ppm [i.e., 1.5-wt%] identified in the TSCR design basis. Although the system is not required to make throughput above the nominal solids loading, the testing was intended to provide important information related to potential off normal operations. At off normal levels near the high-solids limit, there are potential implications for TSCR performance in the areas of throughput, DEF pressure drop, filter backflush frequency, and IX column pressure drop. In addition, intrusion of solids into the IX column was postulated to impact the Cs-137 loading behavior by promoting channeling or flow maldistribution in the column; since the magnitude of the postulated effect was unknown, assessing it was also of interest. The testing was performed using representative waste simulants and a prototypic, integrated TSCR system designed and assembled specifically to conduct the high solids performance assessment. Overall, the scaled TSCR testing demonstrated that full-scale unit operations can succeed in fulfilling their processing objectives in the presence of solids up to 3,000 ppm, but there are potential performance challenges to filtration operations at solids loadings as low as ~500 ppm. The severity of the challenge is likely to be dependent on the type and size distribution of solids, of which the current testing only examined a single type and size distribution. To provide some flexibility for future full-scale operations, the results of the testing suggest two possible risk reduction strategies that can be implemented without any changes in TSCR design or configuration. One option would be to enact an administrative limit on the solids loading to protect TSCR from feeds that are likely to require a high DEF swap frequency. Another option is to permit operation of the DEFs at differential pressures greater than 2 psid before swapping filters. The selection of a higher differential pressure target is not anticipated to adversely impact DEF backflushing efficacy and would reduce both swap frequency and the amount of waste sent to AP-108.

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Design and Development of a Robotic Exoskeleton Device for use in Glove Boxes - 20374

Glovebox operators in nuclear facilities often operate under radioactive environment and typically perform monotonous, dull and repetitive work. This can result in muscle fatigue leading to increased levels of accidents. Current research work aims in assisting the glovebox operators by designing a novel upper limb exoskeleton device to reduce injuries and increase worker efficiency. Almost none of the commercially available hand and upper body exoskeletons are directly suitable for use in gloveboxes. This explains the novelty of the present work in developing a wearable powered robotic exoskeleton device. The device consists of electronics, motors and spring systems mounted on adjustable plates and is supported by a stationary frame behind the worker. The device is intended to be portable and adjustable according to the operator height. The elbow has a flexion movement assisted with motor control by way of a Bowden cable system. The system is floated on a spring and pulley combination that allows for extra safety and smoothness of operation. The end points are fixed near the pivot points to help minimize unwanted movement/torque. The extension movement of the upper and lower arm are assisted by gravity and a counter spring. The shoulder and upper arm movement are held in a neutral position with a spring system for counter balance allowing small movements within the glovebox. This initial concept considers only the vertical movement of the arm. The abduction and adduction movement will not be assisted or impeded by the over shoulder plate or at the forearm plate allowing for full range of motion. Several glovebox aspects such as the level of task assistance, ergonomic design, minimum hindrance in dexterity/operator movements and tear resistance for hazardous safety reasons will be considered in the design of the exoskeleton. Computer models will be simulated to test the load levels and stress conditions on the exoskeleton to enhance the designs. The exoskeleton focusses on major muscular involvement reduction through weight offset and active assistance with minimal parts and hence the device is derived with several joint links and contains both passive and active joints. The paper details the various stages in the design development and prototype testing of the exoskeleton device. Upon successful prototype device development, human testing will be conducted in a replica of the SRS-235F hot cell facility constructed at FIU-ARC. The final stage of the project will be potential testing/deployment of the exoskeleton device in one of the glovebox facilities in Savannah River Nuclear Waste site. (authors)

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Real Time, In-line Monitoring of Hanford Tank Wastes - Year 1 Report

The team comprised of students, postdocs, early, mid and senior career scientists from Los Alamos National Laboratory, Savanah River National Laboratory, Georgia Tech and Florida International University, with the guidance of H2C, is developing a suite of in-line instruments for the Hanford high level waste (HLW) and low active waste (LAW) processes to provide near-real-time analysis of waste form physical properties and composition. The work builds on results from the recent DOE-ORP, EM Technology Development and other projects that demonstrated promise for the use of real-time in-line monitoring (RTIM) to measure chemical compositions of slurries of up to 20 weight % solids. The goal is for this instrument suite is to substantially reduce the need for sampling for process control. Sample waste, exposure associated with sample analysis, and the demand for an external laboratory facility would be greatly reduced. The throughput of waste treatment systems would be improved by elimination of the downtime caused by waiting for sample results. This translates into reduced process storage as process knowledge will be continuously updated in near-real-time. These breakthrough technologies would significantly reduce the life cycle cost and accelerate the schedule for the Hanford tank waste mission.

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30 cm Drop Modeling

This report documents modeling and analysis of the 30 cm cask drop scenario for spent nuclear fuel (SNF) casks. The 30 cm drop is an important hypothetical dynamic loading scenario for SNF because 10 CFR 71.71 defines testing requirements for normal conditions of transport (NCT) of SNF packages, and free drop (of 30 cm for the heaviest packages) tests are required. Other NCT conditions specified in 10 CFR 71.71 include vibration normally incident to transport, and DOE’s multimodal transportation test (MMTT) and related analysis concluded that mechanical shock and vibration incident to transportation is negligible for SNF structural integrity because the loads are so small. The SNF cask 30 cm drop has a lot more potential to cause damage to SNF because the kinetic energy and momentum of the system is stopped abruptly during the cask impact event. Test data and model predictions both agree that the SNF assemblies are subjected to significantly higher loads in a cask drop event than normal transportation shock and vibration, but currently there is no reason to expect gross structural damage of the fuel assembly or widespread SNF cladding failure as a result of a 30 cm cask drop. The ultimate goal of this work is to close the knowledge gap of stress profiles on SNF rods and fuel assembly components during NCT, and this report describes the progress toward reaching that goal as well as a plan for finishing this work over the next year or two. Key remaining tasks include refining the fuel assembly model, calculating a selection of specific fuel assembly cases to populate the damage model, and finally running the damage model to predict the full general response range of 30 cm drop loads on SNF.

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Results of a Soil Flushing Treatability Test for Hanford Site Hexavalent Chromium Removal - 20446

U.S. Department of Energy contractor CH2M Hill Plateau Remediation Company has made significant progress remediating hexavalent chromium in groundwater near the Columbia River through the operation of pump and treat facilities and source removal activities. To mitigate risks associated with hexavalent chromium contamination, which was identified in a CERCLA interim action record of decision as the primary groundwater contaminant of concern for the 100-KR-4 Operable Unit from discharging to the Columbia River, three ion-exchange pump and treat systems were installed between 1997 and 2009. Between 2010 and 2016, operation of the K West pump and treat system, located near the former 105K West plutonium production reactor, reduced concentrations below the groundwater remediation target for hexavalent chromium. In 2016, a rebound study was performed to evaluate the effectiveness of the pump and treat remedy and assess the potential for any residual hexavalent chromium sources continuing to contribute to groundwater contamination. As a result of the rebound study, a source of continued groundwater contamination was identified. To address this condition, a soil flushing treatability test was written and approved by the U.S. Department of Energy and the U.S. Environmental Protection Agency. The test included saturating about a 1-acre area of soil with treated water with the goal of forcing the residual hexavalent chromium to groundwater where existing extraction wells could remove it for treatment, accelerating the cleanup process. In spring and summer 2019, CH2M Hill Plateau Remediation Company installed and began operating a shallow underground leach system to support the treatability test. Preliminary sampling data show the test was effective in removing the residual contamination. Following further evaluation and analysis from the initial test, the process will be considered for implementation elsewhere on the Hanford Site, particularly at locations where there is a known or suspected continuing contamination source in the soil column. The paper will cover the goals of the study, as well as more details about the following: - Overview of pump and treat exit strategy (using U.S. Environmental Protection Agency guidance); - The design and installation of the soil flushing technology in the 100-K Area; - The testing process and results of sampling data; - Post-study evaluation and potential for implementation in other locations at the Hanford Site. Innovative solutions for cleanup, such as the soil flushing technology, have resulted in substantial progress in the removal of groundwater contaminants at the Hanford Site, potentially shortening the time frame required to meet cleanup goals and resulting in significant cost savings. (authors)

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