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DOE OSTI2019

Mercury Speciation and Quantification of Hanford 241-AP-107 Tank Waste Feed and Treated Samples

This report describes the second phase of a multi year project to adapt and develop an analytical technique for the low-level determination of total mercury (HgT), elemental mercury (Hg0), and monomethyl mercury (MeHg, hereafter referred to as methyl mercury) in Hanford tank waste. The focus of Phase II was to quantify the mercury (Hg) species total, elemental, and methyl mercury from a raw sample and treated (i.e., filtered and cesium-decontaminated) waste samples originating from Hanford waste tank 241-AP 107. The focus of Phase I work was to determine if mercury species can be preconcentrated from aqueous Hanford tank waste solution onto a solid adsorbent while simultaneously eliminating or very significantly reducing the parent radionuclide activity. This was demonstrated by developing protocols for the preparation, separation, and collection of total mercury and methyl mercury onto solid adsorbents, then measuring the adsorbents for radiological content at Pacific Northwest National Laboratory’s (PNNL’s) Radiochemical Processing Laboratory (RPL) on the Hanford Site. During Phase II, the Hanford tank waste material was preserved, separated (through distillation or digestion), then collected as either solid adsorbents or into liquid distillate form at the RPL and transferred to the PNNL Marine Sciences Laboratory (MSL) in Sequim, Washington, for analysis and quantification. Quantification of total, elemental, and methyl mercury was conducted under radiological control at the MSL using the ultra trace analytical capabilities available there.

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DOE OSTI2025

FY25 Organic Evaporation Testing in Support of Hanford Sample-and-Send

The Hanford Waste Treatment and Immobilization Plant (WTP) is being built to treat and immobilize the approximately 56 million gallons of radioactive mixed waste stored in 156 underground storage tanks. The baseline method for immobilization of Low-Activity Waste (LAW) is vitrification, but additional immobilization capacity is needed to supplement the initial LAW melters. Additionally, efforts are being undertaken to accelerate the disposition of Pre-Treated Waste (PTW) in Hanford’s West Area. An alternative cementitious waste form is being investigated as an alternative immobilization method. However, one impediment to a cementitious waste form is the presence of Land Disposal Restricted (LDR) organic chemicals in tank waste, which are regulated on a concentration-based standard in the final waste form. Hence, if the quantity of organics in PTW and LAW is high enough, treatment may be needed to destroy or remove said organics to make a waste form compatible with disposal in a mixed low level waste landfill. This work evaluates evaporation as a potential avenue for treatment of LDR organics to eliminate the impediment and permit possible use of a cementitious waste form.

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DOE OSTI2021

CACTI ARM Aerial Facility Measurements of Ice Nucleating Particles

The dataset comprises measures, using Colorado State University's Ice Spectrometer (IS, an immersion freezing device with a range from 0°C down to -26 to -29°C) of atmospheric ice nucleating particle (INP) concentrations taken on the Atmospheric Radiation Measurement (ARM) program Aerial Facilty (AAF) G-1 aircraft. INP measurements on the G-1 were collected from varied altitudes on different flights over the region of the Sierras de Córdoba mountain range of north-central Argentina, centred over ARM's Mobile Facility (AMF-1) near Villa Yacanto, where ground-based INP measures were being taken concurrently. Both studies took place as part of the Cloud, Aerosol, and Complex Terrain Interactions (CACTI) Experiment. A filter sampling system was deployed on the G-1 to collect aerosol particles for post-processing of INPs collected on filters (measuring their immersion freezing ability) once returned to Colorado State University (CSU). Filter holders used were pre-cleaned aluminum in-line units loaded with pre-cleaned and pre-sterilized 47 mm dia. Nuclepore polycarbonate filters (0.2 mm pore size). Filters were drawn for varied times, resulting in varied volumes collected (51 to 1667 SL). Mass flow rate was recorded in real-time so that total sampled volume (at standard temperature and pressure) could be determined. A total of 34 sample filters were collected over the IOP, including 5 blanks. Filters were stored at -20°C freezer prior to frozen return to Colorado State University (CSU). . Processing to obtain spectra of INP number concentration active via the immersion freezing mechanism versus temperature was conducted using CSU's IS instrument (McCluskey et al., 2018). For measurment of INPs, collected aerosol particles were re-suspended in 7 mL of 0.02 µm-filtered deionized water. Aliquots of each suspension, and serial dilutions, were dispensed into trays which were fit into aluminum blocks in the IS. Samples are cooled at 0.33°C min-1 and the freezing temperatures of wells recorded automatically. Cumulative INP concentrations were determined by first calculating the INPs per mL of suspension based on Vali (1971) and then converting to concentration per standard liter of air using the proportion of the total liquid sample dispensed and the air sample volumes. Aliquots of suspensions from selected samples were also heat treated (95°C for 20 min) to denature and deactivate biological INPs, and digested in 10% H2O2 at 95°C under UV-B to remove all organic carbon INPs. McCluskey, C. S., J. Ovadnevaite, M. Rinaldi, J. Atkinson, F. Belosi, D. Ceburnis, S. Marullo, T. C. J. Hill, U. Lohmann, Z. A. Kanji, C. O’Dowd, S. M. Kreidenweis, P. J. DeMott, 2018: Marine and Terrestrial Organic Ice Nucleating Particles in Pristine Marine to Continentally-Influenced Northeast Atlantic Air Masses, Journal of Geophysical Research: Atmospheres, 123, 6196–6212, https://doi.org/10.1029/2017JD028033. Vali, G., 1971: Quantitative evaluation of experimental results on the heterogeneous freezing nucleation of supercooled liquids. J. Atmos. Sci., 28, 402–409.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2021

Hanford Supplemental Low Activity Waste Simulant Evaporation Testing for Removal of Organics

The Hanford site has approximately 56 million gallons of radioactive waste stored in 177 underground storage tanks. The Hanford Waste Treatment and Immobilization Plant (WTP) is being built to treat and immobilize the tank waste. The baseline method for immobilization of Low Activity Waste (LAW) is vitrification, but additional immobilization capacity is likely needed to supplement the initial melters. An alternative cementitious waste form is being investigated for that future supplemental immobilization method. However, one impediment to a cementitious waste form is the presence of Land Disposal Restricted (LDR) organic chemicals in tank waste. Developing a method to remove the organics would eliminate that impediment to permit possible use of a cementitious waste form. Savannah River National Laboratory (SRNL) performed testing to examine evaporation as a method to remove some prevalent organics from the Supplemental LAW (SLAW) stream. Samples of product streams from the evaporation were analyzed to determine partitioning of the organics. Modeling was also performed to determine if the experimental and modeling results matched. A description of the experimental details, equipment, and results of that testing are included in this report.The work is intended to inform future SLAW flowsheet development activities and gather useful data about the partitioning of constituents through a possible SLAW feed evaporator.

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DOE OSTI2022

FY2022 Status Update: A Probabilistic Model for Stress Corrosion Cracking of SNF Dry Storage Canisters

Understanding the potential risk of stress corrosion cracking of spent nuclear fuel dry storage canisters has been identified as a knowledge gap for determining the safety of long-term interim storage of spent nuclear fuel. To address this, the DOE is funding a multi-lab DOE effort to understand the timing, occurrence, and consequences of potential canister SCC. Sandia National Laboratories has developed a probabilistic model for canister penetration by SCC. This model has been continuously updated at SNL since 2014. Model uncertainties are treated using a nested loop structure, where the outer loop accounts for uncertainties due to lack of data and the inner aleatoric loop accounts for uncertainties due to variation in nature. By separating uncertainties into these categories, it is possible to focus future work on reducing the most influential epistemic uncertainties. Several experimental studies have already been performed to improve the modeling approach through expanded process understanding and improved model parameterization. The resulting code is physics-based and intended to inform future work by identifying (1) important modeling assumptions, (2) experimental data needs, and (3) necessary model developments. In this document, several of the sub-models in the probabilistic SCC model have been exercised, and the intermediate results, as the model progresses from one sub-model to the next, are presented. Evaluating the sub-models in this manner provides a better understanding of sub-model outputs and has identified several unintended consequences of model assumptions or parameterizations, requiring updates to the modeling approach. The following updates have been made, and future updates have been identified.

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DOE OSTI1989

Nuclear Safety [Vol. 30, No. 3, July-September 1989]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 325 Safety of Framatome Advanced Nuclear Steam Supply Systems Designs by J. A. Charles and D. Lange, 333 Book Review of Nuclear Accidents: Intervention Levels for the Protection of the Public by H. B. Piper; ACCIDENT ANALYSIS: 335 Living PRA Computer Systems by S. C. Dinsmore and H.-P. Balfanz, 343 Summary of ICAP Assessments of RELAP5/MOD2 by W. E. Driskell and R. G. Hanson; CONTROL AND INSTRUMENTATION: 352 Thermal Performance Monitoring System at Maanshan Nuclear Power Plant by H.-J. Chao, Y.-P. Lin, G.-H. Jou, L.-Y. Liao, and Y.-B. Chen; DESIGN FEATURES: 358 Warning Systems for Nuclear Power Plant Emergencies by J. H. Sorensen and D. S. Mileti; WASTE AND SPENT FUEL MANAGEMENT: 371 Activities Related to Waste Management Compiled by E G. Silver; OPERATING EXPERIENCES: 382 Steam Generator Tube Performance: Experience with Water-Cooled Nuclear Power Reactors During 1985 by O. S. Tatone and R. L. Tapping, 400 Systems Interaction Analyses: Concepts and Techniques (Part II) by M. D. Muhlheim and G. A. Murphy, 413 Reactor Shutdown Experience Compiled by J. W. Cletcher, 416 Operating U.S. Power Reactors Compiled by E G. Silver; RECENT DEVELOPMENTS: 440 General Administrative Activities Compiled by E G. Silver, 460 Reports, Standards, and Safety Guides by D. S. Queener, 466 Status of Power-Reactor Licensing Activities Compiled by E G. Silver, 470 Proposed Rule Changes as of Mar. 31, 1989; ANNOUNCEMENTS: 334 Proceedings Published, 351 CEC Seminar on Methods and Codes for Assessing the Off-Site Consequences of Nuclear Accidents, 357 Short Course on Multiphase Flow and Heat Transfer: Bases and Applications in A: The Nuclear Power Industry B: The Process Industries, 357 International Conference on Probabilistic Safety Assessment and Management, 478 International Topical Meeting on the Safety, Status, and Future of Non-Commercial Reactors and Irradiation Facilities, 475 The Authors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI1989

Nuclear Safety [Vol. 30, No. 3, July-September 1989]

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Emerging Contaminants: Identification, Life Cycle, and Key Challenges - 20415

The term 'emerging contaminants' and its multiple variants has come to refer to previously overlooked compounds discovered in the environment which represent a potential threat to human and ecological receptors. These compounds may be discovered in the environment as new analytical techniques are developed and/or as new monitoring programs for drinking water and groundwater are implemented, while the threat to receptors may be realized through advances in understanding of toxicology and pathways to receptors. Emerging contaminants present unique and considerable challenges as demands to address them can outpace the understanding of their toxicity, their need for regulation, their occurrence, and techniques for treating the environmental media they affect. With these challenges in mind, this paper serves as a primer for understanding emerging contaminants. The paper will lay a foundation for understanding emerging contaminants through three key topics: 1. Who identifies emerging contaminants? An overview of the various stakeholder groups around the world working on identifying potential emerging contaminants and developing systems for screening potential candidates, prioritizing, and ultimately generating regulations will be provided. 2. How Chemicals Rise into Awareness as Emerging Contaminants. The process by which contaminants emerge through a cycle of awareness, characterization, management and regulation will be discussed. Current emerging contaminants will be discussed in the context of the life cycle: emerging contaminants where impacts to drinking water supplies are driving public awareness (e.g. per and polyfuoroalkyl substances (PFASs) and 1,4-dioxane), emerging contaminants impacted by changing regulatory standards or development and debate on toxicity (e.g., hexavalent chromium, 1,2,3-trichloropropane, trichloroethylene in vapor intrusion), and those with evolving understanding of impacts and usage (e.g., microplastics). 3. Key Challenges. Unique challenges associated with emerging contaminants, including visibility and public sensitivity, uncertainty and vulnerability, will be discussed. (authors)

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DOE OSTI2025

Direct Extraction of Uranium-Lanthanide Oxides in Tributyl Phosphate

Abstract Direct extraction of used nuclear fuel (UNF) in an organic solution could be more efficient than the previous practice with aqueous solutions. However, the UNF would need to be treated via voloxidiation before being processed using solvent extraction. The voloxidation process can form oxide and/or nitrate compounds. This work investigated the dissolution of uranium/lanthanide oxides in 30 vol % TBP diluted in dodecane (pre-equilibrated with 4 M nitric acid) in a glass reactor with air sparging to ascertain the uranium/lanthanide oxide dissolution behavior prior to scaling the process. The uranium/lanthanide oxides were prepared by co-precipitating uranium/lanthanide nitrates with hydroxide and then calcined to form a mixed oxide. While the relative concentrations of the lanthanides are not representative of used nuclear fuel, the neodymium and erbium allowed ease of tracking dissolution with visible spectroscopy. Cerium was used as a surrogate for plutonium. The dissolution rate of the oxides was similar but incomplete. A miniscule amount of cerium, as cerium oxide, took several months to slowly dissolve; however, when co-precipitated with uranium and other lanthanides, a significant amount of cerium dissolved readily.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Information on Technical Impracticability for Remediation of Iodine-129 Contamination

Groundwater in the 200-UP-1 operable unit (OU) is contaminated with carbon tetrachloride, uranium, nitrate, chromium (total and hexavalent), iodine-129, technetium-99, and tritium, associated with past nuclear weapons production activities at Hanford. The preferred alternative described in the 2012 Record of Decision (ROD) for the 200-UP-1 OU Interim Remedial Action includes 35 years of active remediation using a combination of groundwater pump-and-treat and monitored natural attenuation (MNA) for portions of the contaminated groundwater, followed by institutional controls until cleanup levels are met for unrestricted use. As noted in the 200-UP-1 OU interim ROD, no treatment technology for iodine-129 had been found that could achieve the drinking water standard (DWS) of 1 pCi/L for the iodine-129 concentrations present in the 200-UP-1 OU groundwater. Therefore, the 200-UP-1 OU interim ROD specified hydraulic containment of the iodine-129 plume, update of the conceptual model for iodine-129, and further evaluation of potentially applicable iodine-129 treatment technologies. The 200-UP-1 OU interim ROD further stated that in the event a viable treatment technology is not available, the use of a technical impracticability (TI) waiver may need to be considered as part of the final remedy and that this information be gathered concurrently with the technology evaluation. Hence, this report provides information relevant to a TI waiver consideration on the current federal 1 pCi/L DWS for iodine-129 in the 200-UP-1 groundwater OU of the Hanford Site.

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DOE OSTI2023

Estimate of Gamma Dose Rates from Arrays of Fermi-1 Blanket Elements During the MEDE Process

The Enrico Fermi Atomic Power Plant Unit 1 (Fermi-1) was a sodium-cooled fast breeder reactor located in Monroe County, Michigan. The reactor was powered by a core of enriched uranium metal alloy driver fuel, which was enveloped by an axial and radial blanket material consisting of depleted uranium metal alloyed with 3 wt.% molybdenum. There are 406 axial and 559 radial irradiated sodium-bonded Fermi-1 blanket assemblies in storage at INL, totaling 34 metric tons of heavy metal. Disposal of the Fermi-1 blanket material directly into a geological repository is prohibited due to the reactive characteristic of its bond sodium. A melt drain evaporate (MEDE) process can effectively remove bond sodium from Fermi-1 blanket material. Consequently, planning is underway to apply a MEDE process to treat the 34 metric tons of heavy metal of irradiated Fermi-1 blanket material. Given the irradiated Fermi-1 blanket material’s relatively low power history and 50+ years of decay, it can be treated in a shielded glovebox. To assess the requisite shielding, the dose rates of the Fermi-1 blanket elements in various process configurations are needed. The method to perform the dose rate calculations is to first generate an average source term and associated photon source spectra for the Fermi-1 blanket material in SCALE and then use the associated photon spectra to calculate dose rates using MCNP6.2 in seven representative geometries. Dose rates in rem/h were calculated on contact (1 cm from outer geometry surface), 30 cm away from outer geometry surface, and 1 m from outer geometry surface at axial heights spanning the length of the blanket material within the geometry. The maximum average dose rate for the single Fermi-1 radial blanket element is ~0.60 rem/h, for the Fermi-1 radial blanket assembly ~1.10 rem/h, for the single Fermi-1 axial blanket element ~0.5 rem/h, for the Fermi-1 axial blanket assembly ~1.01 rem/h, for the MEDE can ~ 0.98 rem/h, for the Fermi Storage Canister ~ 0.83 rem/h, and for the MEDE cans in the DOE Standard Canister ~0.51 rem/h. Based on these maximum average values, the bounding dose rates are assumed to be 3× the average dose rates calculated for the axial blanket material and 7.6× the average dose rates calculated for the radial blanket material, based on the distribution of Fermi-1 blanket material assembly burnup.

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DOE OSTI2025

Ascorbic Acid Stability and pH Testing to Support RWM-018 Pilot Scale Wellhead Treatment System Feasibility Determination

A multitude of groundwater remediation techniques for chlorinated volatile organic compounds (cVOCs) have been conducted at the A/M-Areas of the Savannah River Site (SRS). Historical in-situ chemical oxidation (ISCO) injections of potassium permanganate and sodium persulfate conducted in 2018 and 2020, have left residual oxidant concentrations in the A/M-Area groundwater system. ISCO events targeted dense non-aqueous phase liquids (DNAPLs) through the injection of strong oxidants within plume zones upgradient of current groundwater recovery wells RWM-008 and RWM-018. These residual oxidants could significantly impede the operation of the mercury removal system needed to meet National Pollutant Discharge Elimination System (NPDES) permit requirements for the ongoing pump-and-treat system that hydraulically controls potions of the A/M-Area plume. Effluent treated groundwater from this system is discharged to surface water at the receiving outfall of the M-1 Air Stripper. To prevent interference of the M-1 Air Stripper system by residual oxidants, groundwater recovery has been ceased at well RWM-018. Modeling of oxidant transport in A/M-Area from ISCO estimated that the potential oxidant load contributed by RWM-018 could be as high as 120 mg/L. It is also possible that oxidants may be contributed via pumping at well RWM-008 in the near future as residual oxidants within the aquifer migrate toward its zone of influence. Savannah River National Laboratory (SRNL) has conducted research to investigate potential pre-treatment chemicals to neutralize these residual oxidants, before reaching the M-1 Air Stripper, to allow for resumed groundwater recovery at RWM-018 and continued pumping at RWM-008. Deployment of ascorbic acid for neutralization of residual oxidants at RWM-018 (and possibly RWM-008) is recommended based on results from previous reductant testing (SRNL, 2021). For a maximum estimated oxidant load at RWM-0018, a well-head treatment system would need to dose about 400 gallons of ascorbic acid each month to ensure complete neutralization. A second system with additional reagent should also be considered for deployment at RWM-008.

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DOE OSTI2025

Microscopic analysis of irradiated UN–Mo–W for space nuclear propulsion

Nuclear propulsion is an effective method of providing thrust in deep space, allowing for shorter mission duration for long-term surface missions to Mars. The fuel for nuclear propulsion systems must exhibit superior mechanical and thermal properties, to decrease the possibility of fuel failure and to increase reactor efficiency. Here, the objective of the Sirius project is to fabricate and irradiate cermet fuels at the Idaho National Laboratory Transient Reactor Facility to determine the properties of the fuel after irradiation at prototypical nuclear propulsion start-up conditions. The Sirius project featured 3 test capsules, Sirius-1 featured uranium nitride fuel dispersed in a matrix of tungsten and rhenium, while Sirius-2A and -2B featured uranium nitride-molybdenum-tungsten fuel (UN-Mo-W). The fuel samples from the Sirius-2A and -2B capsules were examined after irradiation using optical microscopy, scanning electron microscopy, and X-ray diffraction analyses. The mechanical properties of the fuel from Sirius-2A and -2B are discussed in a previous work. In this study, the results of the optical microscopy, scanning electron microscopy, and X-ray diffraction analyses are discussed.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

AOI2 - Initial Engineering, Testing, and Design of a Commercial-Scale Postcombustion CO 2 Capture System on an Existing Coal-Fired Generating Unit (Project Carbon)

The continued long-term use of North Dakota lignite is likely dependent on creating a business case for carbon capture, utilization, and storage (CCUS) that also addresses society’s desire to reduce carbon emissions. CCUS coupled with enhanced oil recovery appears to be the most feasible option for utilities to sustain and grow the lignite industry. Establishing a market where lignite-powered utilities provide carbon dioxide (CO 2 ) to oil producers depends upon having a cost-effective method to capture CO 2 . The application of CO 2 capture technology to lignite-fired flue gas has challenges in addition to the cost of capture, including the effects of the buildup of heat-stable salts, aerosol formation, degradation of solvents or solid sorbents, or subsequent solvent loss. Lignite, especially North Dakota lignite, presents its own set of challenges that include the combination of high sodium, sulfur, and ash contents; NO x reduction; and footprint limitations. Project Carbon was designed to address these technical barriers and provide pre-FEED (preliminary front-end engineering and design) information necessary for additional commercial demonstrations to move forward. A long-term pilot-scale evaluation of Mitsubishi Heavy Industries’ (MHI’s) KS-1™ solvent at Minnkota Power Cooperative’s lignite-fired Milton R. Young Station Unit 2 (MRY2) was a major objective of the project. The purpose of the test was to demonstrate the technology’s compatibility with MRY2’s flue gas over an extended period of time. During the long-term test, aerosol sampling and measurement were performed, and a pilot-scale baghouse was tested to determine the effect of bag material and operation on aerosol concentrations in the flue gas. Project engineering and design work were performed to support the other major project objective: a pre-FEED study of CO 2 capture at MRY2. Finally, a techno-economic assessment (TEA) of the application of CO 2 capture to MRY2 was performed. This project was performed over a period of 28 months, with a total budget of $12,700,000. The U.S. Department of Energy’s (DOE’s) National Energy Technology Laboratory, through the Energy & Environmental Research Center’s (EERC’s) Cooperative Agreement, the North Dakota Industrial Commission, and industry partners ALLETE and Minnkota Power Cooperative, funded the project. The pilot-scale test system at MRY2 performed very well for the continuous test period of 83 days. The CO 2 capture plant achieved stable operation while utilizing MHI’s KS-1™ solvent and the MHI amine emission reduction unit. No major concerns were encountered when capturing CO 2 from lignite-fired flue gas. Challenges were encountered because the cooling water temperature was warmer than needed for the capture system to operate efficiently. This affected the ability of the solvent to capture CO 2 and promoted solvent loss through solvent entrainment caused by high CO 2 gas temperature within a cyclone separator. The KS-1™ solvent and the lignite-derived flue gas were analyzed during the long-term test. Analysis of the solvent indicates that, although some metals and heat-stable salts increased over the course of the long-term test, they were at levels that can be treated with filtration and reclaiming and do not pose a concern for effective solvent operation. Fourier transform infrared (FTIR) spectrometry was successfully used to determine the composition of the lignite-derived flue gas as well as concentrations of select components of the flue gas at various locations within the EERC’s CO 2 capture system. The formation of aerosols during the capture process is of interest, as are approaches for their reduction, especially in a baghouse. Testing of aerosol removal was performed using the EERC’s portable baghouse. It was found that both Ryton and polytetrafluoroethylene (PTFE) membrane bags were effective at reducing the level of aerosols during steady-state operation. During backpulse events, however, the amount of aerosols passing through the bags was an order of magnitude lower for the Ryton bags than the PTFE membrane bags. The testing showed that proper design of the baghouse air-to-cloth ratio as well as minimization of backpulsing can dramatically reduce the amount of aerosols that pass through the bags and remain in the flue gas stream exiting a baghouse system. A technology maturation plan (TMP) was performed to enable successful development of the Kansai Mitsubishi Carbon Dioxide Recovery® (KM CDR®) Process for application to a combustion flue gas for which the process has not yet been demonstrated. In the TMP, new challenges to the MHI CO 2 capture technology during its application to MRY2 were identified. These challenges included capture from lignite-fired (as opposed to subbituminous coal-fired) flue gas, a larger scope than previous applications of the technology, and implementation in a colder climate that could affect process and equipment operability. Because MRY2 is located where stack icing in the winter could be problematic, a rime icing study was performed. Rime icing modeling indicated that, under both 100% and 50% operating loads, there is a potential for many icing events during any given part of the year in which below-freezing weather conditions exist. The models showed that a critical amount of rime ice could be formed at the stack top during two or more events per year. CO 2 capture plant stack height and location will need to be determined in order to minimize the formation of rime ice. It was found that the integration of the MHI CO 2 capture process into the normal operation of the MRY2 steam turbine is a viable alternative to a cogeneration facility to provide steam and power for the carbon capture facility. The research performed under this project showed that CO 2 capture at a lignite-fired power plant can successfully be performed. No known fatal flaws that would prevent the continued development of the MRY2 CO 2 capture facility were identified during the pre-FEED study. A TEA performed following DOE methodology shows that applying capture technology to MRY2 may not only meet but exceed the DOE-recognized state-of-the-art baseline.

20 FOSSIL-FUELED POWER PLANTS↗

DOE OSTI2020

A White Paper: Disposition Options for a High-Temperature Gas-Cooled Reactor

The high-temperature gas-cooled reactor (HTGR) is a uranium-fueled, graphite-moderated, gas-cooled nuclear reactor design concept capable of producing very high core outlet temperatures. Both types of HTGR have the tristructural isotropic (TRISO) fuel kernel at the heart of the fuel design. For the prismatic block-type HTGR, the TRISO particles are overcoated with a resinated graphitic matrix and pressed into fuel compacts, which are then heat treated and placed in the fuel channels of the prismatic-block-shaped fuel assemblies. For the pebble-bed-type HTGR, the TRISO particles are dispersed in a graphitic-matrix sphere, which is the basic unit for the reactor core. Despite having very different fuel designs, both types of HTGR are graphite-moderated, gas-cooled, thermal reactors using many of the same materials. As a result, both prismatic-block-type and pebble-bed-type HTGRs have similar radioactive waste streams, all of which require safe and secure storage and eventual disposition. Modern HTGR designs are based on a long and rich operating history of several different graphite-moderated, gas-cooled, thermal reactors. Several of these reactors have been shut down, the fuel has been placed in safe storage, and they have undergone some degree of decommissioning. As such, there is significant experience in the management of the spent nuclear fuel (SNF) and radioactive wastes associated with operating these reactors. This white paper will, (1) identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste; and (2) identify the key radioactive waste streams from HTGRs and their characteristics. Idaho National Laboratory (INL) has significant experience in the management of SNF from HTGR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new HTGR-type small modular reactor, or microreactor intended for deployment at the INL site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Benchmarking of Different Inverse Point Kinetics Implementations for an Autocorrected Reactimeter Algorithm

In November 2017, the Transient Reactor Test Facility returned to operation. Since that time, many transient test series have been completed; each has provided valuable data for materials performance, reactor safety that can be applied in future designs. During each experimental series, detector count rates provided important information on the core behavior during transients. However, a limitation of these data is that variations in the neutron distribution during experiments cause errors when attempting to infer reactivity evolution from detector signals. Neutron physics codes can be used to compute the flux shape variations. However, this is a poor solution when the experimental data is used to do verification, validation and uncertainty quantification (VVUQ) on codes. Indeed, if the output of the code is used both as a reference and to correct what the reference is compared to, the circular dependency limits the quality of the VVUQ approach. To overcome this problem, an Autocorrected Reactimeter Algorithm (ACRA) has been developed. This approach infers time-dependent reactivity evolution by testing different spatial corrections and selecting the one that minimizes reactivity variations when the core is in a frozen configuration (i.e. when there is no variation in parameters affecting reactivity). However, the scope of this method was limited to transients where there were negligible thermal feedback. Indeed, the core is never in a frozen configuration when the fuel temperature varies during the whole transient. This is our motivation for the development of an improved version of the ACRA which does not require frozen configurations

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

Reduced-Order CFD Modeling to Support Waste Loading Optimization in Hanford WTP Vitrification

The U.S. DOE Hanford Site stores over 56 million gallons of radioactive liquid tank waste that must be treated and immobilized for long-term disposal The Waste Treatment and Immobilization Plant (WTP) will vitrify this waste by feeding it into Joule-heated melters, where it is incorporated into a stable borosilicate glass Computational fluid dynamics (CFD) simulations of glass melters can provide insight into the maximum achievable waste loading under varying melter operating conditions Fully resolved VOF multiphase simulations were used as the reference model to capture bubble-driven convection in the melter, including bubble formation, rise behavior, and induced glass melt circulation Effective bubble column diameter and rise velocity were extracted from the resolved simulations, compared with empirical correlations, and refit across relevant viscosity and gas flow rate conditions Explicit gas–liquid interface tracking was replaced with a single-phase momentum source term model, enabling faster steady-state CFD simulations while preserving the dominant hydrodynamic effects of bubbling New empirical correlations were developed for effective bubble column diameter and bubble rise velocity by fitting resolved simulation data across expected melter viscosity and gas flow rate ranges, providing improved inputs for the momentum source term model compared with existing literature correlations The momentum source term model reduced fluid-domain mesh size by 89% and achieved an 8.4× computational speedup relative to resolved bubbling simulations The validated momentum source term approach enables prediction of process-relevant heat transfer behavior in the integrated melter model, including heat transfer from the molten glass to the cold cap, plenum, refractory walls, and surrounding structural regions under varying melter operating conditions

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

DOE OSTI2019

Kinetics of Co-Mingled 99 Tc and Cr Removal during Mineral Transformation of Ferrous Hydroxide

In this work, four simulated waste streams relevant to the vitrification of Hanford nuclear waste were studied to evaluate the removal kinetics of technetium-99 (Tc) and co-mingled Cr(VI) during treatment with solid ferrous hydroxide (Fe(OH) 2 (s)). Simulants treated with Fe(OH) 2 (s) were reacted for 24 h and sub-sampled periodically to monitor Tc and Cr removal. Solution sample analysis during the reaction was coupled with solid phase characterization, for example, X-ray absorption spectroscopy (XAS) and X-ray diffraction (XRD), to establish influence of the solid phase product on Tc and Cr removal rates. Based on these results, the majority of Tc and Cr removal occurs within the first 5 min of simulant contact with Fe(OH) 2 (s). However, the order in which Tc and Cr are completely removed from each simulant depends on the simulant chemistry, the preferred reduction pathway, and the solid phase product, for example, magnetite (Fe 3 O 4 ) versus goethite (α-FeOOH). Low pH and low Cr concentrations favor rapid Tc removal, with XRD and XAS confirming that Fe 3 O 4 readily incorporates reduced Tc(IV) into its structure. High pH, high Cr concentrations, and the presence of other co-mingled constituents favor heterogeneous removal of Tc early in the reaction (<1 h) with removal rates often faster than those determined for Cr. At reaction times of >1 h, Tc removal slows as homogeneous removal of Cr begins to dominate, concurrent with an increase in the formation of FeOOH as the solid phase reaction product. These results suggest that for complex, high pH waste streams, Tc removal within the first hour after Fe(OH) 2 (s) treatment is necessary for complete Tc reduction and improved mineral immobilization.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗