Method for removing oxygen impurities from cesium Patent
Heated tungsten filter for removing oxygen impurities from cesium
Anderson, J. R.↗
SEARCH · Engineering Papers
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Extending Magnetic Core Shell Nanoparticle Extraction Technology to Cesium and Antimony Removal from Geothermal Brines in New Zealand - CRADA 440
Geo40 Limited has developed and deployed a process to remove silica from geothermal fluids and produce a high-margin specialty colloidal silica product comparable to market leaders. Geo40 now wishes to explore opportunities to extend their mineral extraction operations to other elements that are present in these brines. Geo40 has identified Cs present in Ohaaki brines at ppm level amounts that could be sold to known customers if it could be produced at an attractive price. With support from DOE’s Geothermal Technologies Office, a simple and highly cost-effective magnetic nanofluid method for extraction of rare earth elements (REEs) from geothermal brine solutions has been developed and demonstrated at the laboratory bench scale at Pacific Northwest National Laboratory (PNNL). Moselle Technologies has acquired a license to the background IP associated with this technology and wishes to foster commercial deployment by supporting applications of the technology beyond REEs, including Cs.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Density of Next-Generation Caustic-Side Solvent Extraction Solvent
This report recommends the appropriate modifier concentration to employ in the Next Generation Solvent (NGS) to be used in the Salt Waste Processing Facility (SWPF) at the Savannah River Site for removal of cesium from legacy tank waste. The Next Generation Caustic-Side Solvent Extraction (NG-CSSX) process employing the NGS is a marked improvement over the original Caustic-Side Solvent Extraction (CSSX) process, as it provides more effective removal of cesium from highly alkaline radioactive wastes and concentrates this removed cesium into an aqueous solution that is easy to vitrify for final disposal. Considerable effort has been devoted to optimizing the chemistry and process conditions of both systems, and their effectiveness has been demonstrated at the pilot scale at the Modular CSSX Unit (MCU). Due to the success of the demonstration and pilot-scale tests, the Salt Waste Processing Facility (SWPF) has been built to process larger volumes of waste with the CSSX process. To accelerate the waste-processing rate of the SWPF beyond its design throughput, plans call for the replacement of the CSSX process by the NG-CSSX process. To implement the NG-CSSX process in equipment optimized for CSSX, adjusting the density of the NGS to match the CSSX solvent density of 0.851 ± 0.0008 g/mL is considered advisable to obtain comparable hydraulic performance. Toward this objective, the density of the NGS has been measured at different temperatures and modifier concentrations to determine that a concentration of 0.650 M Cs-7SB modifier provides the best match for the desired density.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Small to Full Height Scale Comparisons of Cesium Ion Exchange Performance with Crystalline Silicotitanate
The U.S. Department of Energy’s (DOE) Hanford Site houses 56 million gallons of high-level radioactive waste generated from plutonium production from 1944 to 1988. The supernatant waste, currently stored in underground tanks, is intended to be vitrified following filtration and 137 Cs removal at the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Vitrification Facility. The WTP Pretreatment Facility will not be operational for several years. The Tank Side Cesium Removal (TSCR) system is a technology demonstration that will remove cesium from tank waste supernate to support directly feeding LAW to the vitrification facility. The 137 Cs removal is important to meet the WTP LAW contract specification and ultimately for creating a contact-handled waste form. The waste acceptance criteria (WAC) limit for the WTP LAW Facility is <3.18E-5 Ci 137 Cs per mole Na. The TSCR system is skid mounted and employs two key technologies: dead end filtration for solids removal and ion exchange (IX) for cesium removal. Filtration is necessary to protect the functionality of the IX columns. The IX process utilizes sodium form crystalline silicotitanate (CST) IX media produced by UOP Honeywell, LLC (Des Plaines, IL) under the product name IONSIV R9140-B, 18 x 50 mesh, in a lead-lag-polish column configuration. Each column contains a CST bed height of 92 inches and a volume of approximately 157 gallons of CST IX media. The full-scale TSCR operation will run at a residence time of 1.9 bed volumes (BV)/h, which results in a superficial velocity of 7.3 cm/min and a flowrate of ~5 gal/min. Column testing at small (2.5% of the full bed height), medium (12% of the full bed height), and full-height scales has been previously conducted to evaluate process variables and scale up performance of Cs exchange onto the CST. Cesium load performances from various sieve cuts at the small scale indicated that a 30-mesh sieve cut be tested to determine if it better reflects the 12% and full-height column performances at the small scale. Two process flowrates were tested in the small-column configuration with <30-mesh CST and simulant solutions. These tests were compared to full-height column tests at the same residence times to assess CST particle size effect on column scaling. Table S.1 summarizes the observed column performance determined for the two flowrates juxtaposed to the previous work with 5.6 M Na simulant at higher scales (used as benchmarks). The WAC breakthroughs between the small and full-height tests at 1.3 BV/h differed by ~59 BVs. The WAC breakthroughs for small, medium, and full-height tests at 1.8 BV/h were consistent at 240 BVs. The 50% Cs breakthroughs were nominally equivalent for all column tests. The common transition zones and onsets of Cs breakthrough at the 1.8 BV/h tests indicated that the Cs mass transfers were equivalent and thus the <30-mesh CST Cs load performance at the small scale successfully modeled that of the full-height system. It is recommended that the <30-mesh CST be used in subsequent 10-mL CST bed tests.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Evaluation of Material Balance Approaches for Hanford Direct-Feed Low Activity Waste Processing - 20022
The Hanford Site has accumulated millions of gallons of tank waste from reprocessing spent fuel to recover plutonium, uranium, cesium, and strontium. The supernatant from the accumulated tank waste will be treated using a Direct Feed Low-Activity Waste approach. The supernatant will be treated to remove solids and cesium in the Tank-Side Cesium Removal process in the Hanford tank farm, then vitrified in a semi-batch process in the Hanford Waste Treatment and Immobilization Plant (WTP). Currently, each batch of feed is sampled at three locations prior to being fed to the melter: the feed qualification tank in the Hanford tank farm as well as the concentrate receipt vessel (CRV) and melter feed preparation vessel (MFPV) in the WTP. The feed qualification sample is taken from a large batch of accumulated feed, only two to three samples are expected each year. Approximately 275 samples from the CRVs and 1100 samples from the MFPV are expected each year. An evaluation was performed to determine if a material balance based on the feed qualification sample could replace most of the sampling in the CRVs and MFPVs. The evaluation consisted of three elements: (1) determination of the practicality of using a material balance to estimate the stream composition of the CRV and MFPV contents, (2) evaluation of whether the material balance could be automated using the existing process control system, and (3) determination of the uncertainty in glass composition using the material balance approach. It is assumed that periodic sampling at the CRV and MFPV would be performed periodically to re-baseline the material balance, evaluations are in progress to determine the frequency of this periodic sampling. Process sample locations downstream of the melter were reviewed as well, but the partitioning of semi-volatile species in the melter was determined to preclude extending the material balance approach past the melter. It was determined that replacement of the CRV sample location was feasible and did not increase process uncertainty or significantly impact waste loading. Replacement of the MFPV sample was also determined to be feasible, but that measurement of the glass former chemical addition may be needed prior to addition of these chemicals to the MFPV. This measurement could be performed by an in situ laser-induced breakdown spectroscopy (LIBS) system. Limited tests were performed to evaluate LIBS for direct measurements of the low-activity waste melter feed. The use of a material balance would eliminate over 1200 samples each year if only 10% of the CRV and MFPV batches are sampled and could likely allow the WTP laboratory to operate on days only versus 24/7 operation. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Evaluation of Load Behavior for Select Analytes in Hanford Tank Waste
Crystalline silicotitanate (CST) inorganic ion exchanger is a candidate material for remediation of highly alkaline (pH > 14) aqueous nuclear waste streams containing high sodium concentrations (>5 M). In this work, ion exchange column testing with wastes from Hanford tanks AP-105, AP-107, and AW-102 was carried out to study the uptake of 137 Cs to estimate the decontamination factor (DF) value. Additionally, DF values for uptake of Al, Ca, Pb, Np, Pu, U, and Sr were determined and the ion exchange capacity of CST toward these analytes was estimated. Limited data is available on the load behavior of other minor and trace elements with recent CST production lots and this characterization will help improve understanding of the behavior of CST and assist in identifying potential disposition pathways as well as assessing removal capabilities of CST for other components.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Tank Waste LDR Organics Data Summary for Sample-and-Send (Rev.1A)
The presence of organic chemicals regulated under the Resource Conservation and Recovery Act (RCRA) Land Disposal Restrictions (LDR) adds complexity to treating and disposing of the low activity fraction of Hanford tank waste if a low temperature treatment method such as grouting is used (SRNL-STI-2020-00228). The complexity arises from the fact that the baseline vitrification method is considered by the Washington State Department of Ecology (Ecology) as providing adequate thermal treatment for organics; a status not automatically extended to a lowtemperature process, such as solidifying the waste in a cementitious waste form. In addition, the Environmental Protection Agency (EPA) LDR program is intended to ensure that wastes are properly treated prior to disposal. Proper treatment makes hazardous waste less harmful to groundwater by reducing the mobility and/or toxicity of the hazardous constituents in the waste. EPA guidance indicates that stabilization/solidification of waste for organics could be considered impermissible dilution under the LDR dilution prohibition. In addition, waste storage activities at Hanford have required transferring and blending waste within the tank system and these activities have potentially altered the concentrations of the hazardous constituents. The LDR dilution prohibition found in 40 Code of Federal Regulations (CFR) 268.3 states that “… no generator, transporter, handler, or owner or operator of a treatment, storage, or disposal facility shall in any way dilute a restricted waste or the residual from treatment of a restricted waste as a substitute for adequate treatment …”. Hence, if LAW is to be treated using low temperature stabilization (such as cementation), then it is important to demonstrate both how past storage activities have contributed to the removal (by vacuum evaporation), or destruction (by in situ decomposition) of the LDR organics and how future retrieval and waste feed preparation will contribute to their removal (by filtration and ion exchange). Demonstrating these processes helps validate that cementation without additional organic treatment does not necessarily represent impermissible dilution. To aid in implementing cementitious solidification of Low Activity Waste (LAW), WRPS has been developing a regulatory and processing LDR treatment variance strategy termed “Sampleand-Send” that relies, in part, on demonstrating that in situ decomposition reactions along with historic evaporation of tank waste has destroyed or removed most of the LDR organics possibly associated with Hanford Tank Waste (SRNL-STI-2020-00582, SRNL-STI-2021-00453, SRNL-STI-2022-00391). Under the Sample-and-Send concept, Hanford tank waste would be retrieved, processed through a Tank-Side Cesium Removal-like system, and staged as a candidate feed that would then be sampled to confirm the waste acceptance criteria is met for solidification in an LAW cementitious treatment facility. If it can be shown that LDR organics are at concentrations below the waste acceptance criteria (WAC) for cementitious stabilization and have been sufficiently removed (by historic evaporation or by filtration and ion exchange during Cs removal), destroyed (by historic in situ decomposition), or are not soluble in LAW above the WAC then additional organic treatment is not needed prior to creating a cementitious final waste form and the concept of Sample-and-Send would be proposed to establish a non-rulemaking site-specific treatment variance using the specified method of treatment “STABL” to remove sampling requirements of the waste form after treatment. Waste not meeting the WAC could either be routed to the Hanford Waste Treatment and Immobilization Plant for LAW vitrification, or further processed by evaporation or chemical oxidation before solidifying in a cementitious waste form. A key component in implementing the Sample-and-Send strategy is identifying which of the 207 LDR organic compounds associated with the RCRA Part A permit application waste codes for the Double Shell Tanks (DSTs) and Single Shell Tanks (SSTs) and any applicable Underlying Hazardous Constituents (UHCs) from 40 CFR 268.48 should be considered as potentially present and thus subject to regulation. In addition, it is also necessary to understand the solubility volatility, and reactivity of these compounds in LAW to identify which of the potentially present LDR organic compounds are not soluble above regulatory levels or are likely to have been removed by historic evaporation or destroyed by in situ decomposition reactions. If there are potentially present LDR organic compounds that have not been removed or destroyed and are soluble above regulatorily significant concentrations then a treatability variance may be needed for these species to eliminate any concerns pertaining to impermissible dilution. The spreadsheet accompanying this calculation report contains the data and logic computations needed to screen the list of 207 LDR organics associated with Hanford tank waste to identify those potentially present and to indicate which compounds may need to be included in a treatability variance.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Ion Exchange Processing of AP-105 Hanford Tank Waste through Crystalline Silicotitanate in a Staged 2- then 3-Column System
The U.S. Department of Energy (DOE) is working to expedite processing of Hanford tank waste supernate at the Hanford Waste Treatment and Immobilization Plant (WTP). To support this goal, Washington River Protection Solutions, LLC (WRPS, Richland, WA) is designing a system for suspended solids and cesium (Cs/ 137 Cs) removal from Hanford tank waste supernate. The effluent will then be sent to the WTP Low-Activity Waste (LAW) Facility for vitrification. The Cs removal is critical for eliminating the high dose rate associated with 137 Cs and facilitating a contact maintenance philosophy for the LAW Facility. The maximum 137 Cs concentration in the LAW sent to the WTP is targeted to be below the 3.18E-5 Ci 137 Cs/mole of Na waste acceptance criteria (WAC) limit. The filtration and ion exchange systems will be placed near the Hanford tanks and are collectively termed the Tank Side Cesium Removal (TSCR) system.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
The Effects of Contaminant-Aging on Decontamination Efficacy for Rapid Remediation of Concrete Surfaces - 20412
Understanding the relationship between decontamination efficacy and contaminant aging is imperative for developing effective remediation strategies following a large-scale contamination event. Non-destructive decontamination of concrete following a contamination incident becomes more difficult with increased time between contamination and decontamination because contaminants may chemically bind to concrete and/or penetrate into the concrete subsurface. In this work, we evaluated the decontamination efficacy of two decontamination methods on concrete samples contaminated with soluble Cesium-137 (Cs-137) and silica particles with 0.5 μm and 2 μm diameters. Concrete samples were aged between 1 and 59 days, with half of the coupons receiving 1 mL of artificial rainwater about once every three days. After aging, coupons were either decontaminated or analyzed to determine contaminant penetration depths. Coupons were decontaminated with 0.1 M Potassium Chloride (KCI) solution applied either by flowing solution across the contaminated coupon face or non-destructive power washing. Contaminant depth profiles were created by removing the top surface of the coupon at least 15 times, measuring the activity in each removed layer, normalizing the total activity removed to the measured decontamination efficacy following the last layer removal, and determining the penetration thickness using the coupon dimensions, bulk density, and mass of material removed for each layer. Pressurized washing effectively removed particles, but not soluble cesium, despite the majority of Cs-137 being located within the first millimeter of the subsurface. Soluble cesium removal by flowing 0.1 M KCI solution across the contaminated surface dropped to less than 12% within the first ten days. This drop in Cs-137 removal efficacy occurred within the first five days for coupons experiencing artificial rainfall events. Flowing 0.1 M KCI solution across the contaminated face was ineffective at removing particles, likely because particles had localized in surface depressions. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Management of Alkali and Alkaline Earth Fission Products in Used Pyroprocessing Salt
Pyroprocessing of spent nuclear fuel (SNF) involves dissolving metallic fuel into a molten salt electrolyte (typically eutectic LiCl-KCl) and then preferentially depositing actinides onto inert cathodes. Subsequent operations include drawdown of residual actinides and lanthanides from the electrolyte prior to re-using the salt. The recovered actinides are recycled and the recovered lanthanides are disposed as waste. Alkali and alkaline earth metal fission products in the fuel, such as Cs, Sr and Ba, dissolve into the salt during electrorefining. The concentrations of these elements buildup over time in the molten salt electrolyte, which may change the freezing point. The radioactive decay of 137 Cs and 90 Sr (half life 30 and 29 years) generates significant heat and produces strong ionizing radiation fields (β and γ). The increasing heat load and radioactivity as these elements build up in the molten salt requires frequent replacement and disposal of the electrolyte salt. Alternatively, the salt can be treated to remove these and other elements and then recycled to the electrorefiner. An effective strategy to manage these alkali and alkaline earth metal fission products in the molten salt electrolyte would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal. Alkali and alkaline earth metal fission products are extremely stable in molten salt as chlorides--even more stable than the LiCl-KCl eutectic base salt--making them challenging to remove. They are not removed during drawdown operations to recover residual actinides and lanthanides and a separate operation is required to sufficiently purify the salt for reuse. This work is focused on selecting a method for separating Cs, Sr and Ba from the salt recovered from the lanthanide drawdown operation prior to recycling the cleaned salt back to the electrorefiner. Not addressed in this work is the management of the waste stream produced by the separation. This report summarizes the issues to be addressed when developing removal strategies for cesium, strontium, and barium and reviews existing methods to identify suitable methods and any technological gaps in their application.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Closed cycle MHD power generation experiments using a helium-cesium working fluid in the NASA Lewis Facility
A MHD channel, which was previously operated for over 500 hours of thermal operation, ten thermal cycles, and 200 cesium injection tests, was removed from the facility and redesigned. The cross sectional dimensions of the channel were reduced to 5 by 16.5 cm to allow operation over a variety of conditions. The redesigned channel has been operated for well over 300 hours, 10 thermal cycles, and 150 cesium injection tests with no problems. Experiments have been run at temperatures of 1900-2100 K and Mach numbers from 0.3 to 0.55 in argon and 0.2 in helium. The best results to date have been obtained in the helium tests. Power outputs of 2.2 kw for tests with 28 electrodes and 2.1 kw for tests with 17 electrodes were realized. Power densities of 0.6 MW/cu m and Hall fields of about 1,100 V/m were obtained in the tests with 17 electrodes.
Sovie, R. J.↗
THERMAL MODELING OF HANFORD CESIUM AND STRONTIUM CANISTERS DURING SIMULATED LOADING
A computational fluid dynamics (CFD) model was built to simulate planned testing of heater assemblies within a canister and overpack for the Hanford Lead Canister (HLC) project. The HLC is a canister storage system that will contain heaters to simulate the decay heat of nuclear material and provide the canister storage system with environmental conditions equivalent to the operating conditions on a dry storage pad. The HLC will be equipped with long-term data collection and monitoring systems to provide an early warning of corrosion, pitting, cracking, or other signs of canister degradation that might threaten the integrity of the containment boundary over the potentially long term of dry storage. An important part of the HLC development is to make pretest numerical predictions for the behavior of the heated canister during the simulated radiolytic decay heat testing, which simulates the dry storage system during loading operations. The simulated radiolytic decay heat test is planned for mid-2024 in a configuration that includes the heater assembly, overpack, and canister, but with the lids removed to allow loading cesium and strontium capsules into the canister. One of the goals of the test is to evaluate the thermal behavior of the canister and overpack assembly in the ambient air of the test facility, which will provide data critical to validating the thermal models and understanding how the HLC will perform as a system once deployed. To best approximate real-world conditions, the CFD model includes the full air volume of the mock-up truck bay the heated canister test will be performed in, enabling detailed investigation of how the heated canister affects airflow around it. Rigorous pre-deployment testing of the complete HLC cask and canister system is intended to be completed before the HLC is deployed in the 2028 timeframe. This study presents the pre-test temperature predictions of the simulated radiolytic decay heat test. A description of the heater assembly, canister, and overpack system is presented. The model was developed with the commercial CFD software STAR-CCM+. An uncertainty analysis was run with the CFD model to determine the uncertainty in the temperature predictions and provide a range over which the predicted temperatures are expected to vary. The uncertainty analysis was preformed by coupling STAR-CCM+ with the software Dakota, which provides advanced parametric analyses, including quantification of margins and uncertainty with computational models. This work is expected to provide insight into SNF canister behavior.
Carpenter-Graffy, Dina E.↗
Some services of the Time and Frequency Division of the National Bureau of Standards
The Time and Frequency Division of the National Bureau of Standards (NBS) provides several services to the general public. The radio broadcasts of WWV, WWVH, and WWVB supply reliable, unambiguous time signals to many users. The NBS telephone time-of-day service attracts several hundreds of thousands of calls each year. Periodically, the NBS provides courses on specific topics relating to time and frequency technology. In addition to numerous technical papers published each year, the NBS has prepared the first volume of a comprehensive monograph on time and frequency. The results of research in the Time and Frequency Division of the NBS have had significant impact. An active TV time system capable of serving most of the U.S. currently awaits a ruling by the FCC on a petition filed last year on behalf of the NBS by the Department of Commerce. Three more recent developments are: (1) a TV frequency comparator (patent applied for); (2) a method to perform an independent (absolute) frequency evaluation of commercial cesium beam oscillators; and (3) a method of removing one source of frequency drift in commercial cesium beam oscillators.
Barnes, J. A.↗
Identification and Quantification of Technetium Species in Hanford Waste Tank AN-102
Technetium-99 (Tc) generated from the fission of 235U and 239Pu in high yields is one of the most difficult contaminants to be addressed at the U.S. Department of Energy Hanford Site. In strongly alkaline solutions typifying Hanford tank waste, Tc exists as pertechnetate (TcO4-) (oxidation state VII) as well as in reduced forms (oxidation state < VII) collectively known as non-pertechnetate species. Designing strategies for effective Tc management, including separation and immobilization, necessitates understanding the molecular structure of the non- pertechnetate species and their identification in the actual tank waste samples, which would facilitate development of new treatment technologies effective for dissimilar Tc species. Toward this objective, a spectroscopic library of the Tc(I) [fac-Tc(CO)3]+ and Tc(IV, VII) compounds was generated using a range of techniques and applied to the characterization of the actual tank waste supernatant collected from the tank 241-AN-102 at Hanford, WA. A sample of the 241-AN-102 tank waste supernatant was processed to adjust Na concentration to about 5.6 M and remove 137Cs by spherical resorcinol-formaldehyde (sRF) ion exchange resin. Cesium-loaded sRF column was eluted with 0.5 M HNO3. As received AN-102, Cs-depleted AN-102 effluent, and sRF eluate fractions were comprehensively characterized for chemical composition and speciation of Tc using 99Tc nuclear magnetic resonance spectroscopy and X-ray absorption spectroscopy. It was demonstrated for the first time that non-pertechnetate Tc present in the 241-AN-102 tank waste is composed of several low-valent Tc species, including the Tc(I) [fac-Tc(CO)3]+ and Tc(IV) compounds. This is the second experimental observation of the [fac-Tc(CO)3]+ species in the Hanford tank waste and the first demonstration of multiple forms of non-pertechnetate species existing simultaneously in the waste, cumulatively highlighting their importance for the waste processing.
Low Activity Waste (LAW), High Level Waste, nuclea↗
Managing Hanford's Direct Feed Low Activity Waste Program - 20002
The Office of River Protection (ORP) is responsible for the management and completion of the River Protection Project (RPP) mission, which comprises both the Hanford Site tank farms operations and the completion and operation of the Waste Treatment and Immobilization Plant Project. The RPP mission is to safely retrieve, treat, and immobilize Hanford's high-level and low-activity tank waste and close the tank farms to protect the Columbia River. The Direct-Feed Low-Activity Waste (DFLAW) Program is a major subset of the overall RPP mission, providing disposition of low-activity waste. The scope is to retrieve, treat, and immobilize the low-activity waste at the Low-Activity Waste Facility. DFLAW will begin operations no later than December 2023. The DFLAW Program overarches a suite of individual projects, activities, and infrastructure upgrades. The major projects and activities include multiple contractors under both ORP and the Richland Operations Office, underscoring the need for effective integration, coordination, and collaboration. The objectives of the DFLAW Program leadership team are to: - Effectively coordinate and integrate the projects that comprise the DFLAW Program - Manage the interfaces between the projects so that the integrated DFLAW Program is completed successfully - Ensure the DFLAW portfolio of projects operate as required without gaps or conflicts at the interfaces. DOE is accomplishing these objectives through a newly established leadership model that integrates the DOE and contractor work streams and promotes a teamwork model to enable collaborative success. Long considered one of the most formidable cleanup challenges at Hanford, the Department of Energy and its contractors are on the verge of achieving a cleanup commitment that has been decades in the making. DOE is preparing to vitrify (turn to glass) Hanford's chemical and radioactive tank waste using the DFLAW process. The Modified Consent Decree milestone for completing hot commissioning of the Hanford Site's Low- Activity Waste Facility is Dec. 31, 2023. While this milestone may seem distant under normal project management circumstances, it is a relatively short period of time for the operational and cultural transformation necessary to successfully begin treating tank waste at Hanford. Direct feed means separating the waste at a tank farm to remove the more radioactive portion (i.e., solids and cesium) so that the resulting low-activity (less radioactive) waste can be fed directly to the Waste Treatment and Immobilization Plant's Low-Activity Waste Facility. Starting DFLAW will require a singular, intensive leadership focus, along with a sustained, collective commitment to excellence and teamwork by the Department and its contractors. Major upgrades to Hanford's infrastructure will occur prior to startup. Supporting DFLAW operations requires significant integration between the DFLAW facilities - the Waste Treatment Plant, Hanford's tank farms, effluent treatment facilities, disposal facilities - and between all of the site contractors to achieve success. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Disulfonamide Ligands as f-Element Extractants from Alkaline High-Level Waste
Almost 100 million gallons of high level waste (HLW) has been generated from defense reprocessing programs to support nuclear weapons production. HLW is currently stored at the Hanford and Savannah River Sites (SRS) [1]. It contains radioactive components, such as {sup 137}Cs, {sup 99}Tc and {sup 90}Sr, as well as large amounts of non-radioactive species, including solvated cations and insoluble metal hydroxides. HLW has high concentration of soluble hydroxides ([OH{sup -}] = 2 M at SRS [2]), complexing inorganic anions ([NO{sub 3}{sup -}] ∼ 0.65 - 3.7 M at SRS [2]) and a high ionic strength. Even though the majority of actinide component in alkaline HLW is precipitated, studies of An(III) and Ln(III) complexation in highly alkaline solutions in the presence of high nitrate concentrations have showed that soluble nitrate complexes can be formed [3]. The current treatment of HLW at SRS currently consists of two processes: The Actinide Removal Process (ARP) and the Next-Generation Caustic-Side Solvent Extraction process (NG-CSSX). Strontium and actinides are removed by sorption on monosodium titanate through the Actinide Removal Process (ARP) commonly referred to as the alpha-strike step [4]. Then, cesium is selectively extracted from the alkaline media via the Caustic-Side Solvent Extraction (CSSX) process [5]. The CSSX solvent consists of a calix[4]arene-crown-6 extractant dissolved in an inert hydrocarbon matrix containing i) a solvent modifier (alkylphenoxy alcohol) which increases extractant solubility and prevents third phase formation, and ii) a suppressor (trioctylamine or guanidine) - that mitigates surfactant effects. The CSSX process removes {sup 137}Cs selectively and rapidly, yet the post- CSSX aqueous stream may still contain high amounts of An, thus requiring a second ARP treatment (referred to as 'alpha-finishing') for some tanks. Despite the success of the ARP process in removing Sr and An, it often represents the kinetic bottleneck of integrated processing, as it is slower than solvent extraction. Understanding the complexation of actinides by organic ligands that are compatible with the CSSX process could eventually lead to a combined caustic-side Cs/Sr/actinide extraction process with better economics due to a reduced amount of monosodium titanate and/or a shorter required contact time with titanate during ARP. This would ensure a low-activity waste (LAW) stream with no actinides without additional ARP processing post-CSSX. Disulfonamides were studied for extraction of Sm(III) from alkaline aqueous media of pH 10-14 into dichloromethane. Up to 82% of Sm(III) was extracted from solutions of pH 12.5 -13.5 and up to 84% from solutions of pH 10.5 - 11.5. These results show some resemblance with Am extraction results by calixarene ligands previously reported [10]. Kinetic studies demonstrated that even 5 min is enough to complete stripping, whereas extraction is time-limiting process and requires up to 20 h for efficient removal of Sm(III) from alkaline aqueous media. Determination of composition of complexes in solution after extraction by the equilibrium shift method showed a 1:1 Sm(III):dsa-2 complexation ratio for the extracted species.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Development of High Quantum Efficiency UV/Blue Photocathode Epitaxial Semiconductor Heterostructures for Scintillation and Cherenkov Radiation Detection
The primary goal of this research project was to further extend the use of advanced heteroepitaxial-semiconductor crystal growth techniques such as molecular beam epitaxy (MBE) and to demonstrate significant gains in UV/blue photonic detection by designing and fabricating atomically-tailored heteroepitaxial GaAlN/GaInN photocathode device structures. This NASA Explorer technology research program has focused on the development of photocathodes for Cherenkov and scintillation radiation detection. Support from the program allowed us to enhance our MBE system to include a nitrogen plasma source and a magnetic bearing turbomolecular pump for delivery and removal of high purity atomic nitrogen during GaAlN/GaInN film growth. Under this program we have also designed, built and incorporated a cesium activation stage. In addition, a connected UHV chamber with photocathode transfer/positioner components as well as a hybrid phototube stage was designed and built to make in-situ quantum efficiency measurements without ever having to remove the photocathodes from UHV conditions. Thus we have constructed a system with the capability to couple atomically-tailored MBE-grown photocathode heterostructures with real high gain readout devices for single photon detection evaluation.
Leopold, Daniel J.↗
Removal and Deactivation of Bond Sodium from Fast Reactor Blanket Materials
The disposition of sodium-bonded spent nuclear fuel and blanket materials in a repository is complicated by the presence of sodium metal that is used as a thermal bond between the uranium metal fuel and blanket slugs and their cladding. The concern is that the metallic sodium could react with water, producing explosive hydrogen gas, or could exhibit a pyrophoric character. Thus, experimental studies were performed to investigate and demonstrate the removal and deactivation of bond sodium from blanket material in a dry environment. Specifically, bond sodium was removed from unirradiated Fermi-1 blanket elements and an assembly via a melt-drain-evaporate process using elevated temperature and reduced pressure. The effectiveness of sodium metal removal from the blanket materials and their associated cladding was =99.9998%, based on post-test quantitative analyses. The separated sodium metal was collected and subsequently deactivated by reacting it in a molten state with a controlled addition of ammonium chloride particles atop a molten salt medium. In this process sodium chloride is formed and assimilates into the salt pool. The subsequent deactivation of the bond sodium produced a solid ingot of sodium chloride, potassium chloride, lithium chloride, and cesium chloride that was devoid of sodium metal based on post-test analyses of the salt product. Both the sodium removal and deactivation operations were conducted within a dry inert atmosphere enclosure. The results of this study substantiate a path forward for the disposition of sodium-bonded blanket materials, including 34 metric tons heavy metal in irradiated Fermi-1 blanket material currently stored at Idaho National Laboratory.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗