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At least 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Historic Documentation of TA-03-0016. Smashing Atoms & Modern Physics at the Ion Beam Facility, Volume 1

The U. S. Department of Energy, National Nuclear Security Administration, Los Alamos Field Office (Field Office) requested the State Historic Preservation Officer (SHPO) to concur with the National Register of Historic Places eligibility determination contained in the report The Decommissioning of TA-3-16, the Ion Beam Facility (IBF) (LA-UR-05-0775). In this report, Building Technical Area (TA) 3-16 was determined eligible to the National Register of Historic Places. The SHPO concurred with the eligibility determination on March 25, 2005. However, the Field Office later proposed to decontaminate, decommission, and ultimately demolish this property, an adverse effect to Building TA-3-16.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2019

Fuel Fabrication Facility Study for FCF HALEU

Nuclear energy technology developers seek to bring to market new advanced reactor designs and new light water reactor fuel designs, many of which would deploy fuel made from uranium with 235 U enrichment ranging from 5% to 19.75%. There is currently no commercial supply of uranium in this enrichment range, which is higher than used in today’s light water reactors but still lower than the safeguards limit of 20% 235 U. This limitation is a barrier to development of demonstration of these new nuclear energy technologies. As a means to address this near-term need, DOE and Idaho National Laboratory (INL) has considered how the HALEU product from sodium-bonded spent fuel treatment in the INL Fuel Conditioning Facility (FCF), located at the Materials and Fuels Complex (MFC), might be provided and fabricated into some of the new fuel designs. The evaluation presented in this report addresses the potential establishment of two types of fuel fabrication capability at INL, represented generically as a ceramic-type (or pellet-type) fuel fabrication line and a metallic fuel fabrication line. Because the FCF HALEU feedstock contains residual transuranic contamination and a relatively small amount of residual fission products, fabrication with that material is necessarily performed inside of engineered confinement barriers, such as gloveboxes, which would be filled with inert gas for most of the fuel types considered. The throughput rate of the fabrication lines envisioned is 2.5 metric tons of HALEU fuel per year, which is judged to be sufficiently representative of engineering-scale. Space in three existing buildings was identified for consideration in this study of pre-conceptual equipment layouts. The scope and nature of the work needed to prepare each of the three buildings for each type of fuel fabrication was identified (though not in engineering detail), and three-dimensional models of envisioned fabrication equipment were laid out in each of the buildings to assess adequacy of space. The study concluded that either type of fuel fabrication line can feasibly be installed and operated in either of the three buildings. However, each of the three buildings presents different challenges in preparation, due to accommodating current missions that would need to be relocated or necessary decontamination and decommissioning work to free-up space for fuel fabrication, or due to complications of accessibility. Rough order-of-magnitude (ROM) cost estimates for preparing buildings begin at $10M ( 20%/+50%) and would increase with increased complexity of building modifications. The cost of fabrication equipment and gloveboxes is roughly estimated to be $21M (-20%/+50%) for a metallic fuel fabrication line and $27M (-20%/+50%) for a ceramic/intermetallic fuel fabrication line. Nominal operating costs are estimated to be $7M per year (in 2019 dollars), increasing from lower amounts needed as the operating crew is being assembled and trained.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Methylmercury Speciation and Retention Evaluation to Support Saltstone Waste Acceptance Criteria

Mercury is a contaminant of concern in Savannah River Site (SRS) high level waste (HLW). For five decades elemental mercury was used in the catalytic dissolution of aluminum cladding from the enrichedalloy uranium recovery process. Over 60,000 kg of mercury are now distributed in the SRS HLW tanks. The mercury speciation in the liquid waste streams resulting from conditioning HLW sludge for vitrification and in the decontaminated low activity sodium salt solution removed from the tanks was initially thought to be primarily ionic with a minor amount of elemental mercury. Recent sampling detected organic mercury, primarily methylmercury, HgCH 3 + (MeHg + ), and a trace amount of ethylmercury (EtHg) in these streams in addition to the ionic and elemental forms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Fundamental Forces, Lasting Impressions: Historical Documentation of the Sigma Press Building (TA-3-35) Volume 1

The US Department of Energy, National Nuclear Security Administration, Los Alamos Field Office (Field Office) has prepared final documentation for the resolution of adverse effects to Building 35 in Technical Area (TA) 3 at Los Alamos National Laboratory (LANL or the Laboratory). This documentation is being submitted to the New Mexico State Historic Preservation Officer (SHPO). TA-3-35, a Cold War-era press building, was determined eligible for listing in the National Register of Historic Places (Register) in the report Pressing Matters: An Evaluation of the Cold War Buildings at the Sigma Complex (LA-UR-15-28961). This report also identified Building TA-3-35 as excess property to be decontaminated, decommissioned, and demolished by the LANL Footprint Reduction Program in Fiscal Year 2016, an adverse effect requiring mitigation. The SHPO concurred with the both the eligibility determination of, and adverse impact to, Building TA-3-35 on February 29, 2016.

42 ENGINEERING↗
DOE OSTI2018

Municipal and Commercial Equipment for Radiological Response and Recovery in an Urban Environment: State of Science, Research Needs, and Evaluation of Implementation Towards Critical Infrastructure Resilience

This report provides a thorough evaluation of technologies and equipment assets that may be effective at mitigating the effects of a wide-area radioactive contamination event and to restore critical functions to the urban environment. Moreover, equipment that could facilitate decontamination efforts during a recovery phase are evaluated. The summary of methods has been critically evaluated in terms of equipment availability, potential advantages and limitations, and research needs in order to justify and predict its efficacy. By ranking the host of techniques, a concise list of those technologies is provided that might be most impactful to a response and recovery effort and how to direct future effort. The totality of this effort and future work anticipates providing early responders and stakeholders with practical guidance information to expedite an effective response and recovery effort and mitigate the potentially devastating effects of a wide-area radioactive contamination event.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2018

Identification of Potential Waste Processing and Waste Form Options for Molten Salt Reactors

The overall summary of the waste management envelope discussed in this report is represented by the diagram shown in Figure S1. The streams that are included in this discussion include the following from reactor systems in which chlorine-based salts with and without processing and fluorine-based salts with and without processing are being considered: 1) Off-gas streams (e.g., particulates, aerosols, reactive gases, 3 H, water, residual halides, N 2 , O 2 , noble gases); 2) Salt waste streams (e.g., chloride- or fluoride-based); 3) Separated salt streams (e.g., 7 Li recycle, 37 Cl recycle, eutectic salts, fission products); 4) Metal waste streams (e.g., reactor components, plenums); 5) Carbon waste streams (e.g., graphite, SiC); 6) Decommissioning and decontaminating (D&D) waste streams; 7) Operating waste streams

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DOE OSTI2019

U.S. Efforts in Support of Examinations at Fukushima Daiichi: 2019 Evaluations

Much is still not known about the end-state of core materials in each unit that was operating on March 11, 2011 at the Fukushima Daiichi Nuclear Power Station (Daiichi). Information obtained from Daiichi is required to inform Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include Fiscal Year 2019 (FY2019) information, summarizes results from a U.S. effort to review Daiichi information and extract insights to enhance the safety of existing and future nuclear power plant designs. This U.S. effort, which was initiated in 2014 by the Department of Energy Office of Nuclear Energy (DOE-NE), is completed by a group of experts in reactor safety and plant operations that identify examination needs and evaluate recent Daiichi examination data to address these needs. Each year, annual reports are issued that document significant safety insights being obtained in areas of special emphasis. For FY2019, the areas emphasized by the expert panel include component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties related to severe accident modeling progression, these insights have and continue to be used to update guidance for severe accident prevention, mitigation, and emergency planning. Reduced uncertainties in modeling the events at Daiichi improve the realism of reactor safety evaluations that inform future D&D activities. In addition, the process established by this U.S. effort provides an example that Japan may wish to implement in future international efforts to support D&D. As emphasized in this FY2019 report, important insights are being gained from examinations of the affected units at Daiichi. Although many results from prior years have been moved to appendices, key areas of interest are highlighted in the main document. For example, larger intact components and fuel containing materials observed on the floor of the primary containment and retained on ex-vessel structures continue to be emphasized because of insights that may be gained with respect to vessel failure and ex-vessel interactions with structures and concrete. In addition, the main report presents insights gained by focused evaluations of plant data because of potential use of systems and components during severe accident conditions and alternate perspectives regarding accident progression that may be inferred from such data. The report also illustrates how systems analysis codes can be applied to reduce uncertainties about the progression of events at Daiichi and how practical insights obtained from Daiichi can be used to enhance radiation safety for the existing fleet. Recent and planned updates to systems analysis codes are also discussed. As in prior years, this report contains an appendix with a list of examination needs that was updated after U.S. experts reviewed recently obtained information from Daiichi. Additional details for higher priority, near-term, examination activities are also provided. This report also includes an appendix with a description of a website that was developed to assist U.S. experts by providing information in an archived retrievable location.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

MELCOR Modeling of Accident Scenarios at a Facility for Aqueous Reprocessing of Spent Nuclear Fuel

The work presented in this report applies the MELCOR code to evaluate potential accidents in non-reactor nuclear facilities, focusing on Design Basis Accidents. Ten accident scenarios were modeled using NRC's best-estimate severe accident analysis code, MELCOR 2.2. The accident scenarios simulated a range of explosions and/or fires related to a nuclear fuel reprocessing facility. The objective was to evaluate the radionuclide source term to the environment following initiating explosion and/or fire events. The simulations were performed using a MELCOR model of the Barnwell Nuclear Fuel Plant, which was decommissioned before beginning reprocessing operations. Five of the accident scenarios were based on the Class 5 Design Basis Accidents from the Final Safety Analysis Report. Three of the remaining accident scenarios include sensitivity studies on smaller solvent fires. The final two accidents included an induced fire from an initial explosion. The radionuclide inventory was developed from ORIGEN calculations of spent PWR fuel with an initial enrichment of 4.5% U-235 by weight. The fuel aged for five years after a final 500-day irradiation cycle. The burn-up was conservatively increased to 60 GWd/MTU to bound current US operations. The results are characterized in terms of activity release to the environment and the building decontamination factor, which is related the leak path factor used in Department of Energy safety analyses. The MELCOR 2.2 results consider adverse consequences to the filters, ventilation system, and structures as a result of the explosions and fires. The calculations also include best-estimate models for aerosol transport, agglomeration, and deposition. The new calculations illustrate best-estimate approaches for predicting the source term from a reprocessing facility accident.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Dose Modeling for Harborview Research and Training Facility

The Harborview Research and Training (R&T) Facility at the University of Washington (UW) was accidently contaminated from a breached sealed source of Cesium-137. The release occurred during removal of the sealed source from its holder within a portable hot cell in the loading dock area near room 220. The loading dock and northeast corner of the building above the ground floor were impacted by the release, whereas the remaining portions of the R&T Facility were minimally affected. Detailed descriptions of the R&T Facility and the affected area/rooms are included in Section 2. The RESRAD Team in the Environmental Science Division of Argonne National Laboratory visited the site on August 12, 2019, and interviewed the R&T Facility building manager and staff working in the building at that time. Site- and building-specific data were collected including building floor plan CAD drawings, building occupants’ working hours, contaminated rooms photos and their dimensions, building air exchange rates, building materials and “curtain wall” details, and so forth. These data are being used in the dose modeling for potential occupants of the building after it is remediated. The dose modeling was performed using the RESRAD-BUILD computer code, which is designed for the evaluation of radiologically contaminated buildings (Yu et al. 2003). Two hypothetical building occupants of the Harborview R&T Facility were evaluated: a researcher occupying room 726 (a research laboratory on the seventh floor) and a custodian occupying room 220 (loading dock and trash compactor room with a rollup door). A maintenance worker that will be spending less time in impacted areas was not modeled because it will be bounded by other two hypothetical building occupants modeled. A summary description of the RESRAD-BUILD is provided in Appendix A, and the modeling of potential doses to a researcher and a custodian is described in detail in Section 3. A summary and a discussion of the dose modeling results are included in Section 4. The parameters used in the RESRAD-BUILD modeling are included in Appendix B. This report documents the dose modeling results; it does not describe measurements of the actual radioactivity levels at the Harborview R&T Facility, nor does it set any expectation for the actual final radioactivity levels. The radiation doses presented are modeling results based on the generic screening values of Cs-137 and several conservative assumptions and are meant only to assist in decision making regarding decontamination of the Harborview R&T Facility.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2019

Analytical Results from Salt Batch 10 Routine DSSHT and SEHT Monthly Samples

Strip Effluent (SE) Hold Tank (SEHT) and Decontaminated Salt Solution (DSS) Hold Tank (DSSHT) samples from several “microbatches” of the Integrated Salt Disposition Project (ISDP) “macrobatch” Salt Batch 10 have been analyzed for 238 Pu, 90 Sr, 137 Cs, cations (via Inductively Coupled Plasma Emission Spectroscopy - ICPES), and in some cases anions (via Ion Chromatography Anions - IC-A). The analytical results from the current microbatch samples are similar to those from previous macrobatches. In the Modular Caustic-Side Solvent Extraction Unit (MCU), the Cs removal continues to exceed process requirements. The bulk chemistry of the DSSHT and SEHT samples do not show any signs of unusual behavior except for slightly elevated amounts of sodium in the SE samples. This summary represents the final Salt Batch processed through MCU prior to commencement of process and transfer line tie-ins to support the startup of the Salt Waste Processing Facility (SWPF).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Impact of Crystalline Silicotitanate Particle Size on Cesium Removal Efficiency

Washington River Protection Solutions is working on a Tank-Side Cesium Removal (TSCR) system to remove Cs from Hanford tank waste supernate. Removal of 137 Cs from tank waste supernate by TSCR results in a low-activity waste (LAW) form that can be fed directly to the Hanford Waste Treatment and Immobilization Plant (WTP) LAW vitrification facility, where the supernate can be vitrified into a glass form. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIV TM R9140-B), was selected for use as the ion exchange media to remove Cs at TSCR. The particle size used in testing is expected to impact the kinetic performance of the ion exchange media. Therefore, two single-column systems were tested with CST Lot 2002009604. One column contained <25 mesh sieve cut fraction and the other column contained <35 mesh sieve cut fraction. Flowrate through both test systems was adjusted to match the CST contact time expected for the full-scale operation, i.e., matched bed volumes per hour (BV/h) at 1.83 BV/h. The feed was processed downflow through the columns until ~1000 BVs of feed were processed. The waste acceptance criteria (WAC) limit for the WTP LAW vitrification facility is <3.18E-5 Ci 137 Cs per mole Na. For a tank waste containing 156 µCi/mL 137 Cs and 5.6 M Na, up to 0.114% of the influent 137 Cs concentration may be delivered to the WTP (based on 137 Cs content in AP-107 tank waste); this required a Cs decontamination factor of 876. The WAC limit was used as one of the relevant process parameters, along with the 50% breakthrough point, Cs loading per gram of CST, and the mass transfer zone.

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

U.S. Effort Support to Examinations at Fukushima - Meeting Notes with Updated Information Requests (FY2020)

Much is still not known about the end-state of core materials in each unit that was operating on March 11, 2011 at the Fukushima Daiichi Nuclear Power Station (Daiichi). Information obtained from Daiichi is required to inform Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include Fiscal Year 2020 (FY2020) information, summarizes results from a U.S. effort to review Daiichi information and extract insights to enhance the safety of existing and future nuclear power plant designs. This U.S. effort, which was initiated in 2014 by the Department of Energy Office of Nuclear Energy (DOE-NE), is completed by a group of experts in reactor safety and plant operations that identify examination needs and evaluate recent Daiichi examination data to address these needs. Since its inception, comprehensive annual reports were issued that document significant safety insights being obtained in areas of special emphasis: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties related to severe accident modeling progression, these insights have and continue to be used to update guidance for severe accident prevention, mitigation, and emergency planning. Reduced uncertainties in modeling the events at Daiichi improve the realism of reactor safety evaluations that inform future D&D activities. For FY2020, it was decided that the program would gain more benefit from a more concise report that emphasizes new information and insights that affect prior findings and recommendations from the U.S. experts participating in this effort. A key aspect of prior U.S. efforts, the updated list of information requests, is included in this letter report to ensure that they are transmitted to organizations within Japan. In addition, findings and associated recommendations are provided regarding information presented by TEPCO Holdings, Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF), and the Japan Atomic Energy Agency (JAEA). This letter report also continues to emphasize how information obtained from the affected reactors at Daiichi has been and will continue to be used to update severe accident management strategies, improve maintenance activities (especially in areas of radiation protection) and reduce uncertainties in systems analysis code models. In addition, recommendations are included that would expand the use of this information to provide insights regarding maintenance, radiation protection, design, and siting activities of existing and new reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Technology Summary DNATRAX

DNATrax (DNA Tagged Reagents for Aerosol eXperiments) was developed to provide a safe simulant for understanding the transport and dispersion of pathogens. Specifically, it was developed as a safe surrogate for the Bacillus anthrasis spore bioagent, since most naturally occurring spores are not safe for public release. As DNA-tagged sugar particles, DNATrax aerosols are safe for public release. Scientists can add unique DNA barcode sequences to the sugar particles to produce a wide variety of test particles that can be easily differentiated, making it possible to release and test them simultaneously. The unique barcodes also eliminate the need to conduct decontamination between tests, dramatically reducing the cost and time to run experiments. Additionally, the surrogate offers a high-sensitivity detection method to understand key public health issues. We conducted both lab-based and largescale field studies using DNATrax to validate its capabilities for use as a pathogen surrogate. We have demonstrated the ability to produce DNATrax in a range of sizes, adjusting the microparticles to simulate different sized pathogens with ability to create unique sequences of DNA bases, enabling them to create a nearly unlimited variety of test particles.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2020

Remote Handled Hot Cell Facility and Other Research Transuranic Waste (Debris)

This acceptable knowledge (AK) Summary Report has been prepared for the Central Characterization Program (CCP) for remote-handled (RH) transuranic (TRU) waste generated and managed by Sandia National Laboratories/New Mexico (SNL/NM) in Albuquerque, New Mexico. The waste described in this report was predominately generated in the SNL/NM Hot Cell Facility (HCF) during the removal and packaging of experimental material and decontamination operations in Building 6580 at Technical Area (TA)-V. The waste stream also includes a very small amount of waste (estimated at less than 1 gram of fuel similar to that used for research at the HCF) originating from classified research at TA-I. In addition, the waste also includes secondary waste termed by SNL/NM as process generated waste (PGW), created during repackaging operations for this waste at the Auxiliary Hot Cell Facility (AHCF). The waste has been stored at the Sandia Pulsed Reactor (SPR) Dense Pack Storage Facility within TA-V or the Manzano Bunkers located at Manzano Base within Kirtland Air Force Base. All of the waste is being repackaged at the AHCF. This report was prepared in accordance with CCP-TP-005, CCP Acceptable Knowledge Documentation (Reference 1), to implement the AK requirements of DOE/WIPP-02-3214, Remote-Handled TRU Waste Characterization Program Implementation Plan (WCPIP) (Reference 2); Waste Isolation Pilot Plant Hazardous Waste Facility Permit, Waste Analysis Plan (WIPP-WAP) (Reference 3); and DOE/WIPP-02-3122, Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WIPP-WAC) (Reference 4).

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

Tank 50H Mixing Pump Run Time Reassessment

Tank 50H is required to operate a single rotating standard slurry pump for 4.5 hours prior to transfers to the Salt Solution Receipt Tanks (SSRTs) in the Saltstone Production Facility (SPF). This mixing time is required to adequately mix miscible and immiscible liquids within Tank 50H and meet the Saltstone Waste Acceptance Criteria (WAC). The miscible liquids are aqueous solutions of dissolved salts, while the immiscible liquids include organic droplets such as Isopar® L. To support the accelerated salt processing rates from the Salt Waste Processing Facility (SWPF), a desire to reduce the transfer time of Decontaminated Salt Solution (DSS) downstream of SWPF was identified. This document describes the analysis performed to reassess the technical basis for the 4.5 hour mixing requirement for Tank 50H.

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

Extraction, Scrub, and Strip Test Results for the Salt Waste Processing Facility Caustic Side Solvent Extraction Solvent Sample

An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D (Cs) ), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D (Cs) measured 13.2, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges

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

Molten Salt Reactor Engineering Study for Off-Gas Management

In previous work published by the current team, the waste processing and waste form options were summarized for molten salt reactors (MSR) (Riley et al., 2018b; Riley et al., 2019). The primary types of waste from an MSR are summarized in Figure S1 and include (1) off-gas streams, (2) salt waste streams, (3) separated salt streams, (4) metal waste streams, (5) carbon waste streams, (6) decommissioning and decontaminating (D&D) waste streams, and (7) operating waste streams, or those that are generated from maintenance procedures. The primary focus of this report will be on management of the off-gas stream as it represents the pressure boundary for fuel-salt MSRs and, thus, is required for fission product confinement during reactor operation and reduction of the source term during a reactor accident.

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

ZAM Modeling Study to Support the Tank Closure Cesium Removal (TCCR) 1A Unit

Currently at the Savannah River Site (SRS), the Tank Closure Cesium Removal (TCCR) is an “at-tank” process designed to remove cesium from aqueous tank waste. Cesium will be removed by ion exchange using the engineered IONSIV® R9120 form of Crystalline Silicotitanate (CST) media. The current TCCR design has two columns online in a lead-lag configuration to optimize media usage and achieve the target decontamination. Once the lead column is saturated with cesium, it will be removed from service, the lag column will rotate into the lead position, and a new column with fresh ion-exchange media will be placed into the lag position. The TCCR process for cesium removal from Tank 10H is detailed in X-SOW-H- 00002. Demonstration of the system began in early calendar year 2019 with two batches of salt solution generated by dissolving saltcake in Tank 10H, followed by processing of these batches through the TCCR system. A third TCCR Tank 10H dissolved saltcake batch is scheduled for processing soon. Upon completion of the demonstration with Tank 10H dissolved saltcake, Tank 9H salt solution will be transferred to Tank 10H and subsequently processed through the TCCR unit with new CST media (referred to as R9120-B 30x60) added to new IX columns. The TCCR processing campaign of Tank 9H salt solution is referred to as TCCR-1A .

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