SEARCH · Engineering Papers

Results for “RADIOACTIVE WASTE PROCESSING”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 163 records · Page 9

All sources NASA NTRS DOE OSTI

DOE OSTI2016

Progress Updates for the Hanford Waste Treatment and Immobilization Plant Project - 16281

The Waste Treatment and Immobilization Plant (WTP) Project is vital to the U.S. Department of Energy's (DOE) mission to clean up radioactive waste at the Hanford Site, located in southeastern Washington State. The overall WTP Project objective is to design, build, and commission the facilities and systems that will treat and immobilize approximately 56 million gallons of radioactive waste stored in Hanford's 177 underground storage tanks. The combination of technologies in the WTP makes it a first-of-a-kind facility at a scale larger than ever built for radioactive waste processing. The WTP includes several facilities that will contribute to waste processing: (1) the Pretreatment (PT) Facility, which is designed to receive waste retrieved from the Hanford tank farms and separate it into low-activity and high-level waste streams; (2) the Low-Activity Waste (LAW) Facility, which will vitrify the low-activity waste stream; (3) the High-Level Waste (HLW) Facility, which will vitrify the high-level waste stream; (4) the Analytical Laboratory (Lab), which will support both vitrification facilities' operations; and (5) the Balance of Facilities (BOF), which will provide industrial support capabilities and infrastructure, like electricity and steam, needed for WTP operations. In 2012, DOE suspended production work on the WTP's Pretreatment (PT) Facility and to a lesser degree the High-Level Waste (HLW) Facility because of unresolved technical issues associated with those facilities. DOE has established an approach that would facilitate immobilization of tank waste to begin as early as practicable without waiting for completion of work to resolve the technical issues associated with the PT and HLW Facilities. This approach enables LAW treatment to begin before the PT and HLW facilities are completed by transferring low activity waste directly from the Hanford tank farms to the LAW Facility - referred to as Direct Feed LAW (DFLAW). ORP has placed the highest priority on completing the Low Activity Waste Facility, Balance of Facilities, and Analytical Laboratory (collectively known as LBL) in support of the planned phased start to the Hanford Tank waste treatment mission. DOE has requested, and the WTP contractor has delivered contract modifications proposals for: (1) design of a permanent capability to accommodate DFLAW in WTP; (2) procurement and construction of modifications to enable DFLAW operations in WTP, and (3) completion of LBL work scope in the current contract through hot commissioning in a manner to accommodate a phased start toward completing the WTP mission. DOE is reviewing and approving contract modifications in a sequence and schedule that allows for an orderly completion of the LBL Facilities and to begin treating low-activity tank waste as soon as practicable. In August 2014, ORP authorized the WTP contractor to resume all engineering work necessary to finalize the design of the HLW Facility. Authorization for HLW equipment procurements and construction is forthcoming, based on conditions that must be satisfied for a full release of these activities, including completion of an update to the Preliminary Documented Safety Analysis. DOE has also established a comprehensive plan for resuming production activities for the PT Facility that is similar to the process established and implemented for the HLW Facility. Resolution of the technical issues and establishment of a revised safety basis are key prerequisites to resuming engineering, procurements, and construction. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Progress updates for the Hanford waste treatment and immobilization plant project - 15593

The Waste Treatment and Immobilization Plant (WTP) Project is vital to the U.S. Department of Energy's (DOE) mission to clean up radioactive waste at the Hanford Site in southeastern Washington State. The overall WTP Project objective is to design, build, and commission the facilities and systems that will treat and immobilize approximately 211 983 m{sup 3} (56 million gallons) of radioactive waste stored in 177 underground storage tanks. The combination of technologies in the WTP makes it a first-of-a-kind facility at a scale larger than ever built for radioactive waste processing. In recent years, the project has encountered issues that affected project design completion, especially in the Pretreatment and High-Level Waste Facilities. These included technical issues, changes in requirements, changes in funding, weaknesses in the nuclear supply chain, identification of new hazards, and new information about the tank waste. The combination of these issues led to a suspension of construction of the Pretreatment Facility and a partial suspension of construction of the High-Level Waste Facility in mid-2012. Since then, the WTP Project has made significant progress in establishing and implementing a strategy for resolving these issues. This strategy focuses on resuming full production work in a sequence that fully supports the DOE plan for a phased approach for starting treatment of the Hanford tank waste, beginning with the low-activity waste streams. Concurrent with the technical and programmatic issues, the DOE Office of River Protection has made numerous changes in the processes it uses to manage and oversee the WTP Project. These changes have had a beneficial impact on the progress the project has made in establishing a path forward for resuming production activities, and on planning for the necessary contract and baseline changes needed to complete the project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Corrosion Testing of Carbon Steel Exposed to Sludge Heel Chemical Cleaning Solutions - 16109

As efforts continue to treat and dispose of millions of gallons of legacy radioactive materials from the production of nuclear weapons, non-compliant waste storage tanks will gradually be emptied of the bulk waste volume leaving heel materials requiring removal prior to tank closure. The waste heel slurries are distributed on the floor of large, million gallon tanks, which frequently contain numerous obstructions that limit the effectiveness of mechanical removal methods. As a result, chemical cleaning methods are needed for the effective removal of the heels as well as chemical scales that are present on the tank walls and other interior surfaces. Two tank cleaning technologies have already been implemented at the Savannah River Site (SRS): Low Temperature Aluminum Dissolution (LTAD) and Bulk Oxalic Acid Cleaning (BOAC). Recent chemical cleaning efforts on SRS Tank 12 were very successful with regard to bulk sludge heel (especially for Al, Fe, and U phases) and beta/gamma radionuclide removal. Although chemical cleaning using these technologies has been shown to be effective, no disposition path has been identified for oxalate added during BOAC, and insoluble oxalate salts are accumulating within the SRS tank farm and waste processing facilities (evaporators, etc.). Extensive sludge washing is also required to remove moderately soluble sodium oxalate salts prior to sludge vitrification in the SRS Defense Waste Processing Facility (DWPF). As a result, oxalate additions to the tank farm need to be minimized by the use of supplementary acids to assist sludge removal in OA or the use of other cleaning reagents or processing strategies. Previous SRNL testing revealed the importance of pH control for BOAC and recommended the use of a supplementary acid (dilute HNO{sub 3}) with OA to minimize oxalate additions. Primary drivers in SRS Tank Closure Performance Assessments, which evaluate the fate and impact of tank sludge residuals on a geological timescale, are the removal of alpha emitting radionuclides present at low concentrations such as Pu, Am, and Np, which are not highly soluble in currently utilized chemical cleaning reagents. Scoping studies conducted at SRNL revealed promising methods to dissolve the actinides within the HLW tank heels. Oxidation of the actinides with permanganate in either strong caustic (10 M NaOH) or dilute acidic (0.2 M HNO{sub 3}) solutions was shown to result in dissolution of oxy/hydroxide phases of these metals in the absence of major sludge phases. The waste tanks at SRS and Hanford are constructed of carbon steel. Carbon steel corrodes rapidly in most acidic and/or oxidizing environments. Previous electrochemical corrosion testing indicated that the corrosion rates in dilute HNO{sub 3} with OA may result in manageable corrosion rates, over short exposure periods. Additional corrosion testing was performed to provide data for the nitric acid/OA blend as well as the acid/permanganate and caustic permanganate based chemical cleaning solutions. Corrosion rates from the passive coupon tests for the nitric acid/oxalic acid blend were significantly greater than those observed previously with the electrochemical testing. This result may be due to simultaneous electrochemical and chemical dissolution of the steel. The corrosion rate data from the coupon tests in the nitric acid/oxalic acid blend were evaluated to determine the degree of potential structural damage. The corrosion rates, although relatively high, would not be expected to cause damage that would reduce the capacity of the tank primary wall if the process is completed within a month. The corrosivity of sodium permanganate in four proposed cleaning solutions, 1 nitric acid solution and 3 sodium hydroxide solutions, was studied by electrochemical methods at room temperature. The corrosion rates were significantly less aggressive than in the nitric acid/oxalic acid blend corrosion tests. Passive coupon tests needs to be performed in these solutions to ensure that chemical dissolution of the metal is not occurring simultaneously. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Nuclear Fuel Traces Definition in Storage Ponds of Research VVR-2 and OR Reactors in NRC 'Kurchatov Institute' - 16022

The Gas Plant complex is the experimental base of the Institute of Nuclear Reactors, which is part of the Kurchatov Institute. In 1954 the commissioning of the first Soviet water-cooled water-moderated research reactor VVR-2 on enriched uranium, and until 1983 the complex operated two research water-cooled water-moderated reactors 3 MW (VVR-2) and 300 kW (OR) capacity, which were dismantled in connection with the overall upgrades of the complex. The complex has three storage ponds in the reactor building. They are sub-surface vessels filled with water (the volume of water in each is about 6 m{sup 3}). In 2007-2013 the spent nuclear fuel from storages was removed for processing to 'Mayak'. Survey of Storage Ponds by Underwater Collimated Spectrometric System shows a considerable layer of slime on the bottom of ponds and traces of spent nuclear fuel in one of the storage pond. For determination qualitative and the quantitative composition of radionuclide we made complex α-, β-, γ - spectrometric research of water and bottom slimes from Gas Plant complex storage ponds. We found the spent nuclear fuel in water and bottom slime in all storage ponds. Specific activity of radionuclides in the bottom slime exceeded specific activity of radionuclides in the ponds water and was closed to levels of high radioactive waste. Analysis of the obtained data and data from earlier investigation of reactor MR storage ponds showed distinctions of specific activity of uranium and plutonium radionuclides. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Career Development through Discreet Simulation Modeling - 19009

The cleanup mission of the Department of Energy (DOE) Office of River Protection (ORP) at the Hanford Site is to immobilize 212,000 m3 (56 million gallons) of nuclear waste, currently stored in 177 underground tanks, at the Waste Treatment and Immobilization Plant (WTP) and associated supplemental treatment facilities. Waste transfers to and from the WTP utilize the intricate Hanford Waste Feed Delivery (WFD) system, a network composed of: double-shell tanks (DSTs), single shell tanks (SSTs), the effluent treatment facility (ETF), a large number of pumps, valves, and leak detectors, and thousands of feet of underground transfer lines. This equipment must be prepared to safely and reliably transfer waste to the WTP. This study used discrete event simulation modeling to explore the behavior of the waste transfer system to identify areas of improvement. Two models and their subsequent results are presented: the Effluent Treatment Facility and the next-to-be-retrieved single-shell tanks (A/AX Farms). This study focused on identifying throughput bottlenecks, operational constraints, choke-point analysis, and redundancy in equipment, process design, and operational strategy. Analysis performed on the ETF OR Model indicated that the existing facility operation strategy will not result in a sufficient throughput to prevent WTP melter idling. Furthermore, in the absence of equipment failures, the facility was still shown to be unable to reach the required throughput to prevent melter idling. It is concluded that ETF system maintenance events are currently the main bottlenecks in the ETF facility. Analysis performed on the SST (A/AX) OR model indicate that the main drivers of the retrieval schedule are the hard retrieval start dates for each SST tank. In addition, it was concluded that operating the ventilation trains in parallel results in 15 times less retrieval interruptions and shortens the retrieval durations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Progress on Resolution of Technical Issues at the Hanford Waste Treatment And Immobilization Plant Project - 16285

The U.S. Department of Energy's (DOE) Waste Treatment and Immobilization Plant (WTP) Pretreatment (PT) Facility has a number of first-of-a-kind technical challenges that have required additional expert analysis and testing activities to resolve. In 2006 an External Flowsheet Review Team identified 28 technical vulnerabilities with the plant design or future operability. Over a period of three months starting in September 2012, a number of top scientists and engineers conducted a review of the major technical issues associated with the design and operability of the WTP. These experts also provided independent advice on some of the broader challenges associated with completing the Hanford tank waste mission - highlighting the linkage between the physical and chemical properties of the tank waste and the PT Facility technical issues. In 2012 and early 2013 DOE's Office of River Protection (ORP) directed the WTP contractor to suspend production work on the Pretreatment Facility until the outstanding technical issues could be resolved and the nuclear safety basis could be updated and brought into alignment with any updates to the facility design and/or process flowsheet. Specific plans for resolving each the technical issues have been developed and significant progress has been made over the past year in identifying solution sets for each of the issues. The eight remaining technical issues (referred to as T1 to T8) are as follows: (T1) Hydrogen Gas Release from Vessel Solids; (T2) Potential Criticality in Pretreatment Facility Vessels; (T3) Hydrogen in Piping and Ancillary Vessels; (T4) Pulse-Jet Mixing and Control; (T5) Erosion and Localized Corrosion in WTP Vessels and Piping; (T6) Design Redundancy in Black Cells/In-Service Inspection; (T7) Black Cell Vessel Structural Integrity; (T8) Facility Ventilation. This paper provides a synopsis of the progress that has been made on resolving the Pretreatment technical issues over the past year, and how resolution of the issues will enable DOE to make an informed decision to resume design and production engineering for the Pretreatment Facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Progress Updates for the Hanford Waste Treatment and Immobilization Plant Project - 17275

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Integrity Assessment of Hanford 242-A Evaporator Facility E-A-1 Re-boiler Vessel via Fluorescein Liquid Tracer Test Methods - 17357

The 242-A Evaporator Facility is critical to the United States Department of Energy's (DOE) Hanford Site clean-up mission due to limited space within the existing Double-Shell Tanks (DSTs) used to store nuclear and chemical waste until future processing. The 242-A Evaporator Facility is currently the sole waste volume reduction process used at Hanford, and the E-A-1 Re-boiler is essential to the Evaporator process. Due to the potential for cross-contamination between the Re-boiler tubes through which waste passes and the tube sheets of uncontaminated steam, an assessment of the Re-boiler vessel's integrity is required for continuing operations of the facility. In order to meet these requirements, an independent integrity assessment was designed to use fluorescein liquid tracer (LT) to verify the leak-tightness of the Re-boiler vessel in accordance with Washington River Protection Solutions, LLC (WRPS) tightness test standards, as well as, the ASME Boiler and Pressure Vessel (B and PV) Code, Section III, Division 1, NB-6300. LT test methodologies were evaluated and determined to be appropriate for a vessel with restricted access, such as the E-A-1 Re-boiler, as it provided the needed sensitivity, reliability, and cost; as well as, exceeded the leak-tightness specifications required by both WRPS and ASME B and PV Code standards. The integrity test requires water laced with fluorescein be circulated through the waste processing loop of the 242-A Evaporator and samples of steam condensate be collected and analyzed for traces of fluorescein to determine leak-tightness. Liquid condensate sampling was conducted prior to and during testing and the samples were analyzed for fluorescein tracer to determine the amount of fluorescein that flowed out of the tubes of the Re-boiler into the steam side of the vessel. Given the established criteria for 'leak-tight,' test results determined that the Re-boiler passed the integrity assessment. After further scrutiny of the 2014 test and data results, it was determined that although the Re-boiler was 'leak-tight,' as established by the test criteria, there was concern of future leakage. The DOE and the DOE Office of River Protection (ORP) requested that WRPS repeat the integrity assessment in two and a half years to determine if the potential leak was growing, which could threaten continuing operations of the 242-A Evaporator. The 2017 test has been redesigned using lessons learned from the 2014 test and was modified to account for ergonomic and safety considerations, ease of repeatability, and accuracy of results. The liquid tracer methodology and test plan developed for the Re-boiler will continue to be implemented in 5 year increments to monitor current and future integrity. The method may also prove useful for other aging components across the Hanford Site, where appropriate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Preliminary Preclosure Accident Consequence Analysis for the APM Conceptual Design - 18090

The Nuclear Waste Management Organization (NWMO) is implementing Adaptive Phased Management (APM), Canada's plan for the long-term management of used nuclear fuel. The APM approach encompasses centralized containment and isolation of the used fuel in a Deep Geological Repository (DGR) in a suitable rock formation, such as crystalline rock or sedimentary rock, in an informed and willing host. This paper describes a preliminary analysis of the potential dose consequences to the public under accident conditions associated with the operation of a DGR and related surface processing facilities. These results are for a generic site and considers exposure to a person standing at a potential fenceline location under conservative atmospheric conditions. Radiological dose consequences to the public were calculated for accidents identified in the preliminary hazard identification. The presence or absence of ventilation system HEPA filters was also considered in combination with specific accident scenarios. A Gaussian dispersion model considering the most stable atmospheric condition (Pasquill stability category F) was used to estimate the exposure of a member of the public in the direct plume path at the location of maximum airborne radionuclide concentrations. The public dose was calculated via the inhalation, air immersion and ground exposure pathways. For all accidents considered, the calculated public dose was low, up to about 0.8 mSv for a fall of an UFTP with container breach. The total dose to the public was also calculated at various distances from the Used Fuel Packaging Plant, Ventilation Shaft, and Main Shaft. For all accidents considered, the calculated dose to the public was low, below 1 mSv even at the minimum 100 m distance evaluated. Sensitivity cases were also carried out to determine the effect of stack release height, the effluent exit velocity, and the release orientation on the calculated minimum site boundary distance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

System Health and Status Reporting is Used to Improve Infrastructure Systems at the Hanford Site - 20449

Reliable physical infrastructure systems are vital to completing the clean-up mission and the start-up and commissioning of new nuclear waste processing facilities at the Hanford Site. Key infrastructure systems at Hanford include water, sewer, and electrical utilities, roads, telecommunications/network systems, and fire systems. All of the infrastructure systems are non-nuclear, General Service systems that are being maintained under formal engineering configuration control and none are specifically credited in any documented safety analysis (DSA). Starting in calendar year 2017, the Hanford Site's infrastructure and services contractor started producing system health and status reports for a total of 15 critical infrastructure systems as means to get a better understanding of each system's health and how it could be improved to ensure better and more reliable service delivery. As the reporting, tracking, and use of the metrics has matured, the number of systems for which reports are being developed has been expanded to a total of 23. Each of the critical infrastructure systems are assigned primary and alternate design authorities that complete qualifications on each assigned system. The design authorities actively manage and maintain configuration control on each of their systems and track metrics associated with availability, maintenance (corrective and preventive), and configuration of the technical baseline. The metrics are based on the principles established in Chapter V of DOE O 420.1C, Facility Safety, [1] and have been tailored as appropriate to each of the key infrastructure systems. An overall assessment of system health is performed and documented on a routine basis (e.g., quarterly). Included in this documentation is a quantitative score for each of the metrics (i.e., availability, maintenance, configuration management) along with results of system walk-downs, description of any permanent design modifications, identification of new or emerging issues and vulnerabilities affecting the system, and an assessment by the design authority of actions necessary to improve the system health score. The overall status scores are calculated using a formula and based on a pre-determined grading scale, a color is assigned for simple presentation. The information gained from developing the reports is being used to (1) assist with making timely and better informed decisions through the use of solid technical bases that are increasing the overall reliability and capabilities of the Site's infrastructure systems as new expense and capital projects in the Reliability Projects Investment Portfolio (RPIP) are prioritized and executed; (2) feed the risk management program, and (3) integrate the technical content of Master Plans that describe the Site's needs for the system and the system's capabilities to meet those future needs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

CST Batch Contact Equilibrium Testing of TCCR Tank 10H Batch 1 and 1A Waste Supernate Samples

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate media to process radioactive waste solutions for the removal of Cs-137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium saltcake solids) within the tank followed by at-tank ion exchange column treatment. Equilibrium batch contact tests conducted in the SRNL Shielded Cells laboratory and associated analyses indicated that both the Tank 10H Batch 1 (pre-production) and 1A (production) waste supernates are unique solutions for which CST cesium removal performance is lower than predicted by ZAM modeling. Cesium distribution coefficients (K d ), percent removal, and CST loading data for the TCCR production supernate batch are provided in Table ES-1. ZAM-based cesium loading predictions are compared to the observed loading for the production batch in Figure ES-1, where significantly lower performance was observed than predicted by the model. Even lower cesium loading results (versus ZAM) were observed previously for the in-tank “teabag” batch contact samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Steam Reforming Process for Treating Radioactive Waste - 18172

In 2005 the THOR steam reforming fluidized-bed reactor (FBR) technology was selected by the DOE for treatment of sodium-bearing waste (SBW) at the INL. During commissioning a number of equipment reliability and process control challenges were encountered preventing the start of radiological operations. To address these challenges the team adopted a phased systematic mechanistic approach, underpinned by scientific principles and supported by some of the world's leading experts in FBR technology. The Phase 1 effort broadly addressed the entire plant and associated modifications. Phase 2 focused on the most critical issues associated with the denitration and mineralization reformer (DMR) vessel: specifically, mitigation of wall scale (mineral deposits) that formed on the inside of the DMR; mitigation of friable agglomerates (sand-castles) that caused operational instabilities; repair of damaged lower fluidizing gas distributor (ring header), and resolution of auger-grinder (A/G) reliability issues. Basic chemistry and fluidization, as well as operational issues were evaluated using engineering analysis, modeling, bench-scale, pilot scale, and full-scale runs to validate the resolutions. The bench scale testing revealed the need to ensure at least a 5 vol% concentration of CO{sub 2} in the fluidizing gas to minimize undesirable reaction intermediates that could form in less fluidized regions of the bed. Engineering analysis and modeling results pointed to the need to redesign the fluidizing gas distributor to overcome fluidization problems. Chemistry evaluations guided efforts to minimize the deposition of wall-scale. The process modifications were confirmed on the pilot level and are being validated in the full-scale unit (Phase 3). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Cesium Removal Performance Comparisons of Crystalline Silicotitanate Media Batches with Savannah River Site Waste Simulant

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with Crystalline Silicotitanate (CST) media to process radioactive waste solutions for the removal of Cs137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium salt cake solids) within the tank followed by at-tank ion exchange column treatment. Plans are underway to prepare and install a second TCCR unit at SRS. Capacity and particle size differences exist between archived (IE-911) and more recently-prepared CST media batches (9120- B and 9140-B). Side-by-side comparison testing was performed to evaluate the cesium removal performance of each batch to aid in selecting the preferred CST batch and media characteristics to load into the second TCCR unit. Batch contact equilibrium and flow-through column tests have been conducted with three CST batches using an SRS Average Simulant. Simulant batch contact equilibrium cesium loading results for the three CST batches (including two lots of one batch) are provided in Table ES-1. The 35 °C data indicates that the archived IE-911 batch has a higher cesium capacity than recently-prepared CST media and that the minor TCCR CST 9120-B media lot (2099000035) has similar cesium removal performance to the major TCCR lot (2099000034). Tests conducted at 25 ºC for the archived IE-911 CST batch indicated lower cesium removal performance with this simulant batch than was observed recently with a different SRS Average Simulant batch. A dilution factor (DF) is typically utilized when modeling engineered CST media cesium loading performance to account for mass contributions from the binder material. In cases where CST performance is lower than expected, this factor includes corrections for the binder and for low performance. The ZAM (Zheng, Anthony, and Miller) Isotherm Model DF values are provided in Table ES-1 for each simulant batch contact result. DF values near 0.5 were determined for the 9120-B and 9140-B CST batch contact tests while DF values near 0.6 (20% higher) were determined for the tests with IE-911. These dilution factors are lower than recently observed with a different SRS Average Simulant batch (9140-B DF = 0.68; IE-911 DF = 1.0). Based on these results, it appears that some component in the simulant solution used for equilibrium testing may have resulted in reduced cesium loading on the CST media.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

WTP Modifications to Support DFLAW: LAW, BOF, Lab - 16219

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is a one-of-a-kind nuclear waste treatment facility being designed and constructed for the U.S. Department of Energy by Bechtel National, Inc. and principal subcontractor URS to process and vitrify radioactive waste stored in underground tanks at the Hanford Site. The Tank Operations Contractor, Washington River Protection Solutions, has responsibility to retrieve, stage, pretreat, and deliver the tank waste feed to WTP. To treat tank waste as soon as practicable, a scenario is being considered to deliver pretreated feed straight to WTP's Low-Activity Waste (LAW) Vitrification Facility. This scenario is called the Direct Feed LAW (DFLAW) program. Balance of Facility utility and system facilities are designed for full WTP operation. Thirteen of these are needed to support reduced throughput during DFLAW operation, of which eight require moderate to extensive changes. Extensive changes include permanent plant modifications that allow DFLAW to operate if the Pretreatment (PT) Facility is unavailable. Under full WTP operation, effluents from the LAW Facility will be returned to the PT Facility for further processing and/or transfer to the Liquid Effluent Retention Facility/Effluent Treatment Facility. The PT Facility also handles all flush water from the tank farm transfers. To address these requirements, a Value Improvement Project was performed. The recommended approach is to provide a 'split stream' Effluent Management Facility (EMF) with an evaporator to support DFLAW. The required modifications for DFLAW facilities are discussed in this paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Cesium Removal Performance Comparisons of Crystalline Silicotitanate Media Batches with Savannah River Site Waste Simulant

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Characterization and CST Batch Contact Equilibrium Testing of Modified Tank 9H Process Supernate Samples in Support of TCCR

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate (CST) media to process radioactive waste solutions for the removal of ¹³⁷Cs. TCCR currently focuses on dissolving Savannah River Site (SRS) radioactive tank waste (primarily sodium saltcake solids) within Tank 10H followed by at-tank ion exchange column treatment. Two supernate batches from Tank 10H have been processed through the TCCR unit and processing of a third batch is expected soon. After processing of this supernate batch, plans are to replace the CST columns and process dissolved salt solution from Tank 9H through Tank 10H and then through the new CST columns installed in the TCCR unit. The new columns are expected to contain either a media similar to an archived CST batch (IE-911) or the R9120-B CST media used in the current TCCR columns (the two materials are fundamentally the same; just different specifications, product names, and preconditioning steps). Samples of Tank 9H dissolved saltcake were received at the Savannah River National Laboratory (SRNL) and characterized. The Tank 9H supernate contained a high sodium concentration (~9.6 M Na⁺) and will require dilution to near 6 M [Na⁺] prior to processing through the TCCR unit. The cesium concentration in Tank 9H is currently significantly higher than was observed with Tank 10H. Three dilutions of the Tank 9H supernate were conducted to mimic possible dilutions that could be conducted in the tank farm prior to TCCR processing using inhibited water, sodium hydroxide, and sodium nitrate solutions. Dilution #1 was prepared by diluting the Tank 9H supernate by a factor of 1.6 with inhibited water, Dilution #2 was prepared by diluting the Tank 9H supernate by a factor of 3.7 with a mixed sodium hydroxide/sodium nitrate diluent, and Dilution #3 was prepared by also diluting by a factor of 3.7 but with sodium hydroxide only. All three dilutions have similar Na⁺ concentrations (~6 M), but Dilutions #2 and #3 contain significantly less cesium. Major and key minor components of the diluted Tank 9H supernate samples are provided in Table ES-1. Batch contact equilibrium tests were conducted with the Tank 9H dilutions and the two different CST media batches being considered for use in the new TCCR columns. Results are summarized in Table ES-2 and compared to ZAM model predictions. The highest cesium distribution coefficient and percent removal were observed with Dilution #3. IE-911 CST (an archived CST media batch) was more effective at removing cesium than the more recently prepared R9120-B CST media, though both media samples removed >88% of the cesium and the differences may not be statistically different considering the overall uncertainty. Based on the results, maximum cesium loadings were calculated for each CST media type and Tank 9H dilution. In general, maximum cesium loadings from dilutions of this supernate batch are quite high (approaching 0.1 mmol total Cs⁺/g CST for Dilution #1). The highest calculated maximum ¹³⁷Cs loading for the Tank 9H dilutions using the ZAM model with input of the tank compositions and batch contact results was 207 Ci/kg CST (Dilution #1 with IE-911 CST). In all cases, higher maximum cesium loading values were predicted for IE-911 CST versus R9120-B, although the differences varied considerably between the dilutions. The maximum loading value for IE-911 CST with Tank 9H Dilution #1 was only 7% higher than the maximum loading for R9120-B. The maximum loading value for IE-911 CST with Dilution #3 was 24% higher than the maximum loading for R9120-B. Dilution #2 was intermediate between these values. Note that these maximum loading values are the theoretical calculated values, and actual operating conditions will cause differences. A CST binder dilution (correction) factor near 0.7 (relative to pure powder CST) was required for each batch contact test with IE-911 engineered CST using the three Tank 9H dilutions. Correction factors calculated for R9120-B CST ranged from 0.56 to 0.66 for the three dilutions. Following the batch contact tests, the CST samples were isolated from the Tank 9H solution, washed, dried, and digested in acid following established procedures. The analysis results are provided in Table ES-3. Total cesium loading values determined by CST digestion were similar to the calculated loading values based on solution analysis (Table ES-2) for all samples. In addition, the CST was observed to load calcium (R9120-B sample only), iron, strontium, lead, uranium, and plutonium after contact with the waste supernate, as has been observed previously.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Literature Review: Assessment of DWPF Melter and Melter Off-gas System Lifetime

A glass melter for use in processing radioactive waste is a challenging environment for the materials of construction (MOC) resulting from a combination of high temperatures, chemical attack, and erosion/corrosion; therefore, highly engineered materials must be selected for this application. The focus of this report is to review the testing and evaluations used in the selection of the Defense Waste Processing Facility (DWPF), glass contact MOC specifically the Monofrax ® K-3 refractory and Inconel ® 690 alloy. The degradation or corrosion mechanisms of these materials during pilot scale testing and in-service operation were analyzed over a range of oxidizing and reducing flowsheets; however, DWPF has primarily processed a reducing flowsheet (i.e., Fe 2+ /ΣFe of 0.09 to 0.33) since the start of radioactive operations. This report also discusses the materials selection for the DWPF off-gas system and the corrosion evaluation of these materials during pilot scale testing and non-radioactive operations of DWPF Melter #1. Inspection of the off-gas components has not been performed during radioactive operations with the exception of maintenance because of plugging.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Literature Review: Assessment of DWPF Melter and Melter Off-gas System Lifetime

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗