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DOE OSTI2025

SaltStone Wastewater Cement Study Using Isothermal Calorimetry, Standard Concrete Characterization Techniques, and CemGEMS - 25169

Cementitious reagents are used to solidify/stabilize aqueous radioactive, hazardous, and mixed salt solutions, and sludges to meet low-level radioactive waste (LLW) and Resource Conservation and Recovery Act (RCRA) requirements for disposal at Department of Energy (DOE). It is flexible enough to solidify radioactive wastewater saturated in complex species that include but are not limited to radioactive isotopes from the bombardment of neutrons in reactor operation, corrosion products from metallic components, and a variety of soluble organic compounds. [1,2]. Waste form testing typically includes processing or fresh properties, cured properties, compressive strength and hydraulic properties, porosity, density, saturated and unsaturated moisture transport, and leachability of contaminants in the waste form pore solution. Properties are collected over a relatively limited time, typically 28 to 365 days [3]. In addition, changes in the waste form as the result of time and changing conditions are important for concrete engineers to predict overall performance of the forms and potential release of contaminants in the disposal process via unintended filtration into the environment [4]. These predictions are determined/calculated characterizing young waste forms (relative to the standard age of concrete) and are based on transport through soluble ions in pore solutions. Characterization methods include X-ray, SEM, and isothermal calorimetry among other methods used to define the composition, amorphous vs. crystalline nature of the components, and the energetic formation mechanisms for multi-phase mineral systems. [5–7] Isothermal calorimetry is a well standardized technique for cements and concretes and can be used to predict the timing and nature of the hydration reactions.[8] The technique can measure long term energetic

Bustamante, Michael E. [Savannah River National La↗

DOE OSTI2017

Immobilization in Cement of Rice Husk used as Bio-sorbent of Radionuclides Present in Radioactive Liquid Waste - 17225

The use of new techniques for the treatment of radioactive liquid wastes has been studied in the last few years by many countries. In this context, the bio-sorption has been an attractive process. In Brazil the Radioactive Waste Management Laboratory of Nuclear and Energy Research Institute (IPEN-CNEN/SP), studied some bio-masses to be used as bio-sorbents to remove radionuclides of the radioactive liquid waste stored in this institute. The rice husk is a low cost and wide available byproduct of the agriculture and its bio-sorption capacity has been proved as feasible. Thus, it is necessary to evaluate if its immobilization in a cementitious matrix will result in a waste form that meets the requirements of Brazilian regulations for final disposal of low and intermediate level waste. This work aimed at evaluating the properties of the immobilization of rice husk in a cementitious matrix. Some properties of the waste form with incorporated rice husk were evaluated, as free water, workability, setting time and mechanical strength. The results indicated that cement matrices with 5 and 10% of biomass and water/cement ratio of 0.30 and 0.35 are viable for the production of waste forms that meets the Brazilian regulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

A Case for Molecular Recognition in Nuclear Separations: Sulfate Separation from Nuclear Wastes

In this paper, we present the case for molecular-recognition approaches for sulfate removal from radioactive wastes via the use of anion-sequestering systems selective for sulfate, using either liquid liquid extraction or crystallization. Potential benefits of removing sulfate from the waste include improved vitrification of the waste, reduced waste-form volume, and higher waste-form performance, all of which lead to potential cleanup schedule acceleration and cost savings. The need for sulfate removal from radioactive waste, especially legacy tank wastes stored at the Hanford site, is reviewed in detail and primarily relates to the low solubility of sulfate in borosilicate glass. Traditional methods applicable to the separation of sulfate from radioactive wastes are also reviewed, with the finding that currently no technology has been identified and successfully demonstrated to meet this need. Fundamental research in the authors laboratories targeting sulfate as an important representative of the class of oxoanions is based on the hypothesis that designed receptors may provide the needed ability to recognize sulfate under highly competitive conditions, in particular where the nitrate anion concentration is high. Receptors that have been shown to have promising affinity for sulfate, either in extraction or in crystallization experiments, include hexaurea tripods, tetraamide macrocycles, cyclo[8]pyrroles, calixpyrroles, and self-assembled urea-lined cages. Good sulfate selectivity observed in the laboratory provides experimental support for the proposed molecular-recognition approach.

Moyer, Bruce A↗

DOE OSTI2015

Cementitious Barriers Partnership - FY2015 End-Year Report

The DOE-EM Office of Tank Waste Management Cementitious Barriers Partnership (CBP) is chartered with providing the technical basis for implementing cement-based waste forms and radioactive waste containment structures for long-term disposal. Therefore, the CBP ultimate purpose is to support progress in final treatment and disposal of legacy waste and closure of High-Level Waste (HLW) tanks in the DOE complex. This status report highlights the CBP 2015 Software and Experimental Program efforts and accomplishments that support DOE needs in environmental cleanup and waste disposal. DOE needs in this area include: Long-term performance predictions to provide credibility (i.e., a defensible technical basis) for regulator and DOE review and approvals, Facility flow sheet development/enhancements, and Conceptual designs for new disposal facilities. In 2015, the CBP developed a beta release of the CBP Software Toolbox – “Version 3.0”, which includes new STADIUM carbonation and damage models, a new SRNL module for estimating hydraulic properties and flow in fractured and intact cementitious materials, and a new LeachXS/ORCHESTRA (LXO) oxidation module. In addition, the STADIUM sulfate attack and chloride models have been improved as well as the LXO modules for sulfate attack, carbonation, constituent leaching, and percolation with radial diffusion (for leaching and transport in cracked cementitious materials). These STADIUM and LXO models are applicable to and can be used by both DOE and the Nuclear Regulatory Commission (NRC) end-users for service life prediction and long-term leaching evaluations of radioactive waste containment structures across the DOE complex.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

In-service inspection of radioactive waste tanks at the Savannah River Site - 15410

Liquid radioactive wastes from the Savannah River Site (SRS) separation process are stored in large underground carbon steel tanks. The high level wastes are processed in several of the tanks and then transferred by piping to other site facilities for further processing before they are stabilized in a vitrified or grout waste form. Based on waste removal and processing schedules, many of the tanks, will be required to be in service for times exceeding the initial intended life. Until the waste is removed from storage, transferred, and processed, the materials and structures of the tanks must maintain a confinement function by providing a barrier to the environment and by maintaining acceptable structural stability during design basis events, which include loadings from both normal service and abnormal (e.g., earthquake) conditions. A structural integrity program is in place to maintain the structural and leak integrity functions of these waste tanks throughout their intended service life. In-service inspection (ISI) is an essential element of a comprehensive structural integrity program for the waste tanks at the Savannah River Site (SRS). The ISI program was developed to determine the degree of degradation the waste tanks have experienced due to service conditions. As a result of the inspections, an assessment can be made of the effectiveness of corrosion controls for the waste chemistry, which precludes accelerated localized and general corrosion of the waste tanks. Ultrasonic inspections (UT) are performed to detect and quantify the degree of general wall thinning, pitting and cracking as a measure of tank degradation. The results from these inspections through 2013, for the 27 Type III/IIIA tanks, indicate no reportable in-service corrosion degradation in the primary tank (i.e., general, pitting, or cracking). The average wall thickness for all tanks remains above the manufactured nominal thickness minus 0.25 millimeter and the largest pit identified is approximately 1.70 millimeter deep (i.e., less than 10% through-wall). Improvements to the inspection program were recently instituted to provide additional confidence in the degradation rates. Thickness measurements from a single vertical strip along the accessible height of the primary tank have been used as a baseline to compare historical measurements. Changes in wall thickness and pit depths along this vertical strip are utilized to estimate the rate of corrosion degradation. An independent review of the ISI program methodology, results, and path forward was held in August 2009. The review recommended statistical sampling of the tanks to improve the confidence of the single strip inspection program. The statistical sampling plan required that SRS increase the amount of area scanned per tank. Therefore, in addition to the baseline vertical strip that is obtained for historical comparisons, four additional randomly selected vertical strips are inspected. To date, a total of 104 independent vertical strips along the height of the primary tank have been completed. A statistical analysis of the data indicates that at this coverage level there is a 99.5% confidence level that one of the worst 5% of all the vertical strips was inspected. That is, there is a relatively high likelihood that the SRS inspection program has covered one of the most corroded areas of any of the Type III/IIIA waste tanks. These data further support the conclusion that there are no significant indications of wall thinning or pitting. Random sampling will continue to increase the confidence that one of the worst 5% has been inspected. In order to obtain the additional vertical strips, and minimize budget and schedule impacts, data collection speed for the UT system was optimized. Prior to 2009, the system collected data at a rate of 32 square centimeters per minute. The scan rate was increased to 129 - 160 square centimeters per minute by increasing the scanner step and pixel sizes in the data acquisition set-up. Laboratory testing was utilized to optimize the scan index/pixel size such that the requirements for wall thinning and pit detection were still maintained. SRS continues to evaluate improvements to ultrasonic equipment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Waste Chemistry Envelope for the Determination of Pitting Corrosion Susceptibility - 19554

Liquid radioactive wastes at the Hanford and Savannah River sites are stored in large underground carbon steel tanks. The high-level wastes are processed in several of the tanks and then transferred by piping to other site facilities for further processing before they are stabilized in a vitrified or grout waste form. Based on waste removal and processing schedules, many of the tanks, will be required to be in service for times exceeding the initial intended life (i.e., greater than 60 years). Until the waste is removed from storage, transferred, and processed, the materials and structures of the tanks must maintain a confinement function by providing a barrier to the environment and by maintaining acceptable structural stability during design basis events, which include loadings from both normal service and abnormal (e.g., earthquake) conditions. Administrative programs are in place to mitigate potential corrosion mechanisms and thereby maintain the structural and leak integrity functions of these waste tanks throughout their intended service life. Waste chemistry and temperature control are the means by which corrosion degradation of the waste tanks is minimized. Chemistry controls are based on in-situ coupon tests, waste tank experience, and laboratory testing. These requirements comprise the technical standards for corrosion control. The standards are utilized then as guidelines for operational procedures. The present waste chemistry controls for the Hanford DSTs were established in the 1980's in response to tank failures caused by stress corrosion cracking (SCC) at the DOE Savannah River Site (SRS). Portions of the SRS chemistry limits, which were developed at the Savannah River National Laboratory (SRNL) were coupled with corrosion testing results from Pacific Northwest National Laboratories (PNNL) and adopted for corrosion control of the double-shell tank (DST) waste at the Hanford Site. Since that time, the temperatures of the waste in the DSTs have decreased to less than 50 deg. C in most cases as the radioactive material has decayed. As a consequence, new chemistry controls were formulated for SCC in 2010. This testing however, did not address that potential for pitting in the supernatant and at the liquid-air interface. Indeed, initial tests indicated that the inhibitor requirements for SCC would be insufficient for the prevention of pitting. In addition to lower waste temperatures, future retrieval and process changes may impact the DST waste chemistry. Initial projections suggest that the waste chemistry of the DSTs may shift to a broader range of pH and higher aggressive anion concentrations (e.g., chloride, sulfate, etc.) than the wastes that are presently stored. Additionally, processes within the vitrification facility may result in streams that are returned to the tank farm for evaporation. These streams may also contain high concentrations of aggressive species. SRNL investigated the propensity for pitting of the carbon steel waste tanks at current and projected DST waste chemistries. A statistically designed series of tests provided a pitting factor that relates the ratio of the inhibitor species to the aggressive species and the probability of observing pitting. An acceptable pitting factor was determined and was utilized for the development of a new chemistry envelope. The new chemistry envelope was compared to the envelope that was established for SCC and was determined to be bounding. Validation of the new chemistry envelope was performed via model simulations and comparison with historical experimental results and approaches to establishing chemistry envelopes. The new chemistry envelope will eventually be part of an overall assessment of the condition of the tanks and any changes made to the overall structural integrity program for the Hanford DSTs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Natural and Engineered Features Supporting Environmental Performance of Idaho National Laboratory's Remote-Handled Low-Level Waste Disposal Facility - 19436

The Idaho National Laboratory's (INL's) Remote-Handled Low-Level Waste (RH LLW) Disposal Facility has been designed and constructed to receive waste generated at the INL site in support of the US DOE Office of Nuclear Energy (NE) and Naval Reactors (NR) missions. The facility has been designed to receive legacy RH LLW currently in storage at INL facilities and new waste generated by nuclear research programs. RH LLW to be received at the facility is solid waste consisting of surface contaminated materials; ion exchange resins used to purify cooling and canal water; and activated metal components generated during reactor core change outs and Naval spent fuel management; and nuclear R and D. Disposal of INL's RH LLW poses unique disposal challenges due to high radiation levels (contact rates up to 600 Sv/hr [60,000 Rem/hr]) that require remote handling and adequate shielding for transportation and disposal. The facility has been designed to meet the requirements of DOE Order 435.1, Radioactive Waste Management, including the all-pathways dose, air pathway dose, inadvertent intruder doses, and radon emissions limits. This paper presents an overview of natural and engineered design features that support the safety case and enhance the environmental performance of the facility. These features include location selection, use of reinforced pre-cast concrete vaults; a robust hydraulic drainage system; use of steel waste canisters for waste disposal; the facility monitoring system; facility layout; and crediting waste forms. The facility performance assessment (PA) developed to meet the requirements of DOE Order 435.1, credits each of these protective features. Releases from waste forms disposed of at this facility are limited by corrosion of the steel waste canisters which will ultimately allow contact with infiltrating water followed by transport of radionuclides from the vault system. The steel waste canisters were credited for their ability to limit water contact with the waste based on the demonstrated ability of the reinforced pre-caste concrete vaults to provide long-term structural protection of the steel waste canisters. Verification of hydraulic and concrete performance was conducted through laboratory data, field infiltration tests supported by a robust monitoring system, and numerical modeling. Releases into the vault system from activated metals are based on site-specific corrosion data. Radionuclide transport from the vault system from all waste forms is based on site-specific climatological, hydrologic, and stratigraphic data. At INL, the groundwater pathway historically and appropriately has received the most attention. By crediting aspects of the natural system, the engineered vault system, and the waste, estimated radiologic doses via the groundwater pathway are minimal. Air pathway doses were reduced by differentially crediting corrosion of aluminum and stainless steel activated metals resulting in the pathway being screened out of the PA. Groundwater and inadvertent intruder doses were reduced through selective facility layout. This paper provides an overview of the system-wide protective features incorporated into the facility design along with an extensive reference list to supporting technical papers and analyses that provide data and interpretation. (authors)

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DOE OSTI2016

Characterization of Aged Radioactive Pollucite

The objective of this task was to provide information to better understand the long-term effects of decay on the stability of crystalline pollucite (CsAlSi 2 O 6 , also known as Cs-leucite). This decay is a key scientific issue in the assessment of the long-term stability, and hence performance, of pollucite as a waste form for 137 Cs. Most previous work on radiation effects in waste forms focused on alpha radiation, which produces more displacements per decay than beta radiation. However, the results of decay with beta radiation, which changes both the ionic radius and the valence of the element undergoing decay (0.160 nm for Ba 2+ vs. 0.188 nm for Cs + ), are likely to predominate over cascade damage. The resulting changes in coordination chemistry may destabilize the waste form, permitting release of radionuclides to the accessible environment.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2013

Determination Of Reportable Radionuclides For DWPF Sludge Batch 7B (Macrobatch 9)

The Waste Acceptance Product Specifications (WAPS) 1.2 require that “The Producer shall report the inventory of radionuclides (in Curies) that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115”. As part of the strategy to comply with WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type, all radionuclides (with half-lives greater than 10 years) that have concentrations greater than 0.01 percent of the total inventory from time of production through the 1100 year period from 2015 through 3115. The initial listing of radionuclides to be included is based on the design-basis glass as identified in the Waste Form Compliance Plan (WCP) and Waste Form Qualification Report (WQR). However, it is required that this list be expanded if other radionuclides with half-lives greater than 10 years are identified that may meet the greater than 0.01% criterion for Curie content. Specification 1.6 of the WAPS, International Atomic Energy Agency (IAEA) Safeguards Reporting for High Level Waste (HLW), requires that the ratio by weights of the following uranium and plutonium isotopes be reported: U-233, U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. Therefore, the complete set of reportable radionuclides must also include this set of U and Pu isotopes. The DWPF is receiving radioactive sludge slurry from HLW Tank 40. The radioactive sludge slurry in Tank 40 is a blend of the heel from Sludge Batch 7a (SB7a) and Sludge Batch 7b (SB7b) that was transferred to Tank 40 from Tank 51. The blend of sludge in Tank 40 is also referred to as Macrobatch 9 (MB9). This report develops the list of reportable radionuclides and associated activities as a function of time. The DWPF will use this list and the activities as one of the inputs for the development of the Production Records that relate to radionuclide inventory.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Determination of Reportable Radionuclides for DWPF Sludge Batch 7b (Macrobatch 9)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Potential benefit of amorphization in the retention of gaseous species in irradiated pyrochlores

Understanding the structure-property relationship for materials destined for irradiation extremes is a key step in developing materials with reliable, long-term performance. One crucial relationship is the ability of a material to withstand or accommodate amorphization, as this dictates its potential use as a nuclear waste form. Pyrochlores are one such class of materials for consideration as waste forms and there has been significant work examining how both the crystal structure and chemistry impacts amorphization resistance, leading to the important conclusion that the amorphization resistance of pyrochlores (A 2 B 2 O 7 ) is very sensitive to the nature of the B cations. For example, pyrochlores with B=Ti amorphize much more readily than B=Zr compounds. However, there are still questions regarding how these types of materials respond to prolonged or high-dose irradiation conditions. Gd 2 Ti 2 O 7 (GTO) and Gd 2 Zr 2 O 7 (GZO) pyrochlores were implanted with 400 keV Kr++ ions at room temperature to calculated peak damages of 119 and 135 displacements per atom (dpa), respectively. As expected, GTO amorphized completely under irradiation. However, discrete bubbles of Kr coalesced within the amorphous matrix without micro-cracking or spallation. In contrast, GZO transforms to a disordered fluorite structure under irradiation with no indications of localized amorphization. But, the accumulation of Kr within the host material leads to sub-grain structures, extended defects, and the development of micro-cracks. Thus, while GTO readily amorphizes even under low dose irradiations, the resistance of the amorphous GTO matrix to micro cracking and gas release, even in the presence of large bubble formation, suggests an enhanced propensity to retain gaseous species. Finally, consideration of long-term dose accumulation effects in nuclear waste forms would suggest reconsideration of amorphization processes in pyrochlores and related materials as a potential beneficial effect for immobilization and long-term storage of actinide materials.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Alternative Hanford ETF Grout Evaluation Report

The Hanford Effluent Treatment Facility (ETF) ammonia tolerant grout waste form was developed by the Vitreous State Laboratory (VSL) to stabilize ammonium and thereby prevent emission of ammonia vapor during processing of a final cementitious waste form. This ETF evaporator bottom waste is a concentrated sodium sulfate aqueous solution that also contains dissolved ammonium. The VSL waste form was produced in a two-step process in which struvite (MgNH 4 PO 4.6 H 2 O) is precipitated and then the resulting slurry is solidified in a cementitious matrix. SRNL researchers were tasked to evaluate the effect of compositional and struvite stability in simulated ETF evaporator brine. In 2022 to 2023, samples prepared by SRNL experienced significant expansion (up to 20 volume %) which was not observed by VSL.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Deployment of Performance Management Methodology as part of Liquid Waste Program at Savannah River Site - 12178

In 2009, Savannah River Remediation LLC (SRR) assumed the management lead of the Liquid Waste (LW) Program at the Savannah River Site (SRS). The four SRR partners and AREVA, as an integrated subcontractor are performing the ongoing effort to safely and reliably: - Close High Level Waste (HLW) storage tanks; - Maximize waste throughput at the Defense Waste Processing Facility (DWPF); - Process salt waste into stable final waste form; - Manage the HLW liquid waste material stored at SRS. As part of these initiatives, SRR and AREVA deployed a performance management methodology based on Overall Equipment Effectiveness (OEE) at the DWPF in order to support the required production increase. This project took advantage of lessons learned by AREVA through the deployment of Total Productive Maintenance and Visual Management methodologies at the La Hague reprocessing facility in France. The project also took advantage of measurement data collected from different steps of the DWPF process by the SRR team (Melter Engineering, Chemical Process Engineering, Laboratory Operations, Plant Operations). Today the SRR team has a standard method for measuring processing time throughout the facility, a reliable source of objective data for use in decision-making at all levels, and a better balance between engineering department goals and operational goals. Preliminary results show that the deployment of this performance management methodology to the LW program at SRS has already significantly contributed to the DWPF throughput increases and is being deployed in the Saltstone facility. As part of the liquid waste program on Savannah River Site, SRR committed to enhance production throughput of DWPF. Beyond technical modifications implemented at different location of the facility, SRR deployed performance management methodology based on OEE metrics. The implementation benefited from the experience gained by AREVA in its own facilities in France. OEE proved to be a valuable tool in order to support the enhancement program in DWPF by providing unified metrics to measure plant performances, identify bottleneck location, and rank the most time consuming causes from objective data shared between the different groups belonging to the organization. Beyond OEE, the Visual Management tool adapted from the one used at La Hague were also provided in order to further enhance communication within the operating teams. As a result of all the initiatives implemented on DWPF, achieved production has been increased to record rates from FY10 to FY11. It is expected that thanks to the performance management tools now available within DWPF, these results will be sustained and even improved in the future to meet system plan targets. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

WastePD, an innovative center on materials degradation

The US Department of Energy recently awarded funds to create the Center for Performance and Design of Nuclear Waste Forms and Containers (WastePD) as part of the Energy Frontier Research Center (EFRC) program. EFRCs are multi-investigator collaborations of universities, national labs and companies that “conduct fundamental research focusing on one or more “grand challenges” and use-inspired “basic research needs” identified in major strategic planning efforts by the scientific community.” The major performance parameter of nuclear waste forms is their ability to isolate the radionuclides by withstanding degradation in a repository environment over very long periods of time. So WastePD is at heart a center focused on materials degradation.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Technetium incorporation into goethite (α-FeOOH): An atomic-scale investigation

During the processing of low-activity radioactive waste to generate solid waste forms (e.g., glass), technetium-99 (Tc) is of concern because of its volatility. A variety of materials are under consideration to capture Tc from waste streams, including the iron oxyhydroxide, goethite (α-FeOOH), which was experimentally shown to sequester Tc(IV). This material could ultimately be incorporated into glass or other low-temperature waste form matrices. However, questions remain regarding the incorporation mechanism for Tc(IV) in goethite, which has implications for predicting the long-term stability of Tc in waste forms under changing conditions. Here, quantum-mechanical calculations were used to evaluate the energy of five different charge-compensated Tc(IV) incorporation scenarios in goethite. The two most stable incorporation mechanisms involve direct substitution of Tc(IV) onto Fe(III) lattice sites and charge balancing either by removing one nearby H+ (i.e., within 5 Å), or by creating an Fe(III) vacancy when substituting 3 Tc(IV) for 4 Fe(III), with the former being preferred over the latter relative to gas-phase ions. When corrections for hydrated references phases are applied, the Fe(III)-vacancy mechanism becomes more energetically competitive. Calculated incorporation energies and optimized bond-lengths are presented. Proton movement is observed to satisfy under-coordinated bonds surrounding vacancies in the goethite structure.

atomic-scale modeling↗

DOE OSTI2021

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

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

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Natural and Engineered Features Supporting Environmental Performance of Idaho National Laboratory’s Remote-Handled Low-Level Waste Disposal Facility

The Idaho National Laboratory’s (INL’s) Remote-Handled Low-Level Waste (RH LLW) Disposal Facility has been designed and constructed to receive waste generated at the INL site in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) and Naval Reactors (NR) missions. The facility has been designed to receive legacy RH LLW currently in storage at INL facilities and new waste generated by nuclear research programs. RH LLW to be received at the facility is solid waste consisting of surface contaminated debris; ion exchange resins used to purify cooling and canal water; activated metals generated during reactor core change outs and Naval spent fuel management; and nuclear Research & Development. Disposal of INL's RH LLW poses unique disposal challenges due to high radiation levels (contact rates up to 600 Sv/hr [60,000 Rem/hr]) that require remote handling and adequate shielding for transportation and disposal. The facility has been designed to meet the requirements of DOE Order 435.1 [1], including the all-pathways dose, air pathway dose, inadvertent intruder doses, and radon emissions limits. This paper presents an overview of natural and engineered design features that support the safety case and enhance the environmental performance of the facility. These include location selection, use of reinforced pre-cast concrete vaults; a robust hydraulic drainage system; use of steel waste canisters for waste disposal; the facility monitoring system; facility layout; and crediting waste forms. The facility performance assessment (PA) [2] developed to meet the requirements of DOE Order 435.1 “Radioactive Waste Management” credits each of these protective features. Releases from waste forms disposed of at this facility are limited by corrosion of the steel waste canisters which will ultimately allow contact with infiltrating water followed by transport of radionuclides from the vault system. The steel waste canisters were credited for their ability to limit water contact with the waste based on the demonstrated ability of the reinforced pre-caste concrete vaults to provide long-term structural protection of the steel waste canisters. Verification of hydraulic and concrete performance was conducted through laboratory data, field infiltration tests supported by a robust monitoring system, and numerical modeling. Releases into the vault system from activated metals are based on site-specific corrosion data. Radionuclide transport from the vault system from all waste forms is based on site-specific climatological, hydrologic, and stratigraphic data. At INL, the groundwater pathway historically and appropriately has received the most attention. By crediting aspects of the natural system, the engineered vault system, and the waste, estimated radiologic doses via the groundwater pathway are minimal. Air pathway doses were reduced by differentially crediting corrosion of aluminum and stainless steel activated metals resulting in the pathway being screened out of the PA. Groundwater and inadvertent intruder doses were reduced through selective facility layout. This paper provides an overview of the system-wide protective features incorporated into the facility design along with an extensive reference list to supporting technical papers and analyses that provide data and interpretation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Saltstone Second Quarter Calendar Year 2018 (2qcy18) Toxicity Characteristic Leaching Procedure (Tclp) Results

The aqueous waste from Tank 50 (salt solution) is sampled quarterly for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). A SDF waste form (saltstone) was prepared in the Savannah River National Laboratory (SRNL) from the Tank 50 Waste Acceptance Criteria (WAC) sample and Z-Area premix material for the second quarter of calendar year 2018 (2QCY18). Results from this memorandum support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Remediation (SRR). After a 28 day cure, a sample of the SDF waste form was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The 2QCY18 saltstone sample met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79-268.48 requirements for a non-hazardous waste form with respect to Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and also met the SPF WAC.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗