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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

DFLAW Glass and Feed Qualifications for AP-107 to Support WTP Start-Up and Flow-Sheet Development (VSL-18R4500-1, Rev 0)

About 50 million gallons of high-level mixed waste is currently stored in underground tanks at The United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed in an engineered facility on the Hanford site while the IHLW product is designed for acceptance into the national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal. The Low Activity Waste (LAW) at Hanford will be vitrified in a Joule Heated Ceramic Melter (JHCM) to produce a stable product for safe disposal. Although present in relatively small amounts, a few key constituents in the LAW streams play particularly important roles. Sulfur, chlorine, and fluorine can have significant impacts on LAW glass formulation and waste loadings, while technetium and iodine are important for waste form performance, with technetium being the constituent of most concern with respect to waste disposal in the Integrated Disposal Facility (IDF). The IDF is designed to accept ILAW, secondary waste associated with ILAW production, and other on-site Hanford Non-Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) waste. As a result of volatilization of these species during the vitrification process, they are also present in significant amounts in the liquid effluents from the off-gas treatment system. Understanding the fraction of each constituent incorporated in the LAW glass or in WTP liquid effluents is critical for assessing the performance of waste forms to be disposed in the IDF. Liquid effluents from the LAW vitrification process that are not recycled will likely be solidified to produce a secondary waste form that is not as chemically durable as glass. Therefore, understanding partitioning of the key constituents between the glass and non-glass waste forms is a significant factor in the Performance Assessment (PA) for the Hanford IDF. Clearly, it is preferable to incorporate as much of the technetium into the glass as possible because of the superior chemical durability of that waste form. The objective of the present work is to improve the understanding of the partitioning of technetium and other key waste constituents between the glass and non-glass waste forms.

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

Performance Criteria for Capture and/or Immobilization Technologies (Revision 1)

The capture and subsequent immobilization of regulated volatile radionuclides from the off-gas streams of a used nuclear fuel (UNF) reprocessing facility has been a topic of substantial research interest for the US Department of Energy and its international counterparts. Removal of specific radionuclides from the plant effluent streams before discharge to the environment is required to meet regulations set forth by the US Environmental Protection Agency. Upon removal, the radionuclides, as well as associated sorbents that cannot be regenerated in a cost-effective manner, are destined for conversion to a waste form. Research in separation and capture methodologies has included a wide range of technology types, and studies of waste forms are correspondingly diverse. In considering the future development and implementation of both sorbents and waste forms, it is necessary to identify benchmark measures of performance to objectively evaluate each sorbent system or waste form. Sets of performance criteria and associated metrics have been developed for sorbent and waste form evaluation. These criteria address physical, radiological, and chemical characteristics, technical practicality, technical maturity, cost, and, for sorbents, system performance. The criteria and metrics appear to be robust and should be applicable despite the eventual waste classification (as either high- or low-level waste). They are flexible enough to address both aqueous reprocessing and electrochemical reprocessing of UNF. These criteria sets can serve as tools to evaluate performance at multiple stages within the development process, and in this revision (Revision 1) they have been used to assess technologies relating to krypton/xenon separations and iodine capture from off-gas streams arising from UNF reprocessing. Assessment of krypton/xenon separations using engineered forms of two zeolite minerals (silver mordenite and hydrogen mordenite in a polyacrylonitrile-based binder [AgZ-PAN/HZ-PAN]) found that the zeolite-based separation is relatively advanced in its development, but several key issues require resolution. First, desorption processes for both krypton and xenon require refinement to provide an understanding of the product purity that can be achieved. Second, adsorption rate data is needed in order to calculate the bed depth required for effective separation. Finally, it is strongly recommended that a technical review of krypton/xenon separation by AgZ-PAN/HZ-PAN be performed to synergize available data and assess the cost savings and operational benefits that may be realized from implementation of this technology. Assessment of metal organic frameworks (MOFs) for their use in the separation of krypton/xenon found that the ideal separation would be performed using a single-column system with a MOF selective for krypton over xenon. A robust research effort should work to identify a krypton-selective MOF designed to operate at temperatures of approximately 0°C or higher, which could be preferred over cryogenic krypton/xenon capture for used fuel reprocessing off-gas streams. In the case of the CaSDB-MOF (the most well-understood xenon sorbent to date), two issues are judged of high importance. First, xenon breakthrough capacity for the CaSDB-MOF in prototypical conditions should be determined. Preliminary research indicates that breakthrough may be near immediate, presenting a substantial obstacle in separative system design. Second, development of desorption methodology should be performed to determine regeneration time, energy requirements, and the product stream composition. Silver-based sorbents (AgZ and AgAero) for use in iodine capture from the dissolver off-gas were evaluated against the established criteria. These sorbents are significantly better understood for this application as a result of research efforts over the past decade. The potential implementation of AgAero at a large scale is hindered by its physical degradation by components of the dissolver off-gas stream. Less is known about the adsorption of iodine by these sorbents from other off-gas streams in the plant. Initial experimental efforts have been closely coordinated in an effort to understand organic iodine (such as would be found in the vessel off-gas) adsorption by AgZ and AgAero. Future work should expand this experimental program, and analysis of other reprocessing facility off-gas streams such as the vitrification off-gas stream should be conducted to better understand other potential applications for iodine sorbents. A review of iodine waste form development shows that this area is diverse and that multiple promising waste forms have been identified for the immobilization of radioactive iodine. Efforts related to the direct conversion of iodine sorbents (including AgZ and AgAero) should be continued because of the advantages of direct conversion in a waste management strategy and other sorbents should continue to be advanced as merited.

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

Nuclear Criticality Safety Assessment of Criticality Control Containers without Moderation Control at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) provides for safe, permanent disposal of government-owned transuranic (TRU) and TRU mixed wastes. Receipt and disposal of waste at the WIPP site began in March 1999. The Sandia report, Consideration of Nuclear Criticality When Disposing of Transuranic Waste at the Waste Isolation Pilot Plant, addressed potential nuclear criticality safety issues based on the projected inventory characteristics known at the time [1]. As designs for inventory, waste forms, and disposal packages have changed, new analyses have been performed, and updates have been made to address any potential effects to the WIPP safety basis. New analyses performed include Saylor 2017 [2] and Brickner 2019 [3], which address certain waste containers with specified loadings under post-closure conditions. Both examined several hypothetical scenarios and included analyses to bound (from a criticality potential standpoint) credible configurations that could occur at WIPP during the repository regulatory post-closure disposal time period for feature, event, and process (FEP) considerations—10,000 years. During this post-closure period at WIPP, the screening of FEPs is governed by the risk-based standards and implementing regulations of the US Environmental Protection Agency (EPA) (i.e., 40 CFR 191 and 40 CFR 194, respectively) [4,5]. An FEP screening can be based on either a low-consequence or low-probability rationale. A low-probability rationale includes either (a) a qualitative rationale that the FEP is not credible or (b) a quantitative demonstration that the probability is less than 10-4 in 104 years. In this evaluation, a qualitative lowprobability rationale of not credible is used by demonstrating that bounding configurations of the waste are not critical. The demonstration of subcriticality is through quantitative calculations, but a probability of criticality is not evaluated. Rather, the rationale for this evaluation is that bounding configurations with an effective neutron multiplication factor (keff) well below the upper subcriticality limit (USL) make criticality incredible. Reference [2] documented a nuclear criticality assessment of the WIPP repository for disposal of dilute surplus plutonium materials using the Dilute and Dispose Approach and packaging in criticality control overpacks (CCOs). The CCO is the waste disposal container recently designed to allow for up to 380 fissile gram equivalent (FGE) 239 Pu per drum, which is a higher fissile loading than typical waste containers. The CCO consists of a criticality control container (CCC) positioned by upper and lower plywood spacers within a standard 55 gal drum. The CCC is used to establish a geometry control for fissile materials during transportation and WIPP emplacement operations. The current WIPP waste acceptance criteria for CCO payloads limit beryllium to less than or equal to 1% by weight of the waste contents and require the waste form to be non-machine compacted. Reference [2] considered two scenario progressions—room closure from salt creep, hereafter referred to as the reconfigured dry scenario, and flooding with brine, hereafter referred to as the reconfigured wet scenario. The subsequent drying out of the reconfigured wet scenarios was also considered. For all scenarios, subcriticality was maintained when 50 g of B 4 C (acting as a neutron absorber) per CCC was intermixed within the plutonium disposition waste form. The analysis used a waste form description that limits the amount of moderation that could be present within the waste form (i.e., it limits the amount of water and polyethylene that could be present based on planned processing conditions). This analysis to evaluate increased limits on the amount of moderation that could be present was performed as a companion to Reference [2] to address concerns associated with verifying moisture and/or plastic contents of waste materials following packaging of dilute surplus plutonium in the CCO. To that end, this analysis used the models and methods from Reference [2] to evaluate a more generic base waste form consisting of water and polyethylene that is more similar (and nearly identical) to the generic waste forms utilized in other models/analyses supporting the TRU Package Transporter Model II (TRUPACTII) safety analysis [6] (all are without moderation controls). The waste form in this analysis uses a base mixture of 75% water and 25% polyethylene, the total amount of which is varied to determine the optimum moderation to fissile material (H/Pu) ratio. The fissile loading is maintained at up to 380 FGE 239 Pu (modeled as PuO 2 ) per CCO with an additional 545 g of beryllium (to bound the 1% by weight contents restriction) and 50 g of B 4 C intermixed per CCO. The beryllium content (1% by weight) is based on the total allowed waste weight (this does not include packaging and container weights). Figures ES-1 and ES-2 display summary results, showing that with this model including 50 g of B 4 C per CCO, the system keff remains under 0.85 for all moderator amounts and provides a significant margin against post-closure criticality under postulated bounding conditions for compaction. Figure ES-1 compares an infinite model with a room model at the initial emplacement spacing and under full radial compaction. Full radial compaction places each CCC in direct contact and does not credit any anticipated spacing associated with current post-closure geomechanical modeling of the repository [7]. The effects of variations in the H/Pu ratio were evaluated by varying the amount of the water/polyethylene component of the waste model, with fissile loading maintained at 380 239 Pu FGE. Similarly, Figure ES-2 illustrates how various amounts of B 4 C per CCO influence k eff at different radial compactions, all at the H/Pu ratio of 200 (in the room array model). Therefore, while the results from Saylor 2017 [2] modeled more realistic process limits associated with packaging of dilute surplus plutonium, this analysis demonstrates that limits on moderation (plastic and water content) are not necessary to ensure subcriticality in the WIPP repository, provided the requisite B 4 C absorber is present.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Optimization of sintered AgI-Mordenite composites for {sup 129}I storage

The thermal processing of a proposed durable waste form for {sup 129}I was investigated. The waste form is a composite with a matrix of low-temperature sintering glass that encapsulates particles of AgI-mordenite. Ag-mordenite, an ion-exchanged zeolite, is being considered as a capture medium for gaseous {sup 129}I{sub 2} as part of a spent nuclear fuel reprocessing scheme under development by the US Department of Energy/Nuclear Energy (NE). The thermal processing of the waste form is necessary to increase the density of the glass matrix by viscous sintering so that the final waste form does not have any open porosity. Other processes that can also occur during the thermal treatment include desorption of chemisorbed I{sub 2}, volatilization of AgI and crystallization of the glass matrix. We have optimized the thermal processing to achieve the desired high density with higher AgI-mordenite loading levels and with minimal loss of iodine. Using these conditions, 625 C. degrees for 20 minutes, the matrix crystallizes to form a eulytite phase. Results of durability tests indicate that the matrix crystallization does not significantly decrease the durability in aqueous environments. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Secondary Waste Simulant Development for Cast Stone Formulation Testing

Washington River Protection Solutions, LLC (WRPS) funded Pacific Northwest National Laboratory (PNNL) to conduct a waste form testing program to implement aspects of the Secondary Liquid Waste Treatment Cast Stone Technology Development Plan (Ashley 2012) and the Hanford Site Secondary Waste Roadmap (PNNL 2009) related to the development and qualification of Cast Stone as a potential waste form for the solidification of aqueous wastes from the Hanford Site after the aqueous wastes are treated at the Effluent Treatment Facility (ETF). The current baseline is that the resultant Cast Stone (or grout) solid waste forms would be disposed at the Integrated Disposal Facility (IDF). Data and results of this testing program will be used in the upcoming performance assessment of the IDF and in the design and operation of a solidification treatment unit planned to be added to the ETF. The purpose of the work described in this report is to 1) develop simulants for the waste streams that are currently being fed and future WTP secondary waste streams also to be fed into the ETF and 2) prepare simulants to use for preparation of grout or Cast Stone solid waste forms for testing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Evaluation of the Waste Streams from Management of Reprocessing Off-Gas Streams - 18483

Nuclear fuel reprocessing is being considered in the United States as part of an effort to help meet US energy needs and as a means to improve waste management. During the chemical processing of UNF to recover the valuable components for reuse, a number of volatile radionuclides will be released into the off-gas streams of the reprocessing plant. These volatile radionuclides include H-3, C-14, Kr-85, and I-129. Off-gas abatement systems for the volatile radionuclides will be required to ensure that air emissions from a UNF reprocessing facility can meet or exceed emission requirements set forth by the US Environmental Protection Agency and the US Nuclear Regulatory Commission. An initial engineering evaluation of offgas abatement systems was completed recently. The evaluation addressed the designs of the head-end offgas treatment system, which typically includes both the shear off-gas and the dissolver off-gas and vessel off-gas treatment system. Based on this study, the waste volumes for each of the four primary volatile radionuclides (H-3, C-14, Kr-85, and I-129) were estimated. Included in these estimates were the nonradioactive isotopes and the 'tramp' elements found in air (e.g., Kr and C-12 as CO{sub 2}) and process chemicals (e.g., I and Cl). Because the final waste form for these radionuclides has not been selected, estimated volumes were calculated for various alternative waste forms as appropriate. The resulting waste volume for I was found to be highly dependent on the final waste form selected and increased by a factor of 2 to 4 by tramp halogens that would be released into the off-gas stream from the process chemicals. The volume of tritiated water from the fuel was quite small, but the estimated volume of the final waste form depended on the tritium management approach employed within the reprocessing plant. The Kr waste volume was calculated assuming pure Kr (i.e., assuming complete separation of Xe and Ar). For the case of no Xe separation, the estimated waste volume increased by factor of ∼10. Argon from the air, if not removed, increased the estimated volume by a factor of 425. The estimated volume of C-14 waste was dominated by the CO{sub 2} from the air leaking into the processing cell. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Organic Evaporation, Oxidation, and Hydrolysis Testing in Support of Hanford Sample-and-Send

The Hanford site has approximately 56 million gallons of radioactive mixed waste stored in 156 unretrieved 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) through the WTP is vitrification, but additional immobilization capacity is needed to supplement the initial LAW melters. An alternative cementitious waste form is being investigated for that future immobilization method. However, one impediment to a cementitious waste form is the presence of Land Disposal Restricted (LDR) organic chemicals in tank waste, which are regulated on a concentration-based standard in the final waste form. Hence, if the quantity of organics in LAW is high enough, they must be destroyed or removed to make a waste form compatible with disposal in a mixed low level waste landfill. This work evaluates potential avenues for treatment of LDR organics to eliminate the impediment and permit possible use of a cementitious waste form.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Cementation of Nuclear Wastes: From Past to Future in a Sustainable Approach - 18266

Cementation of nuclear wastes is a widespread technology due to several factors such as excellent feedback from past experience, simplicity of implementation and versatility. Indeed, as nuclear wastes come in many physical states (bulk wastes, powders, sludges), cementation is the immobilisation process that can easily accommodate this large range of wastes provided the appropriate mixing technology is used: heterogenous cementation (grouting of bulky wastes) or homogenous cementation (intermixing process between the powder or the sludge and the cementitious material). However, nuclear wastes equally present an important chemical variability. This is especially true as New Areva's range of nuclear activities covers the entire fuel cycle: mining, conversion, enrichment and recycling and dismantling and decommissioning. The chemical interactions between the waste and the cement matrix are carefully studied by New Areva's R and D teams in order to provide comprehensive solutions by developing cement formulations which are adapted to wastes characteristics. The aim of the studies is to obtain cement waste-forms that meet the waste acceptance criteria, the final product being thus in perfect adequacy with the disposal requirements. For over 30 years New Areva has been developing cement formulations adapted to the physical and chemical specificities of the wastes accounting for constrains imposed by waste characteristics. The overall approach includes the definition of the appropriate cement composition at laboratory scale, engineering design, equipment manufacturing, testing and commissioning of the cementation facilities. With this strong technical background, a specific methodology has been developed by New Areva teams in order to allow a rapid screening of possible solutions for a given waste (in terms of cement chemistry adapted to the waste) in a learn-fast approach that is based on theoretical analysis and small-scale tests. The methodology was applied for wastes presenting a complex nature (such as reactive metals, ion exchange resins, oils and solvents, mercury, etc), known to interact with Portland cement from a chemical point of view and leading to durability issues of the wasteform (swelling, strength reduction, leaching) and thus reducing the final waste loading. Once the screening process is over, the main drivers for choosing the most appropriate solution to be further developed at pilot scale are: reducing the overall volume and the impact of the final waste package, ensure the waste package performance as defined by the disposal requirements and control the long-term behaviour in final disposal. Moreover, a modelling approach is addressed, aiming to develop a thermo-chemo-mechanical (THM) concrete models that are able to predict the behaviour of the structure (waste form) in time. The aim of the present paper is to describe the main principles within the developed methodology and illustrate it by using a few specific examples. Several distinct cement formulations are developed for complex wastes, limiting the chemical interaction between the cement matrix and the given wastes, obtaining better waste loadings and better overall durability behaviour which facilitates long term behaviour demonstration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Organic Evaporation, Oxidation, and Hydrolysis Testing in Support of Hanford Sample-and-Send

The Hanford site has approximately 54 to 56 million gallons of radioactive mixed waste stored in 156 unretrieved 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) through the WTP is vitrification, but additional immobilization capacity is needed to supplement the initial LAW melters. An alternative cementitious waste form is being investigated for that future immobilization method. However, one impediment to a cementitious waste form is the presence of Land Disposal Restricted (LDR) organic chemicals in tank waste, which are regulated on a concentration based standard in the final waste form. Hence, if the quantity of organics in LAW is high enough, they must be destroyed or removed to make a waste form compatible with disposal in a mixed low level waste landfill. This work evaluates potential avenues for treatment of LDR organics to eliminate the impediment and permit possible use of a cementitious waste form. Vacuum evaporation testing to remove LDR organics consisted of preparing a non-radioactive LAW simulant, spiking that simulant with organic chemicals, and evaporating the mixture via differential distillation. The apparatus was a laboratory-scale vacuum evaporator operated at 60 ±5 torr absolute (vacuum evaporation). The LAW simulant represented the liquid expected to be retrieved from the Hanford tank farms at approximately 4.0 M [Na+] total sodium ion concentration. The concentration of the organic chemicals added was significantly higher than typically found in the tank waste samples since the higher levels were necessary to assist in analytical measurement and tracking of the spiked species.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

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

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

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Salt-stone Oxidation Study: Leaching Method - 13092

Cementitious waste forms can be designed to chemically stabilize selected contaminants, such as Tc{sup +7} and Cr{sup +6}, by chemically reduction to lower valance states, Tc{sup +4} and Cr{sup +3}, respectively, and precipitation of these species in alkaline media as low solubility solid phases. Data for oxidation of this type of cementitious waste form cured under field conditions as a function of time is required for predicting the performance of the waste form and disposal facility. The rate of oxidation (oxidation front advancement) is an important parameter for predicting performance because the solubilities of some radionuclide contaminants, e.g., technetium, are a function of the oxidation state. A non-radioactive experiment was designed for quantifying the oxidation front advancement using chromium, as an approximate redox-sensitive surrogate (Cr{sup +6} / Cr{sup +3}) for technetium (Tc{sup +7} / Tc{sup +4}). Nonradioactive cementitious waste forms were prepared in the laboratory and cured under both laboratory and 'field conditions'. Laboratory conditions were ambient temperature and sealed sample containers. Field conditions were approximated by curing samples in open containers which were placed inside a plastic container stored outdoors at SRS. The container had a lid and was instrumented with temperature and humidity probes. Sub-samples as thin as 0.2 mm were taken as a function of distance from the exposed surface of the as-cast sample. The sub-samples were leached and the leachates were analyzed for chromium, nitrate, nitrite and sodium. Nitrate, nitrite, and sodium concentrations were used to provide baseline data because these species are not chemically retained in the waste form matrix to any significant extent and are not redox sensitive. 'Effective' oxidation fronts for Cr were measured for samples containing 1000, 500 and 20 mg/kg Cr added as soluble sodium chromate, Na{sub 2}CrO{sub 4}. For a sample cured for 129 days under field conditions, leachable Cr (assumed to be the oxidized form, i.e., Cr{sup +6} was detected in the top 15 to 20 mm of the sample spiked with 1000 mg/kg Cr. Below about 20 mm, the Cr concentrations in leachates were below the detection limit (< 0.010 mg/L) which indicates that the oxidation as the result of exposure to air was limited to the top 20 mm of the sample after exposure for 129 days and that the bulk of the waste form was not affected, i.e., the Cr was stabilized and insoluble. For samples cured in the laboratory, leachable Cr was detected in the top 8 mm of the Cr 1000 sample cured in the laboratory for 37 days. Between 8 and 14 mm, the concentration Cr in the leachate dropped by a factor of about 20 to just above the detection limit. These depth of penetration results indicate that the rate of advancement of the oxidation front for samples spiked with 1000 mg/kg Cr cured under 'field conditions' for 129 days is less than that for the sample cured in the laboratory for 37 days, i.e., 0.156 and 0.216 mm/day, respectively. Additional data are presented for samples spiked with 500 and 20 mg/kg Cr. In summary, cementitious waste forms are porous solids with a network of interconnected pores ranging in diameter from 10 E-10 m to greater than a few mm. The oxidation process is assumed to occur as the result of oxygen transport through the interconnected porosity which may be filled with air, aqueous salt solution, or both. Upon oxidation, the Cr becomes soluble and can be transported in solution through aqueous pore fluid or leachate. (authors)

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

Modeling Interfaces to Support Low-Level Waste Disposal System Performance Assessments - 20366

In low-level waste (LLW) disposal sites, interfaces between cementitious materials, used as waste forms and/or engineered barriers, and the surrounding soil or backfill material are often encountered. Reactions across these interfaces may lead to chemical and structural alteration of the cementitious and backfill materials that may ultimately affect long-term performance. For example, the ingress of carbon dioxide from soil gas into a waste tank concrete shell at the Hanford Site or Savannah River Site (SRS) or a low-level waste disposal vault at SRS leads to carbonation of cement hydration products (e.g. portlandite, calcium-silicate hydrate or C-S-H, and ettringite). The result of carbonation is a decrease in pH in the cement paste portion of the tank wall or vault concrete structure that could ultimately lead to de-passivation of the embedded structural steel and, eventually, to cracking. Cracking can lead to increased ingress of water into the structure, corresponding increased release of constituents of concern, and to subsequent increased transport of these constituents to the surrounding environment. The mobility of trace constituents (e.g., radionuclides of concern) in waste forms may increase in response to changes in pH, pore structure, and mineralogical gradients. Depletion of portlandite results in subsequent decalcification of C-S-H that leads to changes in the cement strength and may lead to structural failure. At the boundary between waste forms and concrete vaults, the migration of sulfate ions from the salt waste form (e.g., the saltstone waste form used for LLW disposal at the Savannah River Site in Saltstone Disposal Units, or SDUs) into the barrier or vault concrete has been predicted to result in expansive mineral phase formation (or 'sulfate attack'). Expansive mineral formation could result in cracking and potential loss of structural integrity in the vault concrete. In the performance assessment (PA) for the SRS SDUs, prediction of both the carbonation and sulfate attack fronts are critical to assessing the long-term performance of these LLW disposal vaults. A general purpose geochemical reactive transport model has been developed in LeachXS/ORCHESTRA to evaluate typical interfaces for LLW disposal environments, including SDUs at SRS. The model may be used to predict the long-term performance of interfaces between cementitious materials and saltstone or backfill materials with respect to primary phases by considering mass transport, cement chemistry, geochemical speciation, and multi-ionic diffusion over interfaces between different materials. The present study compares the performance and interactions between vault concrete and saltstone waste form for a representative SDU scenario. The materials were characterized using pH-dependence leaching test (EPA 1313) and semi-dynamic transport tests (EPA 1315) with test results used to parameterize the interface model. The development of a credible inter model provides an important basis for supporting future PAs where carbonation and sulfate attack are important aging and degradation mechanisms. (authors)

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

Solid Waste Release Calculations for the Composite Analysis Baseline Assessment

The purpose of the Hanford Site Composite Analysis (CA) is to provide an estimate of the cumulative radiological impacts from active and planned low-level radioactive waste disposals and other potentially interacting radioactive waste disposal sources that will remain following Hanford Site closure. The need for a waste form release model addressing multiple solid waste forms is documented in CP-60410, Hanford Site Composite Analysis Technical Approach Description: Waste Form Release.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Organic Acid-Assisted Thermal Dehalogenation of Halide Salt Nuclear Wastes: From Waste Salts to Borosilicate Glass

Only a handful of high-halide salt waste forms have been demonstrated for vitrification-based immobilization strategies for halide-salt nuclear waste streams (e.g., pyroprocessing wastes, molten salt reactor wastes) and they all have low waste loading potential and most have low chemical durabilities for high-alkali streams. An alternative approach to direct salt immobilization is salt partitioning prior to waste form fabrication and one option for partitioning is halide removal (called dehalogenation). Removing the halogen fraction through dehalogenation can significantly reduce the waste volume required for disposal in the primary waste form. Furthermore, when dehalogenation is performed using organic acids, the dehalogenation reagent can decompose during high-temperature vitrification, reducing waste loading limitations in the waste form. In the current work, different organic acids (i.e., oxalic, formic, acetic, oxamic, and citric) were evaluated for dehalogenation efficiency of a simple chloride salt simulant (7.19% LaCl 3 , 53.77% LiCl, and 39.04% KCl, by mole) and a more complex chloride salt simulant called ERV3 (electrorefiner version 3) at 150 °C–300 °C and using H + /Cl – molar ratios of 1:1, 2:1, and 3:1. Additionally, a borosilicate glass waste form called TARS (or the average of refined specifications) was formulated, produced, and characterized for dehalogenated ERV3.

Amorphous materials↗
DOE OSTI2024

Nitrate and Nitrite at Hanford – From Tanks to Natural Attenuation

Washington River Protection Solutions, LLC (WRPS) contracted Pacific Northwest National Laboratory (PNNL) in support of their maturation testing and analysis to develop and deploy an ambient temperature solidification/stabilization process for low-activity waste (LAW) (referred to as “grout” in this report). The intent of this effort is to support a sample-and-send regulatory and processing strategy in the Hanford 200 West Area, specifically considering grout waste forms disposal at Hanford in the Integrated Disposal Facility (IDF). Currently the disposal of grouted LAW is not an option because the IDF is not permitted to receive this waste form. The material presented in this report is intended to help inform the U.S. Department of Energy (DOE), the site operating contractors, regulatory agencies, and stakeholders of the future implications of IDF disposal of a grouted LAW waste form, specifically for the behavior of nitrate (NO 3 - ) and nitrite (NO 2 - ) at the Hanford. The only calculation of nitrate release from a hypothetical grouted LAW inventory in the IDF showed that the overall release was slightly above the compliance limit of 45 mg/L nitrate and 3.32 mg/L nitrate after 1000 years (Asmussen et al. 2019). As such, uncertainty around the impact of nitrate and nitrite release from the IDF was identified in the recent National Academies of Sciences study of Hanford supplemental LAW by both the national laboratory team and Hanford stakeholders (Bates et al. 2023). However, these calculations (and others for the IDF involving secondary waste grout) may be conservative as no nitrate and nitrite retention in the grout waste form was assumed, nor any attenuation in the subsurface. There is evidence from the literature of processes that can attenuate nitrate migration both in the waste form and in the Hanford subsurface. If found to be technically defensible and likely to occur in the IDF and associated subsurface, then these processes should be included in modeling of the IDF to represent the behavior of nitrate and nitrite more accurately. Doing so would remove unnecessary conservatism in the modeling projections and reduce the uncertainty in assessments of facility compliance of a grouted LAW waste form in the IDF.

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DOE OSTI2013

Saltstone Oxidation Study: Leaching Method

Cementitious waste forms can be designed to chemically stabilize selected contaminants, such as Tc{sup +7} and Cr{sup +6}, by chemically reduction to lower valance states, Tc{sup +4} and Cr{sup +3}, respectively, and precipitation of these species in alkaline media as low solubility solid phases. Data for oxidation of this type of cementitious waste form cured under field conditions as a function of time is required for predicting the performance of the waste form and disposal facility. The rate of oxidation (oxidation front advancement) is an important parameter for predicting performance because the solubilities of some radionuclide contaminants, e.g., technetium, are a function of the oxidation state. A non-radioactive experiment was designed for quantifying the oxidation front advancement using chromium, as an approximate redox-sensitive surrogate (Cr{sup +6} / Cr{sup +3}) for technetium (Tc{sup +7} / Tc{sup +4}). Nonradioactive cementitious waste forms were prepared in the laboratory and cured under both laboratory and ?field conditions.? Laboratory conditions were ambient temperature and sealed sample containers. Field conditions were approximated by curing samples in open containers which were placed inside a plastic container stored outdoors at SRS. The container had a lid and was instrumented with temperature and humidity probes. Subsamples as thin as 0.2 mm were taken as a function of distance from the exposed surface of the as-cast sample. The subsamples were leached and the leachates were analyzed for chromium, nitrate, nitrite and sodium. Nitrate, nitrite, and sodium concentrations were used to provide baseline data because these species are not chemically retained in the waste form matrix to any significant extent and are not redox sensitive. ?Effective? oxidation fronts for Cr were measured for samples containing 1000, 500 and 20 mg/kg Cr added as soluble sodium chromate, Na{sub 2}CrO{sub 4}. For a sample cured for 129 days under field conditions, leachable Cr (assumed to be the oxidized form, i.e., Cr{sup +6} was detected in the top 15 to 20 mm of the sample spiked with 1000 mg/kg Cr. Below about 20 mm, the Cr concentrations in leachates were below the detection limit (< 0.010 mg/L) which indicates that the oxidation as the result of exposure to air was limited to the top 20 mm of the sample after exposure for 129 days and that the bulk of the waste form was not affected, i.e., the Cr was stabilized and insoluble. For samples cured in the laboratory, leachable Cr was detected in the top 8 mm of the Cr 1000 sample cured in the laboratory for 37 days. Between 8 and 14 mm, the concentration Cr in the leachate dropped by a factor of about 20 to just above the detection limit. These depth of penetration results indicate that the rate of advancement of the oxidation front for samples spiked with 1000 mg/kg Cr cured under ?field conditions? for 129 days is less than that for the sample cured in the laboratory for 37 days, i.e., 0.156 and 0.216 mm/day, respectively. Additional data are presented for samples spiked with 500 and 20 mg/kg Cr. In summary, cementitious waste forms are porous solids with a network of interconnected pores ranging in diameter from 10E-10 m to greater than a few mm. The oxidation process is assumed to occur as the result of oxygen transport through the interconnected porosity which may be filled with air, aqueous salt solution, or both. Upon oxidation, the Cr becomes soluble and can be transported in solution through aqueous pore fluid or leachate.

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DOE OSTI2018

Performance Assessment of a Generic Nuclear Waste Repository in Shale - 18392

A performance assessment (PA) is an important component of a comprehensive safety analysis for a nuclear waste repository. In a PA, probabilistic simulations of the total repository system are performed, and results are evaluated against performance metrics. Uncertainty and sensitivity analyses help prioritize additional research and model development. The United States Department of Energy (DOE) has been developing a state of the art PA simulation software tool-kit, Geologic Disposal Safety Assessment (GDSA) Framework, that couples increasingly higher fidelity models of subsystem processes into total system PA simulations. Over the past several years, PAs of generic repositories in three geologic media (salt, shale, and crystalline rock) have demonstrated ongoing developments in capability. The current PA of a nuclear waste repository in a generic shale formation showcases GDSA Framework, including capabilities in domain discretization (Cubit), multi-physics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). The generic shale reference case considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel (SNF) in a generic shale formation. PA simulations account for the thermal load and radionuclide inventory of the waste form, components of the engineered barrier system including waste package and buffer, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Two repository layouts are considered, one for emplacement of waste packages containing 12 pressurized water reactor (PWR) assemblies, and one for 4-PWR waste package emplacement. Model domains are half-symmetry and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high-performance computing (HPC) environment on 512 or 2048 processes. The governing equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation; waste form dissolution; radioactive decay and ingrowth in aqueous, solid, and adsorbed phases; and a simple biosphere based on a pumping well. Simulations are run to 106 y; dose and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Given the assumptions of the reference case, results of 12-PWR and 4-PWR simulations are very similar. Dose at the pumping well 5-km downgradient of the repository does not exceed 10-9 Sv/y. I-129 is the largest contributor to dose. The dose due to Cl-36, the next largest contributor, is orders of magnitude smaller. At the pumping well, the concentration of I-129 5-km downgradient does not exceed 10-14 mol/L. Rank correlation coefficients indicate that the maximum concentration of I-129 within the aquifer is sensitive to shale (repository host rock) porosity and aquifer permeability. These results affirm that HPC-capable codes can be used to simulate important multi-physics couplings directly in a total system performance assessment of a geologic repository, and can be used in prioritization of future research and model development. (authors)

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