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DOE OSTI2019

Solid Secondary Waste Immobilization in Cementitious Waste Forms at the Hanford Site - 19081

During operations of the Hanford Tank Waste Treatment and Immobilization Plant, solid secondary wastes (SSW) will be generated. These materials include HEPA filters, granular activated carbon (GAC), and silver mordenite (AgM). The planned disposal pathway for SSW is through solidification (blending) or encapsulation in a cementitious waste form and then disposal on site at the Hanford Site's Integrated Disposal Facility. Performance assessment evaluations of SSW forms have been performed; however, limited experimental data on simulated SSW is available as input. This paper will cover efforts to fill this gap in experimental data through testing of simulated SSW materials. Two grout formulations were used in the tests: 1) Hanford grout mix 5, a mix of 25 wt% ordinary Portland cement (OPC) and 75 wt% fly ash (FA); and 2) the Cast Stone dry blend of 47 wt% blast furnace slag, 45 wt% FA, and 8 wt% OPC. The sorption (Kd) of potential contaminants (iodide, iodate, technetium) to the grouts and SSW materials was investigated in simulated grout pore water. Waste forms containing GAC and AgM were prepared using the two grout formulations and characterized using scanning electron microscopy energy dispersive X-ray spectroscopy and X-ray diffraction. Acquiring this information will further support the candidacy of cementitious waste forms to be used for disposal of SSW at Hanford. (authors)

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DOE OSTI2019

Decommissioning, Decontamination, and Disposal of LLW-LL, and ILW-LL at a Former Rare Earths Processing Facility in Malaysia - 19087

Geosyntec Consultants, Inc. (Geosyntec) was the design-build Contractor for a project involving the remediation and disposal of low level (LLW-LL and ILW-LL) NORM and TENORM wastes at a former rare earth processing facility in Malaysia. These sites were associated with a plant that extracted rare earth elements from local tin ores. Because of the naturally occurring radioactivity contained in the raw materials, these productions resulted in by-products and residues containing LLW-LL and ILW-LL thorium hydroxide and uranium hydroxide. The total mass of uranium (∼120 tons) and thorium (∼1,200 tons) managed in a single repository required the project design incorporate requirements related to inventory, security and safeguards around nuclear precursor materials. The requirement to consider a performance over a period of more than 10,000 years led to significant challenges in design and modeling of the disposal facility as this disposal facility was purposely designed and permitted as a final disposal for the LLW-LL and ILW-LL. The first phase of the project was the decommissioning, decontamination, and demolition of the 14- acre former rare earth processing plant and excavation of contaminated soils, construction of an engineered cell at the repository for waste disposal, and transportation and disposal of the waste materials into the cell. Final status surveys were carried out to allow the site to be successfully released by the Malaysian regulatory authorities as a clean site for re-development as common industrial use. Over 11,000 loads of material, including approximately 75,000 cubic meters of excavated soils and plant debris, were transported in specially designed steel transport boxes on public roads for disposal in the engineered cell. The second phase of the project included the design and construction of a second engineered cell, enhanced with protective barrier systems and a final cover system constructed of natural materials. Ancillary facilities such as substations, offices, processing buildings, and wastewater treatment plants were constructed to support retrieval and processing of the materials from more than 87,000 individual drums using a semi-automated process. The drums, waste packages, and other materials in the storage facility were systematically extracted, sorted, inventoried, solidified using Portland cement, transported, and disposed in the cell. All temporary facilities were decontaminated, demolished, and disposed in the cell. Radiological inventory and radiation safety programs were implemented to quantify the waste activity and to control exposure to workers and the public during the project. The total mass of radioactive contaminated waste disposed in the two projects is nearly 200,000 tons. The projects were carried out under a high level of scrutiny from Malaysian and international regulatory bodies. Between 2014 and 2017, performance monitoring and modeling activities were carried out to demonstrate that the disposal facility was performing in accordance with design requirements established by regulatory agencies. (authors)

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DOE OSTI2019

Treatment of Foreign Radioactive Waste at the BELGOPROCESS Plant - 19345

Belgium has used a large number of nuclear technologies during the last 60 years. Power generation, reprocessing, fuel fabrication, isotope production and research activities have produced a large variety of radioactive waste streams. BELGOPROCESS is the operator of the centralised waste treatment facilities located in the Campine region in Flanders. Over the years BELGOPROCESS has developed the know-how and facilities to treat each one of these waste streams. BELGOPROCESS is a subsidiary of NIRAS, the governmental institution responsible for the safe management of radioactive waste and enriched fissile materials in Belgium. A number of these waste streams are only produced during a short period of time. After processing the existing stockpile the treatment facility is put into operational standby and eventually prepared for decommissioning. In order to treat the stockpile of some types of legacy waste within an acceptable time frame, the treatment facilities needed to be designed at a certain treatment capacity. Since the majority of the legacy waste is processed nowadays, a significant amount of treatment capacity becomes available. In June 2006 the Belgian government decided that this know-how and facilities should become available to foreign waste producers. This would allow BELGOPROCESS to fill up the available capacity of its facilities and consequently lower the cost for the Belgian waste producers. Nowadays two facilities are open to accept foreign radioactive waste. The CILVA-facility can be used for sorting, cutting, incineration, precompaction, supercompaction and cementation of low level solid waste. The PAMELA-facility - a former vitrification plant - can be used for the treatment of medium level and alpha-bearing waste. In this paper we will discuss the technical characteristics of the CILVA-incinerator and a showcase for the treatment of foreign waste at our facilities. In 2016 BELGOPROCESS started a cooperation with MATE-FIN for the incineration of low level burnable waste from the Cernavoda NPP in Romania. (authors)

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DOE OSTI2019

The Durability of Containers for the Disposal of Radioactive Waste - Swiss and Belgian R and D Studies - 19392

In this paper, we present a summary of the results from research and development (R and D) studies associated with the durability of steel-based disposal canisters for high-level waste (HLW) and spent nuclear fuel (SNF), focusing on experimental studies carried out in support of the Swiss and Belgian disposal programmes. We present the results of laboratory studies carried out for periods of the order of ten to twenty years, and compare them with more recent observations from in situ experiments carried out at the Mont Terri and Grimsel underground laboratories. For the Belgian programme, which is currently envisaging relatively high gamma dose rates outside the steel overpack, we also show the results of studies focusing on gamma radiation at dose rates compatible with those expected in the disposal system. These studies show how, in bentonite systems, corrosion rates decrease to values lower than 1 μm year{sup -1} after an initial period. In the case of cement-based systems, corrosion rates lower than 0.1 μm year{sup -1} are typically observed, with modest effects of gamma irradiation. (authors)

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DOE OSTI2019

Immobilization of Radioactive Wastes Using Geopolymers - 19483

This work explores an encapsulation and immobilization technology for spent radioactive ion exchange (IX) resins using geopolymer materials as a matrix. Geopolymer-IX resin composites with different IX resin loadings were prepared by mixing an aluminosilicate material (metakaolin), sodium hydroxide and inactive IX resin (Amberlite IRN 150) and curing the mixture at room temperature with or without γradiation applied using an external radiation source. The metakaolin and sodium hydroxide reacted to form a geopolymer framework encapsulating the IX resin beads. The IX resin loading capacity and the encapsulation efficiency of the geopolymer material were investigated by examining the morphology and distribution of resins within the geopolymer framework using optical microscopy, and by measuring the compressive strengths of the geopolymer-IX resin composites. These properties of the geopolymer-IX resin composites were compared with those of ordinary Portland cement (OPC)-IX resin composites. The geopolymer matrix allowed for a higher IX resin loading capacity than OPC. The geopolymer-IX resin composite was stable under γ-radiation for 28 d. In leaching tests performed with encapsulated Sr-loaded IX resins, the ability of the geopolymer to prevent leaching of Sr{sup 2+} into solution was greatly superior to that of OPC. (authors)

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DOE OSTI2019

How do Early Age Geopolymers Respond to Gamma Irradiation Exposure? - 19515

Geopolymers have been proposed by the UK nuclear industry as a potential replacement for the currently accepted blend of Portland cement and supplementary cementitious materials, such as fly ash and blast furnace slag. This preferred grouting system is not ideal for the encapsulation of some problematic waste streams, such as oils and reactive metals. Metakaolin based geopolymers have been selected as the alternative grouting system, as this mitigates the supply risk which is currently associated with the supplementary cementitious materials described above. To allow these systems to be used as an alternative nuclear grouting system, the effect of irradiation on the 'neat' grout needs to be understood. The majority of the literature appears to present studies of the irradiation of aged grouts (3 - 7 days). This study will assess samples that have been irradiated at early age curing (20 hours). The effects of clay type and total dose will also be studied, and a comparison between the effects of rotary and flash calcined clays, irradiated to doses up to 1 MGy will be presented. The thermogravimetry data indicates that irradiation promotes the liberation of water and also the formation of carbonate species. The rotary calcined metakaolin liberates more water than the flash calcined metakaolin. Therefore, it is expected that the flash calcined system is more stable to irradiation. (authors)

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DOE OSTI2019

The Use of Geopolymers to Immobilise Sellafield Historic, Intermediate Level Wastes - 19547

A number of geopolymer formulations have been investigated for application to Sellafield ILWs, from sludges to solid items. Trials using Magnox sludge and clinoptilolite simulants have demonstrated that higher incorporation rates of simulated wastes are possible when using a geopolymer, compared to traditional OPC based cement formulations. Trials on scaled zeolite packages, containing simulated AW500 zeolite, have indicated that encapsulation with geopolymers are possible, with mild agitation. This has been possible even when left fully saturated with over standing water and provides a simple method to immobilize ion exchange materials inside sealed containers. The base geopolymers and encapsulated products have been characterised for viscosity, leach rate, physical properties, gas permeability and resistance to radiation. (authors)

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DOE OSTI2019

Development of Polymer Waste Form for Spent Ion-Exchange Resins

A large amount of radioactive waste streams has been treated by ion-exchange resins at most of nuclear facilities for many years. These activities leave radionuclides as well as chemical contaminants on the spent resins, and the spent resins are considered as low- and intermediate-level radioactive waste. Spent ion-exchange resins require special immobilization treatment to meet the waste acceptance criteria for disposal. Immobilization of spent ion-exchange resins can be processed by mixing or incorporating them with suitable solidifying materials, such as cement, bitumen, organic polymers, and other materials or a combination of them. The selection of immobilization options for spent ion-exchange resins must consider their physical, chemical, and radiological characteristics. In this study, direct immobilization method was used to solidify the spent ion-exchange resins using organic polymer for improving the waste content. The objective of this study is to test the immobilization of spent ion-exchange resins using polymer waste form and enhance the solidified waste content. The specific objective of this study was to develop polymer waste form to solidify spent ion-exchange resins. Leaching (ANSI/ANS 16.1) and compressive strength tests were performed to meet the acceptance criteria of waste form for disposal. Ion-exchange resin has been solidified physically and chemically stable in the polymer waste form. The leaching test (ANSI/ANS16.1) for two polymer waste form specimens were conducted. The leachability indexes of sample A and B showed to meet the acceptance criteria for disposal. The compressive strength tested showed higher values in polymer waste form prepared with ion exchange resin compared to CoCl{sub 2} polymer form (sample B). Although spent resin was successively solidified in polymer matrix, the future experiment will develop a polymer waste form as a pre-treatment of spent resin for size reduction using a ball milling (500 RPM, 280 min)

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DOE OSTI2019

The Effect of Glass Chemistry on Caesium Volatility

The UK currently has 99,000 m3 of intermediate level waste (ILW). The current preferred treatment route is cementation. However, thermal treatment could offer a better long term option due to higher waste loading and lower storage costs. This project will focus on volatile retention in the vitrification process to make thermal treatment a more viable option. A series of iron phosphate base glasses were prepared with the nominal molar composition xM{sub y}O{sub z}(40 - x)Fe{sub 2}O{sub 3}-P{sub 2}O{sub 5} (M{sub y}O{sub z} = B{sub 2}O{sub 3}, MnO, ZnO with x = 5%). Samples are given names of the form IPG-1-X5 where X = B, Mn or Zn. For all batches, reagents were weighed and mechanically mixed to produce 150 g of glass. melted at 1150 deg. C for 3 h, Samples were immediately transferred to an annealing furnace and annealed at 450 deg. C for 1 h before being cooled to room temperature at 1 deg. C min{sup -1}. 75 g batches was produced with a target of 2 wt.% caesium oxide. Each batch was placed into a mullite crucible and placed in the sealed system seen below. Each canister was placed into a top-loading resistance furnace and ramped to 950 deg. C at a rate of 3 deg. C min{sup -1} under a nitrogen atmosphere with a flowrate of ∼0.9 L min{sup -1}. The samples were held at this temperature for 3 h before power to the furnace was removed and the system allowed to cool naturally. Once cool, the steel canister was opened and the entire apparatus (steel canister, mullite crucible and tubing) was washed through with ∼50 ml of 0.5 M HNO{sub 3} to allow for a complete mass balance of the caesium species. A series of Iron Phosphate glasses have been produced and characterised to investigate the influence of glass chemistry on the volatility of caesium. All glasses were doped with ∼2 wt.% Cs{sub 2}O and re-melted at 950 deg. C for 3 hours in a specially designed furnace set-up. Results of these experiments showed that the addition of 5 mol. % ZnO greatly increased the retention of Cs{sub 2}O whilst B{sub 2}O{sub 3} was detrimental to the Cs{sub 2}O retention. A series of iron phosphate based glasses have been produced with 5 mol.% additions of B{sub 2}O{sub 3}, MnO and ZnO, substituted in place of Fe{sub 2}O{sub 3}. All samples were analysed before and after being doped and re-melted with a 2 wt.% addition of Cs{sub 2}O. The effect of each additive on the base glass has been reported showing that both B{sub 2}O{sub 3} and ZnO increase the glass transition temperature, whilst MnO additions decrease the thermal stability of the glass. The Q-speciation of the glasses was investigated, and the results showed that the additives did not significantly influence the Q-speciation in comparison to the base glass. It has been shown that the addition of ZnO dramatically increases the retention of Cs during re-melting by almost a factor of 10 in comparison to the base glass, whilst the addition of MnO addition increased the caesium retention but only by a factor of ∼3. The addition of B{sub 2}O{sub 3} had a detrimental effect on the volatility in comparison to the 40Fe{sub 2}O{sub 3}-60P{sub 2}O{sub 5} base glass.

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DOE OSTI2020

The TRANSCEND University Consortium: Theme 2 - Site Decommissioning, Deconstruction and Remediation - 20329

This paper will discuss the aims, objectives and progress to-date under Theme 2 of the TRANSCEND consortium project: Site decommissioning, deconstruction and remediation. Decommissioning nuclear sites involves waste retrieval, decontamination, deconstruction and, where necessary, containment and/or remediation of the remaining structure and surrounding land. Critical to management of these processes is limiting radiation exposure for the workforce, restricting the spread of radionuclides in groundwater, surface water and airborne particulates, and minimising the volume of contaminated waste for disposal. The aim of Theme 2 research is to develop new technologies for monitoring, remediation and containment that serve to minimise the volume of radioactively contaminated waste for disposal, for application prior to, during and after retrieval, deconstruction and decontamination operations. Prior research, conducted under the previous DISTINCTIVE project, demonstrated that colloidal silica grout can penetrate low permeability materials (including cement) for hydraulic barrier formation, and improved sorption capacity. The silica grout can be injected at surface using extremely small (potentially gravity-driven) fluid pressures, without the need for borehole drilling. Current work is investigating the erodibility of silica-grouted soils for inhibition of airborne and water-borne particulates as well as enhancing the grout's sorption capacity by addition of other materials to provide a chemical, as well as hydraulic, barrier to subsurface migration. EK remediation uses low voltage DC current to control migration of contaminants in porous media as well as to remove or degrade them. Researchers in the consortium have already demonstrated that low-energy ex-situ EK techniques can be used to provide remediation and volume minimisation for AWE legacy wastes in the UK. Under the current project researchers are building EK test cells containing simulated site materials at laboratory and intermediate(m)-scales to: remove, focus or degrade contaminants (remediation or waste minimisation); and direct subsurface water, chemical and colloid flow (fencing/containment or forced migration). This lab-based research is being informed by numerical models of EK processes that can subsequently be used to design full-scale on-site applications by nuclear site holders. Research under theme 2 of TRANSCEND will ultimately combine novel non-invasive detection technologies with EK techniques and colloidal silica grout barriers, to optimise the containment of radionuclide contamination in soils. Thus, allowing us to detect in-situ contamination, mobilise it to a selected location and grout it in-situ, prior to the initiation of decommissioning and deconstruction operations. (authors)

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DOE OSTI2020

Adhesion Capabilities of Permanent Foaming Fixatives

DOE-EM have identified an operational requirement for a fixative that can immobilize and/or encapsulate residual contamination in 3D void volumes (pipes, gloveboxes, waste containers, etc.) during D and D activities. Failure to safely and effectively immobilize residual contamination can: Put workers at risk; Contaminate the public and environment; Drive up costs. Commercial-off-the-shelf (COTS) polyurethane (PU) foams is one possible solution and is currently being investigated by Savannah River National Lab (SRNL). Adhesion testing will dictate how well PU foams will adhere and immobilize residual contamination onto a substrate. Results can provide a performance criteria for Section 5 of ASTM E3191: Standard Specification for Permanent Foaming Fixatives Used to Mitigate Spread of Radioactive Contamination. Initial mechanical testing showed intumescent rigid PU foams were best in class (tensile, compression, TGA/DSC) Application of intumescent rigid PU foam in pipe scenario with 'contamination' showed excellent adhesion capabilities as long as foam had contact with pipe. The foam, however, was not able to penetrate 'contamination' but created pockets where the contamination interacted with the substrate. Pretreatment options of pipe were considered to produce better immobilization capabilities. Pretreatment options included COTS products like baking spray, rubber cement, and soap. Foam was still able to immobilize on a global sense, but not a local sense. Essentially acted as a mechanical plug. A tensile tester (MTS Criterion Series 43) was utilized to evaluate the tensile adhesion strength of 6 COTS PU foams. ASTM D1623: Tensile and Tensile Adhesion Properties of Rigid Cellular Plastics. Procedures: PU foams were cured between two smooth 2'' x 2'' 304 stainless steel coupons with hinges glued on plates for gripping support. Parameters Used: Pull rate of 0.1 in/min. Calculations: Tensile adhesion strength, elongation, and ImageJ analysis of how PU foam is left on substrate. Outcome: The PU foam with the best adhesion capabilities would be downselected for further evaluation to be used as a permanent foaming fixative. The rigid PU foams had the best adhesion capabilities with the R1 foam having a max load of 236 N. One of the I-R2 samples reached almost 450 N and another sample reached 2500 N before slipping. The other intumescent foam, I-F4, had the most coverage on the stainless steel coupon's surface from the ImageJ results (61.43% surface coverage). The intumescent foam, I-R2, will be the foam of choice for future testing due its fire retardant and adhesion capabilities. Future directions: Conduct the following experiment in a pipe scenario with the down-selected PU foam. Procedures: PU foam will be cured between in a 304 stainless steel pipe (ID: 4'', OD: 4.5'', Height: 4''). Parameters Used: Compression testing results will be referenced to determine if foam's adhesion strength would be greater its compression strength as its being compressed. Calculations: Compression strength, shear stress, total time elapsed. Conduct further adhesion testing with the down-selected PU foam (I-R2) to be treated as a permanent foaming fixative (PFF). Determine the minimum contact the PFF should have with the substrate for adequate results. Evaluate if PFF's adhesion capabilities is time dependent (3.1.5 of ASTM E3911). Subject PFF to seismic stressors addressed in Safety Basis of Interim Operation documents (SBIO) to further evaluate adhesion capabilities.

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DOE OSTI2020

Influence of Borate on Sorption of Strontium and Barium to Calcium Silicate Hydrate - 20148

Radioactive waste with the decommissioning process at the Fukushima Daiichi Nuclear Power Plant (FDNPP), contains borate with relatively high concentration. As with general low-level wastes, such waste is also assumed the solidification with cementitious materials which calcium silicate hydrate (C-S-H) is one of the main components. In this study, the sorption behavior of strontium (Sr) ions and barium (Ba) ions to C-S-H in the presence of borate was examined by the sorption experiments without drying process. In the sorption experiments, C-S-H was synthesized by the mixture of calcium oxide, fumed silica and a solution with setting the Ca/Si molar ratios to 0.4, 0.8, 1.2 and 1.6. The total amount of a solid phase was 1.5 g, and the total volume of the solution containing borate and adsorbed ions (Sr or Ba) was 30 mL. The concentration of borate was set to 0 M or 0.6 M and the sodium/boron (Na/B) ratios was set to 0.2 or 1.0 by mixing boric acid and sodium hydroxide. The concentration of Sr ions or Ba ions was adjusted to 1 mM. The sorption period was set to 7 days. In the results, the sorption ratios of Sr ions and Ba ions for C-S-H exceeded 80% under all experimental conditions, although the sorption competition of cations with Na ions and the inhibition of the C-S-H formation with generating borate compounds observed by Raman spectroscopy might slightly reduce the sorption of them. These suggest that cationic radionuclides will be immobilized by C-S-H contained in the cements even in the presence of borate. (authors)

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DOE OSTI2020

Borehole Disposal: Update on the Science, Technology, and Potentially Suitable Radioactive Wastes - 20170

A recent Special Issue of the Energies Journal on Deep Borehole Disposal of Nuclear Waste has delivered a timely update on the science and technology of borehole disposal and the types of radioactive wastes it could potentially accommodate. The Special Issue papers discuss (i) circumstances under which a national waste management programme might wish to consider deep borehole disposal [1], (ii) a status report of deep borehole disposal options in Germany [2], (iii) the use of deep horizontal drill-holes in sedimentary, metamorphic or igneous rocks for disposal of spent nuclear fuel and high-level radioactive waste [3], (iv) the analysis of corrosion performance of engineered barrier systems [4], (v) a review of the potential cementing systems suitable for deep borehole disposal [5], (vi) the thermal evolution around heat-generating waste for a wide range of material properties and disposal configurations [6], (vii) a geochemical analysis of deep brines focussed on fluid-rock interactions [7], (viii) post-closure performance assessment calculations for deep borehole disposal of Cs/Sr capsules [8], and (ix) an example safety case for deep borehole disposal of nuclear wastes [9]. This paper provides an overview of the main findings from these publications and highlights some of the technological and scientific challenges that need to be overcome when developing a deep borehole disposal concept. (authors)

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DOE OSTI2020

Estimation of Sorption Behavior of Europium to Calcium Silicate Hydrate by Thermal Analysis - 20172

High-level radioactive waste generated from reprocessing processes will be disposed in a geological disposal that is deep underground. Calcium silicate hydrate is formed as a secondary mineral in the repository via the reaction of calcium ions leaching from the cement and the silicic acid dissolved from the host rocks. In this study, the sorption behavior of europium, which is also known as a chemical analog of americium, on calcium silicate hydrate was examined. The solid phase was analyzed via thermogravimetry- differential thermal analysis, differential scanning calorimetry, Raman spectroscopy, fluorescence spectroscopy, and fluorescence lifetime analysis. To understand the behavior of calcium silicate hydrate under conditions saturated with groundwater, the study samples were analyzed using the above methods, except for deferential scanning calorimetry, without a drying process. The results of the differential scanning calorimetry analyses showed that the temperatures for the endothermic peaks of calcium silicate hydrate for the lower Ca/Si ratios were larger than those for the higher Ca/Si ratios for coexistent conditions of europium ions. Hence, the relationships of the temperatures of the endothermic peaks and the Ca/Si ratios were reversed by the addition of europium ions. These study results suggest that calcium silicate hydrate can immobilize europium ions, thus preventing the production of colloidal forms. (authors)

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DOE OSTI2020

Aging of Concrete for the Evaluation of Repair Materials to Protect the Walls of the HCAEX Tunnel at Savannah River - 20301

H-Canyon, located in the Savannah River site, is a unique facility for chemical reprocessing of plutonium, highly-enriched uranium, and other radioactive materials. The exhaust gases of the H-Canyon are sent through the H-Canyon Exhaust (HCAEX) tunnel for contamination removal. Robotic inspections of the tunnel revealed significant degradation of the reinforced concrete structure that was associated with acid attack, and could compromise the structural stability of the tunnel. Thus, the identification and evaluation of potential repair materials that could be applied on the degraded walls to mitigate and prevent further degradation is of significant interest to the Department of Energy and the Savannah River representatives. This research effort has been divided into two phases: 1) Development and evaluation of aged concrete under accelerated aging conditions (which is the focus of this paper) and 2) Evaluation of potential repair materials applied on aged and non-aged concrete under simulated aggressive conditions. In order to develop and evaluate concrete samples exposed to accelerated aging conditions in simulated aggressive environments, a literature review of the HCAEX tunnel was conducted that included 1) characterization and extent of the concrete damage, 2) environmental conditions inside the tunnel and 3) primary deterioration mechanisms. In addition, potential coatings and/or repair materials for degraded concrete surfaces exposed to aggressive environments (primarily acidic) were selected from the literature review and the most common testing and measurements for evaluating acid attack phenomena, erosion, etc. were reviewed. From the literature review findings, a preliminary bench-scale test plan for the concrete aging was developed, including accelerated aging tests with aggressive conditions (acid fumes, humidity, etc.). Concrete samples were exposed to the aging accelerated conditions (e.g. immersion in acid solutions) and visual inspection, mass loss and pH change (acid solution) were recorded over time. Correlations between the visual inspection, mass loss and pH changes results and the aging time or the aging conditions were developed. Specimens submitted to the highest acid concentration showed the fastest and most intense degradation. The type of coarse aggregate (limestone) used for the concrete seemed to be the cause of the fastest aging observed, compared to the cement paste. The research findings created the foundation for the ongoing investigation, in which new concrete samples with a mix design similar to the HCAEX tunnel will be tested and will serve as the substrate for testing the selected coatings and /or repair materials. In this paper, the literature review and preliminary results from the aging tests are provided. (authors)

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DOE OSTI2020

The Impact of Ethylenediaminetetraacetic Acid (EDTA) on the Sorption of Nd(III), Th{sup (IV)}, and U(VI) onto Dolomite in WIPP-Relevant Brines, GWB and ERDA-6 - 20356

Storage of legacy, transuranic waste at the Waste Isolation Pilot Plant (WIPP) requires accurate risk models for performance assessments. These models must predict the long-term fate and transport of contaminants in the WIPP and surrounding environment. Laboratory support is necessary for the development of an accurate model through better understanding of contaminant behavior in the WIPP. The WIPP is a deep geologic repository that is characterized by high ionic strength due to evaporite deposits in the Salado formation (porewaters up to 7.4 M). High concentrations of sodium chloride (NaCl), calcium chloride (CaCl{sub 2}), and magnesium chloride (MgCl{sub 2}) are particularly of concern due to their potential impact on solubility and ultimately transport of contaminants of concern. Additionally, ligands capable of forming strong complexes with metals are found in significant concentrations in the WIPP environment due to their disposal from waste processing and formation during degradation of repository components (e.g. cellulose degradation by calcium hydroxide in cement). Consequently, there is a need to investigate the fate and transport of contaminants in the WIPP environment with a focus on the effects of high ionic strength and strongly coordinating ligands. Contaminants of high concern include the actinides (U(VI), Pu(IV), and Am(III)). Laboratory support of the WIPP modeling efforts was provided by this research. A series of batch experiments were utilized to study the impact of ionic strength, contaminant concentration, and the presence of ethylenediaminetetraacetic acid (EDTA) on the sorption of Nd(III), Th(IV), and U(VI) onto dolomite, a carbonate mineral found within the Culebra formation. The Culebra formation is the most transmissive layer above the WIPP, making transport of contaminants the most likely route in previous risk assessments. Two brines specifically designed to imitate conditions within the WIPP environment were used in these experiments. The U.S. Energy Research and Development Administration Well 6 (ERDA-6, 95%) brine simulates a low Mg environment, while the generic weep brine (GWB, 95%) simulates the WIPP environment with a high Mg concentration. Mg is added to the repository in the form of MgO as an engineered barrier and is naturally present in some formations. It is expected to scavenge carbonate from solutions decreasing formation of the highly soluble and competitive uranium-carbonate complexes. EDTA is used during manipulations of nuclear materials due to its ability to form strong, stable complexes with the actinides and may be found in high concentrations (up to 0.08 mM) within the WIPP as waste canisters degrade. Further, it may provide additional avenues for unwanted transport of contaminants. Nd(III) and Th(IV) were used as stable analogues of Am(III) and Pu(IV), respectively, to represent the most common oxidation states of actinides in the WIPP environment in addition to U(VI). The potential for colloid formation was also investigated with low and high concentrations of contaminants (10 and 1000 μg/L) using three size-dependent particle separation steps. The results of this research provide a better understanding of potential behavior of actinides in the presence of EDTA and within WIPP-relevant brines. (authors)

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DOE OSTI2020

Grimsel Test Site - A Successful International Underground Research Laboratory for Many Decades - 20429

For more than 35 years, Nagra and its partners from around the world have been conducting underground research projects at the Grimsel Test Site (GTS, www.grimsel.com) to contribute to the development and confirmation of safe geological disposal concepts and for the characterization of suitable host rock formations. Over the years, the results of this internationally recognized research program have been, and continue to be, incorporated directly into exploration programs, modelling, safety, and engineering feasibility studies on options for deep geological repositories. Each project of the GTS program involves field-testing, laboratory studies, design and modelling tasks, and integrates all scientific and technical aspects. Each project phase is planned with a duration of three to five years to facilitate practical and administrative aspects and allow flexibility for updating the overall project plans with the latest findings. Scientific and engineering interaction among the different projects is ensured via an international steering committee meeting. Hosting an IAEA level C radiation- controlled zone, which allows use of radionuclides, including actinides such as thorium, uranium, neptunium, plutonium and americium, in in-situ experiments is one of the reasons why GTS also developed as a center of excellence for work with radioactive tracers under realistic in-situ boundary conditions. Last year, a new five-year program (2019 to 2023) started which includes projects with a planning horizon of decades. The new five-year program includes a new phase of in-situ experiments using radionuclides such as migration experiments in the Colloid Formation and Migration project (CFM), the Long-Term Diffusion experiment (LTD) and the newly established C-14 and I-129 Migration in cement project (CIM). The 'High Temperature effects on Bentonite' (HotBENT) project is starting in the current phase and is studying the effects of elevated temperatures (>175 deg. C) on bentonite materials. As a generic underground research laboratory (URL) it is expected that the GTS will provide in the coming years a platform for international collaboration, knowledge development and knowledge transfer for the next generation of scientists and engineers in the area of radioactive waste disposal and geosciences. A key role regarding knowledge transfer and training is provided by the well-established Grimsel Training Center (GTC), which (beside many URL related issues) also covers many general aspects of radioactive waste management. In this paper we provide an overview of the current program at the GTS, focusing on the experiments that study the migration of radionuclides through engineered barrier materials and the geosphere. (authors)

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

Generic Post-Closure Safety Assessment for Deep Borehole Disposal of Heat-Generating Waste - 20451

The deep borehole disposal (DBD) concept for high-activity waste consists of drilling a large-diameter borehole into crystalline basement rock to a depth of about 5,000 m, placing waste packages in the lower portion of the borehole - the waste emplacement zone - and sealing and plugging the upper portion of the borehole with a combination of bentonite, cement plugs, and sand or crushed rock ballast. The waste emplacement zone is several times deeper than typical mined repositories and is well below the typical maximum depth of fresh groundwater resources. Recent work describes a generic reference case for deep borehole disposal of Cs and Sr capsules stored at the Hanford Site. Two scenarios are described: an undisturbed (nominal) scenario, and disturbed scenario, in which a waste package is stuck adjacent to a transmissive fracture zone in crystalline basement. Simulations of the processes affecting radionuclide release and transport in the two scenarios provide the basis for quantitative post-closure performance assessment (PA) of the reference case. This paper presents a probabilistic PA of the stuck waste package scenario, i.e. uncertainties affecting radionuclide mobilization and transport are propagated through 200 realizations of a predictive model to quantify uncertainties in predicted radionuclide concentrations within the borehole and the fracture zone. Simulations predict no radionuclide transport in the borehole to distances greater than 100 m above the stuck waste package. An assumed lateral pressure gradient results in predictions of limited radionuclide transport in the fracture zone at a lateral distance of 200 m from the stuck waste package. These results complement previous results for the nominal scenario and contribute to the development of a generic safety case for DBD. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗