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Advances in Geologic Disposal Safety Assessment and an Unsaturated Alluvium Reference Case

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Depat ment of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (OFCT) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high level nuclear waste (HLW). Two high priorities for SFWST disposal R&D are design concept development and disposal system modeling (DOE 2011, Table 6). These priorities are directly addressed in the SFWST Geologic Disposal Safety Assessment (GDSA) work package, which is charged with developing a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic media. This report describes specific GDSA activities in fiscal year 2018 (FY 2018) toward the development of GDSA Framework, an enhanced disposal system modeling and analysis capability for geologic disposal of nuclear waste. GDSA Framework employs the PFLOTRAN thermal-hydrologic-chemical multiphysics code (Hammond et al. 2011a; Lichtner and Hammond 2012) and the Dakota uncertainty sampling and propagation code (Adams et al. 2012; Adams et al. 2013). Each code is designed for massivelyparallel processing in a high-performance computing (HPC) environment. Multi-physics representations in PFLOTRAN are used to simulate various coupled processes including heat flow, fluid flow, waste dissolution, radionuclide release, radionuclide decay and ingrowth, precipitation and dissolution of secondary phases, and radionuclide transport through engineered barriers and natural geologic barriers to the biosphere. Dakota is used to generate sets of representative realizations and to analyze parameter sensitivity.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Integrated pre-feasibility study for CO 2 geological storage in the Cascadia Basin, offshore Washington State, British Columbia

The Cascadia CarbonSAFE project conducted a pre-feasibility study to evaluate technical and nontechnical aspects of storing 50 million metric tons (MMT) of carbon dioxide in a safe, ocean basalt reservoir offshore Washington State and British Columbia. Sub-seafloor basalts are very common on Earth and enable geological mineralization as a long-term storage mechanism, permanently sequestering the carbon in solid rock form. This project evaluated the offshore storage complex, developed potential industrial source/transport scenarios, built an inventory of existing geophysical/geological data and environmental monitoring capabilities, conducted laboratory studies and reservoir modeling to determine storage capacity, and analyzed economic factors, regulatory requirements, and project management risks. Our team included researchers at Columbia University in New York, Pacific Northwest National Laboratory (PNNL) in Washington, University of Victoria in British Columbia, University of California at Santa Cruz, and GHG Underground in Maine, as well as collaborators at University of Iceland in Reykjavik and University of Washington in Saint Louis. In the study region, large emitters (>100,000 MT/year) generate a total of approximately 40 MMT annually from stationary sources, such as power plants, refineries, ammonia production, and mineral processing plants. Five potential industry source/transport scenarios were identified (three in U.SA S. and two in Canada) to potentially provide 50 MMT of carbon dioxide to the offshore reservoir. One scenario included net carbon-negative sources that would reduce atmospheric levels. The inventory of existing data illustrated that basalt properties in the region are potentially beneficial for long-term storage. Permeability is on the order of 0.1 to 1 Darcies within the uppermost 600 m of ocean crust and density-derived porosity reaches 10–20% in thin layers that provide act as flow channels and provide access to porous and permeable basalt. Fine-grained sediments overlying basalt provide a low-permeability barrier separating the storage reservoir from the overlying ocean. Subsurface monitoring of the storage complex benefits from a unique opportunity close proximity of to the nearby NEPTUNE cabled seafloor observatory, which actively transmits data to shore and could be used to monitor reservoir properties. Injection pressure changes, reservoir chemistry, seismicity, well head corrosion, and potential leakage monitoring would be feasible in real time. Using existing data and published models, the STOMP-CO 2 modeling code developed at PNNL was used to simulate carbon dioxide injection, storage, and mineralization in the reservoir. Preliminary simulations indicate that a 50 MMT carbon dioxide plume injected over a 20-year period remains well within the reservoir area, both laterally and vertically, for at least a 50-year period of post-injection monitoring. Laboratory studies on core samples from Cascadia measured dissolution rates in CO 2 –basalt–seawater mixtures and indicated large variability in rates with the potential for complete mineralization. Compared to published data, the dissolution rate measured in these samples is faster, decreasing the time until full conversion to carbonate minerals would occur. Adding geochemical reactivity in the STOMP-CO 2 model indicates that carbon dioxide would be fully converted to carbonate minerals in 135 years or less after injection ceases. The project assessed the viability and costs for the five scenarios using either dedicated offshore pipelines or shipping vessels. Carbon dioxide capture costs were not considered. Shipping incurred lower upfront CAPEX costs, and because most viable sources were located near the shore and waterways, shipping is preferred; no dedicated carbon dioxide pipelines exist in the region. The project also considered potential risks from mixed-use impacts on surrounding lands and ocean areas, such as active use by commercial and recreational fisherman, and other community considerations in planning for public engagement in the future. Other project risks were identified, including project management, permitting and compliance, and financing, as well as technical issues such as long-term instrumentation, plume migration, and model development. Geological concerns like reservoir permeability, seal, and induced seismicity held relatively low risk. The regulatory framework for offshore storage in the U.S. and Canada may be covered by general legal/regulatory programs, but is not yet fully defined. Existing programs appear to restrict the transportation of carbon dioxide across national boundaries (e.g., Canada to U.S.) and its sub-seabed disposal in some circumstances. Our assessment suggests that developing specific and separate pilot projects will help to advance the technical feasibility and regulatory requirements in each country, and lessons learned from such activities may be transferable elsewhere around the globe.

42 ENGINEERING↗

Separation and Speciation of Mercury from Hanford 241-AP-107 Tank Waste Supernatant

Mercuric nitrate, Hg(NO 3 ) 2 , was introduced into the Hanford flowsheet when it was used to catalyze nitric acid dissolution of certain aluminum alloy fuels and to suppress radioiodine volatilization during nitric acid dissolution of short-cooled uranium metal fuel. Inventory estimates, based primarily on Hanford process-knowledge, suggest a total mercury inventory of 2,100 kg 1 in the Hanford Tank Farms. Aside from the obvious human exposure concerns via tank vapors, mercury species in the tank waste liquid and solid fraction pose risk to the implementation and execution of the River Protection Project mission due to mercury limits imposed under waste acceptance criteria, toxicity characteristic leaching procedure restrictions for secondary waste disposal, and air emission permits. Although there is currently no processing of tank waste at Hanford, waste processing is imminent and representative mercury inventory tracking will be important to anticipate and mitigate stack emissions and secondary liquid waste disposal issues like those experienced at the Savannah River Site.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Influences of Neutron Irradiation on Aggregate Induced Degradation of Concrete

Reactor cavity concrete, the primary support structure for the reactor pressure vessel (RPV), is exposed to chronic neutron/gamma radiation during normal operation. During long term operation, the interaction of neutron irradiation with the concrete’s constituents may cause premature degradation of the concrete structure, as highlighted by a review of concrete properties in nuclear power plant environments. At high fluence levels, as in biological shielding, neutron irradiation may cause amorphization of the crystalline components of concrete, i.e., mineral aggregates, and volume changes, which are typically expansive in nature. Volume expansion can generate tensile stresses within concrete which can induce microcracking and/or result in debonding of aggregate with the matrix. On the other hand, structural disordering can result in changes in the stability of the phase in aqueous environments. This report presents the findings of investigations performed over 3 years through the collaboration of researchers from University of California, Los Angeles, Arizona State University, and Oak Ridge National Laboratory. The first part of the study addresses the effect of neutron irradiation on the durability of aggregates. Comparisons of aqueous dissolution rates, as a function of pH and temperature, were performed for aggregates that are either pristine (i.e., non-irradiated) or exposed to Ar+ ion irradiation at specified doses. These experimental results were coupled with molecular dynamics simulations of atomic scale changes in the crystallographic structure of the mineral and any resulting changes in physical properties (e.g., density) that result. The second section focuses on macroscopic effects of such changes in aggregate behavior in concrete degradation, primarily through changes in the concrete’s microstructure. Particularly, the effects akin to those originating from alkali-silica reaction (ASR) were evaluated. The project is structured to encompass 6 main tasks, as follows: Task 1: Selection of aggregates in relation to their abundance and mineralogy, and their irradiation Task 2: Molecular dynamics (MD): The effects of irradiation on the atomic disordering of aggregates Task 3: The influences of irradiation on aggregate reactivity Task 4: Reactive-transport simulation of expansive product formation in concrete microstructures Task 5: X-FEM-based modeling of damage evolution due to irradiation-induced aggregates reactions Task 6: Benchmarking, verification and validation of the simulation modules Thus, a rigorous understanding of irradiation-assisted mineral alterations was used to develop a mechanistic degradation model for concrete exposed to neutrons, while considering aspects of: chemical reaction (dissolution and precipitation), mass and ion transport, and damage evolutions. The models developed were embedded in a multi-physics aging code for nuclear components and structures called Grizzly, which allows for benchmarking, verification, and validation against expansion data acquired on concrete undergoing alkali-silica reaction, a physically/chemically equivalent form of degradation that develops in the absence of radiation. These findings provide insight into the sensitivities of minerals to irradiation, by highlighting characteristics of concretes and power plants subject to such risks. The research will aid the nuclear power industry to anticipate and mitigate such risks in light of second license renewals.

36 MATERIALS SCIENCE↗

Graphite Steel and Chromium-Nickel Steel Corrosion: Solution Flow Through Corrosion Experiments

The corrosion of steel, when exposed to various compositions of brines is a complex, heterogeneous process involving dissolution and precipitation of multiple solids. The rate at which elements will be released from the corrosion process under these conditions will depend, in part, on effects associated with secondary alteration phases (passivating film) that may potentially form under these conditions. Understanding these processes has required the development of sophisticated methods to sample and quantitatively characterize the composition of the steel as it corrodes and releases elements. The determination of the concentrations of both brine and steel corrosion components in solutions sampled from these tests is required to fully understand the process that may eventually lead to dissolution of corrosion products.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Conversion of Uranium Metal to Uranyl Sulfate Solution

Experiments involving conversion of uranium (U) metal to U 3 O 8 , followed by dissolution to form uranyl sulfate solution, were performed with batches of up to 300 g of U metal. The focus was to identify major U species formed during the oxidation process, determine the particle size distribution of the uranium oxide product, and identify the size of the dissolution vessel needed to accommodate the exothermic reaction during the conversion of U 3 O 8 to uranyl sulfate using 30% hydrogen peroxide and H 2 SO 4 . The information obtained in this study will be used for designing full-scale processes to convert kilogram quantities of U metal to uranyl sulfate solution with a concentration of 300 g U/L.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Application of a Dual-Column Method to Selectively Extract and Purify Berkelium-249 from Other Actinides and Impurities

The super heavy element (SHE) Tennessine (Ts, Element 117) was discovered in 2016 a collaboration of US and Russian scientists vis bombardment reaction of 249 Bk + 48 Ca $\rightarrow$ 117 Ts, while the target element 249 Bk (half-life 330d) was produced and provided by the Radiochemical Engineering Development Center (REDC), Oak Ridge National Laboratory (ORNL). The production of 249 Bk at REDC, ORNL requires a multi-step chemical process of irradiated curium feedstock target to remove all fission product (FP) elements and to separate 249 Bk from other actinide (An) elements (Am, Cm, Cf, Fm and Es), including: Caustic Dissolution, Acid Dissolution, Cleanex, two LiCl Anion Exchanges, LiOH Precipitation, Cation Exchange with AHIB, and with Bk Finishing as the final Bk purification step. The major product of this campaign is 252 Cf, while 249 Bk is one of important by-products. The current Cf campaign for recovery of c.a. 15 mg of 249 Bk per campaign (not including the time for Finishing of Cf, Fm and Es) takes 6 or 7 months, including ~3 months spent on Bk Finishing. Bk Finishing includes two major processes: (1) BERKEX, operated inside a hot cell with two cycles of batch solvent extraction (SX) using di-2-ethylhexyl phosphoric acid (HDEHP) as the extractant for tetravalent 249 Bk from an 8 $M$ HNO 3 -0.3 $M$ NaBrO 3 solution; and (2) Cation Exchange (CX) with AHIB (α-Hydroxy isobutyric acid), operated inside a glove box laboratory. Entire Bk Finishing takes five SX, more than five column runs and several drying down processes for aqueous solutions including one from 500 mL to 1 mL for volume reduction. To shorten the Bk Finishing period and simplify the Bk purification operation steps, possibilities of using other resin columns were explored.Recent trials using columns of either $LN$ resin or $MP$-1 anion exchange resin adsorbing the tetravalent 249 Bk 4+ in 8 $M$ HNO 3 -0.5 $M$ NaBrO 3 , followed by eluting impurities (e.g. Ce 4+ ,Cf 3+ ) and stripping 249 Bk resulted in a new Dual Column Method of selectively separating 249 Bk from other trans-plutonium elements and fission product (FP) impurities. This new method is proposed to replace the current Bk Finishing and provide increased Bk purification effeciency and a shorter processing period of 4 to 5 months (less decay loss for higher 249 Bk productivity). This TM Report discusses the trial experimental results, describes the dual column method with relevant chemistry issues and possibilities being used in further process steps ahead of the Bk Finishing.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Geochemical Changes in Response to CO 2 Injection in a CO 2 -EOR Complex in Northern Michigan (Volume II.1)

The Midwest Regional Carbon Sequestration Partnership (MRCSP) was founded in 2003 as part of the U.S. Department of Energy’s (DOE’s) Regional Carbon Sequestration Partnership initiative. Since its founding, MRCSP has made significant strides toward making CCUS a viable option for states in the region. The public/private consortium, funded through the DOE Regional Carbon Sequestration Initiative, brings together nearly 40 industry partners and 10 states. Battelle, as the project lead, oversees research, development and operations and coordinates activities among the partners. The incremental, phased approach has built a valuable knowledge base for the industry and paved the way for commercial-scale adoption of CCUS technologies. From 2008 to 2020, MRCSP Phase III focused on the development of large-scale injection projects. This report is part of a series of reports prepared under the Midwestern Regional Carbon Sequestration Partnership (MRCSP) Phase III (Development Phase). These reports summarize and detail the findings of the work conducted under the Phase III project. The report describes the geochemical monitoring program under MRCSP to use stable and radiogenic isotope geochemistry in concert with analysis of general geochemical parameters of fluids and gases and analysis of core samples to determine geochemical processes occurring in the reef structure because of CO 2 injection. Specifically, brine and gas samples were collected and analyzed to determine changes occurring between reefs prior to and following CO 2 injection. The analytical results for general geochemical parameters were modeled with chemical equilibrium models to determine if the injection of CO 2 resulted in the mineral dissolution or precipitation. Finally, core samples were collected and analyzed to determine if there was evidence of dissolution features or mineral precipitation.

01 COAL, LIGNITE, AND PEAT↗

Optimized Microwave Digestion and Quantification Procedure for Boron Carbide samples

A microwave-assisted HNO3-H2SO4-HF digestion system was explored/optimized for the total dissolution of Boron Carbide samples followed by multi-elemental determination using ICP-MS and ICP OES, in order to improve the methods used in a previous try using a microwave-assisted H2O2-HNO3-HF digestion system. The samples were microwave digested to accomplish complete dissolution needed to perform quantitative analysis of their metal content. Two Boron Carbide samples were provided to us: JM10 and JM11. The Boron Carbide samples were completely dissolved at the end of the optimized microwave digestion procedure that was developed. The digested samples were analyzed for metals mostly by ICP MS; few elements were analyzed by ICP-OES. Microwave digestion reactions were performed in a Titan MPS instrument (Perkin Elmer, USA). Titan MPS instrument is equipped with Temperature and Pressure regulations and controlled by software. Each microwave digestion batch contained the samples of interest along with a reagent blank and one certified standard (1632e) in order to (i) compensate for the contaminations present in the reagents, (ii) ensure that the digestion was complete and (iii) that there was a good recovery of all the constituents. The certified standard used (1632e) was provided with USGS certificates of analysis.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

The Precision of Product Consistency Tests with Reference Glasses ARM-1 and SRM 623

An interlaboratory study (ILS) was conducted in 2014 to measure the precision with which the ASTM-International product consistency test method A (PCT-A) can be performed using properly prepared samples of reference glass ARM-1 or SRM 623 and test parameter values near the centers of ranges specified in procedure ASTM C1285. The ILS did not evaluate the full range of test parameter values allowed in the PCT-A method and the results do not represent the precision that should be expected if extreme values of the ranges are used. However, the precision measured in the ILS quantified the precision of tests conducted with mean parameters that were used to estimate the effects of individual parameters. Follow-on tests were conducted to assess the effects of several individual test parameter values on the test response and uncertainties in the derived glass dissolution rates. The effects of particle size distributions in glass samples prepared by wet- and dry-sieving methods, methods to remove fines, and the limiting parameter values for the water-to-glass (w/g) mass ratio and test duration allowed in PCT-A were measured. Test responses for all variations except test duration exceeded the precision measured in the ILS for conducting triplicate PCT-A tests within a single laboratory. The accuracy of glass dissolution rates derived from PCT results benefits from using a tighter control of these test variables than required in ASTM C1285. Comparison with previous analyses of the method shows how use of the relative standard deviation to represent the test precision can be misleading, and that the mean response and standard deviation should both be considered.

36 MATERIALS SCIENCE↗

Laboratory Evaluation to Increase Effectiveness of Field-Scale Soil Flushing in the Hanford 100 Areas

This laboratory study was initiated to develop an improved technical understanding of the key controlling geochemical and physical processes of mobilizing residual Cr(VI) in Hanford sediments to provide increased effectiveness in soil flushing activities planned for Hanford 100 Areas. This was accomplished by (1) quantifying Cr mass, release rate, and surface phase changes as Cr is leached from different Hanford sediments; (2) maximizing Cr leaching during soil flushing by evaluating different leach solutions and leach strategies; and (3) minimizing leaching of remaining residual Cr after soil flushing ends. To quantify geochemical controls on Cr leaching from sediments, sequential liquid extractions were used to identify aqueous, adsorbed, and solid Cr surface phases (precipitates or Cr incorporated into other phases) before and after water-saturated and unsaturated leach experiments with pH 8 artificial groundwater and with other amendment solutions. Changes in the Cr release rate from sediments were also correlated to changes in Cr surface phases. During unsaturated infiltration, water advection occurs primarily through larger pores, with less movement through smaller grain layers leading to a slower release of Cr trapped in smaller pores next to smaller grains. To quantify these physical controls on Cr leaching from sediments, 10 ft high 1-D infiltration experiments were conducted at differing leach solution application rates, and Cr leaching during application and subsequent residual water flow was quantified. Qualitative 2-D experiments were also conducted to evaluate surfactant addition to increase flow in low-permeability zones. Cr-contaminated Hanford sediments from the 100-H, 100-K, 100-D, and 200 East Areas exhibited a wide variety of leach behavior from rapid release of a small amount of Cr to fast (i.e., minutes) and slow (i.e., hours or more) release of high Cr. Sediments with low labile Cr (< 0.2 µg/g) exhibited fast Cr release but less overall release, and most (>95%) leached from the sediments within a few pore volumes in water-saturated columns. With nearly no sorption of chromate, aqueous and sorbed Cr (as chromate) were released quickly from sediment within the first few pore volumes, resulting in high leach concentrations, fast initial Cr release rates, and changes in pre- and post-leach extractions. In contrast, sediments with high labile Cr (0.2 to 20 µg/g) exhibited a combination of fast and slow Cr release due to multiple Cr surface phases contributing to Cr release from the sediment at different rates with greater overall release. This resulted in elevated Cr effluent concentrations even after dozens of pore volumes were leached. In addition, the Cr release rate decreased with increasing time during leaching. A decrease in labile Cr (aqueous Cr, adsorbed Cr, pH 5 acetate dissolved precipitates) in post-leach extractions compared to pre-leach extractions showed initial aqueous and adsorbed Cr release, followed by dissolution of high-solubility precipitates such as CaCrO 4 (if present), then slow dissolution of calcite that may contain some chromate and possibly BaCrO 4 . The labile Cr in all sediments correlated well with the Cr release rate (r 2 = 0.81). Labile Cr also correlated well with the leached mass in water-saturated columns (r 2 = 0.98), and therefore can be used for prediction of potential removal with flushing applications.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Recombination of Hydrogen in the Iodine Reactors

The H-Canyon facility is currently dissolving spent nuclear fuel, including Material Test Reactor (MTR) and High Flux Isotope Reactor (HFIR) fuel. Dissolution of aluminum spent nuclear fuel produces hydrogen and other NOx gases. A theory long held by H-Canyon Engineering and Facility Technical Advisors that the silver nitrate-coated berl saddles, present in the iodine reactor, used for off-gas treatment were acting as a catalytic hydrogen recombiner was captured during an external review of the Accelerated Basin Deinventory program. Results from a separate SRNL project using a Raman Spectrometer to monitor the offgas from the dissolution process indicated a lack of the expected hydrogen in the off-gas stream. To test this hypothesis, a laboratory scale iodine reactor was assembled and filled with silver nitrate-coated berl saddles. Testing with this laboratory scale reactor confirmed the recombination of hydrogen when a simulated dissolver off-gas was passed through the reactor containing silver nitrate-coated berl saddles at 173-188 °C. Control experiments performed with uncoated berl saddles resulted in no change to the hydrogen concentration after passing through the reactor. The residence time of the gas in the reactor was varied to determine if residence time had an impact on the amount of hydrogen recombination occurring. Results from these experiments indicated that at the shortest residence times tested (~5 seconds) recombination of the hydrogen still exceeded 90%; however, the percent recombination did increase at longer residence times, reaching 97.9% or greater for residence times over 10 seconds. Finally, testing performed with a simplified off-gas composition containing only hydrogen and air gave similar results, indicating that the presence of NO and N 2 O gases in the stream do not play a role in the recombination reaction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASNF Processing and Packaging Technical Study

The ABD program will require development of a strategy for the dispositioning the Non-Aluminum Spent Nuclear Fuel (NASNF), some of which will be very challenging to handle and dissolve as described in this report. Not all the NASNF is suitable for processing through the electrolytic dissolver. The strategy may involve dissolution of the majority of the NASNF inventory in the electrolytic dissolver combined with some other disposition alternative(s) for the fraction that is not suitable for electrolytic dissolution. This report highlights the risks and challenges associated with the baseline approach of using the electrolytic dissolver to process all of the NASNF. In addition, the characteristics of the NASNF and how those characteristics relate to the risks associated with electrolytically dissolving the specific fuel types.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Predicting and Controlling Corrosion (Abbreviated Final Report)

Corrosion accumulates significant replacement costs in the transportation, utility, manufacturing, and infrastructure sectors. It also underpins several of LLNL’s core missions in stockpile stewardship, defense, and energy security. Whereas most corrosion and aging models are empirically parameterized to describe well-defined conditions late in the reaction, the factors that determine the early stages of corrosion—during which mitigation could be most impactful— are poorly understood. This project addressed the critical need for new approaches to predict the kinetics of corrosion initiation based on firm physical and chemical understanding. The activities encompassed degradation of relevant metals in both hydrogen-rich and environmental corrosion scenarios. The team integrated state-of-the-art multiscale simulation, in situ characterization, and data science within three technical thrusts: hydriding of Ti alloys; aqueous corrosion of Al and Ni-Cr alloys; and degradation of additively manufactured 316L stainless steel. In each case, novel capabilities were developed to identify and track the impacts of key atomistic, compositional, and microstructural features on the metal systems. For hydriding, protocols were developed to tightly integrate multiscale models, advanced multimodal characterization, and machine learning to determine how hydrogen interacts with native passivating surface oxides and nucleates new undesired phases, shedding new light on the critical role of grain boundaries, interfaces, and atomically disordered regions. For aqueous corrosion, the project demonstrated methods to predict dissolution rates of metal surfaces in corrosive solutions, to measure and understand microstructural and grain orientation effects on corrosion susceptibility, and to investigate competing growth and dissolution kinetics of surface oxides. For additively manufactured metals, analysis using state-of-the-art microscopy techniques revealed the role of specific heterogeneities invoked during laser processing, including cellular structure, dislocations, and precipitates, on corrosion susceptibility. In addition to new capabilities and understanding, the project provided an avenue for workforce development, as well as key partnerships with stakeholders in corrosion science.

08 HYDROGEN↗

Formation of Zeolites Responsible for Waste Glass Rate Acceleration: An Experimental and Computational Study for Understanding Thermodynamic and Kinetic Processes

Before nuclear waste glasses can be disposed at geological disposal facilities, a performance assessment (PA) must be conducted to provide confidence to the regulators and public that the release rates of radioisotopes and hazardous materials into the environment are below the regulatory limits. Because the release rate of radionuclides is directly related to the dissolution rate of the radionuclide-containing glass, the corrosion behavior of the glass must be well understood for the disposal site lifetime. For simplicity, glass dissolution in static conditions has been categorized into three different stages, referred to as initial rate (Stage I), residual rate (Stage II) and, depending upon the secondary phases formed, a possible delayed resumption in the alteration rate (Stage III). The last of these stages is the least understood and has the possibility to drive radionuclide release rates above regulatory thresholds.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Characterization of Contaminants and Mobility in WMA A-AX Boreholes D0006 and D0008

This study was initiated to determine if subsurface contaminants in the A-AX vadose zone at the Hanford site (boreholes D0006 and D0008) caused corrosion of nearby stainless steel well casings (299-E-24-19, 299-E25-46, and 299-E25-236) and, based on the type and distribution of contaminants, determine if contaminants are from one or more sources including a) the A-104 and/or A-105 tanks, b) the 242A evaporator, and c) the 200-E-286 ditch. A tiered approach was used for characterizing contaminants in vadose zone sediments starting with mobile and total contaminant concentrations (Tier 1), characterizing contaminant speciation (Tier 2), and measuring contaminant leaching (Tier 3). Total and mobile contaminant measurements at seven depths showed elevated sulfate and nitrate in all samples (130 to 280 ft depth), elevated chloride in five depths, and no elevated radionuclides except I-129 in D0006 at 206 to 208 ft depth. If these contaminants were initially acidic, stainless-steel corrosion would occur. However, because sediments likely have a high acid neutralization capacity, an acidic spill would have been neutralized within tens of feet, so casing corrosion would likely occur in shallow sediments. GEL labs data also shows low (but above natural) chromate from shallow to 170 ft depth, then higher chromate to 280 ft, suggesting a surface Cr spill or casing corrosion. GEL labs data also showed evidence of more than one plume, with deep nitrate, sulfate, and chloride migration, but only shallow tritium migration. Tritium should migrate nearly unretarded in the subsurface like nitrate and chloride. Elevated aqueous cations and anions in the D0006 206 to 208' depth pore water (SO 4 2- , NO 3 - , Na + , K + , Ca 2+ , Mg 2+ , Si) could be the result of acid disposal and calcite dissolution at shallow depth, with deeper migration of SO 4 2- , NO 3 - from acids and Ca 2+ and Mg 2+ from calcite. Elevated Si (160 mg/L) indicates dissolution of clays or other silicates, which can also occur in acids. Acid extractable metals did show elevated Cr (but not Fe, Ni, Mo, and Mn) above natural levels, which may indicate stainless-steel well corrosion or transport of metals from a shallow spill. The elevated I-129 in D0006 at 206 to 208' depth was low with 0.34 μg/g in pore water and 1.5 μg/g total extractable I-129 in the sediment, which indicates significant fraction of I-129 was bound in one or more precipitates. Although total I-129 was below detection limits in most extractions due to high molybdate interference, I-127 analysis provided some insight into potential precipitate phases. Total I-127 was mainly extracted by acetate or acetic acid (67%), which may be iodine in calcite. Inorganic carbon measurement identified 1.65% calcite, and total organic carbon (TOC) was below detection limits, so iodine was not associated with organic matter. Nearly all the I-127 was present as iodide. In 1-D leach experiments, I-127 as iodide leached from the sediment within 2 pore volumes. Due to high and dynamic molybdate leaching from the sediment and interference with I-129 analysis, there were few quantifiable I-129 measurements. The calculated molybdate release rates from the sediment suggested Mo was being released from the same precipitate phase, such as molybdate incorporated into calcite. Overall, because there was little radioactive contamination measured in D0006 and D0008 boreholes, potential releases from A-104 and A-105 did not reach this location south of the A tank farm. Elevated cations and anions found in D0006 and D0008 at depth as the result of acidic disposal could cause shallow stainless steel casing corrosion. The nearby 242A evaporator generally contained alkaline waste and tritium, but the nearby 200-E-286 ditch generally contained high Cl - liquid effluent, so these are unlikely to be acidic sources.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development of Corrosion Resistant Coatings for Structural Materials for Liquid Fueled Molten Salts Reactors Applications (Final Report)

Current structural alloys, code certified for the MSRs temperature ranges, contain high levels of chromium, making them highly susceptible to molten halide salt corrosion. One potential solution to circumvent the need for code certification of novel alloys, which is expensive and time consuming, is to design claddings that will protect the underlying code certified materials from corrosion damage during operation. In this work, we examined the corrosion of Ni and Cu electroplated, Ni and CuNi weld overlay, Mo-laser clad, and carburized claddings on SS316H for use in molten salt reactor environments. After characterization of the cladded materials (Task 1), static corrosion tests were used to assess corrosion resistance of the claddings in typical molten fluoride salts (Task 2). The corrosion tests were performed in molten FLiNaK at 700°C up to 1000 hours. Pre- and post-corrosion Scanning Electron Microscopy (SEM), energy-dispersive-spectroscopy (EDS), Transmission Electron Microscopy (TEM), X-ray Diffraction (XRD) and glow-discharge-optical-emission-spectroscopy (GDOES) were performed on cladding cross-sections and surfaces to evaluate degradation. The Cu and Ni electroplated samples, as well as the carburized samples, showed excellent corrosion resistance relative to the bare SS316H. To assess high temperature cladding stability, ageing experiments were performed at temperatures up to 900 C in inert atmosphere for the electroplated and carburized samples and a diffusion model was developed to predict long term cladding behavior (Task 3). It was found that the Cu cladding was basically insensitive to the high temperature ageing, except for small secondary phases forming at the interface. On the other hand, the Ni electroplated cladding experienced significant interdiffusion. Nevertheless, the gain in corrosion resistance for a 100μm Ni electroplated cladding is phenomenal, with more than 50% reduction in chromium dissolution from the substrate material for the first 15 years of salt exposure at 700°C. To assess radiation resistance and phase stability, high-temperature 4MeV Ni heavy ion irradiation was performed across the cladding/substrate interface up to 50 displacement-per-atom (DPA) at 500°C and 700°C to assess the phase stability and irradiation behaviors of the electroplated systems (Task 4). The Ni and Cu electroplated systems did not experience void swelling at the contrary to the SS316H substrate due to their nanocrystalline nature. Some level of recrystallization was observed in the cladding, as well as radiation induced segregation and enhanced diffusion. The interface acted as a potent sink for point defects with the presence of a void denuded zone. The mechanical properties of the claddings were assessed using thermal shock, micro-indentation, nano-indentation, and four-point bend testing experiments (Task 5). These experiments were performed on the claddings in as-received, corroded, irradiated, and thermally aged states to determine the effects of typical molten salt reactor environments on cladding integrity. No significant mass loss was observed after repeated thermal shocks. While the electroplated samples softened after high-temperature ageing, irradiation hardening compensate this effect, such that there is little different with the as-received materials. The results of these experiments suggest that the electroplated (copper/nickel) show the greatest promise for application in molten salt reactor development. These claddings prevented any chromium dissolution from occurring during the static corrosion experiments. Additionally, they demonstrated favorable properties for high temperature diffusion, phase stability, interfacial mechanical properties, and irradiation resistance. Weld-overlay cladding are also of great interest and could reach properties similar to the electroplated samples upon optimization (especially with multiple weld passes). The carburized SS316H showed excellent behavior as well, but more work needs to be done to assess their long-term stability. Finally, the Mo-laser clad system was not pursued further due to the manufacturing challenges. It is also worth noting that the Ni cladding systems should behave relatively well in terms of weldability since the Ni-weld overlay microstructure and associated corrosion rate are sound. While the NiCu weld-overlay has even lower corrosion rates than the Ni-weld overlay, more studies on welding of Cu-electroplated systems are necessary since Cu clusters are known to embrittle steels.

36 MATERIALS SCIENCE↗

Assessment of Carbonated Brine Injection as Low-risk Strategy for Geologic Carbon Storage

The main objective of this early-phase research was to evaluate the techno-economic feasibility and risk associated with combined brine and CO 2 storage in SWD wells using brine dissolution in the North Dakota portion of the Williston Basin. Three simulation studies were conducted to investigate: (1) CO 2 phases at different conditions, (2) wellbore compatibility, and (3) long-term storage fate in reservoir. (1) A simple reservoir model and injection simulations were created using data to represent the BEST (brine extraction and storage test) site, an operational SWD facility located near Watford City, North Dakota. The pressure evolution caused by CO 2 comingled in produced water injectate in a layer cake reservoir was then modeled while tracking aqueous CO 2 throughout the project. The salinity of the injection water, the salinity of the reservoir brine, and the amount of dissolved CO 2 comingled in the injection water were varied. (2) A wellbore corrosion model was performed using the CO 2 concentrations selected based on the reservoir modeling to examine the carbonated produced water impact on wellbore. (3) Reactive transport modeling was conducted with the optimal CO 2 concentration for this injection site to study the rock-fluid interactions and CO 2 fate in the reservoir. Results suggest that CO 2 dissolved in produced water can be injected without appreciably increasing subsurface pressure or leakage risks. Pressure buildup was found to vary with salinity but not with CO 2 mass fraction. Simulation results show that lower CO 2 percent mass fraction leads to a higher amount of CO 2 that can be dissolved at a higher injection salinity. Furthermore, the long-term goal of dissolution trapping in a traditional carbon storage project is accomplished from the start, mitigating risks associated with potential migration of buoyant CO 2 , so long as the reservoir pressure and temperature are used to determine the maximum mass fraction of the dissolved CO 2 .

54 ENVIRONMENTAL SCIENCES↗