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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

On the dissolution of a borosilicate glass with the use of isotopic tracing – Insights into the mechanism for the long-term dissolution rate

An understanding of the mechanisms responsible for controlling the long-term corrosion rate of nuclear waste glass is paramount if reliable glass dissolution models are to be used to calculate the controlled release of radionuclides from nuclear waste glass under geologic disposal conditions. Additionally, understanding silicate glass dissolution is also important for natural glasses to understand the role from the dissolution of these glasses has on the composition of natural aquatic systems. Two general mechanisms appear to be responsible for the elemental release from glass to the surrounding biosphere - ion exchange (release principally of alkalis) and matrix dissolution (release of the structural components of the glass). Key unknowns related to these mechanisms are the impact of surface layers on the altered glass and how these layers participate in corrosion. To better understand the impact of these layers on glass corrosion, we made two glasses of the same elemental composition, but with selected elements enriched in specific isotopes, e.g. more 10 B than 11 B. We reacted these glasses for 1 y then the solutions from each were exchanged and the chemistry of the solutions and the solids were followed for another 2 y and in some cases 3 y. This allowed us to follow changes in the glasses with respect to these elements and how these elements were transported through the alteration layers with time and, for at least one element, into the glass. Results from these experiments demonstrate that the release behavior of different elements is strongly dependent on their structural role in the glass (e.g. network formers and modifiers), but, more importantly, the role of water transport and subsequent ion exchange has in the long-term dissolution of the glass studied here. In this article, we highlight the behavior of four elements: lithium, sodium, silicon, and boron. Lithium and sodium, network modifying elements, have similar chemistries. The behavior of Li could be tracked in more detail by following both 6 Li and 7 Li; Li penetrates through the gel layer in both directions without hinderance and into and released from the glass deeper than sodium. Silicon, a network forming element, reacts with the silica-rich alteration layer. Boron, a network forming element, does not accumulate in the gel or the pristine glass and has a very sharp elemental profile between the pristine glass and alteration layer. Boron, lithium, and sodium elemental profiles suggest that there is little if any transport control within the alteration layer. The correlation between the Li and Na profiles and that of a water species suggest that the limiting release is a steady state between the diffusion of a water species and matrix dissolution that results in a low release rate for alkali and this provides a steady driving force for the long-term dissolution of glass.

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

Modulation of Uptake and Reactivity of Nitrogen Dioxide in Metal-Organic Framework Materials

We report the modulation of reactivity of nitrogen dioxide (NO 2 ) in a charged metal–organic framework (MOF) material, MFM-305-CH 3 in which unbound N-centres are methylated and the cationic charge counter-balanced by Cl - ions in the pores. Uptake of NO 2 into MFM-305-CH 3 leads to reaction between NO 2 and Cl - to give nitrosyl chloride (NOCl) and NO 3 - anions. A high dynamic uptake of 6.58 mmol g -1 at 298 K is observed for MFM-305-CH 3 as measured using a flow of 500 ppm NO 2 in He. In contrast, the analogous neutral material, MFM-305, shows a much lower uptake of 2.38 mmol g -1 . The binding domains and reactivity of adsorbed NO 2 molecules within MFM-305-CH 3 and MFM-305 have been probed using in situ synchrotron X-ray diffraction, inelastic neutron scattering and by electron paramagnetic resonance, high-field solid-state nuclear magnetic resonance and UV/Vis spectroscopies. The design of charged porous sorbents provides a new platform to control the reactivity of corrosive air pollutants.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

High-temperature chromium diffusion in austenitic stainless steel: Ab initio molecular dynamics simulations

Chromium self-diffusion through stainless steel (SS) matrix and along grain boundaries is an important mechanism controlling SS structural materials corrosion. For this work, Cr diffusion in austenitic SS was simulated using canonical ab initio molecular dynamics with realistic models of type-316 SS bulk, with and without Cr vacancies, and a low-energy Σ3 twin boundary typically observed at active corrosion sites. Cr self-diffusion coefficients at 750 and 850 °C calculated using Einstein’s diffusion equation are 4.2 × 10 –6 and 8.1 × 10 –6 Å 2 ps –1 in pristine bulk, 3.8 × 10 –3 and 5.5 × 10 –3 Å 2 ps –1 in bulk including Cr vacancies, and 9.5 × 10 –2 and 1.0 × 10 –1 Å 2 ps –1 at a Σ3[1 1 1]60° twin boundary.

36 MATERIALS SCIENCE↗
DOE OSTI2023

In-situ measurement and control of the tool-workpiece interface temperature during friction stir processing of 304/304L stainless steel

Friction stir processing of 304/304L stainless steel is an area of interest for repair of spent nuclear fuel canisters. To enable repair without canister deflection or fracture, it is necessary to define the limitations for the depth of the processing zone which will be critically dependent on the thermal history of the material. Here, this work provides the first direct measurement of the interface temperature for friction stir processing of steel. The measured temperature is in the range of 850–1050 °C which is 20–30 % higher than previously reported values. Notably, the peak temperature occurs at the pin tip which is attributed to a higher sticking fraction and subsequently higher strain rate near the tool axis of rotation. Furthermore, a clear correlation between the temperature, grain size, and hardness is established within a single weld nugget and across different processing conditions. In the future, this data may lead to improved process modelling through direct validation and allow for property optimization through direct microstructure control.

304 stainless steel↗
DOE OSTI2023

Application of NEAMS Multiphysics Framework for Species Tracking in Molten Salt Reactors

This report from Idaho National Laboratory (INL) summarizes the key modeling and simulation activities conducted under the Department of Energy (DOE) Molten Salt Reactor (MSR) Campaign during the Fiscal Year 2023 (FY23). The focus of the work was to leverage state-of-the-art modeling capabilities from the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) codes to enable novel multiphysics and multiscale modeling and simulation of MSRs. Through collaboration with NEAMS code developers, advanced multiphysics analysis capabilities for MSR systems were demonstrated by coupling depletion, thermal-hydraulics, and thermochemistry into an innovative framework for chemical species transport in MSRs. As a result, the framework can track nuclides throughout their lifetimes in the core, from production (depletion) to advection throughout the salt volume (thermal-hydraulics) and off-gassing or precipitation outside of the salt (thermochemistry). This work supports the near-term deployment of MSRs by integrating the synergistic efforts between the DOE’s MSR Campaign and NEAMS program. The resulting framework will help better connect system design modelers with experimentalists to better understand and predict complex physical behaviors in MSRs. Researchers and MSR developers alike can now leverage these new modeling and simulation capabilities to perform novel analyses with applications including: • MSR dynamics during normal operational transients and accident scenarios • Off-gas system design and performance for fuel cycle and depletion analysis • Corrosion and active chemistry control for reactor component health and lifetime determination • Source term, decay heat and activity determination in accident scenarios • Special nuclear material accountancy and chemical forensic analysis for safeguards • Digital twin development of experiments and experimental reactor demonstrations • Measurement requirements for instrumentation and control design • Uncertainty and sensitivity analysis of missing data to inform future experimental data collection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1999

Design Through Simulation of a Molecular Sieve Column for Treatment of MON-3

The presence of water in propellant-grade MON-3 is a concern in the Aerospace Industry. NASA Johnson Space Center (JSC), White Sands Test Facility (WSTF) Propulsion Department has evaluated many types of molecular sieves for control of iron, the corrosion product of water in Mixed Oxides of Nitrogen (MON-3). In 1995, WSTF initiated laboratory and pilot-scale testing of molecular sieve type 3A for removal of water and iron. These tests showed sufficient promise that a series of continuous recycle tests were conducted at WSTF. Periodic samples of the circulating MON-3 solution were analyzed for water (wt %) and iron (ppm, wt). This test column was modeled as a series of transfer units; i. e., each unit represented the height equivalent of a theoretical plate. Such a model assumes there is equilibrium between the adsorbent material and the effluent stream from the unit. Operational and design parameters were derived based on the simulation results. These parameters were used to predict the design characteristics of a proposed molecular sieve column for removal of water and iron from MON-3 at the NASA Kennedy Space Center (KSC). In addition, these parameters were used to simulate a small, single-pass operation column at KSC currently used for treating MON-3. The results of this work indicated that molecular sieve type 3A in 1/16 in. diameter pellets, in a column 2.5 ft. in diameter, 18 ft. in height, and operated at 25 gpm is adequate for the required removal of water and iron from MON-3.

Swartz, A. Ben↗
NASA NTRS2014

Raising Nuclear Thermal Propulsion (NTP) Technology Readiness Above 3

NTP development is currently supported by the NASA program office "Advanced Exploration Systems". The concept is a main propulsion option being considered for human missions to Mars in the 2030's. Major NTP development took place in the 1960's and 1970's under the Rover/NERVA program. The technology had matured to TRL 6 and was preparing to go to TRL 7 with a prototype flight engine before the program was cancelled. Over the last 40 years, a variety of continuations started, but only lasted a few years each. The Rover/NERVA infrastructure is almost all gone. The only remains are a few pieces of hardware, final reports and a few who worked the Rover/NERVA. Two types of nuclear fuel are being investigated to meet the current engine design specific impulse of 900 seconds compared to approximately 850 seconds demonstrated during Rover/NERVA. One is a continuation of composite fuel with new coatings to better control mid-band corrosion. The other type is a CERMET fuel made of Tungsten and UO2. Both fuels are being made from Rover/NERVA lessons learned, but with slightly different recipes to increase fuel endurance at higher operating temperatures. The technology readiness level (TRL) of these current modified reactor fuels is approximately TRL 3. To keep the development cost low and help mature the TRL level past 4 quickly, a few special non-nuclear test facilities have been made to test surrogate fuel, with depleted uranium, as coupons and full length elements. Both facilities utilize inductive heating and are licensed to handle depleted uranium. TRL 5 requires exposing the fuel to a nuclear environment and TRL 6 requires a prototype ground or flight engine system test. Currently, three different NTP ground test facility options are being investigated: exhaust scrubber, bore hole, and total exhaust containment. In parallel, a prototype flight demonstration test is also being studied. The first human mission to Mars in the 2030's is currently 2033. For an advanced propulsion concept to be seriously considered for use, the engine development plans need to show it is feasible and affordable to reach TRL 8 by 2027 and can be qualified for human mission use.

Gerrish, Harold P., Jr.↗
DOE OSTI2020

Coupled Thermal-Hydraulic Analysis and Species Mass Transport in a Versatile Experimental Salt Irradiation Loop (VESIL) using COBRA-TF

With the resurgence of interest in Molten Salt Reactor (MSR) technology among commercial and governmental agencies, there is a need for new experiments and new modelling capabilities to characterize the unique phenomena present in this fluid fuel system. A Versatile Experimental Salt Irradiation Loop (VESIL) is currently under investigation at Idaho National Laboratory. It is designed to be a dynamic fuel-bearing salt loop placed in the Advanced Test Reactor. One of the key phenomena this proposed experiment plans to elucidate is fission product speciation in the fuel-salt and the subsequent effects this can have on fuel-salt properties, fission product transport and interactions, source term generation, corrosion and redox potential control. Specifically, noble gases (Xe & Kr) will bubble out to a plenum or off-gas system and noble metals (Mo, Tc, Te etc.) will precipitate and deposit in specific zones in the loop. This work extends the mass transfer and species interaction models in the general species transport capability of COBRA-TF (CTF) and applies these models to give a preliminary estimation of fission product behavior in the proposed VESIL design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

The corrosion effects of neutron activation of 2LiF-BeF 2 (FLiBe)

The redox effect of activation reactions in molten salts has been previously qualitatively defined but never quantified. Here, we quantify the corrosion effects of neutron activation of FLiBe. Rates of change of activation products are computed for FHR and ARC. The impact of activation reactions on salt chemistry is quantitatively compared to other environmental effects of similar consequences on the salt. The results are applied to a redox control case-study and an irradiation-corrosion experiment. The results are presented in a manner useful to the reactor chemist and designer concerned with corrosion and chemistry control and tritium management in irradiated FLiBe, and design of salt-irradiation experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Rapid synthesis of phase-engineered tungsten carbide electrocatalysts via flash joule heating for high-current-density hydrogen evolution

Fabricating durable and high-performance electrocatalysts operating at high current densities for industrial acidic hydrogen evolution remains a daunting challenge. Tailoring the phase composition of electrocatalysts is a promising strategy to harness synergistic effects and improve charge transfer, thereby optimizing their performance. Here, this work presents a fast, green method based on flash joule heating (FJH) to synthesize phase-engineered tungsten carbide electrocatalysts for the acidic hydrogen evolution reaction (HER) at high current densities. Tungsten carbide electrodes with varying FJH treatment durations (3, 10, 30, and 60 s) are fabricated to fine-tune the mixture of tungsten monocarbide (WC) and tungsten semicarbide (W 2 C) phases. Results show that samples with a 30-s treatment (TC-3) exhibit an optimal balance between these phases, leading to a low overpotential of 180.97 and 387 mV at current densities of 10 mA/cm 2 and 4 A/cm 2 , respectively. TC-3 exhibits significantly lower charge transfer resistance compared to the other electrocatalysts, which can be attributed to its optimal phase ratio. Notably, the TC-3 electrocatalyst remains stable for over 9 days at 4 A/cm 2 due to their controlled phases and excellent corrosion-resistant properties. This work highlights a new method to fabricate cost-effective, high-performance tungsten carbide electrocatalysts with well-controlled phase compositions.

Acidic hydrogen evolution reaction↗
DOE OSTI2021

Progress Report on Performance of A709 and G91 Steels in Sodium

Specimens in six different processing and heat treatment conditions of A709 H58776 were exposed to sodium at 550, 600, and 650°C, respectively for various exposure times. G91 base metal was tested in sodium at 550, 600, and 650°C, and G91 weldment at 550 and 600°C, respectively. Parallel thermal aging experiments were conducted on G91 and A709 to obtain thermal aging data for comparison with sodium exposure results to separate the thermal and sodium effects. The corrosion data obtained on A709 steel continue to show low corrosion rates, which indicate good compatibility of A709 with sodium when oxygen content is controlled. G91 has also shown acceptable corrosion rates over the temperature range of 550-650°C in sodium. Thermal aging or sodium exposures reduced the tensile strength, uniform elongation, and total elongation for A709 ESR, AOD specimens. These specimens consistently showed higher yield stress and tensile strength and lower uniform and total elongations after sodium exposure than after thermal aging under comparable testing conditions. It implies that there was an additional effect of sodium exposure that gave rise to an increase in tensile strength of A709. Dynamic strain aginginduced flow serrations were completely removed by either thermal aging or sodium exposure at 650°C, which imply that carbon or nitrogen in solution was removed from the solution and formed precipitates during aging or sodium exposures. The tensile data ruled out the possibility of decarburization under the sodium exposure conditions. The HOMO specimens behaved somewhat differently from the ESR and AOD specimens in sodium. While thermal aging and sodium exposures at 550 and 600°C decreased the yield stress and the ultimate tensile strength of G91, there was virtually no additional effect resulting from sodium exposures at these two temperatures. In contrast, sodium exposures at 650°C had a drastic effect on the yield stress and the ultimate tensile strength of G91. It is suggested that G91 experienced decarburization in the 650°C sodium environments. Carbon concentrations in sodium in the SMT-1 and SMT-2 loops were determined by a foil equilibration method. The estimated carbon concentration was in the range of 0.8-1.2 ppm in the SMT-1 loop and 0.3-0.7 ppm in the SMT-2 loop. The carbon activity in sodium in the SMT-2 loop was estimated to be 0.03-0.08 at 600°C, and 0.08-0.2 at 550°C. The carbon activity in sodium at 650°C in the SMT-1 loop was 0.04-0.07. Equilibrium simulation of the carburization – decarburization processes was conducted for A709 and G91 steels exposed in sodium environments at temperatures of 550-700°C. The carbon activity-concentration relationship for G91 was re-evaluated by considering four phases in G91, i.e. bcc ferrite, M 23 C 6 , NbC and VC carbides. It was found that the carburization-decarburization process in G91 steel was dictated by M 23 C 6 carbides at high carbon activities, while NbC carbides dominated the process at low carbon activities. Formation of NbC increases the decarburization resistance of G91 steel. It remains to be understood whether the decarburization resistance provided by MC carbides in G91 can be maintained during long-term operations of SFRs. The carbon concentration-activity relationship for A709 was calculated based on the equilibrium of fcc-austenite and M 23 C 6 carbide phase. The calculations showed that A709 decarburizes at 650°C in the SMT-1 loop environments, which is different from the experimental findings. The carbon concentration-activity relationship in A709 was further evaluated by considering four phases, including fcc-austenite, M 23 C 6 , NbC, and TiC carbide phases. TiC is the most stable carbides among the three carbide phases. The carburization-decarburization process in A709 is dictated by M 23 C 6 carbides at high carbon activities and by TiC at low carbon activities. Formation of TiC increases the decarburization resistance of A709 steel. Because of the complexity of the precipitation process in A709 during sodium exposures, detailed characterization of precipitates in sodium-exposed specimens and improved thermodynamic models are needed to understand the carburization-decarburization behavior of A709. Kinetic analysis of carbon transfer will be investigated for G91 and A709 in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Hydro-chemo-mechanical phase field formulation for corrosion induced cracking in reinforced concrete

Steel corrosion induced cracking in reinforced concrete structures is often caused by volumetric straining at the steel-concrete interface. This volumetric strain is mainly controlled by the rate of generation of corrosion product, which depends on the diffusion of Cl{sup −} and dissolved O{sub 2} as well as the chemical reaction at the steel-concrete interface. While many numerical modeling techniques have been proposed to tackle this complex phenomenon, most have been limited to using staggered solution schemes, where the diffusion, chemical reaction, mechanical strain, and crack fields are solved for separately. Unfortunately, such approaches introduce numerical errors that can lead to unrealistic predictions. Therefore, a coupled solution, is needed to remove the formulation induced errors and understand this complex phenomenon. To provide a unifying solution framework for corrosion induced cracking in reinforced concrete, a phase field formulation is presented and implemented with FEM algorithms. The governing equations are derived from the system's free energy and Fick's law of diffusion. The proposed formulation was first validated in 1D by comparing the modeling results with the semi-analytical solution approximated via Laurent series to avoid singularities in the complex domain. 2D plane strain solutions to the embedded rebar undergoing corrosion induced cracking with pre-cracks are also presented and compared with the approximated solutions from linear elastic fracture mechanics theory (LEFM). The modeling of crack propagation enabled by the proposed formulation is also presented along with parametric studies to reveal the roles of the fracture toughness and permeability of the concrete on the corrosion induced cracking. A case study was also conducted on H-pile steel with concrete jackets to demonstrate the feasibility of modeling corrosion induced fracture in reinforced concrete structures with complex steel-concrete interfaces.

36 MATERIALS SCIENCE↗
DOE OSTI2023

FY23 Status Report: SNF Interim Storage Canister Corrosion and Surface Environment Investigations

Work evaluating spent nuclear fuel (SNF) dry storage canister surface environments and canister corrosion progressed significantly in FY23, with the goal of developing a scientific understanding of the processes controlling initiation and growth of stress corrosion cracking (SCC) cracks in stainless steel canisters in relevant storage environments. The results of the work performed at Sandia National Laboratories (SNL) will guide future work and will contribute to the development of better tools for predicting potential canister penetration by SCC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2011

External Tank Program - Legacy of Success

The largest single element of Space Shuttle is the External Tank (ET), which serves as the structural backbone of the vehicle during ascent and provides liquid propellants to the Orbiter s three Main Engines. The ET absorbs most of the seven million pounds of thrust exerted by the Solid Rocket Boosters and Main Engines. The design evolved through several block changes, reducing weight each time. Because the tank flies to orbital velocity with the Space Shuttle Orbiter, minimization of weight is mandatory, to maximize payload performance. The initial configuration, the standard weight tank, weighed 76,000 pounds and was an aluminum 2219 structure. The light weight tank weighed 66,000 pounds and flew 86 missions. The super light weight tank weighed 58,500 pounds and was primarily an aluminum-lithium structure. The final configuration and low weight enabled system level performance sufficient for assembly of the International Space Station in a high inclination orbit, vital for international cooperation. Another significant challenge was the minimization of ice formation on the cryogenic tanks. This was essential due to the system configuration and the choice of ceramic thermal protection system materials on the Orbiter. Ice would have been a major debris hazard. Spray on foam insulation materials served multiple functions including thermal insulation, conditioning of cryogenic propellants, and thermal protection for the tank structure during ascent and entry. The tank is large, and unique manufacturing facilities, tooling, and handling, and transportation operations were developed. Weld processes and tooling evolved with the design as it matured through several block changes. Non Destructive Evaluation methods were used to assure integrity of welds and thermal protection system materials. The aluminum-lithium alloy was used near the end of the program and weld processes and weld repair techniques had to be refined. Development and implementation of friction stir welding was a substantial technology development incorporated during the Program. Automated thermal protection system application processes were developed for the majority of the tank surface. Material obsolescence was an issue throughout the multi-decade program. Process controls were implemented to assure cleanliness in the production environment, to control contaminants, and to preclude corrosion. Each tank was accepted via rigorous inspections, including non-destructive evaluation techniques, proof testing, and all systems testing. In the post STS-107 era, the project focused on ascent debris risk reduction. This was accomplished via stringent process controls, post flight assessment using substantially improved imagery, and selective redesigns. These efforts were supported with a number of test programs to simulate combined environments. The debris risk was reduced by two orders of magnitude. During this time a major natural disaster was overcome when hurricane Katrina damaged the manufacturing facility. Numerous lessons from these efforts, the manufacturing and material processing issues, the key design features, and evolution of the design will be discussed.

Pilet, Jeffery C.↗
NASA NTRS2011

External Tank - The Structure Backbone

The External Tank forms the structural backbone of the Space Shuttle in the launch configuration. Because the tank flies to orbital velocity with the Space Shuttle Orbiter, minimization of weight is mandatory, to maximize payload performance. Choice of lightweight materials both for structure and thermal conditioning was necessary. The tank is large, and unique manufacturing facilities, tooling, handling, and transportation operations were required. Weld processes and tooling evolved with the design as it matured through several block changes, to reduce weight. Non Destructive Evaluation methods were used to assure integrity of welds and thermal protection system materials. The aluminum-lithium alloy was used near the end of the program and weld processes and weld repair techniques had to be refined. Development and implementation of friction stir welding was a substantial technology development incorporated during the Program. Automated thermal protection system application processes were developed for the majority of the tank surface. Material obsolescence was an issue throughout the 40 year program. The final configuration and tank weight enabled international space station assembly in a high inclination orbit allowing international cooperation with the Russian Federal Space Agency. Numerous process controls were implemented to assure product quality, and innovative proof testing was accomplished prior to delivery. Process controls were implemented to assure cleanliness in the production environment, to control contaminants, and to preclude corrosion. Each tank was accepted via rigorous inspections, including non-destructive evaluation techniques, proof testing, and all systems testing. In the post STS-107 era, the project focused on ascent debris risk reduction. This was accomplished via stringent process controls, post flight assessment using substantially improved imagery, and selective redesigns. These efforts were supported with a number of test programs to simulate combined environments. Processing improvements included development and use of low spray guns for foam application, additional human factors considerations for production, use of high fidelity mockups during hardware processing with video review, improved tank access, extensive use of non destructive evaluation, and producibility enhancements. Design improvements included redesigned bipod fittings, a bellows heater, a feedline camera active during ascent flight, removal of the protuberance airload ramps, redesigned ice frost ramps, and titanium brackets replaced aluminum brackets on the liquid oxygen feedline. Post flight assessment improved due to significant addition of imagery assets, greatly improving situational awareness. The debris risk was reduced by two orders of magnitude. During this time a major natural disaster was overcome when Katrina damaged the manufacturing facility. Numerous lessons from these efforts are documented within the paper.

Welzyn, Kenneth↗
NASA NTRS1985

Hot corrosion of sintered alpha-SiC at 1000 C

The hot corrosion of sintered alpha-SiC by thin films of Na2SO4 and Na2CO3 was studied at 1000 C in controlled gas atmospheres. Under all conditions, corrosion led to 10 to 20 times the amount of SiO2 formed in pure oxidation after a 48-h exposure. In addition, small amounts of sodium silicate formed. Melts of Na2SO4/SO3 caused uniform pitting of the SiC substrate; Na2CO3/CO2 melts caused localized pitting and grain-boundary attack. In all cases, the protective SiO2 layer dissolved to form silicate, leading to corrosion. In the sulfate case, free carbon in the SiC promotes this process. In all cases the presence of liquid films is responsible for rapid transport rates and the subsequent rapid reaction.

Jacobson, N. S.↗
DOE OSTI2024

Feasibility Study for Cathodic Protection of Waste Tanks at Hanford

The Hanford Nuclear Reservation site contains approximately 211 million liters of radioactive and chemically hazardous waste arising from nuclear weapons production, beginning with World War II, and continuing through the Cold War. The waste is stored in 177 carbon-steel underground storage tanks, of which 149 are single-shell tanks (SSTs) and the remaining are double-shell tanks (DSTs). The mission of an ongoing River Protection Project is to retrieve the waste from the underground storage tanks and then treat and immobilize (i.e., vitrify) it for disposal. Waste from the older SSTs is being progressively retrieved into the newer DSTs for storage pending treatment, immobilization, and disposal. Waste chemistry controls are in place to mitigate general corrosion, pitting corrosion, and stress corrosion cracking (SCC) and minimize corrosion risks to the carbon steel tank liners of DST at the Hanford Site. The chemistry control program focuses on preserving liner integrity while balancing the need for caustic/nitrite additions that have large downstream impacts on vitrification. Some of the DSTs store salt cake or sludge waste layers contain out-of-specification interstitial liquids that contact the primary tank bottom or sidewall. Remediation of such layers through chemical additions is not practical. At the present time, the most significant leak integrity concern at Hanford is for corrosion of the tank bottoms of DSTs where the solids layer is out of specification because of hydroxide ion depletion and consequently, pH values that are less than 12. Cathodic protection (CP) has been proposed as a possible way to protect the liners of such tanks against corrosion. The objective of this project was to investigate the feasibility of applying CP to Hanford DSTs to mitigate internal corrosion of the carbon steel tank primary liners. The scope of the project included (1) a literature review on the application of CP to storage tank systems, in general, and to waste tanks at the Savannah River Site (SRS) and Hanford, (2) a review of the finite element analysis (FEA) model recently developed to evaluate potentials for a DST, and (3) the development of a boundary element analysis (BEA) model for CP for a DST. Tank AN-107, one of the Hanford DSTs, was selected for this work because (1) the bottommost waste layer in the tank is out-of-specification, and (2) electrochemical and resistivity data, based on recent waste sampling and testing, are available for this tank. The overarching conclusion of this study is that, based on the available information, it is feasible to apply CP to mitigate internal corrosion of the primary liners of the Hanford DSTs. However, significant additional information and work are required before a CP system can be designed. Recommendations for next steps in the design of the CP system will be presented.

WIERSMA, BRUCE↗