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DOE OSTI2024

Kinetics of the reaction of ferrous ions with hydroxyl radicals in the temperature range 25–300 °C

Here, the kinetics and mechanism of the reaction between OH radicals and ferrous ions in the temperature range 25–300 °C were studied using pulse radiolysis. At temperatures <150 °C the rate of reaction is essentially independent of temperature, while at temperatures >150 °C the activation energy is 45.8 ± 3.0 kJ mol —1 . The change in activation energy is attributed to a change in the dominant mechanism from hydrogen atom transfer (HAT) to dissociative ligand interchange. The kinetic isotope effect (KIE) was measured by repeating experiments in heavy water. A value of 2.9 was measured at room temperature where HAT is the dominant mechanism. The KIE decreases to zero at temperatures > 150 °C as ligand interchange becomes dominant and the O–H bond is no longer involved in the reaction.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2025

Transient methods for understanding the properties of strongly oxidizing radicals

This review discusses the properties of strongly oxidizing radicals in organic and aqueous media and highlights the challenges in obtaining accurate values of their reduction potentials. Transient redox equilibrium methods based on the use of strong photooxidants or initiated by pulse radiolysis are shown to provide versatile approaches for decoupling electron transfer reactions from follow-up reactivity of unstable radical species, resulting in accurate values of reduction potentials of very positive couples, including some solvent radical cations. Here, we also show that correlations of reduction potentials with Hammett Σσ + p parameters, as well as gas phase ionization potentials, can be used to estimate the redox properties of unknown couples within a homologous series of compounds. The effects of ion pairing and hemicolligation on redox properties of organic and inorganic radicals are also discussed.

14 SOLAR ENERGY↗

DOE OSTI2025

Formation of U(VI) peroxide nanoclusters from cascade reactions with a persulfate radical initiator

Radiolysis of water in high radiation fields generates a variety of reactive oxygen species that influence the chemical behavior and complexation of hexavalent uranium. This study investigates the behavior of interaction of a uranyl cation (UO 2 2+ (VI)) with a series of free radicals that are formed in situ via activation of the free radical initiator persulphate (S 2 O 8 2− ), which releases both SO 4 ˙ − and ˙OH species in the solution. Electron Paramagnetic Resonance (EPR) and Raman spectroscopy were used to evaluate the presence of the hydroperoxyl radical (HO 2 ˙) and superoxide radicals (O 2 ˙ − ) that are formed within the solution through radical cascade reactions. In addition, a uranyl peroxide cluster solid (NaU 24 ) was crystallized and characterized using single crystal X-ray diffraction (SCXRD), vibrational spectroscopy, and EPR spectroscopy. The presence of the hydroperoxyl radical (HO 2 ˙) and superoxide radicals (O 2 ˙ − ) was also observed in the solid-state compound, but spectroscopic evidence suggests that it was associated with the Na + network and not the cluster itself. Density functional theory (DFT) calculations were also utilized to further confirm the radical species produced and determine the potential stabilization of radicals detected within the cluster and lattice.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Mechanistic roles of metal- and ligand-protonated species in hydrogen evolution with [Cp*Rh] complexes

Protonation reactions involving organometallic complexes are ubiquitous in redox chemistry and often result in the generation of reactive metal hydrides. However, some organometallic species supported by η 5 -pentamethylcyclopentadienyl (Cp*) ligands have recently been shown to undergo ligand-centered protonation by direct proton transfer from acids or tautomerization of metal hydrides, resulting in the generation of complexes bearing the uncommon η 4 -pentamethylcyclopentadiene (Cp*H) ligand. Here, time-resolved pulse radiolysis (PR) and stopped-flow spectroscopic studies have been applied to examine the kinetics and atomistic details involved in the elementary electron- and proton-transfer steps leading to complexes ligated by Cp*H, using Cp*Rh(bpy) as a molecular model (where bpy is 2,2′-bipyridyl). Stopped-flow measurements coupled with infrared and UV-visible detection reveal that the sole product of initial protonation of Cp*Rh(bpy) is [Cp*Rh(H)(bpy)] + , an elusive hydride complex that has been spectroscopically and kinetically characterized here. Tautomerization of the hydride leads to the clean formation of [(Cp*H)Rh(bpy)] + . Variable-temperature and isotopic labeling experiments further confirm this assignment, providing experimental activation parameters and mechanistic insight into metal-mediated hydride-to-proton tautomerism. Spectroscopic monitoring of the second proton transfer event reveals that both the hydride and related Cp*H complex can be involved in further reactivity, showing that [(Cp*H)Rh] is not necessarily an off-cycle intermediate, but, instead, depending on the strength of the acid used to drive catalysis, an active participant in hydrogen evolution. Identification of the mechanistic roles of the protonated intermediates in the catalysis studied here could inform design of optimized catalytic systems supported by noninnocent cyclopentadienyl-type ligands.

08 HYDROGEN↗

DOE OSTI2021

New high-throughput endstation to accelerate the experimental optimization pipeline for synchrotron X-ray footprinting

Synchrotron X-ray footprinting (XF) is a growing structural biology technique that leverages radiation-induced chemical modifications via X-ray radiolysis of water to produce hydroxyl radicals that probe changes in macromolecular structure and dynamics in solution states of interest. The X-ray Footprinting of Biological Materials (XFP) beamline at the National Synchrotron Light Source II provides the structural biology community with access to instrumentation and expert support in the XF method, and is also a platform for development of new technological capabilities in this field. Hee, the design and implementation of a new high-throughput endstation device based around use of a 96-well PCR plate form factor and supporting diagnostic instrumentation for synchrotron XF is described. This development enables a pipeline for rapid comprehensive screening of the influence of sample chemistry on hydroxyl radical dose using a convenient fluorescent assay, illustrated here with a study of 26 organic compounds. The new high-throughput endstation device and sample evaluation pipeline now available at the XFP beamline provide the worldwide structural biology community with a robust resource for carrying out well optimized synchrotron XF studies of challenging biological systems with complex sample compositions.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2021

Measurement of radiolytic hydrogen generation and impact of drying treatments on reactor exposed and surrogate aluminum materials

Technical challenges associated with dry storage of Aluminum-clad Spent Nuclear Fuel (ASNF) include a need to better understand the potential extent of gaseous molecular hydrogen (H 2 ) production through radiolytic degradation of the aluminum (oxy)hydroxide films present on ASNF. The characterization of radiolytic gas generation from ASNF (oxy)hydroxide layers has been identified as a key knowledge gap which poses a technical challenge to the long-term storage of ASNF. Task 2 of the action plan for the extended (>50 years) dry storage of ASNF addresses this gap. Previous radiolysis studies performed under Task 2 established baseline estimates of H 2 generation rates from the attendant hydrated oxides. The results confirmed that net radiolytic H 2 production has a dependency on absorbed gamma dose, as well as relative humidity and cover gas composition (air, nitrogen, and argon). Further experiments revealed that the physisorbed water on the samples may significantly impact the radiolytic H 2 yield. This phenomenon complicates the determination of H 2 generation rates for hydrated oxides, which may lead to inaccurate modeling predictions of the long-term H 2 yields in sealed storage systems containing ASNF, particularly when compared to spent fuel casks which have undergone some drying process intended to remove physically- and chemically-bound water. This report describes the testing methods utilized and the hydrogen generation results obtained in an investigation of the effects of gamma irradiation on aluminum materials for a variety of (oxy)hydroxide surface compositions and drying conditions. The testing methods included small-area aluminum material testing in ampules, and large-area aluminum material testing in steel vessels. Test material preparation, irradiation, and radiolytic H 2 measurement methods are summarized. The measured H 2 concentrations, which reflect the variable initial hydrated inventory and drying treatments, are compared to one another as well as to previously published data, to identify the primary factors affecting radiolytic H 2 generation rates and potential equilibrium H 2 concentrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Modeling of ATR fuel in DOE Standard Canisters with Helium Backfilled Condition

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel dictates storage within helium backfilled sealed DOE standard canisters. These sealed canisters are intended for extened (>50 year) dry storage). The typical packaging configuration for the 15-foot DOE canisters places 10 ATR elements within a Type 1a basket, and three baskets are loaded within each DOE canister. During in-reactor operations and cooling pond storage conditions, oxyhydroxide layers form on the surface of the aluminum clad fuel. These layers produce hydrogen gas over time due to the fuel’s radiation field. As part of the packing procedure, the ATR fuel should be dried to remove any residual physio-/chemi- sorbed water from the surface. As testing to the effectiveness of the drying procedure is still underway, this modeling will include results at fully saturated and fully dried conditions. In previous modeling efforts, the G-value for the production of hydrogen from the oxyhydroxide layers was assumed to be in argon environments as measured by Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution. Previous experimental testing showed differences in the hydrogen generated based on the gaseous environment. In the associated experimental work, Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution, additional tests were completed in a helium environment, and updated G-values for the radiolytic production of hydrogen from the oxyhydroxide layers were provided. These values are 28% and 58% higher than values for argon. In addition, a change to the modeling of the oxyhydroxide radiolysis has been made from previous reports. This change assumes the dependency of the dose rate on the overall reaction rate is applied to the total weight of the sample, rather than to just the weight of the oxyhydroxide layer. This decreases the dependency of the radiolytic reaction on the thickness of the oxyhydroxide layer. For a nominal scenario of stored ATR fuel, assuming the chemi-/physio- sorbed water have been fully removed, the internal canister pressure increases to 1.61 atm over a 50 year period, with a hydrogen mole percentage of 21%. As in previous modeling, any oxygen present is in negligible amounts (<1 ppt). If a small amount residual air is present, nitric acid can form up to 1300 ppm. For a scenario with high fuel decay heat, the model shows internal pressure increasing to 2.1 atm, with 39.3 mole percentage of hydrogen. In a scenario where significant chemi-/physio- sorbed water is present within the corrosion layer, the nominal scenario shows a pressure increase to 2.54 atm, with 21.1 mole percent hydrogen. The high decay heat case shows a pressure increase to 3.18 atm with 39.9 mole percent hydrogen.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Extended Modeling of DOE Sealed Canisters with Updated Chemistry Models

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. These models have coupled the temperature conditions to both the gas phase radiolysis chemistry and the surface chemistry associated with the oxyhydroxide layer. The previous iteration of the model utilized constant G-values for the hydrogen generation a 50-year period. This new iteration of the model utilizes new experimental data to update the hydrogen generation rate, as well as increase the simulated time to a 200-year period. Given the half-life assumed for the primary Cs-137 isotope responsible for gamma radiation in the ATR spent fuel, the 200-year period decreases the decay heat and dose rate of the spent fuel by a factor of 100, which combined with the updated chemistry substantially decreases new hydrogen generation. Continued experimental work has identified trends for aluminum surrogate samples with oxyhydroxide layers for tests done at higher dose rates. At low initial doses a fast generation rate of hydrogen occurs which starts to roll over to a lower generation as the total dose applied increase. A small-scale chemical model was built to replicate as mini-canister surrogate system at SRNL as well as for the smaller capsule tests performed at INL. A variety of chemical models to capture this effect were tested, and a back reaction of H radical absorbing onto the surface, or inhibition of the reaction by significant H 2 cover gas were both able to fit both the mini-canister and the small capsule test data. Both kinetic fits were used to generate data from the new 200-year simulation. As additional experiments continue, the kinetic fits may be adjusted to adapt to new data. However, updated data shows that hydrogen atmosphere has little effect on actual generation data, so the model was reverted to use a star-stepped G-value for low-dose and high-dose regions. For the undried fuel case, the three models differ significantly with an end concentration of 1% for back reaction, 4.2% for inhibition, and 13.1% for constant G-value for the nominal scenario. For the dried fuel case, the nominal cases showed end concentrations of 0.23% for back reaction, 1.9% for inhibition, and 4.2% for const G-value for the nominal scenario. For the constant G-value case that is less conservative than the other two, the pressure for undried fuel increases to 1.94 atm over 200 years for the nominal case and 2.11 atm for the high decay heat case. The total hydrogen concentration after 200 years is 4.92% for the low decay heat case and 27.4% for the high decay heat case for undried fuel and is 1.4% and 9.6% for low and high decay heat fuel for the dried fuel case. If small amount of residual air is present, the potential for nitric acid formation of 595, 1568, and 2816 ppm for the lower, nominal, and upper fuel decay heat can occur. At long-term timeframes no significant shift in major species present occurs, so no appreciable amount of oxygen is present in the system. The continued rate of hydrogen generation in the canister occurs at an increase of 0.04% by mole over the final 10 years. Increasing the model range out to 1000 years continues to drop the hydrogen generation rate through decreasing dose rate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Modeling DOE Standard Canister Configurations with Updated Surface Chemistry

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Considerations for Defining G-Values for Aluminum-Clad Spent Nuclear Fuel

Sealed-canister dry storage of aluminum-clad spent nuclear fuel (ASNF) generated by research reactors is an alternative to current storage and disposition pathways as directed by the U.S. Department of Energy. The major challenge faced for this storage approach is radiolytic H 2 generation, including from the aluminum (oxy)hydroxide layers on the surface of ASNF. Experimental and modeling activities have been carried out to characterize the radiolytic yield as part of a DOE-sponsored research program to develop the technical basis for ASNF dry storage. The G-value is a commonly way to report results of radiolysis testing and is defined as the radiolytic yield of a species (e.g. molecular hydrogen) per unit radiation energy deposited into the material system. An independent technical review of the ASNF dry storage technical basis performed by Pacific Northwest National Laboratory raised questions about differences in G-value definitions used for experiments on ASNF surrogates consisting of aluminum samples with adherent (oxy)hydroxides compared to G-values reported in prior literature and how the magnitudes compared between different studies. Material systems resembling ASNF pose complications for measuring/defining G-values to predict the evolution of H 2 in a sealed canister, including i) accounting for radiolytic yields potentially arising from multiple sources, i.e., residual free (vapor), physisorbed, and chemisorbed/chemically bound waters; ii) deciding what portions of the multi-material system to include in the absorbed energy (radiation dose) calculation, considering possible energy exchange between materials as well as measurement limitations, and iii) capturing variations in G-value associated with non-linear yield vs. dose curves and/or dependence on the cover gas. This report summarizes previous literature information on radiolytic H 2 generation and associated G-values from mixed-material systems (generally oxides in contact with water or organic compounds) and from (oxy)hydroxides/hydrates to compare with the definitions and values for ASNF surrogate samples containing adherent aluminum (oxy)hydroxides.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Test Plan for DOE-GENIORS Dyanamic Irradiation Tests

Dynamic irradiation tests will be performed within the DOE-GENIORS collaboration, to be finished by the end of March 2021. These tests will be performed using three different setups, one at INL, one at CEA, and one at CIEMAT. The cis-mTDDGA/PTD chemical system will be used, applying both "extraction" (cis-mTDDGA - n-dodecane/HNO3) and "stripping" (cis-mTDDGA - n-dodecane/PTD - HNO3) conditions. This test plan provides the details and parameters to be used to perform irradiation at the INL using the INL Radiolysis Test Loop.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2021

Radiation-Induced Oxidation Reactions of 2-Selenouracil in Aqueous Solutions: Comparison with Sulfur Analog of Uracil

One-electron oxidation of 2-selenouracil (2-SeU) by hydroxyl (●OH) and azide (●N3) radicals leads to various primary reactive intermediates. Their optical absorption spectra and kinetic characteristics were studied by pulse radiolysis with UV-vis spectrophotometric and conductivity detection and by the density functional theory (DFT) method. The transient absorption spectra recorded in the reactions of ●OH with 2-SeU are dominated by an absorption band with an λmax = 440 nm, the intensity of which depends on the concentration of 2-SeU and pH. Based on the combination of conductometric and DFT studies, the transient absorption band observed both at low and high concentrations of 2-SeU was assigned to the dimeric 2c-3e Se-Se-bonded radical in neutral form (2●). The dimeric radical (2●) is formed in the reaction of a selenyl-type radical (6●) with 2-SeU, and both radicals are in equilibrium with Keq = 1.3 × 104 M−1 at pH 4 (below the pKa of 2-SeU). Similar equilibrium with Keq = 4.4 × 103 M−1 was determined for pH 10 (above the pKa of 2-SeU), which admittedly involves the same radical (6●) but with a dimeric 2c-3e Se-Se bonded radical in anionic form (2●−). In turn, at the lowest concentration of 2-SeU (0.05 mM) and pH 10, the transient absorption spectrum is dominated by an absorption band with an λmax = 390 nm, which was assigned to the ●OH adduct to the double bond at C5 carbon atom (3●) based on DFT calculations. Similar spectral and kinetic features were also observed during the ●N3-induced oxidation of 2-SeU. In principle, our results mostly revealed similarities in one-electron oxidation pathways of 2-SeU and 2-thiouracil (2-TU). The major difference concerns the stability of dimeric radicals with a 2c-3e chalcogen-chalcogen bond in favor of 2-SeU.

2-selenouracil↗

DOE OSTI2026

Effects of Dimethylamino Functional Group Substitution on the Physical, Structural and Radiolytic Properties of Pyridinium Ionic Liquids

A diverse range of 4-dimethylaminopyridinium (DMAP) bis(trifluoromethylsulfonyl)-amide ionic liquids with specific functionalities (alkyl, alkoxy, hydroxyalkyl and benzyl) were designed, characterized and compared with their pyridinium analogs in terms of their physical and radiolytic properties. The influence of the dimethylamino group on ionic liquid structure was investigated by X-ray diffraction and molecular dynamics simulations. The influence of the electron-donating ability of the dimethylamino-substituted cation is evident in the differences in the electronic density of states between the DMAP and pyridinium ILs. This leads to substantial changes in the radical transients observed in pulse radiolysis of the neat ILs. It was found that the DMAP salts were higher melting, more viscous and less conducting than their pyridinium analogs. However, the DMAP salts exhibited higher thermal stabilities and could therefore be useful for high-temperature applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2021

Milestone 1.2.9: Radiolytic Gas Generation Measurements from Helium Backfilled Samples of AA1100 and AA6061 Coupons

Nearly 18 metric tons of aluminum-clad spent nuclear fuel (ASNF) is safely managed by the U.S. Department of Energy (DOE). These assemblies are currently in interim storage, with the intention of extended storage (>50 years) until final disposal. Strategies for the continued safe storage of this material are under evaluation, of which a key criterion is the extent of molecular hydrogen gas (H2) formation from the radiolysis of hydrated (oxy)hydroxide aluminum corrosion layers arising from in-reactor and wet storage conditions. Radiation-induced H2 formation has the potential to compromise cladding and storage canister integrity, in addition to promoting the formation of unfavorable gaseous environments. Consequently, understanding this radiation-induced phenomenon is essential for the development of predictive modeling capabilities to support technical considerations and the identification of radiation related challenges for the extended storage of ASNF. Here, we report radiolytic H2 yields (G-values, G(H2)) from the gamma irradiation of ‘pristine’ and pre-corroded aluminum coupons in helium (He) environments as a function of alloy composition (AA1100 and AA6061), relative humidity, and absorbed gamma dose. Measured yields were lower than corresponding values reported for argon environments, a positive result for proposed extended dry storage strategies that would employ helium as a backfill gas. Interestingly, the presented G(H2)He values are comparable to those previously measured in nitrogen environments, suggesting a He mediated H2 inhibition process, attributed here to Penning ionization.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2022

Modeling of ATR Fuel in DOE Standard Canisters with Helium Backfill

One pathway for road-ready and final disposition packaging configurations for the aluminum-clad spent nuclear fuel (ANSF) fuel is storage within helium backfilled sealed Department of Energy (DOE) standard canisters. The typical packaging configuration for the 15-foot DOE standard canisters places 10 advance test reactor (ATR) elements a basket, and three baskets are loaded within each DOE canister. During in-reactor operations and cooling pond storage conditions, oxyhydroxide layers form on the surface of the aluminum clad fuel. These layers produce hydrogen gas over time due to the fuel’s radiation field. As part of the packing procedure, the ATR fuel should be dried to remove any residual physio-/chemi- sorbed water bound to the surface. A 50-year CFD model of the DOE canister packaged with fuel was developed to provide a temperature profile for coupled chemical modeling of the conditions within the canister. The results of this modeling include results at fully saturated and fully dried fuel cladding conditions. In the associated experimental work, radiolysis experiments tests were completed in a helium environment, and G-values for the radiolytic production of hydrogen from the oxyhydroxide layers were provided. That reaction was coupled with the thermal profiles and gas-phase reactions to develop a 50-year model of the conditions within a sealed DOE canister with ATR fuel. For a nominal scenario of stored ATR fuel, after 50 years the model results give a 1.36 atm total pressure, 7% mole percent hydrogen, for the upper decay heat, 1.51 atm total pressure, 16% mole percent hydrogen, and for upper decay heat with undried fuel 2.6 atm total pressure, 15% mole percent hydrogen. No case modeled yields significant oxygen, and for the lower decay heat case that is modeled, hydrogen concentrations are under the 4% flammability limit after 50 years of storage. The modeled pressures for all cases modeled are below the pressure limit for the DOE standard sealed canister.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Impact of Lanthanide Complexation and Temperature on the Chemical Reactivity of N,N,N',N'-tetraoctyl diglycolamide (TODGA) with the Dodecane Radical Cation

The impact of lanthanide (Ln) metal ion complexation and temperature on the chemical reactivity of N,N,N',N'-tetraoctyl diglycolamide (TODGA) with the dodecane radical cation (RH?+) has been measured by electron pulse radiolysis. Complexation of trivalent neodymium (Nd), gadolinium (Gd), and ytterbium (Yb) by TODGA yielded [Ln(TODGA)3(NO3)3] complexes that exhibited significantly increased reactivity (up to 9.3×) with the RH?+ radical cation, relative to the “free” ligand: k([Ln(TODGA)3(NO3)3] + RH?+) = (8.99 ± 0.93) × 1010, (2.88 ± 0.40) × 1010, and (1.53 ± 0.34) × 1010 M–1 s–1, for Nd(III), Gd(III), and Yb(III), respectively. The kinetic enhancement measured for both ligands exhibited a dependence on atomic number. Arrhenius parameters—specifically activation energies (Ea) and pre-exponential factors (A)—were determined for the reaction of “free” TODGA ligand with the RH?+ radical cation, giving: Ea(TODGA) = 17.43 ± 1.64 kJ mol–1, and A(TODGA) = (1.08 ± 0.02) × 1013 M–1 s–1. This draft manuscript has been prepared in fulfillment of Milestone M4FT-22IN030402024.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2022

Recent Advances in Radiation-Induced Actinide Redox Chemistry

The actinide series boasts many unique physical and chemical features worthy of both fundamental and applied study. However, the chemical influence of their inherent radiation field is often overlooked, especially as we begin to explore the late actinides in more detail than ever possible before. From the perspective of used nuclear fuel reprocessing, the absorption of ionizing radiation induces the formation of a variety of transient and steady-state excited states, radicals, ions, and molecular degradation products, many of which are highly redox active and can lead to significant changes in a reprocessing solvent system’s physical and chemical properties. For example, radiolysis of the actinides can drive steady-state redox distributions and the formation of non-traditional oxidation states which can complicate their separation and recovery from fission products. This scenario is further exacerbated when complexation is taken into account. Consequently, a molecular-level understanding of radiation effects on the actinides over multiple time, distance, and material domains is essential for supporting innovation in used nuclear fuel reprocessing technologies. Attaining this knowledge necessitates a firm grasp of actinide radiation chemistry to develop predictive, mechanistic, multiscale models to support engineering efforts. Presented here are several recent studies that highlight recent advances in actinide radiation chemistry, in particular, the effect of actinide complexation on ligand reactivity towards radiation-induced transients.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2020

Key Technical Issues for Greater-Than-Class-C (GTCC) Waste Disposal - 20164

Greater-than-Class C (GTCC) waste is low-level waste (LLW) that exceeds the Class C concentrations tabulated in Title 10, Code of Federal Regulations (CFR), Part 61. Disposal of GTCC waste in the near surface (i.e. upper 30 m of the earth's surface) is prohibited in the United States. Though GTCC waste disposal is generally prohibited, the Commission can approve disposal according to 10 CFR 61.55(a)(2)(iv) on a case-by-case basis. The waste classification tables, developed in the early 1980's, were based on model projections of dose to an inadvertent intruder in agricultural, construction, and discovery exposure scenarios. The assumptions and parameters used were documented in a series of public reports. Though different disposal facility designs and site conditions were considered, the waste classification tables were based on a disposal facility design located in a specific environment. The tables were based on shallow (i.e., top few meters) trench facility designs and did not consider deeper facilities. To determine the suitability of GTCC waste disposal in the near-surface, site-specific analyses must account for differences between GTCC waste and Class A, B, and C LLW (hereafter, traditional LLW). GTCC waste can have concentrations of radionuclides that are much higher than traditional LLW. Because of these higher concentrations, processes that are typically not significant for traditional LLW may be significant with respect to disposal of GTCC waste. These processes include, but are not limited to, heat generation, criticality, and radiolysis. The form of the waste as well as the barriers to release of the waste (e.g. waste package) could be substantially different than they are for traditional LLW. These barriers need to be considered when assessing the impacts of accidents during receipt and placement and in evaluating long-term performance. Some GTCC radioactivity is either embedded in stainless steel or contained in stainless steel barriers. Stainless steel can have very low corrosion rates under a variety of environmental conditions. Finally, and possibly most importantly, GTCC waste would likely need to be disposed deeper than traditional LLW to reduce the probability of disturbance. The waste classification table values of traditional LLW are based on an inadvertent intruder excavating into the waste and bringing some of the material to the land surface. If waste is deeper than approximately 5 m, the excavation scenario becomes very unlikely. Therefore, other intruder scenarios, such as drilling exposure scenarios, need to be evaluated. This paper summarizes key technical issues for the disposal of GTCC waste. Our previous study shows that certain GTCC waste may be suitable for near surface disposal whereas others may not. This study may support ongoing technical analyses assessing potential disposal of GTCC in a near surface disposal facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗