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To support the viability of a wash-down approach to mitigating nuclear contamination, this study presents a characterization of the aggregate of a common concrete by optical microscopy and the sorption-desorption characteristics of cesium from these into potential wash solutions. Various minerals with weathered surfaces displayed strong affinity for 137Cs with an effective partition coefficient Kd = 120 mL/g for micas, >25–90 mL/g for feldspars, and >25–30 mL/g for amphiboles. The desorption Kd into 0.1 M NH4Cl varied greatly but for amphiboles, sandstones, granite, and fine-grained quartzite it was >200 mL/g as a result of irreversible sorption. These same mineral phases are prevalent in all types of building materials, extending our conclusions more broadly to the problem of wide-area urban decontamination. In contrast, ionic solutions desorbed up to 98% of 137Cs from cement, suggesting that fresh concretes with an intact surface layer of cement could be more easily decontaminated if Cs+ interactions with the underlying minerals could be avoided. For practical applications common, non-hazardous chemicals such as sodium, potassium, and ammonium salts are as effective or more effective than harsher chemicals and expensive chelating agents. For example, when treated shortly after exposure, on time-scales commensurate with early response phase activities, 0.5 M KCl could remove nearly 50% of bound 137Cs from concrete aggregate. Statistical analyses showed that desorption from the fine aggregate benefited from higher K+ and NH4+ concentrations. These results suggest that contamination in large areas of the urban environment can be dramatically reduced using common chemicals obtained readily from local stores.

Borosilicate waste glass degradation models must quantify the effects of the solution composition on the dissolution rate. Here, we present results of modified ASTM C1285 tests conducted at 90 °C with AFCI and LRM glasses to determine whether dependencies of dissolution rates on the pH, Al, and Si concentrations must be included. Solution compositions were modified from those generated by glass dissolution alone by adding small amounts of K4SiO4 glass, Al(OH)3•2H2O, and a concentrated NaOH solution when the tests were initiated. Results show rate laws for the initial and resumption regimes must include pH dependences, but the residual rate can be modeled independent of the pH, Al, and Si concentrations. Triggering the resumption rate probably depends on the pH, Si, and Al concentrations and perhaps other aspects of the glass composition. A waste glass degradation model using is being parameterized using tests with a range of waste glass compositions to quantify these dependencies.

A system for rapid reduction of radioactive contamination and recycle of contaminated waters is called the Integrated Wash-Aid, Treatment, and Emergency Reuse System (IWATERS). First developed for cesium contaminations, IWATERS prescribes the use of salt and surfactant additives to decontaminate radionuclides from urban surfaces. The water is collected and recycled after passing through reactive filtration beds containing selective sorbents. To adapt the IWATERS for strontium contaminations, potential additives to enhance its decontamination from urban surfaces are identified. One possible additive is calcium (Ca(2+)). However, its concentration can have a very strong detrimental effect on the ability of selective sorbents to remove strontium from spent wash water. We recognized that studies on off-the-shelf sorbents that include Ca(2+) concentrations at relevant levels (greater than millimolar) are absent in the literature. To understand better the effect of Ca(2+), we completed a literature review, batch tests, and surface complexation modeling to reveal few sorbent options. Only silico-titanate sorbents exhibited high K(d) values in the presence of Ca(2+), but have significant drawbacks in cost and availability. Given the state of the art, it is imperative that alternatives to alkaline earth ions in the IWATERS be identified to permit in situ recycle of the wash waters.

The immobilization of radioiodine produced from reprocessing used nuclear fuel is a growing priority for research and development of nuclear waste forms. This review provides a comprehensive summary of the current issues surrounding processing and containment of 129I, the isotope of greatest concern due to its long half-life of 1.6 × 107 y and potential incorporation into the human body. Strategies for disposal of radioiodine, captured by both wet scrubbing and solid sorbents, are discussed, as well as potential iodine waste streams for insertion into an immobilization process. Next, consideration of direct disposal of salts, incorporation into glasses, ceramics, cements, and other phases is discussed. The bulk of the review is devoted to an assessment of various sorbents for iodine and of waste forms described in the literature, particularly inorganic minerals, ceramics, and glasses. This review also contains recommendations for future research needed to address radioiodine immobilization materials and processes.

The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO 2 and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO 2 and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO 2 and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H 2 O 2 and O 2 ). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package. The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit fuel degradation to chemical dissolution, which results in radionuclide source term values that are four or five orders of magnitude lower than when oxidative dissolution processes are operative. This paper presents the scientific basis of the model, the approach for modeling used fuel in a disposal system, and preliminary calculations to demonstrate the application and value of the model.

A new method is presented here for digesting irradiated low-enriched uranium foil targets in alkaline carbonate media to recover 99Mo. This method consists of the electrolytic dissolution of uranium foil in a sodium bicarbonate solution, followed by the precipitation of carbonate, base-insoluble fission products, activation products, and actinides with calcium oxide; most of the molybdenum, technetium, and iodine remain in solution. An electrochemical dissolver and mixing vessel were designed, fabricated, and tested for the processing of a full-sized irradiated foil under ambient pressure and elevated temperature. Over 92% of the fission-induced 99Mo was recovered in a product solution that was compatible with an anion-exchange column for retaining molybdenum and iodine.

The overall objective of the SFR Regulatory Technology Development Plan (RTDP) effort is to identify and address potential impediments to the SFR regulatory licensing process. In FY14, an analysis by Argonne identified the development of an SFR-specific MST methodology as an existing licensing gap with high regulatory importance and a potentially long lead-time to closure. This work was followed by an initial examination of the current state-of-knowledge regarding SFR source term development (ANLART-3), which reported several potential gaps. Among these were the potential inadequacies of current computational tools to properly model and assess the transport and retention of radionuclides during a metal fuel pool-type SFR core damage incident. The objective of the current work is to determine the adequacy of existing computational tools, and the associated knowledge database, for the calculation of an SFR MST. To accomplish this task, a trial MST calculation will be performed using available computational tools to establish their limitations with regard to relevant radionuclide release/retention/transport phenomena. The application of existing modeling tools will provide a definitive test to assess their suitability for an SFR MST calculation, while also identifying potential gaps in the current knowledge base and providing insight into open issues regarding regulatory criteria/requirements.more » The findings of this analysis will assist in determining future research and development needs.« less

The mineralogy and geochemistry of soils developed over the unmined Coles Hill uranium deposit (Virginia) were studied to determine how phosphorous influences the speciation of uranium in oxidizing soil/saprolite systems typical of the eastern US. Results from this study have implications for both uranium remediation (e.g. in situ stabilization) and uranium resource exploration (e.g. near-surface geochemical sampling). The primary uranium ore (coffinite and uraninite hosted in quartzo-feldspathic gneiss) weathers to saprolites containing the same uranium concentration as the underlying ore (approximately 1000 mg U/kg saprolites). In these water saturated (below water table) saprolites the uranium is retained as uranyl phosphates of the meta-autunite group (mainly meta-uranocircite). Above the water table the soils overlying the deposit contain approximately 200 mg uranium per kg soil (20 times higher than uranium concentrations in similar soils formed from unmineralized rocks adjacent to the deposit). In these unsaturated zone soils uranium is retained by two processes: (1) incorporation into barium–strontium–calcium aluminum phosphate minerals of the crandallite group (mainly gorceixite), and (2) sorption of uranium with phosphorous onto iron oxides that coat the surfaces of other soil minerals. Thermodynamic calculations suggest that the meta-autunite group minerals present in the saprolites below the water table are not stable in the unsaturated zone soils overlying the deposit due to the drop in soil pH from ∼ 6.0 down to ∼ 4.5. Mineralogical observations suggest that, once exposed to the unsaturated environment, the meta-autunite group minerals react to form U(VI)-bearing aluminum phosphates and U(VI) surface complexes or nano-precipitates associated with ferric oxides. These results therefore indicate that models predicting U(VI) speciation in phosphate amended soils must simultaneously account for variations in pH, ion activities (aluminum appears to be particularly important) and surface complexation with iron oxide mineral surfaces.

The aqueous dissolution of irradiated and unirradiated uranium-thorium dioxide, (U,Th)O{sub 2}, fuel pellets in Yucca Mountain well water has been investigated. Whole and crushed pellets were reacted at 25 and 90 deg. C for periods of up to 195 days. The fuel dissolution was measured by analyzing the concentrations of soluble uranium, thorium, and important fission products ({sup 137}Cs, {sup 99}Tc, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Am) in the well water. The surface-area-normalized fractional uranium release rates for unirradiated crushed uranium dioxide (UO{sub 2}) pellets were 10 to 40 times higher than the values for (U,Th)O{sub 2} fuel. Similarly, the dissolution rates of irradiated (U,Th)O{sub 2} pellets with compositions ranging from 2.0 to 5.2% UO{sub 2} were at least two orders of magnitude lower than reported literature values for pure UO{sub 2}. These results demonstrate an advantage of (U,Th)O{sub 2} over UO{sub 2} in terms of matrix dissolution in groundwater and suggest that (U,Th)O{sub 2} fuel is a more stable long-term waste form than conventional UO{sub 2} fuel.

The unmined Coles Hill U deposit of south central Virginia represents a natural setting where U is stabilized by phosphate mineral precipitation in an oxidizing bedrock aquifer. Drill cores from the shallow portion of the deposit preserve a sharp Fe redox front defined by Fe(III) oxide staining. This front is located near a discontinuity in U mineralogy with U(IV) assemblages (e.g. coffinite, uraninite) on the reducing side, and U(VI) assemblages on the oxidizing side. The discontinuity in U mineralogy does not, however, represent a major discontinuity in whole rock U concentrations, with sample groups from both oxidized and reduced sides of the front generally ranging from 500 to 1000 ppm. This observation suggests that the volume of shallow bedrock associated with the deposit has not lost significant amounts of U during the oxidation and incipient chemical weathering. The precipitation of Ba uranyl phosphate (Ba meta-autunite) is responsible for U retention within this zone. Ground waters sampled from the weathered bedrock aquifer associated with the deposit contain less than 15 m gl � 1 dissolved U. This suggests that the low solubility of the Ba meta-autunite limits U concentrations to values lower than the US-EPA maximum contaminant level of 30 m gl � 1 . Ground water speciation and mineral saturation calculations show that, in addition to Eh and pH, the most important factor controlling this U fixation process is the activity ratio of dissolved phosphate to dissolved carbonate. Experimental results suggest that, at the Coles Hill site, the oxidation of U(IV) to U(VI) and subsequent precipitation of uranyl phosphate occurs rapidly (time scale of weeks) relative to ground water transport (e.g. 20 m/a). Furthermore, based on the rate of downward migration of the redox front, it is estimated that the oldest U(VI) phosphate assemblages associated with the Coles Hill U deposit have been stable for up to 150 ka. These observations have important implications for the design and long term performance assessment of phosphate-based stabilization and reactive barrier techniques.

The primary purpose of this report is to describe the strategy for coupling three process level models to produce an integrated Used Fuel Degradation Model (FDM). The FDM, which is based on fundamental chemical and physical principals, provides direct calculation of radionuclide source terms for use in repository performance assessments. The G-value for H2O2 production (Gcond) to be used in the Mixed Potential Model (MPM) (H2O2 is the only radiolytic product presently included but others will be added as appropriate) needs to account for intermediate spur reactions. The effects of these intermediate reactions on [H2O2] are accounted for in the Radiolysis Model (RM). This report details methods for applying RM calculations that encompass the effects of these fast interactions on [H2O2] as the solution composition evolves during successive MPM iterations and then represent the steady-state [H2O2] in terms of an “effective instantaneous or conditional” generation value (Gcond). It is anticipated that the value of Gcond will change slowly as the reaction progresses through several iterations of the MPM as changes in the nature of fuel surface occur. The Gcond values will be calculated with the RM either after several iterations or when concentrations of key reactants reach threshold values determined from previous sensitivity runs. Sensitivity runs with RM indicate significant changes in G-value can occur over narrow composition ranges. The objective of the mixed potential model (MPM) is to calculate the used fuel degradation rates for a wide range of disposal environments to provide the source term radionuclide release rates for generic repository concepts. The fuel degradation rate is calculated for chemical and oxidative dissolution mechanisms using mixed potential theory to account for all relevant redox reactions at the fuel surface, including those involving oxidants produced by solution radiolysis and provided by the radiolysis model (RM). The RM calculates the concentration of species generated at any specific time and location from the surface of the fuel. Several options being considered for coupling the RM and MPM are described in the report. Different options have advantages and disadvantages based on the extent of coding that would be required and the ease of use of the final product.

A mineralogical and geochemical study of soils developed from the unmined Coles Hill uranium deposit (Virginia) was undertaken to determine how phosphorous influences the speciation of uranium in an oxidizing soil/saprolite system typical of the eastern US. This paper presents mineralogical and geochemical results that identify and quantify the processes by which uranium has been sequestered in these soils. It was found that uranium is not leached from the saturated soil zone (saprolites) overlying the deposit due to the formation of a sparingly soluble uranyl phosphate mineral of the meta-autunite group. The concentration of uranium in the saprolites is approximately 1000 mg uranium per kg of saprolite. It was also found that a significant amount of uranium was retained in the unsaturated soil zone overlying uranium-rich saprolites. The uranium concentration in the unsaturated soils is approximately 200 mg uranium per kg of soil (20 times higher than uranium concentrations in similar soils adjacent to the deposit). Mineralogical evidence indicates that uranium in this zone is sequestered by a barium-strontium-calcium aluminum phosphate mineral of the crandallite group (gorceixite). This mineral is intimately inter-grown with iron and manganese oxides that also contain uranium. The amount of uranium associated with both the aluminum phosphates (as much as 1.4 weight percent) has been measured by electron microprobe microanalyses and the geochemical conditions under which these minerals formed has been studied using thermodynamic reaction path modeling. The geochemical data and modeling results suggests the meta-autunite group minerals present in the saprolites overlying the deposit are unstable in the unsaturated zone soils overlying the deposit due to a decrease in soil pH (down to a pH of 4.5) at depths less than 5 meters below the surface. Mineralogical observations suggest that, once exposed to the unsaturated environment, the meta-autunite group minerals react to form U(VI)-bearing aluminum phosphates.

Performance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO22+) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of ∼ 50 micrometers. We find evidence of a thin (∼ 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO2+/Np4+ and UO22+/U4+ couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions.

Batch adsorption experiments in which neptunium bearing solutions were reacted with goethite (alpha-FeOOH) have been performed to study uptake mechanisms in sodium chloride and calcium-bearing sodium silicate solutions. This paper presents results identifying and quantifying the mechanisms by which neptunium is adsorbed as a function of pH and reaction time (aging). Also presented are results from tests in which neptunium is reacted with goethite in the presence of other cations (uranyl and calcium) that may compete with neptunium for sorption sites. The desorption of neptunium from goethite has been studied by resuspending the neptunium-loaded goethite samples in solutions containing no neptunium. Selected reacted sorbent samples were analyzed by x-ray absorption spectroscopy (XAS) to determine the oxidation state and molecular speciation of the adsorbed neptunium. Results have been used to establish the pH adsorption edge of neptunium on goethite in sodium chloride and calcium-bearing sodium silicate solutions. The results indicate that neptunium uptake on goethite reaches 95% at a pH of approximately 7 and begins to decrease at pH values greater than 8.5. Distribution coefficients for neptunium sorption range from less than 1000 (moles/kg)sorbed / (moles/kg)solution at pH less than 5.0 to greater than 10,000 (moles/kg)sorbed / (moles/kg)solution at pH greater than 7.0. Distribution coefficients as high as 100,000 (moles/kg)sorbed / (moles/kg)solution were recorded for the tests done in calcite equilibrated sodium silicate solutions. XAS results show that neptunium complexes with the goethite surface mainly as Np(V) (although Np(IV) is prevalent in some of the longer-duration sorption tests). The neptunium adsorbed to goethite shows Np-O bond length of approximately 1.8 angstroms which is representative of the Np-O axial bond in the neptunyl(V) complex. This neptunyl(V) ion is coordinated to 5 or 6 equatorial oxygens with Np-O bond lengths of 2.45 angstroms. The absence of a clearly recognizable Np-Fe interaction for the sodium chloride sorption tests suggests that neptunium in these solutions adsorbs as an outer-sphere complex. XAS results from the calcium-bearing sodium silicate sorption tests show evidence for a neptunyl(V) inner-sphere surface complex with a Np-Fe interaction at 3.5 angstroms. Desorption tests indicate that samples in which neptunium is bound as inner-sphere complexes show significant sorption hysteresis relative to samples in which neptunium is bound largely as outer-sphere complexes.

Heat generated by radioactive decay of spent fuel represents a potentially important barrier to water accumulation on commercial spent nuclear fuel in breached waste packages. In the absence of water, fuel degradation and radionuclide release will be negligible. Thermal models for the proposed Yucca Mountain Repository suggest that, after a period of approximately 1000-4000 years (depending on loadingand ventilation conditions), the repository drift walls may decline to sub-boiling temperatures, thus allowing humidity in the drift to increase. The question thus arises, is the thermal gradient between the fuel and the drift sufficient to prevent water accumulation in a humid drift environment throughout the regulatory period? The answer depends on the balance between processes that oppose water condensation ontothe fuel (decay heat) and those that promote condensation such as the deliquescence of hygroscopic phaseswithin the fuel.Our experimental results indicate that deliquescence could lead to the condensation of water onto spent fuel despite the thermal “self-drying”effect if the following criteria are met: (1) the fission product salt CsI is present in the fuel or in the fuel-cladding gap, (2) the relative humidity in the driftexceeds 80% while temperatures in the waste package are around 90oC. Previous work suggests that these criteria may be met for some fuel pins within the proposed Yucca Mountain Repository. However,experiments that account for the role of U(VI) alteration phases suggest that deliquescence may be a self-limiting process in the sense that deliquescent components (e.g. Cs, Ba, Sr) may be incorporatedinto nondeliquescent U(VI) phases that form from the corrosion of spent fuel.

The dissolution behavior and fission-product release from irradiated thoria-urania fuel was studied by immersing fuel samples in J-13 well water at 90°C. The samples are from the Shippingport Light Water Breeder Reactor and consist of binary solid solutions of (U,Th)O2 with UO2 contents varying from 2.0 to 5.2 Wt.%. The post-irradiation U isotopic composition of the samples used in our experiments is: 87.3% 233U, 10.4% 234U, 1.8% 235U, and 238U, 236U, 232U. Burn up values for the samples range from 22.3 to 40.9 megawatt-days per kg-metal. Our tests were performed on polished disks and on crushed and sieved samples in stainless-steel reaction vessels with air-filled head-space. After 196 days of reaction, samples showed no evidence for corrosion at the micrometer scale. Concentration ranges (μgL-1) of key radionuclides in filtered (∼5 nm pore size) leachates were: 0.1 – 15 90Sr, 0.9 – 7.0 99Tc, 0.1 – 35.2 137Cs,233U, 232Th. Concentrations of 237Np, 239Pu, 240Pu and 241Am were all -1. The relatively high concentrations of the fission products 90Sr and 137Cs occur early during leaching and decrease for later samplings. Matrix dissolution rates for the irradiated thoria-urania samples range from ∼3x10-3 to -5 mg m-2day-1 and are at least two orders of magnitude lower than those measured for UO2 spent fuels under similar experimental conditions.