Neptunium is found predominantly as Np(IV) in reducing environments, but as Np(V) in aerobic environments. Currently, it is not known how the interplay between biotic and abiotic processes affects Np redox speciation in the environment. To evaluate the effect of anaerobic microbial activity on the fate of Np in natural systems, Np(V) was added to a microcosm inoculated with anaerobic sediments from a metal-contaminated freshwater lake. The consortium included metal-reducing, sulfate-reducing, and methanogenic microorganisms, and acetate was supplied as the only exogenous substrate. Addition of more than 10−5 M Np did not inhibit methane production. Total Np solubility in the active microcosm, as well as in sterilized control samples, decreased by nearly two orders of magnitude. A combination of analytical techniques, including VIS-NIR absorption spectroscopy and XANES, identified Np(IV) as the oxidation state associated with the sediments. The similar results from the active microcosm and the abiotic controls suggest that microbially produced Mn(II/III) and Fe(II) may serve as electron donors for Np reduction.

Transport of bacteria is investigated as part of the Large-Block Test (LBT), a thermally perturbed block of Topopah Spring tuff. Two bacterial species, Bacillus subtilis and Arthrobacter oxydans, were isolated from the Yucca Mountain tuff. Natural mutants that can grow under the simultaneous presence of the two antibiotics, streptomycin and rifampicin, were selected from these species by laboratory procedures, cultured, and injected into the five heater boreholes of the large block hours before heating was initiated. The temperature, as measured 5 cm above one of the heater boreholes, rose slowly over a matter of months to a maximum of 142°C and to 60°C at the top and the bottom of the block. Samples were collected from boreholes located approximately 5 ft below the injection points. Double-drug-resistant bacteria were found in the collection boreholes nine months after injection. Biochemical tests support the identification of B. subtilis in the collection boreholes. Double-drug-resistant microbes also appeared in the heater boreholes where the temperature had been sustainably high throughout the test. The number of double-drugresistant bacteria that were identified in the collection boreholes increased with time until the heater was deactivated. Negative indications in the collection holes after the heater was deactivated support the supposition that these bacteria were the species that were injected. An apparent homogeneous distribution among the collection boreholes suggests no pattern to the migration of bacteria through the block. The relationship between bacterial migration and the movement of water is not yet understood. These observations indicate the possibility of rapid bacterial transport in a thermally perturbed geologic setting. The implications for colloid transport need to be reviewed.

A mathematical model has been developed to predict the extent of sulphate reduction by sulphate-reducing bacteria (SRB) in an underground nuclear waste disposal vault and the consequences for corrosion of the waste package. The model is based on a series of massbalance equations that describe the kinetics of sulphate reduction by two types of SRB (one organotrophic and one chemoheterotrophic), the growth and death of SRB, the supply and consumption of nutrients (acetate and hydrogen) and reactants (SO4 2−) and the consumption of sulphide by precipitation with aqueous Fe(II) or by container corrosion. The disposal vault is described by a series of 1-D mass-transport barriers representing the buffer and backfill materials and fractured rock. The model can also be used to simulate the effects of γ-radiation and desiccation of the buffer and backfill materials on the extent of microbial activity.

The U.S. Department of Energy is contributing to the design of a potential nuclear waste repository at Yucca Mountain, Nevada. A system to predict the contribution of Yucca Mountain (YM) bacteria to overall corrosion rates of candidate waste package (WP) materials was designed and implemented. DC linear polarization resistance techniques were applied to candidate material coupons that had been inoculated with a mixture of YM-derived bacteria with potentially corrosive activities, or left sterile. Inoculated bacteria caused a 5- to 6-fold increase in corrosion rate of carbon steel C 1020 (to approximately 7-8μm/yr), and an almost 100-fold increase in corrosion rate of Alloy 400 (to approximately μm/yr) was observed due to microbiological activities. Microbiologically Influenced Corrosion (MIC) rates on more resistant materials (CRMs: Alloy 625, Type 304 Stainless Steel, and Alloy C22) were on the order of hundredths of micrometers per year (μm/yr). Bulk chemical and surfacial endpoint analyses of spent media and coupon surfaces showed preferential dissolution of nickel from Alloy 400 coupons and depletion of chromium from CRMs after incubation with YM bacteria. Scanning electron microscopy also showed greater damage to the Alloy 400 surface than that indicated by electrochemical detection methods.

Eight perovskites of different compositions and pre-existing vacancy contents were irradiated with 1.5 MeV Kr+ ions using the HVEM-Tandem User Facility at Argonne National Laboratory. The critical dose of 1.5 MeV Kr+ ions for amorphisation (Dc) of NaNbO3, SrTiO3 and two natural perovskites varies with composition and may decrease with Na content. The D,c values of 5 members of the solid solution series from SrTiO3 to La0.67TiO3 do not exhibit a simple relationship with vacancy content: In particular, Dc(Sr0 7La0.2TiO3) is extraordinarily high (>170 × 1014 ions cm−2) compared to those of the other (Sr, La)-perovskites in the series, which have Dc values between 2.4 and 11 × 1014 ions cm−2.

Argonne National Laboratory has developed a glass-bonded sodalite ceramic waste form to immobilize fission products and plutonium that accumulate during the electrometallurgical conditioning of spent nuclear fuel. To investigate the effects of alpha decay damage on the structure and leaching characteristics of the ceramic material, 238Pu has been incorporated into the ceramic waste form. The 238pu,with its high specific activity, significantly increases the rate of alpha damage to the waste form. Long term studies have begun with periodic examination of the 238Pu loaded ceramic material. A number of characterization techniques are used to study the alpha decay damage on the structure of the waste form. In addition, PCT type leachate studies will be performed to determine the effect of alpha decay damage on the durability of the ceramic waste form. Preliminary results from this study are presented.

The generation of hydrogen-gas from metallic waste is an important issue for the safety analysis of geological disposal facilities for transuranic (TRU) radioactive waste in Japan. The objective of this study is to clarify the gas-generation behavior of stainless steel and carbon steel in non-oxidizing alkaline synthetic groundwater (pH 12.8 and 10.5) at 30 °C simulating geological disposal environments. At pH 12.8, the observed gas-generation rate from stainless steel in the initial period of immersion was 1.0 × 102 Nml/m2/y and 1.0 × 10 Nml/m2/y after 200 days (N represents the standard state of gas at 0 °C and 1 atm). At pH 10.5, gas generation was not observed for 60 days in the initial period. At 60 days, the gasgeneration observed was 5.0 × 10 Nml/m2/y. After 250 days, the gas-generation rate approaches zero. At pH 12.8, the observed gas generation rate of carbon steel in the initial period of immersion was 1.5 × 102 Nml/m2/y and the gas generation rate began to decrease after 200 days. After 300 days, it was 25 Nml/m2/y. At pH 10.5, the gas generation rate in the initial period was 5.0 × 102 Nml/m2/y and was 1.0 × 10 Nml/m2/y after 200 days.

Iron phosphate ceramic (IPC) samples were prepared with plutonium-238, and radiolytic gas generation was detennined for exposure times of up to 195 days. The specific activity of the synthesized IPC samples was 1.91 mCi/g. The composition of the generated gases was determined to contain mainly H, and lower amounts of O2, CO2 and CO. The molecular hydrogen G value (molecules of H2 per 100 eV absorbed dose) depended on the sample preparation methodology, ranging from 0.07 at 79 days for sample IPC-2 to 0.34 at 119 days for sample IPC-1. This difference could be due to different total amount (wt.%) of free water resulting in sample preparation. The leach rate data for radioactive and non-radioactive IPC samples obtained by the ANSI/ANS 16.1 procedure resulted in Leachability Indexes of 13.7 for Pu and 15.4 for Ce, above the NRC LLW acceptance minimum index of 6.

The long-lived radioactivity of lead-bismuth coolant (LBC) inside the targets of the accelerator driven systems (ADS) is estimated. It is determined by both spallation and fission products of lead and bismuth nuclides. Based on US safety standards the radiotoxicity of targets was calculated for different cooling down periods. In order to estimate the total radioactivity of 1-MW power target circuit (TC), which is developing now at IPPE, the contribution of its main components are preliminarily analyzed.

An important function of the cement backfill in a nuclear waste repository is to react with aqueous waste species and reduce their solubility. However, to quantify backfill performance it is first necessary to prove the existence and establish the nature of the chemical solubility controls. This can be done by characterising the solubility-limiting phases, determining their solubility and thermodynamic functions, and assessing their stability and persistence and solubility during backfill degradation. Much of the necessary data must be acquired experimentally. The title paper describes briefly the progresss of experimental work on selected species including nickel, chromium(III,VI), tin(IV) molybdenum(VI), uranium(VI), Ce(III,IV), thorium, actinide simulants (III, IV) and chloride. Data needs are assessed and although much experimental work remains to be done, methodologies have been developed which will expedite progress. The expectation of a more quantitative performance assessment of cement barriers is, therefore, attainable.

A model for dissolution of CaO-SiO2-H2O (C-S-H) gel is proposed by considering a Margules type of non-ideal mixture of binary solid solutions at Ca/Si (C/S)<1. The Guggenheim and Prausnitz equations are used to represent the activities of model solids as a function of mole fractions. The Gibbs-Duhem equation, together with the activities of model solids, is applied to express the conditional solubility products of model solids as a function of C/S ratio. The determination of Guggenheim’s empirical parameters is performed with the geochemical code PHREEQE on experimental data. The dissolution model is developed by considering the effect of disorder in the structure of C-S-H gel at C/S<1. The solubility results predicted by the proposed model are comparable with experimental data. The study predicts some structural information of C-S-H gel at C/S<1.

Calculations have been performed to investigate long timescale redox control in a cementitious deep waste repository, once there has been significant corrosion of the steel containers in the near field. The methodology considers reaction of the porewater with redoxsensitive solids. The solids were selected based on considerations of the repository concept, the waste inventory and the processing of the waste into a form suitable for disposal. A number of factors that may affect the redox potential have been considered. Amongst these are the form of the redox-controlling solids, the effect of changes in pH, the effect of sorption onto the cementitious backfill, and the effect of the association of dissolved radionuclides with complexants arising from the degradation of organic wastes. The calculations suggest that for the United Kingdom Nirex Limited (Nirex) disposal concept and inventory, the redox system in the near field of the repository would be controlled by the iron(II)/iron(III) redox couple. This would maintain the redox potential in the repository near field at a low value for timescales in excess of 1 million years.

Phosphate-based chemically bonded ceramics were formed from magnesium potassium phosphate (MKP) binder and either industrial fly ash or steel slag. The resulting ceramics were subjected to solid-particle erosion by a stream of either angular Al2O3 particles or rounded SiO2 sand. Particle impact angles were 30 or 90° and the impact velocity was 50 m/s. Steady-state erosion rates, measured as mass lost from a specimen per mass of impacting particle, were dependent on impact angle and on erodent particle size and shape. Material was lost by a combination of fracture mechanisms. Evolution of H2O from the MKP phase appeared to contribute significantly to the material loss.

Powder and granules of the high density polyethylene (PEHD) were used to prepare mortar based matrices for immobilization of radioactive waste materials containing 137Cs, as well as containers for solidified radioactive waste form. Seven types of matrices, differ due to the percentage of granules and filler material added, were investigated. PEHD powder and granules were added to mortar matrix preparations with the objective of improving physico-chemical characteristics of the radwaste-mortar matrix mixtures, in particular the leach-rate of the immobilized radionuclide, as well as mechanical characteristics either of mortar matrix and container. In this paper, only mechanical strength aspect of the investigated mortar and concrete container formulations, is presented. The equivalent diameter of the PEHD granules used was 2.0 mm. PEHD granules were used to replace 100 volume percent of stone granules, sifted size of 2.0 mm, normally used in the matrix preparation, in order to decrease the porosity and density of the mortar matrix and to avoid segregation of the stone particles at the bottom of the immobilized radioactive waste cylindrical form. PEHD powder, particle size of 250 micrometre, was added as filler to the mortar formulation, replacing 5, 8 and 10 weight percent of the total cement weight in matrix formulation and 15 and 18 weight percent of the total cement weight in container formulation. Cured samples were investigated on mechanical strength, using 150 MPa hydraulic press, in order to determine influence of added polyethylene granules and powder on samples resistance to mechanical forces that solidified waste materials and concrete containers may experience at the disposal site. Results of performed investigations have shown that samples prepared with polyethylene granules, replacing 100 weight percents of the stone granules, have almost twice as much mechanical strength than samples prepared with stone aggregate. Samples prepared with PEHD granules and powder have mechanical strength resistance up to 13.5 percent higher than ones prepared with PEHD granules, solely. Improved Mechanical strength resistance of tested samples accommodates trend that functionally depends on the percentage of PEHD powder added in formulation.

Organic based ion selective resins have some similar attributes: case of synthesis, high metal ion complexation ability, and flexibility for different nuclear waste management applications. For most chelating polymers, the ligand is deemed to be of primary importance for the interaction with the targeted metal ion. The role of the polymer matrix is usually ignored. For ion specific resins, the polymer structure is formed to a specific metal ion. Using the molecular imprinting technique, resins can be formed with functional groups and cavities for a target metal ion. Ion selective resins have been developed for the separation of Cs. The methods and concepts used for the development of the Cs specific resins have been applied to the development of selective resins for Eu (a trivalent actinide model). The resulting resins are characterized by FTIR spectroscopy, moisture regain, and ion exchange capacity. The incorporation of 8-hydroxyquinoline into the resin increases selectivity for Eu over La. The results for the Eu study indicate ion specific resins can be developed for the separation of trivalent actinides from nuclear waste.

A computational modeling approach, based on molecular dynamics and related atomistic simulation techniques, is formulated to analyze the confinement of radionuclides in prepared nuclear wasteform for long-term disposal. While the intent is to exploit the ability of simulation to provide a unified treatment of chemical as well as kinetic processes at the molecular level, it is recognized that the problem is too complex to be studied in its entirety at the outset. To demonstrate the feasibility and effectiveness of the atomistic modeling approach, a first step is to determine an equilibrium isotherm for cation exchange, on the premise that a predominant release mechanism is likely to involve ion exchange. As a prototypical application, Cs-Na exchange on sodalite is studied using explicit interatomic potential models. Initial results for the equilibrium Cs concentration in dehydrated sodalite suggest that the effects of frame relaxation need to be taken into account. The present study will be extended to treat an aqueous environment for the loaded wasteform, as well as the effects of radiation damage to the zeolitic framework.

Crown ether Dibenzo-21-crown-7 (DB-7) has demonstrated a high capability for removing cesium-137 both from acidic and alkaline solution. However, there is a very important problem concerning changes in physical and chemical properties of such mixtures as a result of high irradiation. High radiation doses may cause a decrease in extraction efficiency and selectivity, formation of hazardous radiolysis products, disturbance hydrodynamics, and even decay of extractant. It is necessary to estimate such changes before extraction testing on “hot” solutions.

A special procedure, earlier developed in the RRICT (Russian Ministry for Atomic Energy), was used for testing of radiation and chemical stability of extraction mixtures. This method is based upon continuous control of extraction parameters - distribution coefficients of the specified radionuclides, their hold-up in organic phase, of hydrodynamic properties - kinetics of separation of mixed phases, surface tension, etc., and formation of radiolysis products.

The dose rate, determined using a ferrosulfate dosimeter was 1.7 W*h/l (1.05*1016 eV/ml*s) taking into account the electronic density of organic solution. Maximum total dose of gamma-irradiation was 80 W*h/l (1.6* 1021 eV/ml). The specimens of irradiated organic solution - DB-7 in the mixture of fluorinated alcohol (FA) and higher alcohol - after phase separation were analyzed to determine the hydrodynamic parameters. Both organic and aqueous phases were analyzed to estimate the decay of extractant and the composition of radiolysis products.

The primary emphasis for the first radiolytic experiments was given to the estimation of the FA decay by means of measuring the fluoride-ion concentration in the aqueous phase. It was shown that this value increases through the dose of 60 W*h/l and then decreases again. Such behavior may be attributed to an interaction of the fluoride ion with nitric acid and formation of volatile or organic phase extractable compounds. Our tests performed on real HLW acid solutions in P/A “Mayak” also demonstrated that solutions of DB-7 in fluorinated alcohols have a high resistance to radiolysis.

Coprecipitation of Np and Pu with Co(OH)3 and Fe(OH)3, obtained by the Method of Appearing Reagents from Co(NH3)6 Cl3, K2FeO4, and Na2[Fe(CN)5NO]2H2O from alkaline solutions and from simulants of Hanford Site tank wastes has been studied. Removal of Np and Pu from pH 10.5 to 0.4 M NaOH is semiquantitative with decontamination factors of 80 to 400. The presence of EDTA and HEDTA interferes with the precipitation of Co(OH)3 and Fe(OH)3. The deleterious effect of EDTA was eliminated by Ca2 addition at a 1:1 molar ratio of Ca2:EDTA. The deleterious effect of HEDTA was eliminated by the addition of 0.11 M excess precursor Co(NH3)6Cl3 to increase the content of carrier Co(OH)3.