The Oklo Natural Analogue Project, Phase II was conducted by an international consortium of researchers under the sponsorship of the European Union. An important objective of the project was to examine the potential for extracting quantitative information from the Oklo natural reactors that could be used in performance assessment calculations for high level nuclear waste repositories. This paper discusses the uses and limitations of natural analogues with respect to performance assessment of repositories, the types of data needed in performance assessment calculations, the particular example of spent fuel dissolution rates, and the types of data available from Oklo studies that might contribute to understanding spent fuel behavior in a geologic setting. The main conclusions of this assessment are that important qualitative information has been obtained from studies at Oklo that gives us confidence that major processes of importance to the long-term behavior of a high level waste repository have not been overlooked in previous performance assessment calculations, but that the Oklo natural analogues are very unlikely to provide quantitative data to use as input to PA calculations.

Geochemical models of clay-water interaction are evaluated using data characterizing the hydrochemical, isotopic and mineralogical properties of a natural claystone that is similar to bentonite buffers that have evolved over long periods of time in a deep geologic repository for nuclear wastes. Model predictions in general compare favorably with the mineralogy and hydrochemistry of this formation if increases in pH due to partial CO2(g) exsolution from porewater samples is accounted for. This suggests the models can be used with enhanced confidence to accurately predict the chemistry of pore fluids in bentonite buffers, enabling more defensible estimates to be made of radioelement solubilities that partially define the source term in performance assessments. Modifications are needed, however, to improve the accuracy of thermodynamic data supporting the models and to account for potential effects on porewater compositions of reactions that are kinetically inhibited from attaining equilibrium over experimental time scales.

Natural analog studies provide a means to test assumptions in performance assessment models. Data on the spatial distribution of metals derived from archeological artifacts at Akrotiri, Greece, were used to evaluate performance assessment model assumptions of radionuclide transport through unsaturated volcanic tuffs hosting the proposed repository at Yucca Mountain, Nevada. Processes controlling transport of trace elements from 3600 year old metallic artifacts at Akrotiri are analogous to processes controlling radionuclide transport at Yucca Mountain. Archeological information temporally and spatially constrains the source and chemistry of these trace elements, providing a comparison basis for model validation efforts. Despite these constraints and characterization of the Akrotiri site, the data are open to different interpretations as the extent of the plume emanating from the Minoan artifacts, transport pathways, heterogeneities in parameters, and boundary conditions are uncertain. Rather than validating a conceptual model of flow and transport, the Akrotiri site data indicate that 1-D models of aqueous phase transport in the unsaturated zone are conservative. Different conceptual models, apparently matching site data, produce different predictions illustrating that predictive flow and transport modeling is uncertain due to under-determined parameters and boundary conditions.

Conversion of waste graphite into a stable waste form acceptable for long term storage and disposal was considered both theoretically and experimentally. A self-sustaining transformation process of graphite composited with suitable precursors was studied. The powdered precursors that were used were used were: Al+SiO2 (1), Al+TiO2 (2) and Ti+SiO2 (3).

Numeric thermodynamic simulation was performed. Equilibrium temperatures and chemical compositions of reaction products were determined for a wide range of component ratios in the source mixtures. The highest temperatures (up to 2300 K) were observed for precursor type (2). Precursor type (3) demonstrated a minimal rise of temperature of up to 1900 K.

Regions of compositions with complete binding of all chemical elements as well as production of stable final products were found to be rather narrow. About 10 – 13 wt.% of carbon can be processed in composition with given precursors. The gas phase reaction products were studied to minimize carry over of radionuclides. Carbon monoxide was shown to be the main component of the gas phase.

The self-sustaining synthesis process was conducted in ceramic crucibles at ambient pressure in an air atmosphere. Batch masses ranged between 0.1 – 1 kg. Best results were obtained for processing of graphite composited with Al and TiO2. XRD analysis has identified titanium carbide and corundum in the waste form produced. These experiments confirmed that carbon can be converted completely into a stable waste form.

Waste blocks were produced in a bituminization plant using NPP-operational and other liquid wastes of low and intermediate level activity and tested under laboratory and near-surface wet disposal conditions. Leach rates of radioactive and non-radioactive waste components and depths of radionuclide penetration into the host loamy soil were estimated. Bituminized waste seems to occupy a middle position between cemented and vitrified waste forms in terms of radionuclide retention ability. For certain samples of the bitumen waste form, the testing covers a period of more then a quarter of a century that is of great importance for prediction of the waste form behavior over the required time period of several hundred years.

Argonne National Laboratory has developed a ceramic waste form to immobilize radioactive waste salt produced during the electrometallurgical treatment of spent fuel. The first ceramic waste forms that immobilize the radioactive waste salt have been produced. This study presents the first results from electron microscopy and durability testing of a ceramic waste form produced from that radioactive electrorefiner salt. The waste form consists of two primary phases: sodalite and glass. The sodalite phase appears to incorporate most of the alkali and alkaline earth fission products. Other fission products (rare earths and yttrium) tend to form a separate phase and are frequently associated with the actinides, which form mixed oxides. Seven-day leach test results are also presented.

The corrosion products formed on a stainless steel-15Zr (SS-15Zr) alloy have been characterized by transmission electron microscopy (TEM) and energy dispersive x-ray spectroscopy (EDS). Examination of alloy particles that were immersed in 90'C deionized water for two years revealed that different corrosion products were formed on the stainless steel and intermetallic phases. Two corrosion products were identified on an austenite particle: trevorite (NiFe2O4) in the layer close to the metal and maghemite (Fe2O3) in the outer layer. The corrosion layer formed on the intermetallic was uniform, adherent, and amorphous. The EDS analysis indicated that the layer was enriched in zirconium when compared with the intermetallic composition. High-resolution TEM images of the intermetallic-corrosion layer interface show an interlocking metal-oxide interface which may explain the relatively strong adherence of the corrosion layer to the intermetallic surface. These results will be used to evaluate corrosion mechanisms and predict long-term corrosion behavior of the alloy waste form.

Argonne National Laboratory has developed an electrometallurgical treatment for spent fuel from the experimental breeder reactor II. A product of this treatment process is a metal waste form that incorporates the stainless steel cladding hulls, zirconium from the fuel and the fission products that are noble to the process, i.e., Tc, Ru, Nb, Pd, Rh, and Ag. The nominal composition of this waste form is stainless steel/15 wt% zirconium/1-4 wt% noble metal fission products /1-2 wt % U. Leaching results are presented from crushed sample immersion tests on simulated metal waste form samples. The test results will be presented and their relevance for waste form product consistency testing discussed.

It is difficult to accurately predict actinide behavior during the alkaline leaching of Hanford's radioactive sludges due to the diverse chemical and radiolytic conditions existing in these wastes. The results of Pu dissolution during experimental washing of sludge simulants from the BiPO4, Redox, and PUREX processes shows that ≥2.1% Pu is dissolved during contact with alkaline media, but up to 65.5% Pu may be dissolved in acidic media. The dissolution of Cr, Fe, Nd, and Mn has also been observed, and the results of solid state, radioanalytical, and spectroscopic investigations are detailed.

This report describes results.from a seven-day demonstration of the use of a 3M Empore® membrane filter loaded with ion exchange material (potassium cobalt hexacyanoferrate (CoHex)) for radioactive cesium uptake from the R-Disassembly Basin at the Savannah River Site. The goal of the demonstration was to evaluate the efficacy of the Process Absorber Development Unit (PADU), a water pre-filtration/CoHex configuration on a skid, to remove cesium from R-Disassembly basin water at a linear processing flow rate of 22.7 liters per minute (1,200 liters/minute/mi2). During the seven-day demonstration, over 210,000 liters of R-Disassembly basin water was processed through the PADU without cesium breakthrough.

The Defense Waste Processing Facility (DWPF), at the Savannah River Site (SRS), is currently processing the second million gallon batch (Macro-Batch 2) of radioactive sludge slurry into a durable borosilicate glass for permanent geological disposal. To meet the reporting requirements as specified in the Department of Energy‘s Waste Acceptance Product Specifications (WAPS), for the final glass product, the nonradioactive and radioactive compositions must be provided for a Macro-Batch of material [1]. In order to meet this requirement, sludge slurry samples from Macro-Batch 2 were analyzed in the Shielded Cells Facility of the Savannah River Technology Center (SRTC). This information is used to complete the necessary Production Records at DWPF so that the final glass product, resulting from Macro Batch 2, may be disposed of at a Federal Repository. This paper describes the results obtained from the analyses of the sludge slurry samples taken from Macro-Batch 2 to meet the reporting requirements of the WAPS. Twenty eight elements were identified for the nonradioactive composition and thirty one for the radioactive composition. The reportable radioisotopes range from C-14 to Cm-246.

We demonstrate sensitive mass identification of model organic compounds directly from sodium nitrate by infrared matrix-assisted laser desorption and ionization mass spectrometry. Sensitivity limits of order 10−6 are achieved for resonant desorption, with a linear calibration verified for crown ether molecules over almost three decades in concentration. We also have observed that crown ether molecules can abstract atomic sodium from the sample when irradiated at the resonant frequency of the NO3 stretching vibrations around 7.1 μm. A model for this process is proposed which suggests great promise for this technique on untreated tank waste.

A cold test demonstration of the Geosafe® Non-Traditional In-Situ Vitrification technology (NTISV) was performed at a simulated adsorption bed by using commercial-scale equipment. A successful demonstration of NTISV at Los Alamos National Laboratory (LANL) is expected to provide a useful technology for remediation needs of EM-40 in the U.S. Department of Energy (DOT) complex on legacy buried waste containing organics and radionuclides in geological environments. Therefore, a simulated test bed including surrogates was prepared to evaluate the impact this technology has on immobilizing radionuclides. The total volume of the cold demonstration vitrification was 3.1 m × 4.6 m × 4.9 m resulting in 176,667 kg of glass or glassy constituents. A drill core sample was removed from the glass monolith and divided into 13 samples to gain spatial (vertical) information about (a) vitrification progress, (b) phase distribution, and (c) element distribution. Based upon optical microscopy, least square and Rietveld refinement of the X-ray diffraction data, the phase constitution was determined and a temperature profile estimated. The glass compositions as well as the concentrations of surrogates and trace metals were analyzed by using X-ray Fluorescence Analysis, Electron-beam Microprobe Analysis, and Scanning Electron Microscopy. In the main melting zone, in the depth range 3.7 m to 6.3 m, NTISV successfully vitrified the geological formation (Bandelier Tuffl Unit 2/3 of Tshirege) and convert it mainly into 98-99 wt.-% silicate-based glass and 1-2 wt.-% iron-based metal inclusions. Within this zone the glass compositions have been found to be nearly constant and the surrogate elements (cerium and cesium) are distributed homogeneously.

We have initiated a study of improved methods for implementing dry (AIROX) processing of commercial spent nuclear fuels for recycling back into light-water reactors. In this proliferation-resistant recycle, the spent fuel is converted to a powder, blended with fresh medium-enriched uranium powder, and refabricated into fuel elements. Evaluations of neutronic characteristics show that it will be necessary to remove a substantial portion of the neutronabsorbing fission products in the spent fuel, especially lanthanides and rhodium, in order to achieve efficient utilization of the spent fuel. We have already modeled oxidative vaporization of selected fission products from the powder at 1000 °C. In addition to permanent gases and fission products that are vaporized during pellet sintering, this can remove Tc, Mo, and some Ru and improve the neutronics. A number of approaches are being evaluated for removing lanthanides and rhodium, initially by thermodynamic modeling and review of literature. The lanthanides exist in solid solution with UO2 so separations methods will require conversion to fine powder; rhodium is present as a metallic inclusion in the epsilon phase. Chlorination of finely powdered oxide at 1100-1200 °C would vaporize substantial portions of Nd, Eu, Gd, and Rh. A fraction of the uranium would also vaporize; if significant, it could be recovered for recycle. Magnetic and electrostatic methods were evaluated for separation of lanthanide from spent fuel. They are not likely to be practical. However, static separation techniques may be applicable for removing rhodium in the fine powder metallic inclusions. These technicalconsiderations provide the basis for a suggested experimental program.

Decontamination for on-site maintenance during operation and for decommissioning of end-of-operation plants can be achieved using thixotropic corrosive gels to dissolve the radioactive deposit trapped at the surface of stainless steels. The dissolving agent is a corrosive solution of a Ce(IV) salt stabilized in concentrated nitric acid. Rheological properties of the gels are adjusted to give optimal spraying characteristics and surface adhesion using silica of small particle size and non-ionic surfactants. The formulation of the gels is such that they allow easy decontamination processing, and minimize the liquid and solid waste generated.

This paper describes the results of a joint research program of the Russian institutes at St. Petersburg, Krasnoyarsk and Zheleznogorsk with the Idaho National Engineering and Environmental Laboratory. A new “Gubka” (“sponge” in Russian) material was used to sorb and stabilize surrogate problematic actinide solutions, which contained lanthanide mixtures in nitric acid and tracer americium–241, by using repeated saturation-drying-calcining cycles. These tests resulted in maximum loading up to about 45 wt.% nitrate salts after drying and 33 wt.% oxides after calcination. The rates of americium–241 recovery were measured in 6 M nitric acid at 60°C. Gubka samples loaded with cerium and neodymium oxides were hot pressed at 29 MPa and 20-1000°C, resulting in a 35 % volume reduction.

The goal of this work is to assess the physical transport properties of Gd through an ion exchange column while determining the sorption properties of the resin. By coupling the physical transport with the chemical sorption, further insight into the behavior of the ion exchange resin can be gained. NMR imaging provides a powerful, non-destructive, means to extract spatial information from complex systems on a near real-time basis. An important example is liquid flow through granular media. With the use of a chemically reactive NMR contrast agent, the chemical speciation can be traced along the physical flow path of the granular media. In this study, trivalent gadolinium (Gd3+) was selected based on its chemical similarity to typical high-level waste components, 241Am and 244Cm, and for its paramagnetic contrasting abilities in NMR experiments. NMR imaging results of flow experiments are provided showing a characteristic flow phenomena and resin column loading profiles. ICP-AES data are provided to show resin ion exchange capacities (IECs) and breakthrough curves. The use of NMR imaging with a Gd3+ tracer will lead to a better understanding of the transport and sorption properties of these ion-specific resins. This technique can be applied to other complex flow systems such as environmental transport.

The melt-dilute treatment technology is being developed to facilitate the ultimate disposition of highly enriched Al-Base DOE spent nuclear fuels in a geologic repository such as that proposed for Yucca Mountain. Currently, approximately 28 MTHM is expected to be returned to the Savannah River Site from domestic and foreign research reactors. The melt-dilute treatment technology will melt the fuel assemblies to reduce their volume and alloys them with depleted uranium to isotopically dilute the 235U concentration and reduce the potential for criticality and proliferation concerns. A critical technology element in the development of the melt-dilute process is the development of an offgas system. Experimental tests using both cesium surrogates and radioactive cesium have shown that zeolite 4A is an effective gaseous cesium trap and as a result a preliminary offgas system concept has been developed employing dry zeolite 4A absorber beds as the primary cesium trapping medium. Final, validation of this offgas concept will occur during pilot-scale irradiated testing.

This report provides a summary of the results obtained for a limited variability study for glasses containing Crystalline Silicotitanate (CST), Monosodium Titanate (MST), and either simulated Purex or HM sludge. Twenty-two glasses containing Purex sludge and three glasses containing HM sludge were fabricated and tested. The fabricated glasses were tested for durability using the 7-day Product Consistency Test (PCT) and characterized by measuring the viscosity at 1150°C and by determining an approximate, bounding liquidus temperature. The current models used by Defense Waste Processing Facility (DWPF) for predicting durability, viscosity, and liquidus temperature were applied to all 25 glasses. The goal of this work was to identify any major-problems from a glass perspective, within the scope of this effort, which could potentially preclude the use of CST at DWPF.

Electrical conductivity of three simulated nuclear waste glass melts was measured over a wide range of temperatures, using a high-accuracy, calibration-free, coaxial-cylinders technique. Glass chemistries representing both low-activity and high-level wastes at Hanford and Savannah River sites were used. The temperature dependence of the conductivity of these glasses was presented. The conductivity values measured at 1150°C were compared to the conductivity values predicted from first-order mixture model.