bluuebunny/crossref_metadata_2025_split huggingface dataset:
Use of Fly-Ash for Sealing a Radioactive Waste Repository
Authors: Ollagnier Marc , Tauziede Christian , Olivier Jodl 
Published: March 2011

ABSTRACTThe mining industry currently uses fly-ash from coal-fired power-plants to close access shafts in abandoned mines, in sedimentary formations. The technique consists of installing a plug of fly-ash, about fifty meters high, at the base of the shaft-lining. In this study, fly-ash is considered as a possible sealant for radioactive repositories in sedimentary formations, as well as in hard rock.Specific studies have been conducted in order to assess the feasibility of this technique, and to improve the long-term performance of the fly-ash seals. The hydraulic and mechanical characteristics of fly-ash, taken from the plug of an abandoned shaft, as well as from dumps of various ages, were measured. Laboratory tests showed that it is possible to reduce the hydraulic conductivity of fly-ash by a factor of thousand, with the addition of ten percent bentonite. Moreover, it seems possible to block fine fissures in the surrounding damaged rock by injecting grouts made of cement and fly-ash having maximum diameters of twenty micrometers.
Crevice Corrosion - NaCI Concentration Map for Grade-2 Titanium at Elevated Temperature
Authors: Tsujikawa Shigeo , Kojima Yoichj 
Published: March 2011

ABSTRACTThe repassivation potential, ER, for metal/metal-crevice of Commercially Pure Titanium, C.P.Ti, was determined in NaC1 solutions at temperatures up to 250C. The ER has its least noble value near 100C and becomes more noble as the temperature increases. As shown in previous research[1], the shrinkage of the repassivation region should continue with increasing temperatures. However, in conducting this same experiment at temperatures higher than 100C, an examination of the NaCI concentration - temperature - crevice corrosion map verifies that the repassivation region began to expand again when the temperature exceeded 140C. This expansion continued as the temperature continued to increase.
Corrosion of Stainless Steel for HLW Containers Under Gamma Irradiation
Authors: Osada K. , Muraoka S. 
Published: March 2011

ABSTRACTThe corrosion behavior of type 304 stainless steel was studied under gamma irradiation as part of the evaluation for the long-term durability of high-level radioactive waste (HLW) disposal containers. Gamma rays, generated from fission products in high-level radioactive waste, are considered to change the environment around the canisters and overpacks. The redox potentials for NaCl solutions and corrosion potentials of stainless steel were measured to consider the effects of gamma irradiation, by using an electrochemical method. The pitting potentials of stainless steel for NaCl solutions were also measured to examine the pitting corrosion under gamma irradiation. As a result of this experiment, it is concluded that the oxidizing properties as a result of the formation of H2O2 and H2 produced by gamma irradiation depended on the concentration of Cl−, and that the strength of oxidizing properties of 1M (mol·dm−3) NaCl solution was particularly high. and the pitting corrosion was found for 1M NaCI solution under gamma irradiation at the dose rate of 2.6×102 C/kg·h (1.0×106 R/h) at 60°C, by using an electrochemical method.
Effects of Chloride, Bromide, and Thiosulfate Ions on the Critical Conditions for Crevice Corrosion of Several Stainless Alloys as a Material for Geological Disposal Packages for Nuclear Waste
Authors: Nakayama Guen , Wakamatsu Hisao , Akashi Masatsune 
Published: March 2011

ABSTRACTIn addition to mild steel, several stainless alloys are being proposed as materials for packages for geological disposal of high-level nuclear waste. When buried deep underground, the greatest detriment to the integrity of packages made of these alloys is localized corrosion, for which critical conditions for initiation of crevice corrosion in chloride environments, with or without other ions, need be precisely known.Crevice corrosion behavior of Type 304 stainless steel, Type 316 stainless steel, Alloy 825, Ti-Gr.1, and Ti-Gr.12 in solutions containing ions of chloride, bromide (these two for their ordinary presence in natural waters), or thiosulphate (this for the likelihood of microbially influenced corrosion) to varying concentrations have been empirically examined. All of these alloys exhibit much the same concentration dependency of crevice corrosion sensitivity for chloride and bromide ions, while Type 304 stainless steel is particularly sensitive to the thiosulphate ion. The region of insensitivity for chloride ion is wider in the increasing order of Type 304 stainless steel, Type 316 stainless steel, Ti-Gr. 1, and Ti-Gr. 12, with that of Alloy 825 lying somewhere in between.
Use of Sputter-Deposited 316L Stainless Steel Ultrathin Films for Microbial Influenced Corrosion Studies
Authors: Suci P.A. , Pedraza A.J. , Godbole M.J. , Geesey G.G. 
Published: March 2011

ABSTRACTUltra thin films (12nm) were sputter deposited onto cylindrical germanium internal reflection elements pre-coated with a thin (2nm) layer of Cr2O3. Two crystals were inserted into Circle cell flow-through chambers and mounted on the optical bench of an Fourier Transform Infrared (FT-IR) spectrometer. One chamber was maintained as a sterile control while the other was sequentially inoculated with four bacterial species:Pseudomonas aeruginosa, Bacillus subtillis, Hafnia alvei, andDesulfovibrio gigas, in that order. The water absorption band (1640cm−1) was monitored and used to follow the deterioration of the ultra thin films. In this respect, the sterile control and inoculated films exhibited only slight differences during the 1000h course of the experiment. Assay of the visible biofilm that had accumulated on the surface of the inoculated crystal after 1000h revealed that the film incorporated viable cells from all four strains.
Dealing with Uncertainty in the Chemical Environment in Bentonite Backfill
Authors: Arthur R.C. , Apted M. J. , Conca J. L. 
Published: March 2011

ABSTRACTAnalytical and conceptual deficiencies in understanding compositional variability in the smectite clays are expected to generate uncertainty in models used to simulate the chemical environment in bentonite backfill. Equilibrium models accounting for nonstoichiometry in smectite can nevertheless bound ranges in aqueous solution compositions that are an explicit function of the uncertainty in smectite compositions. In one approach, we quantify uncertainty in terms of ranges in concentrations of octahedral and tetrahedral Al, and exchange-site cations and vacancies. Heterogeneous mass transfer in bentonite-water systems is modeled using conventional mass-action relations and standard Gibbs energies for stoichiometric minerals, and the site-occupancy constraints combined with site-mixing relations for smectite. The resultant bounding conditions in groundwater compositions may be large or small depending on which aqueous species are of interest in a given situation, but they are valid irrespective of whether equilibrium in smectite-water reactions is attained or is inhibited by slow intracrystalline reaction rates.
Diffusion of Uranium in Compacted Sodium Bentonite
Authors: Muurinen Arto , Ollila Kaija , Lehikoinen Jarmo 
Published: March 2011

ABSTRACTIn this study the diffusion of uranium dissolved from uranium oxide fuel was studied experimentally in compacted sodium bentonite (Wyoming bentonite MX-80). The parameters varied in the study were the density of bentonite, the salt content of the solution and the redox conditions. In the studies with non-saline water of total dissolved solids about 300 ppm, uranium was both in aerobic and anaerobic experiments as anionic complexes and followed the anionic diffusion mechanism. Anion exclusion decreased effective diffusion coefficients, especially in more dense samples. In the studies with saline water of total dissolves solids about 35000 ppm, uranium appeared in the aerobic experiments probably as cationic complexes and followed the cationic diffusion mechanism. Uranium in the saline, anaerobic experiment was probably U(OH)4 and followed the diffusion mechanism of neutral species.
Compared Study of Radiolysis-Induced Gas Liberation in Rocksalt from Various Origins
Authors: Akram N. , Blanchard J.C. , Gaudez M.T. , Toulhoat P. 
Published: March 2011

ABSTRACTThe evolution of impure rocksalt samples under irradiation has been investigated and compared to the evolution of pure halite, concerning the radiolytic release of gases. Four different types of rocksalt (pure halite, anhydrite-bearing, sylvinite-bearing, and marneous) have been irradiated with spent fuel at doses ranging from 104Gy to 107Gy, under synthetic air or helium at 50°C, 150°C, and 200°C. Marneous and anhydrite-bearing salt produced higher amounts of gas than the two other types, especially for H2, CO, and CH4. Radiolytic gas production essentially originates from the decomposition of traces of different kind of impurities : fluid inclusions, which can be composed of brine and/or organic fluids and gases; minerals such as carbonates, sulfates, or hydrated minerals; organic matter, in general (kerogen). Pure halite decomposes only at very high doses (more than 107Gy) giving off some Cl2 and other corrosive gases. Gas production increases steadily with dose for H2, CO, and CH4, contrary to what is observed for the pure Asse salt, where H2 and CO, respectively, peak at 106, and 105 Gy. The increase, with dose, in the production of CH4, C2H6, and C3H8 is far higher for marneous salt and higher for anhydrite-bearing salt. The decrease of the O2/N2 (oxygen consumption) ratio in the irradiation atmosphere is in the following order: marneous>anhydrite>halite=sylvinite. The overall effect of γ-irradiation on impure rocksalt can be described as follows : destruction of organic matter that produces chemically reduced gases (H2, CO, hydrocarbons), after an initial stage of oxygen consumption, which is reached more or less quickly, depending on the availability and quantity of organic matter. The use of impure, i.e, organic matter-bearing salt, as host-rock or engineered barrier for nuclear wastes, may be questioned, because of the high amount of gases produced in the immediate vicinity of the container (explosion hazards, pressure build-up). The present study will be followed by a modelling of the in-situ radiolytic gas production, taking into account: 1) the dose distribution with time, 2) the temperature distribution with time, and 3) the geometry of the repository, and will allow us to estimate the potentiel safety impact of radiolytic gas production in the case of a repository emplaced in impure rocksalt.
Effects of Air Oxidation on the Dissolution Rate of LWR Spent Fuel
Authors: Gray W. J. , Thomas L. E. , Einziger R. E. 
Published: March 2011

ABSTRACTDissolution rates for air-oxidized spent fuel were measured in flowthrough tests where U concentrations were kept well below the solubility limit. Results from two types of specimens, separated grains and coarse particles, both in oxidized (U4O9+x) and unoxidized (UO2) conditions indicated only minor effects of oxidation on the surface-area-normalized rates. Similar results were obtained for unirradiated specimens in three different oxidation states (UO2, U3O7, and U3O8). These observations have important practical implications for disposal of spent fuel in a geologic repository as well as implications regarding the oxidative dissolution mechanism of UO2fuel.
The Alteration of Uraninite to Clarkeite
Authors: Finch Robert J. , Ewing Rodney C. 
Published: March 2011

ABSTRACTThe oxidative alteration of uraninite by alkali-bearing, hydrothermal (200-400° C) solutions in granitic rocks produces the rare mineral clarkeite. The general formula for Na-clarkeite is{(Na,K)6-y-z(M2+PM3+qM4+r)y(□1+z-xPbx)}[UO2)7-xOt0]·nH2O. z ≈ p+2q+3r.The terms in square brackets designate the sheet structure; the other elements and vacancies (0) occur in interlayer sites. Clarkeite can accommodate Ca, Sr, Ba, Y, Th, and lanthanides. Thus, clarkeite is a potential actinide and fission product host. The replacement of uraninite by clarkeite occurs in the solid state and results in the loss of one-half of the uranium from uraninite. As U decays to Pb, the radiogenic Pb enters interlayer vacancies in clarkeite. This destabilizes the structure and clarkeite eventually decomposes to wölsendorfite, (Pb,Ca)2U2O7·2H2O, or curite, Pb3U8O27·3H2O.Clarkeite from Spruce Pine, North Carolina is zoned compositionally, with a K-rich core (atomic ratio K:Na:Ca = 1.0 : 0.11 : 0.12) surrounded by Na-clarkeite (Na:K:Ca = 1.0 : 0.03 : 0.03), rimmed and veined by Ca-clarkeite (Ca:Na:K = 1.0 : 1.25 : 0.05). Volumetrically, Na-clarkeite is the most important constituent. The chemical zoning reflects the disparate chemistries of K+, Na+, and Ca2+. Na-clarkeite and Ca-clarkeite are structurally similar, but have different sheet structures. The solubility of K in Na-clarkeite is less than five mole percent and is due to the different sizes of Na+ and K+ ions. The K-phase may not be related structurally to clarkeite.Exposed to ground water at low temperatures, clarkeite alters to the uranyl silicates, uranophane and kasolite. The alkalis are leached by ground water. The fate of Th, Y, and lanthanides from clarkeite is uncertain.
Kinetic Studies of Natural Uranium Minerals for the Long-Term Evolution of Spent Nuclear Fuel Under Oxidizing Conditions
Authors: Casas Ignasi , Cera E. , Bruno J. 
Published: March 2011

ABSTRACTThe time scale of spent fuel dissolution studies is of the order of magnitude of 2 to 10 years, while the performance of a spent fuel repository should be assessed for much longer times (105-106 years). These time scales can be bridged using appropriate natural analogues. Among other important information, the study of natural systems can give insight of which can be the oxidative alteration of spent fuel in granitic environments. However, in studying such systems, thermodynamic and kinetic data of relevant natural solid phases are needed.In this work we present preliminary results of dissolution experiments carried out under oxidizing conditions with selected and well characterized natural samples of the alteration chain of uraninite (i.e., uraninite, schoepite, uranophane).The experiments have been performed using a synthetic granitic groundwater as a leachant, in contact with air and at 25 °C.
A Modelling Study of the Effect of Rock Alteration on the Redistribution of Uranium
Authors: Murakami Takashi , Kimura Hideo 
Published: March 2011

ABSTRACTA modelling study was carried out to understand the effect of rock alteration on uranium radionuclide concentrations in rocks, in the vicinity of the Koongarra ore deposit, Australia. The one-dimensional, advection-dispersion-sorption model considers two important factors resulting from the process, mechanism and rate of chlorite weathering, one type of rock alteration that has occurred over a million-year period; (a) the changes in the distribution coefficients of uranium and thorium over time, and (b) that in rock porosity. The distribution coefficient and rock porosity at a given time, are expressed as average values of those of the coexisting minerals. By assuming a Darcy velocity of 0.9 m/y and an initial uranium concentration in the groundwater of 0.5 ppm, similar to present-day values, we were able to derive, from the finite-element modelling, uranium concentrations in the rock which are in good agreement with observed values. The calculated values of 230Th/234U activity ratios show a similar trend to those observed, although more experimental data are necessary to confirm the similarity. However, in the absence of the rock alteration, the calculated results did not agree with those observed. The present study suggests that rock alteration should be included in models used to predict uranium migration over long, geologic timescales.
The Long-Term Performance of Nuclear Waste Forms: Natural Materials - Three Case Studies
Authors: Ewing Rodney C. 
Published: March 2011

ABSTRACTNatural materials may be used to advantage in the evaluation of the long-term performance of nuclear waste forms. Three case studies are presented: (I) radiation effects in ceramic waste forms; (II) corrosion products of U02 under oxic conditions; (III) corrosion rate of nuclear waste glasses. For each case, a natural phase which is structurally and chemically analogous to the waste form is identified and used to evaluate the long-term behavior of a nuclear waste form. Short-term experimental results are compared to the observations made of analogous natural phases. The three case studies illustrate that results may range between providing fundamental data needed for the long-term evaluation of a waste form to only providing qualitative data of limited use. Although in the most rigorous view the long-term behaviour of a phase cannot be predicted, the correspondence between short-term experimental results and observations made of natural phases provides confidence in the “predicted” behavior of the waste form. The strength of this approach rests with the degree to which a mechanistic understanding of the phenomenon is attained.
Modeling Surface Area to Volume Effects on Borosilicate Glass Dissolution
Authors: Bourcier William L. , Ebert W. L. , Feng X. 
Published: March 2011

ABSTRACTWe simulated the reaction of SRL-131 glass with equilibrated J-13 water in order to investigate the effects of surface area to volume ratio (SAN/) on glass dissolution. We show that glass-fluid ion exchange causes solution pH to rise to progressively higher values as SA/V increases. Because the ion exchange is rapid relative to the duration of the glass dissolution experiment, the pH effect does not scale with (SA/V)*time. Experiments compared at the same (SA/V)*time value therefore have different pHs, with higher pHs at higher SA/V ratios. Both experimental data and our simulation results show similar trends of increasing reaction rate as a function of SAN ratio when scaled to (SA/V)*time. Glasses which react in systems of differing SA/V ratio therefore follow different reaction paths and high SAN ratios cannot be used to generate data which accurately scales to long time periods unless the ion exchange effect is taken into account. We suggest some simple test designs which enable more reliable high SAN accelerated tests.
Observation by Sem/edxs of Leached Layers on SRL-165 Glass from 5-Year MIIT Test
Authors: Macedo Pedro B. , Buechele Andrew C. 
Published: March 2011

ABSTRACTLong-term tests on the durability of nuclear waste glasses, both under laboratory conditions and in actual burial environments, have shown that after periods of several months or years the rates of glass corrosion can show unexpected increases. In burial tests in brine environments the growth of the thickness of the surface layer during years 3 to 5 as revealed by SIMS and SEM measurements was found to be faster than expected on the basis of findings during the first 2 years. In this paper we report on our SEM-EDXS observations of a 5 year burial sample of SRL-165 glass from the MIIT test, and compare our findings to those of other studies, especially SIMS measurements. Our average thickness measurement for the leached layer at the glass-metal interface is 3.70 μm, but wide variation is observed, as well as evidence of partial loss of the leached layer in some areas. A thicker, better retained layer is observed on the glass-brine interface near the core heater averaging 8 to 12 μm in thickness with occasional local attack producing leached layers of greater depth, in one case nearly 40 μm. These findings show that it is important to extend durability tests to long periods in order to establish a reliable basis for predictive modeling of long-term glass durability.
The Potential Role of Diffusion-Induced Grain-Boundary Migration in Extended Life Prediction
Authors: Handwerker C.A. , Blendell J.E. , Interrante C.G. , Ahn T.M. 
Published: March 2011

ABSTRACTThe selection of materials that are suitable for various high-level waste-packaging designs must reflect the need to meet requirements for long-term performance in repository environments that change with time. With this in mind, we examine how grain boundaries in materials are induced to migrate as a result of solute diffusion even at low temperatures, how the composition of the matrix material is changed significantly by this diffusion-induced grain boundary migration (DIGM), and how the changing microstructures and compositions during DIGM lead to major changes in materials performance, such as corrosion or embrittlement. Methods are discussed for prediction of the long-term behavior of materials affected by DIGM.
The Engineered Barrier System: Performance Issues
Authors: Pigford T. H. 
Published: March 2011

ABSTRACTThe engineered barrier system (EBS) for geologic disposal of high-level waste may typically consist of the waste solid, a container, and a porous backfill. Whether the surrounding rock is part of the EBS is arguable, but the properties of the surrounding rock strongly affect the EBS performance. The EBS functions and performance issue are addressed.
Progress in Waste Package and Engineered Barrier System Performance Assessment and Design
Authors: Van LUIK Abraham , Stahl David , Harrison Diane 
Published: March 2011

ABSTRACTAs part of the U.S. Department of Energy's evaluation of site suitability for a potential high-level radioactive waste repository, long-term interactions between the engineered barrier system and the site must be determined. This requires a waste-package/engineered-system design, a description of the environment around the emplacement zone, and models that simulate operative processes describing these engineered/natural system interactions. Candidate designs are being evaluated, including a more robust, multi-barrier waste package, and a drift emplacement mode. Tools for evaluating designs and emplacement mode are the currently available waste-package/engineered-system performance assessment codes developed for the project. For assessments that support site suitability, environmental impact, or licensing decisions, more capable codes are needed. Code capability requirements are being written, and existing codes are to be evaluated against those requirements. Recommendations are being made to focus waste-package/engineered-system code-development.
Key Issues Influencing the Performance of Near-Field Barriers
Authors: Tasker P.W. , Wisbey S.J. , Boyle C.B. 
Published: March 2011

ABSTRACTIn developing a national strategy for the disposal of radioactive wastes, each country will consider isolation options that are appropriate to the types of waste that are produced. The options are developed in response to specific national regulatory requirements, and thus will focus on different aspects of performance. However, there are a number of technical concerns that are common to all programmes. The major issues concerning the behaviour of the chemical and physical barriers in the near field of a radioactive waste repository are discussed in this paper.The description of key issues has been divided into the following categories: barrier design, barrier evolution, scientific understanding through modelling, and validation of performance. The near-field barriers are selected and designed to provide appropriate radionuclide containment and control. Factors affecting the evolution of these barriers, such as container degradation and gas generation, determine the subsequent release of radionuclides to the human environment. Modelling repository evolution is therefore an integral feature of performance assessments, and issues such as the treatment of inhomogeneities and non-equilibrium chemistry may need to be addressed. However, the use of mathematical and computer models implies a requirement for validation. The use of demonstration experiments and natural analogues builds confidence in the predictions of repository performance models, and provides a degree of validation for otherwise inaccessible timescales.