bluuebunny/crossref_metadata_2025_split huggingface dataset:
Applications of Focused Ion Beams to Optoelectronic Device Fabrication - An Overview
Authors: Kubena Randall L. 
Published: March 2011

ABSTRACTFocused-ion-beam (FIB) technology has been applied during the past decade to a wide variety of device and circuit fabrication procedures. The ability to perform maskless implantation, selective sputtering and deposition, and high resolution lithography with a single system has allowed FIB researchers to explore a large number of unique fabrication processes for silicon, GaAs, and heterojunction devices. Currently, exploratory studies in advanced optoelectronic device fabrication employ the largest number of diverse FIB techniques. In this paper, the major application areas of FIB technology to optoelectronic research are reviewed, and possible uses of ultrasmall (≤500 Å) ion beams in the fabrication of optoelectronic device structures with novel properties are described.
Distribution of Boron Atoms in Ion Implanted Compound Semiconductors
Authors: Bowman R. C. , Knudsen J. F. , Downing R. G. , Kremer R. E. 
Published: March 2011

ABSTRACTThe nondestructive neutron depth profiling (NDP) technique has been used to measure the boron (10B) distributions in GaAs, CdTe, Hg0.7 Cd0.3 Te, and Hg0.85 Mn0.15Te after multiple energy ion implants. The NDP results are found to be in good agreement with the theoretical ion ranges obtained from Monte Carlo computer simulations. Only minor changes in the boron profiles were seen for the chosen annealing conditions.
X-Ray Rocking Curve Analysis of S-Implanted GaAs
Authors: Fatemi M. , Thompson P.E. , Chaudhuri J. , Shah S. 
Published: March 2011

ABSTRACTThe effect of rapid thermal annealing on strain reduction in 1.15 MeV S-implanted GaAs wafers irradiated to a dose of 5 × 1014/cm2 has been studied by double-crystal x-ray diffraction technique. X-ray rocking curves exhibit characteristic thin film fringes between the peak of unstrained GaAs and the major peak of the strained region. The maximum strain, i.e., the separation between the two peaks, as well as the number of minor fringes decreases with increasing RTA temperature, while the relative spacing between the fringes remains constant. At temperatures above 900°C, the main peaks begin to overlap; however, a residual positive strain can be measured for temperatures as high as 1100°C.
Characterization of Radiation Damage at the Nb Site in Natural Pyrochlores and Samarskites by X-Ray Absorption Spectroscopy
Authors: Greegor R. B. , Lytle F. W. , Chakoumakos B. C. , Lumpkin G. R. , Warner J. K. , Ewing R. C. 
Published: March 2011

ABSTRACTX-ray absorption spectroscopy has been used to investigate the Nb B-site in pyrochlores (A1.2B2O6Y0–1, Fd3m, Z=8) and samarskites (A3B5O16) in both metamict and annealed condition. The XANES and EXAFS measurements indicate significant changes in pyrochlore and smaller changes in samarskite as a result of radiation damage. In the metamict state the Nb site in both pyrochlores and samarskites is similar to Nb in Nb2O5. Short Nb-O (1.65Å) bonds are not disrupted by alpha-decay/recoil-nuclei events as much as longer bonds (2.00Å). This increases the asymmetry and static disorder at the local Nb site while long range order is greatly diminished resulting in considerable distribution in Nb-M distances and bond angles.
Characterization of Low-Gas-Release LHR Fuels by Transmission Electron Microscopy
Authors: Thomas L. E. , Guenther R. J. 
Published: March 2011

ABSTRACTAnalytical transmission electron microscopy (AEM) was performed on light-water reactor (LWR) spent fuel samples from different radial locations in several low-gas release fuels. The examinations revealed high densities of sub-micrometer gas and solid aggregates. Precipitate coarsening and a change in the nature of the fission gas precipitates was observed near the centers of the fuel pellets. The fission-product phases were: 1) ε-ruthenium (Mo-Ru-Tc-Rh-Pd solid-solution alloy); 2) unidentified gases in intra- and intergranular bubbles from fuel edge to mid-radius locations; 3) dense, highly pressurized xenon/krypton “particles” associated with intragranular °-phase near the fuel centers. All other fission products apparently remained in solution. The nature and distribution of these phases are likely to affect fuel behavior in waste repositories and may be important for understanding fission gas behavior.
Detailed Characterization of LWR Fuel Rods for the U.S. Civilian Radioactive Waste Management Program
Authors: Guenther Richard J. , Blahnik D. E. , Campbell T. K. , Jenquin U. P. , Mendel J. E. , Thornhill C. K. 
Published: March 2011

ABSTRACTCharacterizations are being conducted on light-water reactor (LWR) spent fuel with peak burnups from ∼31 to 46 MWd/kgM and rod-average fission gas releases of ∼0.3 to 11%. Measured concentrations of nuclides agreed within ∼10 to 15% of the predicted amounts. Radiochemistry of deposits on the cladding interior surface, fuel ceramography, beta/gamma autoradiog-raphy, and gamma scan data were consistent with fuel burnups and apparent operating temperatures. Microstructural observations also indicated increased porosity at the fuel center in high-release fuels and at the edge in high-burnup fuels. Cladding oxide thickness increased with exterior cladding temperatures as did the trend in hydriding.
Dissolution of UO2 at Various Parametric Conditions: A Comparison Between Calculated and Experimental Results
Authors: Olliia Kaija 
Published: March 2011

ABSTRACTThe solubilities of uranium measured in deionized water, in sodium bicarbonate solutions as a function of carbonate concentration and in two types of synthetic groundwaters have been compared with those predicted using the reaction pathway program, PHREEQE. All the measurements were carried out under both air-saturated, oxidizing and anoxic, reducing conditions.The experimental solubility values of uranium under oxidizing conditions are, in general, lower when compared to the corresponding theoretical ones calculated by PHREEQE. A critical factor is the choice of solubility constant for the solid phase. The reason for the lower solubility values may also be the mechanism of dissolution leading for example either to a situation where low dissolution rate is a limiting factor, or to formation of some solid phase of uranium with lower solubility. The experimental solubility values under reducing conditions appear to be in good agreement with the theoretical solubility values.
Creep Properties of Welded Joints in Copper Canisters for Nuclear Waste Containment
Authors: Ivarsson Bo , Österberg Jan-Olof , Sandström Rolf , Werme Lars 
Published: March 2011

ABSTRACTCopper canisters for nuclear waste containment can be expected to be exposed to temperatures up to 100°C. Since the material is pure copper, creep properties must be taken into account in particular for the welded joints in the canisters. In the paper creep rupture properties of parent metal, weld metal, and simulated heat affected zone are presented for 110°C. About ten times shorter rupture times were found for the weld metal in comparison to the parent metal. Cross weld specimens showed even shorter rupture times.
Basis for Selecting Cement-Based Waste Forms for Immobilizing Radioactive Waste
Authors: Mcdaniel Earl W. , Tallent Othar K. , Sams Terry L. , Delzer Dianne B. , Bostick William D. 
Published: March 2011

ABSTRACTResearch at Oak Ridge National Laboratory (ORNL) has shown that cement-based waste forms can be tailored to tolerate wide fluctuations in waste-feed compositions and still maintain properties that are compatible with standard equipment and produce a product that meets regulatory requirements. The 20-year operational history of ORNL's grouting program has demonstrated this resilience and reliability on a spectrum of waste feeds encompassing the developmental history of the nuclear fuel cycle.Cement-based materials are the most frequently used waste forms for solidifying non-high-level waste. These materials are also perhaps the most misused for immobilizing wastes. This paper discusses the scientific basis for selecting a suitable cement-based waste form to solidify and immobilize radioactive waste. For instance, when a given amount of cement (the source of calcium) is replaced with class F fly ash, the strontium leachability is reduced proportionally.This paper discusses the use of gel clays such as bentonite and attapulgite to absorb excess fluids and thus increase waste loading. Also, cement-compatible natural materials such as nonswelling clays are used to ion exchange or absorb soluble ions such as cesium and thus reduce leachability.Recently, blast furnace slag has been the subject of much interest in cement-based immobilization technology. Data are presented to show that a waste form containing blast furnace slag and other cementing materials is very effective in reducing the release rates of 99Tc.
Leaching Scale Effect for Cement-Waste Forms
Authors: Nomine J. C. , Billon A. , Courtois G. 
Published: March 2011

The confinement ability of a waste package is one of the major safety characteristics to consider in shallow land burial. In order to determine if the confinement is acceptable, in accordance with local policy, one way is to proceed to leaching tests. The practical method, for sake of simplicity, cost and time limit, is to carry out the leaching tests on laboratory samples which are easier to prepare than full-scale blocks, but the representativity of which needs to be treated with caution; it is in this context, that one of the aspect of our work concerns what is known as the “scale effect”.This study has been conducted using blocks the volumes of which are respectively of 200, 20, 2 and 0, 2 1, and made with the same cement-waste form (13 Cs) system.
Modelling of the Degradation of Cement in a Nuclear Waste Repository
Authors: Haworth A. , Sharland S. M. , Tweed C. J. 
Published: March 2011

ABSTRACTThe current UK concept for a low- or intermediate-level nuclear waste repository includes a largely cementitious backfill. The cement provides a high pH environment in which the general corrosion rate of the metal canisters is reduced and the solubilities of many nuclides low. It has previously been assumed that this high pH will exist for a period of 107 years, however cement will degrade due to leaching of the solid components and attack from aqueous species in groundwater. In this paper we describe the preliminary stages of a model of the degradation of cement in a repository. The modelling involves the incorporation of a thermodynamic description of cement[2] into the static code PHREEQE[5]. This is then used in a coupled chemistry-transport model of simple leaching of cement using the code CHEQMATE[4]. This preliminary modelling also provides a useful verification of CHEQMATE as a direct comparison with a THCCDM (a coupled code based on CHEMTRN) model is possible. Results from this preliminary model suggest that the fall in pH due to leaching is slow. The model is sufficiently flexible to form the basis of more detailed investigations of the effect of groundwater interactions on the degradation of cement.
Hygroscopic Water Uptake and Dry/Wet Cycling Compared with Normal Water Exposure of Cemented Waste Forms
Authors: Brodersen K. 
Published: March 2011

ABSTRACTConsiderable water uptake, formation of free liquid and leaching of activity may take place if cemented waste materials containing soluble salts are exposed to high humidity air. Another feature of importance for the structural stability of some materials is cyclic variations between water saturation and dry-out periods. The phenomena have been investigated on a laboratory scale using simulated waste in the form of cemented sodium nitrate or ion-exchange resins. Spherical samples are found to be valuable for simultaneous determination of the leaching of major components, mainly sodium, and the water uptake and swelling obtained from weight-increase measurements for the samples weighed in air and immersed in water.
Phase Development and Pore Solution Chemistry in Ageing Blast Furnace Slag-Portland Cement Blends
Authors: Macphee D. E. , Atkins M. , Glassar P. P. 
Published: March 2011

ABSTRACTBlast furnace slag (BPS) cement blends have certain advantages for the encapsulation of low and intermediate radioactive wastes. However, their hydration reactions are more complex than for Portland cement because Portland cements fully react within several years whereas slag hydration takes considerably longer. Empirical testing of blends matured for short periods, up to several years, are not therefore adequate to characterise the chemical immobilisation potential of such systems. A modelling approach is required to predict long—term properties.This paper is concerned with defining the internal environment within slag-cement matrices. The important parameters for definition are pH, Eh, solubilities and speciation in the aqueous phase, and solid phase equilibria. Compatibility studies in the CaO-CaSO4-Al2O-SiO2-MgO-H2O system are described and compared with the results of X-ray diffraction on pastes aged up to 2 years. In addition, pore water compositions for slag-rich cements are given, and the potential for predicting their redox level(Eh) by extending the present model is discussed.
Leaching of Chlorine, Cesium, Strontium and Technetium from Cement-Fixed Intermediate Level Liquid Waste
Authors: Brodda Bert-G. , Mingxia Xu 
Published: March 2011

ABSTRACTThe leaching behaviour of Cl-36, Cs-137, Sr-90 and Tc-99 from cement-fixed intermediate level liquid waste was investigated. Bentonite, MicroSili-ca®, acrylic resin and sodium sulfide were used as additives to a blast furnace cement matrix. In some cases portland cement was used. The liquid (water or waste solution) to cement ratio applied was mostly 0.47. Samples were leached with water or quinary brine (Q-brine).After 250 days the sequence of leachability from additive-free specimen in Q-brine was Cl>Cs>Sr>Tc with little modification in water: Cs>Cl>Sr>Tc. Cl and Cs leaching are diffusion-controlled, Sr and Tc are chemically fixed.Bentonite improves the retardation of Cs significantly, but has no significant effect on Cl, Sr and Tc. MicroSilica® deteriorates the retardation of Cs, but has no effect on Cl, Sr and Tc release.Sulfide was expected to reduce the leachability of Tc by forming insoluble TC2S7. No significant effect was observed, however, because either TC2S7 does not form at the pH value of the cement slurry or, due to a solubility competition, insoluble Tc species also form in the absence of sulfide ions.Acrylic resin does not reduce the leachability of Cs significantly. Irradiation with doses up to 4.4–105 Gy slightly deteriorates Cs retardation.Applying a liquid to cement ratio of 0.35 instead of 0.47 improves the retardation of Cs by a factor of 1.5–4, which may be due to a reduction of porosity in the specimen.
Safety Analysis of a Sub-Seabed Repository of HLW
Authors: Karpf Armin-Dietmar 
Published: March 2011

ABSTRACTThis national safety study of a repository in the Atlantic Ocean uses as much input from German research work as possible. It is based on a computer code IMPONADOR which can be run on a personal computer. It also introduces a new way of presenting information about radionuclide confinement to individual compartments to facilitate the understanding of release processes. The work is part of the international cooperation within the Seabed Working Group of the OECD/NEA.
Melodie : A Code for Risk Assessment of Waste Repositories in Deep Geological Formations
Authors: Lewi J. , Mejon-Goula M. J. , Goblet P. , Cernes A. 
Published: March 2011

ABSTRACTIn order to perform the safety evaluation of nuclear waste repositories, a global model, called MELODIE, is currently being developed at the CEA/IPSN, in collaboration with other CEA teams and ENSMP (Ecole Nationale Supérieure des Mines de Paris). The version now in operation allows the assessment of the radiological consequences due to a repository located in a granitic formation for a period of several hundred thousands of years.The calculations are based on models which represent the following physical and chemical phenomena:– release of the radionuclides from the waste form and through engineered barriers,– radionuclide transport through the geosphere,– radionuclide behaviour in the biosphere.Three separate models have been developed for each of these subjects; they are integrated in the code through a modular flexible dataprocessing structure.In addition, a sensitivity and uncertainty analysis algorithm has been implemented into the code, which allows the evaluation of the influence of the parameter values on the result and of its uncertainty.After a quite general description of MELODIE, we intend to present what has been done with it, mainly the calculations performed in the PAGIS (CCE) exercise: global dose calculations and ranking of the most important parameters through the sensitivity analysis.We also note the studies performed with the geosphere module of MELODIE (METIS), especially the participation to the HYDROCOIN (0ECD/NEA) exercise. In addition, the main future development axes of MELODIE are outlined.
In-Situ Stress Measurement - Results of Experiments Performed at the Asse Salt Mine - Federal Republic of Germany
Authors: Feddersen H.-K. 
Published: March 2011

ABSTRACTHigh-level nuclear wastes are heat generating wastes. Heat will be transferred to the surrounding salt formation. This heating of the host rock will result in an increased temperature and in stress changes.From 1983 through 1985 two underground tests were conducted in the Asse Salt Mine (Federal Republic of Germany) in which among others thermally induced stress changes were investigated.In the Temperature Test 6 which was conducted at the 750 m-level only electrical heaters were used to simulate the heat generating high-level waste. Three hydraulic stress monitoring stations were arranged in the un-heated area at a distance of about 40 meters to the central heater. Measurements using AWIO flat jacks were also carried out in the heated region of the Temperature Test Field 6 (TVF 6).The joint US/FRG Brine Migration Test (BMT) - a nuclear waste repository simulation experiment - was performed at the 800 m-level and used radioactive sources and electrical heaters to impose the heat load on the host rock. Stress change measurements during this experiment were performed using hydraulic pressure cells and straingaged stressmeters.During 1985 an experimental area for the first in-situ test disposal of high-level radioactive sources was created at the 800 m-level within the Asse anticline structure. Thirty high-level radioactive canisters will be emplaced in six down-boreholes located in two test galleries in the experimental area. There is one additional borehole in each gallery equipped only with electrical heaters. Hydraulic pressure cells, AWID flat jacks, and straingaged stressmeters were installed in the HAW-test field for stress change measurements.
Retrieval and Analysis of Simulated Defense HLW Package Experiments at the WIPP
Authors: Molecke Martin A. , Rob Sorensen N. 
Published: March 2011

ABSTRACTIn situ waste package performance experiments involving simulated (non-radioactive) defense high-level waste (DHLW) containers have been in progress since late 1984 at the Waste Isolation Pilot Plant (WIPP) facility. These experiments involve full-size, simulated DHLW containers of several metals and designs emplaced in the WIPP bedded rock salt. These test containers are surrounded by granular backfill (packing) materials, have in many cases been intentionally injected with brines, and are heavily instrumented. A majority of the test packages also contain nonradioactive DHLW borosilicate glass waste form, either within the container and/or outside of it. The primary purpose of these WIPP simulated DHLW experiments is to evaluate the in situ durability and performance of all waste package engineered barrier materials, and to perform package concept validation testing.Twelve of the test DHLW containers, emplaced in WIPP test Room B, have been in heated operation since 1985 and had a maximum surface temperature of about 190°C. These containers were recently retrieved, after about 3 years of heated exposure, for detailed posttest laboratory analyses of: general corrosion and metallurgical degradation, waste form and backfill materials alterations, and other rock salt-brine-barrier materials near-field interactions with the “repository” geochemical environment. Test canisters and overpacks made of ASTM Grade-12 titanium showed essentially no visible degradation in either the base metal or welds; cast mild steel A216/WCA over-packs have suffered some uniform corrosion. Significant degradation of the removed instruments and associated test apparatus has been found: pieces of stainless steel (both 304L and 316) apparatus have undergone extensive stress-corrosion cracking failure and non-uniform attack; Inconel 600-sheathed instruments have undergone both extensive uniform and localized (pitting) attack. Granular backfill materials have been significantly compacted by creep closure to about a density of 2 kg/m. Laboratory analyses are still in progress. Further details on these materials results plus instrumentation data and other in situ WIPP waste package test observations are discussed.
Mechanical Properties of Swelling Clays
Authors: Broc D. , Plas F. , Robinet J. C. 
Published: March 2011

ABSTRACTThe safety of vitrified radioactive waste disposal in granite is based on the concept of multiple barriers, which include an engineered clay barrier placed between the waste package and the granite. The mechanical properties of the swelling clays used were studied with a view to practical application for storage facility dimensioning. This involved a macroscopic examination of the clays swelling capacities (for sealing of storage boreholes) and fracture criteria (mechanical stability).
Natural Analogue Study of Volcanic Glass – A Case Study of Basaltic Glasses in Pyroclastic Fall Deposits of Fuji Volcano, Japan –
Authors: Arai T. , Yusa Y. , Sasaki N. , Tsunoda N. , Takano H. 
Published: March 2011

ABSTRACTA considerable range in alteration rates of basaltic glasses in various environments has been reported in previous studies. However, these studies paid only cursory attention to the environmental conditions under which the glass had been altered. In this study, the alteration of basaltic glasses was investigated and the environmental conditions and the alteration rate were discussed. Two sample ages were represented: 280 years and 2800 years. Basaltic glasses and their alteration layers were analyzed by electron probe microanalyzer (EPMA) and the thickness of the alteration layers were measured by scanning electron microscope (SEM). The ground water collected near the sampling point of Zunazawa Scoria and the pore water of both samples were analyzed. The alteration temperature and flow rate of water are estimated to be about 13t and 0.2 1/cm2/y respectively on the basis of meteorological data. The alteration layers of young aged basaltic glasses in freshwater conditions are similar to those of leached borosilicate glasses. The alteration rates of these basaltic glasses are estimated to be several um/1000y. The elemental concentrations in the ground water can be roughly explained as the result of leaching of the glasses.
Diffusion of Chloride and Uranium in Compacted Sodium Bentonite
Authors: Muurinen Arto , Pemtilä-Hiltunen Pirkkd , Uusheimo Kari 
Published: March 2011

ABSTRACTSorption and diffusion of chloride (Cl-36) and uranium in conpacted sodium bentonite MX-80 were measured. No sorption was observed in the Sorption tests, however, in the diffusion tests slight sorption of uranium was noticed. The diffusivities of Cl-36 were found to be strongly dependent on the compaction of bentonite and on the salt concentration of the solution. Ion-exclusion can propably explain these phenomena.The diffusivities of uranium were also strongly dependent on the compaction of bentonite. Uranium shews features of both ion-exclusion and sorption. Farther studies are, however, needed to explain the diffusion mechanisms of uranium.