bluuebunny/crossref_metadata_2025_split huggingface dataset:
Quantitative Measurement of Boron and Phosphorus in Solar Grade Silicon Feedstocks by High Resolution Fast-Flow Glow-Discharge Mass Spectrometry
Authors: Putyera Karol , Su Kenghsien , Liu Changhsiu , Hockett R. S. , Wang Larry 
Published: April 2011

AbstractThe calibration factors are examined for determination of boron (B) and phosphorus (P) in solar grade silicon samples in the new generation of high resolution fast-flow glow-discharge mass spectrometers (FF-GDMS). It is shown that using the generalized calibration factors from the Standard RSF table, the relative errors observed in the determination of these analytes is not acceptably small for photovoltaic applications. The certified B and P values in NIST SRM 57a Si metal sample do not have the confidence intervals, which would be adequate for refining these calibration factors to the acceptable levels. Thus, well-characterized single crystalline Si wafers with known boron and phosphorus contents traceable to NIST reference materials (SRM 2133 and SRM 2137) were used for accurate calibrations and for establishing good analytical procedures for measurements of wide variety of Si sample forms. The obtained results for these important analytes using the new FF-GDMS procedure are compared to other characterization techniques commonly used in this industry.
Computer Simulation of Edge Effects in a Small-Area Mesa N-P Junction Diode
Authors: Appel Jesse , Sopori Bhushan , Ravindra N.M. 
Published: April 2011

AbstractThe influence of edges on the performance of small-area solar cells is determined using a modified commercial, finite-element software package. The n+/p mesa device is modeled as having a sub-oxide layer on the edges, which acquires positive charges that result in development of an electric field within the device. Our computer simulations include generation/ recombination at the diode edges as well as the influence of light on the recombination characteristics of the edges. We present a description of our model, dark and illuminated characteristics of devices with various surface charge concentrations, and the dynamics of carrier generation/recombination. The influence of edge geometry on diode performance is determined.
FASST® Reactive Transfer Printing for Morphology and Structural Control of Liquid Precursor based Inorganic Reactants
Authors: Hersh P. , Taylor M. , Sang B. , van Hest M. , Nekuda J. , Miedaner A. , Curtis C. , Leisch J. , Ginley D. , Stanbery B.J. , Eldada L. 
Published: April 2011

AbstractField-Assisted Simultaneous Synthesis and Transfer (FASST®) process offers a controllable and cost-effective method to produce Copper Indium Gallium Selenide (CIGS) films for high efficiency photovoltaic devices. In the first stage of the two-stage FASST® process two separate precursor films are formed, one deposited on the substrate and the other on a reusable printing plate. In the second stage, the precursors are brought into intimate contact and rapidly reacted under pressure in the presence of an applied electrostatic field, effectively creating a sealed micro-reactor that ensures high material utilization efficiency, direct control of reaction pressure, and low thermal budget. The unique ability to control both precursor films independently allows for composition and deposition technique optimization, eliminating pre-reaction prior to the synthesis of CIGS. This flexibility has proven immensely valuable as is demonstrated in the results of depositing the two-reactant films by various combinations of low-cost solution-based and conventional vacuum-based physical vapor deposition techniques, producing in several minutes' high quality “hybrid” CIGS with large grains on the order of several microns. Cell efficiencies as high as 12.2% have been achieved using the FASST® method.
Materials aspects of advanced repository concepts for higher toxicity waste
Authors: McKinley Ian G. , Scourse Ellie M. 
Published: April 2011

AbstractMost designs of disposal facilities for higher toxicity radioactive wastes were developed for generic feasibility assessment and hence tend to be rather simple in terms of layout and conservative in the choice of engineered barrier materials. Recently, there has been a trend to reassess such designs in the light of moves towards implementation, where robustness in terms of post closure safety needs to be balanced against the requirements to assure safety, quality, and economic practicality during operation, to minimise environmental impact and to gain public acceptance. Studies to date have, however, tended to focus predominantly on variants of layout and engineered barrier geometry. Evaluation of alternative materials has received less consideration, despite the huge advances in materials science over the last couple of decades and likely further developments in the time until repositories become operational. This paper will examine the constraints that led to the choice of the most common materials selected for barriers within “wet” host rocks and examine the extent to which performance could be optimised by use of alternative materials.
A Quantitative Risk Assessment Perspective on the Source Term for the Proposed Nuclear Waste Repository at Yucca Mountain
Authors: Murphy William M. , Garrick B. John , Kirstein Bruce E. 
Published: April 2011

AbstractA quantitative risk assessment methodology is outlined for the source term for the proposed geologic repository for high-level nuclear waste at Yucca Mountain, Nevada. The methodology involves construction of a logical event tree to identify scenarios for the success or failure of the waste isolation system. Uncertainties in the intervening events between initial conditions and consequences can be quantified using probability distributions for steps along each scenario path. Likelihood is quantified according to the frequencies for each split fraction in the event tree. The result is a calculated probability of frequency curve for the release rate(s) of the modeled radionuclide(s).
Liquidus Temperature and Crystallization Behavior of US Waste Glasses Investigated at KRI
Authors: Aloy Albert S. , Trofimenko Alexander V. , Belov Valery Z. , Marra James C. , Fox Kevin M. , Peeler David , Vienna John D. , Kim Dong-Sang 
Published: April 2011

AbstractAs a part of the optimization study for achieving the highest possible Hanford and Savannah River Site waste loading into acceptable borosilicate glasses, thirty glass compositions were selected for testing at KRI. These thirty test matrix glasses were designed to augment existing data within the compositional regions of interest with relatively high concentration of Al2O3 between 10 and 20 wt%.This paper reports experimental data on liquidus temperature (TL) and crystallization behavior of all synthesized glasses as well as durability of quenched and heat-treated glasses. The results of this study will be used to develop glass formulations for specific DOE waste streams to maintain or meet waste loading and/or waste throughput expectations while satisfying critical process and product performance related constraints.
Electrochemical Description of Redox Equilibria in Nuclear Glass
Authors: Pinet O. , Mure S. 
Published: April 2011

AbstractGlass matrices were selected to immobilize fission products because glass is capable of chemically incorporating a wide spectrum of elements within a single matrix. Some of these elements can be found at different oxidation states. The redox equilibrium constants of multivalent species can be used to develop thermodynamic models for a better description of nuclear glasses. Some of the multivalent species loaded in nuclear glass, such as iron and sulfur, have already been a subject of investigation by conventional glassmakers or geochemists in the earth sciences. Other redox species more specifically related to nuclear glass, including cerium and ruthenium, have also been investigated. These studies have demonstrated the advantages of using electrochemical techniques, voltammetry and potentiometry, to determine the equilibrium constants. Oxygen potential measurements are also particularly suitable for characterizing the redox state of the multivalent dissolved species in molten glass.
Inorganic matrices for immobilization of Tc-99
Authors: Korneyko Yulia I. , Garbuzov Vladimir M. , Schmidta Olga V. , Burakov Boris E. 
Published: April 2011

AbstractImmobilization of long-lived 99Tc requires development of chemically resistant inorganic matrices. Samples of ceramics based on crystalline Fe-Mn- and Zr-Mn-oxide compounds were synthesized at 1150°C in air, reducing or inert atmosphere from precursors doped with 5-12 wt.% Tc. All the samples obtained were studied using optical and scanning electron microscopy (SEM); powder X-ray diffraction (XRD) and microprobe analysis (EMPA). Content of Tc varied from 0.5-0.8 to 3-6 wt.% in oxide host phases and from 54 to 93 wt.% in metallic inclusions. It was demonstrated that synthesis of oxide host-phases under oxidizing or reducing conditions was not optimal due to partial Tc volatilization or metallic phase formation, respectively. The use of inert atmosphere for ceramic synthesis supports Tc incorporation into crystalline structure of stable host-phases. Development of optimal methods of precursor preparation and synthesis conditions of Tc-doped ceramic are being discussed.
Modeling of Radioactive Graphite Oxidation in Molten Salts: Computer Experiment
Authors: Barbin Nikolai M. , Terentiev Dmitri I. , Alekseyev Sergei G. , Tuktarov Marat A. , Romenkov A. A. 
Published: April 2011

AbstractGraphite is used as the neutron moderator and reflector in many nuclear reactors. Obsolete graphite nuclear reactors are put out of operation, leading to formation of a large quantity of radioactive graphite waste.It is proposed that irradiated reactor graphite is processed by high-temperature chemical oxidation in salt melts with an oxidant, which is part of the salt melt, leading to formation of exhaust gases: gaseous compounds of carbon and oxygen (CO2 and CO).This study deals with carbon oxidation and physical-chemical transformations of radioactive elements during the interaction between graphite waste of the atomic power industry and salt melts. The method of thermodynamic simulation is used. The carbon melt decreases the transfer of radionuclides to the gaseous phase as compared to incineration of graphite in the atmosphere.
Mineralogical Changes of Cement and Bentonite Accompanied ith Their Interactions
Authors: Hoshino Seiichi , Yamaguchi Tetsuji , Maeda Toshikatsu , Mukai Masayuki , Tanaka Tadao , Nakayama Shinichi 
Published: April 2011

AbstractMineralogical changes of cement and bentonite accompanied with their interaction wereexperimentally studied by mixing granulated hardened cement paste and bentonite, and aging the mixture for91 days at 50° C. Mineralogical changes of cement and bentonite were identified by XRD. Hydratedcalcium-silicate phases (C-S-H), Ca(OH)2, ettringite and monosulfate were identified in the unalteredhardened cement. While Ca(OH)2 and monosulfate decreased with aging and disappeared after 91 days,calcite and katoite (Ca3Al2(SiO4)(OH)8) were formed concurrently. Montmorillonite, quartz (and/orchalcedony), clinoptilolite, plagioclase, calcite, analcime and pyrite were identified in the unaltered bentonite.The XRD pattern showed that diffraction intensities of these minerals decreased with aging. It seems thatthese primary minerals dissolved in the course of the alteration. C-S-H appeared in bentonite during the agingas secondary phases, indicating the participation of silicon dissolved from the bentonite and calcium from thecement formed the C-S-H. The formation of C-S-H that had been predicted by previous modeling studieswas confirmed by the present experiments.In addition, diffusivity of tritiated water in mixed specimen with granulated hardened cement andbentonite was determined by a through-diffusion method. The effective diffusivity of tritiated water decreasedwith aging. The result suggests that the mass diffusivity in the interface of cement-bentonite system willdecrease with their interactions. The results of the diffusion experiments are qualitatively consistent with thediffusivity change in cement-bentonite systems predicted by some computational studies.
Experiments Performed in Substantiation of the Conditioning of BN-350 Spent Cesium Traps Using Lead or Lead-Bismuth Alloy Filling Technology
Authors: Romanenko O.G. , Tazhibaeva I.L. , Yakovlev I.L. , Ivanov A.I. , Wells D. , Herrick A. , Michelbacher J.A. , Shiganakov S.B. 
Published: April 2011

AbstractThe technology of removing cesium radionuclides from sodium coolant at the BN-350 fast reactor was realized in the form of two different types of cesium traps: stationary devices connected to the circuit and in-core devices installed into the core of reactor when it was not under operation. Carbon-graphite materials were used as sorbents in these traps to collect and concentrate radioactive cesium, accumulated in the BN-350 reactor circuits over the decades of their operation. The relatively small volume traps provided effective radiation-safe conditions for personnel working close to the primary circuit coolant and equipment during BN-350 decommissioning. These spent cesium traps represent solid radioactive wastes that must be treated before long term storage. The presence of chemically active sodium, potassium and cesium in the traps results in series of problems related to their long term storage and disposal in the Republic of Kazakhstan. As a consequence, the technology of filling spent cesium traps with lead/lead-bismuth alloy was evaluated. A set of experiments was implemented aimed at verification of calculations performed in substantiation of the proposed technology: filling a full scale cesium trap mock-up with sodium followed by its draining to determine the optimal regimes of draining; filling bench scale cesium trap mock-ups with sodium and cesium followed by sodium draining and filling with lead or lead-bismuth alloy at different temperatures and filling rates to chose the optimal regimes for filling spent cesium traps; implementation of leachability tests to determine the rate of cesium release from the filling materials into water. This paper provides a description of the experimental program carried out and the main results obtained.
The Effect of Cold Work on the Creep Properties of Copper
Authors: Martinsson Asa , Andersson Henrik CM 
Published: April 2011

AbstractSpent nuclear fuel is in Sweden planned to be disposed of by encapsulating in waste packages consisting of a cast iron insert surrounded by a copper canister. The waste package is heavy. Throughout the manufacturing process from the extrusion/pierce-and-draw manufacturing to the final placement in the repository, the copper is subjected to handling which could introduce cold work in the material. It is well known that the creep properties of engineering materials at higher temperatures are affected by cold working.The study includes creep testing of four series of cold worked, oxygen-free, phosphorus doped copper (Cu-OFP) at 75 °C. The results are compared to reference series for as-received material carried out in a recent study. Two series of copper cold worked in tension (12 and 24 %) and two series cold worked in compression (12 % parallel to creep load axis and 15 % perpendicular to creep load axis) were tested.The results show that pre-straining in tension of copper leads to prolonged creep life at 75 °C. The creep rate and ductility are reduced. The influence on the creep properties increases with the amount of cold work. Cold work in compression applied along the creep load axis has no effect on the creep life or the creep rate. Nonetheless the ductility is still impaired. However, cold work in compression applied perpendicular to the creep load direction has a positive effect on the creep life. Cold work in both tension and compression results in a pronounced reduction of the initial strain on loading. Yet the high value of the area reduction, 90 %, is unaffected by the degree of cold work.
Quantification of Exchangeable Cations in Interlayer of Tsukinuno Sodium-Montmorillonite
Authors: Sato Haruo 
Published: April 2011

AbstractIn this study, the cation exchange capacity (CEC) and leached exchangeable cations (LC) of montmorillonite purified from bentonite produced in the Tsukinuno bentonite mine, Yamagata, Japan, were measured, and the exchangeable cations in the interlayer of the montmorillonite were discussed. A montmorillonite, in which the soluble minerals were completely removed, was prepared. Kunipia-F and Kunipia-P, for which both bentonites originally contain approximately 100 wt.% montmorillonite, were used as the initial material. All of the measurements were carried out in a N2 atmosphere-controlled glove-box.The CEC values of montmorillonites for both bentonites (100-110 meq/100g) were similar to data conventionally reported, and the sum of LC was also approximately in good agreement with the CEC values. The share of Na+ in the interlayer of montmorillonite calculated from the LC was about 3/4 of the sum of the LC (≍ CEC), and Mg2+ and Ca2+ occupied about 7 and 19 %, respectively. Although montmorillonite in bentonite produced in the Tsukinuno bentonite mine is known as a Na type, the sum of Ca2+ and Mg2+ occupied about 26 % of all exchangeable cations in the interlayer. Based on these data, the ion exchange reaction constant between Na+ and H+ in the interlayer of montmorillonite was calculated to be -0.07. This is nearly 2 orders of magnitude lower than data that are usually adopted.
Overview of the French research on the evolution of spent fuel rod after discharge from the reactor
Authors: Ferry C. , Cappelaere C. , Jegou C. , Piron J.P. , Firon M. , Ambard A. 
Published: April 2011

AbstractSince 2006, French research on spent fuel has focused on the main issues related to transport and extended in-pool storage of spent fuel assembly. Studies on creep behaviour of irradiated cladding have resulted in a new creep model which is valid over a wide domain of temperature, internal pressure and time. Under nominal conditions, no evolution of the spent fuel rod is expected during in-pool storage. In case of defective fuel rods in the storage pool, the consequences of fuel alteration on the initial defect of the cladding depend on the matrix alteration rate and nature of the secondary phases formed. Considering the optional scenario of direct disposal, the long-term behaviour of the spent fuel is investigated focusing on helium consequences before water contact on the one hand and on the influence of repository conditions on matrix alteration on the other hand. The aim of the on-going studies is to improve the safety margins initially introduced in the radionuclide source term models.
Solid-Phase Transformation of Cs+- and Sr2+-Bearing Zeolite Sorbents Derived From Cenospheres to Mineral-Like Forms
Authors: Vereshchagin Sergei N. , Vereshchagina Tatiana A. , Solovyov Leonid A. , Shishkina Nina N. , Vasilieva Nataly G. , Anshits Alexander G. 
Published: April 2011

AbstractThe paper describes the studies of the transformation of Cs+- and Sr2+-containing zeolite sorbents synthesized from fly ash cenospheres to crystalline mineral composition, suitable for the long-term disposal. Series of Cs+- and Sr2+-exchanged NaP1-containing sorbents were subjected to the thermochemical transformation in the temperature range 40-1100°C at atmospheric pressure in air and the progress of reaction was monitored by DSC and XRD analysis. It was shown that initial sodium zeolite undergoes two-step transformation at 736-785°C and 892-982°C forming nepheline as the principle product, with the conversion temperatures being dependant on the heating rate.The thermal treatment of Cs+-bearing zeolite sorbent led to formation of a complex multiphase system, the principal components of which were nepheline and pollucite. Increasing cesium content in the samples led to a monotonous shift of crystallization peak to the higher temperature range (1005-1006°C). A more complicated behavior was observed for Sr2+-containing samples, for which the crystallization temperature tends to increase (compared with NaP1) at lower Sr contents, but it starts decreasing parallel to the Sr2+ content at Sr2+ loadings >10 mg/g. The principal crystalline phases in Sr-NaP1 sample conversion were nepheline and Sr2+-containing feldspar, the quantity of which increased parallel to the increase of strontium content in zeolite.Apparent activation energies of thermochemical transformations were calculated and possible approaches to reduce transformation temperature are discussed and experimentally illustrated.
Immobilization of radioactive isotopes in fluorapatite matrices
Authors: Macerata Elena , Innocente Piergiuseppe , Mariani Mario , Galletta Michele 
Published: April 2011

AbstractFluorapatite (Fap), Ca10(PO4)6F2, could be a potential candidate for the immobilization of radioactive isotopes thanks both to its physical-chemical properties and to the capability to incorporate mono-, di- and trivalent cations in its crystal structure. In this research work different preparation procedures able to incorporate in a fluorapatite matrix strontium in the form of nitrate or fluoride salt were considered. The experiments were performed by means of stable isotopes, as representative of the radionuclides contained in the nuclear fuel. Subsequently, the Sr-substituted fluorapatites (SrxFap), Ca10-xSrx(PO4)6F2, were characterized by X-ray diffraction, Raman spectroscopy, Scanning Electron Microscopy (SEM) and Energy Dispersive Spectroscopy (EDS). The results suggest that the considered synthetic procedures could be reasonably applied to the confinement of the radionuclides contained in some types of nuclear waste. In order to avoid the presence of unreacted reagents in the products or the formation of undesired phases, an optimization of the synthetic procedures may be considered.
Formation of Ultrafine Microstructure during Uniaxial Warm-Compressive Deformation in an Fe-0.67%C Steel for Railway Wheels
Authors: Yasumoto Yuji , Handa Kazuyuki , Kimura Yoshisato , Mishima Yoshinao 
Published: April 2011

AbstractThe evolution of microstructure in an Fe-0.67%C steel used for railway wheels has been investigated. To elucidate the mechanism of the ultrafine microstructure which is formed on the railway wheels tread surface, we have experimentally reproduced the same microstructure using uniaxial compressive deformation and subsequent annealing at 873 K. The deformation conditions required for ultrafine microstructure formation are the initial strain rate of 1 (=100) s-1 and total strain of 0.7. The mechanism of microstructural refinement is not the primary recrystallization but continuous recrystallization, i.e., the recovery process associated with the rearrangement of accumulated dislocations during deformation and annealing at temperatures between 773 K and 873 K. The recovery process never occurs at temperatures lower than 773 K. Primary recrystallization which involves the nucleation and growth of new ferrite grains takes place at temperatures higher than 873 K.
Influence of Analysis Parameters on the Microstructural Characterization of Nanoscale Precipitates
Authors: Serizawa Ai , Miller Michael K. 
Published: April 2011

AbstractA series of simulated microstructures containing nanometer-scale precipitates was created with an atom probe simulator. These data were then analyzed with the proximity histogram and the maximum separation method to determine the influence of the particular analysis method. For simulated 2-nm-radius spherical precipitates, the optimized voxel size and delocalization were found to be 0.5-0.6 nm and 1.0-1.5 nm, respectively. Under optimum analysis parameters, the voxelization/delocalization process only slightly degrades the interface width determined from the proximity histogram to ˜0.15±0.04 nm.