bluuebunny/crossref_metadata_2025_split huggingface dataset:
Analysis of InP Passivated with Thiourea/Ammonia Solutions and Thin CdS Films
Authors: Dauplaise H. M. , Davis A. , Vaccaro K. , Waters W. D. , Spaziani S. M. , Martin E. A. , Lorenzo J. P. 
Published: September 2012

AbstractThe use of thiourea/ammonia pre-treatments on (100) InP, followed by chemical bath deposition (CBD) of CdS thin films (∼ 30 Å), with low-temperature, low-pressure chemical vapor deposited SiO2 has been shown to produce metal-insulator-semiconductor (MIS) samples with near-ideal capacitance-voltage (C-V) response. Here, we report on x-ray photoelectron spectroscopy (XPS) analysis of the near-surface of InP following pre-treatment and CdS deposition. The pre-treatment was shown by XPS to form an indium sulfide layer and effectively remove native oxides from the InP surface. The subsequent deposition of CdS on a sulfur-passivated surface forms a stable layer which protects the substrate from oxidation during SiO2 chemical vapor deposition. MIS samples prepared using the pre-treatment without CdS deposition showed improved C- V response, while samples prepared with both the pre-treatment and CdS deposition showed a dramatic reduction in the density of interface states.
Surface Oxidation Study of Silicon-Doped GaAs Wafers by Ftir Spectroscopy
Authors: Chang R.-H. , Al-Sheikhly M. , Christou A. , Varmazis C. 
Published: September 2012

AbstractA surface study of Si-doped GaAs (100) oriented wafers treated with NH4OH and HC1 following exposure to fluorine containing plasma was conducted using fourier transform infrared spectroscopy (FTIR). These treatments were observed to produce various oxidation products, such as AS2O5 and GaO. Though inorganic salts, such as (NH4)3GaF6, can be formed on the Si-doped GaAs wafers during cleaning with hydrofluoric acid buffered with ammonium fluoride, the applied cleaning method which consisted of NH4OH and HCl treatments subsequent to exposure to fluorine containing plasma did not induce formation of any inorganic salts. A small amount of hydroxide group was also presented in the samples. Water molecules and ammonium hydroxide can be sources of OH which can then be incorporated interstitially into the wafer surfaces.
Ultra High Vacuum Scanning Tunneling Microscopy Observation of Multilayer Step Structure on GaAs and AlAs Vicinal Surface Grown by Metalorganic Vapor Phase Epitaxy
Authors: Ishizaki Jun-Ya , Ishizaki Yasuhiko , Fukui Takashi 
Published: September 2012

AbstractWe observe the atomic structures at the multilayer step region on MOVPE-grown GaAs (001) vicinal surface using ultra high vacuum scanning tunneling microscopy (UHV-STM), and clarify that (4×2) or (4×3) like reconstruction units are dominant. Oxide free AlAs surfaces grown on GaAs vicinal surface are also successfully observed by UHV-STM. The reconstruction units at the multilayer step region on AlAs surface have the same units on GaAs vicinal surface. GaAs surface has the lack of dimmer rows on the terrace region just below the multilayer step region, while AlAs surface has dimmer rows even on the terrace just below the multilayer step region. GaAs layer growth leads tothe step bunching phenomenon and AlAs surface leads to the step debunching phenomenon.
Transport from the Canister to the Biosphere: Using an Integrated Near-and Far-Field Model
Authors: Gylling B. , Romero L. , Moreno L. , Neretnieks I. 
Published: September 2012

ABSTRACTA coupled model concept which may be used for performance assessment of a nuclear repository is presented. The tool is developed by integration of two models, one near field and one far field model. A compartment model, NUCTRAN, is used to calculate the near field release from a damaged canister. The far field transport through fractured rock is simulated by using CHAN3D, based on a three-dimensional stochastic channel network concept. The near field release depends on the local hydraulic properties of the far field. The transport in the far field in turn depends on where the damaged canister(s) is located. The very large heterogeneities in the rock mass makes it necessary to study both the near field release properties and the location of release at the same time. In order to demonstrate the capabilities of the coupled model concept it is applied on a hypothetical repository located at the Hard Rock Laboratory in Äspö, Sweden. Two main items were studied; the location of a damaged canister in relation to fracture zones and the barrier function of the host rock. In the study of the near field rock as a transport barrier the effect of different tunnel excavation methods which may influence the damage level of the rock around the tunnel was addressed.
Development of a New Mechanistic Low-Level Waste Source Term Model
Authors: Yim Man-Sung 
Published: September 2012

ABSTRACTA new mechanistic low-level waste source term model was developed. Key features of this effort are: use of cumulative probability functions to describe the failures of waste containers; capability to describe diffusion controlled release which is dependent on the conditions of the waste package surroundings; consideration of the effects of gas generation on the source term, and; use of a source inventory characterization routine to provide a built-in capability for defining the distributions of radionuclides in various waste forms and streams. The model is capable of describing the diffusion of radionuclides in waste forms with the use of concentration and flux continuity boundary conditions. Release of radionuclides from various waste streams is modeled by the combination of diffusion, dissolution, and surface release. Failures of waste containers are portrayed by the use of probabilistic failure functions based on Weibull and lognormal distributions. The model for radionuclide release from waste packages is coupled with the near-field transport model to describe the effects of migration and dispersion within the disposal unit. Characteristics of the new model were evaluated through sensitivity analysis and compared with existing source term codes.
Incorporation of “Corrosion-Time” and Effects of
Corrosion-Product Spalling in Waste Package Degradation Simulation in the
Potential Repository at Yucca Mountain
Authors: Lee Joon H. , Atkins Joel E. , Dunlap Bryan 
Published: September 2012

ABSTRACTA two-layer waste package (carbon steel outer barrier and Alloy 825 inner
barrier) is specified to dispose of high-level nuclear waste at the
potential repository at Yucca Mountain. A set of improvements and more
realism have been added to a stochastic waste-package degradation model
which was developed for a recent total system performance assessment of the
potential repository [1]. The waste-package surface is divided into
“patches” to better represent the general corrosion of the carbon-steel
outer barrier. The “corrosion-time” concept is developed to represent the
corrosion of the carbon-steel outer barrier in changing exposure conditions
with time such as those expected in the potential repository. With the
patches approach and the corrosion-time concept implemented into the
waste-package degradation model, sensitivity of the waste package
degradation (failure and pitting degradation) to different threshold
spalling thicknesses of the corrosion products from the carbon-steel outer
barrier is analyzed. The results show that the waste-package pitting
degradation is sensitive to the corrosion-products spalling thickness of the
carbon-steel outer barrier. A greater pitting degradation of the waste
packages is predicted with a smaller spalling thickness. Further
understanding of the corrosion-products spalling in different exposure
conditions (i.e., water chemistry, water contact mode, etc.) and its effects
on carbon steel corrosion is needed to enhance the confidence in the
waste-package performance modeling in the potential repository.
Nuclear Repository Performance Assessment: Insights into
Critical Models and Parameters Affecting Projected Future
Doses
Authors: Mcguire R. K. , Kessler J. H. , Vlasity J. A. 
Published: September 2012

ABSTRACTThe Phase 3 Total System Performance Assessment (TSPA) sponsored by the
Electric Power Research Institute (EPRI) has led to new insights into
critical models and parameters affecting estimated doses to humans from a
potential repository of high-level radioactive wastes at Yucca Mountain,
Nevada. The Phase 3 model has been extended to encompass time-varying
climate and infiltration, detailed modeling of the source term and
hydrology, and detailed specification of possible interaction between
percolating ground water and waste containers. The model estimates doses to
a time of one million years.The three key radionuclides contributing to estimated total doses are
Tc-99,1–129, and Np-237. Five other nuclides contributing to dose in lesser
(but significant) amounts are U-233, Th-229, Pa-231, Ac-227, and Se-79.
These results are consistent with other TSPAs.From sensitivity studies, the most critical models and parameters are as
follows. Infiltration and percolation assumptions, including the amount of
lateral diversion of infiltration water, are important and need verification
with site data and/or more detailed modeling. Parameters of the unsaturated
zone (UZ) and saturated zone (SZ) determine dilution and delay of
concentrations and peak doses downstream. The fraction of containers that
become wet are not critical in our model, but this lack of sensitivity
reflects our coupling of the fraction with a model of focused flow past the
containers; an different model might indicate higher sensitivity. Also, the
degree of coupling between fracture and matrix flow is important in
affecting the times of peak doses but not their magnitudes.Other critical design assumptions that could lead to reduced and/or delayed
doses are a more robust container design, a capillary barrier around each
container, the dilution during hydrologie transport from the repository to a
potential agricultural community downstream, and the characteristics of an
“average” individual in that community who might receive a dose.
Maximum Individual Dose and Vicinity-Average Dose for a Geologic
Repository
Authors: Pigford T. H. , Zwahlen E. D. 
Published: September 2012

ABSTRACTRecent proposals for a new U.S. standard for high-level waste disposal would
limit the average dose to individuals in the vicinity surrounding a geologic
repository. This would be a new approach to protecting the public from
environmental releases of radioactivity. Heretofore, criteria adopted for
geologic disposal have limited the reasonable maximum exposure to a future
hypothetical individual. Here we present quantitative analyses of the
relation between maximum exposure and vicinity-average exposure, resulting
from future human use of ground water contaminated by radioactive releases
from a repository.Estimating the vicinity-average exposure would require postulates and
guesses of location and habits of future people. Exposure probabilities
postulated by others show that proposed dose limit to the vicinity-average
individual would be a far more lenient standard than the traditional dose
limit to reasonably maximally exposed individuals. The proposed
vicinity-average dose limit would allow far greater concentrations of
contaminants in ground water than would be allowed by normal standards of
ground water protection. A safety standard that limits vicinity-average
exposure should also include limits on maximum exposure.
Thermal/Hydrological Modeling of the Radioactive Scrap and Waste
Facility (RSWF) with the Tough2 (Transport of Unsaturated Groundwater and
Heat) Code
Authors: Esh D. W. , Benedict R. W. 
Published: September 2012

ABSTRACTThermal/hydrological modeling of the Radioactive Scrap and Waste Facility
(RSWF) has been completed with the TOUGH2 (Transport of Unsaturated
Groundwater and Heat) Code.[1] The RSWF will be utilized as an interim
storage facility for ceramic and metallic waste forms developed from the
electrometallurgical treatment of spent nuclear fuel. The RSWF is an array
of 1,350 carbon steel liners located at grade near Argonne National
Laboratory-West on the Idaho National Engineering Laboratory (INEL).The primary driving force for this modeling research was to assess thermal
capacity limits for RSWF liners so that heat generating materials can be
safely stored. Maximum wasteform temperatures will be governed by both the
amount of heat the system can dissipate and the orientation and
characteristics of the wasteform. The focus of this report is on the amount
of heat the interim storage system can safely dissipate. The effect of the
temporal variation of soil moisture on the performance of the RSWF is
assessed. The facility was analyzed to determine the maximum allowable
thermal loading of the RSWF liners.
An Analysis of Bias in Groundwater Modelling Due to the
Interpretation of Site Characterization Data
Authors: Clark K. J. , Ikeda T. , Impey M. D. , McEwen T. , White M. 
Published: September 2012

ABSTRACTBias is a difference between model and reality. Bias can be introduced at
any stage of the modelling process during a site characterisation or
performance assessment programme. It is desirable to understand such bias so
as to be able to optimally design and interpret a site characterisation
programme. The objective of this study was to examine the source and effect
of bias due to the assumptions modellers have to make because reality cannot
be fully characterised in the prediction of ground-water fluxes. A
well-defined synthetic “reality” was therefore constructed for this study. A
limited subset of these data were independently interpreted and used to
compute groundwater fluxes across specified boundaries in a cross section.
The modelling results were compared to the “true” solutions derived using
the full dataset. This study clarified and identified the large number of
assumptions and judgements which have to be made when modelling a limited
site characterisation dataset. It is concluded that bias is introduced at
each modelling stage, and that it is not necessarily detectable by the
modellers even if multiple runs with varied parameter values are
undertaken.
Thermal Analysis of Options for Spent Fuel Disposal in
Switzerland
Authors: Sasaki Tetsuo , Ando Kenichi , Amura Hideki Kaw , Schneider Jürg W. , McKinley Ian G. 
Published: September 2012

ABSTRACTIn parallel to studies of disposal of vitrified high-level waste from
reprocessing, projects have been initiated to examine options for direct
disposal of spent fuel in Switzerland. The basic concept involves in-tunnel
emplacement of encapsulated spent fuel in a deep repository which is
backfilled with compacted bentonite. Two possible host rocks are considered
- crystalline basement and Opalinus Clay. This paper reports the results of
a thermal analysis which was carried out to evaluate constraints on
repository layout set by the desire to limit temperatures experienced by the
bentonite backfill.
Reconciliation of Experimental and Modelling Concepts in a
Natural Analogue of Radionuclide Migration
Authors: Suksi J. , Rasilainen K. 
Published: September 2012

ABSTRACTOne of the major problems in testing radionuclide transport models with a
natural analogue is the potential discrepancy between experimental and
modelling concepts. In the matrix diffusion studies at Palmottu the measured
U-series concentration profiles and mathematical simulations showed
disagreement, suggesting discrepancy in respective concepts of radionuclide
attachment on the rock pores. Here, the discrepancy was approached by
studying uranium fixation in more detail within the most loosely-bound
fraction.
On the Accuracy of Uranium-Series Dating: a Comparison with a
Known Matrix Diffusion Case
Authors: Rasilainen K. , Suksi J. 
Published: September 2012

ABSTRACTThe accuracy of uranium-series dating was studied quantitatively using the
natural long-lived decay chains 4n+2 and 4n+3. An unusually well-bounded
matrix diffusion case was used as the reference for USD simulations. The
simulations applied the traditional closed and open system models, as well
as detailed multistage models that incorporated the known accumulation
history of the decay chain members. The results show clearly the improved
accuracy and consistency in dating with detailed mass flow information.
The Crystal Structure of Ianthinite, a Mixed-Valence Uranium Oxide Hydrate
Authors: Burns Peter C. , Finck Robert J. , Hawthorne Frank C. , Miller Mark L. , Ewing Rodney C. 
Published: September 2012

ABSTRACTIanthinite, [U4+2(UO2)4O6(OH)4(H2O)4](H2O)5, is the only known uranyl oxide hydrate mineral that contains U4+, and it has been proposed that ianthinite may be an important Pu4+-bearing phase during the oxidative dissolution of spent nuclear fuel. The crystal structure of ianthinite, orthorhombic,a7.178(2),b11.473(2),c.30.39(1) Å,V2502.7 Å3,Z= 4, space groupP21cn, has been solved by direct methods and refined by least-squares methods to anRindex of 9.7 % and awRindex of 12.6 % using 888 unique observed [ | F | ≥ 5σ | F | ] reflections. The structure contains both U6+and U4+. The U6+cations are present as roughly linear (U6+O2)2+uranyl ions (Ur) that are in turn coordinated by five O2-and OH located at the equatorial positions of pentagonal bipyramids. The U4+cations are coordinated by O2-, OH and H2O in a distorted octahedral arrangement. TheUrφ5and U4+φ6(φ: O2-, OH, H2O) polyhedra link by sharing edges to form two symmetrically distinct sheets atzz ≈ 0.0 and z ≈ 0.25 that are parallel to (001). The sheets have the β-U3O8sheet anion-topology. There are five symmetrically distinct H2O groups located at z ≈ 0.125 between the sheets of Uφnpolyhedra, and the sheets of Uφnpolyhedra are linked together only by hydrogen bonding to the intersheet H2O groups. The crystal-chemical requirements of U4+and Pu4+are very similar, indicating that extensive Pu4+↔ U4+substitution can occur within the sheets of Uφnpolyhedra in the structure of ianthinite.
Proliferation Resistance of Borosilicate Glass as a Host Form
for Weapons-Grade Plutonium
Authors: Cerefice G. S. , Wenzel K. W. 
Published: September 2012

ABSTRACTTo examine the proliferation resistance of borosilicate glass, a process to
extract and recover a plutonium analog (thorium) from borosilicate glass was
developed and examined. The glass matrix examined was a modified standard
frit consisting of the ARM-1 frit (with simulated fission products) loaded
with 2 wt. % thorium (as an analog for plutonium) and 2 wt. % each of three
rare earth elements (Gd, Sm, Eu), which were added for criticality control
and to possibly increase the proliferation resistance of the glass matrix.
The plutonium analog was extracted from the crushed glass with a nitric acid
dissolution process, and subsequently decontaminated using a solvent
extraction process. The acid dissolution process was able to extract 88.4 ±
6.8 % of the plutonium surrogate from the glass host form. The bench top
solvent extraction process was 30.2 ± 10.9 % efficient in recovering the
plutonium analog as a purified product. Overall, this process was able to
extract 26.7 ± 9.9 % of the plutonium analog from the glass as a purified
product. To quantify the proliferation resistance of borosilicate glass as a
host form for weapons-grade plutonium, MCNP was used to determine the
compressed critical mass of a plutonium alloy with the same composition as
the product of the extraction process. For the average product composition,
the compressed critical mass was 4.7 kg of material. On average, one
compressed critical mass could be recovered from 613 kg of borosilicate
glass (2 wt. % Pu loading).
Characterization of a Plutonium-Bearing Zirconolite-Rich Synroc
Authors: Buck E. C. , Ebbinghaus B. , Bakel A. J. , Bates J. K. 
Published: September 2012

ABSTRACTA titanate-based ceramic waste form, rich in phases structurally related to zirconolite (CaZrTi2O7), is being developed as a possible method for immobilizing excess plutonium from dismantled nuclear weapons. As part of this program, Lawrence Livermore National Laboratory (LLNL) produced several ceramics that were then characterized at Argonne National Laboratory (ANL). The plutonium-loaded ceramic was found to contain a Pu-Gd zirconolite phase but also contained plutonium titanates, Gd-polymignyte, and a series of other phases. In addition, much of the Pu was remained as PuO2-x. The Pu oxidation state in the zirconolite was determined to be mainly Pu4+, although some Pu3+was believed to be present.
Characterization and Leaching Behavior of Plutonium-Bearing
Synroc-C
Authors: Smith Katherine L. , Lumpkin Gregory R. , Blackford Mark G. , Hambley Michelle , Arthur Day R. , Hart Kaye P. , Jostsons Adam 
Published: September 2012

ABSTRACTSynroc-C containing 10wt% simulated PW-4b-D HLW including 0.62 wt% 239Pu was
subjected to MCC-1 type leach tests at 70°C in deionised water, silicate and
carbonate leachates for 53 d and deionised water for 2472 d. The normalised
total (i.e. unfiltered leachate + vessel wall) Pu leach rates in deionised
water, silicate and carbonate leachates for periods up to 53 d were found to
be of the order of 10-5, 10-4 and 10-4 g
m-2 d-1 respectively. After 2472 d, the
differential, normalised, Pu leach rate in deionised water dropped to ∼5 ×
10-6 (total) and ∼5 × 10-8 (solution - after
filtration through a 1000NMW filter) g m-2 d-1. SEM
and AEM were used to characterise our starting material and investigate the
secondary phases on the surfaces of leached Synroc-C discs. Calculated and
measured normalised Pu leach rates are compared and the partitioning of Pu
between zirconolite and perovskite is discussed.
Underground Criticality in Geologic Disposal
Authors: Choi J-S , Pigford T. H. 
Published: September 2012

ABSTRACTDisposition of weapons fissile material as once-through MOx fuel in reactors
or immobilization in waste glass would result in end products requiring
geologic disposal. Similar considerations apply to some spent fuels in a
geologic repository. The criticality potential of these end products in a
geologic setting is an important consideration for the final disposal of
these materials. In this document we present calculations that help define
parameters and configurations for criticality, including those that, if the
criticality is achieved, could lead to autocatalytic power transients. It is
not the purpose of this paper to define scenarios that could reasonably lead
to criticality or to establish the probabilities that critical
configurations could occur. We do suggest some parameters that should be
considered by waste package designers and repository analysts to ensure that
criticality cannot occur.
Interaction of UO2 and Zircaloy During the Chernobyl
Accident
Authors: Ushakov S. V. , Burakov B. E. , Shabalev S. I. , Anderson E. B. 
Published: September 2012

ABSTRACTA summary of the results collected during the studies of the products of a
chemical interaction between uranium oxide fuel and Zircaloy cladding in the
Chernobyl accident is presented in this paper. The reaction products are
mainly Zr-U-containing phases with different U/Zr ratio and are described on
the basis of electron microprobe and X-ray diffraction (XRD) analyses. The
Zr-U-bearing phases were discovered among the inclusions in different types
of Chernobyl fuel-containing masses (”lava”) inside the destroyed 4th Unit
and in hot particles collected up to 12 km from the 4th Unit along the West
Plume. A correlation of data on the chemical composition and phase
interrelations obtained in investigated samples with a phase diagram of
Zr(O) - UO2 shows, that a temperature >1900 °C was reached in
a part of the core before the explosion. The detection of hot particle with
segregated morphology points out that liquid immiscibility existed between
U-rich and Zr-rich melts. This and other observations indicate that the core
temperature locally was above 2400–2600°C.