bluuebunny/crossref_metadata_2025 huggingface dataset:
A Three-Dimensional Methodology for the Assessment of Neutron Damage and Nuclear Energy Deposition in Graphite Components of Advanced Gas-Cooled Reactors
Authors: Morgan D. O. , Robinson A. T. , Allen D. A. , Picton D. J. , Thornton D. A , Shaw S. E.
Published: May 2012

This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK’s Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition.
The Current Status of the Shielding Integral Benchmark Archive and Database (SINBAD)
Authors: Kirk B. L. , Grove R. E. , Kodeli I. , Gulliford J. , Sartori E.
Published: May 2012

The shielding integral benchmark archive and database (SINBAD) collection of experiments descriptions was initiated in the early 1990s. SINBAD is an international collaboration between the Organization for Economic Cooperation and Development’s Nuclear Energy Agency Data Bank (OECD/NEADB) and the Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory (ORNL). SINBAD was designed to compile experiments and corresponding computational models with the goal of preserving institutional knowledge and expertise that need to be handed down to future scientists. SINBAD can serve as a learning tool for university students and scientists who need to design experiments or gain expertise in modeling and simulation. The SINBAD database is currently divided into three categories—fission, fusion, and accelerator experiments. Many experiments are described and analyzed using deterministic or stochastic (Monte Carlo) radiation transport software. The nuclear cross sections also play an important role as they are necessary in performing computational analysis.
Development of Neutron Measurement in Intense Gamma Field Using New Type of Nuclear Emulsion
Authors: Kawarabayashi J. , Ishihara K. , Takagi K. , Tomita H. , Iguchi T. , Naka T. , Morishima K. , Maeda S.
Published: May 2012

To precisely measure the neutron emissions from a spent fuel assembly of a fast breeder reactor, we formed nuclear emulsions based on a non-sensitized Oscillation Project with Emulsion tRacking Apparatus (OPERA) film with AgBr grain sizes of 60, 90, and 160 nm. The efficiency for 252Cf neutron detection of the new emulsion was calculated to be 0.7 × 10−4, which corresponded to an energy range from 0.3 to 2 MeV and was consistent with a preliminary estimate based on experimental results. The sensitivity of the new emulsion was also experimentally estimated by irradiating with 565 keV and 14 MeV neutrons. The emulsion with an AgBr grain size of 60 nm had the lowest sensitivity among the above three emulsions but was still sensitive enough to detect protons. Furthermore, the experimental data suggested that there was a threshold linear energy transfer of 15 keV/μm for the new emulsion, below which no silver clusters developed. Further development of nuclear emulsion with an AgBr grain size of a few tens of nanometers will be the next stage of the present study.
Sensitivity studies associated with dosimetry experiment interpretation
Authors: Bourganel S. , Soldevila M. , Ferrer A. , Grégoire G. , Destouches C. , Beretz D.
Published: May 2012

Interpretation of reactor dosimetry experiments with C/E comparison requires precise knowledge of parameters involved in modeling. Some parameters have more weight than others on the calculated values. So, sensitivity studies should be conducted to verify the importance of these parameters. The conclusions of these studies are used to refine the experiment modeling, or to correct uncertainty calculations. The results of these sensitivity studies allow a post-irradiation analysis, which can justify the discarding of some atypical C/M values. Derived uncertainties may be improved by the sensitivity analyses. Beyond classical parameters as geometry or composition, this paper describes some specific sensitivity studies conducted for dosimetry irradiation in reactor, and presents conclusions. These studies are based on dosimeters irradiated in the EOLE reactor facility at Cadarache CEA center. Conclusions drawn from these studies are generic and can be applied to any dosimetry study. Calculations performed for these studies were realized using TRIPOLI-4 Monte Carlo code.
Ringhals Unit 3 and 4—Fluence Determination in a Historic and Future Perspective
Authors: Green Eva-Lena , Roudén Jenny , Efsing Pål
Published: May 2012

The Ringhals site is situated on the Swedish southwest coastline. At the site, there are four operating nuclear power plants. Historically, the Swedish policy has been that the nuclear power plants were to be closed in 2010. The present position is to operate the units until their technical and economic lifetime has run out. The units shall be maintained and invested in to ensure a lifetime of at least 50 years, but the actions taken shall not limit the time to this date. When the initial surveillance capsules were evaluated, it was noted that the material properties of the weld material of unit 3 and 4 showed some deviations from the expected behaviour. Currently there is an extensive project running for re-evaluating the embrittlement situation from a long-term operating perspective. One part of the project is aimed at more accurately determining the fluence levels of the reactor pressure vessels (RPVs). The basis for the early evaluations of the dosimeters in the surveillance capsules and the corresponding fluence evaluation had an operating lifetime of 25 years as a target value. Therefore, the accuracy and refinement of the measurement and calculation were taken to be good enough to suit this life span. Looking back at the results from the dosimetry measurements there are a few discrepancies. Some of the dosimeters were disintegrated and some measurements had comparatively large uncertainties. When starting this project there were some re-evaluations done with the old fluence prediction model. For every new run and refinement there appeared new difficulties, and the decision was to start the evaluation from scratch.

Then there are two questions remaining regarding the fluence: What is the current fluence level? What will the resulting fluence be after 60 years of operation, when we have up-rated output power of both reactors? This paper aims to describe the view of the fluence evaluation in the surveillance program of the RPV, both in a historical and prospective view.
Measurements of Actinide-Fission Product Yields in Caliban and Prospero Metallic Core Reactor Fission-Neutron Fields
Authors: Casoli P. , Authier N. , Laurec J. , Bauge E. , Granier T.
Published: May 2012

In the 1970s and early 1980s, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on 233U, 235U, and 239Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments.
Application of Ex-vessel Neutron Dosimetry Combined with In-core Measurements for Correction of Neutron Source Used for Reactor Pressure Vessel Fluence Calculations
Authors: Borodkin P. G. , Borodkin G. I. , Khrennikov N. N. , Konheiser J.
Published: May 2012

This paper deals with calculated and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Neutron activation measurements analyzed in the paper were carried out in an ex-vessel air cavity at different nuclear power plant units with VVER-1000 during different fuel cycles. The time-integrated neutron source distributions used for DORT calculations were prepared via two different approaches based on (a) calculated fuel burnup (standard routine procedure) and (b) in-core measurements by means of self-powered detectors (SPDs) and thermocouples (TCs) (new approach). Considering that fuel burnup distributions in operating VVER may be evaluated now by the use of analytical methods (calculations) only, it is necessary to develop new approaches for the testing and correction of calculated evaluations of a neutron source. The results presented in this paper allow one to consider the reverse task of the alternative estimation of fuel burnup distributions. The proposed approach is based on the adjustment (fitting) of time-integrated neutron source distributions, and thus fuel burnup patterns, in some part of the reactor core, taking into account neutron leakage measurements, neutron-physical calculations, and in-core SPD and TC measurement data.
The Neutron Standard Fields at the BR1 Reactor at SCK•CEN
Authors: Wagemans J. , Malambu E. , Borms L.
Published: May 2012

The BR1 research reactor at SCK•CEN is characterized by a wide variety of irradiation possibilities, a large reactor core, and strong flexibility in its operation. A full MCNP model of BR1 has been recently developed in order to complement the results that can be obtained from activation dosimetry. After a general presentation of the reactor, this paper pays particular attention to its standard 235U(n,f) fast neutron field MARK III. This irradiation field is a useful tool for integral measurements and for detector calibrations. With the support of MCNP calculations, irradiations in MARK III can be directly referred to the pure 235U(n,f) fast neutron spectrum.
Development and Experimental Validation of a Calculation Scheme for Nuclear Heating Evaluation in the Core of the OSIRIS Material Testing Reactor
Authors: Malouch F.
Published: May 2012

The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron–photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation.
Ex-Vessel Neutron Dosimetry Analysis for Westinghouse 4-Loop XL Pressurized Water Reactor Plant Using 3D Parallel Discrete Ordinates Code RAPTOR-M3G
Authors: Chen J. , Alpan F. A. , Fischer G. A. , Fero A. H.
Published: May 2012

Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the beltline region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locations and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure.
An Alternative Calibration Method for Counting P-32 Reactor Monitors
Authors: Quirk T. J. , Vehar D. W.
Published: May 2012

Radioactivation of sulfur is a common technique used to measure fast neutron fluences in test and research reactors. Elemental sulfur can be pressed into pellets and used as monitors. The 32S(n, p)32P reaction has a practical threshold of about 3 MeV and its cross section and associated uncertainties are well characterized [1]. The product 32P emits a beta particle with a maximum energy of 1710 keV [2]. This energetic beta particle allows pellets to be counted intact. ASTM Standard Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32 (E265) [3] details a method of calibration for counting systems and subsequent analysis of results. This method requires irradiation of sulfur monitors in a fast-neutron field whose spectrum and intensity are well known. The resultant decay-corrected count rate is then correlated to the known fast neutron fluence. The Radiation Metrology Laboratory (RML) at Sandia has traditionally performed calibration irradiations of sulfur pellets using the 252Cf spontaneous fission neutron source at the National Institute of Standards and Technology (NIST) [4] as a transfer standard. However, decay has reduced the intensity of NIST’s source; thus lowering the practical upper limits of available fluence. As of May 2010, neutron emission rates have decayed to approximately 3e8 n/s. In practice, this degradation of capabilities precludes calibrations at the highest fluence levels produced at test reactors and limits the useful range of count rates that can be measured. Furthermore, the reduced availability of replacement 252Cf threatens the long-term viability of the NIST 252Cf facility for sulfur pellet calibrations. In lieu of correlating count rate to neutron fluence in a reference field the total quantity of 32P produced in a pellet can be determined by absolute counting methods. This offers an attractive alternative to extended 252Cf exposures because it can be performed regardless of the characterization of the exposure environment. Count rates produced by sulfur pellets are correlated to the measured quantity of separated 32P. A posteriori spectral and cross section determination can be used to correlate the quantity of phosphorus back to a neutron fluence in a reference field. This paper outlines a method for the setup, calibration, and use of the detector systems, 32P sample preparation, and analysis of the beta spectrum. An uncertainty analysis and comparison to ASTM E265 is also included.
Comparison of Attenuation Coefficients for VVER-440 and VVER-1000 Pressure Vessels
Authors: Marek Milan , Rataj Jan , Vandlik Stanislav
Published: May 2012

The paper summarizes the attenuation coefficient of the neutron fluence with E > 0.5 MeV through a reactor pressure vessel for vodo-vodyanoi energetichesky reactor (VVER) reactor types measured and/or calculated for mock-up experiments, as well as for operated nuclear power plant (NPP) units. The attenuation coefficient is possible to evaluate directly only by using the retro-dosimetry, based on a combination of the measured activities from the weld sample and concurrent ex-vessel measurement. The available neutron fluence attenuation coefficients (E > 0.5 MeV), calculated and measured at a mock-up experiment simulating the VVER-440-unit conditions, vary from 3.5 to 6.15. A similar situation is used for the calculations and mock-up experiment measurements for the VVER-1000 RPV, where the attenuation coefficient of the neutron fluence varies from 5.99 to 8.85. Because of the difference in calculations for the real units and the mock-up experiments, the necessity to design and perform calculation benchmarks both for VVER-440 and VVER-1000 would be meaningful if the calculation model is designed adequately to a given unit.
A Database-Informed Approach to New Plant Shielding Design
Authors: Lloyd Timothy M.
Published: May 2012

To facilitate the definition and description of radiation dose rates in the numerous rooms and areas of a new nuclear power plant, a database approach was developed. This approach offers a number of benefits over more manual methods. A key benefit is that the selection of an appropriate shielding method to use in each area of the plant is greatly facilitated by virtue of the team’s improved ability to grasp the significance of each of the individual sources that are candidates for making a significant contribution to the dose rate in each area. By understanding the level of relevant contribution—if any—of each of these candidate sources, an analyst is able to select a method that will define the contribution without becoming enmired in a model representing inappropriately high degrees of accuracy and modeling time. This database method, by allowing for an evolving understanding of dose rates and sources in the neighboring rooms for each portion of the plant, leads to substantial reductions in the effort of characterizing a plant’s radiation environment. As an additional benefit, the database serves as a tool for documenting the shielding calculations themselves, automatically generating formatted sections including drawing and source term references, shielding calculation types, key dimensions, and results; these sections can form the starting point of a full calculation package. The approach offers a final project management benefit: estimating, tracking, and predicting the effort associated with the many calculations involved in such a project are greatly systematized, leading to more reliable manpower estimates.
Outline of Testing and Evaluation Program Used in Selection of Adhesives for Transparent Adhesive Joints in All-Glass Load-Bearing Structures
Authors: Weller Bernhard , Nicklisch Felix , Prautzsch Volker , Vogt Iris
Published: May 2012

Architects are increasingly demanding all-glass load-bearing structures with fully transparent adhesive joints. Usually, such structures are classed as nonregulated forms of construction. The Institute of Building Construction at Dresden’s Technische Universität has now obtained the first individual approvals in Germany for all-glass structures with transparent adhesive joints for two buildings in Dresden and Grimma. In these cases, the loads are carried via load-bearing glued frames that rely on a material bond between the individual parts without any metal fixings. This solution is based on the results of many years of development. Currently, various frame corners covering a wide range of parameters are being studied in a follow-up project. The aim is to optimize the adhesive joints and the bonding technology. The requirements placed on the joint are being identified and corresponding adhesive systems researched. Preliminary studies of numerous material specimens form the starting point for determining the material parameters. Small-scale specimens are being tested under various boundary conditions and aging scenarios to establish the strengths of a number of suitable adhesives. These results will enable prototypes to be designed for numerical simulation. Numerical calculations and experimental investigations are being carried out in parallel to optimize the geometry, load-carrying capacity, long-term reliability, and durability of the glued all-glass frame corners. Specimen components are loaded in a testing machine to study the structural effect of these glued glass frame corners. The findings will be incorporated in the design of the adhesive joint and in the development of a numerical simulation for the glued connection. Data for executing an adhesive joint, specific to each adhesive and crucial for the quality of the transparency, conclude the project.
New Work on Updating and Extending the Nuclear Data Standards
Authors: Carlson A. D. , Pronyaev V. G. , Capote R. , Hambsch F. -J. , Käppeler F. , Lederer C. , Mannhart W. , Mengoni A. , Nelson R. O. , Schillebeeckx P. , Talou P. , Tagesen S. , Vonach H. , Vorobyev A. , Wallner A.
Published: May 2012

An International Atomic Energy Agency (IAEA) Data Development Project was initiated to provide a mechanism for allowing new experimental data and improvements in evaluation procedures to be incorporated into new evaluations of the standards. The work on this project is ongoing. In the past very long periods sometimes occurred between evaluations of the standards. Through the use of this Project, such long periods should not occur. Work on the cross section standards through this Project has included an update of the experimental data to be used in the cross section standards evaluations, a study of the uncertainties obtained in the international standards evaluation, and improvements in the smoothing procedure for capture cross sections. It was decided that this Project should have a broader range of activities than just the cross section standards and thus encompass standards related activities. The following are being investigated: improvements in the gold cross section at energies below where it is considered a standard and work on certain cross sections that are not as well known as the cross section standards but could be very useful as reference cross sections relative to which certain types of cross section measurements can be made. This work includes prompt gamma-ray production in fast neutron-induced reactions, and work on the 252Cf spontaneous fission neutron spectrum and the 235U thermal neutron fission spectrum. Most of the data investigated through this Project are used in dosimetry applications.
Updating and Extending the IRDF-2002 Dosimetry Library
Authors: Capote Roberto , Zolotarev Konstantin I. , Pronyaev Vladimir G. , Trkov Andrej
Published: May 2012

The International Reactor Dosimetry File (IRDF)-2002 released in 2004 by the IAEA (see http://www-nds.iaea.org/irdf2002/) contains cross-section data and corresponding uncertainties for 66 dosimetry reactions. New cross-section evaluations have become available recently that re-define some of these dosimetry reactions including: (1) high-fidelity evaluation work undertaken by one of the authors (KIZ); (2) evaluations from the US ENDF/B-VII.0 and candidate evaluations from the US ENDF/B-VII.1 libraries that cover reactions within the International Evaluation of Neutron Cross-Section Standards; (3) European JEFF3.1 library; and (4) Japanese JENDL-4.0 library. Additional high-threshold reactions not included in IRDF-2002 (e.g.,59Co(n,3n) and 209Bi(n,3n)) have been also evaluated to characterize higher-energy neutron fields. Overall, 37 new evaluations of dosimetry reactions have been assessed and intercomparisons made with integral measurements in reference neutron fields to determine whether they should be adopted to update and improve IRDF-2002. Benchmark calculations performed for newly evaluated reactions using the ENDF/B-VII.0 235U thermal fission and 252Cf spontaneous fission neutron spectra show that calculated integral cross sections exhibit improved agreement with evaluated experimental data when compared with the equivalent data from the IRDF-2002 library. Data inconsistencies or deficiencies of new evaluations have been identified for 63Cu(n,2n), 60Ni(n,p)60m+g Co, 55Mn(n,γ), and 232Th(n,f) reactions. Compared with IRDF-2002, the upper neutron energy boundary was formally increased from the actual maximum energy of typically 20 MeV up to 60 MeV by using the TENDL-2010 cross sections and covariance matrices. This extension would allow the updated IRDF library to be also used in fusion dosimetry applications. Uncertainties in the cross sections for all new evaluations are given in the form of relative covariance matrices. Newly evaluated excitation functions should be considered as suitable candidates in the preparation of an improved version of the IRDF that was planned to be released for testing in December 2011.
Durability of Cold-Bent Insulating-Glass Units
Authors: Besserud Keith , Bergers Matt , Black Andrew J. , Carbary Lawrence Donald , Mazurek Arkadiusz , Misson Donald , Rubis Kenneth
Published: May 2012

Architectural influences upon the commercial building industry have resulted in the design and development of high-performing complex curtain walls that may include bent or warped glass. Glass on modern day curtain-wall projects is mostly insulating glass (IG). When an IG is bent out of plane, a shearing action occurs to both the primary and secondary seals. The testing described in this paper was done to determine the effect on the durability of the IG as measured by argon retention, frost point change, and visual changes after aging in a displaced condition. Full-size IG units, 5 × 10 ft2 × 1 in. (1.52 × 3.05 m2 × 25.4 mm), were fabricated, bent out of plane at 2–12 in. (50–300 mm) in 2-in. (50-mm) increments, and placed under a 100-psf (4788-Pa) wind load. One unit was displaced 15 in. (380 mm) where breakage occurred. An assessment was made based on the data as to what would be a reasonable amount of bending to avoid excessive glass stress and to avoid breakage. This full-size unit was then modeled and measured to predict and validate the stresses and strains on the primary and secondary seals. Upon completion of the bending tests and the review of modeling, small IG units 14 × 20 in.2 (350 × 500 mm2) were tested according to the protocol specified in ASTM E2188-10 and E2190-10 with an equivalent amount of edge-seal displacement in X, Y, and Z directions. Argon-retention and frost-point measurements were taken before and after the durability testing and results reported. During the durability testing, the glass of several of the units broke in thermal cycling because of high local stresses in the glass. However, the units that did survive the aging protocol passed both frost-point and argon-retention tests, while maintaining their appearance. Additionally, the methodology developed in this research provides a strong foundation for future testing in the area of cold-bent glass durability.
Potential of Dynamic-Mechanical Analysis Toward a Complementary Material and System Testing Approach for Structural Glazing
Authors: Recknagel Christoph
Published: May 2012

Dynamic-mechanical material analysis as a basis for a general performance exploration complemented by system testing under superimposed climatic and mechanical loading seems to be a promising interdependent test approach addressing the performance behavior of construction sealants under more realistic conditions. With this contribution an attempt is made to adapt dynamic-mechanical material analysis, which has been already successfully validated for different construction types of expansion joint systems in road and bridge engineering, to the field of construction sealants for building façades. Test results from dynamic-mechanical material analysis characterizing the temperature-dependent, deformation-dependent, and frequency-dependent behavior of structural sealant materials are presented and exemplarily discussed for three different sealant products. An attempt is made to address unknown material characteristics in the multi-dimensional loading matrix representing practical use conditions. Furthermore, the applicability of this test approach and its various complex test modes for the exploration of technological performance and especially estimation of fatigue behavior is verified in several examples. Based on this fundamental material exploration, it is planned to complement the dynamic-mechanical assessment methodology by means of system tests on a section of a structural glazing system subjected to a simplified but superimposed loading function. The technical fundamentals and the procedure proposed to develop an adequate system test mode are introduced. The motivation for these investigations is to identify the actual mechanical system behavior under load combinations and for specimens that both closer resemble reality. The objective is to achieve a consistent and interdependent test program complementary to the existing methodology. Finally, the study is meant to initiate further progress toward a performance-related methodology which considers the design, specification, material, and system selection.
VVER-440 and VVER-1000 Reactor Dosimetry Benchmark —BUGLE-96 Versus ALPAN VII.0
Authors: Duo J. I.
Published: May 2012

Analytical results of the vodo-vodyanoi energetichesky reactor- (VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Institute Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10 %) between BUGLE-96 and ALPAN VII.0 libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15 % with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements.
Retrospective Dosimetry Analyses of Reactor Vessel Cladding Samples
Authors: Greenwood L. R. , Soderquist C. Z. , Fero A. H.
Published: May 2012

Reactor pressure vessel cladding samples for Ringhals Units 3 and 4 in Sweden were analyzed using retrospective reactor dosimetry techniques. The objective was to provide the best estimates of the neutron fluence for comparison with neutron transport calculations. A total of 51 stainless steel samples consisting of chips weighing approximately 100 to 200 mg were removed from selected locations around the pressure vessel and were sent to Pacific Northwest National Laboratory for analysis. The samples were fully characterized and analyzed for radioactive isotopes, with special interest in the presence of Nb-93 m. The RPV cladding retrospective dosimetry results will be combined with a re-evaluation of the surveillance capsule dosimetry and with ex-vessel neutron dosimetry results to form a comprehensive 3D comparison of measurements to calculations performed with 3D deterministic transport code.
Mechanisms of Asphalt Blistering on Concrete Bridges
Authors: Hailesilassie Biruk W. , Partl Manfred N.
Published: May 2012

Blistering is a major problem in asphalt-covered concrete structures, such as multi-storage parking buildings, built-up roofs, tunnels, pedestrian areas, or concrete bridge decks. In this particular research, a linear viscoelastic finite-element model is developed to simulate time-dependent blister growth in an asphalt layer under uniformly applied pressure with and without temperature and pressure fluctuation. Indirect tensile tests on mastic asphalt (MA) are performed at three different temperatures to characterize and determine the material properties for the model. A three-dimensional thick-plate axisymmetric finite-element model is developed using ABAQUS with linear viscoelastic properties and validated with closed-form solution from first-order shear-deformation theory for thick plates. Elastic–viscoelastic analogy is used to find an analytic solution for the time-dependent deflection of the blister. In addition, the blister test is conducted on different samples of MA in the laboratory and digital image correlation measurement technique is used to capture the three-dimensional vertical deflection of the MA with time. Finally, the results from image correlation are compared with the finite-element simulation and thick-plate theory analytic solution. The finite-element model simulation shows that the daily temperature variations may have a significant influence on blister growth in asphalt pavements. It is found that the blister can grow continuously under repeated loading conditions over subsequent days. The study concludes that temperature fluctuation in the blister has more influence on blister growth than fluctuation of the pressure inside the blister.
Modernization of Existing VVER-1000 Surveillance Programs
Authors: Kochkin V. , Erak D. , Makhotin D.
Published: May 2012

According to generally accepted world practice, evaluation of the reactor pressure vessel (RPV) material behavior during operation is carried out using tests of surveillance specimens. The main objective of the surveillance program consists in insurance of safe RPV operation during the design lifetime and lifetime-extension period. At present, the approaches of pressure vessels residual life validation based on the test results of their surveillance specimens have been developed and introduced in Russia and are under consideration in other countries where vodo-vodyanoi energetichesky reactors- (VVER-) 1000 are in operation. In this case, it is necessary to ensure leading irradiation of surveillance specimens (as compared to the pressure vessel wall) and to provide uniformly irradiated specimen groups for mechanical testing. Standard surveillance program of VVER-1000 has several significant shortcomings and does not meet these requirements. Taking into account program of lifetime extension of VVER-1000 operating in Russia, it is necessary to carry out upgrading of the VVER-1000 surveillance program. This paper studies the conditions of a surveillance specimen’s irradiation and upgrading of existing sets to provide monitoring and prognosis of RPV material properties for extension of the reactor’s lifetime up to 60 years or more.
Evaluation of the Structural Sealant for Use in a Four-Sided Structural Sealant Glazing Curtain-Wall System for a Hospital Building
Authors: Memari A. M. , Fisher S. , Krumenacker C. , Broker K. A. , Modrich R. -U.
Published: May 2012

Cathedral Hill Hospital (the California Pacific Medical Center) is a 15-story building designed to be constructed in downtown San Francisco. The curtain-wall system for this building is primarily of unitized design employing a four-sided structural sealant glazing (SSG) system. This paper initially introduces the building and its curtain-wall design. Next, the summary of the results of full-scale racking tests on stick-built mockups of the curtain-wall system developed to evaluate the structural sealant performance are presented. Stick-built mockup, instead of unitized system, was chosen to initially evaluate the behavior of sealants under racking conditions. The mockups were designed to determine the behavior of the glass, framing, connections, and more importantly, the structural silicone under racking displacements following the AAMA 501.6 protocol. The test results (stress–strain relationships) on the structural sealant to determine tensile adhesion properties at selected temperatures and UV exposures are discussed and comparisons to the finite-element analysis results are presented. The wind load and seismic design of the sealants, including the finite-element modeling and analysis of a typical panel to evaluate the stresses in the structural sealants, are presented as well. The allowable stress issue in structural sealants for seismic design of four-sided SSG systems is discussed in light of the new information generated for this project.
Development of a Non-Destructive Evaluation Practice of Installed Weatherproofing Sealant Joints Using a Rolling Device—An Introduction to ASTM C 1736
Authors: Huff Daniel N.
Published: May 2012

This paper provides an overview of the development of standardized methodology to evaluate joint seal continuity, encompassed in the newly published ASTM C 1736-11, “Standard Practice for Non-Destructive Evaluation of Adhesion of Installed Weatherproofing Sealant Joints Using a Rolling Device.” This standard practice was created under the jurisdiction of ASTM Committee C 24 on Building Seals and Sealants and is the direct responsibility of Subcommittee 30 on Adhesion; it was approved July 1, 2011. Fundamental details contributing to the successful usage of this practice are examined by answering the following questions: Where exactly do the stresses produced by this procedure have an effect upon the sealant-to-substrate bond-line? How does joint geometry impact this bond-line stress? What level of stress on a bond-line provides usable information without damaging the seal?
A Numerical Strip-Yield Model for the Creep Crack Incubation in Steels
Authors: Potirniche G. P.
Published: May 2012

A numerical strip-yield model was developed to simulate creep crack incubation in heat-resistant steels. The model is based on a formulation proposed by Newman (Newman, J. C., Jr., “A Crack-Closure Model for Predicting Fatigue Crack Growth under Aircraft Spectrum Loading,” Methods and Models for Predicting Fatigue Crack Growth under Random Loading, ASTM STP 748, J. B. Chang and C. M. Hudson, Eds., ASTM International, West Conshohocken, PA, 1981, pp. 53–84) for fatigue crack growth under variable amplitude loading. The time evolution of the plastic deformation ahead of a crack loaded in tension is modeled using the Norton law for secondary creep stage, and the primary and tertiary creep stages are neglected. The model assumes a pre-existing crack in a specimen and models the behavior of the material prior to the beginning of crack propagation due to creep loading. The evolution with time of the crack-tip plastic zone, crack-tip opening displacement, and yield strength in the plastic zone are computed at constant temperature for center crack panels. Comparison with two previous strip-yield models and experimental data is performed, and good correlation is obtained for several Cr-Mo-V steels. This approach to modeling creep crack incubation has the potential to be applied to other types of cracked specimens under constant or variable amplitude loading.
Testing Damping Performance and Bending Stiffness of Snowboarding Wrist Protectors
Authors: Schmitt Kai-Uwe , Michel Frank I. , Staudigl Florian
Published: May 2012

Although wrist guards are available to prevent the wrist injuries that are frequently sustained in snowboarding, there is no mandated performance standard for these wrist guards as there is for other sports protection equipment. This study investigates whether the standard specified for safety gear in roller sports (EN 14120) can also be applied to snowboarding equipment. Representatives of the different designs of snowboarding wrist protectors available were tested in a manner similar to that defined in EN 14120. The damping characteristics of the products were investigated in drop tests, and wrist extension properties were studied using bending tests. In general the test results indicate that the roller sports standard is also applicable to snowboarding products; however, adjustments to the test conditions and the performance requirements need to be discussed. This relates, first of all, to the damping test conditions, because most products failed the performance criteria. In addition, a higher upper threshold value in the bending test (currently 55°) seems to be more related to falls observed in snowboarding. Furthermore, the influence of test conditions such as the temperature needs to be checked. Nevertheless, the introduction of a modified standard should be a reasonable undertaking to improve the protective potential of wrist guards.
Interlaboratory Study of ASTM F2731, Standard Test Method for Measuring the Transmitted and Stored Energy of Firefighter Protective Clothing Systems
Authors: Deuser L. , Barker R. , Deaton A. S. , Shepherd A.
Published: May 2012

This paper describes an interlaboratory study conducted using ASTM F2731, Standard Test Method for Measuring the Transmitted and Stored Energy of Firefighter Protective Clothing Systems. Five replications of six different composites representative of firefighting turnout gear materials were tested at six different laboratories equipped to conduct the test. Data collected were used to predict the time to second degree burn for each of the turnout composite test specimens. Statistical analysis showed good agreement between test sites. This interlaboratory study confirmed the repeatability and reliability of ASTM F2731, a test method used to measure an important property associated with the thermal protective performance of firefighter turnout materials.
Temperature and Load Interaction Effects on the Fatigue Crack Growth Rate and Fracture Surface Morphology of IN100 Superalloy
Authors: Adair Benjamin S. , Johnson W. Steven , Antolovich Stephen D. , Staroselsky Alexander
Published: May 2012

A study was conducted to explore some of the load and temperature interaction effects on the fatigue crack growth rate (FCGR) of polycrystalline superalloy IN100. Load interaction testing in the form of single overloads was performed at 316°C and 649°C. Temperature interaction testing was performed by cycling between 316°C and 649°C in blocks of 1, 10, and 100 cycles. After compiling a database of constant temperature, constant amplitude FCGR for IN100, fatigue crack growth predictions assuming no load or temperature interactions were made. Experimental fatigue crack propagation data were then compared with these predictions to assess interaction effects. The fracture mechanisms observed during interaction testing using a scanning electron microscope were compared with the mechanisms present during constant temperature, constant amplitude testing. Overload interaction testing led to full crack retardation at 2.0 × overloads for both 316°C and 649°C testing. Overloading by 1.6 × at both temperatures led to retarded crack growth, whereas 1.3 × overloads at 649°C created accelerated crack growth and at 316°C the crack growth was retarded. One block alternating temperature interaction testing grew significantly faster than the non-interaction prediction, while 10 block alternating temperature interaction testing also grew faster but not to the same extent. One hundred block alternating testing grew slower than non-interaction predictions. Possible explanations for the interaction effects responsible for the observed crack growth acceleration and retardation are discussed.
Use of Cement-Treated Lightweight Soils Made from Dredged Clay
Authors: Watabe Yoichi , Noguchi Takatoshi , Mitarai Yoshio
Published: May 2012

Recently, a new runway designated “D-runway” was constructed at Tokyo Haneda Airport. In this project, a large amount of cement treated lightweight soils made from very soft and mildly contaminated dredged clay was used. In this paper, a seawall structure using a cement treated lightweight soil made from dredged clay is introduced and the quality control of these soils in the construction work is discussed. Two types of lightweight soils, “pneumatic mixing cement treated soil” and “air-foam treated lightweight soil,” were used as backfill in the seawall structures. Design and quality control of the lightweight soils were carried out, and the seawall structures were successfully constructed. The shear strength and bulk density of the lightweight soils placed in the sites met the required specifications in design.
Naphthalene Sorption to Organic Additives in Cement-Treated Soils
Authors: Lake Craig B. , Searle Jill , Bridson-Pateman Evan
Published: May 2012

Cement-based solidification/stabilization (S/S) is a widely used source-controlled treatment method for contaminated sediments. Increasingly, the technology is being used to remediate sites that contain high molecular weight organic compounds. The low level of organic content in cement-based S/S mixtures often creates a need for organic additives in the mixtures in order to improve the level of sorption in the treatment process. Very little work has been published related to the quantification of the sorption of organic contaminants to cement-based S/S mixtures and the level of improvement afforded by additives such as fly ash and organoclays. The objective of this study is to examine the sorption levels of naphthalene to several cement-treated soil mixtures with and without organic additives (i.e., fly ash and organoclay) using batch testing. It is found that the sorption values of naphthalene vary but appear to be dependent on the amount of organic carbon present in the mixture. In order to assess the potential benefit of this improved sorption for field applications, contaminant migration modeling is performed using the results obtained. It is shown that cement-based S/S remediation systems can provide long-term protection against naphthalene contaminant migration, especially cement-based S/S mixtures with organoclay additives, for the assumptions considered in the modeling.
Calculation of Kobasko's Simplified Heat Transfer Coefficients from Cooling Curve Data Obtained with Small Probes
Authors: Simencio Otero Rosa Lucia
Published: May 2012

Although heat transfer coefficient characterization of quench severity is not new, there continues to be a need for the rapid and relatively simple calculation of heat transfer coefficients from time-temperature cooling curve data files obtained via test methods such as ASTM D6200, D6482, D6549, and D7646, which utilize relatively small cylindrical test probes with diameters of ≤12.5 mm. One method that may be readily used is Kobasko’s computational method for effective heat transfer coefficients, which is based on time-temperature data obtained at the geometric center of small test probes during cooling curve analysis. A description of the step-by-step procedure for performing these calculations on actual experimental data is provided here.