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Sample records for evaluation core materials

  1. Evaluation of downmotion time interval molten materials to core catcher during core disruptive accidents postulated in LMFR

    International Nuclear Information System (INIS)

    Voronov, S.A.; Kiryushin, A.I.; Kuzavkov, N.G.; Vlasichev, G.N.

    1994-01-01

    Hypothetical core disruptive accidents are postulated to clear potential of a reactor plant to withstand extreme conditions and to generate measures for management and mitigation of accidents consequence. In Russian advanced reactors there is a core catcher below the diagrid to prevent vessel bottom melting and to localize fuel debris. In this paper the calculation technique and estimation of relocation time of molten fuel and materials are presented in the case of core disruptive accidents postulated for LMFR reactor. To evaluate minimum interval of fuel relocation time the calculations for different initial data are provided. Large mass of materials between the core and the catcher in LMFR reactor hinders molten materials relocation toward the vessel bottom. That condition increases the time interval of reaching core catcher by molten fuel. Computations performed allowed to evaluate the minimum molten materials relocation time from the core to the core catcher. This time interval is in a range of 3.5-5.5 hours. (author)

  2. Spectrophotometric Evaluation of Polyetheretherketone (PEEK as a Core Material and a Comparison with Gold Standard Core Materials

    Directory of Open Access Journals (Sweden)

    Bogna Stawarczyk

    2016-06-01

    Full Text Available This study investigated the colorimetric properties of different veneering materials on core materials. Standardized specimens (10 mm × 10 mm × 1.5 mm reflecting four core (polyetheretherketone (PEEK, zirconia (ZrO2, cobalt–chromium–molybdenum alloy (CoCrMo, and titanium oxide (TiO2; thickness: 1.5 mm and veneering materials (VITA Mark II, IPS e.max CAD, LAVA Ultimate and VITA Enamic, all in shade A3; thickness: 0.5, 1.0, 1.5 and 2 mm, respectively were fabricated. Specimens were superimposed to assemblies, and the color was determined with a spectrophotometer (CieLab-System or a chair-side color measurement device (VITA EasyShade, respectively. Data were analyzed using three-, two-, and one-way ANOVA, a Chi2-test, and a Wilson approach (p < 0.05. The measurements with EasyShade showed A2 for VITA Mark II, A3.5 for VITA Enamic, B2 for LAVA Ultimate, and B3 for IPS e.max CAD. LabE-values showed significant differences between the tested veneering materials (p < 0.001. CieLab-System and VITA EasyShade parameters of the different assemblies showed a significant impact of core (p < 0.001, veneering material (p < 0.001, and thickness of the veneering material (p < 0.001. PEEK as core material showed comparable outcomes as compared to ZrO2 and CoCrMo, with respect to CieLab-System parameters for each veneering material. The relative frequency of the measured VITA EasyShade parameters regarding PEEK cores also showed comparable results as compared to the gold standard CoCrMo, regardless of the veneering material used.

  3. Evaluation of Core Loss in Magnetic Materials Employed in Utility Grid AC Filters

    DEFF Research Database (Denmark)

    Beres, Remus Narcis; Wang, Xiongfei; Blaabjerg, Frede

    2016-01-01

    magnetic materials adopted in utility grid ac filters have been investigated and measured for both sinusoidal and rectangular excitation, with and without dc bias condition. The core loss information can ensure cost effective passive filter designs and may avoid trial-error design procedures of the passive......Inductive components play an important role in filtering the switching harmonics related to the pulse width modulation in voltage source converters. Particularly, the filter reactor on the converter side of the filter is subjected to rectangular excitation which may lead to significant losses...... in the core, depending on the magnetic material of choice and current ripple specifications. Additionally, shunt or series reactors that exists in LCL or trap filters and which are subjected to sinusoidal excitations have different specifications and requirements. Therefore, the core losses of different...

  4. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, Liancheng; Zhang, Bin

    2014-01-01

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  5. Adaptation of adhesive post and cores to dentin after in vitro occlusal loading: evaluation of post material influence.

    Science.gov (United States)

    Dietschi, Dider; Ardu, Stefano; Rossier-Gerber, Anne; Krejci, Ivo

    2006-12-01

    Fatigue resistance of post and cores is critical to the long term behavior of restored nonvital teeth. The purpose of this in vitro trial was to evaluate the influence of the post material's physical properties on the adaptation of adhesive post and core restorations after cyclic mechanical loading. Composite post and cores were made on endodontically treated deciduous bovine teeth using 3 anisotropic posts (made of carbon, quartz, or quartz-and-carbon fibers) and 3 isotropic posts (zirconium, stainless steel, titanium). Specimens were submitted to 3 successive loading phases--250,000 cycles at 50 N, 250,000 at 75 N, and 500,000 at 100 N--at a rate of 1.5 Hz. Restoration adaptation was evaluated under SEM, before and during loading (margins) and after test completion (margins and internal interfaces). Six additional samples were fabricated for the characterization of interface micromorphology using confocal microscopy. Mechanical loading increased the proportion of marginal gaps in all groups; carbon fiber posts presented the lowest final gap proportion (7.11%) compared to other stiffer metal-ceramic or softer fiber posts (11.0% to 19.1%). For internal adaptation, proportions of debonding between dentin and core or cement varied from 21.69% (carbon post) to 47.37% (stainless steel post). Debonding at the post-cement interface occurred only with isotropic materials. Confocal microscopy observation revealed that gaps were generally associated with an incomplete hybrid layer and reduced resin tags. Regardless of their rigidity, metal and ceramic isotropic posts proved less effective than fiber posts at stabilizing the post and core structure in the absence of the ferrule effect, due to the development of more interfacial defects with either composite or dentin.

  6. Evaluation of materials for retention of sodium and core debris in reactor systems. Annual progress report, September 1977-December 1978

    International Nuclear Information System (INIS)

    Swanson, D.G.; Zehms, E.H.; McClelland, J.D.; Meyer, R.A.; van Paassen, H.L.L.

    1978-12-01

    This report considers some of the consequences of a hypothetical core disruptive accident in a nuclear reactor. The interactions expected between molten core debris, liquid sodium, and materials that might be employed in an ex-vessel sacrificial-bed or in the reactor building are discussed. Experimental work performed for NRC by Sandia Laboratories and Hanford Engineering Development Laboratory on the interactions between liquid sodium and basalt concrete is reviewed. Studies of molten steel interactions with concrete at Sandia Laboratories and molten UO 2 interactions with concrete at The Aerospace Corporation are also discussed. The potential of MgO for use in core containment is discussed and refractory materials other than MgO are reviewed. Finally, results from earlier experiments with molten core debris and various materials performed at The Aerospace Corporation are presented

  7. Materials behaviour in PWRs core

    International Nuclear Information System (INIS)

    Barbu, A.; Massoud, J.P.

    2008-01-01

    Like in any industrial facility, the materials of PWR reactors are submitted to mechanical, thermal or chemical stresses during particularly long durations of operation: 40 years, and even 60 years. Materials closer to the nuclear fuel are submitted to intense bombardment of particles (mainly neutrons) coming from the nuclear reactions inside the core. In such conditions, the damages can be numerous and various: irradiation aging, thermal aging, friction wear, generalized corrosion, stress corrosion etc.. The understanding of the materials behaviour inside the cores of reactors in operation is a major concern for the nuclear industry and its long term forecast is a necessity. This article describes the main ways of materials degradation without and under irradiation, with the means used to foresee their behaviour using physics-based models. Content: 1 - structures, components and materials: structure materials, nuclear materials; 2 - main ways of degradation without irradiation: thermal aging, stress corrosion, wear; 3 - main ways of degradation under irradiation: microscopic damaging - point defects, dimensional alterations, evolution of mechanical characteristics under irradiation, irradiation-assisted stress corrosion cracking (IASCC), synergies; 4 - forecast of materials evolution under irradiation using physics-based models: primary damage - fast dynamics, primary damage annealing - slow kinetics microstructural evolution, impact of microstructural changes on the macroscopic behaviour, insight on modeling methods; 5 - materials change characterization techniques: microscopic techniques - direct defects observation, nuclear techniques using a particle beam, global measurements, mechanical characterizations; 6 - perspectives. (J.S.)

  8. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, LianCheng; Zhang, Bin

    2016-01-01

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  9. Armor systems including coated core materials

    Science.gov (United States)

    Chu, Henry S [Idaho Falls, ID; Lillo, Thomas M [Idaho Falls, ID; McHugh, Kevin M [Idaho Falls, ID

    2012-07-31

    An armor system and method involves providing a core material and a stream of atomized coating material that comprises a liquid fraction and a solid fraction. An initial layer is deposited on the core material by positioning the core material in the stream of atomized coating material wherein the solid fraction of the stream of atomized coating material is less than the liquid fraction of the stream of atomized coating material on a weight basis. An outer layer is then deposited on the initial layer by positioning the core material in the stream of atomized coating material wherein the solid fraction of the stream of atomized coating material is greater than the liquid fraction of the stream of atomized coating material on a weight basis.

  10. Reactor core materials research and integrated material database establishment

    International Nuclear Information System (INIS)

    Ryu, Woo Seog; Jang, J. S.; Kim, D. W.

    2002-03-01

    Mainly two research areas were covered in this project. One is to establish the integrated database of nuclear materials, and the other is to study the behavior of reactor core materials, which are usually under the most severe condition in the operating plants. During the stage I of the project (for three years since 1999) in- and out of reactor properties of stainless steel, the major structural material for the core structures of PWR (Pressurized Water Reactor), were evaluated and specification of nuclear grade material was established. And the damaged core components from domestic power plants, e.g. orifice of CVCS, support pin of CRGT, etc. were investigated and the causes were revealed. To acquire more resistant materials to the nuclear environments, development of the alternative alloys was also conducted. For the integrated DB establishment, a task force team was set up including director of nuclear materials technology team, and projector leaders and relevant members from each project. The DB is now opened in public through the Internet

  11. Core design studies on various forms of coolants and fuel materials. 2. Studies on liquid heavy metal and gas cooled cores, small cores and evaluation of 4-type cores

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Sakashita, Yoshiyuki; Naganuma, Masayuki; Takaki, Naoyuki; Mizuno, Tomoyasu; Ikegami, Tetsuo

    2001-01-01

    Alternative concepts to sodium cooled fast reactors, such as heavy metal liquid cooled reactors and gas cooled fast reactors were studied in Phase-1 of the feasibility studies, aiming at simplification of the system, high thermal efficiency and enhancing safety. Fuel and core specifications and nuclear characteristics were surveyed to meet the targets for commercialization of fast reactor cycle. Nuclear characteristics of small fast reactor cores were also surveyed from the perspective of the possibility of multi-purpose use and dispersed power stations. The key points of the design study for each concept in Phase-2 were summarized from the aspect of the screening of the candidates for FR commercialization. (author)

  12. Liquid metal reactor core material HT9

    International Nuclear Information System (INIS)

    Kim, S. H.; Kuk, I. H.; Ryu, W. S. and others

    1998-03-01

    A state-of-the art is surveyed on the liquid metal reactor core materials HT9. The purpose of this report is to give an insight for choosing and developing the materials to be applied to the KAERI prototype liquid metal reactor which is planned for the year of 2010. In-core stability of cladding materials is important to the extension of fuel burnup. Austenitic stainless steel (AISI 316) has been used as core material in the early LMR due to the good mechanical properties at high temperatures, but it has been found to show a poor swelling resistance. So many efforts have been made to solve this problem that HT9 have been developed. HT9 is 12Cr-1MoVW steel. The microstructure of HT9 consisted of tempered martensite with dispersed carbide. HT9 has superior irradiation swelling resistance as other BCC metals, and good sodium compatibility. HT9 has also a good irradiation creep properties below 500 dg C, but irradiation creep properties are degraded above 500 dg C. Researches are currently in progress to modify the HT9 in order to improve the irradiation creep properties above 500 dg C. New design studies for decreasing the core temperature below 500 dg C are needed to use HT9 as a core material. On the contrary, decrease of the thermal efficiency may occur due to lower-down of the operation temperature. (author). 51 refs., 6 tabs., 19 figs

  13. Core Design Concept and Core Structural Material Development for a Prototype SFR

    International Nuclear Information System (INIS)

    Chang, Jinwook

    2013-01-01

    Core design Concept: – Initial core is Uranium metal fueled core, then it will evolve into TRU core; – Tight pressure drop constraint lowers power density; – Trade-off studies with relaxed pressure drop constraint (~0.4MPa) are on-going; – Major feature will be finalized this year. • KAERI is developing advanced cladding for high burnup fuel in Ptototype SFR: – Advanced cladding materials are now developing, which shows superior high temperature mechanical property to the conventional material; – Processing technologies related to tube making process are now developed to enhance high temperature mechanical propertyl – Preliminary HT9 cladding tube was manufactured and out-of pile mechanical properties were evaluated. Advanced cladding tube is now being developed and being prepared for irradiation test

  14. The materials challenge for LFR core design

    International Nuclear Information System (INIS)

    Grasso, Giacomo; Agostini, Pietro

    2013-01-01

    LFR share the main issues of all Fast Reactors, while presenting specific issues due to the use of lead as coolant. A number of constraints impairs the design of a LFR core, possibly resulting in a viability domain not exploitable for producing electricity in an efficient (hence economic) way. In particular, the most restrictive issues to be faced pend on the cladding. The selection of proper cladding materials provides the solution for the issues impairing the resistance of the cladding against stresses and irradiation effects. On the other hand, the protection of the cladding requires surface protections like oxide scales (passivation) or adherent layers (coating). Oxide scales seem not sufficient for a stable and effective protection of the base material. The application of adherent layers seems the only promising solution for protecting the cladding against corrosion. For the short term (i.e.: ALFRED), advanced 15/15Ti with coating is the reference solution for the cladding, allowing a core design complying with all the design constraints and goals. The candidate coatings are already being tested under irradiation to proceed towards qualification. In parallel, new base materials and/or coatings are presently under investigation. For the long term (i.e.: ELFR), the availability of such advanced materials/coatings might allow the extension of the viability domain towards higher and broader ranges (temperature, dpa, etc.), extending the fields of applications of LFRs and resulting in higher performances

  15. New concept of damage evaluation method for core internal materials considering radiation induced stress relaxation (1). Experiments and modeling of radiation effects

    International Nuclear Information System (INIS)

    Miwa, Yukio; Kondo, Keietsu; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi

    2009-01-01

    In order to build the new concept of material damage evaluation method, synergistic effect of radiation and residual stress on material degradation was estimated experimentally, and the effect of radiation induced stress relaxation on retardation of material degradation was observed. (author)

  16. Evaluation of core distortion in FBR

    International Nuclear Information System (INIS)

    Ikarimoto, I.; Tanaka, M.; Okubo, Y.

    1984-01-01

    The analyses of FBR's core distortion are mainly performed in order to evaluate the following items: 1) Change of reactivity; 2) Force at pads on core assemblies; 3) Withdrawal force at refueling; 4) Loading, refueling and residual deviations of wrapper tubes (core assemblies) at the top; 5) Bowing modes of guide tubes for control rods. The analysis of core distortion are performed by using computer program for two-dimensional row deformation analysis or three-dimensional core deformation if necessary, considering these evaluated items which become design conditions. This report shows the relationship between core deformation analysis and component design, a point of view of choosing an analysis program for design considering core characteristics, and computing examples of core deformation of prototype class reactor by the above code. (author)

  17. Materials interaction tests to identify base and coating materials for an enhanced in-vessel core catcher design

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Knudson, D.L.; Condie, K.G.; Swank, W.D. [Idaho National Engineering and Environmental Laboratory, Idaho Falls ID (United States); Cheung, F.B. [Pennsylvania State University, Department of Mechanical and Nuclear Engineering, University Park PA (United States); Suh, K.Y. [Seoul National University, Department of Nuclear Engineering, Seoul (Korea, Republic of); Kim, S.B. [Korea Atomic Energy Research Institute, Severe Accident Research Project, Taejon (Korea, Republic of)

    2004-07-01

    An enhanced in-vessel core catcher is being designed and evaluated, it must ensure In-Vessel Retention of core materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulating oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. Initial evaluations suggest that a thermally-sprayed oxide material is the most promising candidate insulator coating for a core catcher. Tests suggest that 2 coatings can provide adequate protection to a stainless steel core catcher: -) a 500 {mu}m thick zirconium dioxide coating over a 100-200 {mu}m Inconel 718 bond coating, and -) a 500 {mu}m thick magnesium zirconate coating.

  18. Evaluation of learning materials

    DEFF Research Database (Denmark)

    Bundsgaard, Jeppe; Hansen, Thomas Illum

    2011-01-01

    This paper presents a holistic framework for evaluating learning materials and designs for learning. A holistic evaluation comprises investigations of the potential learning potential, the actualized learning potential, and the actual learning. Each aspect is explained and exemplified through...

  19. Lifetime embrittlement of reactor core materials

    International Nuclear Information System (INIS)

    Kreyns, P.H..; Bourgeois, W.F.; Charpentier, P.L.; Kammenzind, B.F.; Franklin, D.G.; White, C.J.

    1994-08-01

    Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of wrought Zircaloy-2, Zircaloy-4, and hafnium; Zircaloy-4 to hafnium butt welds; and hydrogen precharged beta treated and weld metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 10 24 n/M 2 (> 1 Mev). While Zircaloy-4 did not exhibit a decrement in K IC due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes cleavage fracture in hafnium and hafnium-Zircaloy welds, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen, over time, to drift from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to- Zircaloy butt-weld applications due to its absence of a radiation- induced reduction in K IC plus its lower hydrogen absorption characteristics compared with Zircaloy-2

  20. Adhesion of resin composite core materials to dentin.

    Science.gov (United States)

    O'Keefe, K L; Powers, J M

    2001-01-01

    This study determined (1) the effect of polymerization mode of resin composite core materials and dental adhesives on the bond strength to dentin, and (2) if dental adhesives perform as well to dentin etched with phosphoric acid as to dentin etched with self-etching primer. Human third molars were sectioned 2 mm from the highest pulp horn and polished. Three core materials (Fluorocore [dual cured], Core Paste [self-cured], and Clearfil Photo Core [light cured]) and two adhesives (Prime & Bond NT Dual Cure and Clearfil SE Bond [light cured]) were bonded to dentin using two dentin etching conditions. After storage, specimens were debonded in microtension and bond strengths were calculated. Scanning electron micrographs of representative bonding interfaces were analyzed. Analysis showed differences among core materials, adhesives, and etching conditions. Among core materials, dual-cured Fluorocore had the highest bond strengths. There were incompatibilities between self-cured Core Paste and Prime & Bond NT in both etched (0 MPa) and nonetched (3.0 MPa) dentin. Among adhesives, in most cases Clearfil SE Bond had higher bond strengths than Prime & Bond NT and bond strengths were higher to self-etched than to phosphoric acid-etched dentin. Scanning electron micrographs did not show a relationship between resin tags and bond strengths. There were incompatibilities between a self-cured core material and a dual-cured adhesive. All other combinations of core materials and adhesives produced strong in vitro bond strengths both in the self-etched and phosphoric acid-etched conditions.

  1. Molten core material holding device in a nuclear reactor

    International Nuclear Information System (INIS)

    Nakamura, Hisashi; Tanaka, Nobuo; Takahashi, Katsuro.

    1985-01-01

    Purpose: To improve the function of cooling to hold molten core materials in a molten core material holding device. Constitution: Plenum structures are formed into a pan-like configuration, in which liners made of metal having high melting point and relatively high heat conductivity such as tantalum, tungsten, rhenium or alloys thereof are integrally appended to hold and directly cool the molten reactor core materials. Further, a plurality of heat pipes, passing through the plenum structures, facing the cooling portion thereof to the coolants at the outer side and immersing the heating portion into the molten core materials fallen to deposit in the inner liners are disposed radially. Furthermore, heat pipes embodded in the plenum structure are disposed in the same manner below the liners. Thus, the plenum structures and the molten reactor core materials can be cooled at a high efficiency. (Seki, T.)

  2. Scaling of Core Material in Rubble Mound Breakwater Model Tests

    DEFF Research Database (Denmark)

    Burcharth, H. F.; Liu, Z.; Troch, P.

    1999-01-01

    The permeability of the core material influences armour stability, wave run-up and wave overtopping. The main problem related to the scaling of core materials in models is that the hydraulic gradient and the pore velocity are varying in space and time. This makes it impossible to arrive at a fully...... correct scaling. The paper presents an empirical formula for the estimation of the wave induced pressure gradient in the core, based on measurements in models and a prototype. The formula, together with the Forchheimer equation can be used for the estimation of pore velocities in cores. The paper proposes...... that the diameter of the core material in models is chosen in such a way that the Froude scale law holds for a characteristic pore velocity. The characteristic pore velocity is chosen as the average velocity of a most critical area in the core with respect to porous flow. Finally the method is demonstrated...

  3. Radionuclide sorption on granitic drill core material

    International Nuclear Information System (INIS)

    Eriksen, T.E.; Locklund, B.

    1987-11-01

    Distribution ratios were determined for Sr-85, Cs-134 and Eu-152 on crushed granite and fissure coating/filling material from Stripa mines. Measurements were also carried out on intact fissure surfaces. The experimental data for Sr-85, Cs-134 on crushed material can be accomodated by a sorption model based on the assumption that the crushed material consists of porous spheres with outer and inner surfaces available for sorption. In the case of Eu-152 only sorption on the outer surfaces of the crushed material was observed. The absence of sorption on inner surfaces is most probably due to high depletion of the more strongly sorbed Eu-152 in the water phase and very low diffusivity of Eu-152 in the sorbed state. (orig./HP)

  4. Effect of different composite core materials on fracture resistance of endodontically treated teeth restored with FRC posts

    OpenAIRE

    PANITIWAT, Prapaporn; SALIMEE, Prarom

    2017-01-01

    Abstract Objective This study evaluated the fracture resistance of endodontically treated teeth restored with fiber reinforced composite posts, using three resin composite core build-up materials, (Clearfil Photo Core (CPC), MultiCore Flow (MCF), and LuxaCore Z-Dual (LCZ)), and a nanohybrid composite, (Tetric N-Ceram (TNC)). Material and Methods Forty endodontically treated lower first premolars were restored with quartz fiber posts (D.T. Light-Post) cemented with resin cement (Panavia F2...

  5. A finite element thermal analysis of various dowel and core materials

    Directory of Open Access Journals (Sweden)

    Shanti Varghese

    2012-01-01

    Conclusion: Non-metallic dowel and core materials such as fibre reinforced composite dowels (FRC generate greater stress than metallic dowel and core materials. This emphasized the preferable use of the metallic dowel and core materials in the oral environment.

  6. The Common Core State Standards and the Role of Instructional Materials: A Case Study on EdReports.org

    Science.gov (United States)

    Watt, Michael G.

    2016-01-01

    The purpose of this study was to review research studies investigating the role of instructional materials in relation to the Common Core State Standards and to evaluate whether a new organisation, EdReports.org, founded to evaluate the alignment of instructional materials to the Common Core State Standards, has achieved its objectives. Content…

  7. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.; Driscoll, M.J.; Lanning, D.D.

    1979-08-01

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235 U/UO 2 : Pu/ThO 2 : 233 U/ThO 2 - and the conventional recycle-mode uranium system - 235 U/UO 2 : Pu/UO 2 . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

  8. Fatigue Characterization of Fire Resistant Syntactic Foam Core Material

    Science.gov (United States)

    Hossain, Mohammad Mynul

    Eco-Core is a fire resistant material for sandwich structural application; it was developed at NC A&T State University. The Eco-Core is made of very small amount of phenolic resin and large volume of flyash by a syntactic process. The process development, static mechanical and fracture, fire and toxicity safety and water absorption properties and the design of sandwich structural panels with Eco-Core material was established and published in the literature. One of the important properties that is needed for application in transportation vehicles is the fatigue performance under different stress states. Fatigue data are not available even for general syntactic foams. The objective of this research is to investigate the fatigue performance of Eco-Core under three types of stress states, namely, cyclic compression, shear and flexure, then document failure modes, and develop empherical equations for predicting fatigue life of Eco-Core under three stress states. Compression-Compression fatigue was performed directly on Eco-Core cylindrical specimen, whereas shear and flexure fatigue tests were performed using sandwich beam made of E glass-Vinyl Ester face sheet and Eco-Core material. Compression-compression fatigue test study was conducted at two values of stress ratios (R=10 and 5), for the maximum compression stress (sigmamin) range of 60% to 90% of compression strength (sigmac = 19.6 +/- 0.25 MPa) for R=10 and 95% to 80% of compression strength for R=5. The failure modes were characterized by the material compliance change: On-set (2% compliance change), propagation (5%) and ultimate failure (7%). The number of load cycles correspond to each of these three damages were characterized as on-set, propagation and total lives. A similar approach was used in shear and flexure fatigue tests with stress ratio of R=0.1. The fatigue stress-number of load cycles data followed the standard power law equation for all three stress states. The constant of the equation were

  9. Applications of simulation experiments in LMFBR core materials technology

    International Nuclear Information System (INIS)

    Appleby, W.K.

    1976-01-01

    The development of charged particle bombardment experiments to simulate neutron irradiation induced swelling in austenitic alloys is briefly described. The applications of these techniques in LMFBR core materials technology are discussed. It is shown that use of the techniques to study the behavior of cold-worked Type-316 was instrumental in demonstrating at an early date the need for advanced materials. The simulation techniques then were used to identify alloying elements which can markedly decrease swelling and thus a focused reactor irradiation program is now in place to allow the future use of a lower swelling alloy for LMFBR core components

  10. Statistical evaluations of current sampling procedures and incomplete core recovery

    International Nuclear Information System (INIS)

    Heasler, P.G.; Jensen, L.

    1994-03-01

    This document develops two formulas that describe the effects of incomplete recovery on core sampling results for the Hanford waste tanks. The formulas evaluate incomplete core recovery from a worst-case (i.e.,biased) and best-case (i.e., unbiased) perspective. A core sampler is unbiased if the sample material recovered is a random sample of the material in the tank, while any sampler that preferentially recovers a particular type of waste over others is a biased sampler. There is strong evidence to indicate that the push-mode sampler presently used at the Hanford site is a biased one. The formulas presented here show the effects of incomplete core recovery on the accuracy of composition measurements, as functions of the vertical variability in the waste. These equations are evaluated using vertical variability estimates from previously sampled tanks (B110, U110, C109). Assuming that the values of vertical variability used in this study adequately describes the Hanford tank farm, one can use the formulas to compute the effect of incomplete recovery on the accuracy of an average constituent estimate. To determine acceptable recovery limits, we have assumed that the relative error of such an estimate should be no more than 20%

  11. Nanostructured core-shell electrode materials for electrochemical capacitors

    Science.gov (United States)

    Jiang, Long-bo; Yuan, Xing-zhong; Liang, Jie; Zhang, Jin; Wang, Hou; Zeng, Guang-ming

    2016-11-01

    Core-shell nanostructure represents a unique system for applications in electrochemical energy storage devices. Owing to the unique characteristics featuring high power delivery and long-term cycling stability, electrochemical capacitors (ECs) have emerged as one of the most attractive electrochemical storage systems since they can complement or even replace batteries in the energy storage field, especially when high power delivery or uptake is needed. This review aims to summarize recent progress on core-shell nanostructures for advanced supercapacitor applications in view of their hierarchical architecture which not only create the desired hierarchical porous channels, but also possess higher electrical conductivity and better structural mechanical stability. The core-shell nanostructures include carbon/carbon, carbon/metal oxide, carbon/conducting polymer, metal oxide/metal oxide, metal oxide/conducting polymer, conducting polymer/conducting polymer, and even more complex ternary core-shell nanoparticles. The preparation strategies, electrochemical performances, and structural stabilities of core-shell materials for ECs are summarized. The relationship between core-shell nanostructure and electrochemical performance is discussed in detail. In addition, the challenges and new trends in core-shell nanomaterials development have also been proposed.

  12. Radiation quality factor of spherical antennas with material cores

    DEFF Research Database (Denmark)

    Hansen, Troels Vejle; Kim, Oleksiy S.; Breinbjerg, Olav

    2011-01-01

    This paper gives a description of the radiation quality factor and resonances of spherical antennas with material cores. Conditions for cavity and radiating resonances are given, and a theoretical description of the radiation quality factor, as well as simple expressions describing the relative...

  13. Use of stainless steel as structural materials in reactor cores

    International Nuclear Information System (INIS)

    Teodoro, C.A.

    1990-01-01

    Austenitic stainless steels are used as structural materials in reactor cores, due to their good mechanical properties at working temperatures and high generalized corrosion resistance in aqueous medium. The objective of this paper is to compare several 300 series austenitic stainless steels related to mechanical properties, localized corrosion resistance (SCC and intergranular) and content of delta ferrite. (author)

  14. Assessment of core structural materials and surveillance programme of research reactors. Report of the consultants meeting. Working material

    International Nuclear Information System (INIS)

    2009-01-01

    A series of presentations on the assessment of core structural components and materials at their facilities were given by the experts. The different issues related to degradation mechanisms were discussed. The outputs include a more thorough understanding of the specific challenges related to Research Reactors (RRs) as well as proposals for activities which could assist RR organizations in their efforts to address the issues involved. The experts recommend that research reactor operators consider implementation of surveillance programs for materials of core structural components, as part of ageing management program (TECDOC-792 and DS-412). It is recognised by experts that adequate archived structural material data is not available for many RRs. Access to this data and extension of existing material databases could help many operating organisations extend the operation of their RRs. The experts agreed that an IAEA Technical Meeting (TM) on Assessment of Core Structural Materials should be organised in December 2009 (IAEA HQ Vienna). The proposed objectives of the TM are: (i) exchange of detailed technical information on the assessment and ageing management of core structural materials, (ii) identification of materials of interest for further investigation, (iii) proposal for a new IAEA CRP on Assessment of Core Structural Materials, and (iv) identification of RRs prepared to participate in proposed CRP. Based on the response to a questionnaire prepared for the 2008 meeting of the Technical Working Group for Research Reactors, the number of engineering capital projects related to core structural components is proportionally lower than those related to,for example, I and C or electrical power systems. This implies that many operating research reactors will be operating longer using their original core structural components and justifies the assessment and evaluation programmes and activities proposed in this report. (author)

  15. Electrical properties of spherical dipole antennas with lossy material cores

    DEFF Research Database (Denmark)

    Hansen, Troels Vejle; Kim, Oleksiy S.; Breinbjerg, Olav

    2012-01-01

    A spherical magnetic dipole antenna with a linear, isotropic, homogenous, passive, and lossy material core is modeled analytically, and closed form expressions are given for the internally stored magnetic and electric energies, the radiation efficiency, and radiation quality factor. This model...... and all the provided expressions are exact and valid for arbitrary core sizes, permeability, permittivity, electric and magnetic loss tangents. Arbitrary dispersion models for both permeability and permittivity can be applied. In addition, we present an investigation for an antenna of fixed electrical...

  16. The recent advances on carrier materials for microencapsulating lipophilic cores

    Directory of Open Access Journals (Sweden)

    JIN Minfeng

    2014-12-01

    Full Text Available Lipophilic ingredients,such as polyunsaturated fatty acids,play an important role in industrialized foods to fortify the nutrients.However,these materials are normally sensitive to oxygen,light or heat to be oxidized,and hard to flow and mix within the bulk food due to the hydrophobic nature.Microencapsulation of lipophilic materials could effectively extend their shelf lives,mask unsatisfied flavors,change their physicochemical properties,and enhance the mixing capacities.This work reviewed the different carrier materials applied in microencapsulating the lipophilic ingredients,and discussed their characteristics and effects on encapsulation efficiencies and release profiles of lipophilic cores.

  17. [Comparative investigation of compressive resistance of glass-cermet cements used as a core material in post-core systems].

    Science.gov (United States)

    Ersoy, E; Cetiner, S; Koçak, F

    1989-09-01

    In post-core applications, addition to the cast designs restorations that are performed on fabrication posts with restorative materials are being used. To improve the physical properties of glass-ionomer cements that are popular today, glass-cermet cements have been introduced and those materials have been proposed to be an alternative restorative material in post-core applications. In this study, the compressive resistance of Ketac-Silver as a core material was investigated comparatively with amalgam and composite resins.

  18. Holistic evaluations of learning materials

    DEFF Research Database (Denmark)

    Bundsgaard, Jeppe; Hansen, Thomas Illum

    2011-01-01

    The aim of this paper is to present a holistic framework for evaluating learning materials and designs for learning. A holistic evaluation of learning material comprises investigations of - the potential learning potential, i.e. the affordances and challenges of the learning material...

  19. Evaluation of in-place concrete strength by core testing.

    Science.gov (United States)

    2016-11-01

    The overall objective of the work contained in this report is to develop an ALDOT procedure to evaluate core strength results obtained under various conditions. Since there are many factors that influence the apparent strength of cores, strength corr...

  20. Material control evaluation

    International Nuclear Information System (INIS)

    Waddoups, I.G.; Anspach, D.A.; Abbott, J.A.

    1993-01-01

    Changes in the Department of Energy's (DOE) scope of work have stimulated several laboratories and commercial companies to develop and apply technology to enhance nuclear material control. Accountability, inventory, radiation exposure, and insider protection concerns increase as many DOE facilities require increased storage. This paper summarizes a study of the existing material control technologies. The goal of the study is to identify, characterize, and quantify the trade-offs associated with using these technologies to provide real-time information on stored nuclear material that in turn supports decreasing the frequency of inventories conducted by site personnel

  1. Comparative study of mechanical properties of direct core build-up materials

    Directory of Open Access Journals (Sweden)

    Girish Kumar

    2015-01-01

    Full Text Available Background and Objectives: The strength greatly influences the selection of core material because core must withstand forces due to mastication and para-function for many years. This study was conducted to evaluate certain mechanical properties of commonly used materials for direct core build-up, including visible light cured composite, polyacid modified composite, resin modified glass ionomer, high copper amalgam, and silver cermet cement. Materials and Methods: All the materials were manipulated according to the manufacturer′s recommendations and standard test specimens were prepared. A universal testing machine at different cross-head speed was used to determine all the four mechanical properties. Mean compressive strength, diametral tensile strength, flexural strength, and elastic modulus with standard deviations were calculated. Multiple comparisons of the materials were also done. Results: Considerable differences in compressive strength, diametral tensile strength, and flexural strength were observed. Visible light cured composite showed relatively high compressive strength, diametral tensile strength, and flexural strength compared with the other tested materials. Amalgam showed the highest value for elastic modulus. Silver cermet showed less value for all the properties except for elastic modulus. Conclusions: Strength is one of the most important criteria for selection of a core material. Stronger materials better resist deformation and fracture provide more equitable stress distribution, greater stability, and greater probability of clinical success.

  2. The influence of core materials and mix on the performance of a 100 kVA three phase transformer core

    Energy Technology Data Exchange (ETDEWEB)

    Snell, David E-mail: dave.snell@cogent-power.com; Coombs, Alan

    2003-01-01

    Various grades of grain-oriented electrical steel, and the effect of mixing domain refined and non-domain refined materials in the same three phase transformer core have been assessed using a developed computer-based test system. Ball unit domain refined material and non-domain refined material can be successfully mixed in the same core, without degrading performance.

  3. Survey of melt interactions with core retention material

    International Nuclear Information System (INIS)

    Powers, D.A.

    1979-01-01

    A survey of the interactions of up to 220 kg stainless steel melts at 1973 0 K with the candidate core retention materials borax, firebrick, high alumina cement, and magnesia is described. Data collected for the interactions include rates of material erosion, aerosol generation, gas evolution, and upward heat flux. Borax acts as an ablative solid that rapidly quenches the melt. Firebrick is ablated by the steel melt at a rate of 8.2 x 10 -6 m/s. High alumina cement is found to be an attractive melt retention material especially if it can be used in the unhydrated form. Magnesia is also found to be an attractive material though it can be eroded by the molten oxides of steel

  4. Opalescence of all-ceramic core and veneer materials.

    Science.gov (United States)

    Cho, Moon-Sang; Yu, Bin; Lee, Yong-Keun

    2009-06-01

    The enamel of natural teeth is opalescent, where there is light scattering of the shorter wavelengths of the visible spectrum, giving a tooth a bluish appearance in the reflected color and an orange/brown appearance in the transmitted color. The objective of this study was to determine the opalescence of all-ceramic core, veneer and layered specimens with a color measuring spectrophotometer. Colors of core (A2-corresponding shade), veneer (A2- and A3-corresponding shades) and layered (A2- and A3-layered) ceramics for all-ceramic restorations in clinically relevant thicknesses were measured in the reflectance and transmittance modes. The opalescence parameter (OP), which was calculated as the difference in blue-yellow coordinate (Deltab(*)) and red-green coordinate (Deltaa(*)), and the differences in blue-yellow coordinate (Deltab(*)) and in color (DeltaE(ab)(*)) between the reflected and transmitted colors were calculated. One-way ANOVA was performed for the OP values of the core, veneer and layered specimens by the kind of materials. Regression analysis was performed between the OP and Deltab(*), and the OP and DeltaE(ab)(*) values. The range of the OP value was 1.6-6.1, 2.0-7.1, 1.3-5.0 and 1.6-4.2 for the core, veneer, A2- and A3-layered specimens, respectively, all of which were significantly influenced by the kind of materials (pOpalescence varied by kind of ceramics. The OP values of ceramics were lower than those of tooth enamel. All-ceramic materials that can simulate the opalescence of natural teeth should be developed.

  5. New sacrificial material for ex-vessel core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Komlev, Andrei A., E-mail: komlev@kth.se [Kungliga Tekniska Högskolan (KTH), AlbaNova University Centre, Nuclear Power Safety Division, Roslagstullsbacken 21, SE-106 91, Stockholm (Sweden); Almjashev, Vyacheslav I., E-mail: vac@mail.ru [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Bechta, Sevostian V., E-mail: bechta@safety.sci.kth.se [Kungliga Tekniska Högskolan (KTH), AlbaNova University Centre, Roslagstullsbacken 21, SE-106 91, Stockholm (Sweden); Khabensky, Vladimir B., E-mail: vladimirkhabensky@gmail.com [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Granovsky, Vladimir S., E-mail: gran@niti.ru [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Gusarov, Victor V., E-mail: victor.v.gusarov@gmail.com [Ioffe Institute, 26 Polytekhnicheskaya Str., St. Petersburg, 194021 (Russian Federation)

    2015-12-15

    A new functional (sacrificial) material has been developed in the Fe{sub 2}O{sub 3}–SrO–Al{sub 2}O{sub 3}–CaO system based on strontium hexaferrite ceramic in concrete matrix. The method of producing SM has been advanced technologically; this technological effectiveness allows the SM to be used in ex-vessel core catchers with corium spreading as well as in crucible-type core catchers. Critical properties regarding the efficiency of SM in ex-vessel core catchers, such as porosity, pycnometric density, apparent density, solidus and liquidus temperatures, and water content have been measured. Suitable fractions of SrFe{sub 12}O{sub 19} and high alumina cement (HAC) were found in the SM based on thermodynamic analysis of the SM/corium interaction. The use of sacrificial steel as an additional heat adsorption component in the core catcher allowed us to increase the mass fraction range of SrFe{sub 12}O{sub 19} in the SM from 0.3−0.5 to 0.3–0.85. The activation temperature of the SM/corium interaction has been shown to correspond to the liquidus temperature of the local composition at the SM/corium interface. The calculated value of this temperature was 1716 °C. Analysis of phase transformations in the SrO–Fe{sub 2}O{sub 3} system revealed advantages of the SrFe{sub 12}O{sub 19}–based sacrificial material compared with the Fe{sub 2}O{sub 3}-contained material owing to the time proximity of SrFe{sub 12}O{sub 19} decomposition and corium interaction activation. - Highlights: • A sacrificial material (SM) was developed for ex-vessel core catcher. • Suitable proportions in the SrFe{sub 12}O{sub 19}–Al{sub 2}O{sub 3}·CaO–Fe system were determined. • Hydrogen release limitation was shown for ex-vessel corium retention with the SM. • Calculated temperature of the active initiation of corium/SM interaction is 1716 °C. • Functional properties of the SM were measured.

  6. Safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release

    International Nuclear Information System (INIS)

    Pointner, W.; Broecker, A.

    2012-01-01

    The report on safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release covers the following issues: assessment of the relevant status for PWR, evaluation of the national and international (USA, Canada, France) status, actualization of recommendations, transferability from PWR to BWR. Generic studies on the core cooling capability in case of insulation material release in BWR-type reactors were evaluated.

  7. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

  8. Penetration of molten core materials into basaltic and limestone concrete

    International Nuclear Information System (INIS)

    Sutherland, H.J.

    1978-01-01

    In conjunction with the small-scale, melt-concrete interaction tests being conducted at Sandia Laboratories, an acoustic technique has been used to monitor the penetration of molten core materials into basaltic and limestone concrete. Real time plots of the position of the melt/concrete interface have been obtained, and they illustrate that the initial penetration rate of the melt may be of the order of 80 mm/min. Phenomena deduced by the technique include a non-wetted melt/concrete interface

  9. Molten LWR core material interactions with water and with concrete

    International Nuclear Information System (INIS)

    Dahlgren, D.A.; Buxton, L.D.; Muir, J.F.; Murfin, W.B.; Nelson, L.S.; Powers, D.A.

    1977-01-01

    Nuclear power reactors are designed and operated to minimize the possibility of fuel melting. Nevertheless, in order to assess the risks associated with reactor operation, a realistic assessment is required for postulated accident sequences in which melting occurs. To investigate the experimental basis of the fuel melt accident analyses, a comprehensive review was performed at Sandia Laboratories. The results of that study indicated several phenomenological areas where additional experimental data should be gathered to verify common assumptions made in risk studies. In particular, vapor explosions and molten core material/concrete interactions were identified for further study. Results of these studies are presented

  10. Strength evaluation code STEP for brittle materials

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Futakawa, Masatoshi.

    1997-12-01

    In a structural design using brittle materials such as graphite and/or ceramics it is necessary to evaluate the strength of component under complex stress condition. The strength of ceramic materials is said to be influenced by the stress distribution. However, in the structural design criteria simplified stress limits had been adopted without taking account of the strength change with the stress distribution. It is, therefore, important to evaluate the strength of component on the basis of the fracture model for brittle material. Consequently, the strength evaluation program, STEP, on a brittle fracture of ceramic materials based on the competing risk theory had been developed. Two different brittle fracture modes, a surface layer fracture mode dominated by surface flaws and an internal fracture mode by internal flaws, are treated in the STEP code in order to evaluate the strength of brittle fracture. The STEP code uses stress calculation results including complex shape of structures analyzed by the generalized FEM stress analysis code, ABAQUS, so as to be possible to evaluate the strength of brittle fracture for the structures having complicate shapes. This code is, therefore, useful to evaluate the structural integrity of arbitrary shapes of components such as core graphite components in the HTTR, heat exchanger components made of ceramics materials etc. This paper describes the basic equations applying to the STEP code, code system with a combination of the STEP and the ABAQUS codes and the result of the verification analysis. (author)

  11. Evaluating Community Health Advisor (CHA) Core Competencies: The CHA Core Competency Retrospective Pretest/Posttest (CCCRP).

    Science.gov (United States)

    Story, Lachel; To, Yen M

    2016-05-01

    Health care and academic systems are increasingly collaborating with community health advisors (CHAs) to provide culturally relevant health interventions that promote sustained community transformation. Little attention has been placed on CHA training evaluation, including core competency attainment. This study identified common CHA core competencies, generated a theoretically based measure of those competencies, and explored psychometric properties of that measure. A concept synthesis revealed five CHA core competencies (leadership, translation, guidance, advocacy, and caring). The CHA Core Competency Retrospective Pretest/Posttest (CCCRP) resulted from that synthesis, which was administered using multiple approaches to individuals who previously received CHA training (N= 142). Exploratory factor analyses revealed a two-factor structure underlying the posttraining data, and Cronbach's alpha indicated high internal consistency. This study suggested some CHA core competencies might be more interrelated than previously thought, and two major competencies exist rather than five and supported the CCCRP's use to evaluate core competency attainment resulting from training. © The Author(s) 2014.

  12. Static analysis of material testing reactor cores:critical core calculations

    International Nuclear Information System (INIS)

    Nawaz, A. A.; Khan, R. F. H.; Ahmad, N.

    1999-01-01

    A methodology has been described to study the effect of number of fuel plates per fuel element on critical cores of Material Testing Reactors (MTR). When the number of fuel plates are varied in a fuel element by keeping the fuel loading per fuel element constant, the fuel density in the fuel plates varies. Due to this variation, the water channel width needs to be recalculated. For a given number of fuel plates, water channel width was determined by optimizing k i nfinity using a transport theory lattice code WIMS-D/4. The dimensions of fuel element and control fuel element were determined using this optimized water channel width. For the calculated dimensions, the critical cores were determined for the given number of fuel plates per fuel element by using three dimensional diffusion theory code CITATION. The optimization of water channel width gives rise to a channel width of 2.1 mm when the number of fuel plates is 23 with 290 g ''2''3''5U fuel loading which is the same as in the case of Pakistan Reactor-1 (PARR-1). Although the decrease in number of fuel element results in an increase in optimal water channel width but the thickness of standard fuel element (SFE) and control fuel element (CFE) decreases and it gives rise to compact critical and equilibrium cores. The criticality studies of PARR-1 are in good agreement with the predictions

  13. Stress wave nondestructive evaluation of Douglas-fir peeler cores

    Science.gov (United States)

    Robert J. Ross; John I. Zerbe; Xiping Wang; David W. Green; Roy F. Pellerin

    2005-01-01

    With the need for evaluating the utilization of veneer peeler log cores in higher value products and the increasing importance of utilizing round timbers in poles, posts, stakes, and building construction components, we conducted a cooperative project to verify the suitability of stress wave nondestructive evaluation techniques for assessing peeler cores and some...

  14. Influence of different post core materials on the color of Empress 2 full ceramic crowns.

    Science.gov (United States)

    Ge, Jing; Wang, Xin-zhi; Feng, Hai-lan

    2006-10-20

    For esthetic consideration, dentin color post core materials were normally used for all-ceramic crown restorations. However, in some cases, clinicians have to consider combining a full ceramic crown with a metal post core. Therefore, this experiment was conducted to test the esthetical possibility of applying cast metal post core in a full ceramic crown restoration. The color of full ceramic crowns on gold and Nickel-Chrome post cores was compared with the color of the same crowns on tooth colored post cores. Different try-in pastes were used to imitate the influence of a composite cementation on the color of different restorative combinations. The majority of patients could not detect any color difference less than DeltaE 1.8 between the two ceramic samples. So, DeltaE 1.8 was taken as the objective evaluative criterion for the evaluation of color matching and patients' satisfaction. When the Empress 2 crown was combined with the gold alloy post core, the color of the resulting material was similar to that of a glass fiber reinforced resin post core (DeltaE = 0.3). The gold alloy post core and the try-in paste did not show a perceptible color change in the full ceramic crowns, which indicated that the color of the crowns might not be susceptible to change between lab and clinic as well as during the process of composite cementation. Without an opaque covering the Ni-Cr post core would cause an unacceptable color effect on the crown (DeltaE = 2.0), but with opaque covering, the color effect became more clinically satisfactory (DeltaE = 1.8). It may be possible to apply a gold alloy post core in the Empress 2 full ceramic crown restoration when necessary. If a non-extractible Ni-Cr post core exists in the root canal, it might be possible to restore the tooth with an Empress 2 crown after covering the labial surface of the core with one layer of opaque resin cement.

  15. Evaluative Review in Materials Development

    Science.gov (United States)

    Stoller, Fredricka L.; Horn, Bradley; Grabe, William; Robinson, Marin S.

    2006-01-01

    English for Academic Purposes (EAP) professionals know that initial efforts to produce or adapt materials generally require evaluative review and revision. A review process that solicits feedback from teacher and student users is critical because materials writers often find it difficult to envision the problems others may have with their…

  16. Effect of adhesive resin cements on bond strength of ceramic core materials to dentin.

    Science.gov (United States)

    Gundogdu, M; Aladag, L I

    2018-03-01

    The aim of the present study was to evaluate the effects of self-etch and self-adhesive resin cements on the shear bond strength of ceramic core materials bonded to dentin. Extracted, caries-free, human central maxillary incisor teeth were selected, and the vestibule surfaces were cut flat to obtain dentin surfaces. Ceramic core materials (IPS e.max Press and Prettau Zirconia) were luted to the dentin surfaces using three self-etch adhesive systems (Duo-Link, Panavia F 2.0, and RelyX Ultimate Clicker) and two self-adhesive resin systems (RelyX U200 Automix and Maxcem Elite). A shear bond strength test was performed using a universal testing machine. Failure modes were observed under a stereomicroscope, and bonding interfaces between the adhesive resin cements and the teeth were evaluated with a scanning electron microscope. Data were analyzed with Student's t-test and one-way analysis of variance followed by Tukey's test (α = 0.05). The type of adhesive resin cement significantly affected the shear bond strengths of ceramic core materials bonded to dentin (P materials when the specimens were luted with self-adhesive resin cements (P materials.

  17. Effect of Fuel Structure Materials on Radiation Source Term in Reactor Core Meltdown

    International Nuclear Information System (INIS)

    Jeong, Hae Sun; Ha, Kwang Soon

    2014-01-01

    The fission product (Radiation Source) releases from the reactor core into the containment is obligatorily evaluated to guarantee the safety of Nuclear Power Plant (NPP) under the hypothetical accident involving a core meltdown. The initial core inventory is used as a starting point of all radiological consequences and effects on the subsequent results of accident assessment. Hence, a proper evaluation for the inventory can be regarded as one of the most important part over the entire procedure of accident analysis. The inventory of fission products is typically evaluated on the basis of the uranium material (e.g., UO2 and USi2) loaded in nuclear fuel assembly, except for the structure materials such as the end fittings, grids, and some kinds of springs. However, the structure materials are continually activated by the neutrons generated from the nuclear fission, and some nuclides of them (e.g., 14 C and 60 Co) can significantly influence on accident assessment. During the severe core accident, the structure components can be also melted with the melting points of temperature relatively lower than uranium material. A series of the calculation were performed by using ORIGEN-S module in SCALE 6.1 package code system. The total activity in each part of structure materials was specifically analyzed from these calculations. The fission product inventory is generally evaluated based on the uranium materials of fuel only, even though the structure components of the assembly are continually activated by the neutrons generated from the nuclear fission. In this study, the activation calculation of the fuel structure materials was performed for the initial source term assessment in the accident of reactor core meltdown. As a result, the lower end fitting and the upper plenum greatly contribute to the total activity except for the cladding material. The nuclides of 56 Mn, '5 1 Cr, 55 Fe, 58 Co, 54 Mn, and 60 Co are analyzed to mainly effect on the activity. This result

  18. Phenomena in the interaction among a core melt and protective and sacrificial materials

    International Nuclear Information System (INIS)

    Steinwarz, W.; Koller, W.; Dyllong, N.; Fischer, M.; Hellmann, S.; Lansmann, V.; Nie, M.; Haefner, W.; Alkan, Z.; Andrae, P.; Rensing, B.

    2000-01-01

    In a postulated core meltdown accident in a light water reactor there are bound to be interactions, in the ex-vessel phase, among the core melt and the structural materials within and below the reactor cavity. In existing plants, these structural materials normally are structural concrete, while future, evolutionary reactor lines are to have sacrificial and protective materials specially designed for this hypothetical case. To add to the state of knowledge about the phenomena occurring, experiments need to be conducted under conditions as realistic as possible. Within the research programs funded by the European Union, the German Federal Ministry for Economics, and the German nuclear power plant operators, experiments on a laboratory as well as an industrial scale on these problems are being carried out in the two projects called CORESA (COrium on REfractory and SAcrificial materials) and ECOSTAR (Ex-vessel COre melt STAbilization Research). The experiments are accompanied by an extensive analytical theoretical program also serving to advance and validate computer codes on the problems under investigation. The projects, which are carried out with international European participation, are expected to allow a concept to be developed for managing postulated accident scenarios involving core meltdown for innovative nuclear power plants, and to provide findings on risk evaluation of plants now in operation so as to further develop accident management measures. (orig.) [de

  19. Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhujie [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Bartels, David [Univ. of Notre Dame, IN (United States)

    2015-04-02

    This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that the effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.

  20. The influence of core material on transient thermal impedances in transformers

    International Nuclear Information System (INIS)

    Górecki, K; Górski, K

    2016-01-01

    In the paper the results of measurements of thermal parameters of impulse-transformers containing cores made of different ferromagnetic materials are presented. Investigations were performed with the use of methods worked out in Gdynia Maritime University. The obtained results of measurements prove that the material of the core does not influence transient thermal impedance of the winding, whereas this parameter visibly changes with the change of spatial orientation of the transformer. In turn, the material of the core decides about transient thermal impedance of the core. Additionally, the influence of the core material on temperature distribution on the surface of the transformer was analysed. (paper)

  1. Comparison of the fractional power motor with cores made of various magnetic materials

    Science.gov (United States)

    Gmyrek, Zbigniew; Lefik, Marcin; Cavagnino, Andrea; Ferraris, Luca

    2017-12-01

    The optimization of the motor cores, coupled with new core shapes as well as powering the motor at high frequency are the primary reasons for the use of new materials. The utilization of new materials, like SMC (soft magnetic composite), reduce the core loss and/or provide quasi-isotropic core's properties in any magnetization direction. Moreover, the use of SMC materials allows for avoiding degradation of the material portions, resulting from punching process, thereby preventing the deterioration of operating parameters of the motor. The authors examine the impact of technological parameters on the properties of a new type of SMC material and analyze the possibility of its use as the core of the fractional power motor. The result of the work is an indication of the shape of the rotor core made of a new SMC material to achieve operational parameters similar to those that have a motor with a core made of laminations.

  2. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

  3. Behaviour of LWR core materials under accident conditions. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1996-12-01

    At the invitation of the Government of the Russian Federation, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA convened a Technical Committee Meeting on Behaviour of LWR Core Materials Under Accident Conditions from 9 to 13 October 1995 in Dimitrovgrad to analyze and evaluate the behaviour of LWR core materials under accident conditions with special emphasis on severe accidents. In-vessel severe accidents phenomena were considered in detail, but specialized thermal hydraulic aspects as well as ex-vessel phenomena were outside the scope of the meeting. Forty participants representing eight countries attended the meeting. Twenty-three papers were presented and discussed during five sessions. Refs, figs, tabs

  4. Evaluation of thermal margin for HANARO core

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Chae, Hee Taek; Kim Heon Il; Lim, I. C.; Lee, C. S.; Kim, H

    1999-08-01

    During the commissioning and the start-up of the HANARO, various design parameters were confirmed and measured. For safer operation of HANARO and resolution of the CHF penalty issue which is one of unresolved licensing problems, thermal margins for normal and transient conditions were re-evaluated reflecting the commissioning and the start-up test results and the design modifications during operation. The re-evaluation shows that the HANARO meets the design criteria for ONB margin and fuel centerline temperature under normal condition. For upset condition, it also satisfies the safety limits for CHFR and fuel centerline temperature. (Author). 11 refs., 13 tabs., 4 figs.

  5. Comparison of the fractional power motor with cores made of various magnetic materials

    Directory of Open Access Journals (Sweden)

    Gmyrek Zbigniew

    2017-12-01

    Full Text Available The optimization of the motor cores, coupled with new core shapes as well as powering the motor at high frequency are the primary reasons for the use of new materials. The utilization of new materials, like SMC (soft magnetic composite, reduce the core loss and/or provide quasi-isotropic core’s properties in any magnetization direction. Moreover, the use of SMC materials allows for avoiding degradation of the material portions, resulting from punching process, thereby preventing the deterioration of operating parameters of the motor. The authors examine the impact of technological parameters on the properties of a new type of SMC material and analyze the possibility of its use as the core of the fractional power motor. The result of the work is an indication of the shape of the rotor core made of a new SMC material to achieve operational parameters similar to those that have a motor with a core made of laminations.

  6. Effect of different composite core materials on fracture resistance of endodontically treated teeth restored with FRC posts.

    Science.gov (United States)

    Panitiwat, Prapaporn; Salimee, Prarom

    2017-01-01

    This study evaluated the fracture resistance of endodontically treated teeth restored with fiber reinforced composite posts, using three resin composite core build-up materials, (Clearfil Photo Core (CPC), MultiCore Flow (MCF), and LuxaCore Z-Dual (LCZ)), and a nanohybrid composite, (Tetric N-Ceram (TNC)). Forty endodontically treated lower first premolars were restored with quartz fiber posts (D.T. Light-Post) cemented with resin cement (Panavia F2.0). Samples were randomly divided into four groups (n=10). Each group was built-up with one of the four core materials following its manufacturers' instructions. The teeth were embedded in acrylic resin blocks. Nickel-Chromium crowns were fixed on the specimens with resin cement. The fracture resistance was determined using a universal testing machine with a crosshead speed of 1 mm/min at 1350 to the tooth axis until failure occurred. All core materials used in the study were subjected to test for the flexural modulus according to ISO 4049:2009. One-way ANOVA and Bonferroni multiple comparisons test indicated that the fracture resistance was higher in the groups with CPC and MCF, which presented no statistically significant difference (p>0.05), but was significantly higher than in those with LCZ and TNC (paligned with the same tendency of fracture loads. Among the cores used in this study, the composite core with high filler content tended to enhance fracture thresholds of teeth restored with fiber posts more than others.

  7. Analytical methods to characterize heterogeneous raw material for thermal spray process: cored wire Inconel 625

    Science.gov (United States)

    Lindner, T.; Bonebeau, S.; Drehmann, R.; Grund, T.; Pawlowski, L.; Lampke, T.

    2016-03-01

    In wire arc spraying, the raw material needs to exhibit sufficient formability and ductility in order to be processed. By using an electrically conductive, metallic sheath, it is also possible to handle non-conductive and/or brittle materials such as ceramics. In comparison to massive wire, a cored wire has a heterogeneous material distribution. Due to this fact and the complex thermodynamic processes during wire arc spraying, it is very difficult to predict the resulting chemical composition in the coating with sufficient accuracy. An Inconel 625 cored wire was used to investigate this issue. In a comparative study, the analytical results of the raw material were compared to arc sprayed coatings and droplets, which were remelted in an arc furnace under argon atmosphere. Energy-dispersive X-ray spectroscopy (EDX) and X-ray fluorescence (XRF) analysis were used to determine the chemical composition. The phase determination was performed by X-ray diffraction (XRD). The results were related to the manufacturer specifications and evaluated in respect to differences in the chemical composition. The comparison between the feedstock powder, the remelted droplets and the thermally sprayed coatings allows to evaluate the influence of the processing methods on the resulting chemical and phase composition.

  8. Materials considerations for UF6 gas-core reactor. Interim report for preliminary design study

    International Nuclear Information System (INIS)

    Wagner, P.

    1977-04-01

    The limiting materials problem in a high-temperature UF 6 core reactor is the corrosion of the core containment vessel. The UF 6 , the lower fluorides of uranium, and the fluorine that exist at the anticipated reactor operating conditions (1000 K and about one atmosphere UF 6 ) are all corrosive. Because of this, the materials evaluation effort for this reactor design study has concentrated on the identification of a viable system for the containment vessel that meets both the materials and neutronic requirements. A study of the literature has revealed that the most promising corrosion-resistant candidates are Ni or Ni-Al alloys. One of the conclusions of this work is that the containment vessel use a nickel liner or clad since the use of Ni as a structural member is precluded by its relative blackness to thermal neutrons. Estimates of corrosion rates of Ni and Ni-Al alloys, the effects of the pressure and temperature of F 2 on the corrosion rates, calculated equilibrium gas compositions at reactor core operating conditions, suggested methods of fabrication, and recommendations for future research and development are included

  9. MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

    International Nuclear Information System (INIS)

    Relloso, J.M.

    1990-01-01

    This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author) [es

  10. Core and Valence Structures in K beta X-ray Emission Spectra of Chromium Materials

    NARCIS (Netherlands)

    Torres Deluigi, Maria; de Groot, Frank M. F.; Lopez-Diaz, Gaston; Tirao, German; Stutz, Guillermo; Riveros de la Vega, Jose

    2014-01-01

    We analyze the core and valence transitions in chromium in a series of materials with a number of different ligands and including the oxidation states: Cr-II, Cr-III, Cr-IV, and Cr-VI. To study the core-to-core transitions we employ the CTM4XAS program and investigate the shapes, widths,

  11. DWPF MATERIALS EVALUATION SUMMARY REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Gee, T.; Chandler, G.; Daugherty, W.; Imrich, K.; Jankins, C.

    1996-09-12

    To better ensure the reliability of the Defense Waste Processing Facility (DWPF) remote canyon process equipment, a materials evaluation program was performed as part of the overall startup test program. Specific test programs included FA-04 ('Process Vessels Erosion/Corrosion Studies') and FA-05 (melter inspection). At the conclusion of field testing, Test Results Reports were issued to cover the various test phases. While these reports completed the startup test requirements, DWPF-Engineering agreed to compile a more detailed report which would include essentially all of the materials testing programs performed at DWPF. The scope of the materials evaouation programs included selected equipment from the Salt Process Cell (SPC), Chemical Process Cell (CPC), Melt Cell, Canister Decon Cell (CDC), and supporting facilities. The program consisted of performing pre-service baseline inspections (work completed in 1992) and follow-up inspections after completion of the DWPF cold chemical runs. Process equipment inspected included: process vessels, pumps, agitators, coils, jumpers, and melter top head components. Various NDE (non-destructive examination) techniques were used during the inspection program, including: ultrasonic testing (UT), visual (direct or video probe), radiography, penetrant testing (PT), and dimensional analyses. Finally, coupon racks were placed in selected tanks in 1992 for subsequent removal and corrosion evaluation after chemical runs.

  12. Evaluating the core microbiota in complex communities: A systematic investigation.

    Science.gov (United States)

    Astudillo-García, Carmen; Bell, James J; Webster, Nicole S; Glasl, Bettina; Jompa, Jamaluddin; Montoya, Jose M; Taylor, Michael W

    2017-04-01

    The study of complex microbial communities poses unique conceptual and analytical challenges, with microbial species potentially numbering in the thousands. With transient or allochthonous microorganisms often adding to this complexity, a 'core' microbiota approach, focusing only on the stable and permanent members of the community, is becoming increasingly popular. Given the various ways of defining a core microbiota, it is prudent to examine whether the definition of the core impacts upon the results obtained. Here we used complex marine sponge microbiotas and undertook a systematic evaluation of the degree to which different factors used to define the core influenced the conclusions. Significant differences in alpha- and beta-diversity were detected using some but not all core definitions. However, findings related to host specificity and environmental quality were largely insensitive to major changes in the core microbiota definition. Furthermore, none of the applied definitions altered our perception of the ecological networks summarising interactions among bacteria within the sponges. These results suggest that, while care should still be taken in interpretation, the core microbiota approach is surprisingly robust, at least for comparing microbiotas of closely related samples. © 2017 Society for Applied Microbiology and John Wiley & Sons Ltd.

  13. Evaluation of spent fuel properties from a conceptual PEACER core

    International Nuclear Information System (INIS)

    Lim, Jae Yong; Kim, Myung Hyun; Kim, Chang Hyo; Hwang, Il Soon

    2003-01-01

    In this paper, a new conceptual core design, PEACER was evaluated in aspect of core performance and spent fuel properties. The core shape is like a pancake to increase axial neutron leakage. Square lattice array was applied which was suitable to decrease the flow speed of Pb-Bi coolant. Although over 30% TRU produced by pyroprocessing was loaded in U-Zr metal fuel, the cycle length of 1 year was achieved and the relative assembly power peaking was less than 1.3. In order to confirm nuclear performance of PEACER core design, several performance indices were adopted and developed. Simple indices such as FIR and FG were used to evaluate fissile breeding. BCM, TG, SNS, and OR calculated by plutonium composition vectors were chosen to distinguish the competency of proliferation resistance. For the estimation of transmutation capability, D-value and extended effective fission half-life time(T EX ) were used. According to these indices, the PEACER core had the better performance compared with other conventional reactor cores although fissile breeding was not acquired

  14. Evaluation of Treatments to Reduce Hardness of Agave americana Core

    Directory of Open Access Journals (Sweden)

    José A. Ramírez

    2006-01-01

    Full Text Available Agave americana contains inulin as storage carbohydrate. Therefore, agave is interesting to be used for the extraction of inulin by pressing. The yield of the process is low due to the high hardness of the core. The objective of this work was to evaluate pretreatments to reduce the hardness in the process of obtaining inulin by pressing. Treatments with water, sulphuric acid 1 % (by mass or sodium hydroxide 1 % (by mass were tested and optimized. The pretreatment of the core of A. americana with sulphuric acid 1 % allowed the reduction of hardness from 30 000 g to 2000 g of breaking force. The mathematical model obtained predicts an optimum processing at 84 °C during 75 min. The treatment with sulphuric acid 1 % also allows white core of A. americana to be obtained, while the other treatments provide yellow core. These results open a good alternative to obtain value added products from this resource.

  15. Mixing core material into the envelopes of red grants

    International Nuclear Information System (INIS)

    Deupree, R.G.

    1986-01-01

    A discussion is presented of calculations of four core helium flashes in red giant stars. The starting point for these calculations is a point source explosion on the polar axis of a two-dimensional finite difference grid. The amount of residue of the core helium flash mixed into and above the hydrogen shell is calculated at four temperatures for the elements carbon, oxygen, neon, magnesium, silicon, and sulfur. 7 refs., 1 tab

  16. Core Self-Evaluations, life satisfaction, and sport satisfaction

    OpenAIRE

    Antón Aluja

    2014-01-01

    We investigated the association between Core Self-Evaluations (CSE) and life and sport satisfaction to assess whether the Core Self-Evaluations scale was a better predictor of life satisfaction or sport satisfaction. The study included three hundred and thirteen athletes (231 men and 82 women; age range to 47 years (Mage=22.9 years, SDage=5.9 years)). Participants completed the French language version of the CSE scale, the Satisfaction with Life Scale, and the Satisfaction with Sport Scale. A...

  17. Measuring technique of super high temperature thermal properties of reactor core materials

    International Nuclear Information System (INIS)

    Ono, Akira; Baba, Tetsuya; Watanabe, Hideo; Matsumoto, Tsuyoshi

    1998-01-01

    In this study, thermal properties of reactor core materials used for water cooled reactors and FBR were tried to develop a technique to measure their melt states at less than 3,000degC in order to contribute more correct evaluation of the reactor core behavior at severe accident. Then, a thermal property measuring method of high temperature melt by using floating method was investigated and its fundamental design was begun to investigate under a base of optimum judgement on the air flow floating throw-down method. And, in order to measure emissivity of melt specimen surface essential for correct temperature measurement using the throw down method, a spectroscopic emissivity measuring unit using an ellipsometer was prepared and induced. On the thermal properties measurement using the holding method, a specimen container to measure thermal diffusiveness of the high temperature melts by using laser flashing method was tried to prepare. (G.K.)

  18. Development and psychometric evaluation of the Core Nurse Resource Scale.

    Science.gov (United States)

    Simpson, Michelle R

    2010-11-01

    To examine the factor structure, internal consistency reliability and concurrent-related validity of the Core Nurse Resource Scale. A cross-sectional survey study design was used to obtain a sample of 149 nurses and nursing staff [Registered Nurse (RNs), Licensed Practical Nurse (LPNs) and Certified Nursing Assistant (CNAs)] working in long-term care facilities. Exploratory factor analysis, Cronbach's alpha and bivariate correlations were used to evaluate validity and reliability. Exploratory factor analysis yielded a scale with 18 items on three factors, accounting for 52% of the variance in scores. Internal consistency reliability for the composite and Core Nurse Resource Scale factors ranged from 0.79 to 0.91. The Core Nurse Resource Scale composite scale and subscales correlated positively with a measure of work engagement (r=0.247-0.572). The initial psychometric evaluation of the Core Nurse Resource Scale demonstrates it is a sound measure. Further validity and reliability assessment will need to be explored and assessed among nurses and other nursing staff working in other practice settings. The intent of the Core Nurse Resource Scale is to evaluate the presence of physical, psychological and social resources of the nursing work environment, to identify workplaces at risk for disengaged (low work engagement) nursing staff and to provide useful diagnostic information to healthcare administrators interested in interventions to improve the nursing work environment. © 2010 The Author. Journal compilation © 2010 Blackwell Publishing Ltd.

  19. Material sampling for rotor evaluation

    International Nuclear Information System (INIS)

    Mercaldi, D.; Parker, J.

    1990-01-01

    Decisions regarding continued operation of aging rotating machinery must often be made without adequate knowledge of rotor material conditions. Physical specimens of the material are not generally available due to lack of an appropriate sampling technique or the high cost and inconvenience of obtaining such samples. This is despite the fact that examination of such samples may be critical to effectively assess the degradation of mechanical properties of the components in service or to permit detailed examination of microstructure and surface flaws. Such information permits a reduction in the uncertainty of remaining life estimates for turbine rotors to avoid unnecessarily premature and costly rotor retirement decisions. This paper describes the operation and use of a recently developed material sampling device which machines and recovers an undeformed specimen from the surface of rotor bores or other components for metallurgical analysis. The removal of the thin, wafer-like sample has a negligible effect on the structural integrity of these components, due to the geometry and smooth surface finish of the resulting shallow depression. Samples measuring approximately 0.03 to 0.1 inches (0.76 to 2.5 mm) thick by 0.5 to 1.0 inch (1.3 to 2.5 cm) in diameter can be removed without mechanical deformation or thermal degradation of the sample or the remaining component material. The device is operated remotely from a control console and can be used externally or internally on any surface for which there is at least a three inch (7.6 cm) working clearance. Application of the device in two case studies of turbine-generator evaluations are presented

  20. A volatile-rich Earth's core inferred from melting temperature of core materials

    Science.gov (United States)

    Morard, G.; Andrault, D.; Antonangeli, D.; Nakajima, Y.; Auzende, A. L.; Boulard, E.; Clark, A. N.; Lord, O. T.; Cervera, S.; Siebert, J.; Garbarino, G.; Svitlyk, V.; Mezouar, M.

    2016-12-01

    Planetary cores are mainly constituted of iron and nickel, alloyed with lighter elements (Si, O, C, S or H). Understanding how these elements affect the physical and chemical properties of solid and liquid iron provides stringent constraints on the composition of the Earth's core. In particular, melting curves of iron alloys are key parameter to establish the temperature profile in the Earth's core, and to asses the potential occurrence of partial melting at the Core-Mantle Boundary. Core formation models based on metal-silicate equilibration suggest that Si and O are the major light element components1-4, while the abundance of other elements such as S, C and H is constrained by arguments based on their volatility during planetary accretion5,6. Each compositional model implies a specific thermal state for the core, due to the different effect that light elements have on the melting behaviour of Fe. We recently measured melting temperatures in Fe-C and Fe-O systems at high pressures, which complete the data sets available both for pure Fe7 and other binary alloys8. Compositional models with an O- and Si-rich outer core are suggested to be compatible with seismological constraints on density and sound velocity9. However, their crystallization temperatures of 3650-4050 K at the CMB pressure of 136 GPa are very close to, if not higher than the melting temperature of the silicate mantle and yet mantle melting above the CMB is not a ubiquitous feature. This observation requires significant amounts of volatile elements (S, C or H) in the outer core to further reduce the crystallisation temperature of the core alloy below that of the lower mantle. References 1. Wood, B. J., et al Nature 441, 825-833 (2006). 2. Siebert, J., et al Science 339, 1194-7 (2013). 3. Corgne, A., et al Earth Planet. Sc. Lett. 288, 108-114 (2009). 4. Fischer, R. a. et al. Geochim. Cosmochim. Acta 167, 177-194 (2015). 5. Dreibus, G. & Palme, H. Geochim. Cosmochim. Acta 60, 1125-1130 (1995). 6. Mc

  1. Career Commitment: Interplay of Core-Self-Evaluation and Reward ...

    African Journals Online (AJOL)

    This study investigated the joint influence of core self-evaluation (self efficacy, self-esteem, work locus of control and neuroticism) and reward system on carrer commitment of local government employees in Oyo State, Nigeria. The study adopted descriptive survey research design. A total of 2,040 respondents were selected ...

  2. Core Self-Evaluation and Goal Orientation: Understanding Work Stress

    Science.gov (United States)

    Morris, Michael Lane; Messal, Carrie B.; Meriac, John P.

    2013-01-01

    This study investigates the dispositional factors related to work stress. Specifically, previous research has demonstrated a relationship between core self-evaluation (CSE) and general life stress. This article extends past research by examining the relationship between CSE and work stress, and includes goal orientation as a potential mediator of…

  3. Longitudinal Relationships between Core Self-Evaluations and Job Satisfaction

    Science.gov (United States)

    Wu, Chia-Huei; Griffin, Mark A.

    2012-01-01

    Core self-evaluations (CSE) have been proposed as a static personality trait that influences individuals' work experiences. However, CSE can also be influenced by work experiences. Based on the corresponsive principle of personality development, this study incorporated both dispositional and contextual perspectives to examine longitudinal…

  4. The separation and distribution of some Luna 24 core materials

    International Nuclear Information System (INIS)

    Pillinger, C.T.; Fabian, D.M.

    1980-01-01

    Three soil samples from different horizons of a core recovered from the Moon's surface by the Soviet Space mission Luna 24 have been separated according to size, visual appearance, density and magnetic properties. Appropriate samples have been distributed to a number of British laboratories for detailed investigations. (author)

  5. Effect of different composite core materials on fracture resistance of endodontically treated teeth restored with FRC posts

    Directory of Open Access Journals (Sweden)

    Prapaporn PANITIWAT

    Full Text Available Abstract Objective This study evaluated the fracture resistance of endodontically treated teeth restored with fiber reinforced composite posts, using three resin composite core build-up materials, (Clearfil Photo Core (CPC, MultiCore Flow (MCF, and LuxaCore Z-Dual (LCZ, and a nanohybrid composite, (Tetric N-Ceram (TNC. Material and Methods Forty endodontically treated lower first premolars were restored with quartz fiber posts (D.T. Light-Post cemented with resin cement (Panavia F2.0. Samples were randomly divided into four groups (n=10. Each group was built-up with one of the four core materials following its manufacturers’ instructions. The teeth were embedded in acrylic resin blocks. Nickel-Chromium crowns were fixed on the specimens with resin cement. The fracture resistance was determined using a universal testing machine with a crosshead speed of 1 mm/min at 1350 to the tooth axis until failure occurred. All core materials used in the study were subjected to test for the flexural modulus according to ISO 4049:2009. Results One-way ANOVA and Bonferroni multiple comparisons test indicated that the fracture resistance was higher in the groups with CPC and MCF, which presented no statistically significant difference (p>0.05, but was significantly higher than in those with LCZ and TNC (p<0.05. In terms of the flexural modulus, the ranking from the highest values of the materials was aligned with the same tendency of fracture loads. Conclusion Among the cores used in this study, the composite core with high filler content tended to enhance fracture thresholds of teeth restored with fiber posts more than others.

  6. Reactor Structure Materials: Corrosion of Reactor Core Internals

    International Nuclear Information System (INIS)

    Van Dyck, S.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on the corrosion of reactor core internals are: (1) to gain mechanistic insight into the Irradition Assisted Stress Corrosion Cracking (IASCC) phenomenon by studying the influence of separate parameters in well controlled experiments; (2) to develop and validate a predictive capability on IASCC by model description and (3) to define and validate countermeasures and monitoring techniques for application in reactors. Progress and achievements in 1999 are described

  7. Feasibility study of thermal insulation materials for core support of experimental VHTR

    International Nuclear Information System (INIS)

    Kawakami, H.; Nakanishi, T.

    1982-01-01

    Thermal insulation materials for core support of the experimental VHTR, planned by JAERI, should maintain moderate compressive strength and dimensional stability as well as low thermal conductivity at the maximum service temperature of 1100 0 C for 20 years. For selecting materials, we investigate properties of some candidates, and evaluate their feasibility. Preliminary tests, heat treatment test and compressive creep tests for 1000 hours at 900 0 C and 1000 0 C were conducted. In the preliminary tests, EG-38B (carbon baked at 1350 0 C) and Fine Finnex 600 (silicon nitride) showed acceptable physical stability. In the heat treatment tests, silicon nitride showed weight loss probably caused by thermal decomposition. Compressive creep deformation of Fine Finnex 600 was negligible under stress of 100 kg/cm 2 for 1000 hours. Heat treatment at 1200 to 1300 0 C for 50 hours improved dimensional stability of carbon at 1000 0 C

  8. Effect of curing mode on the hardness of dual-cured composite resin core build-up materials

    Directory of Open Access Journals (Sweden)

    César Augusto Galvão Arrais

    2010-06-01

    Full Text Available This study evaluated the Knoop Hardness (KHN values of two dual-cured composite resin core build-up materials and one resin cement exposed to different curing conditions. Two dual-cured core build-up composite resins (LuxaCore®-Dual, DMG; and FluoroCore®2, Dentsply Caulk, and one dual-cured resin cement (Rely X ARC, 3M ESPE were used in the present study. The composite materials were placed into a cylindrical matrix (2 mm in height and 3 mm in diameter, and the specimens thus produced were either light-activated for 40 s (Optilux 501, Demetron Kerr or were allowed to self-cure for 10 min in the dark (n = 5. All specimens were then stored in humidity at 37°C for 24 h in the dark and were subjected to KHN analysis. The results were submitted to 2-way ANOVA and Tukey's post-hoc test at a pre-set alpha of 5%. All the light-activated groups exhibited higher KHN values than the self-cured ones (p = 0.00001, regardless of product. Among the self-cured groups, both composite resin core build-up materials showed higher KHN values than the dual-cured resin cement (p = 0.00001. LuxaCore®-Dual exhibited higher KHN values than FluoroCore®2 (p = 0.00001 when they were allowed to self-cure, while no significant differences in KHN values were observed among the light-activated products. The results suggest that dual-cured composite resin core build-up materials may be more reliable than dual-cured resin cements when curing light is not available.

  9. Materials problems related to the core catcher of sodium cooled reactors

    International Nuclear Information System (INIS)

    Goetzmann, O.

    1975-05-01

    There are in principal two possible solutions for the external core catcher as far as materials are concerned. 1) A barrier consisting of a material with a high melting point, 2) a tray of comparatively low melting material with a high solubility for the fuel. In case of the first concept one has to look for materials whose melting temperatures are above the temperature of the molten core. Based on metallurgical reasons it seems very likely that the molten core does not exceed a temperature in the range between 2,500 and 2,800 0 C. Due to the compatibility situation with the molten core only a few high melting oxides will be suitable as liner materials for a core catcher. In the second case basalt or concrete, if free of water and lime, are suitable materials. Graphite is a high melting material, however, due to its behaviour with the molten core it should be listed under the second group. By the reaction of graphite with the core materials the melt can be kept liquid down to temperatures of around 1,100 0 C. The evolution of CO by this reaction should be supportable. It is an endothermal reaction. Experiments on the behaviour of core catcher materials have shown that sodium is capable of penetrating into sintered bodies of UO 2 with densities of 90% TD at temperatures higher than 200 0 C. This may lead to the desintegration of these bodies. The exposure to moist air has not done much harm to UO 2 pellets of densities from 80 to 90% TD. Even after one year of exposure, swelling or desintegration could not be observed. Sodium is also capable of penetrating into bodies of synthetic carbon and graphite. Only well graphitized material will not be destroyed. (orig.) [de

  10. Relating the structural strength of concrete sewer pipes and material properties retrieved from core samples

    NARCIS (Netherlands)

    Stanic, N.; Langeveld, J.G.; Salet, Theo; Clemens, F.H.L.R.

    2016-01-01

    Drill core samples are taken in practice for an analysis of the material characteristics of concrete pipes in order to improve the quality of the decision-making on rehabilitation actions. Earlier research has demonstrated that core sampling is associated with a significant uncertainty. In this

  11. Development of material balance evaluation technique(2)

    International Nuclear Information System (INIS)

    Lee, Byung Doo

    2000-06-01

    IAEA considers that the evaluation on material balance is one of the important activities for detecting the diversion of nuclear materials as well as measurement uncertainties and measurement bias. Nuclear material accounting reports, the results of DA and NDA, the summarized lists of material stratified by inspector are necessary for the material balance evaluation. In this report, the concepts and evaluation methods of material balance evaluation such as the estimation techniques of random and systematic errors, MUF, D and MUF-D are described. As a conclusion, it is possible for national inspection to evaluate the material balance by applying the evaluation methods of the IAEA such as error estimation using operator-inspector paired data, inspector MUF(IMUF) evaluation

  12. The Relative Impact of Aligning Tier 2 Intervention Materials with Classroom Core Reading Materials in Grades K-2

    Science.gov (United States)

    Foorman, Barbara R.; Herrera, Sarah; Dombek, Jennifer

    2018-01-01

    This randomized controlled trial in 55 low-performing schools across Florida compared 2 early literacy interventions--1 using stand-alone materials and 1 using materials embedded in the existing core reading/language arts program. A total of 3,447 students who were below the 30th percentile in vocabulary and reading-related skills participated in…

  13. Effect of Three Different Core Materials on Masking Ability of a Zirconia Ceramic

    Directory of Open Access Journals (Sweden)

    Farhad Tabatabaian

    2016-12-01

    Full Text Available Objectives: Masking ability of a restorative material plays a role in hiding colored substructures; however, the masking ability of zirconia ceramic (ZRC has not yet been clearly understood in zirconia-based restorations. This study evaluated the effect of three different core materials on masking ability of a ZRC.Materials and Methods: Ten zirconia disc samples, 0.5mm in thickness and 10mm in diameter, were fabricated. A white (W substrate (control and three substrates of nickel-chromium alloy (NCA, non-precious gold alloy (NPGA, and ZRC were prepared. The zirconia discs were placed on the four types of substrates for spectrophotometry. The L*, a*, and b* values of the specimens were measured by a spectrophotometer and color change (ΔE values were calculated to determine color differences between the test and control groups and were then compared with the perceptual threshold. Randomized block ANOVA and Bonferroni test analyzed the data. A significance level of 0.05 was considered.Results: The mean and standard deviation values of ΔE for NCA, NPGA, and ZRC groups were 10.26±2.43, 9.45±1.74, and 6.70±1.91 units, respectively. Significant differences were found in the ΔE values between ZRC and the other two experimental groups (NCA and NPGA; P<0.0001 and P=0.001, respectively. The ΔE values for the groups were more than the predetermined perceptual threshold.Conclusions: Within the limitations of this study, it was concluded that the tested ZRC could not well mask the examined core materials.Keywords: Color; Spectrophotometry; Visual Perception; Yttria Stabilized Tetragonal Zirconia

  14. PGSFR Core Thermal Design Procedure to Evaluate the Safety Margin

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Kim, Sang-Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The Korea Atomic Energy Research Institute (KAERI) has performed a SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal design is to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damage in SFR subassemblies arises from a creep induced failure. The creep limit is evaluated based on the maximum cladding temperature, power, neutron flux, and uncertainties in the design parameters, as shown in Fig. 1. In this work, the core thermal design procedures are compared to verify the present PGSFR methodology based on the nuclear plant design criteria/guidelines and previous SFR thermal design methods. The PGSFR core thermal design procedure is verified based on the nuclear plant design criteria/guidelines and previous methods in LWRs and SFRs. The present method aims to directly evaluate the fuel cladding failure and to assure more safety margin. The 2 uncertainty is similar to 95% one-side tolerance limit of 1.96 in LWRs. The HCFs, ITDP, and MCM reveal similar uncertainty propagation for cladding midwall temperature for typical SFR conditions. The present HCFs are mainly employed from the CRBR except the fuel-related uncertainty such as an incorrect fuel distribution. Preliminary PGSFR specific HCFs will be developed by the end of 2015.

  15. Status of core material development for fast reactor in Japan

    International Nuclear Information System (INIS)

    Ukai, S.; Shibahara, I.; Nagai, S.

    1994-01-01

    In the last two decades, extensive efforts have been devoted to the development of mixed-oxide fuel for LMFBR in Japan. For the fuel of the prototype reactor MONJU, drastic improvement in creep rupture strength and swelling resistance were attained by modification within the compositional specification of the standard Type 316 stainless steel (PNC316). For the fuel of future large-scale reactors, extensive research and development program are under way to realize the long life fuel. The candidate material for demonstration reactor is advanced austenitic stainless steel (PNC1520) which intended to modify the composition beyond the Type 316 stainless steel specification. In order to further improve the swelling resistance, the austenitic stainless steel with higher nickel content (High Ni alloy) and ferritic/martensitic steel (PNC-FMS) are developed. In a prospective cladding material for the long life fuel, the development of oxide dispersion strengthened (ODS) ferritic steel is focused to establish the alloying design and fabrication process toward as high as 250dpa. (author)

  16. Comparison of Magnetic Characteristics of Powder Magnetic Core and Evaluation of Motor Characteristics

    Science.gov (United States)

    Enomoto, Yuji; Ito, Motoya; Masaki, Ryozo; Yamazaki, Katsuyuki; Asaka, Kazuo; Ishihara, Chio; Ohiwa, Syoji

    A magnetic characteristic measurement, a motor characteristic forecast, and an experimental evaluation of various powder magnetic cores were performed aiming at a fixed quantity grasp when the powder magnetic core was applied to the motor core as the magnetic material. The manufacturing conditions were changed, and magnetic characteristic compares a direct current magnetization characteristic and an iron disadvantageous characteristic with the silicon steel board for a different powder magnetic core. Therefore, though some permeabilities are low, characteristics almost equal to those of a silicon steel board were obtained in the maximum saturation magnetic induction, which confirms that the powder magnetic core in disadvantageous iron in a certain frequency domain, and to confirm disadvantageous iron lowers. Moreover, it has been shown to obtain characteristics almost equal to the silicon steel board when compared in terms of motor efficiency, though some disadvantageous iron increases since the effect when applying to the motor is verified the silicon steel board and the comparison evaluation for the surface type permanent magnet motor.

  17. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  18. Design evaluation of emergency core cooling systems using Axiomatic Design

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Gyunyoung [Massachusetts Institute of Technology, Department of Mechanical Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)]. E-mail: gheo@mit.edu; Lee, Song Kyu [Korea Advanced Institute of Science and Technology, Department of Nuclear and Quantum Engineering, 373-1 Guseong-dong, Yuseong-gu, Daejeon (Korea, Republic of)

    2007-01-15

    In designing nuclear power plants (NPPs), the evaluation of safety is one of the important issues. As a measure for evaluating safety, this paper proposes a methodology to examine the design process of emergency core cooling systems (ECCSs) in NPPs using Axiomatic Design (AD). This is particularly important for identifying vulnerabilities and creating solutions. Korean Advanced Power Reactor 1400 MWe (APR1400) adopted the ECCS, which was improved to meet the stronger safety regulations than that of the current Optimized Power Reactor 1000 MWe (OPR1000). To improve the performance and safety of the ECCS, the various design strategies such as independency or redundancy were implemented, and their effectiveness was confirmed by calculating core damage frequency. We suggest an alternative viewpoint of evaluating the deployment of design strategies in terms of AD methodology. AD suggests two design principles and the visualization tools for organizing design process. The important benefit of AD is that it is capable of providing suitable priorities for deploying design strategies. The reverse engineering driven by AD has been able to show that the design process of the ECCS of APR1400 was improved in comparison to that of OPR1000 from the viewpoint of the coordination of design strategies.

  19. Experimental study of the mechanical behaviour of pin reinforced foam core sandwich materials under shear load

    International Nuclear Information System (INIS)

    Dimassi, M A; Brauner, C; Herrmann, A S

    2016-01-01

    Sandwich structures with a lightweight closed cell hard foam core have the potential to be used in primary structures of commercial aircrafts. Compared to honeycomb core sandwich, the closed cell foam core sandwich overcomes the issue of moisture take up and makes the manufacturing of low priced and highly integrated structures possible. However, lightweight foam core sandwich materials are prone to failure by localised external loads like low velocity impacts. Invisible cracks could grow in the foam core and threaten the integrity of the structure. In order to enhance the out-of-plane properties of foam core sandwich structures and to improve the damage tolerance (DT) dry fibre bundles are inserted in the foam core. The pins are infused with resin and co-cured with the dry fabric face sheets in an out-of-autoclave process. This study presents the results obtained from shear tests following DIN 53294-standard, on flat sandwich panels. All panels were manufactured with pin-reinforcement manufactured with the Tied Foam Core Technology (TFC) developed by Airbus. The effects of pin material (CFRP and GFRP) and pin volume fraction on the shear properties of the sandwich structure and the crack propagation were investigated and compared to a not pinned reference. It has been concluded that the pin volume fraction has a remarkable effect on the shear properties and damage tolerance of the observed structure. Increasing the pin volume fraction makes the effect of crack redirection more obvious and conserves the integrity of the structure after crack occurrence. (paper)

  20. Evaluation of Residual Stresses using Ring Core Method

    Directory of Open Access Journals (Sweden)

    Holý S.

    2010-06-01

    Full Text Available The method for measuring residual stresses using ring-core method is described. Basic relations are given for residual stress measurement along the specimen depth and simplified method is described for average residual stress estimation in the drilled layer for known principal stress directions. The estimation of calculated coefficients using FEM is described. Comparison of method sensitivity is made with hole-drilling method. The device for method application is described and an example of experiment is introduced. The accuracy of method is discussed. The influence of strain gauge rosette misalignment to the evaluated residual stresses is performed using FEM.

  1. Evaluation of the influence of bypass flow gap distribution on the core hot spot in a prismatic VHTR core

    International Nuclear Information System (INIS)

    Kim, Min-Hwan; Lim, Hong-Sik

    2011-01-01

    Highlights: → A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. → The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. → The predicted gap size is large enough to affect the flow distribution in the core. → The bypass gap and flow distributions are closely related to the local hot spot temperature and its location. → The core restraint mechanism preventing outward movement of graphite block reduces the bypass gap size and hot spot temperature. - Abstract: Core bypass flow in VHTR is one of the key issues for core thermal margins and efficiency. The bypass flow in the prismatic core varies during core cycles due to the irradiation shrinkage/swelling and thermal expansion of the graphite blocks. A procedure to evaluate the local gap size variation between graphite blocks was developed and applied to a prismatic core VHTR. The influence of the core restraint mechanism on the bypass flow gap was evaluated. The predicted gap size is as much as 8 mm when the graphite block is exposed to its allowable limit of fast neutron fluence. The analysis for the core bypass flow and hot spot was carried out based on the calculated gap distributions. The results indicate that the bypass gap and flow distributions are closely related to the local hot spot and its location and the core restraint mechanism preventing outward movement of the graphite block by a fastening device reduces the bypass gap size, which results in the decrease of maximum fuel temperature not less than 100 deg. C, when compared to the case without it.

  2. Performance evaluation of open core gasifier on multi-fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bhoi, P.R.; Singh, R.N.; Sharma, A.M.; Patel, S.R. [Thermo Chemical Conversion Division, Sardar Patel Renewable Energy Research Institute (SPRERI), Vallabh Vidyanagar 388 120, Gujarat (India)

    2006-06-15

    Sardar Patel renewable energy research institute (SPRERI) has designed and developed open core, throat-less, down draft gasifier and installed it at the institute. The gasifier was designed for loose agricultural residues like groundnut shells. The purpose of the study is to evaluate the gasifier on multi-fuels such as babul wood (Prosopis juliflora), groundnut shell briquettes, groundnut shell, mixture of wood (Prosopis juliflora) and groundnut shell in the ratio of 1:1 and cashew nut shell. The gasifier performance was evaluated in terms of fuel consumption rate, calorific value of producer gas and gasification efficiency. Gasification efficiency of babul wood (Prosopis juliflora), groundnut shell briquettes, groundnut shell, mixture of Prosopis juliflora and groundnut shell in the ratio of 1:1 and cashew nut shell were 72%, 66%, 70%, 64%, 70%, respectively. Study revealed that babul wood (Prosopis juliflora), groundnut shell briquettes, groundnut shell, mixture of wood (Prosopis juliflora) and groundnut shell in the ratio of 1:1 and cashew nut shell were satisfactorily gasified in open core down draft gasifier. The study also showed that there was flow problem with groundnut shell. (author)

  3. Evaluation of irradiated coating material specimens

    International Nuclear Information System (INIS)

    Lee, Yong Jin; Nam, Seok Woo; Cho, Lee Moon

    2007-12-01

    Evaluation result of irradiated coating material specimens - Coating material specimens radiated Gamma Energy(Co 60) in air condition. - Evaluation conditions was above 1 X 10 4 Gy/hr, and radiated TID 2.0 X 10 6 Gy. - The radiated coating material specimens, No Checking, Cracking, Flaking, Delamination, Peeling and Blistering. - Coating system at the Kori no. 1 and APR 1400 Nuclear power plant, evaluation of irradiated coating materials is in accordance with owner's requirement(2.0 X 10 6 Gy)

  4. An evaluation of dental operative simulation materials.

    Science.gov (United States)

    He, Li-Hong; Foster Page, Lyndie; Purton, David

    2012-01-01

    The study was to evaluate the performance of different materials used in dental operative simulation and compare them with those of natural teeth. Three typical phantom teeth materials were compared with extracted permanent teeth by a nanoindentation system and evaluated by students and registered dentists on the drilling sensation of the materials. Moreover, the tool life (machinability) of new cylindrical diamond burs on cutting the sample materials was tested and the burs were observed. Although student and dentist evaluations were scattered and inconclusive, it was found that elastic modulus (E) and hardness (H) were not the main factors in determining the drilling sensation of the materials. The sensation of drilling is a reflection of cutting force and power consumption.An ideal material for dental simulation should be able to generate similar drilling resistance to that of natural tooth, which is the machinability of the material.

  5. Irradiation-accelerated corrosion of reactor core materials

    International Nuclear Information System (INIS)

    Bartels, David; Was, Gary; Jiao, Zhijie

    2012-09-01

    The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, but also applies to most all other GenIV concepts. Of these four drivers, the combination of radiation and corrosion presents a unique and extremely challenging environment for materials, for which an understanding of the fundamental science is essentially absent. Irradiation can affect corrosion or oxidation in at least three different ways. Radiation interaction with water results in the decomposition of water into radicals and oxidizing species that will increase the electrochemical corrosion potential and lead to greater corrosion rates. Irradiation of the solid surface can produce excited states that can alter corrosion, such as in the case of photo-induced corrosion. Lastly, displacement damage in the solid will result in a high flux of defects to the solid-solution interface that can alter and perhaps, accelerate interface reactions. While there exists reasonable understanding of how corrosion is affected by irradiation of the aqueous environment, there is little understanding of how irradiation affects corrosion through its impact on the solid, whether metal or oxide. The reason is largely due to the difficulty of conducting experiments that can measure this effect separately. We have undertaken a project specifically to separate the several effects of irradiation on the mechanisms of corrosion. We seek to answer the question: How does radiation damage to the solution-oxide couple affect the oxidation process differently from radiation damage to either component alone? The approach taken in this work is to closely compare corrosion accelerated by (1) proton irradiation, (2) electron irradiation, and (3) chemical corrosion potential effects alone, under typical PWR operating conditions at 300 deg. C. Both 316 stainless steel and zirconium are to be studied. The proton

  6. Core-Shell Structured Electro- and Magneto-Responsive Materials: Fabrication and Characteristics

    Directory of Open Access Journals (Sweden)

    Hyoung Jin Choi

    2014-11-01

    Full Text Available Core-shell structured electrorheological (ER and magnetorheological (MR particles have attracted increasing interest owing to their outstanding field-responsive properties, including morphology, chemical and dispersion stability, and rheological characteristics of shear stress and yield stress. This study covers recent progress in the preparation of core-shell structured materials as well as their critical characteristics and advantages. Broad emphasises from the synthetic strategy of various core-shell particles to their feature behaviours in the magnetic and electric fields have been elaborated.

  7. Building waste management core indicators through Spatial Material Flow Analysis: Net recovery and transport intensity indexes

    Energy Technology Data Exchange (ETDEWEB)

    Font Vivanco, David, E-mail: font@cml.leidenuniv.nl [Institut de Ciencia i Tecnologia Ambientals (ICTA), Departament d' Enginyeria Quimica, Universitat Autonoma de Barcelona (UAB), 08193 Bellaterra, Barcelona (Spain); Institute of Environmental Sciences (CML), Leiden University, P.O. Box 9518, 2300 RA Leiden (Netherlands); Puig Ventosa, Ignasi [ENT Environment and Management, Carrer Sant Joan 39, First Floor, 08800 Vilanova i la Geltru, Barcelona (Spain); Gabarrell Durany, Xavier [Institut de Ciencia i Tecnologia Ambientals (ICTA), Departament d' Enginyeria Quimica, Universitat Autonoma de Barcelona (UAB), 08193 Bellaterra, Barcelona (Spain)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Sustainability and proximity principles have a key role in waste management. Black-Right-Pointing-Pointer Core indicators are needed in order to quantify and evaluate them. Black-Right-Pointing-Pointer A systematic, step-by-step approach is developed in this study for their development. Black-Right-Pointing-Pointer Transport may play a significant role in terms of environmental and economic costs. Black-Right-Pointing-Pointer Policy action is required in order to advance in the consecution of these principles. - Abstract: In this paper, the material and spatial characterization of the flows within a municipal solid waste (MSW) management system are combined through a Network-Based Spatial Material Flow Analysis. Using this information, two core indicators are developed for the bio-waste fraction, the Net Recovery Index (NRI) and the Transport Intensity Index (TII), which are aimed at assessing progress towards policy-related sustainable MSW management strategies and objectives. The NRI approaches the capacity of a MSW management system for converting waste into resources through a systematic metabolic approach, whereas the TII addresses efficiency in terms of the transport requirements to manage a specific waste flow throughout the entire MSW management life cycle. Therefore, both indicators could be useful in assessing key MSW management policy strategies, such as the consecution of higher recycling levels (sustainability principle) or the minimization of transport by locating treatment facilities closer to generation sources (proximity principle). To apply this methodological approach, the bio-waste management system of the region of Catalonia (Spain) has been chosen as a case study. Results show the adequacy of both indicators for identifying those points within the system with higher capacity to compromise its environmental, economic and social performance and therefore establishing clear targets for policy

  8. Building waste management core indicators through Spatial Material Flow Analysis: Net recovery and transport intensity indexes

    International Nuclear Information System (INIS)

    Font Vivanco, David; Puig Ventosa, Ignasi; Gabarrell Durany, Xavier

    2012-01-01

    Highlights: ► Sustainability and proximity principles have a key role in waste management. ► Core indicators are needed in order to quantify and evaluate them. ► A systematic, step-by-step approach is developed in this study for their development. ► Transport may play a significant role in terms of environmental and economic costs. ► Policy action is required in order to advance in the consecution of these principles. - Abstract: In this paper, the material and spatial characterization of the flows within a municipal solid waste (MSW) management system are combined through a Network-Based Spatial Material Flow Analysis. Using this information, two core indicators are developed for the bio-waste fraction, the Net Recovery Index (NRI) and the Transport Intensity Index (TII), which are aimed at assessing progress towards policy-related sustainable MSW management strategies and objectives. The NRI approaches the capacity of a MSW management system for converting waste into resources through a systematic metabolic approach, whereas the TII addresses efficiency in terms of the transport requirements to manage a specific waste flow throughout the entire MSW management life cycle. Therefore, both indicators could be useful in assessing key MSW management policy strategies, such as the consecution of higher recycling levels (sustainability principle) or the minimization of transport by locating treatment facilities closer to generation sources (proximity principle). To apply this methodological approach, the bio-waste management system of the region of Catalonia (Spain) has been chosen as a case study. Results show the adequacy of both indicators for identifying those points within the system with higher capacity to compromise its environmental, economic and social performance and therefore establishing clear targets for policy prioritization. Moreover, this methodological approach permits scenario building, which could be useful in assessing the outcomes of

  9. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    International Nuclear Information System (INIS)

    Bi, G.; Liu, C.; Si, S.

    2012-01-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade 233 U-Thorium (U 3 ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade 233 U extracted from burnt PuThOX fuel was used to fabrication of U 3 ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U 3 ThOX mixed core, the well designed U 3 ThOX FAs with 1.94 w/o fissile uranium (mainly 233 U) were located on the periphery of core as a blanket region. U 3 ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U 3 ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U 3 ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U 3 ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U 3 ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared

  10. Estimation of irradiation-induced material damage measure of FCM fuel in LWR core

    International Nuclear Information System (INIS)

    Lee, Kyung-Hoon; Lee, Chungchan; Park, Sang-Yoon; Cho, Jin-Young; Chang, Jonghwa; Lee, Won Jae

    2014-01-01

    An irradiation-induced material damage measure on tri-isotropic (TRISO) multi-coating layers of fully ceramic micro-encapsulated (FCM) fuel to replace conventional uranium dioxide (UO 2 ) fuel for existing light water reactors (LWRs) has been estimated using a displacement per atom (DPA) cross section for a FCM fuel performance analysis. The DPA cross sections in 47 and 190 energy groups for both silicon carbide (SiC) and graphite are generated based on the molecular dynamics simulation by SRIM/TRIM. For the selected FCM fuel assembly design with FeCrAl cladding, a core depletion analysis was carried out using the DeCART2D/MASTER code system with the prepared DPA cross sections to evaluate the irradiation effect in the Korean OPR-1000. The DPA of the SiC and IPyC coating layers is estimated by comparing the discharge burnup obtained from the MASTER calculation with the burnup-dependent DPA for each coating layer calculated using DeCART2D. The results show that low uranium loading and hardened neutron spectrum compared to that of high temperature gas-cooled reactor (HTGR) result in high discharge burnup and high fast neutron fluence. In conclusion, it can be seen that the irradiation-induced material damage measure is noticeably increased under LWR operating conditions compared to HTGRs. (author)

  11. Evaluation of Terminated Nuclear Material Licenses

    International Nuclear Information System (INIS)

    Spencer, K.M.; Zeighami, E.A.

    1999-01-01

    This report presents the results of a six-year project that reviewed material licenses that had been terminated during the period from inception of licensing until approximately late-1994. The material licenses covered in the review project were Part 30, byproduct material licenses; Part 40, source material licenses; and Part 70, special nuclear material licenses. This report describes the methodology developed for the project, summarizes the findings of the license file inventory process, and describes the findings of the reviews or evaluations of the license files. The evaluation identified nuclear material use sites that need review of the licensing material or more direct follow-up of some type. The review process also identified licenses authorized to possess sealed sources for which there was incomplete or missing documentation of the fate of the sources

  12. Nondestructive evaluation ultrasonic methods for construction materials

    International Nuclear Information System (INIS)

    Chilibon, I.; Zisu, T.; Raetchi, V.

    2002-01-01

    The paper presents some ultrasonic methods for evaluation of physical-mechanical properties of construction materials (bricks, concrete, BCA), such as: pulse method, examination methods, and direct measurement of the propagation velocity and impact-echo method. Utilizing these nondestructive evaluation ultrasonic methods it can be determined the main material parameters and material characteristics (elasticity coefficients, density, propagation velocity, ultrasound attenuation, etc.) of construction materials. These method are suitable for construction materials because the defectoscopy methods for metallic materials cannot be utilized, due to its rugged and non-homogeneous structures and grate attenuation coefficients of ultrasound propagation through materials. Also, the impact-echo method is a technique for flaw detection in concrete based on stress wave propagation. Studies have shown that the impact-echo method is effective for locating voids, honeycombing, delaminating, depth of surface opening cracks, and measuring member thickness

  13. Relative translucency of six all-ceramic systems. Part I: core materials.

    Science.gov (United States)

    Heffernan, Michael J; Aquilino, Steven A; Diaz-Arnold, Ana M; Haselton, Debra R; Stanford, Clark M; Vargas, Marcos A

    2002-07-01

    All-ceramic restorations have been advocated for superior esthetics. Various materials have been used to improve ceramic core strength, but it is unclear whether they affect the opacity of all-ceramic systems. This study compared the translucency of 6 all-ceramic system core materials at clinically appropriate thicknesses. Disc specimens 13 mm in diameter and 0.49 +/- 0.01 mm in thickness were fabricated from the following materials (n = 5 per group): IPS Empress dentin, IPS Empress 2 dentin, In-Ceram Alumina core, In-Ceram Spinell core, In-Ceram Zirconia core, and Procera AllCeram core. Empress and Empress 2 dentin specimens also were fabricated and tested at a thickness of 0.77 +/- 0.02 mm (the manufacturer's recommended core thickness is 0.8 mm). A high-noble metal-ceramic alloy (Porc. 52 SF) served as the control, and Vitadur Alpha opaque dentin was used as a standard. Sample reflectance (ratio of the intensity of reflected light to that of the incident light) was measured with an integrating sphere attached to a spectrophotometer across the visible spectrum (380 to 700 nm); 0-degree illumination and diffuse viewing geometry were used. Contrast ratios were calculated from the luminous reflectance (Y) of the specimens with a black (Yb) and a white (Yw) backing to give Yb/Yw with CIE illuminant D65 and a 2-degree observer function (0.0 = transparent, 1.0 = opaque). One-way analysis of variance and Tukey's multiple-comparison test were used to analyze the data (P In-Ceram Spinell > Empress, Procera, Empress 2 > In-Ceram Alumina > In-Ceram Zirconia, 52 SF alloy.

  14. Nanocrystalline material in toroidal cores for current transformer: analytical study and computational simulations

    Directory of Open Access Journals (Sweden)

    Benedito Antonio Luciano

    2005-12-01

    Full Text Available Based on electrical and magnetic properties, such as saturation magnetization, initial permeability, and coercivity, in this work are presented some considerations about the possibilities of applications of nanocrystalline alloys in toroidal cores for current transformers. It is discussed how the magnetic characteristics of the core material affect the performance of the current transformer. From the magnetic characterization and the computational simulations, using the finite element method (FEM, it has been verified that, at the typical CT operation value of flux density, the nanocrystalline alloys properties reinforce the hypothesis that the use of these materials in measurement CT cores can reduce the ratio and phase errors and can also improve its accuracy class.

  15. EVALUATION OF CAUSES OF CONSTRUCTION MATERIAL WASTE

    African Journals Online (AJOL)

    Osondu

    factors contributing to construction material waste generation on building sites in Rivers State, ... the studied factors at every level of the construction processes and in their waste management plan. ..... Evaluation of Solid Waste in Building.

  16. Behaviour of contact layer material between cermet fuel element core and can

    International Nuclear Information System (INIS)

    Gavrilin, S.S.; Permyakov, L.N.; Simakov, G.A.; Chernikov, A.S.

    1996-01-01

    The structural state of the contact layer between the shell of the Zr1Nb alloy and cermet fuel element core containing up to 70% of uranium dioxides is experimental studied. The silumin alloy was used as contact material. The results of studies on interaction zones, formed on the Zr1Nb - silumin boundary after fuel elements manufacture and also under temperature conditions, modeling the maximum design and hypothetical accidents accompanied by the contact material melting, are presented [ru

  17. Simulant-material experimental investigation of flow dynamics in the CRBR Upper-Core Structure

    International Nuclear Information System (INIS)

    Wilhelm, D.; Starkovich, V.S.; Chapyak, E.J.

    1982-09-01

    The results of a simulant-material experimental investigation of flow dynamics in the Clinch River Breeder Reactor (CRBR) Upper Core Structure are described. The methodology used to design the experimental apparatus and select test conditions is detailed. Numerous comparisons between experimental data and SIMMER-II Code calculations are presented with both advantages and limitations of the SIMMER modeling features identified

  18. Supplemental materials for the ICDP-USGS Eyreville A, B, and C core holes, Chesapeake Bay impact structure: Core-box photographs, coring-run tables, and depth-conversion files

    Science.gov (United States)

    Durand, C.T.; Edwards, L.E.; Malinconico, M.L.; Powars, D.S.

    2009-01-01

    During 2005-2006, the International Continental Scientific Drilling Program and the U.S. Geological Survey drilled three continuous core holes into the Chesapeake Bay impact structure to a total depth of 1766.3 m. A collection of supplemental materials that presents a record of the core recovery and measurement data for the Eyreville cores is available on CD-ROM at the end of this volume and in the GSA Data Repository. The supplemental materials on the CD-ROM include digital photographs of each core box from the three core holes, tables of the three coring-run logs, as recorded on site, and a set of depth-conversion programs. In this chapter, the contents, purposes, and basic applications of the supplemental materials are briefly described. With this information, users can quickly decide if the materials will apply to their specific research needs. ?? 2009 The Geological Society of America.

  19. Research activities at JAERI on core material behaviour under severe accident conditions

    International Nuclear Information System (INIS)

    Uetsuka, H.; Katanashi, S.; Ishijima, K.

    1996-01-01

    At the Japan Atomic Energy Research Institute (JAERI), experimental studies on physical phenomena under the condition of a severe accident have been conducted. This paper presents the progress of the experimental studies on fuel and core materials behaviour such as the thermal shock fracture of fuel cladding due to quenching, the chemical interaction of core materials at high temperatures and the examination of TMI-2 debris. The mechanical behaviour of fuel rod with heavily embrittled cladding tube due to the thermal shock during delayed reflooding have been investigated at the Nuclear Safety Research Reactor (NSSR) of JAERI. A test fuel rod was heated in steam atmosphere by both electric and nuclear heating using the NSSR, then the rod was quenched by reflooding at the test section. Melting of core component materials having relatively low melting points and their eutectic reaction with other materials significantly influence on the degradation and melt down of fuel bundles during severe accidents. Therefore basic information on the reaction of core materials is necessary to understand and analyze the progress of core melting and relocation. Chemical interactions have been widely investigated at high temperatures for various binary systems of core component materials including absorber materials such as Zircaloy/Inconel, Zircaloy/stainless steel, Zircaloy/(Ag-In-Cd), stainless steel B 4 C and Zircaloy/B 4 C. It was found that the reaction generally obeyed a parabolic rate law and the reaction rate was determined for each reaction system. Many debris samples obtained from the degraded core of TMI-2 were transported to JAERI for numerous examinations and analyses. The microstructural examination revealed that the most part of debris was ceramic and it was not homogeneous in a microscopic sense. The thermal diffusivity data was also obtained for the temperature range up to about 1800K. The data from the large scale integral experiments were also obtained through the

  20. Analytical nondestructive evaluation for materials characterization

    International Nuclear Information System (INIS)

    Raj, Baldev

    1993-01-01

    Science and technology of nondestructive testing and evaluation has contributed immensely to the safety and productivity of industrial plants. In recent years, nondestructive evaluation (NDE) has emerged as a frontline research area of equal if not greater technological relevance, for materials characterization as well. A comprehensive range of techniques from qualitative nondestructive testing for quality control of engineering products and materials to quantitative NDE for materials characterization is being used by the engineering industry and materials researchers, for better understanding of the manufacturing practices and materials behaviour. Quantitative NDE is considered essential for ensuring fitness for purpose at the start of the life in case the component has been designed using fracture mechanics parameters. Quantitative NDE is also vital for assessing degradation of material during service. Moreover, quantitative NDE enables characterization of dynamics of certain phenomenon (not achievable by destructive test methodologies) leading to better understanding of the performance of materials in relation to unavoidable defects in the materials. As the next logical step, the need for an analytical approach to NDE is felt. The need and motivation for such an approach is addressed and the means to achieve this objective are identified. It is argued that analytical NDE is essential to meet the challenges of characterization, intelligent processing of materials and life prediction of components and plants. These requirements are of significant importance in the context of recent developments in materials engineering, and for enhancing the competitive advantage of Indian engineering industry in the international market. (author). 9 refs., 3 figs

  1. Evaluation of the Fitness of Glass-Infiltrated Zirconia Core in Maxillary Central Incisor.

    Science.gov (United States)

    Kim, Ji-Won; Oh, Gye-Jeong; Lim, Hyun-Pil; Yun, Kwi-Dug; Park, Chan; Lee, Kyung-Ku; Ban, Jae-Sam; Park, Sang-Won; Yim, Eun-Kyung

    2018-02-01

    The purpose of this study was to evaluate the fitness of zirconia cores according to the amount and treated surface of glass infiltration. A maxillary right central incisor customized abutment was milled to have a 6° slope and a 1 mm deep chamfer margin and was manufactured in an intaglio mold using silicone impression material. Fifty-six stone dies were produced by injecting high strength dental stone into a mold and then zirconia cores were milled with CAD/CAM systems. The control group (Control) used non glass-infiltrated zirconia, and the experiment group was divided by one with the glass and distilled water ratio of 1:300 and the other with the ratio of 1:100. Each group was divided into subgroups by glasstreated surface: external surface infiltration, internal surface infiltration, and both surface infiltration. The zirconia cores sintered after glass infiltration were attached to the stone dies and then cut. Afterwards, the absolute marginal discrepancies and internal gaps of the buccal and lingual sides were measured. The buccal absolute marginal discrepancies and lingual internal gaps were influenced by the glass infiltration amount (p 0.05). As a result of the above experiments, the glass-infiltrated zirconia cores showed a clinically acceptable fitness, which is within 120 μm. This means that glass infiltration can be clinically used.

  2. Evaluation of BEACON-COLSS Core Monitoring System Benefits

    International Nuclear Information System (INIS)

    Kim, Joon Sung; Park, Young Ho; Morita, Toshio; Book, Michael A.

    2005-01-01

    In Korean Standard Nuclear Power Plant COLSS (Core Operating Limit Supervisory System) is used to monitor the DNBR Power Operating Limit (DNBRPOL) and Linear Heat Rate POL (KWPFPOL). Westinghouse and KNFC have developed an upgraded core monitoring system by combining the BEACON TM core monitoring system 1 (Best Estimate Analyzer for Core Operation . Nuclear) and COLSS into an integrated product that is called BEACON-COLSS. BEACON-COLSS generates the 3-D power distribution corrected by the in-core detectors measurements. The 3-D core power distribution methodology in BEACON-COLSS is significantly better than the synthesis methodology in COLSS. BEACONCOLSS uses the CETOP-D 2 thermal hydraulic code instead of CETOP-1. CETOP-D is a multi-channel thermal hydraulics code that will provide more accurate DNBR calculations than the DNBR calculators currently used in COLSS

  3. Evaluation of a hanging core support concept for LMR application

    International Nuclear Information System (INIS)

    Burelbach, J.P.; Cha, B.K.; Huebotter, P.R.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.; Wu, T.S.

    1985-01-01

    The paper describes an innovative design concept for a liquid metal reactor (LMR) core support structure (CSS). A hanging core support structure is described and analyzed. The design offers inherent safety features, constructability advantages, and potential cost reductions. Some safety considerations are examined which include the in-service inspection (ISI), the backup support system and the structural behavior in a hypothetical case of a broken beam in the core support structure

  4. Evaluation of Learning Materials: A Holistic Framework

    Science.gov (United States)

    Bundsgaard, Jeppe; Hansen, Thomas Illum

    2011-01-01

    This paper presents a holistic framework for evaluating learning materials and designs for learning. A holistic evaluation comprises investigations of the potential learning potential, the actualised learning potential, and the actual learning. Each aspect is explained and exemplified through theoretical models and definitions. (Contains 3 figures…

  5. Self-Healing Many-Core Architecture: Analysis and Evaluation

    Directory of Open Access Journals (Sweden)

    Arezoo Kamran

    2016-01-01

    Full Text Available More pronounced aging effects, more frequent early-life failures, and incomplete testing and verification processes due to time-to-market pressure in new fabrication technologies impose reliability challenges on forthcoming systems. A promising solution to these reliability challenges is self-test and self-reconfiguration with no or limited external control. In this work a scalable self-test mechanism for periodic online testing of many-core processor has been proposed. This test mechanism facilitates autonomous detection and omission of faulty cores and makes graceful degradation of the many-core architecture possible. Several test components are incorporated in the many-core architecture that distribute test stimuli, suspend normal operation of individual processing cores, apply test, and detect faulty cores. Test is performed concurrently with the system normal operation without any noticeable downtime at the application level. Experimental results show that the proposed test architecture is extensively scalable in terms of hardware overhead and performance overhead that makes it applicable to many-cores with more than a thousand processing cores.

  6. Evaluation of stator core loss of high speed motor by using thermography camera

    Science.gov (United States)

    Sato, Takeru; Enokizono, Masato

    2018-04-01

    In order to design a high-efficiency motor, the iron loss that is generated in the motor should be reduced. The iron loss of the motor is generated in a stator core that is produced with an electrical steel sheet. The iron loss characteristics of the stator core and the electrical steel sheet are agreed due to a building factor. To evaluate the iron loss of the motor, the iron loss of the stator core should be measured more accurately. Thus, we proposed the method of the iron loss evaluation of the stator core by using a stator model core. This stator model core has been applied to the surface mounted permanent magnet (PM) motors without windings. By rotate the permanent magnet rotor, the rotating magnetic field is generated in the stator core like a motor under driving. To evaluate the iron loss of the stator model core, the iron loss of the stator core can be evaluated. Also, the iron loss can be calculated by a temperature gradient. When the temperature gradient is measured by using thermography camera, the iron loss of entire stator core can be evaluated as the iron loss distribution. In this paper, the usefulness of the iron loss evaluation method by using the stator model core is shown by the simulation with FEM and the heat measurement with thermography camera.

  7. Results and Prospects of Development of Works on Structural Core Materials for Russian Fast Reactors

    International Nuclear Information System (INIS)

    Nikitina, A.A.; Ageev, V.S.; Leontyeva-Smirnova, M.V.; Mitrofanova, N.M.; Tselishchev, A.V.

    2015-01-01

    The strategy of development of atomic energy in Russia in the first half of XXI century contemplates construction and putting in operation of fast reactors of new generation with different types of coolant: sodium (BN-800, BN-1200, MBIR), lead (BREST-OD-300) and lead-bismuth eutectic (SVBR-100). For assurance of the working capacity of reactors that are under construction and achievement of economically reasonable burn-up of nuclear fuel the structural core materials with necessary level of radiation resistance, heat resistance, corrosion resistance to products of fuel fission, corrosion resistance in coolant and in water must be developed and justified. For sodium cooled reactors the key challenge is creation of radiation resistant and heat resistant cladding materials, which must ensure the achievement of damage doses at least 140 dpa. The solution of this problem is provided by phased use as cladding materials of austenitic steels ChS68 and EK164 (maximum damage doses ~ 92 and ~110-115 dpa, respectively), precipitation-hardening heat resistant ferritic-martensitic steels EK181 and ChS139 (maximum damage dose ~140 dpa) and oxide dispersion strengthened (ODS) steels (maximum damage dose more than 140 dpa). For development of core materials for reactors with lead and lead-bismuth eutectic coolants the most serious challenge is corrosion resistance of materials in coolant. Therefore at present time a very wide range of works on study of corrosion resistance of candidate materials is carrying out. As the basic material for the cladding tubes is considered a ferritic-martensitic steel EP823 with high silicon content. In this report the main results of works on justification of the working capacity of materials of different classes in respect to use it in cores of operating and prospective fast reactors with different types of coolant and prospects of further development of works are presented. (author)

  8. Evaluation of long-term post-accident core cooling of Three Mile Island Unit 2

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-04-15

    On the basis of current understanding of the accident scenario and available data, the staff reports here on its evaluation of the condition of the core and the core flow resistance as it might affect ability to cool the core by natural circulation. The natural circulation cooling capability of TMI-2 for the estimated core flow resistance and a variety of other conditions is evaluated and a comparison of the Base Case and off-nominal plant configurations is presented. The potential for and effects of natural convection core cooling are addressed, and the staff recommendations for reactor performance acceptance criteria upon initiation of natural convection are presented.

  9. Further work on sodium borates as sacrificial materials for a core-catcher

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Roth, A.; Werle, H.

    1982-01-01

    Sodium borates are suitable low melting point sacrificial materials for a core-catcher of a fast reactor. Concept, design and initial development work have been described previously. Here we report on the measurements of density, volumetric thermal expansion coefficients and viscosity of borax and sodium metaborate, pure and with various percentages of dissolved UO 2 . The density of these molten salts was measured with the buoyancy method in the temperature range 850 - 1300 0 C, while the viscosity was measured in the temperature range 700 - 1250 0 C with a Haake viscosity balance. Simulation experiments with low melting point materials were performed to investigate the ratio of the downward to sideward melt velocity. The results of these experiments show that this ratio is equal to 0.34 for a solid to liquid density ratio rho = 1.66. For the real borax core-catcher rho = 4 and this would correspond to a velocity ratio of about one

  10. Standard Test Method for Water Absorption of Core Materials for Structural Sandwich Constructions

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This test method covers the determination of the relative amount of water absorption by various types of structural core materials when immersed or in a high relative humidity environment. This test method is intended to apply to only structural core materials; honeycomb, foam, and balsa wood. 1.2 The values stated in SI units are to be regarded as the standard. The inch-pound units given may be approximate. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  11. Influence of high dose irradiation on core structural and fuel materials in advanced reactors

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA International Working Group on Fast Reactors (IWGFR) periodically organizes meeting to discuss and review important aspects of fast reactor technology. The fifth meeting held in Obninsk, Russian Federation, 16-19 June 1997, was devoted to the influence of high dose irradiation on the mechanical properties of reactor core structural and fuel materials. The proceedings includes the papers submitted at this meeting each with a separate abstract

  12. Thermal interactions of a molten core debris pool with surrounding structural materials

    International Nuclear Information System (INIS)

    Baker, L. Jr.; Cheung, F.B.; Farhadieh, R.; Stein, R.P.; Gabor, J.D.; Bingle, J.D.

    1979-01-01

    Analytical and experimental results on individual aspects of the overall problem of the interaction of a large mass of LMFBR core debris with concrete or other materials are reviewed. Results of recent heat transfer experiments with molten UO 2 have indicated the importance of internal thermal radiation and methods to take account of this are developed. Effects of gas release and density difference are considered. The GROWS-2 Code is used to illustrate the effects of various assumptions

  13. Corrosion evaluation of service water system materials

    International Nuclear Information System (INIS)

    Stein, A.A.; Felder, C.M.; Martin, R.L.

    1994-01-01

    The availability and reliability of the service water system is critical for safe operation of a nuclear power plant. Degradation of the system piping and components has forced utilities to re-evaluate the corrosion behavior of current and alternative system materials, to support assessments of the remaining service life of the service water system, selection of replacement materials, implementation of corrosion protection methods and corrosion monitoring programs, and identification of maintenance and operational constraints consistent with the materials used. TU Electric and Stone and Webster developed a service water materials evaluation program for the Comanche Peak Steam Electric Station. Because of the length of exposure and the generic interest in this program by the nuclear power industry, EPRI joined TU to co-sponsor the test program. The program was designed to evaluate the corrosion behavior of current system materials and candidate replacement materials and to determine the operational and design changes which could improve the corrosion performance of the system. Although the test program was designed to be representative of service water system materials and environments targeted to conditions at Comanche Peak, these conditions are typical of and relevant to other fresh water cooled nuclear service water systems. Testing was performed in raw water and water treated with biocide under typical service water operating conditions including continuous flow, intermittent flow, and stagnant conditions. The test program evaluated the 300 Series and 6% molybdenum stainless steels, copper-nickel, titanium, carbon steel, and a formed-in-place nonmetallic pipe lining to determine susceptibility to general, crevice, and microbiologically influenced corrosion and pitting attack. This report presents the results of the test program after 4 years of exposure

  14. The influence of anisotropy on the core structure of Shockley partial dislocations within FCC materials

    Science.gov (United States)

    Szajewski, B. A.; Hunter, A.; Luscher, D. J.; Beyerlein, I. J.

    2018-01-01

    Both theoretical and numerical models of dislocations often necessitate the assumption of elastic isotropy to retain analytical tractability in addition to reducing computational load. As dislocation based models evolve towards physically realistic material descriptions, the assumption of elastic isotropy becomes increasingly worthy of examination. We present an analytical dislocation model for calculating the full dissociated core structure of dislocations within anisotropic face centered cubic (FCC) crystals as a function of the degree of material elastic anisotropy, two misfit energy densities on the γ-surface ({γ }{{isf}}, {γ }{{usf}}) and the remaining elastic constants. Our solution is independent of any additional features of the γ-surface. Towards this pursuit, we first demonstrate that the dependence of the anisotropic elasticity tensor on the orientation of the dislocation line within the FCC crystalline lattice is small and may be reasonably neglected for typical materials. With this approximation, explicit analytic solutions for the anisotropic elasticity tensor {B} for both nominally edge and screw dislocations within an FCC crystalline lattice are devised, and employed towards defining a set of effective isotropic elastic constants which reproduce fully anisotropic results, however do not retain the bulk modulus. Conversely, Hill averaged elastic constants which both retain the bulk modulus and reasonably approximate the dislocation core structure are employed within subsequent numerical calculations. We examine a wide range of materials within this study, and the features of each partial dislocation core are sufficiently localized that application of discrete linear elasticity accurately describes the separation of each partial dislocation core. In addition, the local features (the partial dislocation core distribution) are well described by a Peierls-Nabarro dislocation model. We develop a model for the displacement profile which depends upon

  15. In-core materials testing under LWR conditions in the Halden reactor

    International Nuclear Information System (INIS)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A.

    2002-01-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  16. In-core materials testing under LWR conditions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  17. Irradiation experiments on materials for core internals, pressure vessel and fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Materials degradation due to the aging phenomena is one of the key issues for the life assessment and extension of the light water reactors (LWRs). This presentation introduces JAERI`s activities in the field of LWR material researches which utilize the research and testing reactors for irradiation experiments. The activities are including the material studies for the core internals, pressure vessel and fuel cladding. These materials are exposed to the neutron/gamma radiation and high temperature water environments so that it is worth reviewing their degradation phenomena as the continuum. Three topics are presented; For the core internal materials, the irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels is the present major concern. At JAERI the effects of alloying elements on IASCC have been investigated through the post-irradiation stress corrosion cracking tests in high-temperature water. The radiation embrittlement of pressure vessel steels is still a significant issue for LWR safety, and at JAERI some factors affecting the embrittlement behavior such as a dose rate have been investigated. Waterside corrosion of Zircaloy fuel cladding is one of the limiting factors in fuel rod performance and an in-situ measurement of the corrosion rate in high-temperature water was performed in JMTR. To improve the reliability of experiments and to extent the applicability of experimental techniques, a mutual utilization of the technical achievements in those irradiation experiments is desired. (author)

  18. French R&D on Materials for the Core Components of SFRs

    International Nuclear Information System (INIS)

    Le Flem, M.; Séran, J.L.; Blat-Yrieix, M.; Garat, V.

    2013-01-01

    ASTRID demonstrator 480-700°C, 110 dpa. • Use of reference materials benefiniting from a large feed-back from the previous French SFRs (Rapsodie, Phénix, SuperPhénix) • Austenitic steels (cladding), Martensitic steels (wrapper tube), B4C (absorbers). • Improving the description of their behavior (swelling, high temperature) • Qualifying the materials regarding the specificities of ASTRID core. Future SFRs 530-750, 180 dpa. • Use of advanced materials with improved properties • ODS ferritic/martensitic steels (cladding), Other metallic solutions as V alloys (cladding), SiC/SiC composites (wrapper tube), Innovative absorbers and reflectors. • R&D to develop/fabricate suitable grades • Qualifying these materials in ASTRID

  19. Research on the Core Competitive Power Elements Evaluation System of Green Hotel

    OpenAIRE

    Hui LIANG

    2013-01-01

    Green hotel is a new type of hospitality industry development model based on the concept of circular economy and sustainable development. This paper makes an analysis and evaluation of the elements of green hotel core competence, on this basis, constructs the Green Hotel core competitive evaluation index system. The construction of the system is conducive to understand the green hotel’s own competitive advantage objectively, and explore ways to enhance its core competitiveness, providing obje...

  20. Experimental evaluation of new NEU cores in the UVAR

    Energy Technology Data Exchange (ETDEWEB)

    Farrar, P.; Hosticka, B.; Krause, D.; Mulder, R.; Rydin, R. [Univ. of Virginia, Charlottesville, VA (United States)

    1997-08-01

    The University of Virginia began working on converting the UVAR reactor to LEU fuel in the Spring of 1986. The Safety Analysis Report was completed and submitted to the NRC in late 1989. After review, the DOE order to manufacture LEU fuel was placed at B&W in March 1992, and the new fuel was received in January 1994. The 4-by-4 fully-graphite-reflected LEU-1 core went critical on April 20, 1994, and the 4-by-5 partially-graphite-reflected operational LEU-2 core went critical on April 29. Full power was achieved on May 12, 1994. Both cores behaved very much as originally predicted. All of the old HEU fuel has been shipped to Savannah River.

  1. Characterization of a Porous Carbon Material Functionalized with Cobalt-Oxide/Cobalt Core-Shell Nanoparticles for Lithium Ion Battery Electrodes

    KAUST Repository

    Anjum, Dalaver H.

    2016-04-18

    A nanoporous carbon (C) material, functionalized with Cobalt-Oxide/Cobalt (CoO/Co) core-shell nanoparticles (NPs), was structurally and chemically characterized with transmission electron microcopy (TEM) while its electrochemical response for Lithium ion battery (LIB) applications was evaluated as well. The results herein show that the nanoporous C material was uniformly functionalized with the CoO/Co core-shell NPs. Further the NPs were crystalline with fcc-Type lattice on the Co2+ oxide shell and hcp-Type core of metallic Co0. The electrochemical study was carried out by using galvanostatic charge/discharge cycling at a current density of 1000 mA g-1. The potential of this hybrid material for LIB applications was confirmed and it is attributed to the successful dispersion of the Co2+/ Co0 NPs in the C support.

  2. Evaluating nuclear physics inputs in core-collapse supernova models

    Science.gov (United States)

    Lentz, E.; Hix, W. R.; Baird, M. L.; Messer, O. E. B.; Mezzacappa, A.

    Core-collapse supernova models depend on the details of the nuclear and weak interaction physics inputs just as they depend on the details of the macroscopic physics (transport, hydrodynamics, etc.), numerical methods, and progenitors. We present preliminary results from our ongoing comparison studies of nuclear and weak interaction physics inputs to core collapse supernova models using the spherically-symmetric, general relativistic, neutrino radiation hydrodynamics code Agile-Boltztran. We focus on comparisons of the effects of the nuclear EoS and the effects of improving the opacities, particularly neutrino--nucleon interactions.

  3. Moving Away from Exhaustion: How Core Self-Evaluations Influence Academic Burnout

    Science.gov (United States)

    Lian, Penghu; Sun, Yunfeng; Ji, Zhigang; Li, Hanzhong; Peng, Jiaxi

    2014-01-01

    Background Academic burnout refers to students who have low interest, lack of motivation, and tiredness in studying. Studies concerning how to prevent academic burnout are rare. Objective The present study aimed to investigate the impact of core self-evaluations on the academic burnout of university students, and mainly focused on the confirmation of the mediator role of life satisfaction. Methods A total of 470 university students accomplished the core self-evaluation scale, Satisfaction with Life, and academic burnout scale. Results Both core self-evaluations and life satisfaction were significantly correlated with academic burnout. Structural equation modeling indicated that life satisfaction partially mediated the relationship between core self-evaluations and academic burnout. Conclusions Core self-evaluations significantly influence academic burnout and are partially mediated by life satisfaction. PMID:24489857

  4. Moving away from exhaustion: how core self-evaluations influence academic burnout.

    Directory of Open Access Journals (Sweden)

    Penghu Lian

    Full Text Available BACKGROUND: Academic burnout refers to students who have low interest, lack of motivation, and tiredness in studying. Studies concerning how to prevent academic burnout are rare. OBJECTIVE: The present study aimed to investigate the impact of core self-evaluations on the academic burnout of university students, and mainly focused on the confirmation of the mediator role of life satisfaction. METHODS: A total of 470 university students accomplished the core self-evaluation scale, Satisfaction with Life, and academic burnout scale. RESULTS: Both core self-evaluations and life satisfaction were significantly correlated with academic burnout. Structural equation modeling indicated that life satisfaction partially mediated the relationship between core self-evaluations and academic burnout. CONCLUSIONS: Core self-evaluations significantly influence academic burnout and are partially mediated by life satisfaction.

  5. Moving away from exhaustion: how core self-evaluations influence academic burnout.

    Science.gov (United States)

    Lian, Penghu; Sun, Yunfeng; Ji, Zhigang; Li, Hanzhong; Peng, Jiaxi

    2014-01-01

    Academic burnout refers to students who have low interest, lack of motivation, and tiredness in studying. Studies concerning how to prevent academic burnout are rare. The present study aimed to investigate the impact of core self-evaluations on the academic burnout of university students, and mainly focused on the confirmation of the mediator role of life satisfaction. A total of 470 university students accomplished the core self-evaluation scale, Satisfaction with Life, and academic burnout scale. Both core self-evaluations and life satisfaction were significantly correlated with academic burnout. Structural equation modeling indicated that life satisfaction partially mediated the relationship between core self-evaluations and academic burnout. Core self-evaluations significantly influence academic burnout and are partially mediated by life satisfaction.

  6. Effects of core-to-dentin thickness ratio on the biaxial flexural strength, reliability, and fracture mode of bilayered materials of zirconia core (Y-TZP) and veneer indirect composite resins.

    Science.gov (United States)

    Su, Naichuan; Liao, Yunmao; Zhang, Hai; Yue, Li; Lu, Xiaowen; Shen, Jiefei; Wang, Hang

    2017-01-01

    Indirect composite resins (ICR) are promising alternatives as veneering materials for zirconia frameworks. The effects of core-to-dentin thickness ratio (C/Dtr) on the mechanical property of bilayered veneer ICR/yttria-tetragonal zirconia polycrystalline (Y-TZP) core disks have not been previously studied. The purpose of this in vitro study was to assess the effects of C/Dtr on the biaxial flexural strength, reliability, and fracture mode of bilayered veneer ICR/ Y-TZP core disks. A total of 180 bilayered 0.6-mm-thick composite resin disks in core material and C/Dtr of 2:1, 1:1, and 1:2 were tested with either core material placed up or placed down for piston-on-3-ball biaxial flexural strength. The mean biaxial flexural strength, Weibull modulus, and fracture mode were measured to evaluate the variation trend of the biaxial flexural strength, reliability, and fracture mode of the bilayered disks with various C/Dtr. One-way analysis of variance (ANOVA) and chi-square tests were used to evaluate the variation tendency of fracture mode with the C/Dtr or material placed down during testing (α=.05). Light microscopy was used to identify the fracture mode. The mean biaxial flexural strength and reliability improved with the increase in C/Dtr when specimens were tested with the core material either up and down, and depended on the materials that were placed down during testing. The rates of delamination, Hertzian cone cracks, subcritical radial cracks, and number of fracture fragments partially depended on the C/Dtr and the materials that were placed down during testing. The biaxial flexural strength, reliability, and fracture mode in bilayered structures of Y-TZP core and veneer ICR depend on both the C/Dtr and the material that was placed down during testing. Copyright © 2016 Editorial Council for the Journal of Prosthetic Dentistry. Published by Elsevier Inc. All rights reserved.

  7. Characterization and damage evaluation of advanced materials

    Science.gov (United States)

    Mitrovic, Milan

    Mechanical characterization of advanced materials, namely magnetostrictive and graphite/epoxy composite materials, is studied in this dissertation, with an emphasis on damage evaluation of composite materials. Consequently, the work in this dissertation is divided into two parts, with the first part focusing on characterization of the magneto-elastic response of magnetostrictlve materials, while the second part of this dissertation describes methods for evaluating the fatigue damage in composite materials. The objective of the first part of this dissertation is to evaluate a nonlinear constitutive relation which more closely depict the magneto-elastic response of magnetostrictive materials. Correlation between experimental and theoretical values indicate that the model adequately predicts the nonlinear strain/field relations in specific regimes, and that the currently employed linear approaches are inappropriate for modeling the response of this material in a structure. The objective of the second part of this dissertation is to unravel the complexities associated with damage events associated with polymeric composite materials. The intent is to characterize and understand the influence of impact and fatigue induced damage on the residual thermo-mechanical properties and compressive strength of composite systems. The influence of fatigue generated matrix cracking and micro-delaminations on thermal expansion coefficient (TEC) and compressive strength is investigated for woven graphite/epoxy composite system. Experimental results indicate that a strong correlation exists between TEC and compressive strength measurements, indicating that TEC measurements can be used as a damage metric for this material systems. The influence of delaminations on the natural frequencies and mode shapes of a composite laminate is also investigated. Based on the changes of these parameters as a function of damage, a methodology for determining the size and location of damage is suggested

  8. Mechanical properties of chemically bonded sand core materials dipped in sol-gel coating impregnated with filter

    DEFF Research Database (Denmark)

    Nwaogu, Ugochukwu Chibuzoh; Tiedje, Niels Skat

    2012-01-01

    A novel sol-gel coating impregnated with filter dust was applied on chemically bonded sand core materials by dipping. After curing, the strengths of the core materials were measured under uniaxial loading using a new strength testing machine (STM). The STM presents the loading history as a force-...... of the chemically bonded sand core materials, a combination of flexural and compression tests is suggested for improving the casting quality. © 2012 W. S. Maney & Son Ltd.......A novel sol-gel coating impregnated with filter dust was applied on chemically bonded sand core materials by dipping. After curing, the strengths of the core materials were measured under uniaxial loading using a new strength testing machine (STM). The STM presents the loading history as a force...... the strengths were increased under compression. The mode of fracture of the chemically bonded sand core materials was observed to be intergranular through the binder. The stiffness of the chemically bonded sand core materials was determined. For better understanding of the mechanical properties...

  9. Kinetic parameters evaluation of PWRs using static cell and core calculation codes

    International Nuclear Information System (INIS)

    Jahanbin, Ali; Malmir, Hessam

    2012-01-01

    Highlights: ► In this study, we have calculated effective delayed neutron fraction and prompt neutron lifetime in PWRs. ► New software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. ► This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. ► The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. - Abstract: In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs, using static cell and core calculation codes, is reported. A new software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. Using the WIMS cell calculation code, multigroup microscopic cross-sections and number densities of different materials can be generated in a binary file. By the use of BORGES code, these binary-form cross-sections and number densities are converted to a format readable by the CITATION core calculation code, by which the kinetic parameters can be finally obtained. This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. Benchmarking of the results against the final safety analysis report (FSAR) of the aforementioned reactors shows very good agreements with these published documents.

  10. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    International Nuclear Information System (INIS)

    Prabhu Gaunkar, N.; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C.; Bulu, I.; Ganesan, K.; Song, Y. Q.

    2015-01-01

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors

  11. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    Energy Technology Data Exchange (ETDEWEB)

    Prabhu Gaunkar, N., E-mail: neelampg@iastate.edu; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C. [Department of Electrical and Computer Engineering, Iowa State University, Ames, Iowa 50011 (United States); Bulu, I.; Ganesan, K.; Song, Y. Q. [Schlumberger-Doll Research, Cambridge, Massachusetts 02139 (United States)

    2015-05-07

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors.

  12. Studies on the core-support carbon material for VHTR, (1)

    International Nuclear Information System (INIS)

    Matsuo, Hideto; Saito, Tamotsu; Fukuda, Yasumasa; Sasaki, Yasuichi; Hasegawa, Takashi.

    1979-11-01

    To obtain information of core-support carbon material for VHTR, thermal conductivity and electrical resistivity of three domestic carbon blocks were measured. Results indicated the need for development of carbon material with lower thermal conductivity for VHTR. These two were also measured of the samples heat-treated between 1000 0 C and 3040 0 C for one hour. Thermal conductivity increased with heat-treatment above 1200 0 C and resistivity stayed constant between 1500 0 C and 2000 0 C. The results should be useful in choosing the final heat-treatment temperature in carbon material production. The changes of Lorentz number with heat treatment were classified into three heat-treatment temperature regions of below 1500 0 C, 1500 0 C - 2500 0 C, and above 2500 0 C; the results are interpreted with a graphitization model. (author)

  13. Evaluation of aseismic integrity in HTTR core-bottom structure. Pt. 1. Aseismic test for core-bottom structure

    International Nuclear Information System (INIS)

    Iyoku, T.; Futakawa, M.; Ishihara, M.

    1994-01-01

    The aseismic tests were carried out using (1)/(5)-scale and (1)/(3)-scale models of the core-bottom structure of the HTTR to quantitatively evaluate the response of acceleration, strain, impact load etc. The following conclusions are obtained. (i) The frequency response of the keyway strain is correlative with that of the impact acceleration on the hot plenum block. (ii) It was confirmed through (1)/(5)-scale and (1)/(3)-scale model tests that the applied similarity law is valid to evaluate the seismic response characteristics of the core-bottom structure. (ii) The stress of graphite components estimated from the scale model test using S 2 -earthquake excitation was sufficiently lower than the allowable stress used as the design criterion. ((orig.))

  14. The bond of different post materials to a resin composite cement and a resin composite core material.

    Science.gov (United States)

    Stewardson, D; Shortall, A; Marquis, P

    2012-01-01

    To investigate the bond of endodontic post materials, with and without grit blasting, to a resin composite cement and a core material using push-out bond strength tests. Fiber-reinforced composite (FRC) posts containing carbon (C) or glass (A) fiber and a steel (S) post were cemented into cylinders of polymerized restorative composite without surface treatment (as controls) and after grit blasting for 8, 16, and 32 seconds. Additional steel post samples were sputter-coated with gold before cementation to prevent chemical interaction with the cement. Cylindrical composite cores were bonded to other samples. After sectioning into discs, bond strengths were determined using push-out testing. Profilometry and electron microscopy were used to assess the effect of grit blasting on surface topography. Mean (standard deviation) bond strength values (MPa) for untreated posts to resin cement were 8.41 (2.80) for C, 9.61(1.88) for A, and 19.90 (3.61) for S. Prolonged grit blasting increased bond strength for FRC posts but produced only a minimal increase for S. After 32 seconds, mean values were 20.65 (4.91) for C, 20.41 (2.93) for A, and 22.97 (2.87) for S. Gold-coated steel samples produced the lowest bond strength value, 7.84 (1.40). Mean bond strengths for untreated posts bonded to composite cores were 6.19 (0.95) for C, 13.22 (1.61) for A, and 8.82 (1.18) for S, and after 32 seconds of grit blasting the values were 17.30 (2.02) for C, 26.47 (3.09) for A, and 20.61 (2.67) for S. FRC materials recorded higher roughness values before and after grit blasting than S. With prolonged grit blasting, roughness increased for A and C, but not for S. There was no evidence of significant bonding to untreated FRC posts, but significant bonding occurred between untreated steel posts and the resin cement. Increases in the roughness of FRC samples were material dependent and roughening significantly increased bond strength values (p<0.05). Surface roughening of the tested FRC posts is

  15. Effect of silica fiber on the mechanical and chemical behavior of alumina-based ceramic core material

    OpenAIRE

    Weiguo Jiang; Kaiwen Li; Jiuhan Xiao; Langhong Lou

    2017-01-01

    In order to improve the chemical leachability, the alumina-based ceramic core material with the silica fiber was injected and sintered at 1100 °C/4 h, 1200 °C/4 h, 1300 °C/4 h and 1400 °C/4 h, respectively. The micrographs of ceramic core materials at sintered and leached state were characterized by scanning electron microscopy (SEM). The phase composition of ceramic core material after sintering and the leaching product after leaching were detected by X-ray diffraction (XRD). The porosity, r...

  16. Assessment on Evaluating Parameters of Rice Core Collections Constructed by Genotypic Values and Molecular Marker Information

    Directory of Open Access Journals (Sweden)

    Jian-cheng WANG

    2007-06-01

    Full Text Available Eleven evaluating parameters for rice core collection were assessed based on genotypic values and molecular marker information. Monte Carlo simulation combined with mixed linear model was used to eliminate the interference from environment in order to draw more reliable results. The coincidence rate of range (CR was the optimal parameter. Mean Simpson index (MD, mean Shannon-Weaver index of genetic diversity (MI and mean polymorphism information content (MPIC were important evaluating parameters. The variable rate of coefficient of variation (VR could act as an important reference parameter for evaluating the variation degree of core collection. Percentage of polymorphic loci (p could be used as a determination parameter for the size of core collection. Mean difference percentage (MD was a determination parameter for the reliability judgment of core collection. The effective evaluating parameters for core collection selected in the research could be used as criteria for sampling percentage in different plant germplasm populations.

  17. Magnetite Core-Shell Nanoparticles in Nondestructive Flaw Detection of Polymeric Materials.

    Science.gov (United States)

    Hetti, Mimi; Wei, Qiang; Pohl, Rainer; Casperson, Ralf; Bartusch, Matthias; Neu, Volker; Pospiech, Doris; Voit, Brigitte

    2016-10-04

    Nondestructive flaw detection in polymeric materials is important but difficult to achieve. In this research, the application of magnetite nanoparticles (MNPs) in nondestructive flaw detection is studied and realized, to the best of our knowledge, for the first time. Superparamagnetic and highly magnetic (up to 63 emu/g) magnetite core-shell nanoparticles are prepared by grafting bromo-end-group-functionalized poly(glycidyl methacrylate) (Br-PGMA) onto surface-modified Fe 3 O 4 NPs. These Fe 3 O 4 -PGMA NPs are blended into bisphenol A diglycidylether (BADGE)-based epoxy to form homogeneously distributed magnetic epoxy nanocomposites (MENCs) after curing. The core Fe 3 O 4 of the Fe 3 O 4 -PGMA NPs endows the MENCs with magnetic property, which is crucial for nondestructive flaw detection of the materials, while the shell PGMA promotes colloidal stability and prevents NP aggregation during curing. The eddy current testing (ET) technique is first applied to detect flaws in the MENCs. Through the brightness contrast of the ET image, surficial and subsurficial flaws in MENCs can be detected, even for MENCs with low content of Fe 3 O 4 -PGMA NPs (1 wt %). The incorporation of Fe 3 O 4 -PGMA NPs can be easily extended to other polymer and polymer-based composite systems and opens a new and very promising pathway toward MNP-based nondestructive flaw detection in polymeric materials.

  18. Exploratory study of molten core material/concrete interactions, July 1975--March 1977

    International Nuclear Information System (INIS)

    Powers, D.A.; Dahlgren, D.A.; Muir, J.F.; Murfin, W.D.

    1978-02-01

    An experimental study of the interaction between high-temperature molten materials and structural concrete is described. The experimental efforts focused on the interaction of melts of reactor core materials weighing 12 to 200 kg at temperatures 1700 to 2800 0 C with calcareous and basaltic concrete representative of that found in existing light-water nuclear reactors. Observations concerning the rate and mode of melt penetration into concrete, the nature and generation rate of gases liberated during the interaction, and heat transfer from the melt to the concrete are described. Concrete erosion is shown to be primarily a melting process with little contribution from mechanical spallation. Water and carbon dioxide thermally released from the concrete are extensively reduced to hydrogen and carbon monoxide. Heat transfer from the melt to the concrete is shown to be dependent on gas generation rate and crucible geometry. Interpretation of results from the interaction experiments is supported by separate studies of the thermal decomposition of concretes, response of bulk concrete to intense heat fluxes (28 to 280 W/cm 2 ), and heat transfer from molten materials to decomposing solids. The experimental results are compared to assumptions made in previous analytic studies of core meltdown accidents in light-water nuclear reactors. A preliminary computer code, INTER, which models and extrapolates results of the experimental program is described. The code allows estimation of the effect of physical parameters on the nature of the melt/concrete interaction

  19. Evaluation Of Oxide And Silicide Mixed Fuels Of The RSG-GAS Core

    International Nuclear Information System (INIS)

    Tukiran; Sembiring, Tagor Malem; Suparlina, Lily

    2000-01-01

    Fuel exchange of the RSG-GAS reactor core from uranium oxide to uranium silicide in the same loading, density, and enrichment, that is 250 gr, 2.98 gr/cm 3 , and 19.75%, respectively, will be performed in-step wise. In every cycle of exchange with 5/1 mode, it is needed to evaluate the parameter of reactor core operation. The parameters of the reactor operation observed are criticality mass of fuels, reactivity balance, and fuel reactivity that give effect to the reactor operation. The evaluation was done at beginning of cycle of the first and second transition core with compared between experiment and calculation results. The experiments were performed at transition core I and II, BOC, and low power. At transition core I, there are 2 silicide fuels (RI-224 and R1-225) in the core and then, added five silicide fuels (R1-226, R1-252, R1-263, and R1-264) to the core, so that there are seven silicide fuels in the transition core II. The evaluation was done based on the experiment of criticality, control rod calibration, fuel reactivity of the RSG-GAS transition core. For inserting 2 silicide fuels in the transition core I dan 7 fuels in the transition core II, the operation of RSG-GAS core fulfilled the safety margin and the parameter of reactor operation change is not occur drastically in experiment and calculation results. So that, the reactor was operated during 36 days at 15 MW, 540 MWD at the first transition core. The general result showed that the parameter of reactor operation change is small so that the fuel exchange from uranium oxide to uranium silicide in the next step can be done

  20. The effects of radiations on materials for core internals of PWRs: EDF-CEA-Framatome joint research programme

    International Nuclear Information System (INIS)

    Mathan, N. de; Buisine, D.; Goltrant, O.; Dubuisson, P.; Scott, P.; Deydier, D.; Trenty, A.

    1998-01-01

    The effects of neutron irradiation on materials for the core internals of PWRs (austenitic stainless steels) are potentially a significant economic and regulatory concern for EDF. The maintenance strategy for EDF relies primarily on in-service inspection, safety analysis and characterization of materials irradiated in-service. In addition, to anticipate likely future behaviour of highly irradiated materials, EDF has initiated, in collaboration with CEA and Framatome, a large R and D programme designed to (i) evaluate the effects of neutron irradiation on mechanical properties and stress corrosion cracking sensitivity (IASCC), and (ii) identify possible replacement materials. The programme, currently in progress, involves mechanical tests (tensile, fracture toughness, irradiation creep), stress corrosion cracking tests (in flux and out of flux) and metallurgical examinations. The test materials are being irradiated in several experimental reactors in France and Russia up to PWR-related end of life doses (∼ 80 dpa) at several PWR-relevant irradiation temperatures (300-400 deg. C). The presentation will describe the objectives and early results of this ongoing R and D programme. (author)

  1. An efficient strategy for designing ambipolar organic semiconductor material: Introducing dehydrogenated phosphorus atoms into pentacene core

    Science.gov (United States)

    Tang, Xiao-Dan

    2017-09-01

    The charge transport properties of phosphapentacene (P-PEN) derivatives were systematically explored by theoretical calculation. The dehydrogenated P-PENs have reasonable frontier molecular orbital energy levels to facilitate both electron and hole injection. The reduced reorganization energies of dehydrogenated P-PENs could be intimately connected to the bonding nature of phosphorus atoms. From the idea of homology modeling, the crystal structure of TIPSE-4P-2p is constructed and fully optimized. Fascinatingly, TIPSE-4P-2p shows the intrinsic property of ambipolar transport in both hopping and band models. Thus, introducing dehydrogenated phosphorus atoms into pentacene core could be an efficient strategy for designing ambipolar material.

  2. Exfoliated BN shell-based high-frequency magnetic core-shell materials.

    Science.gov (United States)

    Zhang, Wei; Patel, Ketan; Ren, Shenqiang

    2017-09-14

    The miniaturization of electric machines demands high frequency magnetic materials with large magnetic-flux density and low energy loss to achieve a decreased dimension of high rotational speed motors. Herein, we report a solution-processed high frequency magnetic composite (containing a nanometal FeCo core and a boron nitride (BN) shell) that simultaneously exhibits high electrical resistivity and magnetic permeability. The frequency dependent complex initial permeability and the mechanical robustness of nanocomposites are intensely dependent on the content of BN insulating phase. The results shown here suggest that insulating magnetic nanocomposites have potential for application in next-generation high-frequency electric machines with large electrical resistivity and permeability.

  3. Evaluation of CCTF Core-II second acceptance Test C2-AC2 (Run 052)

    International Nuclear Information System (INIS)

    Okubo, Tsutomu; Murao, Yoshio

    1984-03-01

    In order to investigate the thermo-hydrodynamic behavior in a PWR during the reflood phase of the LOCA, large scale reflooding tests have been conducted at JAERI using the CCTF Core-I and Core-II facilities. This report presents the investigation on the difference in the thermo-hydrodynamic behavior observed between in the CCTF Core-I and Core-II facilities. For this purpose the test data of the second CCTF Core-II acceptance test C2-AC2 (Run 052) were evaluated by using the data of the Test CL-21 (Run 040) in the Core-I test series. The experimental conditions for these two tests were almost identical. Comparing the data of those two tests, the following is obtained. 1. The system behavior observed in the Core-II facility was nearly identical to that observed in the Core-I facility. 2. The core behavior observed in the Core-II facility was also nearly identical to that observed in the Core-I facility except for the top quenching behavior. 3. The differences in the top quenching behavior between the two facilities were as follows: (1) The selective occurrence of top quenching below the open holes of the upper core support plate observed in the Core-I facility was not observed in the Core-II facility. (2) Top quenching tended to occur less in the Core-II facility in the region where the initial average linear power density was over 1.69 kW/m. (author)

  4. Current R and D status on material motion and interactions relevant to core disruptive accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Satoru [Safety Engineering Division, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, O-arai, Ibaraki (Japan)

    1994-07-01

    In this paper, the current status of research and development activities are briefly reviewed on evaluation of material-coolant interactions and material movement and relocation relevant to the safety of liquid-metal fast reactors. Since the status of European activities are well summarized in other papers submitted to the present meeting, the activities in Japan and the United States are highlighted in this paper. The review includes: out-of-pile experiments, in-pile experiments and relevant computer code development. It is emphasized that improved understanding on material motion has contributed to establishing more realistic and rational safety evaluation methods, where various mitigation mechanisms are inherently effective. (author)

  5. Role of core support material in veneer failure of brittle layer structures.

    Science.gov (United States)

    Hermann, Ilja; Bhowmick, Sanjit; Lawn, Brian R

    2007-07-01

    A study is made of veneer failure by cracking in all-ceramic crown-like layer structures. Model trilayers consisting of a 1 mm thick external glass layer (veneer) joined to a 0.5 mm thick inner stiff and hard ceramic support layer (core) by epoxy bonding or by fusion are fabricated for testing. The resulting bilayers are then glued to a thick compliant polycarbonate slab to simulate a dentin base. The specimens are subjected to cyclic contact (occlusal) loading with spherical indenters in an aqueous environment. Video cameras are used to record the fracture evolution in the transparent glass layer in situ during testing. The dominant failure mode is cone cracking in the glass veneer by traditional outer (Hertzian) cone cracks at higher contact loads and by inner (hydraulically pumped) cone cracks at lower loads. Failure is deemed to occur when one of these cracks reaches the veneer/core interface. The advantages and disadvantages of the alumina and zirconia core materials are discussed in terms of mechanical properties-strength and toughness, as well as stiffness. Consideration is also given to the roles of interface strength and residual thermal expansion mismatch stresses in relation to the different joining methods. Copyright 2006 Wiley Periodicals, Inc.

  6. Implication of irradiation effects on materials data for the design of near core components

    International Nuclear Information System (INIS)

    Dietz, W.; Breitling, H.

    1995-01-01

    For LWR's strict regulations exist for the consideration of irradiation in the design and surveillance of the reactor pressure vessel in the various codes (ASME, RCC-M, KTA) but less for near core components. For FBR's no firm rules exist either for the vessel nor the reactor internals. In this paper the German design practices for the loop type SNR-300 will be presented, and also some information from the surveillance programme of the KNK-reactor. Austenitic stainless steels have been mainly selected for the near core components. For some special applications Ni-alloys and a stabilized 2 1/4 Cr 1 Mo-alloy were specified. Considerations of the irradiation effects on material properties will be made for the various temperature and fluence levels around the core. The surveillance programmes will be described. Both, the consideration of irradiation effects in the elastic and inelastic analysis and the surveillance programmes had been a part of the licensing process for SNR-300. (author). 8 figs, 4 tabs

  7. Modelling mechanical properties of the multilayer composite materials with the polyamide core

    Directory of Open Access Journals (Sweden)

    Talaśka Krzysztof

    2018-01-01

    Full Text Available Due to the wide range of application for belt conveyors, engineers look for many different combinations of mechanical properties of conveyor and transmission belts. It can be made by creating multilayer or fibre reinforced composite materials from base thermoplastic or thermosetting polymers. In order to gain high strength with proper elasticity and friction coefficient, the core of the composite conveyor belt is made of polyamide film core, which can be combined with various types of polymer fabrics, films or even rubbers. In this paper authors show the complex model of multilayer composite belt with the polyamide core, which can be used in simulation analyses. The following model was derived based on the experimental research, which consisted of tensile, compression and shearing tests. In order to achieve the most accurate model, proper simulations in ABAQUS were made and then the results were compared with empirical mechanical characteristics of a conveyor belt. The main goal of this research is to fully describe the perforation process of conveyor and transmission belts for vacuum belt conveyors. The following model will help to develop design briefs for machines used for mechanical perforation.

  8. Evaluation Of Potting Materials For Use In Extreme Cold

    Science.gov (United States)

    Acosta, Ernesto

    1992-01-01

    Tests help identify noncracking combinations of materials. Aid evaluation of potting materials for copper coils used at low temperatures to measure magnetic fields. Also determine effects of distribution of microballoons, voids, and porosity. Materials also evaluated for ease of use.

  9. Nonlinear Thermo-mechanical Finite Element Analysis of Polymer Foam Cored Sandwich Structures including Geometrical and Material Nonlinearity

    DEFF Research Database (Denmark)

    Palleti, Hara Naga Krishna Teja; Thomsen, Ole Thybo; Taher, Siavash Talebi

    In this paper, polymer foam cored sandwich structures with fibre reinforced composite face sheets subjected to combined mechanical and thermal loads will be analysed using the commercial FE code ABAQUS® incorporating both material and geometrical nonlinearity. Large displacements and rotations...

  10. A new mechanism of hydrogen absorption in water-water reactor core materials

    International Nuclear Information System (INIS)

    Gann, V.V.; Gann, A.V.

    2012-01-01

    The spectrum of fast protons, generated in water by fast neutrons of WWER-1000 reactor core, has been calculated using the code MCNPX. The main mechanism of fast proton generation in the moderator is found to be elastic scattering of fast neutrons on hydrogen nuclei. Fast protons with mean energy 1 MeV flow towards the surface of cladding material at flux density ∼ 0.1 μA/cm 2 . Proton range distribution profile in cladding material is calculated. The range of fast protons in zirconium averages 20 μm, the maximal proton range is larger than 200 μm. The rate of hydrogen deposition in 40 μm layer amounts to 5 x 10 -5 H/n/μ. A role of the suggested mechanism in process of zirconium clad hydrogenation during reactor irradiation is discussed.

  11. Relative translucency of six all-ceramic systems. Part II: core and veneer materials.

    Science.gov (United States)

    Heffernan, Michael J; Aquilino, Steven A; Diaz-Arnold, Ana M; Haselton, Debra R; Stanford, Clark M; Vargas, Marcos A

    2002-07-01

    STATEMENT OF PROBLEM All-ceramic core materials with various strengthening compositions have a range of translucencies. It is unknown whether translucency differs when all-ceramic materials are fabricated similarly to the clinical restoration with a veneered core material. This study compared the translucency of 6 all-ceramic materials veneered and glazed at clinically appropriate thicknesses. Core specimens (n = 5 per group) of Empress dentin, Empress 2 dentin, In-Ceram Alumina, In-Ceram Spinell, In-Ceram Zirconia, and Procera AllCeram were fabricated as described in Part I of this study and veneered with their corresponding dentin porcelain to a final thickness of 1.47 +/- 0.01 mm. These specimens were compared with veneered Vitadur Alpha opaque dentin (as a standard), a clear glass disc (positive control), and a high-noble metal-ceramic alloy (Porc. 52 SF) veneered with Vitadur Omega dentin (negative control). Specimen reflectance was measured with an integrating sphere attached to a spectrophotometer across the visible spectrum (380 to 700 nm); 0-degree illumination and diffuse viewing geometry were used. Measurements were repeated after a glazing cycle. Contrast ratios were calculated from the luminous reflectance (Y) of the specimens with a black (Yb) and a white backing (Yw) to give Yb/Yw with CIE illuminant D65 and a 2-degree observer function (0.0 = transparent, 1.0 = opaque). One-way analysis of variance and Tukey's multiple-comparison test were used to analyze the data (P<.05). Significant differences in contrast ratios were found among the ceramic systems tested when they were veneered (P<.0001) and after the glazing cycle (P<.0001). Significant changes in contrast ratios (P<.0001) also were identified when the veneered specimens were glazed. Within the limitations of this study, a range of translucency was identified in the veneered all-ceramic systems tested. Such variability may affect their ability to match natural teeth. The glazing cycle resulted

  12. Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

    International Nuclear Information System (INIS)

    S, Tukiran; MS, Tagor; P, Surian

    2003-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safely to a full-silicide core

  13. A resin composite material containing an eugenol derivative for intracanal post cementation and core build-up restoration.

    Science.gov (United States)

    Almaroof, A; Rojo, L; Mannocci, F; Deb, S

    2016-02-01

    To formulate and evaluate new dual cured resin composite based on the inclusion of eugenyl methacrylate monomer (EgMA) with Bis-GMA/TEGDMA resin systems for intracanal post cementation and core build-up restoration of endodontically treated teeth. EgMA was synthesized and incorporated at 5% (BTEg5) or 10% (BTEg10) into dual-cure formulations. Curing properties, viscosity, Tg, radiopacity, static and dynamic mechanical properties of the composites were determined and compared with Clearfil™DC Core-Plus, a commercial dual-cure, two-component composite. Statistical analysis of the data was performed with ANOVA and the Tukey's post-hoc test. The experimental composites were successfully prepared, which exhibited excellent curing depths of 4.9, 4.7 and 4.2 mm for BTEg0, BTEg5 and BTEg10 respectively, which were significantly higher than Clearfil™DC. However, the inclusion of EgMA initially led to a lower degree of cure, which increased when measured at 24 h with values comparable to formulations without EgMA, indicating post-curing. The inclusion of EgMA also lowered the polymerization exotherm thereby reducing the potential of thermal damage to host tissue. Both thermal and viscoelastic analyses confirmed the ability of the monomer to reduce the stiffness of the composites by forming a branched network. The compressive strength of BTEg5 was significantly higher than the control whilst flexural strength increased significantly from 95.9 to 114.8 MPa (BTEg5) and 121.9 MPa (BTEg10). Radiopacity of the composites was equivalent to ∼3 mm Al allowing efficient diagnosis. The incorporation of EgMA within polymerizable formulations provides a novel approach to prepare reinforced resin composite material for intracanal post cementation and core build-up and the potential to impart antibacterial properties of eugenol to endodontic restorations. Copyright © 2015 Academy of Dental Materials. Published by Elsevier Ltd. All rights reserved.

  14. Assessment of the Possibility of Using Reclaimed Materials for Making Cores by the Blowing Method

    Directory of Open Access Journals (Sweden)

    Dańko R.

    2017-03-01

    Full Text Available The cumulative results of investigations of the possibility of using the reclaimed materials after the mechanical, thermal or mechanical-thermal reclamation for making cores by means of the blowing method in the alkaline CO2 technology, are presented in the paper. Three kinds of spent sands: with furfuryl resin, bentonite and alkaline phenolic resin, obtained from the foundry, were subjected to three kinds of reclamation: mechanical, thermal and combined mechanical-thermal, applying for this aim adequate experimental devices. The obtained reclaims were assessed with regard to the degree of the matrix liberation from the determined binding material. Reclaims of moulding sands with binders of the form of resin were assessed with regard to ignition loss values and pH reaction, while reclaims of moulding sands with bentonite with regard to the residual clay content and pH value. In all cases the results of the performed sieve analyses were estimated and the average characteristic diameter dl was determined. The reclaimed matrix was applied as a full substitute of the fresh high-silica sand in typical procedures of preparing core sands used for making shaped samples for bending strength investigations, Rgu.

  15. Experience and evaluation of advanced on-line core monitoring system 'BEACON' at IKATA site

    International Nuclear Information System (INIS)

    Fujitsuka, Nobumichi; Tanouchi, Hideyuki; Imamura, Yasuhiro; Mizobuchil, Daisuke

    1997-01-01

    Shikoku Electric Power Company installed BEACON core monitoring system into IKATA unit 3 in May 1994. During its first cycle of core operation, various operational data were obtained including data of some anomalous reactor conditions introduced for the test objective of the plant start-up. This paper presents the evaluation of the BEACON system capability based on this experience. The system functions such as core monitoring and anomaly detection, prediction of future reactor conditions and increased efficiency of core management activities are discussed. Our future plan to utilize the system is also presented. (authors)

  16. Evaluation of nonaqueous processes for nuclear materials

    International Nuclear Information System (INIS)

    Musgrave, B.C.; Grens, J.Z.; Knighton, J.B.; Coops, M.S.

    1983-12-01

    A working group was assigned the task of evaluating the status of nonaqueous processes for nuclear materials and the prospects for successful deployment of these technologies in the future. In the initial evaluation, the study was narrowed to the pyrochemical/pyrometallurgical processes closely related to the processes used for purification of plutonium and its conversion to metal. The status of the chemistry and process hardware were reviewed and the development needs in both chemistry and process equipment technology were evaluated. Finally, the requirements were established for successful deployment of this technology. The status of the technology was evaluated along three lines: (1) first the current applications were examined for completeness, (2) an attempt was made to construct closed-cycle flow sheets for several proposed applications, (3) and finally the status of technical development and future development needs for general applications were reviewed. By using these three evaluations, three different perspectives were constructed that together present a clear picture of how complete the technical development of these processes are

  17. In vitro shear bond strength of Y-TZP ceramics to different core materials with the use of three primer/resin cement systems.

    Science.gov (United States)

    Al-Harbi, Fahad A; Ayad, Neveen M; Khan, Zahid A; Mahrous, Amr A; Morgano, Steven M

    2016-01-01

    Durability of the bond between different core materials and zirconia retainers is an important predictor of the success of a dental prosthesis. Nevertheless, because of its polycrystalline structure, zirconia cannot be etched and bonded to a conventional resin cement. The purpose of this in vitro study was to compare the effects of 3 metal primer/resin cement systems on the shear bond strength (SBS) of 3 core materials bonded to yttria-stabilized tetragonal zirconia polycrystalline (Y-TZP) ceramic retainers. Zirconia ceramic (Cercon) disks (5×3 mm) were airborne-particle abraded, rinsed, and air-dried. Disk-shaped core specimens (7×7 mm) that were prepared of composite resin, Ni-Cr, and zirconia were bonded to the zirconia ceramic disks by using one of 3 metal primer/cement systems: (Z-Prime Plus/BisCem, Zirconia Primer/Multilink Automix, or Clearfil Ceramic Primer/Clearfil SA). SBS was tested in a universal testing machine. Stereomicroscopy was used to evaluate the failure mode of debonded specimens. Data were analyzed using 2-way ANOVA and post hoc analysis using the Scheffe procedure (α=.05). Clearfil SA/Clearfil Ceramic Primer system with an Ni-Cr core yielded the highest SBS value (19.03 MPa), whereas the lowest SBS value was obtained when Multilink Automix/Zirconia Primer system was used with the zirconia core group (4.09 MPa). Differences in mean SBS values among the cement/primer groups were statistically significant, except for Clearfil SA and BisCem with both composite resin and zirconia cores. Differences in mean SBS values among the core subgroups were not statistically significant, except for zirconia core with BisCem, Multilink, and Clearfil SA. The predominant failure mode was adhesive, except for Clearfil SA and BisCem luting agents with composite resin cores, which displayed cohesive failure, and Multilink Automix with a composite resin, core as well as Clearfil SA with Ni-Cr cores, where the debonded specimens of each group displayed a mixed

  18. Evaluation of full MOX core capability for a 900 MWe PWR

    International Nuclear Information System (INIS)

    Joo, Hyung-Kook; Kim, Young-Jin; Jung, Hyung-Guk; Kim, Young-Il; Sohn, Dong-Seong

    1996-01-01

    Full MOX capability of a PWR core with 900 MWe capacity has been evaluated in view of plutonium consumption and design feasibility as an effective means for spent fuel management. Three full MOX cores have been conceptually designed; for annual cycle, for 18-month cycle, and for 18-month cycle with high moderation lattice. Fissile and total plutonium quantities at discharge are significantly reduced to 60% and 70% respectively of initial value for standard full MOX cores. It is estimated that one full MOX core demands about 1 tonne of plutonium per year to be reloaded, which is equivalent to reprocessing of spent nuclear fuels discharged from five nuclear reactors operating with uranium fuels. With low-leakage loading scheme, a full MOX core with either annual or 18-month cycle can be designed satisfactorily by installing additional rod cluster control system and modifying soluble boron system. Overall high moderation lattice case promises better core nuclear characteristics. (author)

  19. Basic data generation and pressure loss coefficient evaluation for HANARO core thermal-hydraulic analyses

    International Nuclear Information System (INIS)

    Chae, Hee Taek; Lee, Kye Hong

    1999-06-01

    MATRA-h, a HANARO subchannel analysis computer code, is used to evaluate thermal margin of the HANARO fuel. It's capability includes the assessments of CHF, ONB margin, and fuel temperature. In this report, basic input data and core design parameters required to perform the subchannel analysis with MATRA-h code are collected. These data include the subchannel geometric data, thermal-hydraulic correlations, empirical constants and material properties. The friction and form loss coefficients of the fuel assemblies were determined based on the results of the pressure drop test. At the same time, different form loss coefficients at the end plates and spacers are evaluated for various subchannels. The adequate correlations are applied to the evaluation of the form loss coefficients for various subchannels, which are corrected by measured values in order to have a same pressure drop at each flow channel. These basic input data and design parameters described in this report will be applied usefully to evaluate the thermal margin of the HANARO fuel. (author). 11 refs., 13 tabs., 11 figs

  20. Fracture Resistance of Endodontically Treated Teeth Restored with Biodentine, Resin Modified GIC and Hybrid Composite Resin as a Core Material.

    Science.gov (United States)

    Subash, Dayalan; Shoba, Krishnamma; Aman, Shibu; Bharkavi, Srinivasan Kumar Indu; Nimmi, Vijayan; Abhilash, Radhakrishnan

    2017-09-01

    The restoration of a severely damaged tooth usually needs a post and core as a part of treatment procedure to provide a corono - radicular stabilization. Biodentine is a class of dental material which possess high mechanical properties with excellent biocompatibility and bioactive behaviour. The sealing ability coupled with optimum physical properties could make Biodentine an excellent option as a core material. The aim of the study was to determine the fracture resistance of Biodentine as a core material in comparison with resin modified glass ionomer and composite resin. Freshly extracted 30 human permanent maxillary central incisors were selected. After endodontic treatment followed by post space preparation and luting of Glass fibre post (Reforpost, Angelus), the samples were divided in to three groups based on the type of core material. The core build-up used in Group I was Biodentine (Septodont, France), Group II was Resin-Modified Glass Ionomer Cement (GC, Japan) and Group III was Hybrid Composite Resin (TeEconom plus, Ivoclar vivadent). The specimens were subjected to fracture toughness using Universal testing machine (1474, Zwick/Roell, Germany) and results were compared using One-way analysis of variance with Tukey's Post hoc test. The results showed that there was significant difference between groups in terms of fracture load. Also, composite resin exhibited highest mean fracture load (1039.9 N), whereas teeth restored with Biodentine demonstrated the lowest mean fracture load (176.66 N). Resin modified glass ionomer exhibited intermediate fracture load (612.07 N). The primary mode of failure in Group I and Group II was favourable (100%) while unfavourable fracture was seen in Group III (30%). Biodentine, does not satisfy the requirements to be used as an ideal core material. The uses of RMGIC's as a core build-up material should be limited to non-stress bearing areas. Composite resin is still the best core build-up material owing to its high fracture

  1. Color stability evaluation of aesthetic restorative materials

    Directory of Open Access Journals (Sweden)

    Adriana Postiglione Bührer Samra

    2008-09-01

    Full Text Available Color match is one of the most important characteristics of aesthetic restorative materials. Maintenance of color throughout the functional lifetime of restorations is important for the durability of treatment. This characteristic is not constant among dental materials. The purpose of this research was to assess the color stability of five aesthetic restorative materials when immersed in a coffee solution. Seventy-one 17 mm x 1 mm specimens, divided into five groups, were made using one direct composite resin (Tetric Ceram®, Ivoclar/Vivadent - G1, three indirect composite resins (Targis, Ivoclar/Vivadent - G2; Resilab Master, Wilcos - G3; belleGlassTM HP, Kerr - G4 and one porcelain (IPS Empress® 2, Ivoclar/Vivadent - G5. The specimens were immersed in a coffee staining media for 15 days and stored under a controlled temperature of 37°C ± 1°C in the dark. The evaluations were made after 1, 7 and 15 days by means of reflectance spectrophotometry. The data was submitted to two-way ANOVA (p < 0.005 and post hoc tests. Statistical difference was observed between G1 / G3 and the other groups; G2 / G4 and the other groups; and G5 and all the other groups. It was concluded that G1 and G3 showed significantly higher discoloration than the other groups. G2 and G4 showed intermediary pigmentation, while G5 showed the smallest changes.

  2. Evaluation of powder metallurgy superalloy disk materials

    Science.gov (United States)

    Evans, D. J.

    1975-01-01

    A program was conducted to develop nickel-base superalloy disk material using prealloyed powder metallurgy techniques. The program included fabrication of test specimens and subscale turbine disks from four different prealloyed powders (NASA-TRW-VIA, AF2-1DA, Mar-M-432 and MERL 80). Based on evaluation of these specimens and disks, two alloys (AF2-1DA and Mar-M-432) were selected for scale-up evaluation. Using fabricating experience gained in the subscale turbine disk effort, test specimens and full scale turbine disks were formed from the selected alloys. These specimens and disks were then subjected to a rigorous test program to evaluate their physical properties and determine their suitability for use in advanced performance turbine engines. A major objective of the program was to develop processes which would yield alloy properties that would be repeatable in producing jet engine disks from the same powder metallurgy alloys. The feasibility of manufacturing full scale gas turbine engine disks by thermomechanical processing of pre-alloyed metal powders was demonstrated. AF2-1DA was shown to possess tensile and creep-rupture properties in excess of those of Astroloy, one of the highest temperature capability disk alloys now in production. It was determined that metallographic evaluation after post-HIP elevated temperature exposure should be used to verify the effectiveness of consolidation of hot isostatically pressed billets.

  3. Repairing rabbit radial defects by combining bone marrow stroma stem cells with bone scaffold material comprising a core-cladding structure.

    Science.gov (United States)

    Wu, H; Liu, G H; Wu, Q; Yu, B

    2015-10-05

    We prepared a bone scaffold material comprising a PLGA/β-TCP core and a Type I collagen cladding, and recombined it with bone marrow stroma stem cells (BMSCs) to evaluate its potential for use in bone tissue engineering by in vivo and in vitro experiments. PLGA/β-TCP without a cladding was used for comparison. The adherence rate of the BMSCs to the scaffold was determined by cell counting. Cell proliferation rate was determined by the 3-(4,5-dimethylthiazol-2-yl)-2,5-diphenyltetrazolium bromide method. The osteogenic capability was evaluated by alkaline phosphatase activity. The scaffold materials were recombined with the BMSCs and implanted into a large segmental rabbit radial defect model to evaluate defect repair. Osteogenesis was assessed in the scaffold materials by histological and double immunofluorescence labeling, etc. The adherence number, proliferation number, and alkaline phosphatase expression of the cells on the bone scaffold material with core-cladding structure were significantly higher than the corresponding values in the PLGA/β-TCP composite scaffold material (P structure completely degraded at the bone defect site and bone formation was completed. The rabbit large sentimental radial defect was successfully repaired. The degradation and osteogenesis rates matched well. The bone scaffold with core-cladding structure exhibited better osteogenic activity and capacity to repair a large segmental bone defect compared to the PLGA/β-TCP composite scaffold. The bone scaffold with core-cladding structure has excellent physical properties and biocompatibility. It is an ideal scaffold material for bone tissue engineering.

  4. Core Self-Evaluation and Burnout among Nurses: The Mediating Role of Coping Styles

    OpenAIRE

    Li, Xiaofei; Guan, Lili; Chang, Hui; Zhang, Bo

    2014-01-01

    OBJECTIVES: This study aimed to determine the potential association between core self-evaluation and the burnout syndrome among Chinese nurses, and the mediating role of coping styles in this relationship. METHODS: A cross-sectional survey was conducted in Shenyang, China, from May to July, 2013. A questionnaire which consisted of the Maslach Burnout Inventory-General Survey (MBI-GS), the Core Self-Evaluation Scale (CSE), and the Simplified Coping Style Questionnaire (CSQ), was completed by a...

  5. Analysis of the thermal response of a BWR Mark-I containment shell to direct contact by molten core materials

    International Nuclear Information System (INIS)

    Kress, T.S.; Cleveland, J.C.

    1988-01-01

    This study was undertaken to evaluate the thermal response of a BWR Mark-I containment shell in the event of an accident severe enough for molten core materials to fall into the cavity beneath the rector vessel and eventually come into direct contact with the shell. An existing ORNL three-dimensional transient heat transport computer code, HEATING-6, was used for a specific 2-D case (and variations) for which representative melt/shell boundary conditions required as input were available from other studies. In addition to the use of HEATING-6, a simplified analytical steady-state correlation was developed and given the name BWR Liner Analysis Program (BWRLAP). BWRLAP was ''benchmarked'' by comparison with HEATING-6 and was then used to make a number of parametric calculations to investigate the sensitivities of the results to the inputs. 5 refs., 11 figs., 2 tabs

  6. Development of the evaluation methods in reactor safety analyses and core characteristics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    In order to support the safety reviews by NRA on reactor safety design including the phenomena with multiple failures, the computer codes are developed and the safety evaluations with analyses are performed in the areas of thermal hydraulics and core characteristics evaluation. In the code preparation of safety analyses, the TRACE and RELAP5 code were prepared to conduct the safety analyses of LOCA and beyond design basis accidents with multiple failures. In the core physics code preparation, the functions of sensitivity and uncertainty analysis were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

  7. Research on the Core Competitive Power Elements Evaluation System of Green Hotel

    Directory of Open Access Journals (Sweden)

    Hui Liang

    2013-12-01

    Full Text Available Green hotel is a new type of hospitality industry development model based on the concept of circular economy and sustainable development. This paper makes an analysis and evaluation of the elements of green hotel core competence, on this basis, constructs the Green Hotel core competitive evaluation index system.The construction of the system is conducive to understand the green hotel’s own competitive advantage objectively, and explore ways to enhance its core competitiveness, providing objective basis for sustainable development of China's Hotel industry.

  8. Dependences of optical properties of spherical two-layered nanoparticles on parameters of gold core and material shell

    International Nuclear Information System (INIS)

    Pustovalov, V.K.; Astafyeva, L.G.; Zharov, V.P.

    2013-01-01

    Modeling of nonlinear dependences of optical properties of spherical two-layered gold core and some material shell nanoparticles (NPs) placed in water on parameters of core and shell was carried out on the basis of the extended Mie theory. Efficiency cross-sections of absorption, scattering and extinction of radiation with wavelength 532 nm by core–shell NPs in the ranges of core radii r 00 =5–40 nm and of relative NP radii r 1 /r 00 =1–8 were calculated (r 1 —radius of two-layered nanoparticle). Shell materials were used with optical indexes in the ranges of refraction n 1 =0.2–1.5 and absorption k 1 =0–3.5 for the presentation of optical properties of wide classes of shell materials (including dielectrics, metals, polymers, vapor shell around gold core). Results show nonlinear dependences of optical properties of two-layered NPs on optical indexes of shell material, core r 00 and relative NP r 1 /r 00 radii. Regions with sharp decrease and increase of absorption, scattering and extinction efficiency cross-sections with changing of core and shell parameters were investigated. These dependences should be taken into account for applications of two-layered NPs in laser nanomedicine and optical diagnostics of tissues. The results can be used for experimental investigation of shell formation on NP core and optical determination of geometrical parameters of core and shell of two-layered NPs. -- Highlights: • Absorption, scattering and extinction of two-layered nanoparticles are studied. • Shell materials change in wide regions of materials (metals, dielectrics, vapor). • Effect of sharp decrease and increase of optical characteristics is established. • Explanation of sharp decreasing and increasing optical characteristics is presented

  9. Core Self-Evaluations and Individual Strategies of Coping with Unemployment among Displaced Spanish Workers.

    Science.gov (United States)

    Virkes, Tihana; Maslić Seršić, Darja; Lopez-Zafra, Esther

    2017-10-30

    Unemployment has negative but also positive effects on mental health and general well-being depending on which coping strategies the individual use. Our aim was to determine the contribution of core self-evaluations in explaining the coping strategies of job search and job devaluation, as well as to test the potential moderation effect of job search and mediation effect of job devaluation on the relationship between self core-evaluations and both positive and negative experience of unemployment. One hundred seventy-eight individuals who lost their jobs involuntarily for a longer period than one month completed a questionnaire while attending to employment office. Results show that there is a significant relation between core-self evaluations and job devaluation (.37**). Furthermore, core-self evaluations were positively related to positive experience of unemployment (r = .31; p unemployment (r = .60; p unemployment strategies (job devaluation; β = .26; p unemployment. But, individuals with a longer duration of the current period of unemployment and higher core self-evaluations had a more positive experience of unemployment, and job devaluation partially mediated this relation (SE = .002; p = .038). These results imply that programs interventions should include the improvement of core self-evaluations and the positive experience of unemployed people.

  10. The relationship between core self-evaluations, views of god, and intrinsic/extrinsic religious motivation.

    Science.gov (United States)

    Smither, James W; Walker, Alan G

    2015-04-01

    Core self-evaluations refer to a higher-order construct that subsumes four well-established traits in the personality literature: self-esteem, generalized self-efficacy, (low) neuroticism, and (internal) locus of control. Studies that have examined the relationship between various measures of religiosity and individual components of core self-evaluations show no clear pattern of relationships. The absence of a clear pattern may be due to the failure of most previous studies in this area to use theory to guide research. Therefore, theories related to core self-evaluations, religious motivation, and views of God were used to develop and test four hypotheses. 220 adults completed measures of four religious attitudes (intrinsic religious motivation, extrinsic religious motivation, viewing God as loving, and viewing God as punitive), general religiosity, and core self-evaluations, separated by 6 weeks (with the order of measures counterbalanced). Multivariate multiple regression, controlling for general religiosity, showed that core self-evaluations were positively related to viewing God as loving, negatively related to viewing God as punitive, and negatively related to extrinsic religious motivation. The hypothesis that core self-evaluations would be positively related to intrinsic religious motivation was not supported.

  11. National Undergraduate Medical Core Curriculum in Turkey: Evaluation of Residents

    Directory of Open Access Journals (Sweden)

    Işıl İrem Budakoğlu

    2014-03-01

    Full Text Available Background: There is very little information available on self-perceived competence levels of junior medical doctors with regard to definitions by the National Core Curriculum (NCC for Undergraduate Medical Education. Aims: This study aims to determine the perceived level of competence of residents during undergraduate medical education within the context of the NCC. Study Design: Descriptive study. Methods: The survey was conducted between February 2010 and December 2011; the study population comprised 450 residents. Of this group, 318 (71% participated in the study. Self-assessment questionnaires on competencies were distributed and residents were asked to assess their own competence in different domains by scoring them on a scale of 1 to 10. Results: Nearly half of the residents reported insufficient experience of putting clinical skills into practice when they graduated. In the theoretical part of NCC, the lowest competency score was reported for health-care administration, while the determination of level of chlorine in water, delivering babies, and conducting forensic examinations had the lowest perceived levels of competency in the clinical skills domain. Conclusion: Residents reported low levels of perceived competency in skills they rarely performed outside the university hospital. They were much more confident in skills they performed during their medical education.

  12. Evaluation of ex-core imaging appertures for STF

    International Nuclear Information System (INIS)

    Rhodes, E.A.

    1976-01-01

    The relative merits of some candidate apertures for use in monitoring material motion in reactor transient tests to be conducted at the proposed Safety Test Facility are investigated. It is found that a hodoscope has the greatest probability of success in meeting established requirements, but that a pinhole or pinhole-hodoscope could be advantageous under certain circumstances. A Fresnel zone plate-hodoscope is found to be less suitable, and the two-dimensional nonredundant pinhole arrays and Fresnel zone plates studied are found to be unsuitable. Although based in part on numerical calculations for fuel motion detection by neutron emission only, these conclusions are thought to be valid for material motion detection by γ-ray emission and flash x-radiography as well

  13. Fabrication of Silicon Nitride Dental Core Ceramics with Borosilicate Veneering material

    Science.gov (United States)

    Wananuruksawong, R.; Jinawath, S.; Padipatvuthikul, P.; Wasanapiarnpong, T.

    2011-10-01

    Silicon nitride (Si3N4) ceramic is a great candidate for clinical applications due to its high fracture toughness, strength, hardness and bio-inertness. This study has focused on the Si3N4 ceramic as a dental core material. The white Si3N4 was prepared by pressureless sintering at relative low sintering temperature of 1650 °C in nitrogen atmosphere. The coefficient of thermal expansion (CTE) of Si3N4 ceramic is lower than that of Zirconia and Alumina ceramic which are popular in this field. The borosilicate glass veneering was employed due to its compatibility in thermal expansion. The sintered Si3N4 specimens represented the synthetic dental core were paintbrush coated by a veneer paste composed of borosilicate glass powder (tube furnace between 1000-1200°C. The veneered specimens fired at 1100°C for 15 mins show good bonding, smooth and glossy without defect and crazing. The veneer has thermal expansion coefficient as 3.98×10-6 °C-1, rather white and semi opaque, due to zirconia addition, the Vickers hardness as 4.0 GPa which is closely to the human teeth.

  14. The CERN antiproton target: hydrocode analysis of its core material dynamic response under proton beam impact

    CERN Document Server

    Martin, Claudio Torregrosa; Calviani, Marco; Muñoz-Cobo, José-Luis

    2016-01-01

    Antiprotons are produced at CERN by colliding a 26 GeV/c proton beam with a fixed target made of a 3 mm diameter, 55 mm length iridium core. The inherent characteristics of antiproton production involve extremely high energy depositions inside the target when impacted by each primary proton beam, making it one of the most dynamically demanding among high energy solid targets in the world, with a rise temperature above 2000 {\\deg}C after each pulse impact and successive dynamic pressure waves of the order of GPa's. An optimized redesign of the current target is foreseen for the next 20 years of operation. As a first step in the design procedure, this numerical study delves into the fundamental phenomena present in the target material core under proton pulse impact and subsequent pressure wave propagation by the use of hydrocodes. Three major phenomena have been identified, (i) the dominance of a high frequency radial wave which produces destructive compressive-to-tensile pressure response (ii) The existence of...

  15. Fabrication of Silicon Nitride Dental Core Ceramics with Borosilicate Veneering material

    International Nuclear Information System (INIS)

    Wananuruksawong, R; Jinawath, S; Wasanapiarnpong, T; Padipatvuthikul, P

    2011-01-01

    Silicon nitride (Si 3 N 4 ) ceramic is a great candidate for clinical applications due to its high fracture toughness, strength, hardness and bio-inertness. This study has focused on the Si 3 N 4 ceramic as a dental core material. The white Si 3 N 4 was prepared by pressureless sintering at relative low sintering temperature of 1650 deg. C in nitrogen atmosphere. The coefficient of thermal expansion (CTE) of Si 3 N 4 ceramic is lower than that of Zirconia and Alumina ceramic which are popular in this field. The borosilicate glass veneering was employed due to its compatibility in thermal expansion. The sintered Si 3 N 4 specimens represented the synthetic dental core were paintbrush coated by a veneer paste composed of borosilicate glass powder ( 2 O 3 - partial stabilized zirconia) and 30 wt% of polyvinyl alcohol (5 wt% solution). After coating the veneer on the Si 3 N 4 specimens, the firing was performed in electric tube furnace between 1000-1200 deg. C. The veneered specimens fired at 1100 deg. C for 15 mins show good bonding, smooth and glossy without defect and crazing. The veneer has thermal expansion coefficient as 3.98x10 -6 deg. C -1 , rather white and semi opaque, due to zirconia addition, the Vickers hardness as 4.0 GPa which is closely to the human teeth.

  16. CERN antiproton target: Hydrocode analysis of its core material dynamic response under proton beam impact

    Directory of Open Access Journals (Sweden)

    Claudio Torregrosa Martin

    2016-07-01

    Full Text Available Antiprotons are produced at CERN by colliding a 26  GeV/c proton beam with a fixed target made of a 3 mm diameter, 55 mm length iridium core. The inherent characteristics of antiproton production involve extremely high energy depositions inside the target when impacted by each primary proton beam, making it one of the most dynamically demanding among high energy solid targets in the world, with a rise temperature above 2000 °C after each pulse impact and successive dynamic pressure waves of the order of GPa’s. An optimized redesign of the current target is foreseen for the next 20 years of operation. As a first step in the design procedure, this numerical study delves into the fundamental phenomena present in the target material core under proton pulse impact and subsequent pressure wave propagation by the use of hydrocodes. Three major phenomena have been identified, (i the dominance of a high frequency radial wave which produces destructive compressive-to-tensile pressure response (ii The existence of end-of-pulse tensile waves and its relevance on the overall response (iii A reduction of 44% in tensile pressure could be obtained by the use of a high density tantalum cladding.

  17. Nuclear characteristics evaluation for Kyoto University Research Reactor with low-enriched uranium core

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Ken; Unesaki, Hironobu [Kyoto University Research Reactor Institute, Kumatori-cho Sennan-gun Osaka (Japan)

    2008-07-01

    A project to convert the fuel of Kyoto University Research Reactor (KUR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) is in progress as a part of RERTR program. Prior to the operation of LEU core, the nuclear characteristics of the core have been evaluated to confirm the safety operation. In the evaluation, nuclear parameters, such as the excess reactivity, shut down margin control rod worth, reactivity coefficients, were calculated, and they were compared with the safety limits. The results of evaluation show that the LEU core is able to satisfy the safety requirements for operation, i.e. all the parameters satisfy the safety limits. Consequently, it was confirmed that the LEU fuel core has the proper nuclear characteristics for the safety operation. (authors)

  18. Heat Storage Performance of the Prefabricated Hollow Core Concrete Deck Element with Integrated Microencapsulated Phase Change Material

    DEFF Research Database (Denmark)

    Pomianowski, Michal Zbigniew; Heiselberg, Per; Jensen, Rasmus Lund

    2012-01-01

    The paper presents the numerically calculated dynamic heat storage capacity of the prefabricated hollow core concrete deck element with and without microencapsulated phase change material (PCM). The reference deck is the ordinary deck made of standard concrete material and that is broadly used...

  19. Synthesis and evaluation of energetic materials

    Science.gov (United States)

    Santhosh, G.

    Over the years new generations of propellants and explosives are being developed. High performance and pollution prevention issues have become the subject of interest in recent years. Desired properties of these materials are a halogen-free, nitrogen and oxygen rich molecular composition with high density and a positive heat of formation. The dinitramide anion is a new oxy anion of nitrogen and forms salts with variety of metal, organic and inorganic cations. Particular interest is in ammonium dinitramide (ADN, NH4N(NO 2)2) which is a potentially useful energetic oxidizer. ADN is considered as one of the most promising substitutes for ammonium perchlorate (AP, NH4ClO4) in currently used composite propellants. It is unique among energetic materials in that it has no carbon or chlorine; its combustion products are not detrimental to the atmosphere. Unquestionable advantage of ADN over AP is the significant improvement in the performance of solid rocket motors by 5-15%. The present thesis is centered on the experimental results along with discussion of some of the most pertinent aspects related to the synthesis and characterization of few dinitramide salts. The chemistry, mechanism and kinetics of the formation of dinitramide salts by nitration of deactivated amines are investigated. The evaluation of the thermal and spectral properties along with the adsorption and thermal decomposition characteristics of the dinitramide salts are also explored in this thesis.

  20. Com-scan techniques for material evaluation

    International Nuclear Information System (INIS)

    Jayakumar, T.K.; Naik, A.D.

    1996-01-01

    Presently a variety of products at various stages of production are being tested using NDT methods for ensuring their quality. The conventional NDT methods such as RT, UT, PT, MPT along with acoustic emission techniques are often employed for the purpose. However, the ever increasing demands for a comprehensive quality assurance of products necessitates newer avenues in testing methods to overcome certain inadequacies of conventional testing. This paper proposes Compton back-scatter technique as an additional alternative NDT tool for various measurements. When the radiation strikes material a small percentage of incident radiation scatters back with reduced energy. The back-scattered radiation is picked up by a digital backscatter gauge and analysed. The paper discusses experimental work carried out at the laboratory consisting of parameter evaluation, source detector geometry, back-scatter response for material, area effects, thickness and blockage measurements. The paper briefly discusses on-stream measurements carried out with the above experimental gauge. The paper deals with selection and comparison of measurements with those of ultrasonics. It also discusses the advantages of the radiation back-scatter testing. The paper recommends Com-scan back-scatter technique as a supplementary tool along with conventional testing. (author)

  1. Materials evaluation for a transuranic processing facility

    International Nuclear Information System (INIS)

    Barker, S.A.; Schwenk, E.B.; Divine, J.R.

    1990-11-01

    The Westinghouse Hanford Company, with the assistance of the Pacific Northwest Laboratory, is developing a transuranium extraction process for preheating double-shell tank wastes at the Hanford Site to reduce the volume of transuranic waste being sent to a repository. The bench- scale transuranium extraction process development is reaching a stage where a pilot plant design has begun for the construction of a facility in the existing B Plant. Because of the potential corrosivity of neutralized cladding removal waste process streams, existing embedded piping alloys in B Plant are being evaluated and ''new'' alloys are being selected for the full-scale plant screening corrosion tests. Once the waste is acidified with HNO 3 , some of the process streams that are high in F - and low in Al and zr can produce corrosion rates exceeding 30,000 mil/yr in austenitic alloys. Initial results results are reported concerning the applicability of existing plant materials to withstand expected process solutions and conditions to help determine the feasibility of locating the plant at the selected facility. In addition, process changes are presented that should make the process solutions less corrosive to the existing materials. Experimental work confirms that Hastelloy B is unsatisfactory for the expected process solutions; type 304L, 347 and 309S stainless steels are satisfactory for service at room temperature and 60 degrees C, if process stream complexing is performed. Inconel 625 was satisfactory for all solutions. 17 refs., 5 figs., 8 tabs

  2. Design and evaluation of a multithreaded many-core architecture

    NARCIS (Netherlands)

    Lankamp, M.

    2015-01-01

    Performance improvements for microprocessors have traditionally been achieved by increasing their clock frequency. However, this technique has reached a point where further scaling is impractical. This thesis describes and evaluates a novel System-on-Chip architecture that focuses on exploiting all

  3. Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    Höhne, Thomas; Grahn, Alexander; Kliem, Sören; Rohde, Ulrich; Weiss, Frank-Peter

    2013-01-01

    Highlights: ► Detailed results of a numerical simulation of the insulation material transport to a PWR core are shown. ► The spacer grid is modeled as a strainer which completely retains the insulation material carried by coolant. ► The CFD calculations showed that the fibers at the upper spacer grid plane are not uniformly distributed. ► Furthermore the pressure loss does not exceed a critical limit. ► The PWR core coolablity can be guaranteed all the time during the transient. -- Abstract: In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the

  4. The results of the CCI-3 reactor material experiment investigating 2-D core-concrete interaction and debris coolability with a siliceous concrete crucible

    International Nuclear Information System (INIS)

    Farmer, M.T.; Basu, S.

    2006-01-01

    The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants and provide the technical basis for better containment designs for future plants. Despite years of international research, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion predicted by codes such as MELCOR, WECHSL, and COSACO. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test involved the interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined ablation depth was reached, the cavity was flooded to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a summary description of the test facility and an overview of test results

  5. Evaluation report on CCTF core-II reflood test C2-6 (Run 64)

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Iguchi, Tadashi; Sugimoto, Jun; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu.

    1985-03-01

    In order to evaluate the effect of the radial power profile on the system behavior and the core thermal hydraulic behavior during the reflood phase of a PWR LOCA, a test was performed using the Cylindrical Core Test Facility(CCTF) with the flat radial power profile. The test was conducted with the same total core power as that of the steep radial power test C2-5(Run 63). Through the comparisons of the results from these two tests, the following conclusions were obtained: (1) The radial power profile in the core has weak effect on the thermal hydraulic behavior in the primary system except the core. (2) Almost the same differential pressure was observed at various elevations in the periphery of the core regardless of different radial power profile. The result suggests that the core differential pressure is determined mainly by the total power and the total stored energy rather than by the local power and the local stored energy. (3) The test results support the single channel core model with the average power rod used in the reactor safety analysis codes such as REFLA-1DS, WREM for the evaluation of the overall system behavior. (4) In the steep radial power test, the heat transfer coefficient in the central(high power) region was higher than that in the peripheral(low power) region. The tendency was not explained by the estimation with the heat transfer correlation developed by Murao and Sugimoto assuming that the void fraction was uniform in a horizontal cross section. It is necessary to study more the dependency of core heat transfer on the radial power profile in the wide core. (author)

  6. Magnetization measurement of niobium for superconducting cavity material evaluation

    International Nuclear Information System (INIS)

    Wake, Masayoshi; Saito, Kenji.

    1995-05-01

    A series of magnetization measurements on niobium materials for superconducting cavities was performed, and the method was found to be very useful for material evaluation. The effects of annealing, chemical polishing and machining were clearly observed by this method. The material quality and the processing of the material can be properly evaluated by measuring the magnetization. An observation of the Q-disease effect indicates the possibility of using this method for the studies beyond material evaluation. (J.P.N)

  7. Interfacial characterization of ceramic core materials with veneering porcelain for all-ceramic bi-layered restorative systems.

    Science.gov (United States)

    Tagmatarchis, Alexander; Tripodakis, Aris-Petros; Filippatos, Gerasimos; Zinelis, Spiros; Eliades, George

    2014-01-01

    The aim of the study was to characterize the elemental distribution at the interface between all-ceramic core and veneering porcelain materials. Three groups of all-ceramic cores were selected: A) Glass-ceramics (Cergo, IPS Empress, IPS Empress 2, e-max Press, Finesse); B) Glass-infiltrated ceramics (Celay Alumina, Celay Zirconia) and C) Densely sintered ceramics (Cercon, Procera Alumina, ZirCAD, Noritake Zirconia). The cores were combined with compatible veneering porcelains and three flat square test specimens were produced for each system. The core-veneer interfaces were examined by scanning electron microscopy and energy dispersive x-ray microanalysis. The glass-ceramic systems showed interfacial zones reach in Si and O, with the presence of K, Ca, Al in core and Ca, Ce, Na, Mg or Al in veneer material, depending on the system tested. IPS Empress and IPS Empress 2 demonstrated distinct transitional phases at the core-veneer interface. In the glassinfiltrated systems, intermixing of core (Ce, La) with veneer (Na, Si) elements occurred, whereas an abrupt drop of the core-veneer elemental concentration was documented at the interfaces of all densely sintered ceramics. The results of the study provided no evidence of elemental interdiffusion at the core-veneer interfaces in densely sintered ceramics, which implies lack of primary chemical bonding. For the glass-containing systems (glassceramics and glass-infiltrated ceramics) interdiffusion of the glass-phase seems to play a critical role in establishing a primary bonding condition between ceramic core and veneering porcelain.

  8. ABOUT INDEX EVALUATION OF MATERIAL RESOURCE SUPPLIER SELECTION

    OpenAIRE

    V. A. Skochinskaya

    2008-01-01

    The paper analyzes existing methods for evaluation of material resource supplier selection. It shows advantages and shortcomings of the present evaluation systems. The necessity for application of an index evaluation is justified in the paper. The paper contains rating (index) evaluation for material resource supplier selection which is based on the application of quantitative (index) tool instead of an expert (numerical) evaluation

  9. Evaluation report on CCTF Core-II reflood test C2-9 (Run 68)

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Iguchi, Tadashi; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Sugimoto, Jun.

    1987-02-01

    In order to study the LPCI flow rate effect on the core cooling and system behavior, a test was performed with the LPCI flow rate of 0.025 m 3 /s, which corresponds to the flow rate in case of no pump failure in a PWR system. Through the comparisons of test results with those from the reference test with the LPCI flow rate of 0.011 m 3 /s, the following conclusions were obtained: (1) The higher LPCI flow rate resulted in the worse core-cooling in these two tests. The test results show that the lower LPCI flow rate is not necessarily a conservative assumption for the evaluation of the core cooling during the reflood phase of a PWR LOCA. (2) The worse core-cooling in the high LPCI flow rate test is attributed to the lower core-pressure than in the reference test. It is found that the lower core-pressure results from the lower pressure drop through the broken cold leg. (3) It is expected that the current evaluation model(EM) code is still conservative because it usually predicts the low pressure drop through the broken cold leg. (4) The flow oscillation in the cold leg was not significant even in the high LPCI flow rate test before the whole core quench. (author)

  10. Statistical analysis on hollow and core-shell structured vanadium oxide microspheres as cathode materials for Lithium ion batteries

    Directory of Open Access Journals (Sweden)

    Xing Liang

    2018-06-01

    Full Text Available In this data, the statistical analyses of vanadium oxide microspheres cathode materials are presented for the research article entitled “Statistical analyses on hollow and core-shell structured vanadium oxides microspheres as cathode materials for Lithium ion batteries” (Liang et al., 2017 [1]. This article shows the statistical analyses on N2 adsorption-desorption isotherm and morphology vanadium oxide microspheres as cathode materials for LIBs. Keywords: Adsorption-desorption isotherm, Pore size distribution, SEM images, TEM images

  11. Chemical and physical analysis of core materials for advanced high temperature reactors with process heat applications

    International Nuclear Information System (INIS)

    Nickel, H.

    1985-08-01

    Various chemical and physical methods for the analysis of structural materials have been developed in the research programmes for advanced high temperature reactors. These methods are discussed using as examples the structural materials of the reactor core - the fuel elements consisting of coated particles in a graphite matrix and the structural graphite. Emphasis is given to the methods of chemical analysis. The composition of fuel kernels is investigated using chemical analysis methods to determine the heavy metals content (uranium, plutonium, thorium and metallic impurity elements) and the amount of non-metallic constituents. The properties of the pyrocarbon and silicon carbide coatings of fuel elements are investigated using specially developed physiochemical methods. Regarding the irradiation behaviour of coated particles and fuel elements, methods have been developed for examining specimens in hot cells following exposures under reactor operating conditions, to supplement the measurements of in-reactor performance. For the structural graphite, the determination of impurities is important because certain impurities may cause pitting corrosion during irradiation. The localized analysis of very low impurity concentrations is carried out using spectrochemical d.c. arc excitation, local laser and inductively coupled plasma methods. (orig.)

  12. Building waste management core indicators through Spatial Material Flow Analysis: net recovery and transport intensity indexes.

    Science.gov (United States)

    Font Vivanco, David; Puig Ventosa, Ignasi; Gabarrell Durany, Xavier

    2012-12-01

    In this paper, the material and spatial characterization of the flows within a municipal solid waste (MSW) management system are combined through a Network-Based Spatial Material Flow Analysis. Using this information, two core indicators are developed for the bio-waste fraction, the Net Recovery Index (NRI) and the Transport Intensity Index (TII), which are aimed at assessing progress towards policy-related sustainable MSW management strategies and objectives. The NRI approaches the capacity of a MSW management system for converting waste into resources through a systematic metabolic approach, whereas the TII addresses efficiency in terms of the transport requirements to manage a specific waste flow throughout the entire MSW management life cycle. Therefore, both indicators could be useful in assessing key MSW management policy strategies, such as the consecution of higher recycling levels (sustainability principle) or the minimization of transport by locating treatment facilities closer to generation sources (proximity principle). To apply this methodological approach, the bio-waste management system of the region of Catalonia (Spain) has been chosen as a case study. Results show the adequacy of both indicators for identifying those points within the system with higher capacity to compromise its environmental, economic and social performance and therefore establishing clear targets for policy prioritization. Moreover, this methodological approach permits scenario building, which could be useful in assessing the outcomes of hypothetical scenarios, thus proving its adequacy for strategic planning. Copyright © 2012 Elsevier Ltd. All rights reserved.

  13. Evaluation of Wavelet-based Core Inflation Measures: Evidence from Peru

    OpenAIRE

    Erick Lahura; Marco Vega

    2011-01-01

    Under inflation targeting and other related monetary policy regimes, the identication of non-transitory inflation and forecasts about future inflation constitute key ingredients for monetary policy decisions. In practice, central banks perform these tasks using so-called "core inflation measures". In this paper we construct alternative core inflation measures using wavelet functions and multiresolution analysis (MRA), and then evaluate their relevance for monetary policy. The construction of ...

  14. Family Resources and Flourishing at Work: The Role of Core Self-Evaluations

    Directory of Open Access Journals (Sweden)

    Larissa Maria David Gabardo-Martins

    Full Text Available Abstract: According to the Work-Home Resources Model, contextual family resources increase personal resources, which, in turn, improve work outcomes. The present study investigated the direct effects of two contextual family resources (work- family enrichment and perceived social support from family and one personal resource (core self-evaluations on a work outcome (flourishing at work. The mediational role of core self-evaluations in these relationships was also investigated. The sample was composed of 519 Brazilian psychologists of both sexes. The Structural Equation Modeling showed that the contextual family resources and the personal resource predicted flourishing at work and that core self-evaluations mediated the relationships between contextual resources and flourishing at work. It was concluded that the acquisition of resources within the family and the positive evaluation of one’s own life can promote flourishing at work.

  15. Job characteristics, core self-evaluations, and job satisfaction: what's age got to do with it?

    Science.gov (United States)

    Besen, Elyssa; Matz-Costa, Christina; Brown, Melissa; Smyer, Michael A; Pitt-Catsouphes, Martha

    2013-01-01

    There is a well-established relationship between age and job satisfaction. To date, there is little research about how many well-known predictors of job satisfaction, specifically job characteristics and core self-evaluations, may vary with age. Using a multi-worksite sample of 1,873 employed adults aged 17 to 81, this study evaluated the extent to which several job characteristics and core self-evaluations varied in their relationships with job satisfaction for workers of different ages. Findings suggest that the positive relationships between job satisfaction and skill variety, autonomy, and friendship weaken as employee age increases, while the positive relationships between job satisfaction and dealing with others, task identity, task significance, feedback, and core self-evaluations did not vary with age. The findings extend previous research by examining how the factors important for job satisfaction vary for employees of different ages.

  16. The development of learning materials based on core model to improve students’ learning outcomes in topic of Chemical Bonding

    Science.gov (United States)

    Avianti, R.; Suyatno; Sugiarto, B.

    2018-04-01

    This study aims to create an appropriate learning material based on CORE (Connecting, Organizing, Reflecting, Extending) model to improve students’ learning achievement in Chemical Bonding Topic. This study used 4-D models as research design and one group pretest-posttest as design of the material treatment. The subject of the study was teaching materials based on CORE model, conducted on 30 students of Science class grade 10. The collecting data process involved some techniques such as validation, observation, test, and questionnaire. The findings were that: (1) all the contents were valid, (2) the practicality and the effectiveness of all the contents were good. The conclusion of this research was that the CORE model is appropriate to improve students’ learning outcomes for studying Chemical Bonding.

  17. Job Characteristics, Core Self-Evaluations, and Job Satisfaction: What's Age Got to Do with It?

    Science.gov (United States)

    Besen, Elyssa; Matz-Costa, Christina; Brown, Melissa; Smyer, Michael A.; Pitt-Catsouphes, Martha

    2013-01-01

    There is a well-established relationship between age and job satisfaction. To date, there is little research about how many well-known predictors of job satisfaction, specifically job characteristics and core self-evaluations, may vary with age. Using a multi-worksite sample of 1,873 employed adults aged 17 to 81, this study evaluated the extent…

  18. Critical evaluation of molybdenum and its alloys for use in space reactor core heat pipes

    International Nuclear Information System (INIS)

    Lundberg, L.B.

    1981-01-01

    The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is examined, and the advantages and disadvantages of this material are brought into focus. Even though pure molybdenum heat pipes have been built and tested, this metal's high ductile-brittle transition temperature and modest creep strength place significant design restrictions on a core heat pipe made from it. Molybdenum alloys are examined with regard to their promise as potential replacements for pure molybdenum. The properties of TZM and molybdenum-rhenium alloys are examined, and it appears that Mo-Re alloys with 10 to 15 wt % rhenium offer the most advantage as an alternative to pure molybdenum in space reactor core heat pipes

  19. Fabrication of Silicon Nitride Dental Core Ceramics with Borosilicate Veneering material

    Energy Technology Data Exchange (ETDEWEB)

    Wananuruksawong, R; Jinawath, S; Wasanapiarnpong, T [Research Unit of Advanced Ceramic, Department of Materials Science, Faculty of Science, Chulalongkorn University, Bangkok (Thailand); Padipatvuthikul, P, E-mail: raayaa_chula@hotmail.com [Faculty of Dentistry, Srinakharinwirot University, Bangkok (Thailand)

    2011-10-29

    Silicon nitride (Si{sub 3}N{sub 4}) ceramic is a great candidate for clinical applications due to its high fracture toughness, strength, hardness and bio-inertness. This study has focused on the Si{sub 3}N{sub 4} ceramic as a dental core material. The white Si{sub 3}N{sub 4} was prepared by pressureless sintering at relative low sintering temperature of 1650 deg. C in nitrogen atmosphere. The coefficient of thermal expansion (CTE) of Si{sub 3}N{sub 4} ceramic is lower than that of Zirconia and Alumina ceramic which are popular in this field. The borosilicate glass veneering was employed due to its compatibility in thermal expansion. The sintered Si{sub 3}N{sub 4} specimens represented the synthetic dental core were paintbrush coated by a veneer paste composed of borosilicate glass powder (<150 micrometer, Pyrex) with 5 wt% of zirconia powder (3 wt% Y{sub 2}O{sub 3} - partial stabilized zirconia) and 30 wt% of polyvinyl alcohol (5 wt% solution). After coating the veneer on the Si{sub 3}N{sub 4} specimens, the firing was performed in electric tube furnace between 1000-1200 deg. C. The veneered specimens fired at 1100 deg. C for 15 mins show good bonding, smooth and glossy without defect and crazing. The veneer has thermal expansion coefficient as 3.98x10{sup -6} deg. C{sup -1}, rather white and semi opaque, due to zirconia addition, the Vickers hardness as 4.0 GPa which is closely to the human teeth.

  20. Heat and fission product transport in molten core material pool with crust

    International Nuclear Information System (INIS)

    Yun, J.I.; Suh, K.Y.; Kang, C.S.

    2005-01-01

    Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the reactor vessel during a severe accident. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool is estimated by product of the mass concentration and energy conversion factor of each fission product. Twenty-nine elements are chosen and classified by their chemical properties to calculate heat generation rate in the pool. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis is performed for heat and fission product transport in a molten core material pool during the Three Mile Island Unit 2 (TMI-2) accident. The pool is assumed to be a partially filled hemisphere, whose change in geometry is neglected during the numerical calculation. Calculated results indicate that the peak temperature in the molten pool is significantly lowered, since a substantial amount of the volatile fission products is released from the molten pool during progression of the accident. The results may directly be applied to the existing severe accident analysis codes to more mechanistically determine the thermal load to the reactor vessel lower head during the in-vessel retention

  1. Report on material and fabrication tests of the KUHFR core vessel

    International Nuclear Information System (INIS)

    Yoshida, H.; Kozuka, T.; Achiwa, N.; Mitani, S.; Kawano, S.; Araki, Y.; Shibata, T.

    1983-01-01

    For the material of the cylindrical reactor core vessel of the Kyoto University High Flux Reactor (KUHFR), A6061 alloy is selected because the aged state of the alloy is known to show the highest resistance against void swelling due to high-dose irradiation. The fabrication possibility of the large-scale tubes is also tested because the sizes (40 cmdiameter and 43 cmdiameter x 960 cm with a thickness of 10 mm for the inner- and outer-tubes, respectively) are just over the largest limit of the conventional factory fabrication. The results are summarized as follows. (1) From an ingot of A6061 alloy a raw inner-tube is hot-extruded by the 3,000 ton press machine. The shape of the extruded tubes is effectively corrected by stretch forming and other special methods. (2) The real scale tubes are heat-treated under the various conditions (T1, T4 and T6) and their size changes are measured just after the every heat-treatment. (3) The hydropressure for a pipe prepared by welding from an aged-tube shows a fairly uniform strain distribution and the breaking initiation at the reasonable pressure in the welded part. (4) Each of the welded specimens prepared using three kinds of welding rods shows sufficient strength in both of bending and tensile test for the JIS standard. Their microstructures correspond to the result of the mechanical tests for each welded specimen. The confidence for the fabrication possibility of the real core vessel has been given through the present tests. (author)

  2. Permeability analysis of Asbuton material used as core layers of water resistance in the body of dam

    Science.gov (United States)

    Rahim, H.; Tjaronge, M. W.; Thaha, A.; Djamaluddin, R.

    2017-11-01

    In order to increase consumption of the local materials and national products, large reserves of Asbuton material about 662.960 million tons in the Buton Islands became an alternative as a waterproof core layer in the body of dam. The Asbuton material was used in this research is Lawele Granular Asphalt (LGA). This study was an experimental study conducted in the laboratory by conducting density testing (content weight) and permeability on Asbuton material. Testing of the Asbuton material used Falling Head method to find out the permeability value of Asbuton material. The data of test result to be analyzed are the relation between compaction energy and density value also relation between density value and permeability value of Asbuton material. The result shows that increases the number of blow apply to the Asbuton material at each layer will increase the density of the Asbuton material. The density value of Asbuton material that satisfies the requirements for use as an impermeable core layer in the dam body is 1.53 grams/cm3. The increase the density value (the weight of the contents) of the Asbuton material will reduce its permeability value of the Asbuton material.

  3. HLW Flexible jumper materials compatibility evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Skidmore, T. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-13

    H-Tank Farm Engineering tasked SRNL/Materials Science & Technology (MS&T) to evaluate the compatibility of Goodyear Viper® chemical transfer hose with HLW solutions. The hose is proposed as a flexible Safety Class jumper for up to six months service. SRNL/MS&T performed various tests to evaluate the effects of radiation, high pH chemistry and elevated temperature on the hose, particularly the inner liner. Test results suggest an upper dose limit of 50 Mrad for the hose. Room temperature burst pressure values at 50 Mrad are estimated at 600- 800 psi, providing a safety factor of 4.0-5.3X over the anticipated operating pressure of 150 psi and a safety factor of 3.0-4.0X over the working pressure of the hose (200 psi), independent of temperature effects. Radiation effects are minimal at doses less than 10 Mrad. Doses greater than 50 Mrad may be allowed, depending on operating conditions and required safety factors, but cannot be recommended at this time. At 250 Mrad, burst pressure values are reduced to the hose working pressure. At 300 Mrad, burst pressures are below 150 psi. At a bounding continuous dose rate of 57,870 rad/hr, the 50 Mrad dose limit is reached within 1.2 months. Actual dose rates may be lower, particularly during non-transfer periods. Refined dose calculations are therefore recommended to justify longer service. This report details the tests performed and interpretation of the results. Recommendations for shelf-life/storage, component quality verification, and post-service examination are provided.

  4. Evaluation of Core Bypass Flow in the Prismatic VHTR with a Multi-block Experiment

    International Nuclear Information System (INIS)

    Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl; Kim, Min Hwan

    2010-01-01

    The core of Prismatic Modular Reactor (PMR) consists of assemblies of hexagonal graphite fuel and reflector elements. The core bypass flow of Very High Temperature Reactor (VHTR) is defined as the core flow that does not pass through the coolant channels but passes through the bypass gap between fuel elements. The increase in bypass flow makes the decrease in effective coolant flow. Since the core bypass flow has a negative impact on safety and efficiency of VHTR, core bypass phenomena have to be investigated to improve the core thermal margin of VHTR. For this purpose, the international project, I-NERI project, has been carried out since 2008. I-NERI project is collaborative project that KAERI and SNU of Korea side and INL, ANL and TAMU of U.S side are involved. In order to evaluate the core bypass flow, the multicolumn and multi-layer experimental facility is designed by SNU. In this experiment, the effect of cross-flow and local variation of bypass gap on the bypass flow distribution is investigated. Furthermore, the experimental data will be used for validation of CFD code or thermal hydraulic analysis codes such as GAMMA or GAS-NET

  5. Evaluation report on CCTF Core-II reflood test C2-4 (Run 62)

    International Nuclear Information System (INIS)

    Okubo, Tsutomu; Iguchi, Tadashi; Sugimoto, Jun; Akimoto, Hajime; Murao, Yoshio; Okabe, Kazuharu.

    1985-03-01

    This report presents a data evaluation of the CCTF Core-II test C2-4 (Run 62), which was conducted on May 12, 1983. This test was conducted to investigate the reproducibility of tests in the CCTF Core-II test series. Therefore, the initial and boundary conditions of the present test were determined to be the same as those for the previously performed base case test (Test C2-SH1). Comparing the data of the present test with those of Test C2-SH1, the following results are obtained. (1) The initial and boundary conditions for the two tests were nearly identical except the temperature of the core barrel and the lower plenum fluid. The difference in the latter is considered to result in the difference in the core inlet subcooling of about 6 K at most. (2) The system behavior was almost identical. (3) The core cooling behavior was also nearly identical except a little difference in the rod surface temperature in the upper part of the high power region. (4) Taking account that the difference mentioned above in item (3) is small and can be explained qualitatively to be caused by the difference in the core inlet subcooling mentioned above in item (1), it is considered practically that there is the reproducibility of the thermo-hydrodynamic behavior in the CCTF Core-II tests. (author)

  6. Fabrication of Fe3O4@CuO core-shell from MOF based materials and its antibacterial activity

    International Nuclear Information System (INIS)

    Rajabi, S.K.; Sohrabnezhad, Sh.; Ghafourian, S.

    2016-01-01

    Magnetic Fe 3 O 4 @CuO nanocomposite with a core/shell structure was successfully synthesized via direct calcinations of magnetic Fe 3 O 4 @HKUST-1 in air atmosphere. The morphology, structure, magnetic and porous properties of the as-synthesized nano composites were characterized by using scanning electron microscope (SEM), transmission electron microscopy (TEM), powder X-ray diffraction (PXRD), and vibration sample magnetometer (VSM). The results showed that the nanocomposite material included a Fe 3 O 4 core and a CuO shell. The Fe 3 O 4 @CuO core-shell can be separated easily from the medium by a small magnet. The antibacterial activity of Fe 3 O 4 -CuO core-shell was investigated against gram-positive and gram-negative bacteria. A new mechanism was proposed for inactivation of bacteria over the prepared sample. It was demonstrated that the core-shell exhibit recyclable antibacterial activity, acting as an ideal long-acting antibacterial agent. - Graphical abstract: Fe 3 O 4 @CuO core-shell release of copper ions. These Cu 2+ ions were responsible for the exhibited antibacterial activity. - Highlights: • The Fe 3 O 4 @CuO core-shell was prepared by MOF method. • This is the first study of antibacterial activity of core-shell consist of CuO and Fe 3 O 4 . • The core-shell can be reused effectively. • Core-shell was separated from the reaction solution by external magnetic field.

  7. Evaluating core technology capacity based on an improved catastrophe progression method: the case of automotive industry

    Science.gov (United States)

    Zhao, Shijia; Liu, Zongwei; Wang, Yue; Zhao, Fuquan

    2017-01-01

    Subjectivity usually causes large fluctuations in evaluation results. Many scholars attempt to establish new mathematical methods to make evaluation results consistent with actual objective situations. An improved catastrophe progression method (ICPM) is constructed to overcome the defects of the original method. The improved method combines the merits of the principal component analysis' information coherence and the catastrophe progression method's none index weight and has the advantage of highly objective comprehensive evaluation. Through the systematic analysis of the influencing factors of the automotive industry's core technology capacity, the comprehensive evaluation model is established according to the different roles that different indices play in evaluating the overall goal with a hierarchical structure. Moreover, ICPM is developed for evaluating the automotive industry's core technology capacity for the typical seven countries in the world, which demonstrates the effectiveness of the method.

  8. Toward a Comprehensive Framework for Evaluating the Core Integration Features of Enterprise Integration Middleware Technologies

    Directory of Open Access Journals (Sweden)

    Hossein Moradi

    2013-01-01

    Full Text Available To achieve greater automation of their business processes, organizations face the challenge of integrating disparate systems. In attempting to overcome this problem, organizations are turning to different kinds of enterprise integration. Implementing enterprise integration is a complex task involving both technological and business challenges and requires appropriate middleware technologies. Different enterprise integration solutions provide various functions and features which lead to the complexity of their evaluation process. To overcome this complexity, appropriate tools for evaluating the core integration features of enterprise integration solutions is required. This paper proposes a new comprehensive framework for evaluating the core integration features of both intra-enterprise and inter-enterprise Integration's enabling technologies, which simplify the process of evaluating the requirements met by enterprise integration middleware technologies.The proposed framework for evaluating the core integration features of enterprise integration middleware technologies was enhanced using the structural and conceptual aspects of previous frameworks. It offers a new schema for which various enterprise integration middleware technologies are categorized in different classifications and are evaluated based on their supporting level for the core integration features' criteria. These criteria include the functional and supporting features. The proposed framework, which is a revised version of our previous framework in this area, has developed the scope, structure and content of the mentioned framework.

  9. Evaluation of the oxide and silicide fuels reactivity in the RSG-GAS core

    International Nuclear Information System (INIS)

    S, Tukiran; M S, Tagor; S, Lily; Pinem, S.

    2000-01-01

    Fuel exchange of The RSG-GAS reactor core from uranium oxide to uranium silicide in the same loading, density, and enrichment, that is, 250 gr, 2.98 gr/cm 3 , and 19.75 % respectively, will be performed in-step wise. In every cycle of exchange with 5/l mode, it is needed to evaluate the parameter of reactor core operation. One of the important operation parameters is fuel reactivity that gives effect to the core reactivity. The experiment was performed at core no. 36, BOC, low power which exist 2 silicide fuels. The evaluation was done based on the RSG-GAS control rod calibration consisting of 40 fuels and 8 control rod.s. From 40 fuels in the core, there are 2 silicide fuels, RI-225/A-9 and RI-224/C-3. For inserting 2 silicide fuels, the reactivity effect to the core must be know. To know this effect , it was performed fuels reactivity experiment, which based on control rod calibration. But in this case the RSG-GAS has no other fresh oxide fuel so that configuration of the RSG-GAS core was rearranged by taking out the both silicide fuels and this configuration is used as reference core. Then silicide fuel RI-224 was inserted to position F-3 replacing the fresh oxide fuel RI-260 so the different reactivity of the fuels is obtained. The experiment result showed that the fuel reactivity change is in amount of 12.85 cent (0.098 % ) The experiment result was compared to the calculation result, using IAFUEL code which amount to 13.49 cent (0.103 %) The result showed that the reactivity change of oxide to silicide fuel is small so that the fuel exchange from uranium oxide to uranium silicide in the first step can be done without any significant change of the operation parameter

  10. On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

    International Nuclear Information System (INIS)

    Merk, B.; Weiss, F. P.

    2012-01-01

    The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

  11. On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B. [Helmholtz-Zentrum Dresden-Rossendorf, Institut fuer Sicherheitsforschung, Postfach 51 01 19, 01314 Dresden (Germany); Weiss, F. P. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH Forschungszentrum, Boltzmannstr. 14, 85748 Garching (Germany)

    2012-07-01

    The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

  12. Evaluation of fission spectra and cross sections by zero-leakage core experiments

    International Nuclear Information System (INIS)

    Iijima, T.; Mukaiyama, T.

    1979-01-01

    A series of unit k-infinity core experiments were performed in FCA of JAERI to obtain the information on the equivalence of 239 Pu to 235 U in fast reactors, and to examine the inelastic slowing down cross section of 238 U. Three assemblies were built. Each assembly consists of a test zone (about 44l) of nearly unit k-infinity, a 20% enriched uranium driver and a natural uranium blanket. Assembly IV-1 (first built in 1969 and rebuilt in 1972) is an all uranium system, and Assemblies IV-1-P, IV-1-P' have a plutonium/natural uranium test zone. Three assemblies are nearly the same from the view-point of the slowing down cross section in the main energy region of the neutron spectrum, since 238 U occupies the most part of the composition. The main difference between Assembly IV-1 and the latter two is the difference in the fissile material. Fission rate ratios and k-infinity values were measured to obtain knowledge of the fission spectra and cross sections important for the criticality. In order to evaluate the inelastic slowing down cross section of 238 U, neutron spectra were measured with various methods. The analysis was done with four cross section sets. The agreement of k-infinity values between the experiment and the calculation is unsatisfactory, especially for Pu/NU systems

  13. About the use of approximations, which ensure materials mass balance conservation by spatial meshes, in Sn full core calculations

    International Nuclear Information System (INIS)

    Voloshchenko, A.M.; Russkov, A.A.; Gurevich, M.I.; Olejnik, D.S.

    2008-01-01

    One analyzes a possibility to make use of the geometry approximations conserving the materials mass local balance in every mesh via adding of mixtures in the meshes containing several feed materials to perform the kinetic calculation of the reactor core neutron fields. To set the 3D-geometry of the reactor core one makes use of the combinatorial geometry methods implemented in the MCI Program to solve the diffusivity equations by the Monte Carlo method, to convert the combinatorial prescribing of the geometry into the mesh representation - the ray tracing method. According to the calculations of the WWER-1000 reactor core and the simulations of the spent fuel storage facility, the described procedure compares favorably with the conventional geometry approximations [ru

  14. Evaluation on Core Competitiveness of Wholesale Market of Agricultural Products Based on CWAA Operator

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    According to relevant data,we select five indices,namely management ability,organization and management capability,enterprise culture,development ability and technical equipment ability,to establish the index system of core competitiveness of wholesale market of agricultural products.Based on combination weight arithmetic average(CWAA) operator,we advance an evaluation model of core competitiveness of wholesale market of agricultural products which involves participation of many people.By inviting five exerts,we conduct evaluation in terms of management ability of wholesale market of agricultural products,organization and management capability of leadership,enterprise culture of wholesale market of agricultural products,future development ability of wholesale market of agricultural products,and exiting technical equipment ability of wholesale market of agricultural products.We adopt hundred-mark system to grade and evaluate core competitiveness of wholesale market of agricultural products.The results show that the experts’ evaluation score of core competitiveness of wholesale market of agricultural products is high.The evaluation result is reasonable and authentic and this model is feasible.

  15. Development of the advanced nuclear materials -Materials performance evaluation-

    International Nuclear Information System (INIS)

    Kim, Woo Chul; Noh, Kye Hoh; Han, Jung Hoh; Jung, Han Sub; Kim, Hong Pyo; Lee, Duk Hyun; Lee, Eun Heui; Hwang, Sung Sik; Huh, Doh Haeng

    1995-07-01

    The software for ACPD was modified to use multi-specimens and multi-frequency. The stress corrosion cracking resistance test of Alloy 600 in Pb contained water was performed by slow strain rate tester. The corrosion fatigue test machine was installed, and an autoclave for this test was purchased. The fatigue test was conducted in air. The stability for the long term test in DCPD was evaluated, and the improvement of current source and the revision of potential drop difference according to temperature variation increased the detection accuracy. A Ag/AgCl reference electrode and electrode support were assembled and the performance test was carried out at high temperature under high pressure. The zirconia pH electrode was assembled. The specimens with SUS304, Zr-2.5Nb were machined for irradiation assisted degradation test. The erosion/corrosion for the selected secondary side piping of Kori-1 was evaluated by CHECKMATE code. The chemical analysis and metallurgical inspection of the secondary piping of Kori-1 were conducted, and the erosion/corrosion test loop was made. 29 figs, 12 tabs, 11 refs. (Author)

  16. Evaluation of re-criticality potential in Fukushima Dai-ichi reactors following core damage accidents

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The re-criticality potential of the debris-bed, formed of the degraded core materials, cannot be ruled out during the cooling-down procedure of the Fukushima Dai-ichi NPPs. In this study the re-criticality potential has systematically investigated based on the core disruption phase analysis using a IMPACT-SAMPSON code prepared by The Institute of Applied Energy (IAE). The results obtained for the re-criticality potential, characterized by the eigen-values k-eff dependent on the debris composition formed at the core, RPV bottom, and PCV pedestal, are reflected to the arguments on the re-criticality prevention measures, such as timing and concentration of boron-compounds, during the cooling-down process of the Fukushima Dai-ichi NPPs. (author)

  17. Material dimensionality effects on the nanoindentation behavior of Al/a-Si core-shell nanostructures

    Energy Technology Data Exchange (ETDEWEB)

    Fleming, Robert A. [Department of Mechanical Engineering, University of Arkansas, Fayetteville, AR 72701 (United States); Center for Advanced Surface Engineering, University of Arkansas, Fayetteville, AR 72701 (United States); Goss, Josue A. [Center for Advanced Surface Engineering, University of Arkansas, Fayetteville, AR 72701 (United States); Zou, Min, E-mail: mzou@uark.edu [Department of Mechanical Engineering, University of Arkansas, Fayetteville, AR 72701 (United States); Center for Advanced Surface Engineering, University of Arkansas, Fayetteville, AR 72701 (United States)

    2017-08-01

    Highlights: • Nanoindentation behavior of Al/a-Si core-shell nanostructures were studied. • 3D core confinement enables significant deformation recovery beyond elastic limit. • As the confinement is reduced, the deformation recovery is reduced or suppressed. • Atomistic simulations suggest core confinement affects dislocation dynamics. • 3D confinement has the highest percentage of dislocation removal after unloading. - Abstract: The nanoindentation behavior of hemispherical Al/a-Si core-shell nanostructures (CSNs), horizontally-aligned Al/a-Si core-shell nanorods (CSRs) with various lengths, and an Al/a-Si layered thin film has been studied to understand the effects of geometrical confinement of the Al core on the CSN deformation behavior. When loaded beyond the elastic limit, the CSNs have an unconventional load-displacement behavior with no residual displacement after unloading, resulting in no net shape change after indentation. This behavior is enabled by dislocation activities within the confined Al core, as indicated by discontinuous indentation signatures (load-drops and load-jumps) observed in the load-displacement data. When the geometrical confinement of the core is slightly reduced, as in the case of CSRs with the shortest rod length, the discontinuous indentation signatures and deformation resistance are heavily reduced. Further decreases in core confinement result in conventional nanoindentation behavior, regardless of geometry. Supporting molecular dynamics simulations show that dislocations nucleated in the core of a CSN are more effectively removed during unloading compared to CSRs, which supports the hypothesis that the unique deformation resistance of Al/a-Si CSNs are enabled by 3-dimensional confinement of the Al core.

  18. Material dimensionality effects on the nanoindentation behavior of Al/a-Si core-shell nanostructures

    International Nuclear Information System (INIS)

    Fleming, Robert A.; Goss, Josue A.; Zou, Min

    2017-01-01

    Highlights: • Nanoindentation behavior of Al/a-Si core-shell nanostructures were studied. • 3D core confinement enables significant deformation recovery beyond elastic limit. • As the confinement is reduced, the deformation recovery is reduced or suppressed. • Atomistic simulations suggest core confinement affects dislocation dynamics. • 3D confinement has the highest percentage of dislocation removal after unloading. - Abstract: The nanoindentation behavior of hemispherical Al/a-Si core-shell nanostructures (CSNs), horizontally-aligned Al/a-Si core-shell nanorods (CSRs) with various lengths, and an Al/a-Si layered thin film has been studied to understand the effects of geometrical confinement of the Al core on the CSN deformation behavior. When loaded beyond the elastic limit, the CSNs have an unconventional load-displacement behavior with no residual displacement after unloading, resulting in no net shape change after indentation. This behavior is enabled by dislocation activities within the confined Al core, as indicated by discontinuous indentation signatures (load-drops and load-jumps) observed in the load-displacement data. When the geometrical confinement of the core is slightly reduced, as in the case of CSRs with the shortest rod length, the discontinuous indentation signatures and deformation resistance are heavily reduced. Further decreases in core confinement result in conventional nanoindentation behavior, regardless of geometry. Supporting molecular dynamics simulations show that dislocations nucleated in the core of a CSN are more effectively removed during unloading compared to CSRs, which supports the hypothesis that the unique deformation resistance of Al/a-Si CSNs are enabled by 3-dimensional confinement of the Al core.

  19. Development of on-line core performance evaluation system for 'FUGEN'

    International Nuclear Information System (INIS)

    Natori, Hisahide; Kaneto, Kunikazu; Oteru, Shigeru.

    1982-01-01

    An on-line core performance evaluation system ATROPOS has been developed in order to carry out safe and efficient reactor operation of ''FUGEN''(a heavy water moderated, boiling light water cooled, pressure tube type reactor). This system offers detailed and useful information on such items of core performance as core thermal power, power distribution and thermal operation limits. The power distribution is calculated first by using a three-dimensional nodal coupling model, employing such process data as control rod position and 10 B concentration in the D 2 O moderator. Then the calculated power distribution is corrected by local power monitor readings. An axial one-dimensional nodal coupling model, which considers radial power distribution, and a localized three-dimensional nodal coupling model are used to predict the core thermal power and the power distribution for the region surrounding the control rods respectively, within a short time in advance of control rod operation. The methods employed in this system are verified by comparison with start-up test data from the FUGEN initial core. The estimated power distribution and channel flow agree with values measured by the power calibration monitor and with channel flow converted from measured values of pressure drop, within 3 and 5% respectively, in their root mean square values. The difference in core thermal power between the predicted value and the value measured by the total power monitor is about 1% for control rod operation. (author)

  20. Sport-specific endurance plank test for evaluation of global core muscle function.

    Science.gov (United States)

    Tong, Tom K; Wu, Shing; Nie, Jinlei

    2014-02-01

    To examine the validity and reliability of a sports-specific endurance plank test for the evaluation of global core muscle function. Repeated-measures study. Laboratory environment. Twenty-eight male and eight female young athletes. Surface electromyography (sEMG) of selected trunk flexors and extensors, and an intervention of pre-fatigue core workout were applied for test validation. Intraclass correlation coefficient (ICC), coefficient of variation (CV), and the measurement bias ratio */÷ ratio limits of agreement (LOA) were calculated to assess reliability and measurement error. Test validity was shown by the sEMG of selected core muscles, which indicated >50% increase in muscle activation during the test; and the definite discrimination of the ∼30% reduction in global core muscle endurance subsequent to a pre-fatigue core workout. For test-retest reliability, when the first attempt of three repeated trials was considered as familiarisation, the ICC was 0.99 (95% CI: 0.98-0.99), CV was 2.0 ± 1.56% and the measurement bias ratio */÷ ratio LOA was 0.99 */÷ 1.07. The findings suggest that the sport-specific endurance plank test is a valid, reliable and practical method for assessing global core muscle endurance in athletes given that at least one familiarisation trial takes place prior to measurement. Copyright © 2013 Elsevier Ltd. All rights reserved.

  1. Evaluation of RSG-GAS Core Management Based on Burnup Calculation

    International Nuclear Information System (INIS)

    Lily Suparlina; Jati Susilo

    2009-01-01

    Evaluation of RSG-GAS Core Management Based on Burnup Calculation. Presently, U 3 Si 2 -Al dispersion fuel is used in RSG-GAS core and had passed the 60 th core. At the beginning of each cycle the 5/1 fuel reshuffling pattern is used. Since 52 nd core, operators did not use the core fuel management computer code provided by vendor for this activity. They use the manually calculation using excel software as the solving. To know the accuracy of the calculation, core calculation was carried out using two kinds of 2 dimension diffusion codes Batan-2DIFF and SRAC. The beginning of cycle burn-up fraction data were calculated start from 51 st to 60 th using Batan-EQUIL and SRAC COREBN. The analysis results showed that there is a disparity in reactivity values of the two calculation method. The 60 th core critical position resulted from Batan-2DIFF calculation provide the reduction of positive reactivity 1.84 % Δk/k, while the manually calculation results give the increase of positive reactivity 2.19 % Δk/k. The minimum shutdown margin for stuck rod condition for manual and Batan-3DIFF calculation are -3.35 % Δk/k dan -1.13 % Δk/k respectively, it means that both values met the safety criteria, i.e <-0.5 % Δk/k. Excel program can be used for burn-up calculation, but it is needed to provide core management code to reach higher accuracy. (author)

  2. Evaluation of criteria for developing traffic safety materials for Latinos.

    Science.gov (United States)

    Streit-Kaplan, Erica L; Miara, Christine; Formica, Scott W; Gallagher, Susan Scavo

    2011-03-01

    This quantitative study assessed the validity of guidelines that identified four key characteristics of culturally appropriate Spanish-language traffic safety materials: language, translation, formative evaluation, and credible source material. From a sample of 190, the authors randomly selected 12 Spanish-language educational materials for analysis by 15 experts. Hypotheses included that the experts would rate materials with more of the key characteristics as more effective (likely to affect behavioral change) and rate materials originally developed in Spanish and those that utilized formative evaluation (e.g., pilot tests, focus groups) as more culturally appropriate. Although results revealed a weak association between the number of key characteristics in a material and the rating of its effectiveness, reviewers rated materials originally created in Spanish and those utilizing formative evaluation as significantly more culturally appropriate. The findings and methodology demonstrated important implications for developers and evaluators of any health-related materials for Spanish speakers and other population groups.

  3. Accelerated materials evaluation for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, M., E-mail: malcolm.griffiths@queensu.ca [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, Ontario, K7L 3N6 (Canada); Walters, L. [Canadian Nuclear Laboratories, Chalk River, ON, K0J 1J0 (Canada); Greenwood, L.R. [Pacific Northwest National Laboratory, Richland, WA, 99352 (United States); Garner, F.A. [Radiation Effects Consulting, Richland, WA, 99352 (United States)

    2017-05-15

    This paper addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors.

  4. A Comparison of General and Work-Specific Measures of Core Self-Evaluations

    Science.gov (United States)

    Bowling, Nathan A.; Wang, Qiang; Tang, Han Ying; Kennedy, Kellie D.

    2010-01-01

    During the past decade, considerable research attention has been given to core self-evaluations (CSEs). Although this research has found that CSE is related to several important work-related outcomes (e.g., job satisfaction, job performance), we believe that researchers' reliance on general rather than work-specific CSE has resulted in…

  5. Structural evaluation for the core sampling trucks, RMCS operations, 200 Area

    International Nuclear Information System (INIS)

    Islam, M.A.

    1996-01-01

    This report evaluates the structural adequacy and the integrity of the existing core sampling trucks to withstand impact should the trucks drop off the ramp, either onto the soft ground or onto a non-yielding surface due to operational error, wind, or earthquake. The report also addresses if the allowable tank dome load will be exceeded by the addition of the impact load

  6. Core Self-Evaluations as Causes of Satisfaction: The Mediating Role of Seeking Task Complexity

    Science.gov (United States)

    Srivastava, Abhishek; Locke, Edwin A.; Judge, Timothy A.; Adams, John W.

    2010-01-01

    This study examined the mediating role of task complexity in the relationship between core self-evaluations (CSE) and satisfaction. In Study 1, eighty three undergraduate business students worked on a strategic decision-making simulation. The simulated environment enabled us to verify the temporal sequence of variables, use an objective measure of…

  7. Core self-evaluations and work engagement: Testing a perception, action, and development path

    NARCIS (Netherlands)

    Tims, M.; Akkermans, J.

    2017-01-01

    Core self-evaluations (CSE) have predictive value for important work outcomes such as job satisfaction and job performance. However, little is known about the mechanisms that may explain these relationships. The purpose of the present study is to contribute to CSE theory by proposing and

  8. ABOUT INDEX EVALUATION OF MATERIAL RESOURCE SUPPLIER SELECTION

    Directory of Open Access Journals (Sweden)

    V. A. Skochinskaya

    2008-01-01

    Full Text Available The paper analyzes existing methods for evaluation of material resource supplier selection. It shows advantages and shortcomings of the present evaluation systems. The necessity for application of an index evaluation is justified in the paper. The paper contains rating (index evaluation for material resource supplier selection which is based on the application of quantitative (index tool instead of an expert (numerical evaluation

  9. Material control test and evaluation system at the ICPP

    International Nuclear Information System (INIS)

    Johnson, C.E.

    1979-01-01

    The US DOE is evaluating process monitoring as part of a total nuclear material safeguards system. A monitoring system is being installed at the Idaho Chemical Processing Plant to test and evaluate material control and surveillance concepts in an operating nuclear fuel reprocessing plant. Process monitoring for nuclear material control complements conventional safeguards accountability and physical protection to assure adherence to approved safeguards procedures and verify containment of nuclear materials within the processing plant

  10. Development and evaluation of an X-ray radioscopy device for drill cores study

    International Nuclear Information System (INIS)

    Bertrand, L.; Gentier, S.; Massal, P.

    1993-01-01

    This work is a cost-sharing contract with the European Atomic Energy Community within the framework of research and development program on management, storage and radioactive waste disposal. The aim of this project is to conceive an X-ray radioscopy mobile unit, adapted to the study of cored geological materials. A prestudy based on the X-ray absorption theory by the material has enabled to design the apparatus and specially the X-ray tube power. Then the schematic diagram of the device is presented and the principle on which it works is described. The main components of the XCORE device may be put together into three big sets: - The X-ray part includes the high-voltage generator, the X-ray transmitter tube, the receiver or brightness-amplifying tube and all the acquisition, visualization and recording system for the video images, and at last the X-ray controls rack, -The mechanical part is composed of the handling cores system, the location system of the radioscopied core sections, the control mechanism of the core's motions, - A PC/AT microcomputer and its peripherals fitted out with a digitizing and processing image card makes up the computing part. The equipment is mounted into a container transportable by lorry, 2.5 x 2.5 x 6 m. in size and 9 T. weight. 6 refs., 79 figs., 3 tabs

  11. Material properties of oxide dispersion strengthened (ODS) ferritic steels for core materials of FBR. Tensile properties of sodium exposed and nickel diffused materials

    International Nuclear Information System (INIS)

    Kato, Shoichi; Yoshida, Eiichi

    2002-12-01

    An oxide dispersion strengthened (ODS) ferritic steel is candidate for a long-life core materials of future FBR, because of good swelling resistance and high creep strength. In this study, tensile tests were carried out the long-term extrapolation of sodium environmental effects on the mechanical properties of ODS steels. The tested heats of materials are M93, M11 and F95. The specimens were pre-exposed to sodium for 1,000 and 3,000 hours under non-stress conditions. The pre-exposure to sodium was conducted using a sodium test loop constituted by austenitic steels. For the conditions of sodium exposure test, the sodium temperature was 650 and 700degC, the oxygen concentration in sodium was about 1 ppm and sodium flow rate on the surface of specimen was less than 1x10 -4 m/seconds (nearly static). Further the specimen with the nickel diffused was prepared, which is simulate to nickel diffusing through sodium from the surface of structural stainless steels. The main results obtained were as follows; (1) The tensile strength and the fracture elongation after sodium exposure (maximum 3,000 hours) were same as that of as-received materials. If was considered that the sodium environmental effect is negligible under the condition of this study. (2) Tensile properties of nickel diffused specimens were slightly lower than that of the as-received specimens, but it remains equal to that of thermal aging specimens. (3) The change in microstructure such as a degraded layer was observed on the surface of nickel diffused specimen. In the region of the degraded layer, phase transformations from the α-phase to the γ-phase were recognized. But, the microscopic oxide particles were observed same as that of α-phase base metal. (author)

  12. Crust behavior and erosion rate prediction of EPR sacrificial material impinged by core melt jet

    Energy Technology Data Exchange (ETDEWEB)

    Li, Gen; Liu, Ming, E-mail: ming.liu@mail.xjtu.edu.cn; Wang, Jinshi; Chong, Daotong; Yan, Junjie

    2017-04-01

    Highlights: • A numerical code was developed to analyze melt jet-concrete interaction in the frame of MPS method. • Crust and ablated concrete layer at UO{sub 2}-ZrO{sub 2} melt and concrete interface periodically developed and collapsed. • Concrete surface temperature fluctuated around a low temperature and ablation temperature. • Concrete erosion by Fe-Zr melt jet was significantly faster than that by UO{sub 2}-ZrO{sub 2} melt jet. - Abstract: Sacrificial material is a special ferro-siliceous concrete, designed in the ex-vessel core melt stabilization system of European Pressurized water Reactor (EPR). Given a localized break of RPV lower head, the melt directly impinges onto the dry concrete in form of compact jet. The concrete erosion behavior influences the failure of melt plug, and further affects melt spreading. In this study, a numerical code was developed in the frame of Moving Particle Semi-implicit (MPS) method, to analyze the crust behavior and erosion rate of sacrificial concrete, impinged by prototypic melt jet. In validation of numerical modeling, the time-dependent erosion depth and erosion configuration matched well with the experimental data. Sensitivity study of sacrificial concrete erosion indicates that the crust and ablated concrete layer presented at UO{sub 2}-ZrO{sub 2} melt and concrete interface, whereas no crust could be found in the interaction of Fe-Zr melt with concrete. The crust went through stabilization-fracture-reformation periodic process, accompanied with accumulating and collapsing of molten concrete layer. The concrete surface temperature fluctuated around a low temperature and ablation temperature. It increased as the concrete surface layer was heated to melting, and dropped down when the cold concrete was revealed. The erosion progression was fast in the conditions of small jet diameter and large concrete inclination angle, and it was significantly faster in the erosion by metallic melt jet than by oxidic melt jet.

  13. How can core self-evaluations influence job burnout? The key roles of organizational commitment and job satisfaction.

    Science.gov (United States)

    Peng, Jiaxi; Li, Dongdong; Zhang, Zhenjiang; Tian, Yu; Miao, Danmin; Xiao, Wei; Zhang, Jiaxi

    2016-01-01

    This study aimed to explore how core self-evaluations influenced job burnout and mainly focused on the confirmation of the mediator roles of organizational commitment and job satisfaction. A total of 583 female nurses accomplished the Core Self-Evaluation Scale, Organizational Commitment Scale, Minnesota Satisfaction Questionnaire, and Maslach Burnout Inventory-General Survey. The results revealed that core self-evaluations, organizational commitment, job satisfaction, and job burnout were significantly correlated with each other. Structural equation modeling indicated that core self-evaluations can significantly influence job burnout and are completely mediated by organizational commitment and job satisfaction. © The Author(s) 2014.

  14. The Relationship of Core Self-Evaluations and Life Satisfaction in College Students with Disabilities: Evaluation of a Mediator Model

    Science.gov (United States)

    Smedema, Susan Miller; Chan, Fong; Yaghmaian, Rana A.; Cardoso, Elizabeth DaSilva; Muller, Veronica; Keegan, John; Dutta, Alo; Ebener, Deborah J.

    2015-01-01

    This study examined the factorial structure of the construct core self-evaluations (CSE) and tested a mediational model of the relationship between CSE and life satisfaction in college students with disabilities. We conducted a quantitative descriptive design using exploratory and confirmatory factor analysis and multiple regression analysis.…

  15. Standard practice for radiologic examination of flat panel composites and sandwich core materials used in aerospace applications

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This practice is intended to be used as a supplement to Practices E 1742, E 1255, and E 2033. 1.2 This practice describes procedures for radiologic examination of flat panel composites and sandwich core materials made entirely or in part from fiber-reinforced polymer matrix composites. Radiologic examination is: a) radiographic (RT) with film, b) Computed Radiography (CR) with Imaging Plate, c) Digital Radiology (DR) with Digital Detector Array’s (DDA), and d) Radioscopic (RTR) Real Time Radiology with a detection system such as an Image Intensifier. The composite materials under consideration typically contain continuous high modulus fibers (> 20 GPa), such as those listed in 1.4. 1.3 This practice describes established radiological examination methods that are currently used by industry that have demonstrated utility in quality assurance of flat panel composites and sandwich core materials during product process design and optimization, process control, after manufacture inspection, in service exami...

  16. Propagation of Electromagnetic Waves in Slab Waveguide Structure Consisting of Chiral Nihility Claddings and Negative-Index Material Core Layer

    Science.gov (United States)

    Helal, Alaa N. Abu; Taya, Sofyan A.; Elwasife, Khitam Y.

    2018-06-01

    The dispersion equation of an asymmetric three-layer slab waveguide, in which all layers are chiral materials is presented. Then, the dispersion equation of a symmetric slab waveguide, in which the claddings are chiral materials and the core layer is negative index material, is derived. Normalized cut-off frequencies, field profile, and energies flow of right-handed and left-handed circularly polarized modes are derived and plotted. We consider both odd and even guided modes. Numerical results of guided low-order modes are provided. Some novel features, such as abnormal dispersion curves, are found.

  17. Design and evaluation of materials for space reactors

    International Nuclear Information System (INIS)

    Tavassoli, A.A.; Vrillon, B.; Robert, G.

    1990-01-01

    The French programme envisages a 20 kWe reactor, project ERATO, with three technological options. The first option is a sodium cooled reactor, derived from the fast breeder reactor technology, (upper core outlet temperature of 700 0 C). The second option is based on the High Temperature Gas-cooled Reactor technology (outlet temperature range 700 0 C-900 0 C). The third option is the reference solution, lithium cooled and UN fuelled fast spectrum reactor, (outlet temperature as high as 1200 0 C). The choice is essentially dominated by material considerations, and more specifically by the problems related to the compatibility with the cooling medium and to the high temperature creep resistance. For the first system limited work will be needed as the technology used is well experimented and there is a wealth of information on the austenitic stainless steel Type 316L-SPH. For the second system, most of the work has been concentrated on characterization of existing commercial alloys. This has included the preselection and the testing of a number of superalloys irradiated or not. The results obtained from high temperature tensile and creep tests have allowed selection of Haynes 230 as the primary candidate material and have also permitted calculation of allowable design stresses for this alloy. For the very high temperature system the French R and D programme has focused on Mo-Re alloys. The results obtained to this date from microstructural examinations and mechanical tests performed on different alloy compositions have allowed selection of Mo-25%Re for future optimization work. They have also shown the need for evaluation of creep properties at low stresses where microstructural instabilities are likely to occur as a result of long exposure to high temperature

  18. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour

  19. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour.

  20. CT Performance Evaluation Using Multi Material Assemblies

    DEFF Research Database (Denmark)

    Stolfi, Alessandro; De Chiffre, Leonardo

    2015-01-01

    This paper concerns an investigation of the accuracy of Computed Tomography measurements using multi-material assemblies. In this study, assemblies involving similar densities for elementary parts were considered. The investigation includes dimensional and geometrical measurements of two 10 mm high...

  1. Evaluation of The Value of Core Needle Biopsy in The Diagnosis of a Breast Mass

    Directory of Open Access Journals (Sweden)

    Asieh Sadat Fattahi

    2016-06-01

    Full Text Available Background: Core needle biopsy (CNB with histological findings is regarded as one of the most important diagnostic measures that make preoperative assessment and planning for appropriate treatment possible. The aim of this study was to determine the sensitivity and specificity of core biopsy results in our patients with benign and malignant breast lumps, especially for borderline breast lesions, by using a classification method.Methods: In this study, 116 patients who were referred to the Surgery Clinic of Ghaem Hospital, Mashhad University of Medical Sciences, Mashhad, Iran with breast lump and underwent diagnostic procedures such as mammography and ultrasound were selected. Core needle biopsy (Tru-cut #14 or 16 was performed. After that, excisional biopsy was done. The benign, malignant and unspecified samples obtained by core needle biopsy were evaluated with the samples of the surgical and pathological findings. Then, false positive, false negative, sensitivity, specificity, and diagnostic accuracy of the core needle biopsy method were calculated. Also, the National Health Service Breast Screening Program (NHSBSP classification was employed.Results: The mean age of the participants in this study was 39±13.13 years and the mean tumor size was 2.7 cm. An average of 3.35 biopsies was taken from all patients. Most of the pathology samples taken from CNB and excisional biopsy were compatible with invasive ductal carcinoma. Of the B type classifications, B5 was the most frequent in both methods. Borderline lesions B3 and B4 had a change in their category after surgery. About 2.5% of the samples in core biopsy were inadequate. Skin bruising was the most common core biopsy complication reported. While, the most common complication of excisional biopsy was hematoma. Accuracy, sensitivity, specificity, positive and negative predictive values of the core needle biopsy procedure compared with excisional biopsy was 95.5%, 92.6%, 100%, 100%, and 91

  2. Reaction rate distribution measurement and the core performance evaluation in the prototype FBR Monju

    Energy Technology Data Exchange (ETDEWEB)

    Usami, S.; Suzuoki, Z.; Deshimaru, T. [Monju Construction Office, Japan Nuclear Cycle Development Institute, Fukui-ken (Japan); Nakashima, F. [Tsuruga head Office, Japan Nuclear Cycle Development Institute, Fukui-ken (Japan)

    2001-07-01

    Monju is a prototype fast breeder reactor designed to have an output of 280 MW (714 MWt), fueled with mixed oxides of plutonium and uranium and cooled by liquid sodium. The principal data on plant design and performance are shown in Table 1. Monju attained initial criticality in April 1994 and the reactor physics tests were carried out from May through November 1994. The reaction rate distribution measurement by the foil activation method was one of these tests and was carried out in order to verify the core performance and to contribute to the development of the core design methods. On the basis of the reaction rate measurement data, the Monju initial core breeding ratio and the power distribution were evaluated. (author)

  3. Reaction rate distribution measurement and the core performance evaluation in the prototype FBR Monju

    International Nuclear Information System (INIS)

    Usami, S.; Suzuoki, Z.; Deshimaru, T.; Nakashima, F.

    2001-01-01

    Monju is a prototype fast breeder reactor designed to have an output of 280 MW (714 MWt), fueled with mixed oxides of plutonium and uranium and cooled by liquid sodium. The principal data on plant design and performance are shown in Table 1. Monju attained initial criticality in April 1994 and the reactor physics tests were carried out from May through November 1994. The reaction rate distribution measurement by the foil activation method was one of these tests and was carried out in order to verify the core performance and to contribute to the development of the core design methods. On the basis of the reaction rate measurement data, the Monju initial core breeding ratio and the power distribution were evaluated. (author)

  4. Evaluation report on CCTF Core-II reflood Test C2-15 (Run 75)

    International Nuclear Information System (INIS)

    Okubo, Tsutomu; Iguchi, Tadashi; Akimoto, Hajime; Murao, Yoshio

    1992-01-01

    This report presents an evaluation on the CCTF Core-II Test C2-15 (Run 75). The purpose of the test is to investigate whether the thermo-hydrodynamic behavior is different between the CCTF and the FLECHT-SET reflooding tests. For this purpose, test conditions of the present test were set as close as possible to those of concerned FLECHT-SET 2714B experiment, taking account of differences in facility design. Investigating results of both the tests, the following conclusions are obtained: (1) Some discrepancies were observed in the measured test conditions between the two tests. Out of them, difference in the Acc injection duration was large and affected test results, such as the water accumulation in the downcomer and the core and the core cooling, during the initial period. However, this effect was found to become small with time. (2) Taking account of this difference and the difference in the broken cold leg pressure loss coefficient between the two facilities, the overall reflooding behavior is judged to be similar in the two facilities. (3) The CCTF test results showed the core heat transfer enhancement in the higher power region due to its steep radial power distribution, whereas the FLECHT-SET did not due to its rather flat radial power distribution. This enhancement was observed significantly at 1.83 m but was smaller at the higher elevation. (4) The heat transfer was nearly identical between the two tests and an existing correlation could well predict the heat transfer coefficients of both the tests at the location where the heat transfer enhancement mentioned above (3) were small, during the time period when the effect of the difference in the Acc injection mentioned above (1) were small. (5) Therefore, the core cooling is expected to be almost the same in the CCTF and the FLECHT-SET under the same core boundary conditions and core radial power distribution. (author)

  5. Performance evaluation of seal coat materials and designs.

    Science.gov (United States)

    2011-01-01

    "This project presents an evaluation of seal coat materials and design method. The primary objectives of this research are 1) to evaluate seal coat performance : from various combinations of aggregates and emulsions in terms of aggregate loss; 2) to ...

  6. Study on preparation and microwave absorption property of the core-nanoshell composite materials doped with La.

    Science.gov (United States)

    Wei, Liqiu; Che, Ruxin; Jiang, Yijun; Yu, Bing

    2013-12-01

    Microwave absorbing material plays a great role in electromagnetic pollution controlling, electromagnetic interference shielding and stealth technology, etc. The core-nanoshell composite materials doped with La were prepared by a solid-state reaction method, which is applied to the electromagnetic wave absorption. The core is magnetic fly-ash hollow cenosphere, and the shell is the nanosized ferrite doped with La. The thermal decomposition process of the sample was investigated by thermogravimetry and differential thermal analysis. The morphology and components of the composite materials were investigated by the X-ray diffraction analysis, the microstructure was observed by scanning electron microscope and transmission electron microscope. The results of vibrating sample magnetometer analysis indicated that the exchange-coupling interaction happens between ferrite of magnetic fly-ash hollow cenosphere and nanosized ferrite coating, which caused outstanding magnetic properties. The microwave absorbing property of the sample was measured by reflectivity far field radar cross section of radar microwave absorbing material with vector network analyzer. The results indicated that the exchange-coupling interaction enhanced magnetic loss of composite materials. Therefore, in the frequency of 5 GHz, the reflection coefficient can achieve -24 dB. It is better than single material and is consistent with requirements of the microwave absorbing material at the low-frequency absorption. Copyright © 2013 The Research Centre for Eco-Environmental Sciences, Chinese Academy of Sciences. Published by Elsevier B.V. All rights reserved.

  7. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    Radiation effects on metals or alloys used in fast reactor core components are examined in the papers presented at this conference, the accent being put on swelling and irradiation creep of steels and nickel alloys

  8. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    Nagano, M.; Sakurai, S.; Yamaguchi, H.

    1997-01-01

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  9. Possibility evaluation of eliminating the saturated control fuel element from Tehran research reactor core

    International Nuclear Information System (INIS)

    Mirvakili, S.M.; Keyvani, M.; Arshi, S. Safaei; Khalafi, H.

    2012-01-01

    Highlights: ► We show safe operation of Tehran research reactor without one of its control rods. ► We propose an optimum new core configuration by fuel management calculations. ► We calculate neutronic and thermal hydraulic parameters of the new core. ► Parameters are consistent with the safety criteria. - Abstract: In this study the possibility of safe operation of Tehran research reactor (TRR) providing the elimination of one control rod is evaluated. One of the control fuel elements (CFEs) of TRR has been reached the maximum permissible burn-up and due to the impossibility of fresh fuel assembly provision under current situation, providing an optimum core configuration which satisfies safe operation conditions by applying fuel management calculations is essential. In order to ensure the safe and stable operation of recently proposed configuration for TRR core, neutronic and thermal hydraulic parameters of the new core are calculated and compared with the safety criteria. The results show good compatibility with reactor safety criteria, and provide desired shutdown margin and safety reactivity factor.

  10. Photoelastic stress analysis of different prefabricated post-and-core materials.

    Science.gov (United States)

    Asvanund, Pattapon; Morgano, Steven M

    2011-01-01

    The purpose of this study was to investigate stress developed by a combination of a stainless steel post or a fiber-reinforced resin post with a silver amalgam core or a composite resin core. Two-dimensional photoelastic models were used to simulate root dentin. Posts (ParaPost XT and ParaPost-FiberWhite) were cemented with a luting agent (RelyX Unicem). Silver amalgam cores and composite resin cores were fabricated on the posts. Complete crowns were fabricated and cemented on the cores. Each model was analyzed with 2 force magnitudes and in 2 directions. Fringe orders were recorded and compared using ANOVA (p=0.05) and the Scheffe's test. With vertical force, no stress differences occurred among the 4 groups (p=0.159). With a 30-degree force, there was stress differences among the 4 groups (p<0.001). The combination of a fiber-reinforced post and composite resin core could potentially reduce stresses within the radicular dentin when angled loads are applied.

  11. The Results of the CCI-3 Reactor Material Experiment Investigating 2-D Core-Concrete Interaction and Debris Coolability with a Siliceous Concrete Crucible

    International Nuclear Information System (INIS)

    Farmer, M.T.; Lomperski, S.; Basu, S.

    2006-01-01

    The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program conducted reactor materials experiments and associated analysis to achieve the following two objectives: 1) resolve the ex-vessel debris coolability issue, and 2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs of future plants. With respect to the second objective, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion profiles predicted by codes such as WECHSL, COSACO, TOLBIAC, MEDICIS, and MELCOR. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test investigated the long-term interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined time interval, the cavity was flooded with water to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a description of the facility and an overview of results from this test. (authors)

  12. Study of diluting and absorber materials to control the reactivity during a postulated core meltdown accident in generation IV reactors

    International Nuclear Information System (INIS)

    Plevacova, Kamila

    2010-01-01

    In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B 4 C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic points of view. Concerning B 4 C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B 4 C - UO 2 system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B 4 C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu 2 O 3 or HfO 2 as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al 2 O 3 - HfO 2 and Al 2 O 3 - Eu 2 O 3 were preselected. These systems being completely unknown to date at high temperature in association with UO 2 , first points on the corresponding ternary phase diagrams were researched. Contrary to Al 2 O 3 - Eu 2 O 3 - UO 2 system, the Al 2 O 3 - HfO 2 - UO 2 mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author)

  13. Study of diluting and absorber materials to control reactivity during a postulated core melt down accident in Generation IV reactors

    International Nuclear Information System (INIS)

    Plevacova, K.

    2010-01-01

    In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B 4 C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic point of view. Concerning B 4 C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B 4 C - UO 2 system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, a volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B 4 C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu 2 O 3 or HfO 2 as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al 2 O 3 - HfO 2 and Al 2 O 3 - Eu 2 O 3 were preselected. These systems being completely unknown to date at high temperature in association with UO 2 , first points on the corresponding ternary phase diagrams were researched. Contrary to Al 2 O 3 - Eu 2 O 3 - UO 2 system, the Al 2 O 3 - HfO 2 - UO 2 mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author) [fr

  14. Refractory material crucibles evaluation for U evaporation

    Energy Technology Data Exchange (ETDEWEB)

    Damiao, A.J.; Vasconcelos, G.; Silveira, C.A.B.; Rodrigues, N.A.S. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados

    1996-12-31

    In studies that involve small amounts of U vapor generation, such as spectroscopy or thin films, most of the E-gun power is delivered to the cooling system. Normally crucibles are used as container and thermal insulator. Since liquid U is extremely reactive at evaporation temperatures, the crucibles are seriously attacked, decreasing the insulation efficiency and adding contaminants to the U vapor. There is no complete solution for the problem, however, with a careful choice of materials, one can design crucibles with extended lifetime and reduced contamination. This work reports some preliminary results we have obtained in the assessing of crucible materials and design, such as, graphite, Si C, vitreous carbon and Al{sub 2} O{sub 3}. (author) 1 refs., 3 figs.,2 tabs.

  15. Request from radiation damage evaluation in materials

    International Nuclear Information System (INIS)

    Fukuya, Koji; Kimura, Itsuro

    2003-01-01

    Radiation transport calculations in a PWR using cross-section data sets based on JENDL3.2 showed that the calculated neutron fluence agreed well with the dosimeter measurements and that the fast neutron flux and dpa rate differed within 10% from to those calculated using ENDF/B-IV and ENDF/B-VI based data sets. Calculations of helium generation in structural materials in the PWR using ENDF/B-VI showed that the dominant source of helium is the (n, α) reaction of 59 Ni and that the calculated helium content agreed with the measurements. For accurate estimation of radiation field from a material viewpoint, it is desirable to construct proper cross-section libraries, which have a proper energy group structure and contain sufficient elements including 59 Ni as an indispensable element. (author)

  16. Refractory material crucibles evaluation for U evaporation

    International Nuclear Information System (INIS)

    Damiao, A.J.; Vasconcelos, G.; Silveira, C.A.B.; Rodrigues, N.A.S.

    1996-01-01

    In studies that involve small amounts of U vapor generation, such as spectroscopy or thin films, most of the E-gun power is delivered to the cooling system. Normally crucibles are used as container and thermal insulator. Since liquid U is extremely reactive at evaporation temperatures, the crucibles are seriously attacked, decreasing the insulation efficiency and adding contaminants to the U vapor. There is no complete solution for the problem, however, with a careful choice of materials, one can design crucibles with extended lifetime and reduced contamination. This work reports some preliminary results we have obtained in the assessing of crucible materials and design, such as, graphite, Si C, vitreous carbon and Al 2 O 3 . (author)

  17. Material correlations and models for the irradiation behavior of fissile and fertile material in SNR-300, Mark-II and KNK II, third core

    International Nuclear Information System (INIS)

    Fenneker; Steinmetz; Toebbe

    1986-07-01

    The report contains the material correlations and models used in the fuel pin design code IAMBUS for the irradiation behavior of PuO 2 -UO 2 fissile materials and UO 2 fertile materials of the SNR-300 Mark-II reload and the KNK II third core. They are applicable for pellet densities of more than 90 % of the theoretical density. The presented models of the fuel behavior and the applied material correlations have been derived either from single experiments or from the comparison of theoretically predicted integral fuel behavior with the results of fuel pin irradiation experiments. The material correlations have been examined and extended in the frame of the collaborations INTERATOM/KWU and INTERATOM/KfK. French and British results were included, when available from the European fast reactor knowledge exchange [de

  18. Subjective evaluation of chosen typographical characteristics in marketing materials

    Directory of Open Access Journals (Sweden)

    Petra Talandová

    2008-01-01

    Full Text Available This paper concentrates on the problems of marketing materials quality evaluation and their formal aspect and also customers’ marketing materials evaluation. This area has not been concentrated on very much and nor in the literature is described. The paper presents the results of our own research which queries how the customers subjectively perceive and evaluate the marketing materials. The emphasis was put on the materials quality i.e. on what materials are considered as quality materials by the customers and which attributes mainly influence the quality. The results were aggregated on the basis of customers’ responses an also on the basis of practical examples evaluation which included intentional mistakes. The subjects of the evaluation were marketing materials quality as a general feature, the attributes influencing the quality and marketing materials quality and company quality relation. Also the exam­ples including mistakes were evaluated. According to the questioning results, the respondents’ answers vary much. It is not possible to find unambiguously right or wrong marketing materials eva­lua­tion. This area will be developed in further research which will be concentrated mainly on the typographical aspects.The aim of this paper is to delimit and to define the present situation through the research result exa­mi­na­tion, to define ‘quality’ and to describe the way how marketing materials are perceived by the customers.

  19. Analysis of an out-of-pile experiment for materials redistribution under core disruptive accident condition of fast breeder reactors

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Ninokata, Hisashi; Shimizu, Akinao

    1995-01-01

    Calculation of one of the SIMBATH experiments was performed using the SIMMER-II code. The experiments were intended to simulate the fuel pin disintegration, the molten materials relocation and following materials redistribution that could occur during core disruptive accidents assumed in fast breeder reactors. The calculation by SIMMER-II showed that the incorporated step-wise fuel pin disintegration model and the modified particle jamming model were capable of reproducing the course of materials relocation within the identified ranges of the parameters which governed the blockages formation, i.e. the characteristic radius of solid particles jamming and/or sieving out in the flow and the effective particle viscosity. In particular the final materials redistribution calculated by SIMMER-II very well reproduced the experiment. This fact made it possible to interpret theoretically the mechanisms of flow blockages formation and related materials redistribution. (author)

  20. Hybridization of MOFs and COFs: A New Strategy for Construction of MOF@COF Core-Shell Hybrid Materials.

    Science.gov (United States)

    Peng, Yongwu; Zhao, Meiting; Chen, Bo; Zhang, Zhicheng; Huang, Ying; Dai, Fangna; Lai, Zhuangchai; Cui, Xiaoya; Tan, Chaoliang; Zhang, Hua

    2018-01-01

    The exploration of new porous hybrid materials is of great importance because of their unique properties and promising applications in separation of materials, catalysis, etc. Herein, for the first time, by integration of metal-organic frameworks (MOFs) and covalent organic frameworks (COFs), a new type of MOF@COF core-shell hybrid material, i.e., NH 2 -MIL-68@TPA-COF, with high crystallinity and hierarchical pore structure, is synthesized. As a proof-of-concept application, the obtained NH 2 -MIL-68@TPA-COF hybrid material is used as an effective visible-light-driven photocatalyst for the degradation of rhodamine B. The synthetic strategy in this study opens up a new avenue for the construction of other MOF-COF hybrid materials, which could have various promising applications. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  1. Development of a DNBR evaluation method for the CEA ejection accident in SMART core

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Yoo, Y. J.; In, W. K.; Chang, M. H. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    A methodology applicable to the analysis of the CEA ejection accident in SMART is developed for the evaluation of the fraction of fuel failure caused by DNB. The transient behavior of the core thermal-hydraulic conditions is calculated by the subchannel analysis code MATRA. The minimum DNBR during the accident is calculated by KRB-1 CHF correlation considering the 1/8 symmetry of hot assembly. The variation of hot assembly power during the accident is simulated by the LTC(Limiting transient Curve) which is determined from the analysis of power distribution data resulting from the three-dimensional core dynamics calculations. The initial condition of the accident is determined by considering LOC(Limiting Conditions for Operation) of SMART core. Two different methodologies for the evaluation of DNB failure rate are established; a deterministic method based on the DNB envelope, and a probabilistic method based on the DNB probability of each fuel rod. The methodology developed in this study is applied to the analysis of CEA ejection accident in the preliminary design core of SMART. As the result, the fractions of DNB fuel failure by the deterministic method and the probabilistic method are calculated as 38.7% and 7.8%, respectively. 16 refs., 16 figs., 5 tabs. (Author)

  2. The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

    International Nuclear Information System (INIS)

    Toivonen, A.; Moilanen, P.; Pyykkoenen, M.; Taehtinen, S.; Rintamaa, R.; Saario, T.

    1998-01-01

    Environmentally assisted cracking (EAC) of core materials has become an increasingly important issue of downtime and maintenance costs in nuclear power plants. Small size specimens are necessary in stress corrosion testing of irradiated materials because of difficulties in handling high dose rate materials and because of restricted availability of the materials. The drawback of using small size specimens is that in some cases they do not fulfil the requirements of the relevant testing standards. Recently VTT has developed J-R testing with irradiated and non-irradiated sub size 3 PB specimens, both in inert and in LWR environments. Also, a new materials testing system which will enable simultaneous multiple specimen testing both in laboratory conditions and in operating reactor core is under development. The new testing system will utilize Charpy and sub size 3 PB specimens. The feasibility study of the system has been carried out using different materials. Fracture resistance curves of a Cu-Zr-Cr alloy are shown to be independent of the specimen geometry and size, to some extent. Results gained from tests in simulated boiling water reactor (BWR) water are presented for sensitized SIS 2333 stainless steel. The experimental results indicate that the size of the plastic zone or stress triaxiality must be further studied although no significant effect on the environmentally assisted crack growth rate was observed. (orig.)

  3. The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Toivonen, A.; Moilanen, P.; Pyykkoenen, M.; Taehtinen, S.; Rintamaa, R.; Saario, T. [Valtion Teknillinen Tutkimuskeskus, Espoo (Finland)

    1998-11-01

    Environmentally assisted cracking (EAC) of core materials has become an increasingly important issue of downtime and maintenance costs in nuclear power plants. Small size specimens are necessary in stress corrosion testing of irradiated materials because of difficulties in handling high dose rate materials and because of restricted availability of the materials. The drawback of using small size specimens is that in some cases they do not fulfil the requirements of the relevant testing standards. Recently VTT has developed J-R testing with irradiated and non-irradiated sub size 3 PB specimens, both in inert and in LWR environments. Also, a new materials testing system which will enable simultaneous multiple specimen testing both in laboratory conditions and in operating reactor core is under development. The new testing system will utilize Charpy and sub size 3 PB specimens. The feasibility study of the system has been carried out using different materials. Fracture resistance curves of a Cu-Zr-Cr alloy are shown to be independent of the specimen geometry and size, to some extent. Results gained from tests in simulated boiling water reactor (BWR) water are presented for sensitized SIS 2333 stainless steel. The experimental results indicate that the size of the plastic zone or stress triaxiality must be further studied although no significant effect on the environmentally assisted crack growth rate was observed. (orig.)

  4. Fabrication of Fe3O4@CuO core-shell from MOF based materials and its antibacterial activity

    Science.gov (United States)

    Rajabi, S. K.; Sohrabnezhad, Sh.; Ghafourian, S.

    2016-12-01

    Magnetic Fe3O4@CuO nanocomposite with a core/shell structure was successfully synthesized via direct calcinations of magnetic Fe3O4@HKUST-1 in air atmosphere. The morphology, structure, magnetic and porous properties of the as-synthesized nano composites were characterized by using scanning electron microscope (SEM), transmission electron microscopy (TEM), powder X-ray diffraction (PXRD), and vibration sample magnetometer (VSM). The results showed that the nanocomposite material included a Fe3O4 core and a CuO shell. The Fe3O4@CuO core-shell can be separated easily from the medium by a small magnet. The antibacterial activity of Fe3O4-CuO core-shell was investigated against gram-positive and gram-negative bacteria. A new mechanism was proposed for inactivation of bacteria over the prepared sample. It was demonstrated that the core-shell exhibit recyclable antibacterial activity, acting as an ideal long-acting antibacterial agent.

  5. Evaluation report on CCTF core-I reflood test C1-19 (RUN 38)

    International Nuclear Information System (INIS)

    Murao, Yoshio; Fujiki, Kazuo; Akimoto, Hajime

    1983-02-01

    A test named the Evaluation Model (EM) test was performed, whose test conditions were simulated the reflood phase predicted with the safety evaluation analysis. The test results were compared with the blindfold results predicted by Evaluation Model (EM) codes. The main conclusions are as follows: (1) The core heat transfer model built in the EM codes gives conservative results. (2) The system models in the present EM codes are found to be well balanced integrally over the system. (3) Conservative items and items to be improved are pointed out. The downcomer slow water accumulation observed in the lower flow rate test was not appeared in the EM test. (author)

  6. Feasibility study of passive gamma spectrometry of molten core material from Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy - low-volatile FP and special nuclear material inventory analysis and fundamental characteristics of gamma-rays from fuel debris

    International Nuclear Information System (INIS)

    Sagara, Hiroshi; Tomikawa, Hirofumi; Watahiki, Masaru; Kuno, Yusuke

    2014-01-01

    The technologies applied to the analysis of the Three Mile Island accident were examined in a feasibility study of gamma spectrometry of molten core material from the Fukushima Daiichi Nuclear Power Station unit 1, 2, and 3 cores for special nuclear material accountancy. The focus is on low-volatile fission products and heavy metal inventory analysis, and the fundamental characteristics of gamma-rays from fuel debris with respect to passive measurements. The inventory ratios of the low-volatile lanthanides, "1"5"4Eu and "1"4"4Ce, to special nuclear materials were evaluated by the entire core inventories in units 1, 2, and 3 with an estimated uncertainty of 9%-13% at the 1σ level for homogenized molten fuel material. The uncertainty is expected to be larger locally owing to the use of the irradiation cycle averaging approach. The ratios were also evaluated as a function of burnup for specific fuel debris with an estimated uncertainty of 13%-25% at the 1σ level for units 1 and 2, and most of the fuels in unit 3, although the uncertainty regarding the separated mixed oxide fuel in unit 3 would be significantly higher owing to the burnup dependence approach. Source photon spectra were also examined and cooling-time-dependent data sets were prepared. The fundamental characteristics of high-energy gamma-rays from fuel debris were investigated by a bare-sphere model transport calculation. Mass attenuation coefficients of fuel debris were evaluated to be insensitive to its possible composition in a high-energy region. The leakage photon ratio was evaluated using a variety of parameters, and a significant impact was confirmed for a certain size of fuel debris. Its correlation was summarized with respect to the leakage photopeak ratio of source "1"5"4Eu. Finally, a preliminary study using a hypothetical canister model of fuel debris based on the experience at Three Mile Island was presented, and future plans were introduced. (author)

  7. Core Self-Evaluation and Burnout among Nurses: The Mediating Role of Coping Styles

    Science.gov (United States)

    Li, Xiaofei; Guan, Lili; Chang, Hui; Zhang, Bo

    2014-01-01

    Objectives This study aimed to determine the potential association between core self-evaluation and the burnout syndrome among Chinese nurses, and the mediating role of coping styles in this relationship. Methods A cross-sectional survey was conducted in Shenyang, China, from May to July, 2013. A questionnaire which consisted of the Maslach Burnout Inventory-General Survey (MBI-GS), the Core Self-Evaluation Scale (CSE), and the Simplified Coping Style Questionnaire (CSQ), was completed by a total of 1,559 nurses. Hierarchical linear regression analyses and the Sobel test were performed to determine the mediating role of coping styles on the relationship between CSE and burnout. Results Nurses who had higher self-evaluation characteristics, reported less emotional exhaustion and cynicism, and higher professional efficacy. Coping style had a partial mediating effect on the relationship between CSE and the burnout syndrome among nurses. Conclusions Core self-evaluation had effects on burnout and coping style was a mediating factor in this relationship among Chinese nurses. Therefore, the improvement of coping strategies may be helpful in the prevention of burnout among nurses, thus enhancing professional performance. PMID:25541990

  8. Core self-evaluation and burnout among Nurses: the mediating role of coping styles.

    Directory of Open Access Journals (Sweden)

    Xiaofei Li

    Full Text Available OBJECTIVES: This study aimed to determine the potential association between core self-evaluation and the burnout syndrome among Chinese nurses, and the mediating role of coping styles in this relationship. METHODS: A cross-sectional survey was conducted in Shenyang, China, from May to July, 2013. A questionnaire which consisted of the Maslach Burnout Inventory-General Survey (MBI-GS, the Core Self-Evaluation Scale (CSE, and the Simplified Coping Style Questionnaire (CSQ, was completed by a total of 1,559 nurses. Hierarchical linear regression analyses and the Sobel test were performed to determine the mediating role of coping styles on the relationship between CSE and burnout. RESULTS: Nurses who had higher self-evaluation characteristics, reported less emotional exhaustion and cynicism, and higher professional efficacy. Coping style had a partial mediating effect on the relationship between CSE and the burnout syndrome among nurses. CONCLUSIONS: Core self-evaluation had effects on burnout and coping style was a mediating factor in this relationship among Chinese nurses. Therefore, the improvement of coping strategies may be helpful in the prevention of burnout among nurses, thus enhancing professional performance.

  9. Core self-evaluation and burnout among Nurses: the mediating role of coping styles.

    Science.gov (United States)

    Li, Xiaofei; Guan, Lili; Chang, Hui; Zhang, Bo

    2014-01-01

    This study aimed to determine the potential association between core self-evaluation and the burnout syndrome among Chinese nurses, and the mediating role of coping styles in this relationship. A cross-sectional survey was conducted in Shenyang, China, from May to July, 2013. A questionnaire which consisted of the Maslach Burnout Inventory-General Survey (MBI-GS), the Core Self-Evaluation Scale (CSE), and the Simplified Coping Style Questionnaire (CSQ), was completed by a total of 1,559 nurses. Hierarchical linear regression analyses and the Sobel test were performed to determine the mediating role of coping styles on the relationship between CSE and burnout. Nurses who had higher self-evaluation characteristics, reported less emotional exhaustion and cynicism, and higher professional efficacy. Coping style had a partial mediating effect on the relationship between CSE and the burnout syndrome among nurses. Core self-evaluation had effects on burnout and coping style was a mediating factor in this relationship among Chinese nurses. Therefore, the improvement of coping strategies may be helpful in the prevention of burnout among nurses, thus enhancing professional performance.

  10. Core Self-Evaluations, Worry, Life Satisfaction, and Psychological Well-Being: An Investigation in the Asian Context

    Science.gov (United States)

    Rathi, Neerpal; Lee, Kidong

    2018-01-01

    The concept of core self-evaluations has been extensively investigated in Western and European countries, nonetheless its implications in Asian countries remains relatively unexplored. To void this gap, the current study investigated the association of core self-evaluations with worry, life satisfaction, and psychological well-being among South…

  11. The Role of Personality Traits, Core Self-Evaluation, and Emotional Intelligence in Career Decision-Making Difficulties

    Science.gov (United States)

    Di Fabio, Annamaria; Palazzeschi, Letizia; Bar-On, Reuven

    2012-01-01

    This study examines the role of personality traits, core self-evaluation, and emotional intelligence (EI) in career decision-making difficulties. Italian university students (N = 232) responded to questions on the Big Five Questionnaire, Core Self-Evaluation Scale, Bar-On Emotional Quotient Inventory, and Career Decision-Making Difficulties…

  12. Formation evaluation in Devonian shale through application of new core and log analysis methods

    International Nuclear Information System (INIS)

    Luffel, D.L.; Guidry, F.K.

    1990-01-01

    In the Devonian shale of the Appalachian Basin all porosity in excess of about 2.5 percent is generally occupied by free hydrocarbons, which is mostly gas, based on results of new core and log analysis methods. In this study, sponsored by the Gas Research Institute, reservoir porosities averaged about 5 percent and free gas content averaged about 2 percent by bulk volume, based on analyses on 519 feet of conventional core in four wells. In this source-rich Devonian shale, which also provides the reservoir storage, the rock everywhere appears to be at connate, or irreducible, water saturation corresponding to two or three percent of bulk volume. This became evident when applying the new core and log analysis methods, along with a new plotting method relating bulk volume of pore fluids to porosity. This plotting method has proved to be a valuable tool: it provides useful insight on the fluid distribution present in the reservoir, it provides a clear idea of porosity required to store free hydrocarbons, it leads to a method of linking formation factor to porosity, and it provides a good quality control method to monitor core and log analysis results. In the Devonian shale an important part of the formation evaluation is to determine the amount of kerogen, since this appears as hydrocarbon-filled porosity to conventional logs. In this study Total Organic Carbon and pyrolysis analyses were made on 93 core samples from four wells. Based on these data a new method was used to drive volumetric kerogen and free oil content, and kerogen was found to range up to 26 percent by volume. A good correlation was subsequently developed to derive kerogen from the uranium response of the spectral gamma ray log. Another important result of this study is the measurement of formation water salinity directly on core samples. Results on 50 measurements in the four study wells ranged from 19,000 to 220,000 ppm NaCl

  13. Ex-vessel core catcher design requirements and preliminary concepts evaluation

    International Nuclear Information System (INIS)

    Friedland, A.J.; Tilbrook, R.W.

    1974-01-01

    As part of the overall study of the consequences of a hypothetical failure to scram following loss of pumping power, design requirements and preliminary concepts evaluation of an ex-vessel core catcher (EVCC) were performed. EVCC is the term applied to a class of devices whose primary objective is to provide a stable subcritical and coolable configuration within containment following a postulated accident in which it is assumed that core debris has penetrated the Reactor Vessel and Guard Vessel. Under these assumed conditions a set of functional requirements were developed for an EVCC and several concepts were evaluated. The studies were specifically directed toward the FFTF design considering the restraints imposed by the physical design and construction of the FFTF plant

  14. Effects of core self-evaluations on the job burnout of nurses: the mediator of organizational commitment.

    Science.gov (United States)

    Zhou, Yangen; Lu, Jiamei; Liu, Xianmin; Zhang, Pengcheng; Chen, Wuying

    2014-01-01

    To explore the impact of Core self-evaluations on job burnout of nurses, and especially to test and verify the mediator role of organizational commitment between the two variables. Random cluster sampling was used to pick up participants sample, which consisted of 445 nurses of a hospital in Shanghai. Core self-evaluations questionnaire, job burnout scale and organizational commitment scale were administrated to the study participants. There are significant relationships between Core self-evaluations and dimensions of job burnout and organizational commitment. There is a significant mediation effect of organizational commitment between Core self-evaluations and job burnout. To enhance nurses' Core self-evaluations can reduce the incidence of job burnout.

  15. Core Self-Evaluations and Job Satisfaction: The Role of Organizational and Community Embeddedness.

    OpenAIRE

    Oyler, Jennifer D.

    2007-01-01

    This study extends job embeddedness and job satisfaction theory in several notable directions. As hypothesized, structural equation modeling revealed that community embeddedness was a partial mediator of the relationship between core self-evaluations and job satisfaction. Contrary to job embeddedness theory, this study found that organizational embeddedness and job satisfaction were best represented by a single latent factor. Thus, organizational embeddedness did not act as a mediator of the ...

  16. Core Abilities Evaluation Index System Exploration and Empirical Study on Distributed PV-Generation Projects

    Directory of Open Access Journals (Sweden)

    Lin He

    2017-12-01

    Full Text Available In line with the constraints of environmental problems and economic development, large-scale renewable-generation projects have been planned and constructed in recent years. In order to achieve sustainable power development and improve the power supply structure, China’s government has focused on distributed photovoltaic (PV generation projects due to their advantages of clean emission and local consumption. However, their unstable output power still brings a series of problems concerning reliability, investment income, and available substitution proportion to traditional power, and so on. Therefore, it is imperative to understand the competitive development abilities of distributed PV generation projects and measure them effectively. First, through various investigation methods such as literature reviews, feasibility report analysis and expert interviews, the factors that influence the core abilities of distributed PV-generation projects were explored based on the micro-grid structure. Then, with the indexed exploration results, the factors were classified into 6 dimensions, i.e., investment and earning ability, production and operation ability, power-grid coordination ability, energy-conservation and emission-reduction ability, sustainable development ability, and society-serving ability. Meanwhile, an evaluation index system for core abilities of distributed PV-generation project was constructed using all quantitative indicators. Third, for examining the availability of the evaluation index system, combination weighting and techniques for order preference by similarity to an ideal solution (TOPSIS methods were adopted to assess the practical distributed PV-generation projects. The case study results showed that installed capacity, local economy development, and grid-connected power quantity will influence the core abilities of distributed PV-generation project, obviously. The conclusions of the evaluation analysis on core abilities can

  17. The Effect of Luting Cement and Titanium Base on the Final Color of Zirconium Oxide Core Material.

    Science.gov (United States)

    Capa, Nuray; Tuncel, Ilkin; Tak, Onjen; Usumez, Aslihan

    2017-02-01

    To evaluate the effects of different types of luting cements and different colors of zirconium cores on the final color of the restoration that simulates implant-supported fixed partial dentures (FPDs) by using a titanium base on the bottom. One hundred and twenty zirconium oxide core plates (Zr-Zahn; 10 mm in width, 5 mm in length, 0.5 mm in height) were prepared in different shades (n = 20; noncolored, A2, A3, B1, C2, D2). The specimens were subdivided into two subgroups for the two types of luting cements (n = 10). The initial color measurements were made on zirconium oxide core plates using a spectrometer. To create the cement thicknesses, stretch strips with holes in the middle (5 mm in diameter, 70 μm in height) were used. The second measurement was done on the zirconium oxide core plates after the application of the resin cement (U-200, A2 Shade) or polycarboxylate cement (Lumicon). The final measurement was done after placing the titanium discs (5 mm in diameter, 3 mm in height) in the bottom. The data were analyzed with two-way ANOVA and Tukey's honestly significant differences (HSD) tests (α = 0.05). The ∆E* ab value was higher in the resin cement-applied group than in the polycarboxylate cement-applied group (p zirconium oxide core-resin cement-titanium base, and the lowest was recorded for the polycarboxylate cement-zirconium oxide core (p zirconium are all important factors that determine the final shade of zirconia cores in implant-supported FPDs. © 2015 by the American College of Prosthodontists.

  18. The accretion of solar material onto white dwarfs: No mixing with core material implies that the mass of the white dwarf is increasing

    Directory of Open Access Journals (Sweden)

    Sumner Starrfield

    2014-02-01

    Full Text Available Cataclysmic Variables (CVs are close binary star systems with one component a white dwarf (WD and the other a larger cooler star that fills its Roche Lobe. The cooler star is losing mass through the inner Lagrangian point of the binary and some unknown fraction of this material is accreted by the WD. One consequence of the WDs accreting material, is the possibility that they are growing in mass and will eventually reach the Chandrasekhar Limit. This evolution could result in a Supernova Ia (SN Ia explosion and is designated the Single Degenerate Progenitor (SD scenario. This paper is concerned with the SD scenario for SN Ia progenitors. One problem with the single degenerate scenario is that it is generally assumed that the accreting material mixes with WD core material at some time during the accretion phase of evolution and, since the typical WD has a carbon-oxygen CO core, the mixing results in large amounts of carbon and oxygen being brought up into the accreted layers. The presence of enriched carbon causes enhanced nuclear fusion and a Classical Nova explosion. Both observations and theoretical studies of these explosions imply that more mass is ejected than is accreted. Thus, the WD in a Classical Nova system is losing mass and cannot be a SN Ia progenitor. However, the composition in the nuclear burning region is important and, in new calculations reported here, the consequences to the WD of no mixing of accreted material with core material have been investigated so that the material involved in the explosion has only a Solar composition. WDs with a large range in initial masses and mass accretion rates have been evolved. I find that once sufficient material has been accreted, nuclear burning occurs in all evolutionary sequences and continues until a thermonuclear runaway (TNR occurs and the WD either ejects a small amount of material or its radius grows to about 1012 cm and the evolution is ended. In all cases where mass ejection occurs

  19. Evaluation of materials and design modifications for aircraft brakes

    Science.gov (United States)

    Ho, T. L.; Kennedy, F. E.; Peterson, M. B.

    1975-01-01

    A test program is described which was carried out to evaluate several proposed design modifications and several high-temperature friction materials for use in aircraft disk brakes. The evaluation program was carried out on a specially built test apparatus utilizing a disk brake and wheel half from a small het aircraft. The apparatus enabled control of brake pressure, velocity, and braking time. Tests were run under both constant and variable velocity conditions and covered a kinetic energy range similar to that encountered in aircraft brake service. The results of the design evaluation program showed that some improvement in brake performance can be realized by making design changes in the components of the brake containing friction material. The materials evaluation showed that two friction materials show potential for use in aircraft disk brakes. One of the materials is a nickel-based sintered composite, while the other is a molybdenum-based material. Both materials show much lower wear rates than conventional copper-based materials and are better able to withstand the high temperatures encountered during braking. Additional materials improvement is necessary since both materials show a significant negative slope of the friction-velocity curve at low velocities.

  20. Core II Materials for Rural Agriculture Programs. Units E-H.

    Science.gov (United States)

    Biondo, Ron; And Others

    This curriculum guide includes teaching packets for 21 problem areas to be included in a core curriculum for 10th grade students enrolled in a rural agricultural program. Covered in the four units included in this volume are crop science (harvesting farm crops and growing small grains); soil science and conservation of natural resources…

  1. CFD Validation with a Multi-Block Experiment to Evaluate the Core Bypass Flow in VHTR

    International Nuclear Information System (INIS)

    Yoon, Su Jong; Lee, Jeong Hun; Park, Goon Cherl; Kim, Min Hwan

    2010-01-01

    Core bypass flow of Very High Temperature Reactor (VHTR) is defined as the ineffective coolant which passes through the bypass gaps between the block columns and the crossflow gaps between the stacked blocks. This flows lead to the variation of the flow distribution in the core and affect the core thermal margin and the safety of VHTR. Therefore, bypass flow should be investigated and quantified. However, it is not a simple question, because the flow path of VHTR core is very complex. In particular, since dimensions of the bypass gap and the crossflow gap are of the order of few millimeters, it is very difficult to measure and to analyze the flow field at those gaps. Seoul National University (SNU) multi-block experiment was carried out to evaluate the bypass flow distribution and the flow characteristics. The coolant flow rate through outlet of each block column was measured, but the local flow field was measured restrictively in the experiment. Instead, CFD analysis was carried out to investigate the local phenomena of the experiment. A commercial CFD code CFX-12 was validated by comparing the simulation results and the experimental data

  2. Fracture toughness evaluation of select advanced replacement alloys for LWR core internals

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chen, Xiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Fracture toughness is one of the key engineering properties required for core internal materials. Together with other properties, which are being examined such as high-temperature steam oxidation resistance, radiation hardening, and irradiation-assisted stress corrosion cracking resistance, the alloys will be down-selected for neutron irradiation study and comprehensive post-irradiation examinations. According to the candidate alloys selected under the ARRM program, ductile fracture toughness of eight alloys was evaluated at room temperature and the LWR-relevant temperatures. The tested alloys include two ferritic alloys (Grade 92 and an oxide-dispersion-strengthened alloy 14YWT), two austenitic stainless steels (316L and 310), four Ni-base superalloys (718A, 725, 690, and X750). Alloy 316L and X750 are included as reference alloys for low- and high-strength alloys, respectively. Compact tension specimens in 0.25T and 0.2T were machined from the alloys in the T-L and R-L orientations according to the product forms of the alloys. This report summarizes the final results of the specimens tested and analyzed per ASTM Standard E1820. Unlike the

  3. Evaluation report on CCTF core-I reflood test C1-5 (Run 14)

    International Nuclear Information System (INIS)

    Murao, Yoshio; Akimoto, Hajime; Sudoh, Takashi; Okubo, Tsutomu

    1983-02-01

    A study of a cylindrical core test facility (CCTF) test was performed for modeling the system behavior during the reflood phase of a PWR-LOCA and the following conclusions were obtained: 1) With the exception of some points, the observed phenomena are similar to a model derived from an evaluation model for a PWR safety evaluation. 2) The different points are the water accumulation in the upper plenum, the ECC bypass in the downcomer, the reduction of the effective downcomer head and the pressure drop at the broken cold leg nozzle and in the interconnected pipes. (author)

  4. Influence of Parameters of Core Bingham Material on Critical Behaviour of Three-Layered Annular Plate

    Directory of Open Access Journals (Sweden)

    Pawlus Dorota

    2017-12-01

    Full Text Available The paper presents the dynamic response of annular three-layered plate subjected to loads variable in time. The plate is loaded in the plane of outer layers. The plate core has the electrorheological properties expressed by the Bingham body model. The dynamic stability loss of plate with elastic core is determined by the critical state parameters, particularly by the critical stresses. Numerous numerical observations show the influence of the values of viscosity constant and critical shear stresses, being the Bingham body parameters, on the supercritical viscous fluid plate behaviour. The problem has been solved analytically and numerically using the orthogonalization method and finite difference method. The solution includes both axisymmetric and asymmetric plate dynamic modes.

  5. First evaluation of low frequency noise measurements of in core detector signals in the measuring assembly Rheinsberg

    International Nuclear Information System (INIS)

    Collatz, S.

    1982-01-01

    Reactor noise spectra of in core neutron detectors are measured in the low frequency range (0.03 Hz to 1 Hz) and evaluated. The increase of the effective noise signal value is due to pressure oscillations or oscillations of special steam volume portions. Thus boiling monitoring of reactor cores in PWR type reactors may be possible, if the low frequency noise of the whole set of in core detectors is taken into account

  6. Conceptual core design study for Japan sodium-cooled fast reactor: Review of sodium void reactivity worth evaluation

    International Nuclear Information System (INIS)

    Ohki, Shigeo

    2012-01-01

    The conceptual core design study for a large-scale Japan sodium-cooled fast reactor (JSFR) have been carried out in the framework of the FaCT project. The reference “High-internal conversion” core can satisfy the requirements for enhanced safety, as well as achieving economic competitiveness. In order to increase the design reliability, more rigorous uncertainty evaluation is important. Development of the verification and validation methodology of the core neutronic design method is currently underway. (author)

  7. Influence of reflector materials and core coolant on the characteristics of accelerator driven systems

    OpenAIRE

    Panza, Fabio; Osipenko, Michail; Ricco, Giovanni; Ripani, Marco; Saracco, Paolo

    2016-01-01

    In this paper we simulated the behavior of a simple ADS model, based on MOX fuel embedded in solid lead, in terms of multiplication coefficient keff, thermal power and absolute neutron spectra. In the first part of the paper, we report on the results obtained when modifying the reflector surrounding the fission core, by replacing pure lead with a layered graphite/lead structure. We found that, by appropriately choosing position and thickness of the graphite and lead layers, it is possible to ...

  8. Test and evaluation of pressure vessel materials

    International Nuclear Information System (INIS)

    Choi, Sun Pil; Hong, Jun Hwa; Nho, Kye Hoe; Han, Dae June; Chi, Se Hwan

    1985-01-01

    We have prepared a method for analyzing the Charpy impact test data, which is deduced from ''the standard anelastic solid equation''. The theoretical expression for the absorbed energy is in a form of W=Wsub(U)+(Wsub(R)-Wsub(U))/ [1+(ωtau) 2 ] showing the Debye characteristics and where tau is given by the Arrhenius equation; tau=tau 0 exp(ΔH/ksub(B)T). Four measurable parameters, at the present stage, can characterize the dynamic hehavior of cracking (Charpy impact result). They are the upper shelf energy(Wsub(R), the lower shelf energy (Wsub(U)), the activation energy of crack (ΔH, and wtau(0) where w tau(0) are the resonance frequency of the specimen and the jumping pre-exponential factor of propagating crack respectively. However the states of R (relaxed) and U (un-relaxed) should be defined from reasonable physical conditions in the future and it is possible that Wsub(U) is small enough to be taken as zero. The effects of irradiation, alloying elements, and heat treatment on the impact results should be interpreted as changes in the above characteristic parameters. The present method has been applied for weld metal of SA 508-2 irradiated up to a fluence of 4x10 18 n/cm 2 , E>1.0Mev, resulting in about 29% decrease in Wsub(R), negligible change in Wsub(U), 5.6 times increase in ωtau 0 , and no change in ΔH. This seems to indicate that irradiation degrades an average value of YOUNG's modulus so that cracks propagate more easily and it does not effect on breaking the lattice bond. However much more systematic analyses should be necessary for correct judgment. It is concluded that the present method is quite adequate for analyzing the Charpy impact data even though plastic deformation in the specimen was not considered separately so that the method should be applied for various cases in order to evaluate the proper trend of effects of irradiation, alloying elements, and heat treatment on the Charpy impact results. (Author)

  9. A reactor core/containment status evaluation flowchart for determining protective actions in emergencies

    International Nuclear Information System (INIS)

    Glissman, M.A.

    1988-01-01

    In the event of an emergency at a power reactor station, there might not be adequate time or sufficient data to fully assess radiological implications and make protective action recommendations based on projected population exposures. Thus, decision-making guidance is needed that is based on readily available plant indicators, not just on time-consuming dose calculations. In the United States, this guidance must be compatible with the recommended by the Nuclear Regulatory Commission and the Environmental Protection Agency, and it must include predetermined, measurable, site-specific parameters for assessing conditions in the reactor core and containment. The preparation of this real time guidance calls for the selection of suitable parameters and the determination of the values for these parameters that will correspond to different levels of protective action. This process is illustrated in this paper by selecting parameters and determining appropriate values for constructing a Core/Containment Status Evaluation Flowchart for an example power plant

  10. Evaluation of local power distribution with fine-mesh core model for the HTTR

    International Nuclear Information System (INIS)

    Murata, Isao; Yamashita, Kiyonobu; Maruyama, So; Shindo, Ryuichi; Fujimoto, Nozomu; Sudo, Yukio; Nakata, Tetsuo.

    1991-01-01

    An evaluation method of the local power distribution was developed considering the radial and axial heterogeneity caused by fuel rods, BP rods and block end graphite for the High Temperature Engineering Test Reactor (HTTR) in Japan Atomic Energy Research Institute (JAERI). The evaluation method was verified through the analyses of critical assembly experiments. A good agreement was obtained between calculations and measurements and the difference was less than 3 % on the power distribution. This method was applied to the core design for the HTTR to evaluate the maximum fuel temperature. From these results, it was confirmed that this evaluation method has an enough accuracy and is able to predict the detailed power distribution of the HTTR. (author)

  11. A core laboratory offering full evaluation of new boron compounds. A service to the BNCT community

    International Nuclear Information System (INIS)

    Zamenhof, R.G.; Patel, H.; Palmer, M.R.; Lin, H.C.; Busse, P.M.; Harling, O.; Binns, P.J.; Riley, K.J.; Bernard, J.

    2000-01-01

    A joint project by the Beth Israel Deaconess Medical Center at Harvard Medical School and The Nuclear Reactor Laboratory of the Massachusetts Institute of Technology is proposed which would provide a core laboratory for the evaluation of new boron compounds. Federal agency funding has been applied for to support such a facility. The facility's evaluation of candidate boron compounds will include: quantitative cellular boron uptake; cell survival curve analysis (using a thermal neutron beam); small or large animal pharmacokinetic analysis; macro- and micro boron distribution analysis using high-resolution autoradiography, prompt gamma analysis and ICP-AES; small or large animal in vivo tumor control studies (using thermal or epithermal neutron beams); and pharmacological in vivo toxicity evaluation. The laboratory will include small and large animal surgical facilities and resources for additional boron compound chemistry as required by the evaluation procedure. This facility will be open to the BNCT research community. (author)

  12. BWRVIP-123, Revision 1NP: BWR Vessel and Internals Project Removal and Analysis of Material Samples from Core Shroud and Top Guide at Susquehanna Unit 2

    International Nuclear Information System (INIS)

    Howell, D.; Haertel, T.; Lindberg, J.; Oliver, B.; Greenwood, L.

    2005-01-01

    Fast and thermal fluence were determined by a laboratory analysis of the samples. Fluence in the upper regions of the shroud (between the H1 and H2 welds) was substantially lower than that in the belt line region (near the H4 weld). Fluence in the top guide was significantly higher than fluence on the core shroud. As expected, helium concentrations were highest in regions where fluence was highest. Estimates of the initial boron concentration were similar to measurements made on materials removed from other reactors. A technical justification evaluated the acceptability of the sampling process with respect to structural consequences of material removal and to increased cracking susceptibility due to the as-left condition. It was determined that the sampling process was acceptable on both counts

  13. Trackless tack coat materials : a laboratory evaluation performance acceptance.

    Science.gov (United States)

    2012-06-01

    The purpose of this study was to develop, demonstrate, and document laboratory procedures that could be used by the : Virginia Department of Transportation (VDOT) to evaluate non-tracking tack coat materials. The procedures would be used to : qualify...

  14. evaluation of teachers' use of instructional materials for teaching ...

    African Journals Online (AJOL)

    Global Journal

    The purpose of this survey was to evaluate the teachers' use of the instructional materials for teaching ... These and other benefits justify the teaching of .... The use of ICT is very effective for foreign language teaching and learning. 40. PRISCA ...

  15. Dredged Material Testing and Evaluation for Ocean Disposal

    Science.gov (United States)

    Evaluation and testing of dredged material proposed for ocean dumping is conducted to help protect human health and the marine environment. National guidance is provided by the Green Book. Regional Implementation Manuals are provided.

  16. Ultrasonic and radiographic evaluation of advanced aerospace materials: Ceramic composites

    Science.gov (United States)

    Generazio, Edward R.

    1990-01-01

    Two conventional nondestructive evaluation techniques were used to evaluate advanced ceramic composite materials. It was shown that neither ultrasonic C-scan nor radiographic imaging can individually provide sufficient data for an accurate nondestructive evaluation. Both ultrasonic C-scan and conventional radiographic imaging are required for preliminary evaluation of these complex systems. The material variations that were identified by these two techniques are porosity, delaminations, bond quality between laminae, fiber alignment, fiber registration, fiber parallelism, and processing density flaws. The degree of bonding between fiber and matrix cannot be determined by either of these methods. An alternative ultrasonic technique, angular power spectrum scanning (APSS) is recommended for quantification of this interfacial bond.

  17. Description and hydrogeologic implications of cored sedimentary material from the 1975 drilling program at the Radioactive Waste Management Complex, Idaho

    International Nuclear Information System (INIS)

    Rightmire, C.T.

    1984-08-01

    Samples of sedimentary material from interbeds between basalt flows and from fractures in the flows, taken from two drill cores at the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory were analyzed for (1) particle-size distribution, (2) bulk mineralogy, (3) clay mineralogy, (4) cation-exchange capacity, and (5) carbonate content. Thin sections of selected sedimentary material were made for petrographic examination. These analyses are needed for a characterization of paths and rates of movement of radionuclides transported by infiltrating water. Preliminary interpretations indicate that (1) it may be possible to distinguish the various sedimentary interbeds on the basis of their mineralogy, (2) the presence of carbonate horizons in sedimentary interbeds may be utilized to approximate the time of exposure and the climate while the surface was exposed, and (3) the type and orientation of fracture-filling material may be utilized to determine the mechanism by which fractures were filled. 9 references, 14 figures, 8 tables

  18. Evaluation on electrical resistivity of silicon materials after electron ...

    Indian Academy of Sciences (India)

    Home; Journals; Bulletin of Materials Science; Volume 38; Issue 5. Evaluation on ... This research deals with the study of electron beam melting (EBM) methodology utilized in melting silicon material and subsequently discusses on the effect of oxygen level on electrical resistivity change after EBM process. The oxygen ...

  19. Fracture Resistance of Endodontically Treated Teeth Restored with 2 Different Fiber-reinforced Composite and 2 Conventional Composite Resin Core Buildup Materials: An In Vitro Study.

    Science.gov (United States)

    Eapen, Ashly Mary; Amirtharaj, L Vijay; Sanjeev, Kavitha; Mahalaxmi, Sekar

    2017-09-01

    The purpose of this in vitro study was to comparatively evaluate the fracture resistance of endodontically treated teeth restored with 2 fiber-reinforced composite resins and 2 conventional composite resin core buildup materials. Sixty noncarious unrestored human maxillary premolars were collected, endodontically treated (except group 1, negative control), and randomly divided into 5 groups (n = 10). Group 2 was the positive control. The remaining 40 prepared teeth were restored with various direct core buildup materials as follows: group 3 teeth were restored with dual-cure composite resin, group 4 with posterior composite resin, group 5 with fiber-reinforced composite resin, and group 6 with short fiber-reinforced composite resin. Fracture strength testing was performed using a universal testing machine. The results were statistically analyzed by 1-way analysis of variance and the post hoc Tukey test. Fracture patterns for each sample were also examined under a light microscope to determine the level of fractures. The mean fracture resistance values (in newtons) were obtained as group 1 > group 6 > group 4 > group 3 > group 5 > group 2. Group 6 showed the highest mean fracture resistance value, which was significantly higher than the other experimental groups, and all the fractures occurred at the level of enamel. Within the limitations of this study, a short fiber-reinforced composite can be used as a direct core buildup material that can effectively resist heavy occlusal forces against fracture and may reinforce the remaining tooth structure in endodontically treated teeth. Copyright © 2017 American Association of Endodontists. Published by Elsevier Inc. All rights reserved.

  20. Evaluation of the Leon3 soft-core processor within a Xilinx radiation-hardened field-programmable gate array.

    Energy Technology Data Exchange (ETDEWEB)

    Learn, Mark Walter

    2012-01-01

    The purpose of this document is to summarize the work done to evaluate the performance of the Leon3 soft-core processor in a radiation environment while instantiated in a radiation-hardened static random-access memory based field-programmable gate array. This evaluation will look at the differences between two soft-core processors: the open-source Leon3 core and the fault-tolerant Leon3 core. Radiation testing of these two cores was conducted at the Texas A&M University Cyclotron facility and Lawrence Berkeley National Laboratory. The results of these tests are included within the report along with designs intended to improve the mitigation of the open-source Leon3. The test setup used for evaluating both versions of the Leon3 is also included within this document.

  1. Fracture resistance of endodontically treated teeth restored with Zirconia filler containing composite core material and fiber posts.

    Science.gov (United States)

    Jeaidi, Zaid Al

    2016-01-01

    To assess the fracture resistance of endodontically treated teeth with a novel Zirconia (Zr) nano-particle filler containing bulk fill resin composite. Forty-five freshly extracted maxillary central incisors were endodontically treated using conventional step back preparation and warm lateral condensation filling. Post space preparation was performed using drills compatible for fiber posts (Rely X Fiber Post) on all teeth (n=45), and posts were cemented using self etch resin cement (Rely X Unicem). Samples were equally divided into three groups (n=15) based on the type of core materials, ZirconCore (ZC) MulticCore Flow (MC) and Luxacore Dual (LC). All specimens were mounted in acrylic resin and loads were applied (Universal testing machine) at 130° to the long axis of teeth, at a crosshead speed of 0.5 mm/min until failure. The loads and the site at which the failures occurred were recorded. Data obtained was tabulated and analyzed using a statistical program. The means and standard deviations were compared using ANOVA and Multiple comparisons test. The lowest and highest failure loads were shown by groups LC (18.741±3.02) and MC (25.16±3.30) respectively. Group LC (18.741±3.02) showed significantly lower failure loads compared to groups ZC (23.02±4.21) and MC (25.16±3.30) (pcomposite cores was comparable to teeth restored with conventional Zr free bulk fill composites. Zr filled bulk fill composites are recommended for restoration of endodontically treated teeth as they show comparable fracture resistance to conventional composite materials with less catastrophic failures.

  2. Flowing and freezing of molten core materials during unprotected loss of flow accidents in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Maschek, W.; Royl, P.

    1988-09-01

    Flowing and freezing of mobile core materials change the fissile material distribution and core-inventory under hypothetical accident conditions and determine the path to permanent shutdown of the neutronic events and the energetic potentials. The report classifies the bondary conditions for such flowing and freezing processes by going through the different situations under which these processes can occur in the scenario of the unprotected loss of flow (ULOF) accident. The classification is based on ULOF-accident simulations for a homogeneous reactor core concept of a 300 MWe LMFBR (e. g. SNR-300), but many boundary conditions are also characteristic for other core designs. A review of the relevant experiments is then made to correlate the available experimental information with these classified boundary conditions and to look at the resulting flowing and freezing processes. Boundary conditions that have been experimentally shown to be important are assigned high priorities. The data are specifically valued in relation to these boundary conditions of high priorities. The review includes the major experimental programs with published results. The discussion shows that the results from most clean condition tests for melt relocations are valuable for a better understanding of basic phenomena and analytical model development, but are not directly applicable to real accident conditions. The database for relevant boundary conditions from the ULOF scenario is limited and largely included in integral sequence tests from which quantitative information for modelling is difficult to obtain. Needs for additional investigations are identified. The suggestions are mainly restricted to investigations of the early phase of fuel removal. They are given with reference to candidate facilities and include relocations in the subassemblies and in the inter-subassembly gaps. Particular emphasis is put on the leading edge properties and possible driving forces to which more attention

  3. Evaluation of core compositions for use in breed and burn reactors and limited-separations fuel cycles

    International Nuclear Information System (INIS)

    Petroski, Robert; Forget, Benoit; Forsberg, Charles

    2013-01-01

    Highlights: ► Calculated minimum burnup and irradiation damage for B and B reactor compositions. ► Computed doubling time of fuel cycles using B and B reactors and no chemical separations. ► Determined sensitivity of doubling time to using melt refining vs. direct reuse. ► Examined tradeoff between power density and neutronics for different coolants. - Abstract: Previously developed methods for analyzing breed-and-burn (B and B) reactors are applied to a wide range of core compositions. The compositions studied include different fuel types, steel and silicon carbide structure, and sodium, lead/lead bismuth eutectic (LBE), and gas coolants. These compositions are evaluated for use in “minimum burnup” B and B reactors in which it is assumed that blocks comprising the core can be shuffled in all three dimensions to flatten out non-uniformities in burnup. The two figures of merit evaluated are the minimum irradiation damage requirement and reactor fleet doubling time. To minimize irradiation damage, gas coolants perform best, followed by lead/LBE then sodium. High uranium-content metal fuel outperforms compound fuels, and different types of steel are similar and perform slightly better than silicon carbide. Once-through irradiation damage requirements can be surprisingly modest in minimum burnup B and B reactors, with a wide range of compositions viable at irradiation damage levels 50% higher than existing materials data. Doubling times were calculated for a reactor fleet consisting of B and B reactors operating in a limited-separations fuel cycle; i.e., a fuel cycle with no chemical separation of actinides. The effects of different cooling times and removal of fission products using a melt refining process are evaluated. To minimize doubling time, sodium cooled compositions perform best because they are able to achieve core power densities several times larger than compositions using other coolants. A hypothetical sodium-cooled core composition with high

  4. Fabrication, characterization and comparison of composite magnetic materials for high efficiency integrated voltage regulators with embedded magnetic core micro-inductors

    International Nuclear Information System (INIS)

    Bellaredj, Mohamed L F; Mueller, Sebastian; Davis, Anto K; Swaminathan, Madhavan; Mano, Yasuhiko; Kohl, Paul A

    2017-01-01

    High-efficiency integrated voltage regulators (IVRs) require the integration of power inductors, which have low loss and reduced size at very high frequency. The use of a magnetic material core can reduce significantly the inductor area and simultaneously increase the inductance. This paper focuses on the fabrication, characterization and modeling of nickel zinc (NiZn) ferrite and carbonyl iron powder (CIP)-epoxy magnetic composite materials, which are used as the magnetic core materials of embedded inductors in a printed wiring board (PWB) for a system in package (SIP) based buck type IVR. The fabricated composite materials and process are fully compatible with FR4 epoxy resin prepreg and laminate. For 85% weight loading of the magnetic powder (around 100 MHz at room temperature), the composite materials show a relative permeability of 7.5–8.1 for the NiZn ferrite composite and 5.2–5.6 for the CIP composite and a loss tangent value of 0.24–0.28 for the NiZn ferrite composite and 0.09–0.1 for the CIP-composite. The room temperature saturation flux density values are 0.1351 T and 0.5280 T for the NiZn ferrite and the CIP composites, respectively. The frequency dispersion parameters of the magnetic composites are modeled using a simplified Lorentz and Landau–Lifshitz–Gilbert equation for a Debye type relaxation. Embedded magnetic core solenoid inductors were designed based on the composite materials for the output filter of a high-efficiency SIP based buck type IVR. Evaluation of a SIP based buck type IVR with the designed inductors shows that it can reach peak efficiencies of 91.7% at 11 MHz for the NiZn ferrite-composite, 91.6% at 14 MHz for CIP-composite and 87.5% (NiZn ferrite-composite) and 87.3% (CIP-composite) efficiency at 100 MHz for a 1.7 V:1.05 V conversion. For a direct 5 V:1 V conversion using a stacked topology, a peak efficiency of 82% at 10 MHz and 72% efficiency at 100 MHz can be achieved for both materials. (paper)

  5. Fabrication, characterization and comparison of composite magnetic materials for high efficiency integrated voltage regulators with embedded magnetic core micro-inductors

    Science.gov (United States)

    Bellaredj, Mohamed L. F.; Mueller, Sebastian; Davis, Anto K.; Mano, Yasuhiko; Kohl, Paul A.; Swaminathan, Madhavan

    2017-11-01

    High-efficiency integrated voltage regulators (IVRs) require the integration of power inductors, which have low loss and reduced size at very high frequency. The use of a magnetic material core can reduce significantly the inductor area and simultaneously increase the inductance. This paper focuses on the fabrication, characterization and modeling of nickel zinc (NiZn) ferrite and carbonyl iron powder (CIP)-epoxy magnetic composite materials, which are used as the magnetic core materials of embedded inductors in a printed wiring board (PWB) for a system in package (SIP) based buck type IVR. The fabricated composite materials and process are fully compatible with FR4 epoxy resin prepreg and laminate. For 85% weight loading of the magnetic powder (around 100 MHz at room temperature), the composite materials show a relative permeability of 7.5-8.1 for the NiZn ferrite composite and 5.2-5.6 for the CIP composite and a loss tangent value of 0.24-0.28 for the NiZn ferrite composite and 0.09-0.1 for the CIP-composite. The room temperature saturation flux density values are 0.1351 T and 0.5280 T for the NiZn ferrite and the CIP composites, respectively. The frequency dispersion parameters of the magnetic composites are modeled using a simplified Lorentz and Landau-Lifshitz-Gilbert equation for a Debye type relaxation. Embedded magnetic core solenoid inductors were designed based on the composite materials for the output filter of a high-efficiency SIP based buck type IVR. Evaluation of a SIP based buck type IVR with the designed inductors shows that it can reach peak efficiencies of 91.7% at 11 MHz for the NiZn ferrite-composite, 91.6% at 14 MHz for CIP-composite and 87.5% (NiZn ferrite-composite) and 87.3% (CIP-composite) efficiency at 100 MHz for a 1.7 V:1.05 V conversion. For a direct 5 V:1 V conversion using a stacked topology, a peak efficiency of 82% at 10 MHz and 72% efficiency at 100 MHz can be achieved for both materials.

  6. Streamlined analysis technique for the evaluation of pellet clad interaction in PWR reload cores

    International Nuclear Information System (INIS)

    Beard, Ch.; Morita, T.; Brown, J.

    2007-01-01

    For some applications, an analysis is required to explicitly demonstrate that fuel failure due to pellet-clad interaction (PCI) is prevented by the core limits and the protection system for both Condition I (normal operation) operation and for Condition II (events of moderate frequency) events. This analysis needs to address the entire range of normal operation allowed by the Technical Specifications and all Condition II transients. The obvious approach which has been utilized for many years is a simulation of normal operation power maneuvers followed by explicit Condition II transients as a function of key core parameters. This is a sampling approach and has concerns about the overall coverage of the potential space. An alternative approach is the 3D FAC power distribution analysis methodology that was based upon the Westinghouse Relaxed Axial Offset Control Strategy (RAOC) evaluation process. The 3D FAC methodology uses a parametric representation of variables affecting the power distributions, defining a grid mesh over a space of Condition I and Condition II parameters. The operation space is defined by a power range, temperature range, rod position range, axial offset range, core protection limits and representative xenon distributions. Then the 3D FAC evaluation consists of systematically calculating the 3D power distribution and margin to the core and fuel limits for each mesh point of this multi-dimensional space. The PCI margin is obtained by the comparison of the 3D power distributions over the Condition II space and the 3D maximum allowed power, which is dependent on the fuel rod history. The fuel history model utilizes the power history developed in the 3-dimensional nuclear analysis code to define local powers for the specified fuel rods to be analyzed. It tracks the rod history and provides the maximum allowed power for the point. This model is appropriate for base load operation, extended reduced power operation, return to power operation and

  7. Streamlined analysis technique for the evaluation of pellet clad interaction in PWR reload cores

    Energy Technology Data Exchange (ETDEWEB)

    Beard, Ch.; Morita, T.; Brown, J. [Westinghouse Electric Company, LLC, Nuclear Fuel Div., Pittsburgh, PA (United States)

    2007-07-01

    For some applications, an analysis is required to explicitly demonstrate that fuel failure due to pellet-clad interaction (PCI) is prevented by the core limits and the protection system for both Condition I (normal operation) operation and for Condition II (events of moderate frequency) events. This analysis needs to address the entire range of normal operation allowed by the Technical Specifications and all Condition II transients. The obvious approach which has been utilized for many years is a simulation of normal operation power maneuvers followed by explicit Condition II transients as a function of key core parameters. This is a sampling approach and has concerns about the overall coverage of the potential space. An alternative approach is the 3D FAC power distribution analysis methodology that was based upon the Westinghouse Relaxed Axial Offset Control Strategy (RAOC) evaluation process. The 3D FAC methodology uses a parametric representation of variables affecting the power distributions, defining a grid mesh over a space of Condition I and Condition II parameters. The operation space is defined by a power range, temperature range, rod position range, axial offset range, core protection limits and representative xenon distributions. Then the 3D FAC evaluation consists of systematically calculating the 3D power distribution and margin to the core and fuel limits for each mesh point of this multi-dimensional space. The PCI margin is obtained by the comparison of the 3D power distributions over the Condition II space and the 3D maximum allowed power, which is dependent on the fuel rod history. The fuel history model utilizes the power history developed in the 3-dimensional nuclear analysis code to define local powers for the specified fuel rods to be analyzed. It tracks the rod history and provides the maximum allowed power for the point. This model is appropriate for base load operation, extended reduced power operation, return to power operation and

  8. MODELING THE FORMATION OF GIANT PLANET CORES. I. EVALUATING KEY PROCESSES

    International Nuclear Information System (INIS)

    Levison, Harold F.; Thommes, Edward; Duncan, Martin J.

    2010-01-01

    One of the most challenging problems we face in our understanding of planet formation is how Jupiter and Saturn could have formed before the solar nebula dispersed. The most popular model of giant planet formation is the so-called core accretion model. In this model a large planetary embryo formed first, mainly by two-body accretion. This is then followed by a period of inflow of nebular gas directly onto the growing planet. The core accretion model has an Achilles heel, namely the very first step. We have undertaken the most comprehensive study of this process to date. In this study, we numerically integrate the orbits of a number of planetary embryos embedded in a swarm of planetesimals. In these experiments, we have included a large number of physical processes that might enhance accretion. In particular, we have included (1) aerodynamic gas drag, (2) collisional damping between planetesimals, (3) enhanced embryo cross sections due to their atmospheres, (4) planetesimal fragmentation, and (5) planetesimal-driven migration. We find that the gravitational interaction between the embryos and the planetesimals leads to the wholesale redistribution of material-regions are cleared of material and gaps open near the embryos. Indeed, in 90% of our simulations without fragmentation, the region near those embryos is cleared of planetesimals before much growth can occur. Thus, the widely used assumption that the surface density distribution of planetesimals is smooth can lead to misleading results. In the remaining 10% of our simulations, the embryos undergo a burst of outward migration that significantly increases growth. On timescales of ∼10 5 years, the outer embryo can migrate ∼6 AU and grow to roughly 30 M + . This represents a largely unexplored mode of core formation. We also find that the inclusion of planetesimal fragmentation tends to inhibit growth except for a narrow range of fragment migration rates.

  9. Survival of extensively damaged endodontically treated incisors restored with different types of posts-and-core foundation restoration material.

    Science.gov (United States)

    Lazari, Priscilla Cardoso; de Carvalho, Marco Aurélio; Del Bel Cury, Altair A; Magne, Pascal

    2018-05-01

    analysis (log-rank post hoc test at α=.05 for pairwise comparisons). None of the tested specimen withstood all 140 000 cycles. All specimens without a ferrule were affected by an initial failure phenomenon (wide gap at the lingual margin between the core foundation restoration/crown assembly and the root). NfPfP, NfPt, and NfPtB had similar survival (29649 to 30987 mean cycles until initial failure). NfPfB outperformed NfPt and NfPtB. None of the post-and-core foundation restoration materials were able to match the performance of the ferrule group FPf (72667 cycles). In all groups, 100% of failures were catastrophic. The survival of extensively damaged endodontically treated incisors without a ferrule was slightly improved by the use of a fiber post with a bulk-fill composite resin core foundation restoration. However, none of the post-and-core techniques was able to compensate for the absence of a ferrule. The presence of the posts always adversely affected the failure mode. Copyright © 2017 Editorial Council for the Journal of Prosthetic Dentistry. Published by Elsevier Inc. All rights reserved.

  10. Evaluation of excess nuclear materials suitability for international safeguards

    International Nuclear Information System (INIS)

    Newton, J.W.; White, W.C.; Davis, R.M.; Cherry, R.C.

    1996-01-01

    President Clinton announced in March 1995 the permanent withdrawal of 200 tons of fissile material from the US nuclear stockpile. This action was made possible by the dramatic reduction in nuclear weapons stockpile size and a desire to demonstrate the US'' commitment to nonproliferation goals. To provide further assurance of that commitment, the US is addressing placement of these materials under International Atomic Energy Agency (IAEA) safeguards. An initial step of this overall assessment was evaluation of the nuclear materials'' suitability for international safeguards. US Department of Energy (DOE) field organizations reviewed a detailed listing of all candidate materials with respect to characterization status, security classification, and acceptability for international safeguards compared to specified criteria. These criteria included form, location, environment and safety considerations, measurability, and stability. The evaluation resulted in broad categorizations of all materials with respect to preparing and placing materials under IAEA safeguards and provided essential information for decisions on the timing for offering materials as a function of materials attributes. A plan is being prepared to determine the availability of these materials for IAEA safeguards considering important factors such as costs, processes and facilities required to prepare materials, and impacts on other programs

  11. Impact of wall materials and seeding gases on the pedestal and on core plasma performance

    Directory of Open Access Journals (Sweden)

    E. Wolfrum

    2017-08-01

    Full Text Available Plasmas in machines with all metal plasma facing components have a lower Zeff, less radiation cooling in the scrape-off layer and divertor regions and are prone to impurity accumulation in the core. Higher gas puff and the seeding of low-Z impurities are applied to prevent impurity accumulation, to increase the frequency of edge localised modes and to cool the divertor. A lower power threshold for the transition from low-confinement mode to high confinement mode has been found in all metal wall machines when compared to carbon wall machines. The application of lithium before or during discharges can lead to ELM free H-modes. The seeding of high-Z impurities increases core radiation, reduces the power flux across the separatrix and, if applied in the right amount, does not lead to deterioration of the confinement. All these effects have in common that they can often be explained by the shape or position of the density profile. Not only the peakedness of the density profile in the core but also the position of the edge pressure gradient influences global confinement. It is shown how (i ionisation in the pedestal region due to higher reflection of deuterium from high-Z walls, (ii reduced recycling in consequence of lithium wall conditioning, (iii the fostering of edge modes with lithium dropping, (iv increased gas puff and (v the cooling of the scrape-off layer by medium-Z impurities such as nitrogen affect the edge density profile. The consequence is a shift in the pressure profile relative to the separatrix, leading to improved pedestal stability of H-mode plasmas when the direction is inwards.

  12. Use of Thermoanalytic Methods in the Evaluation of Combusted Materials

    Directory of Open Access Journals (Sweden)

    František Krepelka

    2006-12-01

    Full Text Available The paper describes possibilities of using thermoanalytic methods for the evaluation and comparison of materials designed for a direct combustion. Differential thermal analysis (DTA and thermogravimetric analysis (TGA were both used in the evaluation. The paper includes a description of methods of data processing from analyses for the purposes of comparison of used materials regarding their heating values. The following materials were analysed in the experiments: wooden coal of objectional grain size, fly ash from heating plant exhaust funnels, dendromass waste: spruce sawdust, micro-briquettes of spruce sawdust and fly-ash combined.

  13. Evaluation of Composite Materials for Use on Launch Complexes

    Science.gov (United States)

    Finchum, A.; Welch, Peter J.

    1989-01-01

    Commercially available composite structural shapes were evaluated for use. These composites, fiberglass-reinforced polyester and vinylester resin materials are being used extensively in the fabrication and construction of low maintenance, corrosion resistant structures. The evaluation found that in many applications these composite materials can be successfully used at the space center. These composite materials should not be used where they will be exposed to the hot exhaust plume/cloud of the launch vehicle during the liftoff, and caution should be taken in their use in areas where electrostatic discharge and hypergolic propellant compatibility are primary concerns.

  14. Evaluation of corrosion characteristics of SMART materials (III)

    International Nuclear Information System (INIS)

    Jeong, Y. H.; Park, S. Y.; Baek, J. H.; Choi, B. K.; Park, J. Y.; Lee, M. H.; Kim, J. H.; Bang, J. G.

    2006-02-01

    The corrosion characteristics of materials (Low-Sn Zircaloy-4, Zr-1.0Nb, PT-7M, ASTM Gr. 2 Ti, Inconel-690 alloys) for cladding and heat-exchanger tubes of SMART were evaluated in ammonia aqueous solution contained recirculating loop of pH 9.98 at 360 .deg. C 300 .deg. C. And CEDM materials (ball bearing, ball screw, magnetic material) were evaluated in ammonia aqueous solution contained static autoclave of pH 9.98 at 120 .deg. C

  15. [Self-evaluation of core competencies and related factors among baccalaureate nursing students].

    Science.gov (United States)

    Wu, Chen-Ting; Hsieh, Suh-Ing; Hsu, Li-Ling

    2013-02-01

    Evaluations of higher education programs are increasingly centered on the learner and designed to assess learning effectiveness and core competencies. Although the Taiwan Nursing Accreditation Council (TNAC) has established eight core competencies for college nursing departments, little research has been done to identify the most salient contributors to undergraduate nursing students' perceived competency levels. This paper investigates the influence of student demographic factors and learning experience on students' development in terms of a selected sample of core nursing competencies and then identifies factors that significantly predicts such development. This is a cross-sectional descriptive correlational study. We collected data from a sample of freshmen students currently enrolled in a two-year nursing bachelor degree program at a private vocational university in Taipei, Taiwan. Participants self-assessed abilities in designated core nursing competencies using the Competency Inventory of Nursing Students (CINS). A total of 279 of 290 distributed questionnaires were returned and used in data collection, giving this study a valid return rate of 96.2%. Participants earned a mean CINS score of 5.23 (SD = 0.49). Scale dimensions from highest to lowest mean score rank were: ethics, accountability, caring spirit, communication and cooperation, lifelong learning, general clinical nursing skills, critical thinking, and basic biomedical science. Differentiated analysis revealed that nursing students who expressed a strong interest in nursing, had a clear career plan, held aspirations to pursue higher nursing education, designated "major hospital" as their first workplace of choice, designated a post-college department / workplace preference, had participated in campus activities, were outspoken in classroom discussions and debates, made consistent effort to complete homework assignments and prepare for examinations, and performed relatively strong academically earned

  16. Evaluation of a thermal SCWR core with sub-channel analysis

    International Nuclear Information System (INIS)

    Liu Xiaojing; Cheng Xu

    2008-01-01

    A previous study shows that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behaviour than the existing one-row fuel assemblies. With this new developed two-row fuel assembly, a thermal SCWR core design is proposed Assessment of this design is carried out in this paper. The performance of this new core design is investigated with 3-D coupled thermal-hydraulic/neutronic calculations. During the coupling procedure, the thermal-hydraulic behaviour is analyzed using a single-channel code and the neutron-physical performance is computed with a 3-D reactor physical code. This paper presents the main results achieved so far related to the distribution of some neutronic and thermal-hydraulic parameters. Since the power distribution in some fuel assemblies is extremely uneven, sub-channel analysis is applied to the hottest and most non-uniform assembly in the core. The sub-channel analysis is performed with the power and thermal hydraulic parameters from the coupling results. It provides the hot channel factor and the maximal cladding surface temperature more precisely. The power and mass flux distribution in these assemblies are illustrated in detail for the demonstration purpose. The difference of the results evaluated with two different methods, i.e. sub-channel analysis and single-channel analysis, shows the importance of applying sub-channel analysis. A sensitivity analysis of some important parameters is also carried out. (author)

  17. Estimation of fracture parameters in foam core materials using thermal techniques

    DEFF Research Database (Denmark)

    Dulieu-Barton, J. M.; Berggreen, Christian; Boyenval Langlois, C.

    2010-01-01

    is described. A mode I simulated crack in the form of a machined notch is used to establish the feasibility of the TSA approach to derive stress intensity factors for the foam material. The overall goal is to demonstrate that thermal techniques have the ability to provide deeper insight into the behaviour......The paper presents some initial work on establishing the stress state at a crack tip in PVC foam material using a non-contact infra-red technique known as thermoelastic stress analysis (TSA). A parametric study of the factors that may affect the thermoelastic response of the foam material...

  18. Structural Integrity Evaluation of the KALIMER-600 Reactor Core Support Structure

    International Nuclear Information System (INIS)

    Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

    2005-01-01

    KALIMER-600(Korea Advanced LIquid MEtal Reactor, 600MWe) is a pool type sodium-cooled liquid metal reactor. Since the normal operating temperature of KALIMER-600 is 545 .deg. C, the reactor structures in the hot pool region are designed and evaluated according to the elevated temperature design rules such as the ASME Boiler and Pressure Vessel Code Section III, Subsection NH. Since the core support structure of KALIMER-600 is in the cold pool region under 400 .deg. C, a high temperature inelastic behavior is not expected. Thus the stress and fatigue limits are the main concerns to assure the structural design integrity following the ASME Subsection NG. In this paper, the evaluations of the stress and fatigue damage for the core support structure of KALIMER-600 are carrried out in the case of a normal operation condition using the rules of ASME Subsection NG. To obtain the stress values, a heat transfer analysis and a stress analysis under a combined loading condition are performed. From the stress distribution results, the critical sections are selected and the stress and fatigue limits are evaluated for the selected regions

  19. Single-step generation of fluorophore-encapsulated gold nanoparticle core-shell materials

    International Nuclear Information System (INIS)

    Sardar, R; Shem, P M; Pecchia-Bekkum, C; Bjorge, N S; Shumaker-Parry, J S

    2010-01-01

    We report a simple route to produce fluorophore-encapsulated gold nanoparticles (AuNPs) in a single step under aqueous conditions using the fluorophore 1-pyrenemethylamine (PMA). Different amounts of PMA were used and the resulting core-shell gold nanoparticles were analyzed using UV-visible absorption spectroscopy, fluorescence spectroscopy, and transmission and scanning electron microscopy. Electron microscopy analysis shows nanoparticles consisting of a gold nanoparticle core which is encapsulated with a lower contrast shell. In the UV-visible spectra, we observed a significant red shift (37 nm) of the localized surface plasmon resonance (LSPR) absorption maximum (λ max ) compared to citrate-stabilized AuNPs of a similar size. We attribute the prominent LSPR wavelength shift for PMA-AuNP conjugates to the increase in the local dielectric environment near the gold nanoparticles due to the shell formation. This simple, aqueous-based synthesis is a new approach to the production of fluorophore-encapsulated AuNPs that could be applicable in biological sensing systems and photonic device fabrication.

  20. In core instrumentation for online nuclear heating measurements of material testing reactor

    International Nuclear Information System (INIS)

    Reynard, C.; Andre, J.; Brun, J.; Carette, M.; Janulyte, A.; Merroun, O.; Zerega, Y.; Lyoussi, A.; Bignan, G.; Chauvin, J-P.; Fourmentel, D.; Glayse, W.; Gonnier, C.; Guimbal, P.; Iracane, D.; Villard, J.-F.

    2010-01-01

    The present work focuses on nuclear heating. This work belongs to a new advanced research program called IN-CORE which means 'Instrumentation for Nuclear radiations and Calorimetry Online in REactor' between the LCP (University of Provence-CNRS) and the CEA (French Atomic Energy Commission) - Jules Horowitz Reactor (JHR) program. This program started in September 2009 and is dedicated to the conception and the design of an innovative mobile experimental device coupling several sensors and ray detectors for on line measurements of relevant physical parameters (photonic heating, neutronic flux ...) and for an accurate parametric mapping of experimental channels in the JHR Core. The work presented below is the first step of this program and concerns a brief state of the art related to measurement methods of nuclear heating phenomena in research reactor in general and MTR in particular. A special care is given to gamma heating measurements. A first part deals with numerical codes and models. The second one presents instrumentation divided into various kinds of sensor such as calorimeter measurements and gamma ionization chamber measurements. Their basic principles, characteristics such as metrological parameters, operating mode, disadvantages/advantages, ... are discussed. (author)

  1. Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

    International Nuclear Information System (INIS)

    Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

    1985-01-01

    The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO 2 and UO 2 /metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO 2 crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process

  2. Economic evaluation of closure cap barrier materials study

    Energy Technology Data Exchange (ETDEWEB)

    Serrato, M.G.; Bhutani, J.S.; Mead, S.M.

    1993-09-01

    Volume II of the Economic Evaluation of the Closure Cap Barrier Materials, Revision I contains detailed cost estimates for closure cap barrier materials. The cost estimates incorporate the life cycle costs for a generic hazardous waste seepage basin closure cap under the RCRA Post Closure Period of thirty years. The economic evaluation assessed six barrier material categories. Each of these categories consists of several composite cover system configurations, which were used to develop individual cost estimates. The information contained in this report is not intended to be used as a cost estimating manual. This information provides the decision makers with the ability to screen barrier materials, cover system configurations, and identify cost-effective materials for further consideration.

  3. Evaluation on elution feature of bentonite buffer materials

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Hirohisa; Kanno, Takeshi; Matsumoto, Kazuhiro

    1997-09-01

    In order to evaluate long term physical stability of artificial barrier in land disposal of high level radioactive wastes, it is necessary to know quantitatively elution behavior of buffering materials from disposal road (or cavity) to circumferential rock crack. When elution of the buffer material occurs on large scale, amount of bentonite in the disposal road (or cavity) reduces and reduction of various functions expected to the buffer materials is presumed. According to specification examples of road transverse arrangement and disposal vertical arrangement systems, evaluation on elution amount of the buffer materials at disposal environment was conducted. Opening width of rock crack in the disposal environment was supposed to be 0.5 mm. As a result, obtained mass elution ratios of the buffer materials due to extrusion phenomenon were 0.04 to 0.2% after 10,000 year and 2 to 12% after 1,000,000 years. (G.K.)

  4. Economic evaluation of closure cap barrier materials study

    International Nuclear Information System (INIS)

    Serrato, M.G.; Bhutani, J.S.; Mead, S.M.

    1993-09-01

    Volume II of the Economic Evaluation of the Closure Cap Barrier Materials, Revision I contains detailed cost estimates for closure cap barrier materials. The cost estimates incorporate the life cycle costs for a generic hazardous waste seepage basin closure cap under the RCRA Post Closure Period of thirty years. The economic evaluation assessed six barrier material categories. Each of these categories consists of several composite cover system configurations, which were used to develop individual cost estimates. The information contained in this report is not intended to be used as a cost estimating manual. This information provides the decision makers with the ability to screen barrier materials, cover system configurations, and identify cost-effective materials for further consideration

  5. Uncertainties assessment for safety margins evaluation in MTR reactors core thermal-hydraulic design

    International Nuclear Information System (INIS)

    Gimenez, M.; Schlamp, M.; Vertullo, A.

    2002-01-01

    This report contains a bibliographic review and a critical analysis of different methodologies used for uncertainty evaluation in research reactors core safety related parameters. Different parameters where uncertainties are considered are also presented and discussed, as well as their intrinsic nature regarding the way their uncertainty combination must be done. Finally a combined statistical method with direct propagation of uncertainties and a set of basic parameters as wall and DNB temperatures, CHF, PRD and their respective ratios where uncertainties should be considered is proposed. (author)

  6. Spirobifluorene Core-Based Novel Hole Transporting Materials for Red Phosphorescence OLEDs

    Directory of Open Access Journals (Sweden)

    Ramanaskanda Braveenth

    2017-03-01

    Full Text Available Two new hole transporting materials, named HTM 1A and HTM 1B, were designed and synthesized in significant yields using the well-known Buchwald Hartwig and Suzuki cross- coupling reactions. Both materials showed higher decomposition temperatures (over 450 °C at 5% weight reduction and HTM 1B exhibited a higher glass transition temperature of 180 °C. Red phosphorescence-based OLED devices were fabricated to analyze the device performances compared to Spiro-NPB and NPB as reference hole transporting materials. Devices consist of hole transporting material as HTM 1B showed better maximum current and power efficiencies of 16.16 cd/A and 11.17 lm/W, at the same time it revealed an improved external quantum efficiency of 13.64%. This efficiency is considerably higher than that of Spiro-NPB and NPB-based reference devices.

  7. Evaluation report on CCTF core-II reflood test C2 - 18 (Run 78)

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Murao, Yoshio; Sugimoto, Jun; Hojo, Tsuneyuki.

    1987-03-01

    This report presents the result of the upper plenum injection (UPI) test C2 - 18 (Run 78), which was conducted on November 13, 1984 with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute (JAERI). The CCTF is a 1/21.4 scale model of a 1,100 MWe PWR with four loop active components to provide information on the system and core thermo-hydrodynamics during reflood phase. The objectives of the test are to investigate the refill behavior with UPI condition and to investigate the reflood behavior with UPI Best-Estimate (BE) condition. The test was performed to simulate refill/reflood behavior with UPI and BE conditions (However, the LPCI flow rate was determined based on single failure of LPCI pumps.). The result of the test showed the followings. (1) Little special phenomena were recognized under UPI and BE conditions in comparison with those under UPI and Evaluation-Model (EM) conditions, although certain special phenoma (i.e. significant fluid oscillation) were recognized under Cold-Leg-Injection (CLI) and BE conditions in comparison with those under CLI and EM conditions. (2) Water inventory in lower plenum increased smoothly due to water injected into both upper plenum and cold leg during refill phase, similarly to that in refill-simulation test with CLI condition. Small difference in refill behavior with UPI condition is the existing of steam condensation in upper plenum, resulting in lower steam binding and higher core cooling during early reflood phase. This indicates the conservatism of UPI against CLI during early reflood phase. (3) The good core-cooling capability was confirmed under UPI and BE conditions. (author)

  8. Confocal fluorescence microscopy for rapid evaluation of invasive tumor cellularity of inflammatory breast carcinoma core needle biopsies.

    Science.gov (United States)

    Dobbs, Jessica; Krishnamurthy, Savitri; Kyrish, Matthew; Benveniste, Ana Paula; Yang, Wei; Richards-Kortum, Rebecca

    2015-01-01

    Tissue sampling is a problematic issue for inflammatory breast carcinoma, and immediate evaluation following core needle biopsy is needed to evaluate specimen adequacy. We sought to determine if confocal fluorescence microscopy provides sufficient resolution to evaluate specimen adequacy by comparing invasive tumor cellularity estimated from standard histologic images to invasive tumor cellularity estimated from confocal images of breast core needle biopsy specimens. Grayscale confocal fluorescence images of breast core needle biopsy specimens were acquired following proflavine application. A breast-dedicated pathologist evaluated invasive tumor cellularity in histologic images with hematoxylin and eosin staining and in grayscale and false-colored confocal images of cores. Agreement between cellularity estimates was quantified using a kappa coefficient. 23 cores from 23 patients with suspected inflammatory breast carcinoma were imaged. Confocal images were acquired in an average of less than 2 min per core. Invasive tumor cellularity estimated from histologic and grayscale confocal images showed moderate agreement by kappa coefficient: κ = 0.48 ± 0.09 (p confocal images require less than 2 min for acquisition and allow for evaluation of invasive tumor cellularity in breast core needle biopsy specimens with moderate agreement to histologic images. We show that confocal fluorescence microscopy can be performed immediately following specimen acquisition and could indicate the need for additional biopsies at the initial visit.

  9. First in-core measurement results obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS material testing reactor

    International Nuclear Information System (INIS)

    Carcreff, Hubert; Salmon, Laurent; Courtaux, Cedric

    2014-01-01

    Nuclear heating rate inside an MTR has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. An innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. Development of the calorimetric probe required manufacturing and irradiation of mock-ups in the ex-core area, where nuclear heating rate does not exceed 2 W.g -1 . The calorimeter working mode, the different measurement procedures, main modeling and ex-core experimental results have been already presented in previous papers. In this paper, we present in-core results obtained from 2011 to 2013 with the final device. For the first time, this new experimental measurement system was operated in several experimental locations, with nominal in-core thermal hydraulic conditions, nominal neutron flux and nuclear heating rate up to 6 W.g -1 (in graphite). After a brief presentation of the displacement system specificities, first nuclear heating distributions are presented and discussed. The Finite Element model of the calorimeter was upgraded in order to match calculated temperatures with measured ones. This 'validated' model allowed to estimate a Kc factor which tends to correct small nonlinearities when heating rate is calculated from the 'calibration method'. A comparison is made between nuclear heating rates determined from 'calibration' and 'zero methods'. In addition, an evaluation of the global uncertainty associated to the measurements is detailed. Finally, a comparison is made with available measurements obtained from previous calorimeters. (authors)

  10. Evaluation and development of advanced nuclear materials: IAEA activities

    International Nuclear Information System (INIS)

    Inozemtsev, V.; Basak, U.; Killeen, J.; Dyck, G.; Zeman, A.; )

    2011-01-01

    Economical, environmental and non-proliferation issues associated with sustainable development of nuclear power bring about a need for optimization of fuel cycles and implementation of advanced nuclear systems. While a number of physical and design concepts are available for innovative reactors, the absence of reliable materials able to sustain new challenging irradiation conditions represents the real bottle-neck for practical implementation of these promising ideas. Materials performance and integrity are key issues for the safety and competitiveness of future nuclear installations being developed for sustainable nuclear energy production incorporating fuel recycling and waste transmutation systems. These systems will feature high thermal operational efficiency, improved utilization of resources (both fissile and fertile materials) and reduced production of nuclear waste. They will require development, qualification and deployment of new and advanced fuel and structural materials with improved mechanical and chemical properties combined with high radiation and corrosion resistance. The extensive, diverse, and expensive efforts toward the development of these materials can be more effectively organized within international collaborative programmes with wide participation of research, design and engineering communities. IAEA carries out a number of international projects supporting interested Member States with the use of available IAEA program implementation tools (Coordinated Research Projects, Technical Meetings, Expert Reviews, etc). The presentation summarizes the activities targeting material developments for advanced nuclear systems, with particular emphasis on fast reactors, which are the focal topics of IAEA Coordinated Research Projects 'Accelerator Simulation and Theoretical Modelling of Radiation Effects' (on-going), 'Benchmarking of Structural Materials Pre-Selected for Advanced Nuclear Reactors', 'Examination of advanced fast reactor fuel and core

  11. Evaluation of material dispersion using a nanosecond optical pulse radiator.

    Science.gov (United States)

    Horiguchi, M; Ohmori, Y; Miya, T

    1979-07-01

    To study the material dispersion effects on graded-index fibers, a method for measuring the material dispersion in optical glass fibers has been developed. Nanosecond pulses in the 0.5-1.7-microm region are generated by a nanosecond optical pulse radiator and grating monochromator. These pulses are injected into a GeO(2)-P(2)0(5)-doped silica graded-index fiber. Relative time delay changes between different wavelengths are used to determine material dispersion, core glass refractive index, material group index, and optimum profile parameter of the graded-index fiber. From the measured data, the optimum profile parameter on the GeO(2)-P(2)O(5)-doped silica graded-index fiber could be estimated to be 1.88 at 1.27 microm of the material dispersion free wavelength region and 1.82 at 1.55 microm of the lowest-loss wavelength region in silica-based optical fiber waveguides.

  12. IMPORTANCE OF MATERIAL BALANCES AND THEIR STATISTICAL EVALUATION IN RUSSIAN MATERIAL, PROTECTION, CONTROL AND ACCOUNTING

    International Nuclear Information System (INIS)

    Fishbone, L.G.

    1999-01-01

    While substantial work has been performed in the Russian MPC and A Program, much more needs to be done at Russian nuclear facilities to complete four necessary steps. These are (1) periodically measuring the physical inventory of nuclear material, (2) continuously measuring the flows of nuclear material, (3) using the results to close the material balance, particularly at bulk processing facilities, and (4) statistically evaluating any apparent loss of nuclear material. The periodic closing of material balances provides an objective test of the facility's system of nuclear material protection, control and accounting. The statistical evaluation using the uncertainties associated with individual measurement systems involved in the calculation of the material balance provides a fair standard for concluding whether the apparent loss of nuclear material means a diversion or whether the facility's accounting system needs improvement. In particular, if unattractive flow material at a facility is not measured well, the accounting system cannot readily detect the loss of attractive material if the latter substantially derives from the former

  13. Effect of varying core thicknesses and artificial aging on the color difference of different all-ceramic materials.

    Science.gov (United States)

    Dikicier, Sibel; Ayyildiz, Simel; Ozen, Julide; Sipahi, Cumhur

    2014-11-01

    Clinicians should reserve all-ceramics with high translucency for clinical applications in which high-level esthetics are required. Furthermore, it is unclear whether a correlation exists between core thickness and color change. The aim of this study was to examine the effects of different core thicknesses and artificial aging on the color stability of three all-ceramic systems. Ninety disc-shaped cores with different thicknesses (0.5 mm, 0.8 mm and 1.0 mm) were prepared from three all-ceramic systems, In-Ceram Alumina (IC), IPS e.max Press (EM) and Katana (K). The colors of the samples were measured with a spectrophotometer and the color parameters (L*, a*, b*, ΔE) were calculated according to the CIE L*a*b* (Commission Internationale de L'Eclairage) color system before and after aging. The effects of aging on color parameters were statistically significant (p artificial aging affected color stability of the all-ceramic materials tested.

  14. From harmful Microcystis blooms to multi-functional core-double-shell microsphere bio-hydrochar materials.

    Science.gov (United States)

    Bi, Lei; Pan, Gang

    2017-11-13

    Harmful algal blooms (HABs) induced by eutrophication is becoming a serious global environmental problem affecting public health and aquatic ecological sustainability. A novel strategy for the utilization of biomass from HABs was developed by converting the algae cells into hollow mesoporous bio-hydrochar microspheres via hydrothermal carbonization method. The hollow microspheres were used as microreactors and carriers for constructing CaO 2 core-mesoporous shell-CaO 2 shell microspheres (OCRMs). The CaO 2 shells could quickly increase dissolved oxygen to extremely anaerobic water in the initial 40 min until the CaO 2 shells were consumed. The mesoporous shells continued to act as regulators restricting the release of oxygen from CaO 2 cores. The oxygen-release time using OCRMs was 7 times longer than when directly using CaO 2 . More interestingly, OCRMs presented a high phosphate removal efficiency (95.6%) and prevented the pH of the solution from rising to high levels in comparison with directly adding CaO 2 due to the OH - controlled-release effect of OCRMs. The distinct core-double-shell micro/nanostructure endowed the OCRMs with triple functions for oxygen controlled-release, phosphorus removal and less impact on water pH. The study is to explore the possibility to prepare smarter bio-hydrochar materials by utilizing algal blooms.

  15. A core outcome set for studies evaluating the effectiveness of prepregnancy care for women with pregestational diabetes.

    LENUS (Irish Health Repository)

    Egan, Aoife M

    2017-04-01

    The aim of this study was to develop a core outcome set (COS) for trials and other studies evaluating the effectiveness of prepregnancy care for women with pregestational (pre-existing) diabetes mellitus.

  16. Thermal-hydraulic evaluation study of the effectiveness of emergency core cooling system for light water reactors

    International Nuclear Information System (INIS)

    Sobajima, Makoto

    1985-08-01

    In order to evaluate the core cooling capability of the emergeny core cooling system, which is a safety guard system of light water reactors for a loss-of-coolant accident, a variety of large scale test were performed. Through the results, many phenomena were investigated and the predictabity of analytical codes were examined. The tests conducted were a single-vessel blowdown test, emergency core cooling test in a PWR simulation facility, spray cooling test for a BWR, large scale reflood test and a separate effect test on countercurrent flow. These test results were examined to clarify thermal-hydraulic phenomena and the effect of various test parameters and were utilized to improve predictability of the analytical codes. Some models for flow behavior in the upper core were also developed. By evaluating the effectiveness of various emergency core cooling system configurations, more effective cooling system than the current one was proposed and demonstrated. (author)

  17. Characterization of a Porous Carbon Material Functionalized with Cobalt-Oxide/Cobalt Core-Shell Nanoparticles for Lithium Ion Battery Electrodes

    KAUST Repository

    Anjum, Dalaver H.; Rasul, Shahid; Roldan-Gutierrez, Manuel A.; Da Costa, Pedro M. F. J.

    2016-01-01

    A nanoporous carbon (C) material, functionalized with Cobalt-Oxide/Cobalt (CoO/Co) core-shell nanoparticles (NPs), was structurally and chemically characterized with transmission electron microcopy (TEM) while its electrochemical response

  18. Chemical interactions of reactor core materials up to very high temperatures

    International Nuclear Information System (INIS)

    Hofmann, P.; Hagen, S.; Schanz, G.; Skokan, A.

    1989-01-01

    The paper describes which chemical interactions may occur in a LWR fuel rod bundle containing (Ag, In, Cd) absorber rods or (Al 2 O 3 /B 4 C) burnable poison rods with increasing temperature up to the complete melting of the components and the formed reaction products. The kinetics of the most important chemical interactions has been investigated and the results are described. In most cases the reaction products have lower melting points or ranges than the original components. This results in a relocation of liquefied components often far below their melting points. There exist three distinct temperature regimes in which liquid phases can form in the core in differently large quantities. These temperature regimes are described in detail. The phase relations in the important ternary (U, Zr, O) system have been extensively studied. The effect of steel constituents on the phase relations is given in addition. All the considerations are focused on PWR conditions only. (orig.) [de

  19. The mediating role of nurses' professional commitment in the relationship between core self-evaluation and job satisfaction.

    Science.gov (United States)

    Barać, Ivana; Prlić, Nada; Plužarić, Jadranka; Farčić, Nikolina; Kovačević, Suzana

    2018-05-11

    The aim of this study was to examine the degree to which it is possible to predict job satisfaction in hospital nurses based on core self-evaluation and the nurses' professional commitment. Psychological constructs of nurses' professional commitment could predict a level of job satisfaction. A cross-sectional design was applied. Data were collected between April 2016 and November 2016 from 584 nurses of the University Hospital Osijek. Core Self-Evaluation Scale (CSES), Job Satisfaction Survey (JSS) and Nurses' professional commitment scale (NPCS) were administrated to the study participants. Confirmatory factor analyses were conducted to test the validity of each questionnaire. Structural equation modeling was used to test the prediction of nurses' professional commitment and core self-evaluation on job satisfaction. Nurses' professional commitment is variable, which functions as a mediator between predictor (CSE) and criterion variable (JS). As a mediator, it explains what the effect is, provided that correlations between all variables are significant. The correlation analyses reveal significant positive correlations between job satisfaction and core self evaluation (r = 0.441, p > 0.001) and also between job satisfaction and nurses' professional commitment (r = 0.464, p > 0.001). Furthermore, core self evaluation significantly and positively correlates with nurses' professional commitment (r = 0.402, p > 0.001). The results showed that nurses' professional commitment mediates the relationship between core self evaluation and job satisfaction. Bootstrap analysis showed that core self evaluation partially mediated the relationship between nurses' professional commitment and job satisfaction ( β = 0.78, p core self evaluation on job satisfaction through nurses' professional commitment was also significant (β = 0.17, p < 0.001**). Nurses who are more committed to their work, regardless of the structure of personality, have greater satisfaction in their work. This

  20. Methodology for Evaluating Raw Material Changes to RSRM Elastomeric Insulation Materials

    Science.gov (United States)

    Mildenhall, Scott D.; McCool, Alex (Technical Monitor)

    2001-01-01

    The Reusable Solid Rocket Motor (RSRM) uses asbestos and silicon dioxide filled acrylonitrile butadiene rubber (AS-NBR) as the primary internal insulation to protect the case from heat. During the course of the RSRM Program, several changes have been made to the raw materials and processing of the AS-NBR elastomeric insulation material. These changes have been primarily caused by raw materials becoming obsolete. In addition, some process changes have been implemented that were deemed necessary to improve the quality and consistency of the AS-NBR insulation material. Each change has been evaluated using unique test efforts customized to determine the potential impacts of the specific raw material or process change. Following the evaluations, the various raw material and process changes were successfully implemented with no detectable effect on the performance of the AS-NBR insulation. This paper will discuss some of the raw material and process changes evaluated, the methodology used in designing the unique test plans, and the general evaluation results. A summary of the change history of RSRM AS-NBR internal insulation is also presented.

  1. Structural evaluation of fast reactor core restraint with irradiation creep-swelling opposition effects

    International Nuclear Information System (INIS)

    Kalinowski, J.E.

    1979-01-01

    Irradiation creep and swelling correlations are derived from primary loading in-reactor experiments in which irradiation creep and swelling act in the same direction. When correlation uncertainty bands are applied in core restraint evaluations, significant variability in sub-assembly behavior is predicted. For example, sub-assemblies in the outer core region where neutron flux and duct temperature gradients are significant exhibit bowing responses ranging from a creep dominated outward bow to a swelling dominated inward bow. Furthermore, solutions based on upper bound and lower bound correlation uncertainty combinations are observed to cross-over indicating that such combinations are physically unrealistic in the assessment of creep-swelling opposition effects. In order to obtain realistic upper and lower bound sub-assembly responses, judgement must be applied in the selection of creep-swelling equation uncertainty combinations. Experimental programs have been defined which will provide the needed basic as well as prototypic creep-swelling opposition data for reference and advanced sub-assembly duct alloys. The first of these is an irradiation of cylindrical capsules subjected to a through-wall temperature gradient. This test which is presently underway in the EBR-II reactor will provide the data needed to refine irradiation creep and swelling correlations and their associated uncertainties when applied to core restraint evaluations. Restrained pin and duct bowing experiments in FFTF have also been defined. These will provide the prototypic data necessary to verify irradiated duct bowing methodology. The results of this experimental program are expected to reduce creep and swelling uncertainties and permit better definition of the design window for load plane gaps. (orig.)

  2. Design and preliminary analysis of in-vessel core catcher made of high-temperature ceramics material in PWR

    International Nuclear Information System (INIS)

    Xu Hong; Ma Li; Wang Junrong; Zhou Zhiwei

    2011-01-01

    In order to protect the interior wall of pressure vessel from melting, as an additional way to external reactor vessel cooling (ERVC), a kind of in-vessel core catcher (IVCC) made of high-temperature ceramics material was designed. Through the high-temperature and thermal-resistance characteristic of IVCC, the distributing of heat flux was optimized. The results show that the downward average heat flux from melt in ceramic layer reduces obviously and the interior wall of pressure vessel doesn't melt, keeping its integrity perfectly. Increasing of upward heat flux from metallic layer makes the upper plenum structure's temperature ascend, but the temperature doesn't exceed its melting point. In conclusion, the results indicate the potential feasibility of IVCC made of high-temperature ceramics material. (authors)

  3. Nano-magnetic particles used in biomedicine: core and coating materials.

    Science.gov (United States)

    Karimi, Z; Karimi, L; Shokrollahi, H

    2013-07-01

    Magnetic nanoparticles for medical applications have been developed by many researchers. Separation, immunoassay, drug delivery, magnetic resonance imaging and hyperthermia are enhanced by the use of suitable magnetic nanoparticles and coating materials in the form of ferrofluids. Due to their low biocompatibility and low dispersion in water solutions, nanoparticles that are used for biomedical applications require surface treatment. Various kinds of coating materials including organic materials (polymers), inorganic metals (gold, platinum) or metal oxides (aluminum oxide, cobalt oxide) have been attracted during the last few years. Based on the recent advances and the importance of nanomedicine in human life, this paper attempts to give a brief summary on the different ferrite nano-magnetic particles and coatings used in nanomedicine. Copyright © 2013 Elsevier B.V. All rights reserved.

  4. Core performance of equilibrium fast reactors for different coolant materials and fuel types

    International Nuclear Information System (INIS)

    Mizutani, Akihiko; Sekimoto, Hiroshi

    1998-01-01

    Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

  5. Transformer core

    NARCIS (Netherlands)

    Mehendale, A.; Hagedoorn, Wouter; Lötters, Joost Conrad

    2008-01-01

    A transformer core includes a stack of a plurality of planar core plates of a magnetically permeable material, which plates each consist of a first and a second sub-part that together enclose at least one opening. The sub-parts can be fitted together via contact faces that are located on either side

  6. Transformer core

    NARCIS (Netherlands)

    Mehendale, A.; Hagedoorn, Wouter; Lötters, Joost Conrad

    2010-01-01

    A transformer core includes a stack of a plurality of planar core plates of a magnetically permeable material, which plates each consist of a first and a second sub-part that together enclose at least one opening. The sub-parts can be fitted together via contact faces that are located on either side

  7. Evaluation of the need for stochastic optimization of out-of-core nuclear fuel management decisions

    International Nuclear Information System (INIS)

    Thomas, R.L. Jr.

    1989-01-01

    Work has been completed on utilizing mathematical optimization techniques to optimize out-of-core nuclear fuel management decisions. The objective of such optimization is to minimize the levelized fuel cycle cost over some planning horizon. Typical decision variables include feed enrichments and number of assemblies, burnable poison requirements, and burned fuel to reinsert for every cycle in the planning horizon. Engineering constraints imposed consist of such items as discharge burnup limits, maximum enrichment limit, and target cycle energy productions. Earlier the authors reported on the development of the OCEON code, which employs the integer Monte Carlo Programming method as the mathematical optimization method. The discharge burnpups, and feed enrichment and burnable poison requirements are evaluated, initially employing a linear reactivity core physics model and refined using a coarse mesh nodal model. The economic evaluation is completed using a modification of the CINCAS methodology. Interest now is to assess the need for stochastic optimization, which will account for cost components and cycle energy production uncertainties. The implication of the present studies is that stochastic optimization in regard to cost component uncertainties need not be completed since deterministic optimization will identify nearly the same family of near-optimum cycling schemes

  8. A benchmarking tool to evaluate computer tomography perfusion infarct core predictions against a DWI standard.

    Science.gov (United States)

    Cereda, Carlo W; Christensen, Søren; Campbell, Bruce Cv; Mishra, Nishant K; Mlynash, Michael; Levi, Christopher; Straka, Matus; Wintermark, Max; Bammer, Roland; Albers, Gregory W; Parsons, Mark W; Lansberg, Maarten G

    2016-10-01

    Differences in research methodology have hampered the optimization of Computer Tomography Perfusion (CTP) for identification of the ischemic core. We aim to optimize CTP core identification using a novel benchmarking tool. The benchmarking tool consists of an imaging library and a statistical analysis algorithm to evaluate the performance of CTP. The tool was used to optimize and evaluate an in-house developed CTP-software algorithm. Imaging data of 103 acute stroke patients were included in the benchmarking tool. Median time from stroke onset to CT was 185 min (IQR 180-238), and the median time between completion of CT and start of MRI was 36 min (IQR 25-79). Volumetric accuracy of the CTP-ROIs was optimal at an rCBF threshold of benchmarking tool can play an important role in optimizing CTP software as it provides investigators with a novel method to directly compare the performance of alternative CTP software packages. © The Author(s) 2015.

  9. Evaluation of global synchronization for iterative algebra algorithms on many-core

    KAUST Repository

    ul Hasan Khan, Ayaz; Al-Mouhamed, Mayez; Firdaus, Lutfi A.

    2015-01-01

    © 2015 IEEE. Massively parallel computing is applied extensively in various scientific and engineering domains. With the growing interest in many-core architectures and due to the lack of explicit support for inter-block synchronization specifically in GPUs, synchronization becomes necessary to minimize inter-block communication time. In this paper, we have proposed two new inter-block synchronization techniques: 1) Relaxed Synchronization, and 2) Block-Query Synchronization. These schemes are used in implementing numerical iterative solvers where computation/communication overlapping is one used optimization to enhance application performance. We have evaluated and analyzed the performance of the proposed synchronization techniques using Jacobi Iterative Solver in comparison to the state of the art inter-block lock-free synchronization techniques. We have achieved about 1-8% performance improvement in terms of execution time over lock-free synchronization depending on the problem size and the number of thread blocks. We have also evaluated the proposed algorithm on GPU and MIC architectures and obtained about 8-26% performance improvement over the barrier synchronization available in OpenMP programming environment depending on the problem size and number of cores used.

  10. Evaluation of advanced two-phase flow instrumentation in SCTF Core-1

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Sobajima, Makoto; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka; Sudo, Yukio; Adachi, Hiromichi

    1984-03-01

    In the Slab Core Test Facility (SCTF) Core-I, advanced two-phase flow instruments have been provided by the USNRC to measure the thermohydraulic behavior in the primary system including pressure vessel during the end of blowdown, refill and reflood phases of a postulated loss-of-coolant accident in a pressurized water reactor. The advanced instruments are turbine meters, drag disks, γ-densitometers, spool pieces, liquid level detectors (LLD), fluid distribution grids (FDG), impedance probes (flag, prong and string probes), film probes, and video optical probes. This report presents evaluated results of the data from these instruments. Some instruments are quantitatively evaluated by comparing with the data from the conventional instruments or the other advanced instruments. Main conclusions are as follows: (1) The spool pieces and the γ-densitometers work well and provide satisfactory results; (2) Some of the turbine meters, the impedance probes and the film probes give partially reasonable results, but still more improvements are required; (3) Most of the LLDs, the FDGs, the impedance probes, and the film probes do not work well due to a hard cable corrosion, and (4) The video optical probes give clear image of the flow pattern. (author)

  11. Evaluation of global synchronization for iterative algebra algorithms on many-core

    KAUST Repository

    ul Hasan Khan, Ayaz

    2015-06-01

    © 2015 IEEE. Massively parallel computing is applied extensively in various scientific and engineering domains. With the growing interest in many-core architectures and due to the lack of explicit support for inter-block synchronization specifically in GPUs, synchronization becomes necessary to minimize inter-block communication time. In this paper, we have proposed two new inter-block synchronization techniques: 1) Relaxed Synchronization, and 2) Block-Query Synchronization. These schemes are used in implementing numerical iterative solvers where computation/communication overlapping is one used optimization to enhance application performance. We have evaluated and analyzed the performance of the proposed synchronization techniques using Jacobi Iterative Solver in comparison to the state of the art inter-block lock-free synchronization techniques. We have achieved about 1-8% performance improvement in terms of execution time over lock-free synchronization depending on the problem size and the number of thread blocks. We have also evaluated the proposed algorithm on GPU and MIC architectures and obtained about 8-26% performance improvement over the barrier synchronization available in OpenMP programming environment depending on the problem size and number of cores used.

  12. BDI behavior evaluation of an upgraded Monju core and a demonstration core. (1) Plans for the out of pile bundle compressive tests for large diameter pins

    International Nuclear Information System (INIS)

    Ichikawa, Shoichi; Haga, Hiroyuki; Katsuyama, Kozo; Uwaba, Tomoyuki; Maeda, Koji; Nishinoiri, Kenji

    2012-07-01

    The life of FBR (Fast Breeder Reactor) fuel assembly is restricted by BDI (Bundle-Duct Interaction). Therefore, it is very important to carry out the out pile bundle compressive tests which can imitate BDI, in order to evaluate BDI behavior. The target of the conventional BDI behavior was small diameter pins (φ6.5mm) for fuel pellets which were used with the assembly of Monju (the Monju prototype fast breeder reactor) etc. Furthermore by an upgraded Monju core and a demonstration core, adoption of large diameter pins for the holler annular pellets is planned. Therefore, it was necessary to carry out BDI evaluation of a large diameter pin. Then, the plans for out of pile bundle compressive test for large diameter pins were are reported. (author)

  13. Aspects of unconventional cores for large sodium cooled power reactors; evaluation of a literature survey

    International Nuclear Information System (INIS)

    Kiefhaber, E.

    1978-10-01

    The report gives an overview of a literature study on the application of unconventional cores for sodium cooled fast reactors. Different types of unconventional cores (heterogeneous cores, pancake cores, moderated cores and others) are compared with conventional cores, which are characterized by a cylindrical geometry with two or three fissile zones surrounded by an axial and a radial blanket. The main parameters of interest in this comparison are the neutronic parameters sodium void and Doppler effect, the breeding properties and the steel damage. Consequences for the core safety and the overall plant design are also mentioned

  14. A risk-based evaluation of LMFBR containment response under core disruptive accident conditions

    International Nuclear Information System (INIS)

    Hartung, J.; Berk, S.

    1978-01-01

    Probabilistic risk methodology is utilized to evaluate the failure modes and effects of LMFBR containment systems under Core Disruptive Accident (CDA) conditions. First, the potential causes of LMFBR containment failure under CDA conditions are discussed and categorized. Then, a simple scoping-type risk assessment of a reference design is presented to help place these potential causes of failure in perspective. The highest risk containment failure modes are identified for the reference design, and several design and research and development options which appear capable of reducing these risks are discussed. The degree to which large LMFBR containment systems must mitigate the consequences of CDA's to achieve a level of risk (for LMFBR's) comparable to the already very low risk of contemporary LWR's is explored. Based on the results of this evaluation, several suggestions are offered concerning CDA-related design goals and research and development priorities for large LMFBR's. (author)

  15. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    In this conference are presented papers dealing with swelling of metals and alloys, (and specially ferritic steels), structural evolution and stability under irradiation, modifications of mechanical properties, consequences on the behaviour of fuel elements and the optimization of materials selection, and irradiation creep [fr

  16. Easily oxidizable triarylamine materials with naphthalene and binaphthalene core: structure-properties relationship

    Czech Academy of Sciences Publication Activity Database

    Kerner, L.; Gmucová, K.; Kožíšek, J.; Petříček, Václav; Putala, M.

    2016-01-01

    Roč. 72, č. 44 (2016), s. 7081-7092 ISSN 0040-4020 R&D Projects: GA ČR(CZ) GA14-03276S Institutional support: RVO:68378271 Keywords : triphenylamine * carbazole * regioselective amination * Sonogashira coupling * Negishi alkynylation * hole-transporting materials * OLED * cyclic voltammetry Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 2.651, year: 2016

  17. Evaluation method for change of concentration of nuclear fuel material

    International Nuclear Information System (INIS)

    Kiyono, Takeshi; Ando, Ryohei.

    1997-01-01

    The present invention provides a method of evaluating the change of concentration of compositions of nuclear fuel element materials loaded to a reactor along with neutron irradiation based on analytic calculation not relying on integration with time. Namely, the method of evaluating the change of concentration of nuclear fuel materials comprises evaluating the changing concentration of nuclear fuel materials based on nuclear fission, capturing of neutrons and radioactive decaying along with neutron irradiation. In this case, an optional nuclide on a nuclear conversion chain is determined as a standard nuclide. When the main fuel material is Pu-239, it is determined as the standard nuclide. The ratio of the concentration of the standard nuclide to that of the nuclide as an object of the evaluation can be expressed by the ratio of the cross sectional area of neutron nuclear reaction of the standard nuclide to the cross sectional area of the neutron nuclear reaction of the nuclide as the object of the evaluation. Accordingly, the concentration of the nuclide as the object of the evaluation can be expressed by an analysis formula shown by an analysis function for the ratio of the concentration of the standard nuclide to the cross section of the neutron nuclear reaction. As a result, by giving an optional concentration of the standard nuclide to the analysis formula, the concentration of each of other nuclides can be determined analytically. (I.S.)

  18. Theoretical evaluation of the double U-core switched reluctance machine

    DEFF Research Database (Denmark)

    Jæger, Rasmus; Nielsen, Simon Staal; Rasmussen, Peter Omand

    2017-01-01

    The switched reluctance machine (SRM) has seen a lot of interest due to its simplicity and ruggedness. Much attention have been paid in academia to improve on some of the disadvantages of the technology such as torque ripple, acoustic noise and low torque density. In this paper a topology, namely...... the double U-core SRM, is reviewed. This topology improves on some of the disadvantages of the regular SRM. Torque ripple is reduced and the torque density is increased for the same amount of material, by reconfiguring the topology of the regular SRM and increasing the number of poles. The result...... is a segmented stator structure where each segment can be wound individually and assembled afterwards. Several similar technologies have been demonstrated, and the claimed advantages have been proven in comparison with regular SRMs with a lower pole count. In this paper, the technology will be compared...

  19. Environmental Evaluation of Building Materials of 5 Slovak Buildings

    Science.gov (United States)

    Porhincak, Milan; Estokova, Adriana

    2013-11-01

    Building activity has recently led to the deterioration of environment and has become unsustainable. Several strategies have been introduced in order to minimize consumption of energy and resulting CO2 emissions having their origin in the operational phase. But also other stages of Life Cycle should are important to identify the overall environmental impact of construction sector. In this paper 5 similar Slovak buildings (family houses) were analyzed in terms of environmental performance of building materials used for their structures. Evaluation included the weight of used materials, embodied energy and embodied CO2 and SO2 emissions. Analysis has proven that the selection of building materials is an important factor which influences the environmental profile. Findings of the case study indicated that materials like concrete, ceramic or thermal insulation materials based on polystyrene and mineral wool are ones with the most negative environmental impact.

  20. PIE technology on mechanical tests for HTTR core component and structural materials developed at Research Hot Laboratory

    International Nuclear Information System (INIS)

    Kizaki, Minoru; Honda, Junichi; Usami, Kouji; Ouchi, Asao; Oeda, Etsuro; Matsumoto, Seiichiro

    2001-02-01

    The high temperature engineering test reactor (HTTR) with the target operation temperature of 950degC established the first criticality on November, 1998 based on a large amount of R and D results on fuel and materials. In such R and D works, the development of reactor materials are one of the key issues from the view point of reactor environments such as extremely high temperature, neutron irradiation and so on for the HTTR. The Research Hot Laboratory (RHL) had carried out much kind of post irradiation examinations (PIEs) on core component and pressure vessel materials for during more than a quarter century. And obtained data played an important role in development, characterization and licensing of those materials for the HTTR. This paper describes the PIE technology developed at RHL and typical results on mechanical tests such as elevated temperature tensile and creep rupture tests for Hasteloy-X, Incolloy 800H and so on, and Charpy impact, J IC fracture toughness, K Id fracture toughness and small punch tests for normalized and tempered 2 1/4Cr-1Mo steel from historical view. In addition, an electrochemical test technique established for investigating the irradiation embrittlement mechanism on 2 1/4Cr-1Mo steel is also mentioned. (author)

  1. Evaluation of the toxicity of radiosterilized implantable materials

    International Nuclear Information System (INIS)

    Lewandowska-Szumiel, M.; Kudelska, D.; Mazur, M.; Zimek, Z.

    1997-01-01

    Autoclave and radiation sterilization modes of selected biomaterials and polymers were studied to evaluate the toxicity, if any, induced in the cells grown in vitro. The materials examined induced: crystalline and amorphous biocarbon, alumina, hydroxyapatite, powdered primary PP (radiation-sensitive), and PP modified with a polypropylene/ethylene or an ethylene/vinyl acetate copolymer to enhance its radiation resistance. Results showed no material to be toxic toward the cell examined. The viability of the cells cultivated in the presence of materials examined was found to remain unaffected regardless of the sterilization mode. (author). 12 refs, 3 figs

  2. Proceedings of JSPS-CAS core university program seminar on target materials

    International Nuclear Information System (INIS)

    Gu, Z.Z.; Norimatsu, T.

    2008-02-01

    China-Japan Bilateral Collaboration on the Study of Ultrahigh Density Plasma has been established since 2001 and its second phase is conducting from 2006. Target materials are key issue of the Study of Ultrahigh Density Plasma, and the second of target fabrication was opened at the 2005 Workshop on Ultrahigh Density Plasma Production, Application and theory for Laser Fusion at Nine Village Valley, Sichuan. It achieved great successes in high-level academic exchange and efficient presentation of state-of-the-art development in this research field. However, in order to attract greater attention and participation of more scientists in these fields, the organizing committee decided to further specify and enlarge the scale of the workshop to be China-Japan Bilateral Seminar on Target Materials 2007 in Huang Shan in southern Anhui Province of east China. The seminar had more than 20 participants from 7 universities and 3 institutes in Japan and China. They exchanged state-of-the-art development in nanomaterials, capsule fabrication and low density materials toward target of high power laser. This issue is the collection of the paper presented at the seminar. The 17 of the presented papers are indexed individually. (J.P.N.)

  3. First In-Core Measurement Results Obtained with the Innovative Mobile Calorimeter CALMOS inside the OSIRIS Material Testing Reactor

    International Nuclear Information System (INIS)

    Carcreff, Hubert; Salmon, Laurent; Courtaux, Cedric

    2013-06-01

    Nuclear heating rate inside an MTR has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry [1, 2]. An innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. The development of the calorimetric probe required the manufacturing and the irradiation of mock-ups in the ex-core area, where nuclear heating rate does not exceed 2 W.g -1 . The calorimeter working mode, the different measurement procedures allowed with such a new probe and main modeling and experimental results have been already presented [3, 4]. In this paper, we present the first results obtained during several measurement campaigns carried out in 2012 and 2013 inside the OSIRIS core with the final device. For the first time, this new experimental measurement system was operated in nominal in-core thermo hydraulic conditions with nominal neutron and gamma fluxes (up to 6 W.g -1 ) in several experimental locations. After a brief presentation of the displacement system specificities, first nuclear heating distributions are presented and discussed. Experimental data were also used to upgrade the Finite Element model of the calorimeter in order to match measured temperatures with calculated ones. This model allowed to estimate a Kc correction factor which takes into account small nonlinearities when the heating rate is deduced from the calibration method. A comparison is made between nuclear heating rates determined from the probe calibration and from the zero method. In addition, an evaluation of the global uncertainty associated to the measurements is detailed. Finally, a global comparison is made with available measurements obtained from previous calorimeters. (authors)

  4. Pattern recognition approach to nondestructive evaluation of materials

    International Nuclear Information System (INIS)

    Chen, C.H.

    1987-01-01

    In this paper, a pattern recognition approach to the ultrasonic nondestructive evaluation of materials is examined. Emphasis is placed on identifying effective features from time and frequency domains, correlation functions and impulse responses to classify aluminum plate specimens into three major defect geometry categories: flat, angular cut and circular hole defects. A multi-stage classification procedure is developed which can further determine the angles and sizes for defect characterization and classification. The research clearly demonstrates that the pattern recognition approach can significantly improve the nondestructive material evaluation capability of the ultrasonic methods without resorting to the solution of highly complex mathematical inverse problems

  5. Quantitative Evaluation of Delamination Inside of Composite Materials by ESPI

    International Nuclear Information System (INIS)

    Kim, Koung Suk; Yang, Kwang Young; Kang, Ki Soo; Ji, Chang June

    2004-01-01

    Electronic speckle pattern interferometry (ESPI) for quantitative evaluation of delaminations inside of a composite material plate is described. Delaminations caused by the impact on composite materials are difficult to detect visual inspection and ultrasonic testing due to non-homeogenous structure. This paper proposes the quantitative evaluation technique of the defects made in the composite plates by impact load. Artificial defects are introduced inside of the composite plate for the development of a reliable ESPI inspection technique. Real defects produced by impact tester are inspected and compared with the results of visual inspection which shows a good agreement within 5% error

  6. Stress relaxation and creep of high-temperature gas-cooled reactor core support ceramic materials: a literature search

    International Nuclear Information System (INIS)

    Selle, J.E.; Tennery, V.J.

    1980-05-01

    Creep and stress relaxation in structural ceramics are important properties to the high-temperature design and safety analysis of the core support structure of the HTGR. The ability of the support structure to function for the lifetime of the reactor is directly related to the allowable creep strain and the ability of the structure to withstand thermal transients. The thermal-mechanical response of the core support pads to steady-state stresses and potential thermal transients depends on variables, including the ability of the ceramics to undergo some stress relaxation in relatively short times. Creep and stress relaxation phenomena in structural ceramics of interest were examined. Of the materials considered (fused silica, alumina, silicon nitride, and silicon carbide), alumina has been more extensively investigated in creep. Activation energies reported varied between 482 and 837 kJ/mole, and consequently, variations in the assigned mechanisms were noted. Nabarro-Herring creep is considered as the primary creep mechanism and no definite grain size dependence has been identified. Results for silicon nitride are in better agreement with reported activation energies. No creep data were found for fused silica or silicon carbide and no stress relaxation data were found for any of the candidate materials. While creep and stress relaxation are similar and it is theoretically possible to derive the value of one property when the other is known, no explicit demonstrated relationship exists between the two. For a given structural ceramic material, both properties must be experimentally determined to obtain the information necessary for use in high-temperature design and safety analyses

  7. Synthesis and properties of Pr-substituted MgZn ferrites for core materials and high frequency applications

    Energy Technology Data Exchange (ETDEWEB)

    Mukhtar, Muhammad Waqas; Irfan, Muhammad [Department of Physics, Federal Urdu University of Arts, Science and Technology, Islamabad 44000 (Pakistan); Ahmad, Ishtiaq; Ali, Ihsan [Department of Physics, Bahauddin Zakariya University, Multan 60800 (Pakistan); Akhtar, Majid Niaz [Department of Physics, COMSATS Institute of Information Technology, Lahore (Pakistan); Khan, Muhammad Azhar [Department of Physics, Islamia University, Bahawalpur (Pakistan); Abbas, Ghazanfar [Department of Physics, COMSATS Institute of Information Technology, Islamabad 44000 (Pakistan); Rana, M.U. [Center of Excellence in Solid State Physics, University of the Punjab, Lahore (Pakistan); Ali, Akbar [Department of Basic Sciences, Riphah International University, Islamabad-44000 (Pakistan); Ahmad, Mukhtar, E-mail: ahmadmr25@yahoo.com [Department of Physics, COMSATS Institute of Information Technology, Islamabad 44000 (Pakistan)

    2015-05-01

    A series of single phase spinel ferrites having chemical formula Mg{sub 0.5}Zn{sub 0.5}Pr{sub x}Fe{sub 2−x}O{sub 4} (x=0.00, 0.05, 0.10, 0.15, 0.20, 0.25) were prepared using the sol–gel technique after sintering at 700 °C. The thermal decomposition behavior of an as prepared powder was investigated by means of DTA/TGA analyses. The sintered powders were then characterized by Fourier transform infrared spectroscope, X-ray diffraction, scanning electron microscope, energy dispersive X-ray spectroscope and vibrating sample magnetometer. X-ray diffraction patterns confirm the single phase spinel structure of prepared ferrites without the presence of any impurity phase. The value of lattice parameter (a) increases with the increase of Pr contents (x) into the spinel lattice. The grain size estimated from electron microscope images is in the range of 2.75–5.4 µm which confirms the spinel crystalline nature of the investigated samples. The saturation magnetization (M{sub s}) decreases whereas coercivity (H{sub c}) increases with the increase of Pr contents (x). The measured parameters suggest that these materials are favorable for high frequency applications and as core materials. - Highlights: • Pr-substituted spinel ferrites synthesized by autocombustion route have been investigated. • The average grain size was in the range of 2.75–5.4 µm estimated by SEM technique. • The (M{sub s}) decreases whereas (H{sub c}) increases with the increase of Pr contents (x). • These parameters are favorable for high frequency applications and as core materials.

  8. Synthesis and properties of Pr-substituted MgZn ferrites for core materials and high frequency applications

    International Nuclear Information System (INIS)

    Mukhtar, Muhammad Waqas; Irfan, Muhammad; Ahmad, Ishtiaq; Ali, Ihsan; Akhtar, Majid Niaz; Khan, Muhammad Azhar; Abbas, Ghazanfar; Rana, M.U.; Ali, Akbar; Ahmad, Mukhtar

    2015-01-01

    A series of single phase spinel ferrites having chemical formula Mg 0.5 Zn 0.5 Pr x Fe 2−x O 4 (x=0.00, 0.05, 0.10, 0.15, 0.20, 0.25) were prepared using the sol–gel technique after sintering at 700 °C. The thermal decomposition behavior of an as prepared powder was investigated by means of DTA/TGA analyses. The sintered powders were then characterized by Fourier transform infrared spectroscope, X-ray diffraction, scanning electron microscope, energy dispersive X-ray spectroscope and vibrating sample magnetometer. X-ray diffraction patterns confirm the single phase spinel structure of prepared ferrites without the presence of any impurity phase. The value of lattice parameter (a) increases with the increase of Pr contents (x) into the spinel lattice. The grain size estimated from electron microscope images is in the range of 2.75–5.4 µm which confirms the spinel crystalline nature of the investigated samples. The saturation magnetization (M s ) decreases whereas coercivity (H c ) increases with the increase of Pr contents (x). The measured parameters suggest that these materials are favorable for high frequency applications and as core materials. - Highlights: • Pr-substituted spinel ferrites synthesized by autocombustion route have been investigated. • The average grain size was in the range of 2.75–5.4 µm estimated by SEM technique. • The (M s ) decreases whereas (H c ) increases with the increase of Pr contents (x). • These parameters are favorable for high frequency applications and as core materials

  9. Weightless Environment Training Facility (WETF) materials coating evaluation, volume 1

    Science.gov (United States)

    1995-01-01

    The Weightless Environment Training Facility Material Coating Evaluation project has included preparing, coating, testing, and evaluating 800 test panels of three differing substrates. Ten selected coating systems were evaluated in six separate exposure environments and subject to three tests for physical properties. Substrate materials were identified, the manner of surface preparation described, and exposure environments defined. Exposure environments included immersion exposure, cyclic exposure, and field exposure. Cyclic exposures, specifically QUV-Weatherometer and the KTA Envirotest were found to be the most agressive of the environments included in the study when all three evaluation criteria are considered. This was found to result primarily from chalking of the coatings under ultraviolet (UV) light exposure. Volumes 2 and 3 hold the 5 appendices to this report.

  10. Bidirectional Thermo-Mechanical Properties of Foam Core Materials Using DIC

    DEFF Research Database (Denmark)

    Taher, Siavash Talebi; Thomsen, Ole Thybo; M Dulieu-Barton, Janice

    2011-01-01

    mechanical properties at room and at elevated temperatures. The MAF enables the realization of pure compression or high compression to shear bidirectional loading conditions that is not possible with conventional Arcan fixtures. The MAF is attached to a standard universal test machine equiped...... with an environmental chamber using specially designed grips that allow the specimen to rotate, and hence reduces paristic effects due to misalignment. The objective is to measure the unidirectional and bidirectional mechanical properties of PVC foam materials at elevated tempreature using digital image correlation...

  11. Uncertainty Evaluation of Reactivity Coefficients for a large advanced SFR Core Design

    International Nuclear Information System (INIS)

    Khamakhem, Wassim; Rimpault, Gerald

    2008-01-01

    Sodium Cooled Fast Reactors are currently being reshaped in order to meet Generation IV goals on economics, safety and reliability, sustainability and proliferation resistance. Recent studies have led to large SFR cores for a 3600 MWth power plants, cores which exhibit interesting features. The designs have had to balance between competing aspects such as sustainability and safety characteristics. Sustainability in neutronic terms is translated into positive breeding gain and safety into rather low Na void reactivity effects. The studies have been done on two SFR concepts using oxide and carbide fuels. The use of the sensitivity theory in the ERANOS determinist code system has been used. Calculations have been performed with different sodium evaluations: JEF2.2, ERALIB-1 and the most recent JEFF3.1 and ENDF/B-VII in order to make a broad comparison. Values for the Na void reactivity effect exhibit differences as large as 14% when using the different sodium libraries. Uncertainties due to nuclear data on the reactivity coefficients were performed with BOLNA variances-covariances data, the Na Void Effect uncertainties are near to 12% at 1σ. Since, the uncertainties are far beyond the target accuracy for a design achieving high performance, two directions are envisaged: the first one is to perform new differential measurements or in a second attempt use integral experiments to improve effectively the nuclear data set and its uncertainties such as performed in the past with ERALIB1. (authors)

  12. Evaluation of DNBR calculation methods for advanced digital core protection system

    International Nuclear Information System (INIS)

    Ihn, W. K.; Hwang, D. H.; Pak, Y. H.; Yoon, T. Y.

    2003-01-01

    This study evaluated the on-line DNBR calculation methods for an advanced digital core protection system in PWR, i.e., subchannel analysis and group-channel analysis. The subchannel code MATRA and the four-channel codes CETOP-D and CETOP2 were used here. CETOP2 is most simplified DNBR analysis code which is implemented in core protection calculator in Korea standard nuclear power plants. The detailed subchannel code TORC was used as a reference calculation of DNBR. The DNBR uncertainty and margin were compared using allowable operating conditions at Yonggwang nuclear units 3-4. The MATRA code using a nine lumping-channel model resulted in smaller mean and larger standard deviation of the DNBR error distribution. CETOP-D and CETOP2 showed conservatively biased mean and relatively smaller standard deviation of the DNBR error distribution. MATRA and CETOP-D w.r.t CETOP2 showed significant increase of the DNBR available margin at normal operating condition. Taking account for the DNBR uncertainty, MATRA and CETOP-D over CETOP2 were estimated to increase the DNBR net margin by 2.5%-9.8% and 2.5%-3.3%, respectively

  13. Efficient multiscale magnetic-domain analysis of iron-core material under mechanical stress

    Science.gov (United States)

    Nishikubo, Atsushi; Ito, Shumpei; Mifune, Takeshi; Matsuo, Tetsuji; Kaido, Chikara; Takahashi, Yasuhito; Fujiwara, Koji

    2018-05-01

    For an efficient analysis of magnetization, a partial-implicit solution method is improved using an assembled domain structure model with six-domain mesoscopic particles exhibiting pinning-type hysteresis. The quantitative analysis of non-oriented silicon steel succeeds in predicting the stress dependence of hysteresis loss with computation times greatly reduced by using the improved partial-implicit method. The effect of cell division along the thickness direction is also evaluated.

  14. Evaluation of neutronic characteristics of STACY 80-cm-diameter cylindrical core fueled with 6% enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Yanagisawa, Hiroshi; Sono, Hiroki

    2003-06-01

    For the examination of neutronic safety design of forthcoming experimental core configurations in the Static Experiment Critical Facility (STACY), neutronic characteristics of 80-cm-diameter cylindrical cores fueled with 6% enriched uranyl nitrate solution have been evaluated by computational analyses. In the analyses, the latest nuclear data library, JENDL-3.3, was used as neutron cross section data. The neutron diffusion and transport calculations were performed using a diffusion code, CITATION, in the SRAC code system and a continuous-energy Monte Carlo code, MVP. Critical level heights of the cores were obtained using such parameters as uranium concentration (up to 500 gU/l), free nitric acid concentration (up to 8 mol/l), and concentration of soluble neutron poisons, gadolinium and boron. It has been confirmed from the evaluation that all critical cores comply with safety criteria required in the STACY operation concerning excess reactivity, reactivity addition rates and shutdown margins by safety rods. (author)

  15. The development and evaluation of reference materials for food microbiology

    NARCIS (Netherlands)

    Veld, in 't P.

    1998-01-01

    Since 1986 the National Institute of Public Health and the Environment (RIVM) has worked on the development and evaluation of microbiological reference materials (RMs) with support from the European Communities Bureau of Reference (BCR), now called Standards Measurement and Testing

  16. Evaluation of material fracture energy by its heat content

    International Nuclear Information System (INIS)

    Frolov, G.A.; Pasichnyj, V.V.; Polezhaev, Yu.V.; Frolov, A.A.; Choba, A.V.

    1986-01-01

    Based on published and experimental data it is shown that there is a simple relationship between the heat of evaporation and heat content. This allows in some instances the evaluation of a rate of material fracture by its content. Experimental and theoretical data for quartz glass ceramics, and glass-reinforced plastic are presented

  17. Standard Practice for Evaluating Solar Absorptive Materials for Thermal Applications

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice covers a testing methodology for evaluating absorptive materials used in flat plate or concentrating collectors, with concentrating ratios not to exceed five, for solar thermal applications. This practice is not intended to be used for the evaluation of absorptive surfaces that are (1) used in direct contact with, or suspended in, a heat-transfer liquid, (that is, trickle collectors, direct absorption fluids, etc.); (2) used in evacuated collectors; or (3) used in collectors without cover plate(s). 1.2 Test methods included in this practice are property measurement tests and aging tests. Property measurement tests provide for the determination of various properties of absorptive materials, for example, absorptance, emittance, and appearance. Aging tests provide for exposure of absorptive materials to environments that may induce changes in the properties of test specimens. Measuring properties before and after an aging test provides a means of determining the effect of the exposure. 1.3 Th...

  18. An evaluation of multigroup flux predictions in the EBR-II core

    International Nuclear Information System (INIS)

    Hill, R.N.; Fanning, T.H.; Finck, P.J.

    1991-01-01

    The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required

  19. Interaction of core self-evaluations and perceived organizational support on work-to-family enrichment.

    Science.gov (United States)

    McNall, Laurel A; Masuda, Aline D; Shanock, Linda Rhoades; Nicklin, Jessica M

    2011-01-01

    The purpose of this article was to offer an empirical test of J. H. Greenhaus and G. N. Powell's (2006) model of work-family enrichment by examining dispositional (i.e., core self-evaluations; CSEs) and situational (i.e., perceived organizational support; POS) factors associated with work-to-family enrichment (WFE) and whether these variables interact in predicting WFE. In a survey of 220 employed adults, our hierarchical regression analysis revealed that in highly supportive work environments, individuals reported high WFE regardless of CSE. However, when POS was low, individuals high in CSEs reported higher WFE than those low in CSEs, in support of conservation of resources theory (S. E. Hobfoll, 2002). Implications for research and practice are discussed.

  20. An evaluation of multigroup flux predictions in the EBR-II core

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.N.; Fanning, T.H.; Finck, P.J.

    1991-12-31

    The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required.

  1. An evaluation of multigroup flux predictions in the EBR-II core

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R.N.; Fanning, T.H.; Finck, P.J.

    1991-01-01

    The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required.

  2. In-core instrumentation and in-situ measurement in connection with fuel behaviour. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The subject of this meeting has been touched on briefly in most of the Specialist's and topical meetings related to fuel behaviour. On the basis of the conclusions and recommendations of these meetings the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended the Agency to organize a dedicated Specialist's Meeting on the subject. The twenty one papers covered the instrumentation, sensors, methods and computer codes currently used in Material Test Reactor (MTR) and power reactors as well as improved instrumentation and methods. The meeting acknowledged the fast development of fuel modelling and therefore the growing need of dedicated high burnup fuel experiments carried out in MTR reactors on refabricated rods from power reactors. In order to reduce safety margins in power reactors, thus improving economics, the necessity to develop more sophisticated on-line calculations, based on improved sensors, was recognized, although this development is limited by insufficient knowledge of the mechanisms involved. Refs, figs, tabs

  3. Statistical evaluation of the on line core monitoring effectiveness for limiting the consequences of the fuel assembly misloading event

    International Nuclear Information System (INIS)

    Molnar, A.; Kereszturi, A.; Temesvari, E.; Korpas, L.

    2007-01-01

    In WWER-440 type reactors, on line core monitoring is used for the early indication of such abnormal events like fuel assembly misloading, inadvertent misalignment of Control Assemblies, blockage of coolant channels. The paper is focusing on the assembly misloading, which can not be indicated by other measurements. A Monte Carlo method was developed and applied to evaluate the on line core monitoring effectiveness for the indication of this abnormal event during the power increase in due time, when the consequences are still acceptable. The investigations proved the satisfactory effectiveness of the online core monitoring down to 55 % power even in case when 75 % of the temperature measurements was only available (Authors)

  4. Physical properties of molten core materials: Zr-Ni and Zr-Cr alloys measured by electrostatic levitation

    Energy Technology Data Exchange (ETDEWEB)

    Ohishi, Yuji, E-mail: ohishi@see.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University (Japan); Kondo, Toshiki [Graduate School of Engineering, Osaka University (Japan); Ishikawa, Takehiko [Japan Aerospace Exploration Agency (Japan); SOKEN-DAI (Graduate University for Advanced Studies) (Japan); Okada, Junpei T. [Institute for Materials Research, Tohoku University (Japan); Watanabe, Yuki [Advanced Engineering Services Co. Ltd. (Japan); Muta, Hiroaki; Kurosaki, Ken [Graduate School of Engineering, Osaka University (Japan); Yamanaka, Shinsuke [Graduate School of Engineering, Osaka University (Japan); Research Institute of Nuclear Engineering, University of Fukui (Japan)

    2017-03-15

    It is important to understand the behaviors of molten core materials to investigate the progression of a core meltdown accident. In the early stages of bundle degradation, low-melting-temperature liquid phases are expected to form via the eutectic reaction between Zircaloy and stainless steel. The main component of Zircaloy is Zr and those of stainless steel are Fe, Ni, and Cr. Our group has previously reported physical property data such as viscosity, density, and surface tension for Zr-Fe liquid alloys using an electrostatic levitation technique. In this study, we report the viscosity, density, and surface tension of Zr-Ni and Zr-Cr liquid alloys (Zr{sub 1-x}Ni{sub x} (x = 0.12 and 0.24) and Zr{sub 0.77}Cr{sub 0.23}) using the electrostatic levitation technique. - Highlights: • The physical properties of Zr-Ni and Zr-Cr liquid alloys have been measured Zr{sub 1-x}Ni{sub x} (x = 0.12 and 0.24) and Zr{sub 77}Cr{sub 23}. • The measurement was conducted using the electrostatic levitation technique. • The density, viscosity, and surface tension of each liquid alloy were measured.

  5. Physical properties of molten core materials: Zr-Ni and Zr-Cr alloys measured by electrostatic levitation

    International Nuclear Information System (INIS)

    Ohishi, Yuji; Kondo, Toshiki; Ishikawa, Takehiko; Okada, Junpei T.; Watanabe, Yuki; Muta, Hiroaki; Kurosaki, Ken; Yamanaka, Shinsuke

    2017-01-01

    It is important to understand the behaviors of molten core materials to investigate the progression of a core meltdown accident. In the early stages of bundle degradation, low-melting-temperature liquid phases are expected to form via the eutectic reaction between Zircaloy and stainless steel. The main component of Zircaloy is Zr and those of stainless steel are Fe, Ni, and Cr. Our group has previously reported physical property data such as viscosity, density, and surface tension for Zr-Fe liquid alloys using an electrostatic levitation technique. In this study, we report the viscosity, density, and surface tension of Zr-Ni and Zr-Cr liquid alloys (Zr 1-x Ni x (x = 0.12 and 0.24) and Zr 0.77 Cr 0.23 ) using the electrostatic levitation technique. - Highlights: • The physical properties of Zr-Ni and Zr-Cr liquid alloys have been measured Zr 1-x Ni x (x = 0.12 and 0.24) and Zr 77 Cr 23 . • The measurement was conducted using the electrostatic levitation technique. • The density, viscosity, and surface tension of each liquid alloy were measured.

  6. Experimental investigation of material chemical effects on emergency core cooling pump suction filter performance after loss of coolant accident

    International Nuclear Information System (INIS)

    Park, Jong Woon; Park, Byung Gi; Kim, Chang Hyun

    2009-01-01

    Integral tests of head loss through an emergency core cooling filter screen are conducted, simulating reactor building environmental conditions for 30 days after a loss of coolant accident. A test rig with five individual loops each of whose chamber is established to test chemical product formation and measure the head loss through a sample filter. The screen area at each chamber and the amounts of reactor building materials are scaled down according to specific plant condition. A series of tests have been performed to investigate the effects of calcium-silicate, reactor building spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the filter screen is strongly affected by spray duration and the head loss increase is rapid at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKON TM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

  7. Evaluation of nickel-based materials for VHTR heat exchanger

    International Nuclear Information System (INIS)

    Burlet, H.; Gentzbittel, J.M.; Cabet, C.; Lamagnere, P.; Blat, M.; Renaud, D.; Dubiez-Le Goff, S.; Pierron, D.

    2008-01-01

    Two available conventional nickel-based alloys (617 and 230) have been selected as structural materials for the advanced gas-cooled reactors, especially for the heat exchanger. An extensive research programme has been launched in France within the framework of the ANTARES programme to evaluate the performances of these materials in VHTR service environment. The experimental work is focused on mechanical properties, thermal stability and corrosion resistance in the temperature range (700-1 000 deg C) over long time. Thus the experimental work includes creep and fatigue tests on as-received materials, short- and medium-term thermal exposure tests followed by tensile and impact toughness tests, short- and medium-term corrosion exposure tests under impure He environment. The status of the results obtained up to now is given in this paper. Additional tests such as long-term thermal ageing and long-term corrosion tests are required to conclude on the selection of the material. (author)

  8. Core to College Evaluation: Statewide Networks. Connecting Education Systems and Stakeholders to Support College Readiness

    Science.gov (United States)

    Bracco, Kathy Reeves; Klarin, Becca; Broek, Marie; Austin, Kim; Finkelstein, Neal; Bugler, Daniel; Mundry, Susan

    2014-01-01

    The Core to College initiative aims to facilitate greater coordination between K-12 and postsecondary education systems around implementation of the Common Core State Standards and aligned assessments. Core to College grants have been awarded to teams in Colorado, Florida, Hawaii, Kentucky, Louisiana, Massachusetts, North Carolina, Oregon,…

  9. Evaluation of core modeling effect on transients for multi-flow zone design of SFR

    International Nuclear Information System (INIS)

    Shin, Andong; Choi, Yong Won

    2016-01-01

    SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

  10. Evaluation of core modeling effect on transients for multi-flow zone design of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Choi, Yong Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

  11. Developing a Framework for Qualitative Evaluation of Urban Interventions in Iranian Historical Cores

    Directory of Open Access Journals (Sweden)

    Azadeh Arjomand Kermani

    2017-11-01

    Full Text Available Iranian historic city cores are important parts of modern cities because of their valuable monuments and morphology but are also significant because of their population density, location and the major governmental functions they house. Since 1920, modernisation policies and urban development trends in Iran have justified spatial transformation and redevelopment and the demolition and destruction of traditional urban fabrics as a way to provide contemporary requirements and hygiene improvements for the residents. As the UNESCO recommendation on the Historic Urban Landscape argues, historic urban cores can only sustain their role in the daily life of the city by getting prepared for and participating in this transformation process. Disagreement over the value of historic urban cores on the one hand and inevitable modification of urban areas in a developing country like Iran on the other, creates a problematic condition for the preservation of the historic environment. The Valletta Principles for the Safeguarding and Management of Historic Cities, Towns and Urban Areas states that historic towns and urban areas require an integrated approach including their “protection, conservation, enhancement and management as well as their coherent development and their harmonious adaptation to contemporary life”. In order to support the process of reaching a balance between these spatial targets in Iran, this research discusses the relation between urban transformation projects and their heritage context. In doing so it connects international literature on urban quality and traditional Iranian urban forms to contemporary Iranian urban design practice. To achieve this integration between urban heritage and spatial development, a framework of quality attributes has been developed to evaluate urban interventions in a heritage context. The three main pillars of this framework have been extracted from and inspired by international literature and guidelines

  12. Mathematical modelling of powder material motion and transportation in high-temperature flow core during plasma coatings application

    Science.gov (United States)

    Bogdanovich, V. I.; Giorbelidze, M. G.

    2018-03-01

    A problem of mathematical modelling of powder material motion and transportation in gas thermal flow core has been addressed. Undertaken studies indicate significant impact on dynamics of motion of sprayed particles of phenomenological law for drag coefficient and accounting momentum loss of a plasma jet upon acceleration of these particles and their diameter. It is determined that at great dispersion of spraying particles, they reach detail surface at different velocity and significant particles separation takes place at spraying spot. According to the results of mathematical modelling, requirements for admissible dispersion of diameters of particles used for spraying have been formulated. Research has also allowed reducing separation of particles at the spraying spot due to the selection of the method of powder feed to the anode channel of the plasma torch.

  13. ROP MATHEMATICAL MODEL OF ROTARY-ULTRASONIC CORE DRILLING OF BRITTLE MATERIAL

    Directory of Open Access Journals (Sweden)

    Mera Fayez Horne

    2017-03-01

    Full Text Available The results from the Phoenix mission led scientists to believe it is possible that primitive life exists below the Martian surface. Therefore, drilling in Martian soil in search for organisms is the next logical step. Drilling on Mars is a major engineering challenge due to the drilling depth requirement and extreme environment condition. Mars lacks a thick atmosphere and a continuous magnetic field that shield the planet’s surface from solar radiation and solar flares. As a result, the Martian surface is sterile and if life ever existed, it must be found below the surface. NASA’s Mars Exploration Payload Advisory Group proposed that drilling should be considered as a priority investigation on Mars in an effort of finding evidence of extinct or extant life. The results from the Curiosity mission suggested drilling six meters deep in the red planet in search for life. Excavation tools deployed to Mars so far have been able to drill to a maximum depth of 6.5 cm. Thus, the drilling capabilities need to be increased by a factor of approximately 100 to achieve the goal of drilling six meters deep. This requirement puts a demand on developing new and more effective technologies to reach this goal. Previous research shows evidence of a promising drilling mechanism in rotary-ultrasonic for what it offers in terms of high surface quality, faster rate of penetration and higher material removal rate. This research addresses the need to understand the mechanics of the drill bit tip and rock interface in rotary-ultrasonic drilling performance of one drill bit at a time drilling in three types of rocks that vary in strength. A mathematical model identifying all contributing independent parameters, such as drill bit design parameters, drilling process parameters, ultrasonic wave amplitude and rocks’ material properties, that have effect on rate of penetration is developed. Analytical and experimental results under ambient condition are presented to show

  14. EFL Published Materials: An Evaluation of English Textbooks for Junior High School in Indonesia

    Directory of Open Access Journals (Sweden)

    Rizaldy Hanifa

    2018-04-01

    Full Text Available The use of EFL published materials like textbooks are becoming more widespread as they can bring easiness in the classroom by providing teachers with guidelines comprised of syllabus, methodologies, as well as materials for teaching and learning. However, choosing a suitable textbook for their teaching situation is deemed to be one of the most challenging tasks that EFL teachers often face. To get a good picture of the suitability of a textbook, a careful investigation needs to be undertaken. This study focused on the analysis and evaluation of two different English textbooks addressed to junior high schools grade VII in Indonesia, KTSP and curriculum 2013 textbook. Harmer’s (2007 framework was employed to figure out the strengths and weaknesses of each textbook. The analysis revealed that both textbooks are quite satisfactory as they are very affordable, contain interesting layout, attractive designs, and clear instructions, correspond to current ELT methodology, cover all language skills, and comprise a wide range of topics which are familiar and culturally appropriate for learners. Nonetheless, the KTSP textbook does not have add-ons and extra materials; meanwhile, curriculum 2013 textbook is weak in providing authentic listening materials. Therefore, although the teachers manage to use the textbooks as their core materials, they are supposed to make adjustment and supplement them with other materials according to their learners’ needs and their teaching context.

  15. Intergranular cracking mechanism in baffle former bolt materials for PWR core internals

    Energy Technology Data Exchange (ETDEWEB)

    Yonezawa, Toshio; Arioka, Koji; Kanasaki, Hiroshi; Fujimoto, Koji [Takasago R and D Center, Mitsubishi Heavy Industries Ltd., Takasago, Hyogo (Japan); Ajiki, Kazuhide [Kobe Shipyard and Machinery, Mitsubishi Heavy Industries Ltd., Kobe, Hyogo (Japan); Matsuoka, Takanori [Nuclear Development Corp., Tokai, Ibaraki (Japan); Urata, Sigeru; Mizuta, Hitoshi [Kansai Electric Power Co., Inc., Osaka (Japan)

    2000-03-01

    In this study, the cause of intergranular cracking in baffle former bolts(BFBs) was estimated from metallurgical and chemical viewpoints based upon the experimental data and information published by EdF. At first, five kinds of possibilities were estimated as the cause of intergranular cracking in BFBs. Five possibilities estimated were (1) mechanical cracking caused by high strain in irradiation hardened austenitic stainless steels, (2) O{sub 2} SCC due to residual oxygen in the bolt stagnant region, (3) caustic SCC due to dry and wet phenomenon, (4) low pH SCC due to oxygen concentration cell, and (5) PWSCC due to radiation induced segregation. In this study each possibility was evaluated by the calculation and some out of pile tests. And also, the cause of the intergranular cracking in BFBs was estimated by the data of the post-irradiation examinations and basic out of pile tests for Type 316CW and Type 347 stainless steels in the authors' previous study. From these evaluation, the intergranular cracking in BFBs seems to be caused by the PWSCC, but not caused by mechanical cracking O{sub 2} SCC, caustic SCC or low pH SCC. (author)

  16. A methodology for evaluating weighting functions using MCNP and its application to PWR ex-core analyses

    International Nuclear Information System (INIS)

    Pecchia, Marco; Vasiliev, Alexander; Ferroukhi, Hakim; Pautz, Andreas

    2017-01-01

    Highlights: • Evaluation of neutron source importance for a given tally. • Assessment of ex-core detector response plus its uncertainty. • Direct use of neutron track evaluated by a Monte Carlo neutron transport code. - Abstract: The ex-core neutron detectors are commonly used to control reactor power in light water reactors. Therefore, it is relevant to understand the importance of a neutron source to the ex-core detectors response. In mathematical terms, this information is conveniently represented by the so called weighting functions. A new methodology based on the MCNP code for evaluating the weighting functions starting from the neutron history database is presented in this work. A simultaneous evaluation of the weighting functions in a user-given Cartesian coverage mesh is the main advantage of the method. The capability to generate weighting functions simultaneously in both spatial and energy ranges is the innovative part of this work. Then, an interpolation tool complements the methodology, allowing the generation of weighting functions up to the pin-by-pin fuel segment, where a direct evaluation is not possible due to low statistical precision. A comparison to reference results provides a verification of the methodology. Finally, an application to investigate the role of ex-core detectors spatial location and core burnup for a Swiss nuclear power plant is provided.

  17. CORE SELF-EVALUATIONS, JOB SATISFACTION, TRANSFORMATIONAL AND SERVANT LEADERSHIP MODEL IN THE ROMAN CATHOLIC EDUCATION SYSTEM

    Directory of Open Access Journals (Sweden)

    Mary Caroline N. Castano

    2017-12-01

    Full Text Available Core self-evaluations, effective leadership styles and employee job satisfaction are essential factors for organizational success. This paper aims to determine the relationship of the leader’s core self-evaluations, transformational leadership and servant leadership styles to their follower’s job satisfaction in selected Parochial Schools in Manila, Philippines under the Roman Catholic Education System. The respondents were selected according to certain criteria. Descriptive correlational design was used. In total, 308 individuals from the teaching and non-teaching personnel participated. The data were collected using survey questionnaires. Data were analyzed using Partial Least Squares-Structural Equation Modeling (PLS-SEM. Research findings revealed that a positive relationship exists between leader’s core self-evaluations and transformational leadership; core self-evaluations to servant leadership; transformational leadership to job satisfaction; and servant leadership to job satisfaction. These relationships are statistically significant. The relationship of the leader’s core self-evaluations to the follower’s job satisfaction indicated a direct effect but were statistically non-significant on the basis of its p-value. The major contribution of the current study is to extend the limited literature regarding the antecedents of the four (4 selected variables. The researcher recommends to the school leaders to create a motivating environment through a more transformational and servant leadership behavior that will enhance their follower’s work satisfaction.

  18. Application of Dredged Materials and Steelmaking Slag as Basal Media to Restore and Create Seagrass Beds: Mesocosm and Core Incubation Experiments

    Science.gov (United States)

    Tsukasaki, A.; Suzumura, M.; Tsurushima, N.; Nakazato, T.; Huang, Y.; Tanimoto, T.; Yamada, N.; Nishijima, W.

    2016-02-01

    Seagrass beds stabilize bottom sediments, improve water quality and light conditions, enhance species diversity, and provide habitat complexity in coastal marine environments. Seagrass beds are now experiencing worldwide decline by rapid environmental changes. Possible options of seagrass bed restoration are civil engineering works including mounding to raise the bottom to elevations with suitable light for seagrass growth. Reuse or recycling of dredged materials (DM) and various industrial by-products including steelmaking slags is a beneficial option to restore and create seagrass beds. To evaluate the applicability of DM and dephosphorization slag (Slag) as basal media of seagrass beds, we carried out mesocosm experiments and core incubation experiments in a land-based flow-through seawater tank over a year. During the mesocosm experiment, no difference was found in growth of eelgrass (Zostera marina L.) and macrobenthic community structures between Slag-based sediments and sand-based control experiments, even though Slag-based sediments exhibited substantially higher pH than sand-based sediments. During the core incubation experiment, we investigated detailed variation and distributions of pH and nutrients, and diffusion fluxes of nutrients between the sediment/seawater interface. Though addition of Slag induced high pH up to 10.7 in deep layers (sediments, whereas dissolved phosphate concentration was substantially reduced by the addition of Slag. The low concentrations of phosphate was likely due to precipitation with calcium under high pH condition. Diffusion fluxes of nutrients from the cores were comparable with those reported in natural coastal systems. It was suggested that the mixture of Slag and DM is applicable as basal media for construction of artificial seagrass beds.

  19. EVALUATION OF THE HTA CORE MODEL FOR NATIONAL HEALTH TECHNOLOGY ASSESSMENT REPORTS: COMPARATIVE STUDY AND EXPERIENCES FROM EUROPEAN COUNTRIES.

    Science.gov (United States)

    Kõrge, Kristina; Berndt, Nadine; Hohmann, Juergen; Romano, Florence; Hiligsmann, Mickael

    2017-01-01

    The health technology assessment (HTA) Core Model® is a tool for defining and standardizing the elements of HTA analyses within several domains for producing structured reports. This study explored the parallels between the Core Model and a national HTA report. Experiences from various European HTA agencies were also investigated to determine the Core Model's adaptability to national reports. A comparison between a national report on Genetic Counseling, produced by the Cellule d'expertise médicale Luxembourg, and the Core Model was performed to identify parallels in terms of relevant and comparable assessment elements (AEs). Semi-structured interviews with five representatives from European HTA agencies were performed to assess their user experiences with the Core Model. The comparative study revealed that 50 percent of the total number (n = 144) of AEs in the Core Model were relevant for the national report. Of these 144 AEs from the Core Model, 34 (24 percent) were covered in the national report. Some AEs were covered only partly. The interviewees emphasized flexibility in using the Core Model and stated that the most important aspects to be evaluated include characteristics of the disease and technology, clinical effectiveness, economic aspects, and safety. In the present study, the national report covered an acceptable number of AEs of the Core Model. These results need to be interpreted with caution because only one comparison was performed. The Core Model can be used in a flexible manner, applying only those elements that are relevant from the perspective of the technology assessment and specific country context.

  20. Purity Evaluation of Bulk Single Wall Carbon Nanotube Materials

    International Nuclear Information System (INIS)

    Dettlaff-Weglikowska, U.; Hornbostel, B.; Cech, J.; Roth, S.; Wang, J.; Liang, J.

    2005-01-01

    We report on our experience using a preliminary protocol for quality control of bulk single wall carbon nanotube (SWNT) materials produced by the electric arc-discharge and laser ablation method. The first step in the characterization of the bulk material is mechanical homogenization. Quantitative evaluation of purity has been performed using a previously reported procedure based on solution phase near-infrared spectroscopy. Our results confirm that this method is reliable in determining the nanotube content in the arc-discharge sample containing carbonaceous impurities (amorphous carbon and graphitic particles). However, the application of this method to laser ablation samples gives a relative purity value over 100 %. The possible reason for that might be different extinction coefficient meaning different oscillator strength of the laser ablation tubes. At the present time, a 100 % pure reference sample of laser ablation SWNT is not available, so we chose to adopt the sample showing the highest purity as a new reference sample for a quantitative purity evaluation of laser ablation materials. The graphitic part of the carbonaceous impurities has been estimated using X-ray diffraction of 1:1 mixture of nanotube material and C60 as an internal reference. To evaluate the metallic impurities in the as prepared and homogenized carbon nanotube soot inductive coupled plasma (ICP) has been used

  1. Synthesis and Performance of Highly Stable Star-Shaped Polyaniline Electrochromic Materials with Triphenylamine Core

    Science.gov (United States)

    Xiong, Shanxin; Li, Shuaishuai; Zhang, Xiangkai; Wang, Ru; Zhang, Runlan; Wang, Xiaoqin; Wu, Bohua; Gong, Ming; Chu, Jia

    2018-02-01

    The molecular architecture of conducting polymers has a significant impact on their conjugated structure and electrochemical properties. We have investigated the influence of star-shaped structure on the electrochemical and electrochromic properties of polyaniline (PANI). Star-shaped PANI (SPANI) was prepared by copolymerization of aniline with triphenylamine (TPA) using an emulsion polymerization method. With addition of less than 4.0 mol.% TPA, the resulting SPANI exhibited good solubility in xylene with dodecylbenzenesulfonic acid (DBSA) as doping acid. The structure and thermal stability of the SPANI were characterized using Fourier-transform infrared spectroscopy, Raman spectroscopy, and thermogravimetric analysis, and the electrochemical behavior was analyzed by cyclic voltammetry (CV). The electrochromic properties of SPANI were tested using an electrochemical workstation combined with an ultraviolet-visible (UV-Vis) spectrometer. The results show that, with increasing TPA loading, the thermal stability of SPANI increased. With addition of 4.0 mol.% TPA, the weight loss of SPANI was 36.9% at 700°C, much lower than the value of 71.2% for PANI at the same temperature. The low oxidation potential and large enclosed area of the CV curves indicate that SPANI possesses higher electrochemical activity than PANI. Enhanced electrochromic properties including higher optical contrast and better electrochromic stability of SPANI were also obtained. SPANI with 1.6 mol.% TPA loading exhibited the highest optical contrast of 0.71, higher than the values of 0.58 for PANI, 0.66 for SPANI-0.4%, or 0.63 for SPANI-4.0%. Overdosing of TPA resulted in slow switching speed due to slow ion transport in short branched chains of star-shaped PANI electrochromic material. Long-term stability testing confirmed that all the SPANI-based devices exhibited better stability than the PANI-based device.

  2. Evaluation of Applicability for the Core Protection Method with 4-Channel CEA Positions to OPR-1000 Plants

    International Nuclear Information System (INIS)

    Koo, B. S.; Cho, J. Y.; Song, J. S.

    2008-05-01

    To increase the applicability of research results established during the process of integral reactor development program, a new core protection method with 4-channel CEA positions was applied to the domestic commercial plants and its feasibility was evaluated. To achieve above object, state-of-the-art related to core protection system was analyzed as followings: - Unusual CPC operating experience in Korea - Evaluate the proposals of CPC improvement suggested by CPC Task Force Team - Review the conventional CPC used in Yonggwang and Ulchin plants - Review the Common-Q CPC to be implemented in Shin-Kori Units 1 and 2 - Evaluate the SENTINEL core protection system in US and COPS core protection system in Germany - Examine the applied patents in Korea, US and Japan - Analyze the copyright for computer programs used in core protection system design and license agreement for PWR technology between Westinghouse and Korea. In addition, study for the formation of system, design requirement, algorithm improvement and enhancement of operator interface was performed to apply the newly suggested core protection method to the commercial plants. By adopting this method, it is expected that unnecessary channel of reactor trips will be decreased considerably. Although change of system including CEA must be set forth as a prerequisite to apply this method to the domestic commercial plants as well as scheduled plants, this study suggests the strategy and direction for the development of core protection system in domestic. This study, moreover, will provide the valuable information as a basic data in establishing the detailed development plan for the advanced core protection system in the future

  3. The effect of core material, veneering porcelain, and fabrication technique on the biaxial flexural strength and weibull analysis of selected dental ceramics.

    Science.gov (United States)

    Lin, Wei-Shao; Ercoli, Carlo; Feng, Changyong; Morton, Dean

    2012-07-01

    The objective of this study was to compare the effect of veneering porcelain (monolithic or bilayer specimens) and core fabrication technique (heat-pressed or CAD/CAM) on the biaxial flexural strength and Weibull modulus of leucite-reinforced and lithium-disilicate glass ceramics. In addition, the effect of veneering technique (heat-pressed or powder/liquid layering) for zirconia ceramics on the biaxial flexural strength and Weibull modulus was studied. Five ceramic core materials (IPS Empress Esthetic, IPS Empress CAD, IPS e.max Press, IPS e.max CAD, IPS e.max ZirCAD) and three corresponding veneering porcelains (IPS Empress Esthetic Veneer, IPS e.max Ceram, IPS e.max ZirPress) were selected for this study. Each core material group contained three subgroups based on the core material thickness and the presence of corresponding veneering porcelain as follows: 1.5 mm core material only (subgroup 1.5C), 0.8 mm core material only (subgroup 0.8C), and 1.5 mm core/veneer group: 0.8 mm core with 0.7 mm corresponding veneering porcelain with a powder/liquid layering technique (subgroup 0.8C-0.7VL). The ZirCAD group had one additional 1.5 mm core/veneer subgroup with 0.7 mm heat-pressed veneering porcelain (subgroup 0.8C-0.7VP). The biaxial flexural strengths were compared for each subgroup (n = 10) according to ISO standard 6872:2008 with ANOVA and Tukey's post hoc multiple comparison test (p≤ 0.05). The reliability of strength was analyzed with the Weibull distribution. For all core materials, the 1.5 mm core/veneer subgroups (0.8C-0.7VL, 0.8C-0.7VP) had significantly lower mean biaxial flexural strengths (p Empress and e.max groups, regardless of core thickness and fabrication techniques. Comparing fabrication techniques, Empress Esthetic/CAD, e.max Press/CAD had similar biaxial flexural strength (p= 0.28 for Empress pair; p= 0.87 for e.max pair); however, e.max CAD/Press groups had significantly higher flexural strength (p Empress Esthetic/CAD groups. Monolithic core

  4. Software for MUF evaluating in item nuclear material accounting

    International Nuclear Information System (INIS)

    Wang Dong; Zhang Quanhu; He Bin; Wang Hua; Yang Daojun

    2009-01-01

    Nuclear material accounting is a key measure for nuclear safeguard. Software for MUF evaluation in item nuclear material accounting was worked out in this paper. It is composed of several models, including input model, data processing model, data inquiring model, data print model, system setting model etc. It could be used to check the variance of the measurement and estimate the confidence interval according to the MUF value. To insure security of the data multi-user management function was applied in the software. (authors)

  5. Tribo-performance evaluation of ecofriendly brake friction composite materials

    Science.gov (United States)

    Kumar, Naresh; Singh, Tej; Grewal, G. S.

    2018-05-01

    This paper presents the potential of natural fibre in brake friction materials. Natural fibre filled ecofriendly brake friction materials were developed without Kevlar fibre evaluated for tribo-performance on a chase friction testing machine following SAE J 661a standard. Experimental results indicated that natural fibre enhances the fade performance, but depresses the friction and wear performance, whereas Kevlar fibre improves the friction, wear and recovery performance but depresses the fade performance. Also the results revealed that with the increase in natural fibre content, the friction and fade performances enhanced.

  6. [Fundamental and clinical evaluation of hepatitis B virus core-related antigen assay by LUMIPULSE f].

    Science.gov (United States)

    Tanaka, Yasuhito; Takagi, Kazumi; Hiramatsu, Kumiko; Naganuma, Hatsue; Iida, Takayasu; Takasaka, Yoshimitsu; Mizokami, Masashi

    2006-07-01

    A sensitive chemiluminescence enzyme immunoassay (CLEIA) has been developed for hepatitis B virus (HBV) core-related antigens (HBcrAg) detection. The HBcrAg is designated as the precore/core gene products including HBeAg. The aim of this study is to evaluate reproducibility of HBcrAg and correlation with HBV-DNA in serum using the automatic LUMIPULSE f to estimate an assay suitable for general laboratory use. In this study, we demonstrated that HBcrAg assay had highly intra-assay reproducible [coefficients of variation(CVs); 2.8-5.2%] and inter-assay reproducible [CVs; 3.9-9.1%]. When the cutoff value was tentatively set at 1 kU/ml, all healthy controls (HBsAg/HBV-DNA negative; n=100) and anti-HCV antibody-positive (n=50) sera were identified as negative. The assay showed a detection limit of 0.5 kU/ml using four serially diluted HBV high-titer sera, indicating higher sensitivity than HBV-DNA (transcription-mediated amplification). The HBcrAg concentration correlated positively with serum HBV-DNA (n=125, r = 0.860, p < 0.0001) regardless of HBeAg, although the HBcrAg levels were higher in HBeAg-positive group than in HBeAg-negative group. In the natural course of HBV infection, the HBcrAg concentration usually changed in accordance with HBV-DNA levels, however during lamivudine therapy the change of HBcrAg was more gradual than that of HBV-DNA. In conclusion, HBcrAg concentration provides a reflection of HBV virus load equivalent to HBV-DNA level, and the assay therefore offers a simple method for monitoring hepatitis B patients.

  7. Family incivility and job performance: a moderated mediation model of psychological distress and core self-evaluation.

    Science.gov (United States)

    Lim, Sandy; Tai, Kenneth

    2014-03-01

    This study extends the stress literature by exploring the relationship between family incivility and job performance. We examine whether psychological distress mediates the link between family incivility and job performance. We also investigate how core self-evaluation might moderate this mediated relationship. Data from a 2-wave study indicate that psychological distress mediates the relationship between family incivility and job performance. In addition, core self-evaluation moderates the relationship between family incivility and psychological distress but not the relationship between psychological distress and job performance. The results hold while controlling for general job stress, family-to-work conflict, and work-to-family conflict. The findings suggest that family incivility is linked to poor performance at work, and psychological distress and core self-evaluation are key mechanisms in the relationship.

  8. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aufiero, M.; Cammi, A.; Fiorina, C. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Leppänen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Ricotti, M.E. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy)

    2013-10-15

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  9. On the effect of different placing ZrH moderator material on the performance of a SFR core

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B. [Helmholtz-Zentrum Dresden-Rossendorf, Institut fuer Sicherheitsforschung, Postfach 51 01 19, 01314 Dresden (Germany); Weiss, F. P. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH Forschungszentrum, Boltzmannstr. 14, 85748 Garching (Germany)

    2012-07-01

    This study describes the development of a sodium fast reactor fuel assembly design with reduced void reactivity coefficient, achieved through the use of the ZrH moderating material. In the study the sodium void effect, as well as the major feedback coefficients are analyzed. Besides the feedback coefficients, the influence on the operational parameters like neutron flux distribution, power distribution, and burnup distribution is investigated for the different possibilities of arranging the moderating material in the fuel assembly. Additionally, the fuel cycle parameters - breeding and minor actinide production - are analyzed. For a first evaluation of the behavior during transients the influence of temperature changes in the ZrH is studied. (authors)

  10. In-service irradiated and aged material evaluations

    International Nuclear Information System (INIS)

    Haggag, F.M.; Nanstad, R.K.; Alexander, D.J.

    1995-01-01

    The objective of this task is to provide a direct assessment of actual material properties in irradiated components of nuclear reactors, including the effects of irradiation and aging. Four activities are currently in progress: (1) establishing a machining capability for contaminated or activated materials by completing procurement and installation of a computer-based milling machine in a hot cell; (2) machining and testing specimens from cladding materials removed from the Gundremmingen reactor to establish their fracture properties; (3) preparing an interpretive report on the effects of neutron irradiation on cladding; and (4) continuing the evaluation of long-term aging of austenitic structural stainless steel weld metal by metallurgically examining and testing specimens aged at 288 and 343 degrees C and reporting the results, as well as by continuing the aging of the stainless steel cladding toward a total time of 50,000 h

  11. Development and evaluation of high temperature materials for power plant

    International Nuclear Information System (INIS)

    Nickel, H.; Schubert, F.

    1992-01-01

    The development of high temperature materials requires the evaluation of the interaction of microstructure and mechanical properties, the implementation of the microstructural aspects in the constitutive equations for the analysis of loads in a high temperature component and verification of the materials reactions. In this way the full potential of materials properties can be better used. This fundamental method is the basis for the formulation of the structural design code KTA 3221 'Metallic HTR Components'. The method of 'design by analysis' is also activated for large internally cooled turbine blades for stationary gas turbines in combined cycle power plants. This kind of exploratory analysis during the dimensioning procedure are discussed with two examples: He/He-heat exchanger produced of NiCr23Co12Mo (Alloy 617) and turbine blades made of superalloys (e.g. IN 738 LC). (author)

  12. Methods for evaluation of mechanical stress condition of materials

    Directory of Open Access Journals (Sweden)

    Mirchev Yordan

    2018-01-01

    Full Text Available Primary attention is given to the following methods: method by drilling cylindrical holes (drill method and integrated ultrasonic method using volume (longitudinal and transverse, surface, and sub-surface waves. Drill method allows determination of residual mechanical stress in small depth of material surfaces, assessing type, size, and orientation of principal stresses. For the first time, parallel studies are carried out of mechanical stress in materials using the electroacoustic effect of volume, surface and sub-surface waves on the one hand, and effective mechanical stresses on the other. The experimental results present electroacoustic coefficients for different types of waves in the material of gas pipeline tube of 243 mm diameter and 14 mm thickness. These are used to evaluate mechanical stresses in pipelines, according to active GOST standards.

  13. Evaluation of radiation-shielding properties of the composite material

    International Nuclear Information System (INIS)

    Pavlenko, V.I.; Chekashina, N.I.; Yastrebinskij, R.N.; Sokolenko, I.V.; Noskov, A.V.

    2016-01-01

    The paper presents the evaluation of radiation-shielding properties of composite materials with respect to gamma-radiation. As a binder for the synthesis of radiation-shielding composites we used lead boronsilicate glass matrix. As filler we used nanotubular chrysotile filled with lead tungstate PbWO4. It is shown that all the developed composites have good physical-mechanical characteristics, such as compressive strength, thermal stability and can be used as structural materials. On the basis of theoretical calculation we described the graphs of the gamma-quanta linear attenuation coefficient depending on the emitted energy for all investigated composites. We founded high radiation-shielding properties of all the composites on the basis of theoretical and experimental data compared to materials conventionally used in the nuclear industry - iron, concrete, etc

  14. Evaluating the Aspect of Nuclear Material in Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Takagi, Shunsuke; Pickett, Susan; Oda, Takuji; Choi, Jor-Shan; Kuno, Yusuke; Takana, Satoru [Department of Nuclear Engineering and Management, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8685 (Japan); Nagasaki, Shinya [Nuclear Professional School, The University of Tokyo (Japan)

    2009-06-15

    The increasing number of countries that wish to introduce nuclear power plants raises attention to proliferation resistance in nuclear power plants, and nuclear fuel cycle facilities. In order to achieve adequate proliferation resistance, it is important to evaluate it and to construct effective international institutional frameworks as well as technologies involving high level of proliferation resistance. Although some methods have been proposed for evaluation of the proliferation resistance, their validities have not been investigated in detail. In the present paper, therefore, we compare some of the proposed methodologies. It is essential to detect the abuse or diversion of nuclear material before the nuclear explosive device can be manufactured in order to prevent proliferation. The time needed for the detection of material primary depends on the safeguards that the country applies, and the time needed for fabrication mainly depends on the attributes of the nuclear material. Hence, we divided the proliferation resistance into two parts: the level of safeguards and the material. For examination of evaluation methods such as the one proposed by Charlton [1] or the figure of merit (FOM) [2], sensitivity analysis was performed on weighting factors and scenarios. The validity and characteristics of each method were discussed, focusing on the applicability of each method to the assessment of multi-national approaches such as GNEP. [1] W. S. Charlton, R. L. LeBouf, C. Gariazzo, D. G. Ford, C. Beard, S. Landeberger, M. Whitaker, 'Proliferation resistance assessment methodology for nuclear fuel cycles', Nuclear Technology, 157, 1 (2007). [2] C.G. Bathke et al, 'An assessment of the proliferation resistance of materials in advanced nuclear fuel cycles', 8. International Conference on Facility Operations (2008). (authors)

  15. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

  16. Systematic Evaluation of Uncertainty in Material Flow Analysis

    DEFF Research Database (Denmark)

    Laner, David; Rechberger, Helmut; Astrup, Thomas Fruergaard

    2014-01-01

    Material flow analysis (MFA) is a tool to investigate material flows and stocks in defined systems as a basis for resource management or environmental pollution control. Because of the diverse nature of sources and the varying quality and availability of data, MFA results are inherently uncertain....... Uncertainty analyses have received increasing attention in recent MFA studies, but systematic approaches for selection of appropriate uncertainty tools are missing. This article reviews existing literature related to handling of uncertainty in MFA studies and evaluates current practice of uncertainty analysis......) and exploratory MFA (identification of critical parameters and system behavior). Whereas mathematically simpler concepts focusing on data uncertainty characterization are appropriate for descriptive MFAs, statistical approaches enabling more-rigorous evaluation of uncertainty and model sensitivity are needed...

  17. Environment - sustainable management of radioactive materials and radioactive - report evaluation

    International Nuclear Information System (INIS)

    2006-05-01

    The economic affairs commission evaluated the report of M. Henri Revol on the law project n 315 of the program relative to the sustainable management of the radioactive materials and wastes. It precises and discusses the choices concerning the researches of the three axis, separation and transmutation, deep underground disposal and retrieval conditioning and storage of wastes. The commission evaluated then the report on the law project n 286 relative to the transparency and the security in the nuclear domain. It precises and discusses this text objectives and the main contributions of the Senate discussion. (A.L.B.)

  18. Materials balance area Custodian Performance Evaluation Program at PNL

    International Nuclear Information System (INIS)

    Dickman, D.A.

    1991-07-01

    The material balance area (MBA) custodian has primary responsibility for control and accountability of nuclear material within an MBA. In this role, the custodian operates as an extension of the facility material control and accountability (MC ampersand A) organization. To effectively meet administrative requirements and protection needs, the custodian must be fully trained in all aspects of MC ampersand A related to the MBA, and custodian performance must be periodically evaluated. DOE Policy requires that each facility provide for a program which assures that each facility provide for a program which assures that personnel performing MC ampersand A functions are (1) trained and/or qualified to perform their duties and responsibilities and (2) knowledgeable of requirements and procedures related to their functions. The MBA Custodian Performance Evaluation Program at PNL uses a variety of assessment techniques to meet this goal, including internal and independent MBA audits, periodic custodian testing, conduct of limited scope performance tests, daily monitoring of MC ampersand A documentation, and reviewing custodian performance during physical inventories. The data collected from these sources is analyzed and incorporated into an annual custodian performance evaluation document, given to each custodian and line management. Development of this program has resulted in significantly improved custodian performance and a marked decrease in finding and observations identified during MBA audits

  19. Physical properties of core-concrete systems: Al{sub 2}O{sub 3}-ZrO{sub 2} molten materials measured by aerodynamic levitation

    Energy Technology Data Exchange (ETDEWEB)

    Ohishi, Yuji, E-mail: ohishi@see.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University (Japan); Kargl, F. [Institute of Materials Physics in Space, German Aerospace Center (Germany); Nakamori, F.; Muta, Hiroaki; Kurosaki, Ken [Graduate School of Engineering, Osaka University (Japan); Yamanaka, Shinsuke [Graduate School of Engineering, Osaka University (Japan); Research Institute of Nuclear Engineering, University of Fukui (Japan)

    2017-04-15

    During a molten core–concrete interaction, molten oxides consisting of molten core materials (UO{sub 2} and ZrO{sub 2}) and concrete (Al{sub 2}O{sub 3}, SiO{sub 2}, CaO) are formed. Reliable data on the physical properties of the molten oxides will allow us to accurately predict the progression of a nuclear reactor core meltdown accident. In this study, the viscosities and densities of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} (x = 0.356 and 0.172) were measured using an aerodynamic levitation technique. The densities of two small samples were estimated from their masses and their volumes (calculated from recorded images of the molten samples). The droplets were forced to oscillate using speakers, and their viscosities were evaluated from the damping behaviors of their oscillations. The results showed that the viscosity of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} compared to that of pure molten Al{sub 2}O{sub 3} is 25% lower for x = 0.172, while it is unexpectedly 20% higher for x = 0.356. - Highlights: •The physical properties of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} (x = 0.356 and 0.172) have been evaluated. •The measurement was conducted using an aerodynamic levitation technique. •The density and viscosity were measured.

  20. Evaluation of the use of digital images for a national prostate core external quality assurance scheme.

    Science.gov (United States)

    Harnden, Patricia; Coleman, Derek; Moss, Sue; Kodikara, Sandhya; Griffin, Nick R; Melia, Jane

    2011-10-01

    To evaluate the use of virtual images as an alternative to glass slides to expand the number of participants in the External Quality Assurance Scheme for prostatic biopsies. Benign and neoplastic cases, previously circulated as glass slides, were selected to include cases that had demonstrated a high level of agreement (n = 10) and a lesser degree of agreement (n = 10). Whole slide virtual images were circulated to 68 pathologists; 51 responses were returned. The levels of agreement for the primary diagnosis and for Gleason grading of cancers were analysed using kappa statistics. Responses for glass slides versus images were compared for the 24 pathologists for whom data were available. Levels of agreement for diagnostic categories using virtual slides were moderate to substantial, comparable to those found using glass slides. The level of agreement for Gleason grades 8-10 was substantial, but for lower grades was fair or moderate, poorer than for the glass slide circulation. Circulation of virtual images of biopsy material is a suitable alternative to glass slide-based schemes for the evaluation of diagnostic consistency. The majority of participants agreed that the ability to evaluate limited diagnostic material outweighed the disadvantages of a virtual system. 2011 Blackwell Publishing Limited.

  1. Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

    International Nuclear Information System (INIS)

    Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree

    2006-01-01

    The AGENT (Arbitrary Geometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic sub-meshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Re-balancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such

  2. Evaluation of color stability of different temporary restorative materials

    Directory of Open Access Journals (Sweden)

    José Vitor Quinelli Mazaro

    Full Text Available AbstractIntroductionTemporary restorative materials are widely used, however, little is know about their color stability.Objectiveto evaluate the color stability of the following temporary restorative materials: acrylic and bis-acrylic resins after immersion in pigmenting solutions for different periods of storage.Material and methodFour materials were tested (Dêncor/Clássico, Protemp 4/3M ESPE; Structur 2 SC/Voco; Luxatemp AM Plus/DMG and 30 test specimens (15 mm in diameter and 2 mm thick per material were fabricated. They were divided according to the storage medium (artificial saliva, saliva + cola type soda, and saliva + coffee and storage time intervals (2, 5, 7 and 15 days. Color measurements were made before and after immersions, with use of a spectrophotometer, by means of the CIE L*a*b* system. The data were analyzed by the analysis of variance and the Tukey Test, at a level of significance of 5%.ResultAcrylic resin presented greater color stability in comparison with bis-acrylic resins (p<0.001. When bis-acrylic resins were compared no significant difference was observed between the resins Structur and Luxatemp (p=0.767. As regards solutions tested, coffee showed the highest color change values (p<0.001, and the longer the storage time interval, the greater was the color change in all the temporary restorative materials analyzed (p<0.001.ConclusionAcrylic resin presented greater color stability in comparison with bis-acrylic resins (p<0.001. Coffee caused the greatest color change, and immersion time was determinant in color stability of the temporary materials analyzed.

  3. Evaluation of a mammographic stereotactic device for localization, fine-needle aspiration cytologic studies, and core biopsy of suspicious lesions

    International Nuclear Information System (INIS)

    Fajardo, L.J.; Davey, G.A.

    1988-01-01

    Mammography-guided interventional breast procedures, such as preoperative localization, fine-needle aspiration cytology (FNAC), and core biopsy of suspicious lesions, require accurate three-dimensional localization. The authors have evaluated a prototype stereotactic mammography device for localizing abnormalities with both phantom and clinical studies. Twenty-six localizations on a phantom were within 0.5 mm +- 0.93 (standard deviation) from the lesions; accuracy in clinical procedures was within 0.6 mm +- 0.8. Procedures are performed an average of 8 minutes faster with this device. They are prospectively evaluating mammography-guided FNAC and core biopsy of suspicious lesions in 100 patients. Results of FNAC performed without the stereotactic device agreed with results of open surgical biopsy in six of eight patients; results of core biopsy agreed in seven of eight

  4. Fuel element and full core thermal–hydraulic analysis of the AHTR for the evaluation of the LOFC transient

    International Nuclear Information System (INIS)

    Avigni, P.; Petrovic, B.

    2014-01-01

    Highlights: • We developed MATLAB and RELAP5 models of the single channel of the AHTR. • Single channel analysis indicates design envelope for effective heat removal. • The reactivity feedback evaluated by SCALE supports safe operation of the reactor. • We developed RELAP5 models of the fuel assembly and full core. • The fully passive DRACS protects the reactor during a LOFC accident. - Abstract: The Advanced High Temperature Reactor (AHTR) is a fluoride-cooled and graphite-moderated reactor concept designed by Oak Ridge National Laboratory (Holcomb et al., 2011). The modeling and optimization of the heat removal system and the core structure is required, in order to obtain an adequate heavy metal loading and to provide effective cooling capability. The single channel MATLAB model provides a simple tool to evaluate the steady state conditions for the coolant and the fuel plate and the effects of the power distribution; sensitivity studies on the main design parameters of the fuel element are performed. A RELAP5-3D single channel model is developed for the validation and comparison with the MATLAB model; this model is the starting point for the development of a full core model, enabling the study of transients. A one-third fuel assembly model is then analyzed, consisting of six fuel plates and modeling the heat conduction of graphite through RELAP5-3D conduction enclosures. Since the assembly model is not suitable for the implementation in a full core model with the same level of detail, several simplifications have been evaluated, involving the modeling of the plate through a single heat structure and the modeling of different plates through a single plate. A SCALE model of the fuel assembly was developed for the evaluation of the reactivity feedback and the power distribution in the core. The results from the neutronic evaluations and the assembly model were implemented in a full core model, involving the core, the main reactor structures, the cooling

  5. Advice networks in teams: the role of transformational leadership and members' core self-evaluations.

    Science.gov (United States)

    Zhang, Zhen; Peterson, Suzanne J

    2011-09-01

    This article examines the team-level factors promoting advice exchange networks in teams. Drawing upon theory and research on transformational leadership, team diversity, and social networks, we hypothesized that transformational leadership positively influences advice network density in teams and that advice network density serves as a mediating mechanism linking transformational leadership to team performance. We further hypothesized a 3-way interaction in which members' mean core self-evaluation (CSE) and diversity in CSE jointly moderate the transformational leadership-advice network density relationship, such that the relationship is positive and stronger for teams with low diversity in CSE and high mean CSE. In addition, we expected that advice network centralization attenuates the positive influence of network density on team performance. Results based on multisource data from 79 business unit management teams showed support for these hypotheses. The results highlight the pivotal role played by transformational leadership and team members' CSEs in enhancing team social networks and, ultimately, team effectiveness. PsycINFO Database Record (c) 2011 APA, all rights reserved

  6. The use of electromyography and magnetic resonance imaging to evaluate a core strengthening exercise programme.

    Science.gov (United States)

    Rutkowska-Kucharska, Alicja; Szpala, Agnieszka

    2018-01-01

    The question that was asked in the study was whether a training routine based on curl-up exercises with a load provided by body mass of the person increases local muscle strength or local muscle endurance. The aim of this study was to evaluate the effect of 4 weeks training based on a small load and low movement velocity on electrical activity (EMG), cross-sectional area (CSA) of core stabilisers. The EMG activity was measured in the rectus abdominis (RA), obliquus abdominis externus and erector spinae (ES) muscles. CSA of the muscles: RA, anterolateral abdominal, psoas major, quadratus lumborum, ES, and multifidus at the level of L3-L4 were measured too. The training increased the CSA and thickness in most of the muscles studied. Statistically significant correlation was found only for the ES circumference (left side) and EMG activity for the right side (r= 0.627, p= 0.022) and left side (r= 0.624, p= 0.023). The training programme resulted in a increase in the number of curl-up repetitions revealing an endurance increase in abdominal muscles. Furthermore, there was a increase in the EMG activity of the RA. An increase of the CSA of all tested muscles showed an increase of muscle active force.

  7. Evaluation of effective coolant flow rate in advanced design of the small scale VHTR core

    International Nuclear Information System (INIS)

    Fumizawa, Motoo; Suzuki, Kunihiko; Murakami, Tomoyuki.

    1988-02-01

    This report describes the evaluation of effective coolant flow rate in the advanced design of the small scale VHTR core. The analytical design study was carried out after the 2nd stage of detailed design in order to reduce the cost of construction. The summary of the analytical results are as follows: (1) Crossflow loss coefficient of flange type fuel block having 0.1 mm of sealing gap is about 100 times higher than that of dowel type block adopted in the 2nd stage of detailed design. (2) In case that coolant channel outer diameter is 52 mm and hydraulic diameter is 6 mm, the effective coolant flow rates using flange and dowel type fuel blocks are 80 % and 70 % respectively. Because the crossflow loss coefficients of dowel type are lower than that of flange type. (3) The effective coolant flow rate, when crossflow loss coefficients are distributed along with the axial direction, agrees well with that using mean value of crossflow loss coefficient i.e. 5 x 10 11 m -4 . (author)

  8. Fabrication of Fe{sub 3}O{sub 4}@CuO core-shell from MOF based materials and its antibacterial activity

    Energy Technology Data Exchange (ETDEWEB)

    Rajabi, S.K. [Department of Chemistry, University of Guilan, University Campus 2, Rasht (Iran, Islamic Republic of); Sohrabnezhad, Sh., E-mail: sohrabnezhad@guilan.ac.ir [Department of Chemistry, Faculty of Science, University of Guilan, P.O. Box 1914, Rasht (Iran, Islamic Republic of); Ghafourian, S. [Clinical Microbiology Research Center, Ilam University of Medical Sciences, Ilam (Iran, Islamic Republic of)

    2016-12-15

    Magnetic Fe{sub 3}O{sub 4}@CuO nanocomposite with a core/shell structure was successfully synthesized via direct calcinations of magnetic Fe{sub 3}O{sub 4}@HKUST-1 in air atmosphere. The morphology, structure, magnetic and porous properties of the as-synthesized nano composites were characterized by using scanning electron microscope (SEM), transmission electron microscopy (TEM), powder X-ray diffraction (PXRD), and vibration sample magnetometer (VSM). The results showed that the nanocomposite material included a Fe{sub 3}O{sub 4} core and a CuO shell. The Fe{sub 3}O{sub 4}@CuO core-shell can be separated easily from the medium by a small magnet. The antibacterial activity of Fe{sub 3}O{sub 4}-CuO core-shell was investigated against gram-positive and gram-negative bacteria. A new mechanism was proposed for inactivation of bacteria over the prepared sample. It was demonstrated that the core-shell exhibit recyclable antibacterial activity, acting as an ideal long-acting antibacterial agent. - Graphical abstract: Fe{sub 3}O{sub 4}@CuO core-shell release of copper ions. These Cu{sup 2+} ions were responsible for the exhibited antibacterial activity. - Highlights: • The Fe{sub 3}O{sub 4}@CuO core-shell was prepared by MOF method. • This is the first study of antibacterial activity of core-shell consist of CuO and Fe{sub 3}O{sub 4}. • The core-shell can be reused effectively. • Core-shell was separated from the reaction solution by external magnetic field.

  9. Evaluation report on CCTF CORE-I REFLOOD TEST Cl-1, (Run 010)

    International Nuclear Information System (INIS)

    Sudoh, Takashi; Murao, Yoshio.

    1983-09-01

    This report describes the effects of the loop flow resistance on the thermohydraulic behavior in the primary system during the reflood phase. The investigation is based on the results of the test Cl-1 which was performed with increased loop flow resistance in the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute. The loop flow resistance was about 40% higher in the present test than in the reference test Cl-5. The results of two tests were compared and the following conclusions were obtained: 1) The total loop flow rate and the core flooding rate were reduced by about 20% with the increased loop flow resistance 2) The core heat transfer was also lowered, then, the turnaround and the quench times extended at the locations above the core midplane. 3) The measured maximum temperature in the core was 50 K higher for the present test than for the reference test. (author)

  10. [Effect of core: dentin thickness ratio on the flexure strength of IPS Empress II heat-pressed all-ceramic restorative material].

    Science.gov (United States)

    Liu, Yi-hong; Feng, Hai-lan; Bao, Yi-wang; Qiu, Yan

    2007-02-18

    To evaluate the effect of core:dentin thickness ratio on the flexure strength, fracture mode and origin of bilayered IPS Empress II ceramic composite specimens. IPS Empress II core ceramic, dentin porcelain and bilayered composite specimens with core:dentin thickness ratio of 2:1 and 1:1 were tested in three-point flexure strength. Mean strengths and standard deviations were determined. The optical microscopy was employed for identification of the fracture mode and origin. The flexure strength of dentin porcelain was the smallest(62.7 MPa), and the strength of bilayered composite specimens was smaller than single-layered core ceramic(190.2 MPa). The core: dentin ratio did not influence the strength of bilayered composite specimens. The frequency of occurrence of bilayered specimen delaminations was higher in the group of core: dentin thickness ratio of 1:1 than in the group of 2:1. IPS Empress II core ceramic was significantly stronger than veneering dentin porcelain. Core:dentin thickness ratio could significantly influence the fracture mode and origin, and bilayered IPS Empress II ceramic composite specimens showed little influence in the fracture strength.

  11. Core discrete event simulation model for the evaluation of health care technologies in major depressive disorder.

    Science.gov (United States)

    Vataire, Anne-Lise; Aballéa, Samuel; Antonanzas, Fernando; Roijen, Leona Hakkaart-van; Lam, Raymond W; McCrone, Paul; Persson, Ulf; Toumi, Mondher

    2014-03-01

    A review of existing economic models in major depressive disorder (MDD) highlighted the need for models with longer time horizons that also account for heterogeneity in treatment pathways between patients. A core discrete event simulation model was developed to estimate health and cost outcomes associated with alternative treatment strategies. This model simulated short- and long-term clinical events (partial response, remission, relapse, recovery, and recurrence), adverse events, and treatment changes (titration, switch, addition, and discontinuation) over up to 5 years. Several treatment pathways were defined on the basis of fictitious antidepressants with three levels of efficacy, tolerability, and price (low, medium, and high) from first line to third line. The model was populated with input data from the literature for the UK setting. Model outputs include time in different health states, quality-adjusted life-years (QALYs), and costs from National Health Service and societal perspectives. The codes are open source. Predicted costs and QALYs from this model are within the range of results from previous economic evaluations. The largest cost components from the payer perspective were physician visits and hospitalizations. Key parameters driving the predicted costs and QALYs were utility values, effectiveness, and frequency of physician visits. Differences in QALYs and costs between two strategies with different effectiveness increased approximately twofold when the time horizon increased from 1 to 5 years. The discrete event simulation model can provide a more comprehensive evaluation of different therapeutic options in MDD, compared with existing Markov models, and can be used to compare a wide range of health care technologies in various groups of patients with MDD. Copyright © 2014 International Society for Pharmacoeconomics and Outcomes Research (ISPOR). Published by Elsevier Inc. All rights reserved.

  12. CT-guided core needle biopsy of pleural lesions: Evaluating diagnostic yield and associated complications

    International Nuclear Information System (INIS)

    Niu, Xiang Ke; Bhetuwal, Anup; Yang, Han Feng

    2015-01-01

    The purpose of this study was to retrospectively evaluate the diagnostic accuracy and complications of CT-guided core needle biopsy (CT-guided CNB) of pleural lesion and the possible effects of influencing factors. From September 2007 to June 2013, 88 consecutive patients (60 men and 28 women; mean [+/- standard deviation] age, 51.1 +/- 14.4 years; range, 19-78 years) underwent CT-guided CNB, which was performed by two experienced chest radiologists in our medical center. Out of 88 cases, 56 (63%) were diagnosed as malignant, 28 (31%) as benign and 4 (5%) as indeterminate for CNB of pleural lesions. The final diagnosis was confirmed by either histopathological diagnosis or clinical follow-up. The diagnostic accuracy, sensitivity, specificity, positive predictive value (PPV), negative predictive value (NPV), and complication rates were statistically evaluated. Influencing factors (patient age, sex, lesion size, pleural-puncture angle, patient position, pleural effusion, and number of pleural punctures) were assessed for their effect on accuracy of CT-guided CNB using univariate and subsequent multivariate analysis. Diagnostic accuracy, sensitivity, specificity, PPV, and NPV were 89.2%, 86.1%, 100%, 100%, and 67.8%, respectively. The influencing factors had no significant effect in altering diagnostic accuracy. As far as complications were concerned, occurrence of pneumothorax was observed in 14 (16%) out of 88 patients. Multivariate analysis revealed lesion size/pleural thickening as a significant risk factor (odds ratio [OR]: 8.744, p = 0.005) for occurrence of pneumothorax. Moreover, presence of pleural effusion was noted as a significant protective factor (OR: 0.171, p = 0.037) for pneumothorax. CT-guided CNB of pleural lesion is a safe procedure with high diagnostic yield and low risk of significant complications.

  13. CT-guided core needle biopsy of pleural lesions: Evaluating diagnostic yield and associated complications

    Energy Technology Data Exchange (ETDEWEB)

    Niu, Xiang Ke [Dept. of Radiology, Affiliated Hospital of Chengdu University, Chengdu (China); Bhetuwal, Anup; Yang, Han Feng [Dept. of Radiology, Sichuan Key Laboratory of Medical Imaging, Affiliated Hospital of North Sichuan Medical College, Nanchong (China)

    2015-02-15

    The purpose of this study was to retrospectively evaluate the diagnostic accuracy and complications of CT-guided core needle biopsy (CT-guided CNB) of pleural lesion and the possible effects of influencing factors. From September 2007 to June 2013, 88 consecutive patients (60 men and 28 women; mean [+/- standard deviation] age, 51.1 +/- 14.4 years; range, 19-78 years) underwent CT-guided CNB, which was performed by two experienced chest radiologists in our medical center. Out of 88 cases, 56 (63%) were diagnosed as malignant, 28 (31%) as benign and 4 (5%) as indeterminate for CNB of pleural lesions. The final diagnosis was confirmed by either histopathological diagnosis or clinical follow-up. The diagnostic accuracy, sensitivity, specificity, positive predictive value (PPV), negative predictive value (NPV), and complication rates were statistically evaluated. Influencing factors (patient age, sex, lesion size, pleural-puncture angle, patient position, pleural effusion, and number of pleural punctures) were assessed for their effect on accuracy of CT-guided CNB using univariate and subsequent multivariate analysis. Diagnostic accuracy, sensitivity, specificity, PPV, and NPV were 89.2%, 86.1%, 100%, 100%, and 67.8%, respectively. The influencing factors had no significant effect in altering diagnostic accuracy. As far as complications were concerned, occurrence of pneumothorax was observed in 14 (16%) out of 88 patients. Multivariate analysis revealed lesion size/pleural thickening as a significant risk factor (odds ratio [OR]: 8.744, p = 0.005) for occurrence of pneumothorax. Moreover, presence of pleural effusion was noted as a significant protective factor (OR: 0.171, p = 0.037) for pneumothorax. CT-guided CNB of pleural lesion is a safe procedure with high diagnostic yield and low risk of significant complications.

  14. Digital microscopic evaluation of vertical marginal discrepancies of CAD/CAM fabricated zirconia cores.

    Science.gov (United States)

    Habib, Syed Rashid

    2018-05-18

    The aim of this in vitro research study was to evaluate the vertical marginal discrepancies of zirconia (Zr) cores fabricated by five different computer-aided design and manufacturing (CAD/CAM) systems using a digital microscope. A total of 60 specimens were prepared and randomly divided into five groups (n=12 each) using the following systems: Ceramill Motion 2 (CM, Amanngirrbach, Germany); Weiland (WI, Ivoclar Vivadent, USA); Cerec (CS, Sirona Dental, USA); Zirkonzahn (ZZ, Gmbh Bruneck, Italy) and Cad4dent (CD, Canada). The specimens were numbered and the vertical marginal discrepancies were evaluated with a digital microscope at 50× magnification. A one-way analysis of variance showed a statistically significant difference (p=0.002) between the groups. The CM group exhibited the lowest values for the marginal gaps (31.30±15.12 μm), while the ZZ group exhibited the highest values for the marginal gaps (44.83±28.76 μm) compared to other groups. A post hoc Tukey's test for multiple comparisons between the experimental groups showed a statistically significant difference (p<0.05) between the group CM and group CD with group ZZ. The rest of the groups showed no significant differences between them. Variations in the values were observed for the four sites measured with the highest and the least mean marginal gap value of 43.19±23.84 μm and 32.49±12.21 μm for buccal and lingual sites, respectively. Variations existed in the marginal discrepancy values for the CAD/CAM systems investigated in the study. Vertical marginal discrepancy values observed for various systems investigated in the study were well within the clinically acceptable range.

  15. Material balance area custodian performance evaluation program at PNL

    International Nuclear Information System (INIS)

    Dickman, D.A.

    1991-01-01

    This paper reports that the material balance area (MBA) custodian has primary responsibility for control and accountability of nuclear material within an MBA. In this role, the custodian operates as an extension of the facility material control and accountability (MC and A) organization. To effectively meet administrative requirements and protection needs, the custodian must be fully trained in all aspects of MC and A related to the MBA, and custodian performance must be periodically evaluated. U.S. Department of Energy (DOE) Policy requires that each facility provide for a program which ensures that personnel performing MC and A functions are trained and/or qualified to perform their duties and responsibilities and knowledgeable of requirements and procedures related to their functions. the MBA Custodian Performance Evaluation Program at Pacific Northwest Laboratory (PNL) uses a variety of assessment techniques to meet this goal, including internal and independent MBA audits, periodic custodian testing, limited scope performance tests, daily monitoring of MC and A documentation, and reviewing custodian performance during physical inventories

  16. Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

    2015-08-15

    Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

  17. Investigating the Influences of Core Self-Evaluations, Job Autonomy, and Intrinsic Motivation on In-Role Job Performance

    Science.gov (United States)

    Joo, Baek-Kyoo; Jeung, Chang-Wook; Yoon, Hea Jun

    2010-01-01

    This study investigates the effects of core self-evaluations, job autonomy, and intrinsic motivation on employees' perceptions of their in-role job performance, based on a cross-sectional survey of 283 employees in a Fortune Global 100 company in Korea. The results suggest that employees perceived higher in-role job performance when they had…

  18. The Role of Core Self-Evaluations in the Relationship between Stress and Depression in Persons with Spinal Cord Injury

    Science.gov (United States)

    DeAngelis, Jesse B.; Yaghmaian, Rana; Smedema, Susan Miller

    2016-01-01

    Purpose: To investigate the role of core self-evaluations (CSE) in the relationship between perceived stress and depression in persons with spinal cord injury. Method: Two hundred forty-seven adults with spinal cord injury completed an online survey measuring perceived stress, CSE, and depressive symptoms. Results: A multiple regression analysis…

  19. Turnover Intentions of Employees With Informal Eldercare Responsibilities: The Role of Core Self-Evaluations and Supervisor Support.

    Science.gov (United States)

    Greaves, Claire E; Parker, Stacey L; Zacher, Hannes; Jimmieson, Nerina L

    2015-12-01

    As longevity increases, so does the need for care of older relatives by working family members. This research examined the interactive effect of core self-evaluations and supervisor support on turnover intentions in two samples of employees with informal caregiving responsibilities. Data were obtained from 57 employees from Australia (Study 1) and 66 employees from the United States and India (Study 2). Results of Study 1 revealed a resource compensation effect, that is, an inverse relationship between core self-evaluations and turnover intentions when supervisor care support was low. Results of Study 2 extended these findings by demonstrating resource boosting effects. Specifically, there was an inverse relationship between core self-evaluations and subsequent turnover intentions for those with high supervisor work and care support. In addition, employees' satisfaction and emotional exhaustion from their work mediated the inverse relationship between core self-evaluations and subsequent turnover intentions when supervisor work support and care support were high. Overall, these findings highlight the importance of employee- and supervisor-focused intervention strategies in organizations to support informal caregivers. © The Author(s) 2016.

  20. Negative Attitudes toward Older Workers and Hiring Decisions: Testing the Moderating Role of Decision Makers' Core Self-Evaluations.

    Science.gov (United States)

    Fasbender, Ulrike; Wang, Mo

    2016-01-01

    Organizational hiring practices have been charged for unfair treatment on the grounds of age. Drawing on theories of planned behavior and core self-evaluations, this research investigated the impact of negative attitudes toward older workers on hiring decisions and examined the moderating role of decision-makers' core self-evaluations. We tested our hypotheses based on a structured online questionnaire and a vignette study using a sample of 102 participants working in human resource management across different industries. As predicted, negative attitudes toward older workers were positively related to avoidance of hiring older people, which in turn was negatively related to the likelihood to select the oldest candidate. Because hiring decisions are not only about the hiring subject but also about the decision-maker, we tested the moderating role of decision-makers' core self-evaluations. Results showed that core self-evaluations buffered the relationship between negative attitudes toward older workers and avoidance of hiring older people. Theoretical implications of the findings with regard to hiring decisions about older people and practical recommendations to improve diversity management strategies and age-balanced hiring practices in organizations are discussed.

  1. Evaluation of final vapor pressures in the loss of flow accident in an irradiation device of a pool reactor core

    International Nuclear Information System (INIS)

    Verri, A.

    1987-01-01

    The reliability feature, are described for a device containing samples, at a temperatures of 300 grade centigrades, in a reactor core for a long time. After an examination of the maximum accident event, the maximum vapour pressure originated by the inlet of reactor cooling water into the experimental device, is evaluated

  2. Core-Shell Al-Polytetrafluoroethylene (PTFE) Configurations to Enhance Reaction Kinetics and Energy Performance for Nanoenergetic Materials.

    Science.gov (United States)

    Wang, Jun; Qiao, Zhiqiang; Yang, Yuntao; Shen, Jinpeng; Long, Zhang; Li, Zhaoqian; Cui, Xudong; Yang, Guangcheng

    2016-01-04

    The energy performance of solid energetic materials (Al, Mg, etc.) is typically restricted by a natural passivation layer and the diffusion-limited kinetics between the oxidizer and the metal. In this work, we use polytetrafluoroethylene (PTFE) as the fluorine carrier and the shielding layer to construct a new type of nano-Al based fuels. The PTFE shell not only prevents nano-Al layers from oxidation, but also assists in enhancing the reaction kinetics, greatly improving the stability and reactivity of fuels. An in situ chemical vapor deposition combined with the electrical explosion of wires (EEW) method is used to fabricate core-shell nanostructures. Studies show that by controlling the stoichiometric ratio of the precursors, the morphology of the PTFE shell and the energy performance can be easily tuned. The resultant composites exhibit superior energy output characters than that of their physically mixed Al/PTFE counterparts. This synthetic strategy might provide a general approach to prepare other high-energy fuels (Mg, Si). © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  3. PBDOWN - a computer code for simulating core material discharge and thermal to mechanical energy conversion in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Royl, P.

    1981-01-01

    PBDOWN is a computer code that simulates the blowdown of confined boiling materials ('pools') into a colder upper coolant plenum as time dependent ejection and expansion with consideration of a few selected exchange processes. Its application is restricted to situations resulting from hypothetical loss of flow (LOF) accidents in LMFBR's, where enough voiding has occured, that in core sodium vapor pressures become negligible. PBDOWN considers one working fluid for the discharge process (either fuel or steel) and a maximum of two working fluids (either fuel and sodium or steel and sodium) for the expansion process in the upper coolant plenum. Entrainment of sodium at the accelerated bubble liquid interfaces is mechanistically calculated by a Taylor instability entrainment model. Simulation of a hemispherical expansion form together with this mechanistic entrainment model gives a new integrated calculation of the time dependent sodium mass in the bubble. The paper summarizes the basic equations and assumptions of this computer model. Sample results compare different heat transfer and Na entrainment models during steel and fuel driven discharge processes. Mechanistic sodium entrainment simulation for SNR-type reactors coupled with a realistic heat transfer model is shown to reduce the integral mechanical work potential by a factor of 1.3 to 2.0 over the isentropic energy of the discharge working fluids. (orig.)

  4. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    Science.gov (United States)

    Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

    2011-03-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  5. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F. [CEA, DEN, Cadarache, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lopez, A. Legrand [CEA, DEN, Saclay, SIREN/LECSI, F-91400 Saclay (France)

    2011-03-15

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10{sup 20} n/cm{sup 2}. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  6. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    International Nuclear Information System (INIS)

    Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F.; Vermeeren, L.; Lopez, A. Legrand

    2011-01-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10 20 n/cm 2 . A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  7. Molecular evaluation of genetic variability of wheat elite breeding material

    Directory of Open Access Journals (Sweden)

    Brbaklić Ljiljana

    2009-01-01

    Full Text Available Estimation of genetic variability of breeding material is essential for yield improvement in wheat cultivars. Modern techniques based on molecular markers application are more efficient and precise in genetic variability evaluation then conventional methods. Variability of 96 wheat cultivars and lines was analyzed using four microsatellite markers (Gwm11, Gwm428, Psp3200, Psp3071. The markers were chosen according to their potential association with important agronomical traits indicated in the literature. Total of 31 alleles were detected with maximum number of alleles (11 in Xgwm11 locus. The highest polymorphism information content (PIC value (0,831 was found in the locus Xpsp3071. The genotypes were grouped into three subpopulations based on their similarity in the analyzed loci. The results have indicated wide genetic variability of the studied material and possibility of its application in further breeding process after validation of marker-trait association. .

  8. Using ABAQUS Scripting Interface for Materials Evaluation and Life Prediction

    Science.gov (United States)

    Powers, Lynn M.; Arnold, Steven M.; Baranski, Andrzej

    2006-01-01

    An ABAQUS script has been written to aid in the evaluation of the mechanical behavior of viscoplastic materials. The purposes of the script are to: handle complex load histories; control load/displacement with alternate stopping criteria; predict failure and life; and verify constitutive models. Material models from the ABAQUS library may be used or the UMAT routine may specify mechanical behavior. User subroutines implemented include: UMAT for the constitutive model; UEXTERNALDB for file manipulation; DISP for boundary conditions; and URDFIL for results processing. Examples presented include load, strain and displacement control tests on a single element model. The tests are creep with a life limiting strain criterion, strain control with a stress limiting cycle and a complex interrupted cyclic relaxation test. The techniques implemented in this paper enable complex load conditions to be solved efficiently with ABAQUS.

  9. Evaluation of backfill materials for a shallow-depth repository

    International Nuclear Information System (INIS)

    Buckley, L.P.; Arbique, G.M.; Tosello, N.B.; Woods, B.L.

    1986-11-01

    The focus of laboratory research effort on the disposal of low- and intermediate-level radioactive waste is to determine what conditions will dominate and which engineered barriers will be most effective for the retention of radionuclides. Initial studies have concentrated on the evaluation of a flooded repository and the assessment of backfill materials suitable for the adsorption of radioactivity, yet permeable enough to allow excess water to pass through the repository and into the underlying water table. Both physical and adsorption studies have been performed. Based on these preliminary experiments, it is felt that a mixture of 10 wt% clay and the remainder sand would satisfy the above criteria. Since both are available within the Ottawa Valley, they also have the added advantage of being more cost effective to use than imported materials

  10. Biomechanical evaluation of bending strength of spinal pedicle screws, including cylindrical, conical, dual core and double dual core designs using numerical simulations and mechanical tests.

    Science.gov (United States)

    Amaritsakul, Yongyut; Chao, Ching-Kong; Lin, Jinn

    2014-09-01

    Pedicle screws are used for treating several types of spinal injuries. Although several commercial versions are presently available, they are mostly either fully cylindrical or fully conical. In this study, the bending strengths of seven types of commercial pedicle screws and a newly designed double dual core screw were evaluated by finite element analyses and biomechanical tests. All the screws had an outer diameter of 7 mm, and the biomechanical test consisted of a cantilever bending test in which a vertical point load was applied using a level arm of 45 mm. The boundary and loading conditions of the biomechanical tests were applied to the model used for the finite element analyses. The results showed that only the conical screws with fixed outer diameter and the new double dual core screw could withstand 1,000,000 cycles of a 50-500 N cyclic load. The new screw, however, exhibited lower stiffness than the conical screw, indicating that it could afford patients more flexible movements. Moreover, the new screw produced a level of stability comparable to that of the conical screw, and it was also significantly stronger than the other screws. The finite element analysis further revealed that the point of maximum tensile stress in the screw model was comparable to the point at which fracture occurred during the fatigue test. Copyright © 2014 IPEM. Published by Elsevier Ltd. All rights reserved.

  11. Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Okubo, Tsutomu

    2009-01-01

    The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the channel flow response characteristics of single fuel assembly were evaluated. In the evaluation, the appropriate single-phase pressure drop setting at the inlet orifice was determined in terms of response stability from the design viewpoint. In addition, from the investigation on the relation of the response and transit time of coolant, it is confirmed that the channel flow response of RMWR is dominated by the transit time of vapor phase resulting from a high void fraction operation condition. As for a natural circulation flow response, it is clarified that the response is strongly influenced by the effect of two-phase pressure loss owing to a high void fraction condition. The reactor power response with reactivity feedback shows quite stable response characteristics on account of the small absolute value of void reactivity coefficient.

  12. Evaluating the MEDLINE Core Clinical Journals filter: data-driven evidence assessing clinical utility.

    Science.gov (United States)

    Klein-Fedyshin, Michele; Ketchum, Andrea M; Arnold, Robert M; Fedyshin, Peter J

    2014-12-01

    MEDLINE offers the Core Clinical Journals filter to limit to clinically useful journals. To determine its effectiveness for searching and patient-centric decision making, this study compared literature used for Morning Report in Internal Medicine with journals in the filter. An EndNote library with references answering 327 patient-related questions during Morning Report from 2007 to 2012 was exported to a file listing variables including designated Core Clinical Journal, Impact Factor, date used and medical subject. Bradford's law of scattering was applied ranking the journals and reflecting their clinical utility. Recall (sensitivity) and precision of the Core Morning Report journals and non-Core set was calculated. This study applied bibliometrics to compare the 628 articles used against these criteria to determine journals impacting decision making. Analysis shows 30% of clinically used articles are from the Core Clinical Journals filter and 16% of the journals represented are Core titles. When Bradford-ranked, 55% of the top 20 journals are Core. Articles sources used. Among the 63 Morning Report subjects, 55 have <50% precision and 41 have <50% recall including 37 subjects with 0% precision and 0% recall. Low usage of publications within the Core Clinical Journals filter indicates less relevance for hospital-based care. The divergence from high-impact medicine titles suggests clinically valuable journals differ from academically important titles. With few subjects demonstrating high recall or precision, the MEDLINE Core Clinical Journals filter may require a review and update to better align with current clinical needs. © 2014 John Wiley & Sons, Ltd.

  13. Evaluation of nuclear power plant component failure probability and core damage probability using simplified PSA model

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2000-01-01

    It is anticipated that the change of frequency of surveillance tests, preventive maintenance or parts replacement of safety related components may cause the change of component failure probability and result in the change of core damage probability. It is also anticipated that the change is different depending on the initiating event frequency or the component types. This study assessed the change of core damage probability using simplified PSA model capable of calculating core damage probability in a short time period, which is developed by the US NRC to process accident sequence precursors, when various component's failure probability is changed between 0 and 1, or Japanese or American initiating event frequency data are used. As a result of the analysis, (1) It was clarified that frequency of surveillance test, preventive maintenance or parts replacement of motor driven pumps (high pressure injection pumps, residual heat removal pumps, auxiliary feedwater pumps) should be carefully changed, since the core damage probability's change is large, when the base failure probability changes toward increasing direction. (2) Core damage probability change is insensitive to surveillance test frequency change, since the core damage probability change is small, when motor operated valves and turbine driven auxiliary feed water pump failure probability changes around one figure. (3) Core damage probability change is small, when Japanese failure probability data are applied to emergency diesel generator, even if failure probability changes one figure from the base value. On the other hand, when American failure probability data is applied, core damage probability increase is large, even if failure probability changes toward increasing direction. Therefore, when Japanese failure probability data is applied, core damage probability change is insensitive to surveillance tests frequency change etc. (author)

  14. Material properties of oxide dispersion strengthened (ODS) ferritic steels for core materials of FBR. Mechanical strength properties of sodium exposed and Nickel diffused materials. Interim report

    International Nuclear Information System (INIS)

    Kato, Shoichi; Yoshida, Eiichi

    2004-02-01

    An oxide dispersion strengthened (ODS) ferritic steel have excellent resistance to swelling and superior creep strength, they are expected to be used as a long-life cladding material in future advanced fast reactor. In this study, sodium environmental effects on the ODS steel developed by JNC were clarified through tensile test after sodium exposure for maximum 10,000hrs and creep-rupture test in sodium at elevated temperature. The exposure to sodium was conducted using a sodium test loop constituted by austenitic steels. For the conditions of sodium exposure test, the sodium temperatures were 923 K and 973 K, the oxygen concentration in sodium was below 2ppm and sodium flow rate on the surface of specimen was less than 1x10 -4 m/s. Further the specimen with the nickel diffused was prepared, which is simulate to nickel diffusing through sodium from the surface of structural stainless steels. The main results obtained were as follows; (1) The results showed excellent sodium-resistance up to a high temperature of about 973 K in stagnant sodium conditions, and its considered that the effects of sodium environment of tensile properties were negligible. In case of stagnant sodium condition, creep-rupture strength in sodium was equal to the in argon gas, and no sodium environmental effect was observed. The same is true for the creep-rupture ductility. (2) The tensile properties of nickel diffused test specimens at high temperatures simulating microstructure change were equal to that of the thermal aging process specimens. These tensile tests suggest that sodium environmental effects can be ignored. However, the effect of nickel diffusion on creep strength are not clear at present and experimental investigation are being conducted. (3) The coefficient of nickel diffusion in the ODS steel can be estimated based on the results of nickel concentration measurement. This value is larger than that of the diffusion coefficient for typical α-Fe steel at temperature below 973 K

  15. Evaluation of dredged material proposed for ocean disposal from Westchester Creek project area, New York

    Energy Technology Data Exchange (ETDEWEB)

    Pinza, M.R.; Gardiner, W.W.; Barrows, E.S.; Borde, A.B.

    1996-11-01

    The objective of the Westchester Creek project was to evaluate proposed dredged material from this area to determine its suitability for unconfined ocean disposal at the Mud Dump Site. Westchester Creek was one of five waterways that the US Army Corps of Engineers- New York District (USACE-NYD) requested the Battelle/Marine Sciences Laboratory (MSL) to sample and evaluate for dredging and disposal in May 1995. The evaluation of proposed dredged material from the Westchester Creek project area consisted of bulk sediment chemical analyses, chemical analyses of dredging site water and elutriate, benthic acute and water-column toxicity tests, and bioaccumulation studies. Thirteen individual sediment core samples were collected from this area and analyzed for grain size, moisture content, and total organic carbon (TOC). One composite sediment sample representing the Westchester Creek area to be dredged, was analyzed for bulk density, specific gravity, metals, chlorinated pesticides, polychlorinated biphenyl (PCB) congeners, polynuclear aromatic hydrocarbons (PAHs), and 1,4-dichlorobenzene. Dredging site water and elutriate water, which is prepared from the suspended- particulate phase (SPP) of the Westchester Creek sediment composite, was analyzed for metals, pesticides, and PCBS.

  16. Preliminary evaluation of SACI-O code for the analysis of transients in a pressurized water reactor core

    International Nuclear Information System (INIS)

    Soares, P.A.; Sirimarco, L.F.; Veloso, M.A.F.

    1979-03-01

    SACI-O is a computer code which deals with the dynamics of the core of pressurized light water reactors (PWR). Its applicability is determined by the evaluation of the models used in the simulation of the several phenomena and processes which occur in the core during transients. This report presents a comparison between the results obtained with SACI-O and those presented in the Final Safety Analysis Report (FSAR) of Angra dos Reis Nuclear Station, Unit 1. Although some data used in the calculations done by Westinghouse are not known, there was a good agreement between the mentioned results. (Author) [pt

  17. Corrosive wear. Evaluation of wear and corrosive resistant materials; Noetningskorrosion. Utvaerdering av noetnings- och korrosionsbestaendiga material

    Energy Technology Data Exchange (ETDEWEB)

    Persson, H.; Hjertsen, D.; Waara, P.; Prakash, B.; Hardell, J.

    2007-12-15

    With a new purchase of a waste conveyer screw at hand, for the 'A-warehouse' at the combined power and heating plant at E.ON Norrkoeping, the request for improved construction materials was raised. The previous screw required maintenance with very short intervals due to the difficult operation conditions. With the new screw the expectation is to manage 6 months of operation without interruption. The environment for the screw has two main components that sets the demand on the materials, on one hand the corrosive products that comes along and which forms at digestion of the waste and on the other hand the abrasive content in the waste. The term of the mechanism is wear-corrosion and can give considerably higher material loss than the two mechanisms wear and corrosion separately. Combination of a strong corrosive environment together with extensive wear is something that we today have limited knowledge about. The overall objective of the project has been to establish better wear and corrosive resistant construction materials for a waste conveyer screw that will lead to reduced operational disturbance costs. The evaluation has been performed in both controlled laboratory environments and in field tests, which has given us a better understanding of what materials are more suitable in this tough environment and has given us a tool for future predictions of the wear rate of the different material. The new conveyer screw, installed in February 2007 and with which the field test have been performed, has considerably reduced the wear of the construction and the target of 6 month maintenance-free operation is met with this screw for all the evaluated materials. The wear along the screw varies very much and with a clear trend for all the materials to increase towards the feeding direction of the screw. As an example, the wear plate SS2377 (stainless duplex steel) has a useful life at the most affected areas that is calculated to be 1077 days of operation with the

  18. In-situ synthetize multi-walled carbon nanotubes@MnO2 nanoflake core-shell structured materials for supercapacitors

    Science.gov (United States)

    Zheng, Huajun; Wang, Jiaoxia; Jia, Yi; Ma, Chun'an

    2012-10-01

    A new type of core-shell structured material consisting of multi-walled carbon nanotubes (MWCNTs) and manganese dioxide (MnO2) nanoflake is synthesized using an in-situ co-precipitation method. By scanning electron microscopy and transition electron microscope, it is confirmed that the core-shell nanostructure is formed by the uniform incorporation of birnessite-type MnO2 nanoflake growth round the surface of the activated-MWCNTs. That core-shell structured material electrode presents excellent electrochemical capacitance properties with the specific capacitance reaching 380 F g-1 at the current density of 5 A g-1 in 0.5 M Na2SO4 electrolyte. In addition, the electrode also exhibits good performance (the power density: 11.28 kW kg-1 at 5 A g-1) and long-term cycling stability (retaining 82.7% of its initial capacitance after 3500 cycles at 5 A g-1). It mainly attributes to MWCNTs not only providing considerable specific surface area for high mass loading of MnO2 nanoflakes to ensure effective utilization of MnO2 nanoflake, but also offering an electron pathway to improve electrical conductivity of the electrode materials. It is clearly indicated that such core-shell structured materials including MWCNTs and MnO2 nanoflake may find important applications for supercapacitors.

  19. Influence of material non-linearity on the thermo-mechanical response of polymer foam cored sandwich structures - FE modelling and preliminary experiemntal results

    DEFF Research Database (Denmark)

    Palleti, Hara Naga Krishna Teja; Thomsen, Ole Thybo; Fruehmann, Richard.K

    In this paper, the polymer foam cored sandwich structures with fibre reinforced composite face sheets will be analyzed using the commercial FE code ABAQUS/Standard® incorporating the material and geometrical non-linearity. Large deformations are allowed which attributes geometric non linearity...

  20. Evaluation of neutron shielding made of cement type material

    International Nuclear Information System (INIS)

    Seshimo, Takuya; Nagai, Takayuki; Onose, Atsushi; Takuma, Yasuhisa; Tanuma, Hiroyuki; Otagawa, Masaaki

    1998-01-01

    We prepared boron-containing cement and evaluated the characteristics of this new cement. This is the material of neutron shielding which is lighter than existing one. The quality we aimed is: H ≥ 0.025 g/cm 3 , B ≥ 0.065 g/cm 3 , density ≤ 1.70 g/cm 3 . We made test pieces changing water powder ratio (W/P), adding amount of air entraining agent, adding amount of water reducing agent, and time of vibration, and then, evaluated the characteristics. The measured parameters are the air content, mortar flow and homogeneity for cement mortar, homogeneity and compressive strength for hardened one. From the results of these tests, we confirmed the possibility of making neutron shielding that can satisfy the aimed quality using this boron-containing cement. After all, we established the method of making the neutron shielding, and this method was used in the construction of RETF. (author)

  1. Process evaluations for uranium recovery from scrap material

    International Nuclear Information System (INIS)

    Westphal, B.R.; Benedict, R.W.

    1992-01-01

    The integral Fast Reactor (IFR) concept being developed by Argonne National Laboratory is based on pyrometallurgical processing of spent nuclear metallic fuel with subsequent fabrication into new reactor fuel by an injection casting sequence. During fabrication, a dilute scrap stream containing uranium alloy fines and broken quartz (Vycor) molds in produced. Waste characterization of this stream, developed by using present operating data and chemical analysis was used to evaluate different uranium recovery methods and possible process variations for the return of the recovered metal. Two methods, comminution with size separation and electrostatic separation, have been tested and can recover over 95% of the metal. Recycling the metal to either the electrochemical process or the injection casting was evaluated for the different economic and process impacts. The physical waste parameters and the important separation process variables are discussed with their effects on the viability of recycling the material. In this paper criteria used to establish the acceptable operating limits is discussed

  2. The permafrost carbon inventory on the Tibetan Plateau: a new evaluation using deep sediment cores.

    Science.gov (United States)

    Ding, Jinzhi; Li, Fei; Yang, Guibiao; Chen, Leiyi; Zhang, Beibei; Liu, Li; Fang, Kai; Qin, Shuqi; Chen, Yongliang; Peng, Yunfeng; Ji, Chengjun; He, Honglin; Smith, Pete; Yang, Yuanhe

    2016-08-01

    The permafrost organic carbon (OC) stock is of global significance because of its large pool size and the potential positive feedback to climate warming. However, due to the lack of systematic field observations and appropriate upscaling methodologies, substantial uncertainties exist in the permafrost OC budget, which limits our understanding of the fate of frozen carbon in a warming world. In particular, the lack of comprehensive estimates of OC stocks across alpine permafrost means that current knowledge on this issue remains incomplete. Here, we evaluated the pool size and spatial variations of permafrost OC stock to 3 m depth on the Tibetan Plateau by combining systematic measurements from a substantial number of pedons (i.e. 342 three-metre-deep cores and 177 50-cm-deep pits) with a machine learning technique (i.e. support vector machine, SVM). We also quantified uncertainties in permafrost carbon budget by conducting Monte Carlo simulations. Our results revealed that the combination of systematic measurements with the SVM model allowed spatially explicit estimates to be made. The OC density (OC amount per unit area, OCD) exhibited a decreasing trend from the south-eastern to the north-western plateau, with the exception that OCD in the swamp meadow was substantially higher than that in surrounding regions. Our results also demonstrated that Tibetan permafrost stored a large amount of OC in the top 3 m, with the median OC pool size being 15.31 Pg C (interquartile range: 13.03-17.77 Pg C). 44% of OC occurred in deep layers (i.e. 100-300 cm), close to the proportion observed across the northern circumpolar permafrost region. The large carbon pool size together with significant permafrost thawing suggests a risk of carbon emissions and positive climate feedback across the Tibetan alpine permafrost region. © 2016 John Wiley & Sons Ltd.

  3. SpaceCubeX: A Framework for Evaluating Hybrid Multi-Core CPU FPGA DSP Architectures

    Science.gov (United States)

    Schmidt, Andrew G.; Weisz, Gabriel; French, Matthew; Flatley, Thomas; Villalpando, Carlos Y.

    2017-01-01

    The SpaceCubeX project is motivated by the need for high performance, modular, and scalable on-board processing to help scientists answer critical 21st century questions about global climate change, air quality, ocean health, and ecosystem dynamics, while adding new capabilities such as low-latency data products for extreme event warnings. These goals translate into on-board processing throughput requirements that are on the order of 100-1,000 more than those of previous Earth Science missions for standard processing, compression, storage, and downlink operations. To study possible future architectures to achieve these performance requirements, the SpaceCubeX project provides an evolvable testbed and framework that enables a focused design space exploration of candidate hybrid CPU/FPGA/DSP processing architectures. The framework includes ArchGen, an architecture generator tool populated with candidate architecture components, performance models, and IP cores, that allows an end user to specify the type, number, and connectivity of a hybrid architecture. The framework requires minimal extensions to integrate new processors, such as the anticipated High Performance Spaceflight Computer (HPSC), reducing time to initiate benchmarking by months. To evaluate the framework, we leverage a wide suite of high performance embedded computing benchmarks and Earth science scenarios to ensure robust architecture characterization. We report on our projects Year 1 efforts and demonstrate the capabilities across four simulation testbed models, a baseline SpaceCube 2.0 system, a dual ARM A9 processor system, a hybrid quad ARM A53 and FPGA system, and a hybrid quad ARM A53 and DSP system.

  4. Core self-evaluations and work engagement: Testing a perception, action, and development path.

    Directory of Open Access Journals (Sweden)

    Maria Tims

    Full Text Available Core self-evaluations (CSE have predictive value for important work outcomes such as job satisfaction and job performance. However, little is known about the mechanisms that may explain these relationships. The purpose of the present study is to contribute to CSE theory by proposing and subsequently providing a first test of theoretically relevant mediating paths through which CSE may be related to work engagement. Based on approach/avoidance motivation and Job Demands-Resources theory, we examined a perception (via job characteristics, action (via job crafting, and development path (via career competencies. Two independent samples were obtained from employees working in Germany and The Netherlands (N = 303 and N = 404, respectively. When taking all mediators into account, results showed that the perception path represented by autonomy and social support played a minor role in the relationship between CSE and work engagement. Specifically, autonomy did not function as a mediator in both samples while social support played a marginally significant role in the CSE-work engagement relationship in sample 1 and received full support in sample 2. The action path exemplified by job crafting mediated the relationship between CSE and work engagement in both samples. Finally, the development path operationalized with career competencies mediated the relationship between CSE and work engagement in sample 1. The study presents evidence for an action and development path over and above the often tested perception path to explain how CSE is related to work engagement. This is one of the first studies to propose and show that CSE not only influences perceptions but also triggers employee actions and developmental strategies that relate to work engagement.

  5. Core self-evaluations and work engagement: Testing a perception, action, and development path

    Science.gov (United States)

    Akkermans, Jos

    2017-01-01

    Core self-evaluations (CSE) have predictive value for important work outcomes such as job satisfaction and job performance. However, little is known about the mechanisms that may explain these relationships. The purpose of the present study is to contribute to CSE theory by proposing and subsequently providing a first test of theoretically relevant mediating paths through which CSE may be related to work engagement. Based on approach/avoidance motivation and Job Demands-Resources theory, we examined a perception (via job characteristics), action (via job crafting), and development path (via career competencies). Two independent samples were obtained from employees working in Germany and The Netherlands (N = 303 and N = 404, respectively). When taking all mediators into account, results showed that the perception path represented by autonomy and social support played a minor role in the relationship between CSE and work engagement. Specifically, autonomy did not function as a mediator in both samples while social support played a marginally significant role in the CSE–work engagement relationship in sample 1 and received full support in sample 2. The action path exemplified by job crafting mediated the relationship between CSE and work engagement in both samples. Finally, the development path operationalized with career competencies mediated the relationship between CSE and work engagement in sample 1. The study presents evidence for an action and development path over and above the often tested perception path to explain how CSE is related to work engagement. This is one of the first studies to propose and show that CSE not only influences perceptions but also triggers employee actions and developmental strategies that relate to work engagement. PMID:28787464

  6. Core self-evaluations and work engagement: Testing a perception, action, and development path.

    Science.gov (United States)

    Tims, Maria; Akkermans, Jos

    2017-01-01

    Core self-evaluations (CSE) have predictive value for important work outcomes such as job satisfaction and job performance. However, little is known about the mechanisms that may explain these relationships. The purpose of the present study is to contribute to CSE theory by proposing and subsequently providing a first test of theoretically relevant mediating paths through which CSE may be related to work engagement. Based on approach/avoidance motivation and Job Demands-Resources theory, we examined a perception (via job characteristics), action (via job crafting), and development path (via career competencies). Two independent samples were obtained from employees working in Germany and The Netherlands (N = 303 and N = 404, respectively). When taking all mediators into account, results showed that the perception path represented by autonomy and social support played a minor role in the relationship between CSE and work engagement. Specifically, autonomy did not function as a mediator in both samples while social support played a marginally significant role in the CSE-work engagement relationship in sample 1 and received full support in sample 2. The action path exemplified by job crafting mediated the relationship between CSE and work engagement in both samples. Finally, the development path operationalized with career competencies mediated the relationship between CSE and work engagement in sample 1. The study presents evidence for an action and development path over and above the often tested perception path to explain how CSE is related to work engagement. This is one of the first studies to propose and show that CSE not only influences perceptions but also triggers employee actions and developmental strategies that relate to work engagement.

  7. Evaluation of issues around road materials for sustainable transport

    CSIR Research Space (South Africa)

    Steyn, WJVDM

    2009-07-01

    Full Text Available In addition to a number of other factors (social, economic, etc) sustainable transport requires the sustainable supply and use of construction materials. This includes the use of marginal materials, waste materials, novel / innovative materials...

  8. Evaluation of the pressure difference across the core during PWR-LOCA reflood phase

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Murao, Yoshio

    1979-03-01

    The flooding rate of the core influences largely cooling of the core during the reflood phase of a PWR-LOCA. Since the void fraction of two-phase flow in the core is important determining the flooding rate, it is essential to examine this void fraction. The void fraction in the core during the reflood phase obtained by experiment was compared with those predicted by the correlations respectively of Akagawa, Nicklin, Zuber, Yeh, Griffice, Behringer and Jhonson. Only Yeh's correlation was found to be usable for the purpose. The pressure difference of the core during the reflood phase was calculated by reflood analyzing code REFLA-1D using Yeh's correlation. Following are the results: (1) During the steady-state period after quenching of the heaters, the prediction agrees within +-15% with the experiment. (2) During the transient period when the quench front is advancing, the prediction is not in agreement with the experiment, the difference being about +-40%. Influence of the advancing quench front upon the void fraction in the core must further be studied. (author)

  9. Preparation and evaluation of reference materials for accountancy analysis. (2) Evaluation results

    International Nuclear Information System (INIS)

    Sumi, Mika; Abe, Katsuo; Kageyama, Tomio; Nakazawa, Hiroaki; Takamatsu, Mai; Kacchi, Tomokazu; Murakami, Toshiki; Ai, Hironobu

    2009-01-01

    Destructive analysis for accountancy at nuclear fuel facilities should attain international target values for measurement uncertainties in safeguarding nuclear materials (ITVs). Since measurement uncertainties of isotope dilution mass spectrometry depend on uncertainties of spikes (standard materials) used, utilizing highly reliable standard material is essential. The LSD spikes prepared under collaboration work with JAEA and JNFL has different Pu/U ratio and smaller nuclear material in a spike compared with the LSD spikes used a safeguard laboratories, and the value of Pu which separated and purified from MOX and used as raw material for one of the LSD spike prepared at JAEA were measured at JAEA. Uncertainties of the prepared LSD spikes and the measurement results of actual samples with these LSD spikes were evaluated based on ISO-GUM and compared with ITVs. (author)

  10. Evaluation of the gravity-injection capability for core cooling after a loss-of-SDC event

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Bang, Young Seok; Kim, Hho Jung

    1999-01-01

    In order to evaluate the gravity-drain capability to maintain core cooling after a loss-of-shutdown-cooling event during shutdown operation, the plant conditions of the Young Gwang Units 3 and 4 were reviewed. The six cases of possible gravity-drain paths using the water of the refueling water storage tank (RWST) were identified and the thermal hydraulic analyses were performed using RELAP5/MOD3.2 code. The core cooling capability was dependent on the gravity-drain paths and the drain rate. In the cases with the injection path and opening on the different leg side, the system was well depressurized after gravity-injection and the core boiling was successfully prevented for a long-term transient. However, in the cases with the injection path and opening on the cold leg side, the core coolant continued boiling although the system pressure remains atmospheric after gravity-injection because the cold water injected from the RWST was bypassed the core region. In the cases with the higher pressurizer opening than the RWST water level, the system was also pressurized by the water-hold in the pressurizer and the core was uncovered because the gravity-injection from the RWST stopped due to the high system pressure. In addition, from the sensitivity study on the gravity-injection flow rates, it was found that about 54 kg/s of RWST drain rate was required to maintain the core cooling. Those analysis results would provide useful information to operators coping with the event

  11. Evaluation of WWER-1000 vessel materials fracture toughness

    International Nuclear Information System (INIS)

    Grinik, Eh.U.; Revka, V.N.; Chirko, L.I.; Chajkovskij, Yu.V.

    2007-01-01

    The lifetime of WWER-1000-type reactor vessels is finally conditioned by the fracture toughness (crack growth resistance) of RPV materials. Up to now in line with the regulations the fracture toughness is characterized by the critical temperature of brittleness determined by the results of the Charpy specimen impact testing. Such approach is typical for all countries operating the water pressure reactors. However, regulatory approach is known from the western specialists not always to characterize adequately the crack growth resistance of the vessel materials and in some cases to underestimate their characteristics in the reference state that leads to unreasonably high conservatism. Excessive conservatism may lead to the invalid restrictions in the operating modes and the service life of the reactor vessel. Therefore there appeared the necessity to apply another approaches based on the state-of-the-art experimental methods of the fracture mechanics and allowing evaluating the fracture toughness parameters sufficiently. The paper presents the results of the comparison of the regulatory approach and the Master curve approach from the point of view of the adequate determination of the vessel material crack growth resistance parameters. Analysis of the experimental data of the surveillance specimens illustrated the potential possibility of applying the new statistical method for the WWER-1000- type reactor vessel lifetime extension

  12. Small punch test evaluation methods for material characterisation

    Energy Technology Data Exchange (ETDEWEB)

    Janča, Adam, E-mail: adam.janca@fjfi.cvut.cz; Siegl, Jan, E-mail: jan.siegl@fjfi.cvut.cz; Haušild, Petr, E-mail: petr.hausild@fjfi.cvut.cz

    2016-12-01

    The Small Punch Test (SPT) is one of the most widespread mechanical testing methods using miniaturized specimens. The paper presented deals with the time independent SPT, in which a flat specimen is bent by means of a (hemi)spherical punch moving at a constant velocity. The main goal is to relate the measured data to deformation processes taking place during specimen loading. Understanding of such relations is crucial for characterizing a material using any non-standardized experimental procedure. Using enhanced instrumentation, not only traditional load-displacement or load-deflection curves could be obtained, but also specimen thinning could be continuously measured and evaluated. Five alloys having a broad range of mechanical properties were tested. The results obtained were evaluated using both traditional and newly proposed methods and they were correlated with results of the conventional tensile test. The methods proposed seem to lead to a universal correlation between SPT results and tensile characteristics. - Highlights: • The newly proposed methodology significantly improved results of SPT. • Plastic deformation starts inside the specimen from the very beginning of loading. • Specimen thinning = punch displacement−specimen deflection. • Material response to loading is well illustrated by the novel load-thinning curve.

  13. Method for evaluating leaching from LSA-III material

    International Nuclear Information System (INIS)

    Abe, H.; Satoh, K.; Ozaki, S.; Watabe, N.; Iida, T.; Akamatsu, H.

    1989-01-01

    The IAEA transport regulations are scheduled to be introduced in Japan. New regulations are supposed to be set forth for low specific activity (LSA) material and industrial packaging (IP) as solidified concentrated waste water should correspond to the LSA material. Solidified concentrated waste water should be transported in accordance with the new transport regulations which reflect the IAEA transport regulations. As one of the regulations for LSA material, the leaching test for LSA-III materials states that the radioactive loss due to leaching without the packaging should not exceed 0.1 A 2 when left in the water for 7 days. This test method is called Transport regulations hereafter. Since the test had not been conducted in Japan before now, there was no available data. Consequently, it is necessary to make an assessment on whether the current solidified concentrated waste water can satisfy the leaching amount of radioactive nuclide specified in the IAEA transport regulations. If the test is performed in accordance with the IAEA transport regulations, however, it is necessary to measure the amount of radioactive nuclide actually leached from the solidified concentrated waste water. Since the solidified concentrated waste water is put in a drum cam, it is necessary to prepare large-scale hot test equipment. In this study, therefore, the leaching test was conducted on the solidified concentrated waste water to propose the means of a leaching assessment which can be conducted with ordinary equipment to evaluate the leaching for assessment of the adaptability to IAEA transport regulations. In addition, the leaching test was performed in accordance with the IAEA method to examine the co-relation between the transport regulations and the IAEA method. Many test results have been reported for the IAEA method in Japan, which will be detailed later on

  14. Evaluating the internalisation of core values at a South African public service organisation

    Directory of Open Access Journals (Sweden)

    Susanna M. O’Neil

    2012-09-01

    Research purpose: This article presents an effort to describe a value internalisation effort within a South African public service organisation as well as the results of a subsequent evaluation to ascertain to what extent those efforts actually led to internalisation throughout the organisation. A set of actions and practices were implemented within the public service organisation; the intent was that they should enhance value internalisation in the organisation. A long-term strategy of value internalisation was followed that focussed mainly on the clear articulation and communication of the values through different communication mediums and platforms, such as road shows and branded value material hand-outs, as well as through extensive value internalisation training. Motivation for the study: Documentation of value internalisation processes and its evaluation, especially in South African public service organisations is extremely rare. To ensure that public service organisations do not repeat the same mistakes in their value internalisation practices and implementation processes, proper documentation of these processes in the public and research domains are needed. The need for the evaluation of value internalisation programmes should also be propagated as in many instances, programmes are implemented, but the subsequent success thereof is never evaluated. Research design, approach and method: A survey questionnaire consisting of a 5-point rating scale was developed to measure the extent of value internalisation after the implementation of long-term internalisation strategies. Employees at different levels and in different units of the organisation participated in the survey. Main findings: Results (N = 941 reflected lower than expected mean scores for each value component. In addition, differences in internalisation extent were found between two demographic variables, namely population groupings and organisational units. Practical/managerial implications: The

  15. Corrosion evaluation of materials in sulfur compound environments

    International Nuclear Information System (INIS)

    Maoying Teng; Iuanjou Yang

    1993-01-01

    The para-toluene sulfonic acid (PTSA) serves as a catalyst in producing diethylene glycol dibenzoate (DEGDB) and decomposes with increasing time at elevated temperature. Due to the presence of bisulfite ion, it is important to evaluate the corrosion properties of materials in this metastable environments. A potentiodynamic method was used to screen materials' properties in a PTSA solution. A surface analysis technique was also performed to investigate the oxide films. The critical current density and passive current density were substantially reduced when Fe alloyed with Cr and/or Ni. With the addition of Mo in Fe-Ni-Cr alloys, the critical current density was lowered further to show the beneficial effect of alloyed Mo. A plot of the corrosion rate of materials in DEGDB as a function of Ni/Cr ratio shows the linearity with increasing Ni/Cr ratio, disregard the type of materials. The corrosion rate of pure chromium can be estimated as ∼ 2.0 mpy by extrapolation of the linearity to Ni/Cr = 0. This is also the minimum corrosion rate that even Fe-Ni-Cr alloys were alloyed with Mo. Surface analysis results showed that the dissolution of Fe and/or Ni leads to a higher surface chromium content and results in the formation of chromium oxide on metal surface. This chromium oxide then prevents metal from corrosion. It is concluded that the higher the nickel content the higher the corrosion rate of materials. The composition potential-pH diagrams for Fe-S-H 2 O and Ni-S-H 2 O show that the stability fields of FeS and NiS cover a wide range of pH. The effect of sulfur or sulfide ions in promoting dissolution of Fe and/or Ni are highly possible. The activating influence of sulfur compounds on Ni is stronger than that of Fe, although the highly electronic conductivity of iron sulfides can catalyze the cathodic reaction. Undoubtedly, sulfur compound strongly depassivates high Ni contents materials

  16. Influence of core thickness and artificial aging on the biaxial flexural strength of different all-ceramic materials: An in-vitro study.

    Science.gov (United States)

    Dikicier, Sibel; Ayyildiz, Simel; Ozen, Julide; Sipahi, Cumhur

    2017-05-31

    The purpose of this study was to investigate the flexural strength of all-ceramics with varying core thicknesses submitted to aging. In-Ceram Alumina (IC), IPS e.max Press (EM) and Katana (K) (n=40), were selected. Each group contained two core groups based on the core thickness as follows: IC/0.5, IC/0.8, EM/0.5, EM/0.8, K/0.5 and K/0.8 mm in thickness (n=20 each). Ten specimens from each group were subjected to aging and all specimens were tested for strength in a testing machine either with or without being subjected aging. The mean strength of the K were higher (873.05 MPa) than that of the IC (548.28 MPa) and EM (374.32 MPa) regardless of core thickness. Strength values increased with increasing core thickness for all IC, EM and K regardless of aging. Results of this study concluded that strength was not significantly affected by aging. Different core thicknesses affected strength of the all-ceramic materials tested (p<0.05).

  17. Influence of Nonfused Cores on the Photovoltaic Performance of Linear Triphenylamine-Based Hole-Transporting Materials for Perovskite Solar Cells.

    Science.gov (United States)

    Wu, Yungen; Wang, Zhihui; Liang, Mao; Cheng, Hua; Li, Mengyuan; Liu, Liyuan; Wang, Baiyue; Wu, Jinhua; Prasad Ghimire, Raju; Wang, Xuda; Sun, Zhe; Xue, Song; Qiao, Qiquan

    2018-05-18

    The core plays a crucial role in achieving high performance of linear hole transport materials (HTMs) toward the perovskite solar cells (PSCs). Most studies focused on the development of fused heterocycles as cores for HTMs. Nevertheless, nonfused heterocycles deserve to be studied since they can be easily synthesized. In this work, we reported a series of low-cost triphenylamine HTMs (M101-M106) with different nonfused cores. Results concluded that the introduced core has a significant influence on conductivity, hole mobility, energy level, and solubility of linear HTMs. M103 and M104 with nonfused oligothiophene cores are superior to other HTMs in terms of conductivity, hole mobility, and surface morphology. PSCs based on M104 exhibited the highest power conversion efficiency of 16.50% under AM 1.5 sun, which is comparable to that of spiro-OMeTAD (16.67%) under the same conditions. Importantly, the employment of M104 is highly economical in terms of the cost of synthesis as compared to that of spiro-OMeTAD. This work demonstrated that nonfused heterocycles, such as oligothiophene, are promising cores for high performance of linear HTMs toward PSCs.

  18. Evaluation report on CCTF Core-II reflood test C2-1 (Run 55)

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Sugimoto, Jun; Akimoto, Hajime; Okubo, Tsutomu; Murao, Yoshio

    1991-10-01

    A high pressure test (0.42 MPa) on the reflood phenomena was performed with the CCTF. The result of the test was compared with the experimental result of the base case test (0.2 MPa). (1) The overall flow characteristics in the high pressure test were qualitatively similar to that of the base case test. Any qualitatively different phenomena were not recognized during reflood phase. This indicates that it is reasonable to utilize the physical reflood model developed from the result of the base case test to the high pressure condition at least up to 0.42 MPa for prediction of reflood behavior of PWRs. (2) On the other hand, following quantitative influence of high pressure on reflood phenomena was observed. The core cooling was better, and the mass flow rate of the steam generated in the core was larger. However, the steam velocity was smaller due to higher density of the steam. Therefore, the steam discharge through loops was easier and hence the so-called steam binding effect was weaker. And, the water accumulation rate in the core was larger. Consequently the core flooding mass flow rate was larger. Since the core cooling was better, the maximum core temperature was lower and the last quenching was earlier. This result was the same as that previously observed in CCTF tests in the scope of the pressure upto 0.3 MPa. (3) The higher pressure leads to the better