Development of core design and analyses technology for integral reactor

International Nuclear Information System (INIS)

Zee, Sung Quun; Lee, C. C.; Kim, K. Y.

2002-03-01

In general, small and medium-sized integral reactors adopt new technology such as passive and inherent safety concepts to minimize the necessity of power source and operator actions, and to provide the automatic measures to cope with any accidents. Specifically, such reactors are often designed with a lower core power density and with soluble boron free concept for system simplification. Those reactors require ultra long cycle operation for higher economical efficiency. This cycle length requirement is one of the important factors in the design of burnable absorbers as well as assurance of shutdown margin. Hence, both computer code system and design methodology based on the today's design technology for the current commercial reactor cores require intensive improvement for the small and medium-sized soluble boron free reactors. New database is also required for the development of this type of reactor core. Under these technical requirements, conceptual design of small integral reactor SMART has been performed since July 1997, and recently completed under the long term nuclear R and D program. Thus, the final objectives of this work is design and development of an integral reactor core and development of necessary indigenous design technology. To reach the goal of the 2nd stage R and D program for basic design of SMART, design bases and requirements adequate for ultra long cycle and soluble boron free concept are established. These bases and requirements are satisfied by the core loading pattern. Based on the core loading pattern, nuclear, and thermal and hydraulic characteristics are analyzed. Also included are fuel performance analysis and development of a core protection and monitoring system that is adequate for the soluble boron free core of an integral reactor. Core shielding design analysis is accomplished, too. Moreover, full scope interface data are produced for reactor safety and performance analyses and other design activities. Nuclear, thermal and

In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

International Nuclear Information System (INIS)

Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

2010-01-01

According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

Core journals in library and information science: measuring the level of specialization over time

DEFF Research Database (Denmark)

Nicolaisen, J.; Frandsen, T. F.

2013-01-01

years. The method is applied to a selection of core journals in library and information science (1990-2012). The reference lists of each journal are compared year by year, and the percentage of re-citations is calculated by dividing the number of re-citations with the total number of citations each year......Introduction. Specialization in science is a process that occurs over time. The present paper presents a bibliometric method for measuring the degree of specialization over time. Methods. The method is based on bibliographic coupling, and counts the percentage of recitations given in subsequent...

Development of the computer code system for the analyses of PWR core

International Nuclear Information System (INIS)

Tsujimoto, Iwao; Naito, Yoshitaka.

1992-11-01

This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)

Analyses on the BFS critical experiments. An analysis on the BFS-62-1 and 62-2 cores

International Nuclear Information System (INIS)

Sugino, Kazuteru; Shono, Akira

2002-04-01

In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 and BFS-62-2 cores. The BFS-62-1 core models the present BN-600, and contains the enriched UO 2 fuel surrounded by the UO 2 blanket. The BFS-62-2 core has the same layout as the BFS-62-1 but the blanket region was replaced with stainless steel shied. For core parameter analyses, the 3-D Hexagonal-Z or XYZ geometry model was applied by not only diffusion calculation but also transport calculation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality, the reaction rate ratio and reaction rate distribution in BFS-62-1. In the reaction rate distribution of BFS-62-2 calculation without cross-section adjustment produced big radial dependency of calculation over experiment value (C/E value) in the core region and overestimation in the shield region. Cross-section adjustment technique procedure improved those estimation, however alternation of cross-section of Iron, which was dominant in above improvement, compared to the cross-section error, and further investigation was required. Concerning the control rod worth of BFS-62-1, radial dependency of the C/E value was observed whether cross-section adjustment technique was applied or not, therefore comparison with results of other BFS-62 cores analyses is

Development of Special Tools for the Straightness Measurement of JRTR Core Inner Shell

International Nuclear Information System (INIS)

Sinjlawi, Abdullah; Cho, Yeong-Garp; Chung, Jong-Ha

2014-01-01

Jordan Research and Training Reactor (JRTR) is an open pool type nuclear research reactor, 5 MW power, JRTR core made from Zircaloy. The JRTR will be used for nuclear applications such as isotopes production, nuclear researches, neutron transmutation doping (NTD), and training. JRTR core structures will be exposed to a large amount of neutron irradiation during the life time of the reactor. The core inner shell also will be exposed to a pressure that comes from heavy water system. JRTR core inner shell will deform due to the neutron irradiation and the mechanical stress. Therefore, the dimensional change of the core inner shell should be periodically (every 10 years) measured as an in-service inspection to confirm the structural integrity. As a result of neutron irradiation, pressure difference of the heavy water vessel, and the mechanical stress, the reactor core will deform as shown in figure 2 to figure 4. The maximum deformation to the normal direction of inner shell wall is 0.75 mm as shown in figure 3. This study discusses development of special tools that will be used for pre-service and in-service inspection of JRTR inner shell. The performance and procedure for the measurements tools will be verified using by the real inner shell of the heavy water vessel at factory before shipping to Jordan.. There will be very delicate working procedure for the measurement in the limited space in JRTR core. Therefore, we will develop the detail procedures to cover the removal of the core components, installation of the measurement tools, measurement, and re-installation of the core components. The measurement of the inner shell at JAEC site during commissioning stage will be the first remote measurement at the same conditions of pool water and heavy water system

Analyses of the stability and core taxonomic memberships of the human microbiome.

Directory of Open Access Journals (Sweden)

Kelvin Li

Full Text Available Analyses of the taxonomic diversity associated with the human microbiome continue to be an area of great importance. The study of the nature and extent of the commonly shared taxa ("core", versus those less prevalent, establishes a baseline for comparing healthy and diseased groups by quantifying the variation among people, across body habitats and over time. The National Institutes of Health (NIH sponsored Human Microbiome Project (HMP has provided an unprecedented opportunity to examine and better define what constitutes the taxonomic core within and across body habitats and individuals through pyrosequencing-based profiling of 16S rRNA gene sequences from oral, skin, distal gut (stool, and vaginal body habitats from over 200 healthy individuals. A two-parameter model is introduced to quantitatively identify the core taxonomic members of each body habitat's microbiota across the healthy cohort. Using only cutoffs for taxonomic ubiquity and abundance, core taxonomic members were identified for each of the 18 body habitats and also for the 4 higher-level body regions. Although many microbes were shared at low abundance, they exhibited a relatively continuous spread in both their abundance and ubiquity, as opposed to a more discretized separation. The numbers of core taxa members in the body regions are comparatively small and stable, reflecting the relatively high, but conserved, interpersonal variability within the cohort. Core sizes increased across the body regions in the order of: vagina, skin, stool, and oral cavity. A number of "minor" oral taxonomic core were also identified by their majority presence across the cohort, but with relatively low and stable abundances. A method for quantifying the difference between two cohorts was introduced and applied to samples collected on a second visit, revealing that over time, the oral, skin, and stool body regions tended to be more transient in their taxonomic structure than the vaginal body region.

Oxygen and carbon isotope analyses of a Late Quaternary core in the Zaire (Congo) fan

International Nuclear Information System (INIS)

Olausson, E.

1984-01-01

Oxygen and carbon isotope analyses have been carried out on samples from a core of the Angola Basin (6 0 50'S, 10 0 45'E, depth 2100 m). The pelagic foraminifer Globigerinoides ruber, a species with a shallow water habitat, and two benthic species Uvigerina peregrina and Bulimina aculeata have been analysed. The data are given relative to PDB. (Auth.)

Transient analyses for a molten salt fast reactor with optimized core geometry

Energy Technology Data Exchange (ETDEWEB)

Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

2015-10-15

Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-11-01

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

A Preliminary Core Domain Set for Clinical Trials of Shoulder Disorders: A Report from the OMERACT 2016 Shoulder Core Outcome Set Special Interest Group.

Science.gov (United States)

Buchbinder, Rachelle; Page, Matthew J; Huang, Hsiaomin; Verhagen, Arianne P; Beaton, Dorcas; Kopkow, Christian; Lenza, Mario; Jain, Nitin B; Richards, Bethan; Richards, Pamela; Voshaar, Marieke; van der Windt, Danielle; Gagnier, Joel J

2017-12-01

The Outcome Measures in Rheumatology (OMERACT) Shoulder Core Outcome Set Special Interest Group (SIG) was established to develop a core outcome set (COS) for clinical trials of shoulder disorders. In preparation for OMERACT 2016, we systematically examined all outcome domains and measurement instruments reported in 409 randomized trials of interventions for shoulder disorders published between 1954 and 2015. Informed by these data, we conducted an international Delphi consensus study including shoulder trial experts, clinicians, and patients to identify key domains that should be included in a shoulder disorder COS. Findings were discussed at a stakeholder premeeting of OMERACT. At OMERACT 2016, we sought consensus on a preliminary core domain set and input into next steps. There were 13 and 15 participants at the premeeting and the OMERACT 2016 SIG meeting, respectively (9 attended both meetings). Consensus was reached on a preliminary core domain set consisting of an inner core of 4 domains: pain, physical function/activity, global perceived effect, and adverse events including death. A middle core consisted of 3 domains: emotional well-being, sleep, and participation (recreation and work). An outer core of research required to inform the final COS was also formulated. Our next steps are to (1) analyze whether participation (recreation and work) should be in the inner core, (2) conduct a third Delphi round to finalize definitions and wording of domains and reach final endorsement for the domains, and (3) determine which instruments fulfill the OMERACT criteria for measuring each domain.

Array analyses of SmKS waves and the stratification of Earth's outermost core

Science.gov (United States)

Kaneshima, Satoshi

2018-03-01

We perform array analyses of SmKS waves in order to investigate the Vp structure of the Earth's outermost core. For earthquakes recorded by broadband seismometer networks in the world, we measure differential travel times between S3KS and S2KS, between S4KS and S3KS, and between S5KS and S3KS by array techniques. The differential times are well fit by a Vp model of the Earth's outermost core, KHOMC (Kaneshima and Helffrich, 2013). Differential slownesses of S4KS and S2KS relative to S2KS are also measured for the highest quality data. The measured slownesses, with unique sensitivity to the outer core 200-400 km below the CMB, are matched by KHOMC. These observations consolidate the evidence for the presence at the top of the outer core of a layer that has a distinctively steeper Vp gradient than the bulk of the outer core. We invert new SmKS differential time data set by a tau-p method and attempt to refine the Vp profile of KHOMC. The essential features of KHOMC are preserved after the model refinement. However, the newly estimated layer thickness is nearly 450 km, which is thicker than that of KHOMC. The Vp anomalies relative to PREM for the depths 400-800 km below the CMB are less than 0.03 km/s, consistent with the degree of agreement between different Vp models for the depth range.

Reentry safety for the Topaz II Space Reactor: Issues and analyses

International Nuclear Information System (INIS)

Connell, L.W.; Trost, L.C.

1994-03-01

This report documents the reentry safety analyses conducted for the TOPAZ II Nuclear Electric Propulsion Space Test Program (NEPSTP). Scoping calculations were performed on the reentry aerothermal breakup and ground footprint of reactor core debris. The calculations were used to assess the risks associated with radiologically cold reentry accidents and to determine if constraints should be placed on the core configuration for such accidents. Three risk factors were considered: inadvertent criticality upon reentry impact, atmospheric dispersal of U-235 fuel, and the Special Nuclear Material Safeguards risks. Results indicate that the risks associated with cold reentry are very low regardless of the core configuration. Core configuration constraints were therefore not established for radiologically cold reentry accidents

ESFR core optimization and uncertainty studies

International Nuclear Information System (INIS)

Rineiski, A.; Vezzoni, B.; Zhang, D.; Marchetti, M.; Gabrielli, F.; Maschek, W.; Chen, X.-N.; Buiron, L.; Krepel, J.; Sun, K.; Mikityuk, K.; Polidoro, F.; Rochman, D.; Koning, A.J.; DaCruz, D.F.; Tsige-Tamirat, H.; Sunderland, R.

2015-01-01

In the European Sodium Fast Reactor (ESFR) project supported by EURATOM in 2008-2012, a concept for a large 3600 MWth sodium-cooled fast reactor design was investigated. In particular, reference core designs with oxide and carbide fuel were optimized to improve their safety parameters. Uncertainties in these parameters were evaluated for the oxide option. Core modifications were performed first to reduce the sodium void reactivity effect. Introduction of a large sodium plenum with an absorber layer above the core and a lower axial fertile blanket improve the total sodium void effect appreciably, bringing it close to zero for a core with fresh fuel, in line with results obtained worldwide, while not influencing substantially other core physics parameters. Therefore an optimized configuration, CONF2, with a sodium plenum and a lower blanket was established first and used as a basis for further studies in view of deterioration of safety parameters during reactor operation. Further options to study were an inner fertile blanket, introduction of moderator pins, a smaller core height, special designs for pins, such as 'empty' pins, and subassemblies. These special designs were proposed to facilitate melted fuel relocation in order to avoid core re-criticality under severe accident conditions. In the paper further CONF2 modifications are compared in terms of safety and fuel balance. They may bring further improvements in safety, but their accurate assessment requires additional studies, including transient analyses. Uncertainty studies were performed by employing a so-called Total Monte-Carlo method, for which a large number of nuclear data files is produced for single isotopes and then used in Monte-Carlo calculations. The uncertainties for the criticality, sodium void and Doppler effects, effective delayed neutron fraction due to uncertainties in basic nuclear data were assessed for an ESFR core. They prove applicability of the available nuclear data for ESFR

Addressing Pre-Service Teachers' Understandings and Difficulties with Some Core Concepts in the Special Theory of Relativity

Science.gov (United States)

Selcuk, Gamze Sezgin

2011-01-01

The aim of this study is to investigate pre-service teachers' understanding of and difficulties with some core concepts in the special theory of relativity. The pre-service teachers (n = 185) from the Departments of Physics Education and Elementary Science Education at Dokuz Eylul University (in Turkey) participated. Both quantitative and…

TCA UO2/MOX core analyses

International Nuclear Information System (INIS)

Tahara, Yoshihisa; Noda, Hideyuki

2000-01-01

In order to examine the adequacy of nuclear data, the TCA UO 2 and MOX core experiments were analyzed with MVP using the libraries based on ENDF/B-VI Mod.3 and JENDL-3.2. The ENDF/B-VI data underpredict k eff values. The replacement of 238 U data with the JENDL-3.2 data and the adjustment of 235 ν-value raise the k eff values by 0.3% for UO 2 cores, but still underpredict k eff values. On the other hand, the nuclear data of JENDL-3.2 for H, O, Al, 238 U and 235 U of ENDF/B-VI whose 235 ν-value in thermal energy region is adjusted to the average value of JENDL-3.2 give a good prediction of k eff . (author)

Structural and magnetic properties of multi-core nanoparticles analysed using a generalised numerical inversion method

Science.gov (United States)

Bender, P.; Bogart, L. K.; Posth, O.; Szczerba, W.; Rogers, S. E.; Castro, A.; Nilsson, L.; Zeng, L. J.; Sugunan, A.; Sommertune, J.; Fornara, A.; González-Alonso, D.; Barquín, L. Fernández; Johansson, C.

2017-01-01

The structural and magnetic properties of magnetic multi-core particles were determined by numerical inversion of small angle scattering and isothermal magnetisation data. The investigated particles consist of iron oxide nanoparticle cores (9 nm) embedded in poly(styrene) spheres (160 nm). A thorough physical characterisation of the particles included transmission electron microscopy, X-ray diffraction and asymmetrical flow field-flow fractionation. Their structure was ultimately disclosed by an indirect Fourier transform of static light scattering, small angle X-ray scattering and small angle neutron scattering data of the colloidal dispersion. The extracted pair distance distribution functions clearly indicated that the cores were mostly accumulated in the outer surface layers of the poly(styrene) spheres. To investigate the magnetic properties, the isothermal magnetisation curves of the multi-core particles (immobilised and dispersed in water) were analysed. The study stands out by applying the same numerical approach to extract the apparent moment distributions of the particles as for the indirect Fourier transform. It could be shown that the main peak of the apparent moment distributions correlated to the expected intrinsic moment distribution of the cores. Additional peaks were observed which signaled deviations of the isothermal magnetisation behavior from the non-interacting case, indicating weak dipolar interactions. PMID:28397851

Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

International Nuclear Information System (INIS)

DeVault, G.P.; Bell, C.R.

1985-01-01

Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

Drilling equipment for difficult coring conditions: a new type of core lifter and triple tube core barrel

Energy Technology Data Exchange (ETDEWEB)

Robinson, J B

1968-08-01

Although considerable improvements in diamond drilling equipment have been made since the early 1950's, deficiencies in existing equipment led to the development of a new type core lifter and special 20 ft triple tube core barrel designed to operate in bad coring conditions. It is claimed that although developed essentially for coal drilling, the new equipment could be adapted to other fields of diamond drilling with the cost advantage of increased life of the core lifter.

JENDL-3.2 performance in analyses of MISTRAL critical experiments for high-moderation MOX cores

International Nuclear Information System (INIS)

Takada, Naoyuki; Hibi, Koki; Ishii, Kazuya; Ando, Yoshihira; Yamamoto, Toru; Ueji, Masao; Iwata, Yutaka

2001-01-01

NUPEC and CEA have launched an extensive experimental program called MISTRAL to study highly moderated MOX cores for the advanced LWRs. The analyses using SRAC system and MVP code with JENDL-3.2 library are in progress on the experiments of the MISTRAL and the former EPICURE programs. Various comparisons have been made between calculation results and measurement values. (author)

Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

International Nuclear Information System (INIS)

Heard, F.J.

1997-01-01

The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

International Nuclear Information System (INIS)

Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F.

2009-01-01

The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U 235 (typically Pu 242 , Np 237 , U 238 , Th 232 ). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

Core design and fuel rod analyses of a super fast reactor with high power density

International Nuclear Information System (INIS)

Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

2009-01-01

A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. One of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak(MPP), Maximum Discharge Burnup(MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be

Introduction to the Special Issue “Sport for Social Inclusion: Critical Analyses and Future Challenges”

Directory of Open Access Journals (Sweden)

Reinhard Haudenhuyse

2015-06-01

Full Text Available “Sport for Social Inclusion: Critical Analyses and Future Challenges” brings together a unique collection of papers on the subject of sport and social inclusion. The special issue can be divided into three major parts. The first part consists of three papers tacking on a broad perspective on sport and social exclusion, with specific attention to austerity policies, sport-for-change and exclusion in youth sports. The second part of the special issue tackles specific themes (e.g., group composition and dynamics, volunteering, physical education, youth work, equality, public health and groups (e.g., people with disabilities, disadvantaged girls, youth in society in relation to sport and social exclusion. The third part consists of three papers that are related to issues of multiculturalism, migration and social inclusion. The special issue is further augmented with a book review on Mike Collins and Tess Kay’s Sport and social exclusion (2nd edition and a short research communication. The editors dedicate the special issue to Mike Collins (deceased.

Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F. [CEA, DEN, Dosimetry Command Control and Instrumentation Laboratory, F-13109 Saint-Paul-lez-Durance (France)

2009-07-01

The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U{sup 235} (typically Pu{sup 242}, Np{sup 237}, U{sup 238}, Th{sup 232}). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

Development of a 3-dimensional calculation model of the Danish research reactor DR3 to analyse a proposal to a new core design called ring-core

Energy Technology Data Exchange (ETDEWEB)

Nonboel, E

1985-07-01

A 3-dimensional calculation model of the Danish research reactor DR3 has been developed. Demands of a more effective utilization of the reactor and its facilities has required a more detailed calculation tool than applied so far. A great deal of attention has been devoted to the treatment of the coarse control arms. The model has been tested against measurements with satisfying results. Furthermore the model has been used to analyse a proposal to a new core design called ring-core where 4 central fuel elements are replaced by 4 dummy elements to increase the thermal flux in the center of the reactor. (author)

Core-annular flow through a horizontal pipe : Hydrodynamic counterbalancing of buoyancy force on core

NARCIS (Netherlands)

Ooms, G.; Vuik, C.; Poesio, P.

2007-01-01

A theoretical investigation has been made of core-annular flow: the flow of a high-viscosity liquid core surrounded by a low-viscosity liquid annular layer through a horizontal pipe. Special attention is paid to the question of how the buoyancy force on the core, caused by a density difference

Review of accident analyses of RB experimental reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-01-01

The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VINCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62; yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin) consisting of 2% enriched uranium metal and 80% enriched U0 2 , dispersed in aluminum matrix, have been available since 1962 and 1976, respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements, as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINCA Institute, an independent regulator)' body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety' Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed) to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given. (author)

Some Examples of Accident Analyses for RB Reactor

International Nuclear Information System (INIS)

Pesic, M.

2002-01-01

The RB reactor is heavy water critical assembly operated in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, since April 1959. The first Safety Analysis Report of the RB critical assembly was prepared in 1961/62. But, the first accidental analysis was done in late 1958 in aim the examine power transient and total equivalent doses received by the staff during the reactivity accident occurred on October 15, 1958. Since 1960, the RB reactor is modified few times. Beside initial natural uranium metal fuel rods, new fuel (TVR-S types) from 2% enriched metal uranium and 80% enriched UO 2 were available since 1962 and 1976, respectively. Also, modifications in control and safety systems of the reactor were done occasionally. Special reactor cores were created using all three types of fuel elements, among them, the coupled fast-thermal ones. Nuclear Safety Committee of the Vinca Institute, an independent regulatory body approved for usage all these modifications of the RB reactor. For those decisions of the Committee, the Preliminary Safety Analysis Reports were prepared that, beside proposed technical modifications and new regulation rules had included analyses of various possible accidents. Special attention is given and new methodology was proposed for thoroughly analyses of design based accidents related to coupled fast-thermal cores, that include reactor central zones filled by fuel elements without moderator. In these accidents, during assumed flooding of the fast zone by moderator, a very high reactivity could be inserted in the system with very high reactivity rate. It was necessary to provide that the safety system of the reactor had fast response to that accident and had enough high (negative) reactivity to shut down the reactor timely. In this paper, a brief overview of some accidents, methodology and computation tools used for the accident analyses at RB reactor are given. (author)

Review of accident analyses of RB experimental reactor

Directory of Open Access Journals (Sweden)

Pešić Milan P.

2003-01-01

Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

Scientific Drilling of Impact Craters - Well Logging and Core Analyses Using Magnetic Methods (Invited)

Science.gov (United States)

Fucugauchi, J. U.; Perez-Cruz, L. L.; Velasco-Villarreal, M.

2013-12-01

Drilling projects of impact structures provide data on the structure and stratigraphy of target, impact and post-impact lithologies, providing insight on the impact dynamics and cratering. Studies have successfully included magnetic well logging and analyses in core and cuttings, directed to characterize the subsurface stratigraphy and structure at depth. There are 170-180 impact craters documented in the terrestrial record, which is a small proportion compared to expectations derived from what is observed on the Moon, Mars and other bodies of the solar system. Knowledge of the internal 3-D deep structure of craters, critical for understanding impacts and crater formation, can best be studied by geophysics and drilling. On Earth, few craters have yet been investigated by drilling. Craters have been drilled as part of industry surveys and/or academic projects, including notably Chicxulub, Sudbury, Ries, Vredefort, Manson and many other craters. As part of the Continental ICDP program, drilling projects have been conducted on the Chicxulub, Bosumtwi, Chesapeake, Ries and El gygytgyn craters. Inclusion of continuous core recovery expanded the range of paleomagnetic and rock magnetic applications, with direct core laboratory measurements, which are part of the tools available in the ocean and continental drilling programs. Drilling studies are here briefly reviewed, with emphasis on the Chicxulub crater formed by an asteroid impact 66 Ma ago at the Cretaceous/Paleogene boundary. Chicxulub crater has no surface expression, covered by a kilometer of Cenozoic sediments, thus making drilling an essential tool. As part of our studies we have drilled eleven wells with continuous core recovery. Magnetic susceptibility logging, magnetostratigraphic, rock magnetic and fabric studies have been carried out and results used for lateral correlation, dating, formation evaluation, azimuthal core orientation and physical property contrasts. Contributions of magnetic studies on impact

HYDRATE CORE DRILLING TESTS

Energy Technology Data Exchange (ETDEWEB)

John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell

2002-11-01

The ''Methane Hydrate Production from Alaskan Permafrost'' project is a three-year endeavor being conducted by Maurer Technology Inc. (MTI), Noble, and Anadarko Petroleum, in partnership with the U.S. DOE National Energy Technology Laboratory (NETL). The project's goal is to build on previous and ongoing R&D in the area of onshore hydrate deposition. The project team plans to design and implement a program to safely and economically drill, core and produce gas from arctic hydrates. The current work scope includes drilling and coring one well on Anadarko leases in FY 2003 during the winter drilling season. A specially built on-site core analysis laboratory will be used to determine some of the physical characteristics of the hydrates and surrounding rock. Prior to going to the field, the project team designed and conducted a controlled series of coring tests for simulating coring of hydrate formations. A variety of equipment and procedures were tested and modified to develop a practical solution for this special application. This Topical Report summarizes these coring tests. A special facility was designed and installed at MTI's Drilling Research Center (DRC) in Houston and used to conduct coring tests. Equipment and procedures were tested by cutting cores from frozen mixtures of sand and water supported by casing and designed to simulate hydrate formations. Tests were conducted with chilled drilling fluids. Tests showed that frozen core can be washed out and reduced in size by the action of the drilling fluid. Washing of the core by the drilling fluid caused a reduction in core diameter, making core recovery very difficult (if not impossible). One successful solution was to drill the last 6 inches of core dry (without fluid circulation). These tests demonstrated that it will be difficult to capture core when drilling in permafrost or hydrates without implementing certain safeguards. Among the coring tests was a simulated hydrate

Risk reduction of core-melt accidents in advaned CAPRA burner cores

International Nuclear Information System (INIS)

Maschek, W.; Struwe, D.; Eigemann, M.

1997-01-01

As part of the CAPRA Program (Consommation Accrue de Plutonium dans les RApides) the feasibility of fast reactors is investigated to burn plutonium and also to destruct minor actinides. The design of CAPRA cores shows significant differences compared to conventional cores. Especially the high Pu-enrichment has an important influence on the core melt-down behavior and the associated recriticality risk. To cope with this risk, inherent design features and special measures/devices are investigated for their potential of early fuel discharge to reduce the criticality of the reactor core. An assessment of such measures/devices is given and experimental needs are formulated. 11 refs., 5 figs

Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

Energy Technology Data Exchange (ETDEWEB)

Zwermann, W.; Aures, A.; Bernnat, W.; and others

2013-06-15

This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

Core story creation: analysing narratives to construct stories for learning.

Science.gov (United States)

Petty, Julia; Jarvis, Joy; Thomas, Rebecca

2018-03-16

Educational research uses narrative enquiry to gain and interpret people's experiences. Narrative analysis is used to organise and make sense of acquired narrative. 'Core story creation' is a way of managing raw data obtained from narrative interviews to construct stories for learning. To explain how core story creation can be used to construct stories from raw narratives obtained by interviewing parents about their neonatal experiences and then use these stories to educate learners. Core story creation involves reconfiguration of raw narratives. Reconfiguration includes listening to and rereading transcribed narratives, identifying elements of 'emplotment' and reordering these to form a constructed story. Thematic analysis is then performed on the story to draw out learning themes informed by the participants. Core story creation using emplotment is a strategy of narrative reconfiguration that produces stories which can be used to develop resources relating to person-centred education about the patient experience. Stories constructed from raw narratives in the context of constructivism can provide a medium or an 'end product' for use in learning resource development. This can then contribute to educating students or health professionals about patients' experiences. ©2018 RCN Publishing Company Ltd. All rights reserved. Not to be copied, transmitted or recorded in any way, in whole or part, without prior permission of the publishers.

Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations. Interim report; Weiterentwicklung moderner Verfahren zu Neutronentransport und Unsicherheitsanalysen fuer Kernberechnungen. Zwischenbericht

Energy Technology Data Exchange (ETDEWEB)

Zwermann, Winfried; Aures, Alexander; Bostelmann, Friederike; Pasichnyk, Ihor; Perin, Yann; Velkov, Kiril; Zilly, Matias

2016-12-15

This report documents the status of the research and development goals reached within the reactor safety research project RS1536 ''Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations'' as of the 3{sup rd} quarter of 2016. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts, in particular fast reactors cooled by liquid metal. The contributing individual goals are the further optimization and validation of deterministic calculation methods with high spatial and energy resolution, the development of a coupled calculation system using the Monte Carlo method for the neutron transport to describe time-dependent reactor core states, the processing and validation of nuclear data, particularly with regard to covariance data, the development, validation, and application of sampling-based methods for uncertainty and sensitivity analyses, the creation of a platform for performing systematic uncertainty analyses for fast reactor systems, as well as the description of states of severe core damage with the Monte Carlo method. Moreover, work regarding the European NURESAFE project, started in the preceding project RS1503, are being continued and completed.

48{sup th} Annual meeting on nuclear technology (AMNT 2017). Key topic / Enhanced safety and operation excellence. Focus session: Uncertainty analyses in reactor core simulations

Energy Technology Data Exchange (ETDEWEB)

Zwermann, Winfried [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Forschungszentrum

2017-12-15

The supplementation of reactor simulations by uncertainty analyses is becoming increasingly important internationally due to the fact that the reliability of simulation calculations can be significantly increased by the quantification of uncertainties in comparison to the use of so-called conservative methods (BEPU- ''Best-Estimate plus Uncertainties''). While systematic uncertainty analyses for thermo-hydraulic calculations have been performed routinely for a long time, methods for taking into account uncertainties in nuclear data, which are the basis for neutron transport calculations, are under development. The Focus Session Uncertainty Analyses in Reactor Core Simulations was intended to provide an overview of international research and development with respect to supplementing reactor core simulations with uncertainty and sensitivity analyses, in research institutes as well as within the nuclear industry. The presented analyses not only focused on light water reactors, but also on advanced reactor systems. Particular emphasis was put on international benchmarks in the field. The session was chaired by Winfried Zwermann (Gesellschaft fuer Anlagen- und Reaktorsicherheit).

Specialized languages

DEFF Research Database (Denmark)

Mousten, Birthe; Laursen, Anne Lise

2016-01-01

Across different fields of research, one feature is often overlooked: the use of language for specialized purposes (LSP) as a cross-discipline. Mastering cross-disciplinarity is the precondition for communicating detailed results within any field. Researchers in specialized languages work cross...... science fields communicate their findings. With this article, we want to create awareness of the work in this special area of language studies and of the inherent cross-disciplinarity that makes LSP special compared to common-core language. An acknowledgement of the importance of this field both in terms...... of more empirical studies and in terms of a greater application of the results would give language specialists in trade and industry a solid and updated basis for communication and language use....

The Late Pliocene Eltanin Impact - Documentation From Sediment Core Analyses

Science.gov (United States)

Gersonde, R.; Kuhn, G.; Kyte, F. T.; Flores, J.; Becquey, S.

2002-12-01

The expeditions ANT-XII/4 (1995) and ANT-XVIII/5a (2001) of the RV POLARSTERN collected extensive bathymetric and seismic data sets as well as sediment cores from an area in the Bellingshausen Sea (eastern Pacific Southern Ocean) that allow the first comprehensive geoscientific documentation of an asteroid impact into a deep ocean (~ 5 km) basin, named the Eltanin impact. Impact deposits have now been recovered from a total of more than 20 sediment cores collected in an area covering about 80,000 km2. Combined biomagnetostratigraphic dating places the impact event into the earliest Matuyama Chron, a period of enhanced climate variability. Sediment texture analyses and studies of sediment composition including grain size and microfossil distribution reveal the pattern of impact-related sediment disturbance and the sedimentary processes immediately following the impact event. The pattern is complicated by the San Martin Seamounts (~57.5 S, 91 W), a large topographic elevation that rises up to 3000 m above the surrounding abyssal plain in the area affected by the Eltanin impact. The impact ripped up sediments as old as Eocene and probably Paleocene that have been redeposited in a chaotic assemblage. This is followed by a sequence sedimented from a turbulent flow at the sea floor, overprinted by fall-out of airborne meteoritic ejecta that settled trough the water column. Grain size distribution reveals the timing and interaction of the different sedimentary processes. The gathered estimate of ejecta mass deposited over the studied area, composed of shock-melted asteroidal matrial and unmelted meteorites including fragments up to 2.5 cm in diameter, point to an Eltanin asteroid larger than the 1 km in diameter size originally suggested as a minimum based on the ANT-XII/4 results. This places the energy released by the impact at the threshold of those considered to cause environmental disturbance at a global scale and it makes the impact a likely transport mechanism

Current directions in core-shell nanoparticle design

Science.gov (United States)

Schärtl, Wolfgang

2010-06-01

Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems.Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems

HTR core physics analysis at NRG

International Nuclear Information System (INIS)

Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

2002-01-01

Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

Results of neutron physics analyses of WWER-440 cores with modified reactor protection and control systems

International Nuclear Information System (INIS)

Lehmann, M.; Pecka, M.; Rocek, J.; Zalesky, K.

1993-12-01

Detailed results are given of neutron physics analyses performed to assess the efficiency and acceptability of modifications of the WWER-440 core protection and control system; the modifications have been proposed with a view to increasing the proportion of mechanical control in the compensation of reactivity effects during reactor unit operation in the variable load mode. The calculations were carried out using the modular MOBY-DICK macrocode system together with the SMV42G36 library of two-group parametrized diffusion constants, containing corrections which allow new-design WWER-440 fuel assemblies to be discriminated. (J.B). 37 tabs., 18 figs., 5 refs

Interpretation of Actinide-Distribution Data Obtained from Non-Destructive and Destructive Post-Test Analyses of an Intact-Core Column of Culebra Dolomite

International Nuclear Information System (INIS)

LUCERO, DANIEL A.; PERKINS, W. GEORGE

1999-01-01

The US DOE, with technical assistance from Sandia National Laboratories, has successfully received EPA certification and opened the Waste Isolation Pilot Plant (WIPP), a nuclear waste disposal facility located approximately 42 km east of Carlsbad, New Mexico. Performance assessment analyses indicate that human intrusions by inadvertent, intermittent drilling for resources provide the only credible mechanisms for releases of radionuclides from the disposal system. In modeling long-term brine releases, subsequent to a drilling event, potential migration pathways through the permeable layers of rock above the Salado formation were analyzed. Major emphasis is placed on the Culebra Member of the Rustler Formation because this is the most transmissive geologic layer overlying the WIPP site. In order to help quantify parameters for the calculated releases, radionuclide transport experiments have been earned out using intact-core columns obtained from the Culebra dolomite member of the Rustler Formation within the WIPP site. This paper deals primarily with results of analyses for 241 Pu and 241 Am distributions developed during transport experiments in one of these cores. Transport experiments were done using a synthetic brine that simulates Culebra brine at the core recovery location (the WIPP air-intake shaft--AIS). Hydraulic characteristics (i.e., apparent porosity and apparent dispersion coefficient) for intact-core columns were obtained via experiments using the conservative tracer 22 Na. Elution experiments carried out over periods of a few days with tracers 232 U and 239 Np indicated that these tracers were weakly retarded as indicated by delayed elution of the species. Elution experiments with tracers 241 Pu and 241 Am were attempted, but no elution of either species has been observed to date, including experiments of many months' duration. In order to quantify retardation of the non-eluted species 241 Pu and 241 Am after a period of brine flow, non-destructive and

Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

Energy Technology Data Exchange (ETDEWEB)

Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-03-01

System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

Geochemistry of mercury and other constituents in subsurface sediment—Analyses from 2011 and 2012 coring campaigns, Cache Creek Settling Basin, Yolo County, California

Science.gov (United States)

Arias, Michelle R.; Alpers, Charles N.; Marvin-DiPasquale, Mark C.; Fuller, Christopher C.; Agee, Jennifer L.; Sneed, Michelle; Morita, Andrew Y.; Salas, Antonia

2017-10-31

Cache Creek Settling Basin was constructed in 1937 to trap sediment from Cache Creek before delivery to the Yolo Bypass, a flood conveyance for the Sacramento River system that is tributary to the Sacramento–San Joaquin Delta. Sediment management options being considered by stakeholders in the Cache Creek Settling Basin include sediment excavation; however, that could expose sediments containing elevated mercury concentrations from historical mercury mining in the watershed. In cooperation with the California Department of Water Resources, the U.S. Geological Survey undertook sediment coring campaigns in 2011–12 (1) to describe lateral and vertical distributions of mercury concentrations in deposits of sediment in the Cache Creek Settling Basin and (2) to improve constraint of estimates of the rate of sediment deposition in the basin.Sediment cores were collected in the Cache Creek Settling Basin, Yolo County, California, during October 2011 at 10 locations and during August 2012 at 5 other locations. Total core depths ranged from approximately 4.6 to 13.7 meters (15 to 45 feet), with penetration to about 9.1 meters (30 feet) at most locations. Unsplit cores were logged for two geophysical parameters (gamma bulk density and magnetic susceptibility); then, selected cores were split lengthwise. One half of each core was then photographed and archived, and the other half was subsampled. Initial subsamples from the cores (20-centimeter composite samples from five predetermined depths in each profile) were analyzed for total mercury, methylmercury, total reduced sulfur, iron speciation, organic content (as the percentage of weight loss on ignition), and grain-size distribution. Detailed follow-up subsampling (3-centimeter intervals) was done at six locations along an east-west transect in the southern part of the Cache Creek Settling Basin and at one location in the northern part of the basin for analyses of total mercury; organic content; and cesium-137, which was

Thermal-hydraulic analysis for wire-wrapped PWR cores

Energy Technology Data Exchange (ETDEWEB)

Diller, P. [General Electric Company, 3901 Castle Hayne Rd., Wilmington, NC 28401 (United States)], E-mail: pdiller@gmail.com; Todreas, N. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)], E-mail: todreas@mit.edu; Hejzlar, P. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

2009-08-15

This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH{sub 1.6} (referred to as U-ZrH{sub 1.6}) or UO{sub 2} fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA). The thermal-hydraulic performance of U-ZrH{sub 1.6} and UO{sub 2} were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.

Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

Energy Technology Data Exchange (ETDEWEB)

Parish, T.A.

1995-03-02

This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

Application of the SPH method in nodal diffusion analyses of SFR cores

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety; Mikityuk, K. [Paul Scherrer Institut, Villigen (Switzerland)

2016-07-01

The current study investigated the potential of the SPH method, applied to correct the few-group XS produced by Serpent, to further improve the accuracy of the nodal diffusion solutions. The procedure for the generation of SPH-corrected few-group XS is presented in the paper. The performance of the SPH method was tested on a large oxide SFR core from the OECD/NEA SFR benchmark. The reference SFR core was modeled with the DYN3D and PARCS nodal diffusion codes using the SPH-corrected few-group XS generated by Serpent. The nodal diffusion results obtained with and without SPH correction were compared to the reference full-core Serpent MC solution. It was demonstrated that the application of the SPH method improves the accuracy of the nodal diffusion solutions, particularly for the rodded core state.

34 CFR 300.18 - Highly qualified special education teachers.

Science.gov (United States)

2010-07-01

... 34 Education 2 2010-07-01 2010-07-01 false Highly qualified special education teachers. 300.18... SPECIAL EDUCATION AND REHABILITATIVE SERVICES, DEPARTMENT OF EDUCATION ASSISTANCE TO STATES FOR THE... special education teachers. (a) Requirements for special education teachers teaching core academic...

A Simple, Reliable Precision Time Analyser

Energy Technology Data Exchange (ETDEWEB)

Joshi, B. V.; Nargundkar, V. R.; Subbarao, K.; Kamath, M. S.; Eligar, S. K. [Atomic Energy Establishment Trombay, Bombay (India)

1966-06-15

A 30-channel time analyser is described. The time analyser was designed and built for pulsed neutron research but can be applied to other uses. Most of the logic is performed by means of ferrite memory core and transistor switching circuits. This leads to great versatility, low power consumption, extreme reliability and low cost. The analyser described provides channel Widths from 10 {mu}s to 10 ms; arbitrarily wider channels are easily obtainable. It can handle counting rates up to 2000 counts/min in each channel with less than 1% dead time loss. There is a provision for an initial delay equal to 100 channel widths. An input pulse de-randomizer unit using tunnel diodes ensures exactly equal channel widths. A brief description of the principles involved in core switching circuitry is given. The core-transistor transfer loop is compared with the usual core-diode loops and is shown to be more versatile and better adapted to the making of a time analyser. The circuits derived from the basic loop are described. These include the scale of ten, the frequency dividers and the delay generator. The current drivers developed for driving the cores are described. The crystal-controlled clock which controls the width of the time channels and synchronizes the operation of the various circuits is described. The detector pulse derandomizer unit using tunnel diodes is described. The scheme of the time analyser is then described showing how the various circuits can be integrated together to form a versatile time analyser. (author)

Nuclear reactor core modelling in multifunctional simulators

International Nuclear Information System (INIS)

Puska, E.K.

1999-01-01

The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

Nuclear reactor core modelling in multifunctional simulators

Energy Technology Data Exchange (ETDEWEB)

Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

1999-06-01

The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

International Nuclear Information System (INIS)

Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

1996-09-01

The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations

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)

An integrated software system for core design and safety analyses: Cascade-3D

International Nuclear Information System (INIS)

Wan De Velde, A.; Finnemann, H.; Hahn, T.; Merk, S.

1999-01-01

The new Siemens program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of the most advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management. (authors)

Development of the temperature field at the WWER-440 core outlet monitoring system and application of the data analyses methods

International Nuclear Information System (INIS)

Spasova, V.; Georgieva, N.; Haralampieva, Tz.

2001-01-01

On-line internal reactor monitoring by 216 thermal couples, located at the reactor core outlet, is carried out during power operation of WWER-440 Units 1 and 2 at Kozloduy NPP. Automatic monitoring of technology process is performed by IB-500MA, which collects and performs initial data processing (discrediting and conversion of analogue signals into digital mode). The paper also presents the results and analyses of power distribution monitoring during the past 21-th and current 22-th fuel cycle at Kozloduy NPP, Unit 1 by using archiving system capacity and related software. The possibility to perform operational assessment and analysis of power distribution in the reactor core in each point of the fuel cycle is checked by comparison of the neutron-physical calculation results with reactor coolant system parameters. Paper shows that the processing and analysis of accumulated significant amount of data in the archive files increases accuracy and reliability of power distribution monitoring in the reactor core in each moment of the fuel cycle of WWER-440 reactors at Kozloduy NPP

Under sodium reliability tests on core components and in-core instrumentation

International Nuclear Information System (INIS)

Ruppert, E.; Stehle, H.; Vinzens, K.

1977-01-01

A sodium test facility for fast breeder core components (AKB), built by INTERATOM at Bensberg, has been operating since 1971 to test fuel dummies and blanket elements as well as absorber elements under simulated normal and extreme reactor conditions. Individual full-scale fuel or blanket elements and arrays of seven elements, modelling a section of the SNR-300 reactor core, have been tested under a wide range of sodium mass flow and isothermal test conditions up to 925K as well as under cyclic changed temperature transients. Besides endurance testing of the core components a special sodium and high-temperature instrumentation is provided to investigate thermohydraulic and vibrational behaviour of the test objects. During all test periods the main subassembly characteristics could be reproduced and the reliability of the instrumentation could be proven. (orig.) [de

Post-facta Analyses of Fukushima Accident and Lessons Learned

Energy Technology Data Exchange (ETDEWEB)

Tanabe, Fumiya [Sociotechnical Systems Safety Research Institute, Ichige (Japan)

2014-08-15

Independent analyses have been performed of the core melt behavior of the Unit 1, Unit 2 and Unit 3 reactors of Fukushima Daiichi Nuclear Power Station on 11-15 March 2011. The analyses are based on a phenomenological methodology with measured data investigation and a simple physical model calculation. Estimated are time variation of core water level, core material temperature and hydrogen generation rate. The analyses have revealed characteristics of accident process of each reactor. In the case of Unit 2 reactor, the calculated result suggests little hydrogen generation because of no steam generation in the core for zirconium-steam reaction during fuel damage process. It could be the reason of no hydrogen explosion in the Unit 2 reactor building. Analyses have been performed also on the core material behavior in another chaotic period of 19-31 March 2011, and it resulted in a re-melt hypothesis that core material in each reactor should have melted again due to shortage of cooling water. The hypothesis is consistent with many observed features of radioactive materials dispersion into the environment.

A Core Knowledge Architecture of Visual Working Memory

Science.gov (United States)

Wood, Justin N.

2011-01-01

Visual working memory (VWM) is widely thought to contain specialized buffers for retaining spatial and object information: a "spatial-object architecture." However, studies of adults, infants, and nonhuman animals show that visual cognition builds on core knowledge systems that retain more specialized representations: (1) spatiotemporal…

Special analyses reveal coke-deposit structure

International Nuclear Information System (INIS)

Albright, L.F.

1988-01-01

A scanning electron microscope (SEM) and an energy dispersive X-ray analyzer (EDAX) have been used to obtain information that clarifies the three mechanisms of coke formation in ethylene furnaces, and to analyze the metal condition at the exit of furnace. The results can be used to examine furnace operations and develop improved ethylene plant practices. In this first of four articles on the analyses of coke and metal samples, the coking mechanisms and coke deposits in a section of tube from an actual ethylene furnace (Furnace A) from a plant on the Texas Gulf Coast are discussed. The second articles in the series will analyze the condition of the tube metal in the same furnace. To show how coke deposition and metal condition dependent on the operating parameters of an ethylene furnace, the third article in the series will show the coke deposition in a Texas Gulf Coast furnace tube (Furnace B) that operated at shorter residence time. The fourth article discusses the metal condition in that furnace. Some recommendations, based on the analyses and findings, are offered in the fourth article that could help extend the life of ethylene furnace tubes, and also improve overall ethylene plant operations

Core characteristics on a hybrid type fast reactor system combined with proton accelerator

International Nuclear Information System (INIS)

Kowata, Yasuki; Otsubo, Akira

1997-06-01

In our study on a hybrid fast reactor system, we have investigated it from the view point of transmutation ability of trans-uranium (TRU) nuclide making the most effective use of special features (controllability, hard neutron spectrum) of the system. It is proved that a proton beam is superior in generation of neutrons compared with an electron beam. Therefore a proton accelerator using spallation reaction with a target nucleus has an advantage to transmutation of TRU than an electron one. A fast reactor is expected to primarily have a merit that the reactor can be operated for a long term without employment of highly enriched plutonium fuel by using external neutron source such as the proton accelerator. Namely, the system has a desirable characteristic of being possible to self-sustained fissile plutonium. Consequently in the present report, core characteristics of the system were roughly studied by analyses using 2D-BURN code. The possibility of self-sustained fuel was investigated from the burnup and neutronic calculation in a cylindrical core with 300w/cc of power density without considering a target material region for the accelerator. For a reference core of which the height and the radius are both 100 cm, there is a fair prospect that a long term reactor operation is possible with subsequent refueling of natural uranium, if the medium enriched (around 10wt%) uranium or plutonium fuels are fully loaded in the initial core. More precise analyses will be planed in a later fiscal year. (author)

Mining microsatellites in the peach genome: development of new long-core SSR markers for genetic analyses in five Prunus species.

Science.gov (United States)

Dettori, Maria Teresa; Micali, Sabrina; Giovinazzi, Jessica; Scalabrin, Simone; Verde, Ignazio; Cipriani, Guido

2015-01-01

A wide inventory of molecular markers is nowadays available for individual fingerprinting. Microsatellites, or simple sequence repeats (SSRs), play a relevant role due to their relatively ease of use, their abundance in the plant genomes, and their co-dominant nature, together with the availability of primer sequences in many important agricultural crops. Microsatellites with long-core motifs are more easily scored and were adopted long ago in human genetics but they were developed only in few crops, and Prunus species are not among them. In the present work the peach whole-genome sequence was used to select 216 SSRs containing long-core motifs with tri-, tetra- and penta-nucleotide repeats. Microsatellite primer pairs were designed and tested for polymorphism in the five diploid Prunus species of economic relevance (almond, apricot, Japanese plum, peach and sweet cherry). A set of 26 microsatellite markers covering all the eight chromosomes, was also selected and used in the molecular characterization, population genetics and structure analyses of a representative sample of the five diploid Prunus species, assessing their transportability and effectiveness. The combined probability of identity between two random individuals for the whole set of 26 SSRs was quite low, ranging from 2.30 × 10(-7) in peach to 9.48 × 10(-10) in almond, confirming the usefulness of the proposed set for fingerprinting analyses in Prunus species.

Analysing Student Performance Using Sparse Data of Core Bachelor Courses

Science.gov (United States)

Saarela, Mirka; Karkkainen, Tommi

2015-01-01

Curricula for Computer Science (CS) degrees are characterized by the strong occupational orientation of the discipline. In the BSc degree structure, with clearly separate CS core studies, the learning skills for these and other required courses may vary a lot, which is shown in students' overall performance. To analyze this situation, we apply…

Special Needs: A Philosophical Analysis

Science.gov (United States)

Vehmas, Simo

2010-01-01

This paper attempts to illuminate a central concept and idea in special education discourse, namely, "special needs". It analyses philosophically what needs are and on what grounds they are defined as "special" or "exceptional". It also discusses whether sorting needs into ordinary and special is discriminatory. It is argued that individualistic…

TMI-2 core boring machine

International Nuclear Information System (INIS)

Croft, K.M.; Helbert, H.J.; Laney, W.M.

1986-01-01

An important and essential aspect of the TMI-2 defueling effort is to determine what occurred in the core region during the accident. Remote cameras and probes only portray a portion of the overall picture. What lies beneath the rubble bed and solidified sublayer is, as yet, unknown. This paper discusses the TMI-2 Core Boring Machine, which has been developed to drill into the damaged core of the TMI-2 reactor and extract stratified samples of the core. This machine, its unique support structure, positioning and leveling systems, and specially designed drill bits, combine to provide a unique mechanical system. In addition, the machine is controlled by a microprocessor; which actually controls the drilling operation, allowing relatively inexperienced operators to drill the core samples. A data acquisition system is data integral with the controlling system and collects data relative to system conditions and monitored parameters during drilling. Data obtained during the actual drilling operations are collected in a data base which will be used for actual mapping of the core region, identifying materials and stratification levels that are present

Heterogeneous gas core reactor

International Nuclear Information System (INIS)

Han, K.I.

1977-01-01

Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

SEA LEVEL AND PALAEOCLIMATIC CHANGES IN THE SOUTH AND MIDDLE CASPIAN SEA REGION SINCE THE LATEGLACIAL FROM PALYNOLOGICAL ANALYSES OF MARINE SEDIMENT CORES

Directory of Open Access Journals (Sweden)

Suzanne Leroy

2010-01-01

Full Text Available A review of pollen, spores, non-pollen palynomorphs and dinocyst analyses made in the last two decades is proposed here. Building on spare palynological analyses before 1990, a series of new projects have allowed taking cores in the deeper parts of the Caspian Sea, hence providing access to low-stand sediment. However, still nowadays no complete record exists for the Holocene. The first steps towards quantification of the palynological spectra have been taken. Some of the most urgent problems to solve are the uncertainties related to radiocarbon dating, which are especially acute in the Caspian Sea.

JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Tabuchi, Shiro; Aoyama, Takafumi; Nagasaki, Hideaki; Kato, Yuichi

1998-12-01

The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, extensive data were accumulated from the core characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database. The code system MAGI has been developed and used for core management of JOYO MK-II, and the core characteristics and the irradiation test conditions were calculated using MAGI on the basis of three dimensional diffusion theory with seven neutron energy groups. The core management data include extensive data, which were recorded on CD-ROM for user convenience. The data are specifications and configurations of the core, and for about 300 driver fuel subassemblies and about 60 uninstrumented irradiation subassemblies are core composition before and after irradiation, neutron flux, neutron fluences, fuel and control rod burn-up, and temperature and power distributions. MK-II core characteristics and test conditions were stored in the database for post analysis. Core characteristics data include excess reactivities, control rod worths, and reactivity coefficients, e.g., temperature, power and burn-up. Test conditions include both measured and calculated data for irradiation conditions. (author)

Reactor core structure

International Nuclear Information System (INIS)

Higashinakagawa, Emiko; Sato, Kanemitsu.

1992-01-01

Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

Solving the uncommon reactor core neutronics problems

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.

1980-01-01

The common reactor core neutronics problems have fundamental neutron space, energy spectrum solutions. Typically the most positive eigenvalue is associated with an all-positive flux for the pseudo-steady-state condition (k/sub eff/), or the critical state is to be effected by selective adjustment of some variable such as the fuel concentration. With sophistication in reactor analysis has come the demand for solutions of other, uncommon neutronics problems. Importance functionss are needed for sensitivity and uncertainty analyses, as for ratios of intergral reaction rates such as the fuel conversion (breeding) ratio. The dominant higher harmonic solution is needed in stability analysis. Typically the desired neutronics solution must contain negative values to qualify as a higher harmonic or to satisfy a fixed source containing negative values. Both regular and adjoint solutions are of interest as are special integrals of the solutions to support analysis

Overview of core disruptive accidents

International Nuclear Information System (INIS)

Marchaterre, J.F.

1977-01-01

An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The methods are used to analyze core-disruptive accidents from initiating event to complete core disruption, the effects of the accident on reactor structures and the resulting radiological consequences are described

Opportunities and Challenges of Linking Scientific Core Samples to the Geoscience Data Ecosystem

Science.gov (United States)

Noren, A. J.

2016-12-01

Core samples generated in scientific drilling and coring are critical for the advancement of the Earth Sciences. The scientific themes enabled by analysis of these samples are diverse, and include plate tectonics, ocean circulation, Earth-life system interactions (paleoclimate, paleobiology, paleoanthropology), Critical Zone processes, geothermal systems, deep biosphere, and many others, and substantial resources are invested in their collection and analysis. Linking core samples to researchers, datasets, publications, and funding agencies through registration of globally unique identifiers such as International Geo Sample Numbers (IGSNs) offers great potential for advancing several frontiers. These include maximizing sample discoverability, access, reuse, and return on investment; a means for credit to researchers; and documentation of project outputs to funding agencies. Thousands of kilometers of core samples and billions of derivative subsamples have been generated through thousands of investigators' projects, yet the vast majority of these samples are curated at only a small number of facilities. These numbers, combined with the substantial similarity in sample types, make core samples a compelling target for IGSN implementation. However, differences between core sample communities and other geoscience disciplines continue to create barriers to implementation. Core samples involve parent-child relationships spanning 8 or more generations, an exponential increase in sample numbers between levels in the hierarchy, concepts related to depth/position in the sample, requirements for associating data derived from core scanning and lithologic description with data derived from subsample analysis, and publications based on tens of thousands of co-registered scan data points and thousands of analyses of subsamples. These characteristics require specialized resources for accurate and consistent assignment of IGSNs, and a community of practice to establish norms

Organizational Models for Non-Core Processes Management: A Classification Framework

Directory of Open Access Journals (Sweden)

Alberto F. De Toni

2012-12-01

The framework enables the identification and the explanation of the main advantages and disadvantages of each strategy and to highlight how a company should coherently choose an organizational model on the basis of: (a the specialization/complexity of the non‐core processes, (b the focus on core processes, (c its inclination towards know‐how outsourcing, and (d the desired level of autonomy in the management of non‐core processes.

Using a Core Vocabulary Intervention to Improve Communication of Students Who Use Augmentative and Alternative Communication (AAC)

Science.gov (United States)

Riccelli-Sherman, Angela

2017-01-01

This study measured the impact of core vocabulary selection and core vocabulary use on overall communication effectiveness and literacy. In this study, 30 kindergarten special education students, both male and female, who were enrolled in the Developmental Kindergarten program (a self-contained special education classroom) and Inclusive…

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)

Palaeohydrology of the Southwest Yukon Territory, Canada, based on multiproxy analyses of lake sediment cores from a depth transect

Science.gov (United States)

Anderson, L.; Abbott, M.B.; Finney, B.P.; Edwards, M.E.

2005-01-01

Lake-level variations at Marcella Lake, a small, hydrologically closed lake in the southwestern Yukon Territory, document changes in effective moisture since the early Holocene. Former water levels, driven by regional palaeohydrology, were reconstructed by multiproxy analyses of sediment cores from four sites spanning shallow to deep water. Marcella Lake today is thermally stratified, being protected from wind by its position in a depression. It is alkaline and undergoes bio-induced calcification. Relative accumulations of calcium carbonate and organic matter at the sediment-water interface depend on the location of the depositional site relative to the thermocline. We relate lake-level fluctuations to down-core stratigraphic variations in composition, geochemistry, sedimentary structures and to the occurrence of unconformities in four cores based on observations of modern limnology and sedimentation processes. Twenty-four AMS radiocarbon dates on macrofossils and pollen provide the lake-level chronology. Prior to 10 000 cal. BP water levels were low, but then they rose to 3 to 4 m below modern levels. Between 7500 and 5000 cal. BP water levels were 5 to 6 m below modern but rose by 4000 cal. BP. Between 4000 and 2000 cal. BP they were higher than modern. During the last 2000 years, water levels were either near or 1 to 2 m below modern levels. Marcella Lake water-level fluctuations correspond with previously documented palaeoenvironmental and palaeoclimatic changes and provide new, independent effective moisture information. The improved geochronology and quantitative water-level estimates are a framework for more detailed studies in the southwest Yukon. ?? 2005 Edward Arnold (Publishers) Ltd.

Pollutant plume delineation from tree core sampling using standardized ranks

DEFF Research Database (Denmark)

Wahyudi, Agung; Bogaert, Patrick; Trapp, Stefan

2012-01-01

There are currently contradicting results in the literature about the way chloroethene (CE) concentrations from tree core sampling correlate with those from groundwater measurements. This paper addresses this issue by focusing on groundwater and tree core datasets in CE contaminated site, Czech...... Republic. Preliminary analyses revealed strongly and positively skewed distributions for the tree core dataset, with an intra-tree variability accounting for more than 80% of the total variability, while the spatial analyses based on variograms indicated no obvious spatial pattern for CE concentration...... groundwater and tree core measurements. Nonetheless, tree core sampling and analysis proved to be a quick and inexpensive semi-quantitative method and a useful tool....

Hypothetical core disruptive accident analysis of a 2000 MWsub(e) liquid metal cooled fast breeder reactor

International Nuclear Information System (INIS)

Struwe, D.

1977-12-01

A structural phase diagram for hypothetical core disruptive accidents (HCDA) has been developed based on a variety of analyses for different LMFBR's. The intention was to identify the strategic phases of HCDA's important with regard to safety aspects of the plant. These phases are investigated in detail for a 2,000 MWsub(e) LMFBR (SNR-2,000). Characteristic data of SNR-2,000 are discussed concerning their influence on safety analysis. Reasons for the choice of model parameters for special phenomena as fuel coolant interaction, fuel pin failure mechanisms and sodium voiding are given. The results of calculations with CAPRI-2, HOPE and KADIS are analyzed for possibilities to enter energetic core disassembly with consequences, making power values below 2,000 MWsub(e) necessary. Investigation of these results shows that the expected consequences do not lead to design requirements, restricting the magnitude of the electrical power output of LMFBR's to values below 2,000 MWsub(e). Therefore, consequences of HCDA's are principal not expected to limit the feasibility of conventional core design of this order of magnitude. (orig.) [de

The core regulatory network of the abscisic acid pathway in banana: genome-wide identification and expression analyses during development, ripening, and abiotic stress.

Science.gov (United States)

Hu, Wei; Yan, Yan; Shi, Haitao; Liu, Juhua; Miao, Hongxia; Tie, Weiwei; Ding, Zehong; Ding, XuPo; Wu, Chunlai; Liu, Yang; Wang, Jiashui; Xu, Biyu; Jin, Zhiqiang

2017-08-29

Abscisic acid (ABA) signaling plays a crucial role in developmental and environmental adaptation processes of plants. However, the PYL-PP2C-SnRK2 families that function as the core components of ABA signaling are not well understood in banana. In the present study, 24 PYL, 87 PP2C, and 11 SnRK2 genes were identified from banana, which was further supported by evolutionary relationships, conserved motif and gene structure analyses. The comprehensive transcriptomic analyses showed that banana PYL-PP2C-SnRK2 genes are involved in tissue development, fruit development and ripening, and response to abiotic stress in two cultivated varieties. Moreover, comparative expression analyses of PYL-PP2C-SnRK2 genes between BaXi Jiao (BX) and Fen Jiao (FJ) revealed that PYL-PP2C-SnRK2-mediated ABA signaling might positively regulate banana fruit ripening and tolerance to cold, salt, and osmotic stresses. Finally, interaction networks and co-expression assays demonstrated that the core components of ABA signaling were more active in FJ than in BX in response to abiotic stress, further supporting the crucial role of the genes in tolerance to abiotic stress in banana. This study provides new insights into the complicated transcriptional control of PYL-PP2C-SnRK2 genes, improves the understanding of PYL-PP2C-SnRK2-mediated ABA signaling in the regulation of fruit development, ripening, and response to abiotic stress, and identifies some candidate genes for genetic improvement of banana.

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

Possibilities of instrumental neutron activation and X-ray fluorescence analyses of sedimentary-magmatic metamorphosed rocks from deep borehole drill cores

International Nuclear Information System (INIS)

Gurevich, A.L.; Drynkin, V.I.; Lejpunskaya, D.I.

1977-01-01

The possibilities for instrumental neutron-activation and X-ray fluorescence analyses of rocks of metamorphized sedimentary magmatic complexes have been studied with the aid of deep-hole core. The principal characteristics of the conditions of irradiation and of sample measurement ensuring the determination of the content of 26 elements are presented. The use of X-ray fluorescence analysis enables one to determine additionally the content of stron-tium and niobium. Standard specimens of the composition of rocks and complex reference compounds based on phenol formaldehyde resins are used as metrolo.o.ical auxiliaries in the calibration system and in evaluating the correctness of the techniques of instrumental neutron activation and fluorescence analysis. The ranges of the contents to be determined, the sensitivity and relative standard deviation are given. The contribution from the nonuniformity of the specimens to the summary error of element determination is estimated. It is shown that the accuracy and error of analyses are within the allowable range

Analyses of PWR boron dilution consequences with the Arrotta code

International Nuclear Information System (INIS)

Johanson, E.; Cheng, H.W.; Sehgal, B.R.

1998-03-01

During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced

National Sexuality Education Standards: Core Content and Skills, K-12. A Special Publication of the Journal of School Health. Special Report

Science.gov (United States)

American School Health Association (NJ1), 2012

2012-01-01

The goal of this paper, "National Sexuality Education Standards: Core Content and Skills, K-12," is to provide clear, consistent and straightforward guidance on the "essential minimum, core content" for sexuality education that is developmentally and age-appropriate for students in grades K-12. The development of these standards is a result of an…

Studies on WWER core diagnostics

International Nuclear Information System (INIS)

Lunin, G.L.; Mitin, V.I.; Bulavin, V.V.

1987-01-01

The reliability and safety of nuclear power plants have decisive meaning under the situation that nuclear power generation steadily increases, and among various measures aiming at ensuring the reliability and safety in the operation of nuclear power plants, the countermeasures for protecting reactor core, main process equipment and high pressure circuits from damage have the important role, and the monitoring of condition and the organization of forecast, which are carried out continuously or periodically during the operation of nuclear power stations using the diagnostic expert system specially developed for the purpose, are included in them. Such monitoring enables the early detection of mechanical damage, increase of vibration, defects caused during operation and so on in reactor cores and primary and secondary circuits, and the continuous watching of defect developments. Also boiling in a core is detected, the place of abnormality occurrence is identified, and the intensity and characteristics of boiling are determined, thus the occurrence of dangerous condition is prevented. The developments of an in-core monitoring system and noise diagnostic systems are reported. (Kako, I.)

Transmission in Optically Transparent Core Networks

Science.gov (United States)

Kilper, Dan; Jensen, Rich; Petermann, Klaus; Karasek, Miroslav

2007-03-01

Call for Papers: Transmission in Optically Transparent Core Networks Guest Feature Editors Dan Kilper and Rich Jensen, Coordinating Associate Editors Klaus Petermann and Miroslav Karasek, Guest Feature Editors Submission deadline: 15 June 2007 Optically transparent networks in which optical transport signals are routed uninterrupted through multiple nodes have long been viewed as an important evolutionary step in fiber optic communications. More than a decade of research and development on transparent network technologies together with the requisite traffic growth has culminated in the recent deployment of commercial optically transparent systems. Although many of the traditional research goals of optical transmission remain important, optical transparency introduces new challenges. Greater emphasis is placed on system efficiency and control. The goal of minimizing signal terminations, which has been pursued through increasing reach and channel capacity, also can be realized through wavelength routing techniques. Rather than bounding system operation by rigid engineering rules, the physical layer is controlled and managed by automation tools. Many static signal impairments become dynamic due to network reconfiguration and transient fault events. Recently new directions in transmission research have emerged to address transparent networking problems. This special issue of the Journal of Optical Networking will examine the technologies and theory underpinning transmission in optically transparent core networks, including both metropolitan and long haul systems. Scope of Submission The special issue editors are soliciting high-quality original research papers related to transmission in optically transparent core networks. Although this does not include edge networks such as access or enterprise networks, core networks that have access capabilities will be considered in scope as will topics related to the interworking between core and edge networks. The core network

Bench top and portable mineral analysers, borehole core analysers and in situ borehole logging

International Nuclear Information System (INIS)

Howarth, W.J.; Watt, J.S.

1982-01-01

Bench top and portable mineral analysers are usually based on balanced filter techniques using scintillation detectors or on low resolution proportional detectors. The application of radioisotope x-ray techniques to in situ borehole logging is increasing, and is particularly suited for logging for tin and higher atomic number elements

Design of radiation shields in nuclear reactor core

International Nuclear Information System (INIS)

Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

2008-01-01

This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

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.

Applying CLSM to increment core surfaces for histometric analyses: A novel advance in quantitative wood anatomy

OpenAIRE

Wei Liang; Ingo Heinrich; Gerhard Helle; I. Dorado Liñán; T. Heinken

2013-01-01

A novel procedure has been developed to conduct cell structure measurements on increment core samples of conifers. The procedure combines readily available hardware and software equipment. The essential part of the procedure is the application of a confocal laser scanning microscope (CLSM) which captures images directly from increment cores surfaced with the advanced WSL core-microtome. Cell wall and lumen are displayed with a strong contrast due to the monochrome black and green nature of th...

Attaining Success for Beginning Special Education Teachers.

Science.gov (United States)

McCabe, Marjorie; And Others

1993-01-01

Three case studies are presented that highlight problem scenarios relating to beginning special education intern teachers and explain how the teachers attained success. The cases focus on classroom management, adaptation of the core curriculum, and knowledge of instructional practices. (JDD)

System for Automated Geoscientific Analyses (SAGA) v. 2.1.4

Science.gov (United States)

Conrad, O.; Bechtel, B.; Bock, M.; Dietrich, H.; Fischer, E.; Gerlitz, L.; Wehberg, J.; Wichmann, V.; Böhner, J.

2015-07-01

The System for Automated Geoscientific Analyses (SAGA) is an open source geographic information system (GIS), mainly licensed under the GNU General Public License. Since its first release in 2004, SAGA has rapidly developed from a specialized tool for digital terrain analysis to a comprehensive and globally established GIS platform for scientific analysis and modeling. SAGA is coded in C++ in an object oriented design and runs under several operating systems including Windows and Linux. Key functional features of the modular software architecture comprise an application programming interface for the development and implementation of new geoscientific methods, a user friendly graphical user interface with many visualization options, a command line interpreter, and interfaces to interpreted languages like R and Python. The current version 2.1.4 offers more than 600 tools, which are implemented in dynamically loadable libraries or shared objects and represent the broad scopes of SAGA in numerous fields of geoscientific endeavor and beyond. In this paper, we inform about the system's architecture, functionality, and its current state of development and implementation. Furthermore, we highlight the wide spectrum of scientific applications of SAGA in a review of published studies, with special emphasis on the core application areas digital terrain analysis, geomorphology, soil science, climatology and meteorology, as well as remote sensing.

Thermal hydraulic model validation for HOR mixed core fuel management

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Vries, J.W. de; Leege, P.F.A. de

1997-01-01

A thermal-hydraulic core management model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type university research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and for in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, and model validation are discussed. (author)

RAtional Mapping (RAM) of in-core data

International Nuclear Information System (INIS)

Bonalumi, R.A.; Kherani, N.P.

1983-01-01

The paper describes and demonstrates a unique processing of in-core flux detector data, such that the detailed in-core power distribution can be derived with great accuracy by combining a specially 'smoothed-out' set of in-core data with neutron diffusion theory. RAM is designed in such a way that erratic detector signals are recognized very efficiently and can be eliminated from the experimental data set: this is achieved by modal expansion of the difference between theoretical fluxes and experimental fluxes at the detector sites. Sensitivity studies have shown that RAM is quite stable, does not absorb the 'wild' detector errors in the mapping procedure and results in mapped fluxes with errors about three times smaller than would be obtained by direct interpolation of detector readings

Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

Energy Technology Data Exchange (ETDEWEB)

Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-10-01

The Transient Reactor Test Facility (TREAT), located at Idaho National Laboratory (INL), is a test facility designed to evaluate the performance of reactor fuels and materials under transient accident conditions. The facility, an air-cooled, graphite-moderated reactor designed to utilize fuel containing high-enriched uranium (HEU), has been in non-operational standby status since 1994. Currently, in support of the missions of the Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program, a new core design is being developed for TREAT that will utilize low-enriched uranium (LEU). The primary objective of this conversion effort is to design an LEU core that is capable of meeting the performance characteristics of the existing HEU core. Minimal, if any, changes are anticipated for the supporting systems (e.g. reactor trip system, filtration/cooling system, etc.); therefore, the LEU core must also be able to function with the existing supporting systems, and must also satisfy acceptable safety limits. In support of the LEU conversion effort, a range of ancillary safety analyses are required to evaluate the LEU core operation relative to that of the existing facility. These analyses cover neutronics, shielding, and thermal hydraulic topics that have been identified as having the potential to have reduced safety margins due to conversion to LEU fuel, or are required to support the required safety analyses documentation. The majority of these ancillary tasks have been identified in [1] and [2]. The purpose of this report is to document the ancillary safety analyses that have been performed at Argonne National Laboratory during the early stages of the LEU design effort, and to describe ongoing and anticipated analyses. For all analyses presented in this report, methodologies are utilized that are consistent with, or improved from, those used in analyses for the HEU Final Safety Analysis

Development of JOYO MK-II core characteristics database

International Nuclear Information System (INIS)

Tabuchi, Shiro; Aoyama, Takafumi

2000-01-01

The MK-II core of the experimental fast reactor JOYO served as the irradiation bed for testing fuels and materials for FBR development since 1982 for 15 years. During the MK-II operation, extensive data were accumulated from the core management calculations and characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database recorded on CD-ROM for user convenience. The calculated core management data are the text style data. The 'Configuration Data' include the history of the fuel exchange and core arrangement for each cycle. The Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of about 300 fuel subassemblies, and 60 irradiation subassemblies. The 'Output Data' include the neutron fluxes, gamma fluxes, power density, linear heat rates, coolant and fuel temperature distributions of each core position at the beginning and end of each cycle. The measured core characteristics data, such as the excess reactivity, control rod worths, temperature coefficient, power coefficient, and burn-up coefficient are also included along with the measurement conditions. (J.P.N.)

Preliminary core design of IRIS-50

International Nuclear Information System (INIS)

Petrovic, Bojan; Franceschini, Fausto

2009-01-01

IRIS-50 is a small, 50 MWe, advanced PWR with integral primary system. It evolved employing the same design principles as the well known medium size (335 MWe) IRIS. These principles include the 'safety-by-design' philosophy, simple and robust design, and deployment flexibility. The 50 MWe design addresses the needs of specific applications (e.g., power generation in small regional grids, water desalination and biodiesel production at remote locations, autonomous power source for special applications, etc.). Such applications may favor or even require longer refueling cycles, or may have some other specific requirements. Impact of these requirements on the core design and refueling strategy is discussed in the paper. Trade-off between the cycle length and other relevant parameters is addressed. A preliminary core design is presented, together with the core main reactor physics performance parameters. (author)

Status of reactor core design code system in COSINE code package

Energy Technology Data Exchange (ETDEWEB)

Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

2014-07-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Status of reactor core design code system in COSINE code package

International Nuclear Information System (INIS)

Chen, Y.; Yu, H.; Liu, Z.

2014-01-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Measurement of 89Y(n,2n) spectral averaged cross section in LR-0 special core reactor spectrum

Science.gov (United States)

Košťál, Michal; Losa, Evžen; Baroň, Petr; Šolc, Jaroslav; Švadlenková, Marie; Koleška, Michal; Mareček, Martin; Uhlíř, Jan

2017-12-01

The present paper describes reaction rate measurement of 89Y(n,2n)88Y in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor and compares this value with results of simulation. The reaction rate is derived from the measurement of activity of 88Y using gamma-ray spectrometry of irradiated Y2O3 sample. The resulting cross section value averaged in spectrum is 43.9 ± 1.5 μb, averaged in the 235U spectrum is 0.172 ± 0.006 mb. This cross-section is important as it is used as high energy neutron monitor and is therefore included in the International Reactor Dosimetry and Fusion File. Calculations of reaction rates were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. The agreement with uranium description by CIELO library is very good, while in ENDF/B-VII.0 description of uranium, underprediction about 10% in average can be observed.

Schumpeter's core works revisited

DEFF Research Database (Denmark)

Andersen, Esben Sloth

2012-01-01

This paper organises Schumpeter’s core books in three groups: the programmatic duology,the evolutionaryeconomic duology,and the socioeconomic synthesis. By analysing these groups and their interconnections from the viewpoint of modern evolutionaryeconomics,the paper summarises resolved problems a...... and points at remaining challenges. Its analyses are based on distinctions between microevolution and macroevolution, between economic evolution and socioeconomic coevolution, and between Schumpeter’s three major evolutionary models (called Mark I, Mark II and Mark III)....

Analyses of systems availability and operator actions to support the development of severe accident procedures

International Nuclear Information System (INIS)

Lutz, R.J. Jr.; Scobel, J.H.

1989-01-01

This paper reports on traditional analyses of severe accidents, such as those presented in Probabilistic Risk Assessment (PRA) studies of nuclear power stations, that have generally been performed on the assumption that all means of cooling the reactor core are lost and that no operator actions to mitigate the consequences or progression of the severe accident are performed. The assumption to neglect the availability of safety systems and operator actions which do not prevent core melting can lead to erroneous conclusions regarding the plant severer accident profile. Recent work in severe accident management has identified the need to perform analyses which consider all systems availabilities and operator actions, irrespective of their contribution to the prevention of core melting. These new analyses indicate that the traditional analyses result in overfly pessimistic predictions of the time of core melting and the subsequent potential for recovery of core cooling prior to core melting. Additionally, since the traditional analyses do not model all of the operator actions which are prescribed, the impact of additional severe accident operator actions on the progression and consequences of the accident cannot be reliably identified. Further, the more detailed analysis can change the focus of the importance of various system to the prevention of core damage and the mitigation of severe accident consequences. Finally, the simplicity of the traditional analyses can have a considerable impact on severe accident decision making, particularly in the evaluation of alternate plant design features and the priorities for research studies

Emergency response guide-B ECCS guideline evaluation analyses for N reactor

International Nuclear Information System (INIS)

Chapman, J.C.; Callow, R.A.

1989-07-01

INEL conducted two ECCS analyses for Westinghouse Hanford. Both analyses will assist in the evaluation of proposed changes to the N Reactor Emergency Response Guide-B (ERG-B) Emergency Core System (ECCS) guideline. The analyses were a sensitivity study for reduced-ECCS flow rates and a mechanistically determined confinement steam source for a delayed-ECCS LOCA sequence. The reduced-ECCS sensitivity study established the maximum allowable reduction in ECCS flow as a function of time after core refill for a large break loss-of-coolant accident (LOCA) sequence in the N Reactor. The maximum allowable ECCS flow reduction is defined as the maximum flow reduction for which ECCS continues to provide adequate core cooling. The delayed-ECCS analysis established the liquid and steam break flows and enthalpies during the reflood of a hot core following a delayed ECCS injection LOCA sequence. A simulation of a large, hot leg manifold break with a seven-minute ECCS injection delay was used as a representative LOCA sequence. Both analyses were perform using the RELAP5/MOD2.5 transient computer code. 13 refs., 17 figs., 3 tabs

Review of pertinent thermal-hydraulic data for LMFBR core natural circulation analyses

International Nuclear Information System (INIS)

Bishop, A.A.; Coffield, R.D. Jr.; Markley, R.A.

1980-01-01

A literature review and summary of significant data is presented relative to LMFBR core natural convection cooling analysis. First, a brief review of computer codes and respective input data needs is made, significant data areas are then addressed and data for verifying the code calculations are described. Recommendations and conclusions with regard to the data are included

Fuel and Core Design Verification for Extended Power Up-rate in Ringhals Unit 3

International Nuclear Information System (INIS)

Gabrielsson, Petter; Stepniewski, Marek; Almberger, Jan

2006-01-01

Vattenfall's Westinghouse 3-loop PWR Ringhals 3 at the western coast of Sweden is scheduled for an extended power up-rate from 2783 to 3160 MWt in 2007, in the frame of the so called GREAT-project. The project will realize an up-rating initially planned and analysed back in 1995, but with a number of significant improvements outlined in this paper. For the licensing of the up-rated power level, a complete revision of the safety analyses, radiological analyses and systems verifications in FSAR is being performed by Westinghouse Electrics Belgium. The work is performed in close cooperation with Vattenfall in the areas of core calculations and input data. For more than a decade, Vattenfall has performed all core design and reload safety evaluations (RSE) for Ringhals, independent of fuel vendors and safety analysts. In GREAT all core parameters in the safety analysis checklist (SAC) used for the safety analyses are determined based upon a set of nine reference loading patterns designed by Vattenfall covering a wide range of fuel and core designs and extreme cycle-to-cycle variations. To facilitate the calculation of SAC parameters Westinghouse has provided a Reload Safety Evaluation Procedure report (RSEP) with detailed specifications for the calculation of all core parameters used in the analyses. The procedure has been automatized by Vattenfall in a set of scripts executing 3D core simulator calculations and extracting the key results. The same tools will be used in Vattenfall's future RSE for Ringhals 3. This approach is taken to obtain consistency between core designs and core calculations for the safety analyses and the cycle specific calculations, to minimize the risk for future violations of the safety analyses. (authors)

Reactivity analysis of core distortion effects in the FFTF

International Nuclear Information System (INIS)

Knutson, B.J.

1982-01-01

An improved technique for evaluating core distortion reactivity effects was developed using reactivity analyses of two core geometry models (R-Z and HEX). This technique is incorporated into a new processor code called CORDIS. The advantages of this technique over existing reactivity models are that is preserves core heterogeneity, provides a control rod insertion effect model, uses row-dependent axial shape functions, and provides a flexible and cost efficient core distortion reactivity analysis method

Rational mapping (RAM) of in-core data

International Nuclear Information System (INIS)

Bonalumi, R.A.; Kherani, N.P.

1985-01-01

A unique processing of in-core flux detector data is described and demonstrated, such that the detailed in-core power distribution can be derived with great accuracy by combining a speciall ''smoothed-out'' set of in-core data with neutron diffusion theory. Rational Mapping (RAM) is designed in such a way that erratic detector signals are recognized very efficiently and can be eliminated from the experimental data set: This is achieved by modal expansion of the difference between theoretical fluxes and experimental fluxes at the detector sites. Sensitivity studies have shown that RAM is quite stable, does not absorb the ''wild'' detector error in the mapping procedure, and results in mapped fluxes with errors about three times smaller than would be obtained by direct interpolation of detector readings. A new method is described to infer corrections to theoretical core parameters based on the difference between the RAM fluxes and the theoretical fluxes

Development of a BWR core burn-up calculation code COREBN-BWR

International Nuclear Information System (INIS)

Morimoto, Yuichi; Okumura, Keisuke

1992-05-01

In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

Reactor physics special problem in 11. ENFIR

International Nuclear Information System (INIS)

Leszczynski, Francisco

1997-01-01

In this report, the computation method and the results of the work performed of the special topic on reactor physics proposed for the 11. ENFIR is presented. MCNP 4.2 has been adopted as the only code to perform the calculations. The full core of the IPEN-MB-1 critical unit has been modelled without important approximations. The specifications given by the Organizer Commission of the Special Topic were followed. The nuclear libraries adopted were those included on the MCNPDAT package, mainly from ENDF/B-V, except indium data, not included in this package. For indium, data obtained from LANL, based on ENDF/B-VI were used. The results are: critical position of the control banks assuming simultaneous movement: percent of extraction: (49±1)% ; excess of reactivity of the core: ρ =( 3590 ±50)pcm ; total reactivity of the one control rod bank: ρ= (4000±50) pcm. The reactivity curve of the control rods is included also. (author)

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Basehore, K.L.; Todreas, N.E.

1980-08-01

Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

International Nuclear Information System (INIS)

Basehore, K.L.; Todreas, N.E.

1980-08-01

Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

RELAP5/MOD3.3 Analyses of Core Heatup Prevention Strategy During Extended Station Blackout in PWR

International Nuclear Information System (INIS)

Prosek, A.

2016-01-01

The accident at the Fukushima Dai-ichi nuclear power plant demonstrated the vulnerability of the plants on the loss of electrical power for several days, so called extended station blackout (SBO). A set of measures have been proposed and implemented in response of the accident at the Fukushima Dai-ichi nuclear power plant. The purpose of the study was to investigate the application of the deterministic safety analysis for core heatup prevention strategy of the extended SBO in pressurized water reactor, lasting 72 h. The prevention strategy selected was water injection into steam generators using turbine driven auxiliary feedwater pump (TD-AFW) or portable water injection pump. Method for assessment of the necessary pump injection flowrate is developed and presented. The necessary injection flowrate to the steam generators is determined from the calculated cumulative water mass injected by the turbine driven auxiliary feedwater pump in the analysed scenarios, when desired normal level is maintained automatically. The developed method allows assessment of the necessary injection flowrates of pump, TD-AFW or portable, for different plant configurations and number of flowrate changes. The RELAP5/MOD3.3 Patch04 computer code and input model of a two-loop pressurized water reactor is used for analyses, assuming different injection start times, flowrates and reactor coolant system losses. Three different reactor coolant system (RCS) coolant loss pathways, with corresponding leakage rate, can be expected in the pressurized water reactor (PWR) during the extended SBO: normal system leakage, reactor coolant pump seal leakage, and RCS coolant loss through letdown relief valve unless automatically isolated or until isolation is procedurally directed. Depressurization of RCS was also considered. In total, six types of RCS coolant loss scenarios were considered. Two cases were defined regarding the operation of the emergency diesel generators. Different delays of the pump

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.

Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

International Nuclear Information System (INIS)

Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

2009-01-01

The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

Progress of full MOX core design in ABWR

International Nuclear Information System (INIS)

Izutsu, S.; Sasagawa, M.; Aoyama, M.; Maruyama, H.; Suzuki, T.

2000-01-01

Full MOX ABWR core design has been made, based on the MOX design concept of 8x8 bundle configuration with a large central water rod, 40 GWd/t maximum bundle exposure, and the compatibility with 9x9 high-burnup UO 2 bundles. Core performance on shutdown margin and thermal margin of the MOX-loaded core is similar to that of UO 2 cores for the range from full UO 2 core to full MOX core. Safety analyses based on its safety parameters and MOX property have shown its conformity to the design criteria in Japan. In order to confirm the applicability of the nuclear design method to full MOX cores, Tank-type Critical Assembly (TCA) experiment data have been analyzed on criticality, power distribution and β eff /l measurements. (author)

COMSORS: A light water reactor chemical core catcher

International Nuclear Information System (INIS)

Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.

1997-01-01

The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B 2 O 3 ) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation

Separative analyses of a chromatographic column packed with a core-shell adsorbent for lithium isotope separation

International Nuclear Information System (INIS)

Sugiyama, T.; Sugura, K.; Enokida, Y.; Yamamoto, I.

2015-01-01

Lithium-6 is used as a blanket material for sufficient tritium production in DT fueled fusion reactors. A core-shell type adsorbent was proposed for lithium isotope separation by chromatography. The mass transfer model in a chromatographic column consisted of 4 steps, such as convection and dispersion in the column, transfer through liquid films, intra-particle diffusion and and adsorption or desorption at the local adsorption sites. A model was developed and concentration profiles and time variation in the column were numerically simulated. It became clear that core-shell type adsorbents with thin porous shell were saturated rapidly relatively to fully porous one and established a sharp edge of adsorption band. This is very important feature because lithium isotope separation requires long-distance development of adsorption band. The values of HETP (Height Equivalent of a Theoretical Plate) for core-shell adsorbent packed column were estimated by statistical moments of the step response curve. The value of HETP decreased with the thickness of the porous shell. A core-shell type adsorbent is, then, useful for lithium isotope separation. (authors)

Conceptual design of PFBR core

International Nuclear Information System (INIS)

Lee, S.M.; Govindarajan, S.; Indira, R.; John, T.M.; Mohanakrishnan, P.; Shankar Singh, R.; Bhoje, S.B.

1996-01-01

The design options selected for the core of the 500 MWe Prototype Fast Breeder Reactor are presented. PFBR has a conventional mixed oxide fuel core of homogeneous type with two enrichment zones for power flattening and with radial and axial blankets to make the reactor self-sustaining in fissile material. Pin diameter has been selected for minimization of fissile inventory. Considerations for the choice of number of pins per subassembly, integrated versus separate axial blankets, and other pin and subassembly parameters are discussed. As the core size is moderate, no special schemes for reducing the maximum positive sodium voiding coefficient is envisages. Two independent, diverse fast acting shutdown systems working in fail-safe mode are selected. The number of absorber rods has been minimized by choosing a layout for maximum antishadow effect. Nine control and safety rods are distributed in two rods for power flattening by differential insertion. Three Diverse Safety Rods, are also provided which are normally fully withdrawn. The optimization of layout of radial and axial shielding and adequacy of flux at detector location are also discussed. (author). 2 figs

Planning for health care transitions: results from the 2005-2006 National Survey of Children With Special Health Care Needs.

Science.gov (United States)

Lotstein, Debra S; Ghandour, Reem; Cash, Amanda; McGuire, Elizabeth; Strickland, Bonnie; Newacheck, Paul

2009-01-01

Many youth with special health care needs have difficulties transferring to adult medical care. To address this, the Maternal and Child Health Bureau has made receipt of transition services a core performance outcome for community-based systems of care for youth with special health care needs. In this article we describe the results for the transition core outcome from the 2005-2006 National Survey of Children With Special Health Care Needs. We also describe changes in the measurement strategy for this outcome since the first National Survey of Children With Special Health Care Needs in 2001. In the nationally representative, cross-sectional 2005-2006 National Survey of Children With Special Health Care Needs, parent or guardian respondents of 18198 youth with special health care needs (aged 12-17) were asked if they have had discussions with their child's health care providers about (1) future adult providers, (2) future adult health care needs, (3) changes in health insurance, and (4) encouraging their child to take responsibility for his or her care. All 4 components had to be met for the youth to meet the overall transition core outcome. Those who had not had transition discussions reported if such discussions would have been helpful. Overall, 41% of youth with special health care needs met the core performance outcome for transition. Forty-two percent had discussed shifting care to an adult provider, 62% discussed their child's adult health care needs, and 34% discussed upcoming changes in health insurance. Most (78%) respondents said that providers usually or always encourage their child to take responsibility for his or her health. Non-Hispanic black or Hispanic race/ethnicity, lower income level, not speaking English, and not having a medical home reduced the odds of meeting the transition core outcome. Current performance on the transition core outcome leaves much room for improvement. Many parents feel that having transition-related discussions with their

Geochemistry of Mariano lake-lake valley cores, McKinley County, New Mexico

International Nuclear Information System (INIS)

Leventhal, J.S.; Lichte, F.E.; Gent, C.A.

1990-01-01

The primary goal of the U.S. Geological Survey-Bureau of Indian Affairs drilling project in the Upper Jurassic Morrison Formation in McKinley County, New Mexico, was to better understand the relationship between host-rock stratigraphy and uranium mineralization. As part of this project, geochemical studies of approximately 280 samples from 8 cores and 1 outcrop were undertaken; samples from 4 cores show uranium enrichment. Geochemical relationships between samples of weathered outcrop, oxidized core, reduced (unmineralized) core, and ore-bearing core were contrasted by comparison of element abundances. Special comparative studies of sandstone and clay chemistry were made using results from x-ray diffraction, optical petrography, and chemical analysis. Results of these studies are discussed

Tank 241-AY-101 Privatization Push Mode Core Sampling and Analysis Plan

International Nuclear Information System (INIS)

TEMPLETON, A.M.

2000-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AY-101. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AY-101 required to satisfy Data Quality Objectives For RPP Privatization Phase I: Confirm Tank T Is An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO) (Nguyen 1999a), Data Quality Objectives For TWRS Privatization Phase I : Confirm Tank T Is An Appropriate Feed Source For Low-Activity Waste Feed Batch X (LAW DQO) (Nguyen 1999b), Low Activity Waste and High-Level Waste Feed Data Quality Objectives (L and H DQO) (Patello et al. 1999), and Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO) (Bloom 1996). Special instructions regarding support to the LAW and HLW DQOs are provided by Baldwin (1999). Push mode core samples will be obtained from risers 15G and 150 to provide sufficient material for the chemical analyses and tests required to satisfy these data quality objectives. The 222-S Laboratory will extrude core samples; composite the liquids and solids; perform chemical analyses on composite and segment samples; archive half-segment samples; and provide subsamples to the Process Chemistry Laboratory. The Process Chemistry Laboratory will prepare test plans and perform process tests to evaluate the behavior of the 241-AY-101 waste undergoing the retrieval and treatment scenarios defined in the applicable DQOs. Requirements for analyses of samples originating in the process tests will be documented in the corresponding test plans and are not within the scope of this SAP

Tank 241-AY-101 Privatization Push Mode Core Sampling and Analysis Plan

International Nuclear Information System (INIS)

TEMPLETON, A.M.

2000-01-01

This sampling and analysis plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for samples obtained from tank 241-AY-101. The purpose of this sampling event is to obtain information about the characteristics of the contents of 241-AY-101 required to satisfy ''Data Quality Objectives For RPP Privatization Phase I: Confirm Tank T Is An Appropriate Feed Source For High-Level Waste Feed Batch X(HLW DQO)' (Nguyen 1999a), ''Data Quality Objectives For TWRS Privatization Phase I: Confirm Tank T Is An Appropriate Feed Source For Low-Activity Waste Feed Butch X (LAW DQO) (Nguyen 1999b)'', ''Low Activity Waste and High-Level Waste Feed Data Quality Objectives (L and H DQO)'' (Patello et al. 1999), and ''Characterization Data Needs for Development, Design, and Operation of Retrieval Equipment Developed through the Data Quality Objective Process (Equipment DQO)'' (Bloom 1996). Special instructions regarding support to the LAW and HLW DQOs are provided by Baldwin (1999). Push mode core samples will be obtained from risers 15G and 150 to provide sufficient material for the chemical analyses and tests required to satisfy these data quality objectives. The 222-S Laboratory will extrude core samples; composite the liquids and solids; perform chemical analyses on composite and segment samples; archive half-segment samples; and provide sub-samples to the Process Chemistry Laboratory. The Process Chemistry Laboratory will prepare test plans and perform process tests to evaluate the behavior of the 241-AY-101 waste undergoing the retrieval and treatment scenarios defined in the applicable DQOs. Requirements for analyses of samples originating in the process tests will be documented in the corresponding test plans and are not within the scope of this SAP

Examining Secondary Special Education Teachers' Literacy Instructional Practices

Science.gov (United States)

Leko, Melinda M.; Handy, Tamara; Roberts, Carly A.

2017-01-01

This study presents findings from a survey of secondary special education teachers who teach reading. Respondents were 577 special education teachers from a large Midwestern state who completed an online or mail survey. Results based on quantitative and qualitative analyses indicate predominant foci of secondary special education teachers' reading…

Integrated CFD investigation of heat transfer enhancement using multi-tray core catcher in SFR

International Nuclear Information System (INIS)

Rakhi; Sharma, Anil Kumar; Velusamy, K.

2017-01-01

Highlights: • Heat transfer enhancement using multi-tray core catcher for SFR is investigated. • The capability of a single core collector tray is estimated. • Double and triple collector trays with innovative designs is discussed. • Provision of openings in the trays contributed to enhanced natural circulation. - Abstract: To render future SFR more robust and safe, certain BDBE have been considered in the recent years. A Core Disruptive Accident leading to a whole core meltdown scenario has gained the interest of researchers. Various design concepts and safety measures have been suggested and incorporated in design to address such a low probability scenario. A core catcher concept, in particular, has proved to be inevitable as an in-vessel core retention device in SFR for safe retention of core debris arising out after the severe accident. This study aims to analyse the cooling capability of the innovative design concept of core catcher to remove decay heat of degraded core after the accident. First, the capability of single collection tray is established and then the study is extended to two and three collection trays with different design concepts. Transient forms of governing equations of mass, momentum and energy conservations along with k-ε turbulence model are solved by finite volume based CFD solver. Boussinesq approximation is invoked to model buoyancy in sodium. The study shows that a single collection tray is capable of removing up to 20 MW decay heat load in a typical 500 MWe pool type SFR. Further, studies are carried out to improve the natural circulation of sodium around the source, in the lower plenum and to distribute core debris of the whole core to multiple collection trays. It is found that the double and triple collection trays can accommodate decay loads up to 29 MW. Provision of openings in the collection trays has proved to be effective in improving the heat transfer and sodium flow as well as in distributing the core debris to the

Science Education for Students with Special Needs

Science.gov (United States)

Villanueva, Mary Grace; Taylor, Jonte; Therrien, William; Hand, Brian

2012-01-01

Students with special needs tend to show significantly lower achievement in science than their peers. Reasons for this include severe difficulties with academic skills (i.e. reading, math and writing), behaviour problems and limited prior understanding of core concepts background knowledge. Despite this bleak picture, much is known on how to…

Low-Power Embedded DSP Core for Communication Systems

Science.gov (United States)

Tsao, Ya-Lan; Chen, Wei-Hao; Tan, Ming Hsuan; Lin, Maw-Ching; Jou, Shyh-Jye

2003-12-01

This paper proposes a parameterized digital signal processor (DSP) core for an embedded digital signal processing system designed to achieve demodulation/synchronization with better performance and flexibility. The features of this DSP core include parameterized data path, dual MAC unit, subword MAC, and optional function-specific blocks for accelerating communication system modulation operations. This DSP core also has a low-power structure, which includes the gray-code addressing mode, pipeline sharing, and advanced hardware looping. Users can select the parameters and special functional blocks based on the character of their applications and then generating a DSP core. The DSP core has been implemented via a cell-based design method using a synthesizable Verilog code with TSMC 0.35[InlineEquation not available: see fulltext.]m SPQM and 0.25[InlineEquation not available: see fulltext.]m 1P5M library. The equivalent gate count of the core area without memory is approximately 50 k. Moreover, the maximum operating frequency of a[InlineEquation not available: see fulltext.] version is 100 MHz (0.35[InlineEquation not available: see fulltext.]m) and 140 MHz (0.25[InlineEquation not available: see fulltext.]m).

German (GRS) approach to accident analysis (part I). German licensing basis for accident analyses. Applicants accident analyses in second part license for Konvoi-plants. Appendix 1. Assessor accident analyses in second part license for Konvoi-plants. Appendix 2. Reference list of DBA to be considered in the safety status analysis of a PSR. Appendix 3a. Reference list of special very rare and BDB plant conditions to be considered in the safety status analysis of a PSE. Appendix 3b

International Nuclear Information System (INIS)

Velkov, K.

2002-01-01

Appendix 1: The Safety Analysis Report (S.A.R.) is presented from 3 Handbooks - ECC Handbook (LOCA), Plant Dynamics Handbook (Transients incl. ATWS), and Core Design Handbook. The first one Conceived as Living handbook, Basis for design, catalogue of transients, specifications and licensing. Handbook contains LOCA in primary system, it contains also core damage analysis, and description of codes, description of essential plant data and code input data. The second one consists of Basis for design, commissioning, operation, and catalogue of transients, specifications and licensing, as well as specified operation, disturbed operation, incidents, non-LOCA, SS-procedures and Code description. The third book consists of Reactivity balance and reactivity coefficients, efficiency of shutdown systems. Calculation of burn up cycle, power density distribution, and critical boron concentration. Also Codes used, as SAV79A standard analysis methodology including FASER for nuclear data generation, MEDIUM and PANBOX for static and transient core calculations. Appendix 2: The three TUEV (Technical Inspection Agencies) responsible for the three individual plants of type KONVOI: TUEV Bayern for ISAR-2, TUV-Hanover for KKE, TUEV-Stuttgart for GKN-2 and GRS performed the safety assessment. TUV-Bayern for disturbance and failure of secondary heat sink without loss of coolant (failure of main heat sink, erroneous operation of valves in MS and in FW system, failure of MFW supply), long term LONOP, performance of selected SBLOCA analyses. TUV Hanover for disturbances due to failure of MCPs, short term LONOP, damages of SG tubes incl. SGTR, performance of selected LOCA analyses (blowdown phase of LBLOCA). TUV-Stuttgart for breaks and leaks in MS and FW system with and without leaks in SG tubes. GRS for ATWS, sub-cooling transients due to disturbances on secondary side, initial and boundary conditions for transients with opening of pressurizer valves with and without stuck-open, most of the

Legal Protection on IP Cores for System-on-Chip Designs

Science.gov (United States)

Kinoshita, Takahiko

The current semiconductor industry has shifted from vertical integrated model to horizontal specialization model in term of integrated circuit manufacturing. In this circumstance, IP cores as solutions for System-on-Chip (SoC) have become increasingly important for semiconductor business. This paper examines to what extent IP cores of SoC effectively can be protected by current intellectual property system including integrated circuit layout design law, patent law, design law, copyright law and unfair competition prevention act.

Modification of unsaturated polyester resins using nano-size core ...

African Journals Online (AJOL)

Modification of unsaturated polyester resins using nano-size core-shell particles. MO Munyati, PA Lovell. Abstract. No Abstract Available Journal of Science and Technology Special Edition 2004: 24-31. Full Text: EMAIL FULL TEXT EMAIL FULL TEXT · DOWNLOAD FULL TEXT DOWNLOAD FULL TEXT.

Conceptual study of advanced PWR core design

International Nuclear Information System (INIS)

Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

1997-09-01

The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

Conceptual study of advanced PWR core design

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

1997-09-01

The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

LWR-core behaviour project

International Nuclear Information System (INIS)

Paratte, J.M.

1982-07-01

The LWR-Core behaviour project concerns the mathematical simulation of a light water reactor in normal operation (emergency situations excluded). Computational tools are assembled, i.e. programs and libraries of data. These computational tools can likewise be used in nuclear power applications, industry and control applications. The project is divided into three parts: the development and application of calculation methods for quantisation determination of LWR physics; investigation of the behaviour of nuclear fuels under radiation with special attention to higher burnup; simulation of the operating transients of nuclear power stations. (A.N.K.)

Analytical Chemistry Core Capability Assessment - Preliminary Report

International Nuclear Information System (INIS)

Barr, Mary E.; Farish, Thomas J.

2012-01-01

The concept of 'core capability' can be nebulous one. Even at a fairly specific level, where core capability equals maintaining essential services, it is highly dependent upon the perspective of the requestor. Samples are submitted to analytical services because the requesters do not have the capability to conduct adequate analyses themselves. Some requests are for general chemical information in support of R and D, process control, or process improvement. Many analyses, however, are part of a product certification package and must comply with higher-level customer quality assurance requirements. So which services are essential to that customer - just those for product certification? Does the customer also (indirectly) need services that support process control and improvement? And what is the timeframe? Capability is often expressed in terms of the currently utilized procedures, and most programmatic customers can only plan a few years out, at best. But should core capability consider the long term where new technologies, aging facilities, and personnel replacements must be considered? These questions, and a multitude of others, explain why attempts to gain long-term consensus on the definition of core capability have consistently failed. This preliminary report will not try to define core capability for any specific program or set of programs. Instead, it will try to address the underlying concerns that drive the desire to determine core capability. Essentially, programmatic customers want to be able to call upon analytical chemistry services to provide all the assays they need, and they don't want to pay for analytical chemistry services they don't currently use (or use infrequently). This report will focus on explaining how the current analytical capabilities and methods evolved to serve a variety of needs with a focus on why some analytes have multiple analytical techniques, and what determines the infrastructure for these analyses. This information will be

Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Concrete Interaction

International Nuclear Information System (INIS)

Robb, Kevin R; Farmer, Mitchell; Francis, Matthew W

2015-01-01

Lower head failure and corium concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis was carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input. MELTSPREAD was then used to predict the spatially dependent melt conditions and extent of spreading during relocation from the vessel. The results of the MELTSPREAD analysis are reported in a companion paper. This information was used as input for the long-term debris coolability analysis with CORQUENCH.

Final report for Tank 241-B-101, push mode cores 90 and 91. Revision 1

International Nuclear Information System (INIS)

Schreiber, R.D.

1995-11-01

This is the final report for tank 241-at sign 101, cores 90 and 91. Samples from these cores were analyzed for safety screening purposes in accordance with the Tank 241-B-101 Tank Characterization Plan (TCP) (Reference 1). This final report contains the results for three sets of TGA and gravimetric analyses performed after the 90-day report was issued. Two of these TGA/gravimetric percent water sets of analyses were done because low original TGA results were obtained for the lower half segment of core 90, segment 2 and the facie of core 91, segment 2; the third set of analyses were performed because the TGA and gravimetric percent water results for the upper half segment of core 90, segment 2 differed by approximately a factor of three and further investigation was warranted

Core calculational techniques and procedures

International Nuclear Information System (INIS)

Romano, J.J.

1977-10-01

Described are the procedures and techniques employed by B and W in core design analyses of power peaking, control rod worths, and reactivity coefficients. Major emphasis has been placed on current calculational tools and the most frequently performed calculations over the operating power range

Special nuclear material simulation device

Science.gov (United States)

Leckey, John H.; DeMint, Amy; Gooch, Jack; Hawk, Todd; Pickett, Chris A.; Blessinger, Chris; York, Robbie L.

2014-08-12

An apparatus for simulating special nuclear material is provided. The apparatus typically contains a small quantity of special nuclear material (SNM) in a configuration that simulates a much larger quantity of SNM. Generally the apparatus includes a spherical shell that is formed from an alloy containing a small quantity of highly enriched uranium. Also typically provided is a core of depleted uranium. A spacer, typically aluminum, may be used to separate the depleted uranium from the shell of uranium alloy. A cladding, typically made of titanium, is provided to seal the source. Methods are provided to simulate SNM for testing radiation monitoring portals. Typically the methods use at least one primary SNM spectral line and exclude at least one secondary SNM spectral line.

Multi-physics design and analyses of long life reactors for lunar outposts

Science.gov (United States)

Schriener, Timothy M.

event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete

Developing an undergraduate curriculum in Special Care Dentistry - by consensus.

LENUS (Irish Health Repository)

Dougall, A

2013-02-01

It has been reported that healthcare providers often lack the skills set to provide care for people with disabilities, leading to inequalities in health and reduced access to health care. Newly graduating dentists are likely to see a significant number of patients with special healthcare needs in the course of their practicing lives. However, there is evidence of national and international variation in the availability of education and training at the undergraduate level in this important, emerging area. The quality and content of undergraduate education in Special Care Dentistry has been shown to correlate with students\\' confidence and their expressed willingness, towards providing care for patients with special healthcare needs in their future practice. The aim of this study was to use information from a three-round Delphi process, continued into a face-to-face meeting, to establish consensus on what constitutes the essential core knowledge, skills and attitudes required by a newly qualified dentist so that they are able to deliver patient care to diverse populations following graduation. A high level of agreement was established amongst an international panel of experts from 30 countries. The final core items identified by the panel showed a paradigm shift away from the traditional emphasis on medical diagnosis within a curriculum towards an approach based on the International Classification of Functioning (ICF) with patient-centred treatment planning for people with disabilities and special healthcare needs according to function or environment. Many of the core skills identified by the panel are transferable across a curriculum and should encourage a person-centred approach to treatment planning based on the function, needs and wishes of the patient rather than their specific diagnosis.

Core Promoter Structure in the Oomycete Phytophthora infestans

Science.gov (United States)

McLeod, Adele; Smart, Christine D.; Fry, William E.

2004-01-01

We have investigated the core promoter structure of the oomycete Phytophthora infestans. The transcriptional start sites (TSS) of three previously characterized P. infestans genes, Piexo1, Piexo3, and Piendo1, were determined by primer extension analyses. The TSS regions were homologous to a previously identified 16-nucleotide (nt) core sequence that overlaps the TSS in most oomycete genes. The core promoter regions of Piexo1 and Piendo1 were investigated by using a transient protoplast expression assay and the reporter gene β-glucuronidase. Mutational analyses of the promoters of Piexo1 and Piendo1 showed that there is a putative core promoter element encompassing the TSS (−2 to + 5) that has high sequence and functional homology to a known core promoter element present in other eukaryotes, the initiator element (Inr). Downstream and flanking the Inr is a highly conserved oomycete promoter region (+7 to + 15), hereafter referred to as FPR (flanking promoter region), which is also important for promoter function. The importance of the 19-nt core promoter region (Inr and FPR) in Piexo1 and Piendo1 was further investigated through electrophoretic mobility shift assays (EMSA). The EMSA studies showed that (i) both core promoters were able to specifically bind a protein or protein complex in a P. infestans whole-cell protein extract and (ii) the same mutations that reduced binding of the EMSA complex also reduced β-glucuronidase (GUS) levels in transient expression assays. The consistency of results obtained using two different assays (GUS transient assays [in vivo] and EMSA studies [in vitro]) supports a convergence of inference about the relative importance of specific nucleotides within the 19-nt core promoter region. PMID:14871940

Evidence of fire resistance of hollow-core slabs

DEFF Research Database (Denmark)

Hertz, Kristian Dahl; Sørensen, Lars Schiøtt; Giuliani, Luisa

is therefore going on in the Netherlands about the fire resistance of hollow-core slabs. In 2014 the producers of hollow-core slabs have published a report of a project called Holcofire containing a collection of 162 fire tests on hollow-core slabs giving for the first time an overview of the fire tests made....... The present paper analyses the evidence now available for assessment of the fire resistance of extruded hollow-core slabs. The 162 fire tests from the Holcofire report are compared against the requirements for testing from the product standard for hollow-core slabs EN1168 and knowledge about the possible......Hollow-core slabs have during the past 50 years comprised a variety of different structures with different cross-sections and reinforcement. At present the extruded hollow-core slabs without cross-reinforcement in the bottom flange and usually round or oval longitudinal channels (holes...

Common Core Preparation in Special Education Teacher Education Programs: Beginning the Conversation

Science.gov (United States)

Murphy, Michelle R.; Marshall, Kathleen J.

2015-01-01

The Common Core State Standards (CCSS) were developed to encourage a common focus of instruction and evaluation in the areas of mathematics, reading/language arts, writing, speaking, and listening. As of 2011, all but five states have adopted CCSS for math and English Language Arts (ELA), with another adopting only the standards for ELA. With…

The development of direct core monitoring in Nuclear Electric plc

International Nuclear Information System (INIS)

Curtis, R.F.; Jones, S. Reed, J.; Wickham, A.J.

1996-01-01

Monitoring of graphite behaviour in Nuclear Electric Magnox and AGR reactors is necessary to support operating safety cases and to ensure that reactor operation is optimized to sustain the necessary core integrity for the economic life of the reactors. The monitoring programme combines studies for pre-characterized ''installed'' samples with studies on samples trepanned from within the cores and also with studies of core and channel geometry using specially designed equipment. Nuclear Electric has two trepanning machines originally designed for Magnox-reactor work which have been used for a substantial programme over many years. They have recently been upgraded to improve sampling speed, safety and versatility - the last being demonstrated by their adaptation for a recently-won contract associated with decommissioning the Windscale piles. Radiological hazards perceived when the AGR trepanning system was designed resulted in very cumbersome equipment. This has worked well but has been inconvenient in operation. The development of a smaller and improved system for deploying the equipment is now reported. Channel dimension monitoring equipment is discussed in detail with examples of data recovered from both Magnox and AGR cores. A resolution of ± 2 of arc (tilt) and ± 0.01 mm change in diameter in attainable. It is also theoretically possible to establish brick stresses by measuring geometry changes which result from trepanning. Current development work on a revolving scanning laser rangefinder which will enable the measurement of diameters to a resolution of 0.001 mm will also be discussed. This paper also discusses non-destructive techniques for crack detection employing ultrasound or resistance networks, the use of special manipulators to deliver inspection and repair equipment and recent developments to install displacement monitors in peripheral regions of the cores, to aid the understanding of the interaction of the restraint system with the core - the region

Thermal behavior of molten corium during TMI-2 core relocation event

International Nuclear Information System (INIS)

Anderson, J.L.; Sienicki, J.J.

1988-01-01

During the TMI-2 accident, a pool of molten corium formed in the central region of the core and was contained by solidified crusts. Failure of the crust surrounding the molten material, at approximately 224 min, resulted in a relocation of an estimated 20-25 tons of molten corium through peripheral fuel assemblies in the east side of the vessel, as well as through the core barrel assembly (CBA) at the periphery of the core. This paper presents the results of an analyses carried out to investigate the thermal interactions of molten corium with the CBA structures during the relocation event. The principal objectives of the analyses are: (a) to assess the potential for relocation to take place through the CBA versus the flow of molten core material directly downward through the core via the fuel assemblies; and (b) to understand the distribution of prior molten corium observed during vessel defueling examinations. 5 refs., 1 fig

Research reactor core conversion guidebook. V.2: Analysis (Appendices A-F)

International Nuclear Information System (INIS)

1992-04-01

Volume 2 consists of detailed Appendices, covering safety analyses for generic 10 MW reactor, safety analysis - probabilistic methods, methods for preventing LOCA, radiological consequence analyses, examples of safety report amendments and safety specifications. Included in Volume 2 are example analyses for cores with with highly enriched uranium and low enriched uranium fuels showing differences that can be expected in the safety parameters and radiological consequences of postulated accidents. There are seven examples of licensing documents related to core conversion and two examples of methods for determining power limits for safety specifications in the document. Refs, figs, bibliographies and tabs

The code DYN3DR for steady-state and transient analyses of light water reactor cores with Cartesian geometry

International Nuclear Information System (INIS)

Grundmann, U.

1995-11-01

The code DYN3D/M2 was developed for 3-dimensional steady-state and transient analyses of reactor cores with hexagonal fuel assemblies. The neutron kinetics of the new version DYN3DR is based on a nodal method for the solution of the 3-dimensional 2-group neutron diffusion equation for Cartesian geometry. The thermal-hydraulic model FLOCAL simulating the two phase flow of coolant and the fuel rod behaviour is used in the two versions. The fundamentals for the solution of the neutron diffusion equations in DYN3DR are described. The 3-dimensional NEACRP benchmarks for rod ejections in LWR with quadratic fuel assemblies were calculated and the results were compared with the published solutions. The developed algorithm for neutron kinetics are suitable for using parallel processing. The behaviour of speed-up versus the number of processors is demonstrated for calculations of a static neutron flux distribution using a workstation with 4 processors. (orig.) [de

Korrelasjon mellom core styrke, core stabilitet og utholdende styrke i core

OpenAIRE

Berg-Olsen, Andrea Marie; Fugelsøy, Eivor; Maurstad, Ann-Louise

2010-01-01

Formålet med studien var å se hvilke korrelasjon det er mellom core styrke, core stabilitet og utholdende styrke i core. Testingen bestod av tre hoveddeler hvor vi testet core styrke, core stabilitet og utholdende styrke i core. Innenfor core styrke og utholdende styrke i core ble tre ulike tester utført. Ved måling av core stabilitet ble det gjennomført kun en test. I core styrke ble isometrisk abdominal fleksjon, isometrisk rygg ekstensjon og isometrisk lateral fleksjon testet. Sit-ups p...

Comparison of advanced mid-sized reactors regarding passive features, core damage frequencies and core melt retention features

International Nuclear Information System (INIS)

Wider, H.

2005-01-01

New Light Water Reactors, whose regular safety systems are complemented by passive safety systems, are ready for the market. The special aspect of passive safety features is their actuation and functioning independent of the operator. They add significantly to reduce the core damage frequency (CDF) since the operator continues to play its independent role in actuating the regular safety devices based on modern instrumentation and control (I and C). The latter also has passive features regarding the prevention of accidents. Two reactors with significant passive features that are presently offered on the market are the AP1000 PWR and the SWR 1000 BWR. Their passive features are compared and also their core damage frequencies (CDF). The latter are also compared with those of a VVER-1000. A further discussion about the two passive plants concerns their mitigating features for severe accidents. Regarding core-melt retention both rely on in-vessel cooling of the melt. The new VVER-1000 reactor, on the other hand features a validated ex-vessel concept. (author)

Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Werner, James Elmer [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; McKellar, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Hummel, Andrew John [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Kennedy, John Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Wright, Richard Neil [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Biersdorf, John Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division

2017-04-01

The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heat exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.

Core supervision methods and future improvements of the core master/presto system at KKB

International Nuclear Information System (INIS)

Lundberg, S.; Wenisch, J.; Teeffelen, W.V.

2000-01-01

Kernkraftwerk Brunsbuettel (KKB) is a KWU 806 MW e BWR located at the lower river Elbe, in Germany. The reactor has been in operation since 1976 and is now operating in its 14. cycle. The core supervision at KKB is performed with the ABB CORE MASTER system. This system mainly contains the 3-D simulator PRESTO supplied by Studsvik Scandpower A/S. The core supervision is performed by periodic PRESTO 3-D evaluations of the reactor operation state. The power distribution calculated by PRESTO is adapted with the ABB UPDAT program using the on-line LPRM readings. The thermal margins are based on this adapted power distribution. Related to core supervision, the function of the PRESTO/UPDAT codes is presented. The UPDAT method is working well and is capable of reproducing the true core power distribution. The quality of the 3-D calculation is, however, an important ingredient of the quality of the adapted power distribution. The adaptation method as such is also important for this quality. The data quality of this system during steady state and off-rate states (reactor manoeuvres) are discussed by presenting comparisons between PRESTO and UPDAT thermal margin utilisation from Cycle 13. Recently analysed asymmetries in the UPDAT evaluated MCPR values are also presented and discussed. Improvements in the core supervision such as the introduction of advanced modern nodal methods (PRESTO-2) are presented and an alternative core supervision philosophy is discussed. An ongoing project with the goal to update the data and result presentation interface (GUI) is also presented. (authors)

Neutronics calculation of RTP core

Science.gov (United States)

Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

2017-01-01

Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

Accidents and transients analyses of a super fast reactor with single flow pass core

International Nuclear Information System (INIS)

Sutanto,; Oka, Yoshiaki

2014-01-01

Highlights: • Safety analysis of a Super FR with single flow pass core is conducted. • Loss of feed water flow leads to a direct effect on the loss of fuel channel flow. • The core pressure is sensitive to LOCA accidents due to the direct effect. • Small LOCA introduces a critical break. • The safety criteria for all selected events are satisfied. - Abstract: The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST

The Healthy People 2010 outcomes for the care of children with special health care needs: an effective national policy for meeting mental health care needs?

Science.gov (United States)

Spears, Amanda P

2010-05-01

To assess the effectiveness of the Maternal and Child Health Bureau's (MCHB) Six Core Outcomes for children with special health care needs (CSHCN) as indicators in measuring the degree to which mental health care needs are met. This study analyzes data from the 2001 National Survey of Children with Special Health Care Needs for 9,748 CSHCN who needed mental health care. Bivariate and logistic analyses were employed to investigate the impact of the MCHB's Six Core Outcomes on the probability of having an unmet need for mental health services. Of the 2.3 million CSHCN in the U.S. who needed mental health care in 2001, almost one-fifth did not receive all of the mental health services that they needed. Ultimately, eight Outcomes and sub-categories of Outcomes were considered. Sixty-one percent of CSHCN with a need for mental health care had care that fulfills six of the eight considered Outcomes. Logistic analysis indicates that individual fulfillment of each of the Core Outcomes and fulfillment of additional Outcomes have a significant association with reducing the probability of having an unmet mental health care need for CSHCN. This study is the first attempt to apply the Six Core Outcomes to meeting the needs for mental health care among CSHCN. Estimates of unmet need for mental health care suggest that efforts can be made to improve access for CSHCN. The initial estimates generated by this study indicate that the MCHB Outcomes are important in meeting children's mental health needs and are important indicators for informing MCHB policy.

Use of compensation assemblies in the first core of SNR-300

International Nuclear Information System (INIS)

Billaux, M.; De Wouters, R.; Pilate, S.; Vandenberg, C.

1975-01-01

For the SNR-300 reactor, the use of thin fuel pins was limited to the first core. A direct consequence of changing from the cycle reloading scheme to a complete irradiation without refueling operation is an increase of the initial excess reactivity and plutonium investment. The new system of special assemblies conceived to compensate for the too high reactivity of the first core is described: fixed absorbers, made of B 4 C pins, and sodium diluents, consisting simply of hollow wrapper tubes [fr

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

Conceptual core designs for a 1200 MWe sodium cooled fast reactor

International Nuclear Information System (INIS)

Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

2008-01-01

The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

Mount Elbert Gas Hydrate Stratigraphic Test Well, Alaska North Slope: Coring operations, core sedimentology, and lithostratigraphy

Science.gov (United States)

Rose, K.; Boswell, R.; Collett, T.

2011-01-01

In February 2007, BP Exploration (Alaska), the U.S. Department of Energy, and the U.S. Geological Survey completed the BPXA-DOE-USGS Mount Elbert Gas Hydrate Stratigraphic Test Well (Mount Elbert well) in the Milne Point Unit on the Alaska North Slope. The program achieved its primary goals of validating the pre-drill estimates of gas hydrate occurrence and thickness based on 3-D seismic interpretations and wireline log correlations and collecting a comprehensive suite of logging, coring, and pressure testing data. The upper section of the Mount Elbert well was drilled through the base of ice-bearing permafrost to a casing point of 594??m (1950??ft), approximately 15??m (50??ft) above the top of the targeted reservoir interval. The lower portion of the well was continuously cored from 606??m (1987??ft) to 760??m (2494??ft) and drilled to a total depth of 914??m. Ice-bearing permafrost extends to a depth of roughly 536??m and the base of gas hydrate stability is interpreted to extend to a depth of 870??m. Coring through the targeted gas hydrate bearing reservoirs was completed using a wireline-retrievable system. The coring program achieved 85% recovery of 7.6??cm (3??in) diameter core through 154??m (504??ft) of the hole. An onsite team processed the cores, collecting and preserving approximately 250 sub-samples for analyses of pore water geochemistry, microbiology, gas chemistry, petrophysical analysis, and thermal and physical properties. Eleven samples were immediately transferred to either methane-charged pressure vessels or liquid nitrogen for future study of the preserved gas hydrate. Additional offsite sampling, analyses, and detailed description of the cores were also conducted. Based on this work, one lithostratigraphic unit with eight subunits was identified across the cored interval. Subunits II and Va comprise the majority of the reservoir facies and are dominantly very fine to fine, moderately sorted, quartz, feldspar, and lithic fragment-bearing to

Number of core samples: Mean concentrations and confidence intervals

International Nuclear Information System (INIS)

Jensen, L.; Cromar, R.D.; Wilmarth, S.R.; Heasler, P.G.

1995-01-01

This document provides estimates of how well the mean concentration of analytes are known as a function of the number of core samples, composite samples, and replicate analyses. The estimates are based upon core composite data from nine recently sampled single-shell tanks. The results can be used when determining the number of core samples needed to ''characterize'' the waste from similar single-shell tanks. A standard way of expressing uncertainty in the estimate of a mean is with a 95% confidence interval (CI). The authors investigate how the width of a 95% CI on the mean concentration decreases as the number of observations increase. Specifically, the tables and figures show how the relative half-width (RHW) of a 95% CI decreases as the number of core samples increases. The RHW of a CI is a unit-less measure of uncertainty. The general conclusions are as follows: (1) the RHW decreases dramatically as the number of core samples is increased, the decrease is much smaller when the number of composited samples or the number of replicate analyses are increase; (2) if the mean concentration of an analyte needs to be estimated with a small RHW, then a large number of core samples is required. The estimated number of core samples given in the tables and figures were determined by specifying different sizes of the RHW. Four nominal sizes were examined: 10%, 25%, 50%, and 100% of the observed mean concentration. For a majority of analytes the number of core samples required to achieve an accuracy within 10% of the mean concentration is extremely large. In many cases, however, two or three core samples is sufficient to achieve a RHW of approximately 50 to 100%. Because many of the analytes in the data have small concentrations, this level of accuracy may be satisfactory for some applications

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

Core Seismic Tests for a Sodium-Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Koo, Gyeong Hoi; Lee, J. H

2007-01-15

This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

Core designs for new VVER reactors and operational experience of immediate prototypes

International Nuclear Information System (INIS)

Vasilchenko, I.; Mokhov, V.; Ryzhov, S.

2011-01-01

The paper covers the recent improvements analyzed in order to implement the enhanced core performances. AES-2006 reactor core design is considered from the point of view of its application and improvement in the planned VVER-TOI project and of the possibilities of using the basic engineering solutions for the cores with spectral control. The discussion of several types of mixing grids considered in the paper involves a preliminary assessment of their efficiency and the information on their introduction into pilot operation at the VVER-1000 Units. Special attention is given to the results of the operation of immediate prototypes (TVS-2 and TVS-2M) that corroborate the reliability of the design both with regard for the core geometrical stability and fuel cladding tightness

Recommended analysis plan for the borehole plugging program potash core test

International Nuclear Information System (INIS)

Lambert, S.J.

1980-05-01

A four-year old plugged potash core hole near the Waste Isolation Pilot Plant (WIPP) site in southeastern New Mexico has been proposed for overcoring, in order to examine the behavior of known grout mix constituents in contact with a variety of rock types during an extended grout-curing interval. This report recommends that various geochemical analyses be applied to the core samples containing both grout and rock and the interface between the two. The methods to be used include optical petrography, electron microscopy, electron probe microanalysis, x-ray diffraction, thermal analysis (TGA, DSC, DTA) with gas chromatography/mass spectrometry, and bulk chemical analysis. These analyses would allow identification of phases which have developed during grout curing, and provide evidence of reactions which may have taken place among constituents in the system grout-rock-groundwater. These reactions, and their sequence of occurrence will be compared with reactions predicted by thermodynamic modeling as the system seeks its lowest Gibbs' free energy. Identification of reactions which have the potential for compromising the integrity of a grout plug will receive special attention. Since not all such detrimental reactions can be observed directly in a human lifetime, due to kinetic inhibitions, and since a capability of time-dependent prediction of their degree of occurrence cannot be developed, thermodynamic modeling is the only known way of evaluating the long-term stability of a grout plug. The analysis of the plug-rock system will give an indication of in situ curing history of grout plug, and will allow an early occurrence of potentially detrimental reactions to be detected. Thus, this activity will be a case-study of suitability of certain grout mixtures for use in evaporites, as an example of evaluation of grouts for long-term compatability with a variety of rock types

Steady state thermal hydraulic analysis of LMR core using COBRA-K code

Energy Technology Data Exchange (ETDEWEB)

Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol

1997-02-01

A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.

Further HTGR core support structure reliability studies. Interim report No. 1

International Nuclear Information System (INIS)

Platus, D.L.

1976-01-01

Results of a continuing effort to investigate high temperature gas cooled reactor (HTGR) core support structure reliability are described. Graphite material and core support structure component physical, mechanical and strength properties required for the reliability analysis are identified. Also described are experimental and associated analytical techniques for determining the required properties, a procedure for determining number of tests required, properties that might be monitored by special surveillance of the core support structure to improve reliability predictions, and recommendations for further studies. Emphasis in the study is directed towards developing a basic understanding of graphite failure and strength degradation mechanisms; and validating analytical methods for predicting strength and strength degradation from basic material properties

Analysis of hepatitis C virus core/NS5A protein co-localization using novel cell culture systems expressing core-NS2 and NS5A of genotypes 1-7

DEFF Research Database (Denmark)

Galli, Andrea; Scheel, Troels K H; Prentoe, Jannick C

2013-01-01

Hepatitis C virus (HCV) is an important human pathogen infecting hepatocytes. With the advent of infectious cell culture systems, the HCV particle assembly and release processes are finally being uncovered. The HCV core and NS5A proteins co-localize on cytoplasmic lipid droplets (c......LDs) or on the endoplasmic reticulum (ER) at different stages of particle assembly. Current knowledge on assembly and release is primarily based on studies in genotype 2a cell culture systems; however, given the high genetic heterogeneity of HCV, variations might exist among genotypes. Here, we developed novel HCV strain...... JFH1-based recombinants expressing core-NS2 and NS5A from genotypes 1-7, and analysed core and NS5A co-localization in infected cells. Huh7.5 cells were transfected with RNA of core-NS2/NS5A recombinants and putative adaptive mutations were analysed by reverse genetics. Adapted core-NS2/NS5A...

Genome Sequence of Azospirillum brasilense CBG497 and Comparative Analyses of Azospirillum Core and Accessory Genomes provide Insight into Niche Adaptation

Science.gov (United States)

Wisniewski-Dyé, Florence; Lozano, Luis; Acosta-Cruz, Erika; Borland, Stéphanie; Drogue, Benoît; Prigent-Combaret, Claire; Rouy, Zoé; Barbe, Valérie; Mendoza Herrera, Alberto; González, Victor; Mavingui, Patrick

2012-01-01

Bacteria of the genus Azospirillum colonize roots of important cereals and grasses, and promote plant growth by several mechanisms, notably phytohormone synthesis. The genomes of several Azospirillum strains belonging to different species, isolated from various host plants and locations, were recently sequenced and published. In this study, an additional genome of an A. brasilense strain, isolated from maize grown on an alkaline soil in the northeast of Mexico, strain CBG497, was obtained. Comparative genomic analyses were performed on this new genome and three other genomes (A. brasilense Sp245, A. lipoferum 4B and Azospirillum sp. B510). The Azospirillum core genome was established and consists of 2,328 proteins, representing between 30% to 38% of the total encoded proteins within a genome. It is mainly chromosomally-encoded and contains 74% of genes of ancestral origin shared with some aquatic relatives. The non-ancestral part of the core genome is enriched in genes involved in signal transduction, in transport and in metabolism of carbohydrates and amino-acids, and in surface properties features linked to adaptation in fluctuating environments, such as soil and rhizosphere. Many genes involved in colonization of plant roots, plant-growth promotion (such as those involved in phytohormone biosynthesis), and properties involved in rhizosphere adaptation (such as catabolism of phenolic compounds, uptake of iron) are restricted to a particular strain and/or species, strongly suggesting niche-specific adaptation. PMID:24705077

Genome Sequence of Azospirillum brasilense CBG497 and Comparative Analyses of Azospirillum Core and Accessory Genomes provide Insight into Niche Adaptation

Directory of Open Access Journals (Sweden)

Victor González

2012-09-01

Full Text Available Bacteria of the genus Azospirillum colonize roots of important cereals and grasses, and promote plant growth by several mechanisms, notably phytohormone synthesis. The genomes of several Azospirillum strains belonging to different species, isolated from various host plants and locations, were recently sequenced and published. In this study, an additional genome of an A. brasilense strain, isolated from maize grown on an alkaline soil in the northeast of Mexico, strain CBG497, was obtained. Comparative genomic analyses were performed on this new genome and three other genomes (A. brasilense Sp245, A. lipoferum 4B and Azospirillum sp. B510. The Azospirillum core genome was established and consists of 2,328 proteins, representing between 30% to 38% of the total encoded proteins within a genome. It is mainly chromosomally-encoded and contains 74% of genes of ancestral origin shared with some aquatic relatives. The non-ancestral part of the core genome is enriched in genes involved in signal transduction, in transport and in metabolism of carbohydrates and amino-acids, and in surface properties features linked to adaptation in fluctuating environments, such as soil and rhizosphere. Many genes involved in colonization of plant roots, plant-growth promotion (such as those involved in phytohormone biosynthesis, and properties involved in rhizosphere adaptation (such as catabolism of phenolic compounds, uptake of iron are restricted to a particular strain and/or species, strongly suggesting niche-specific adaptation.

Petrographic and geochemical characteristics of the Cypress Creek salt core

International Nuclear Information System (INIS)

1983-07-01

Law Engineering Testing Company supervised the drilling of a corehole into the stock of Cypress Creek Dome, located in Perry County, Mississippi. A total of 170 ft of caprock and 501 ft of salt stock was recovered for physical examination and chemical analysis. This report describes the types of analyses performed and summarizes the data developed. The entire caprock and salt core were described and photographed prior to selection of samples for petrologic and geochemical analysis. Transmitted light techniques were used to determine gross structural and compositional variations in the core. The core lithologies are presented graphically, at a scale of 1 in. to 2 ft. In addition to the detailed field descriptions and photographs, petrologic studies performed on selected caprock and salt samples included: thin-section examination, scanning-electron microscope studies, energy-dispersion analysis, and x-ray-diffraction analysis. Geochemical analyses were performed to determine the average elemental composition of the salt core and amounts of methane and carbon dioxide gases contained within the salt grains. Except for two thin (3 and 6 ft thick) gypsum zones in the top 27 ft of the caprock, the core is predominantly anhydrite (generally 80%). Minor amounts of dolomite and calcite are also present. The salt core consists predominantly of crystalline halite, fine- to medium-grained (0.25 to 1 in.) with few megacrysts. Anhydrite occurs in the salt core as disseminated grains, ranging in length from <0.1 in. to 12 in. Discrete zones exist within the salt core, distinguished from one another primarily by the character of the anhydrite inclusions

Comparison of thermohydraulic and nuclear aspects in a standard HEU core and a typical LEU core for the HFR Petten. A case study

International Nuclear Information System (INIS)

Pruimboom, H.; Tas, A.

1985-01-01

Within the framework of the RERTR program various HEU-LEU core calculations have been performed by ANL in a cooperative effort with ECN and JRC Petten. The main purpose of this work has been to gain competence in analysing HEU-LEU core conversion for high power Materials Testing Reactors and to assist in a possible HEU-LEU conversion of the HFR Petten. For reference purposes the present HFR standard core (HEU) in the 'old' vessel geometry was calculated at first. As a next step the new vessel geometry and the increased fuel weights were taken into account. Subsequently various LEU HFR core options have been analysed. Main parameters in the LEU study were the uranium loading in the meat, the fuel type, the thickness of the meat, the number of fuel plates per element and the type of burnable poison applied. Though the study has not yet been completed, one of its striking preliminary results concerns the increased power peaking in the LEU fuel elements as compared with the HEU situation. A preliminary analysis of the thermal characteristics of a typical LEU core as compared with a standard HEU core has been made and is presented in the paper. A short survey of the various HEU and LEU calculations is given. The thermal safety analysis procedure for the HFR, as based on the flow instability criterion, is clarified. Finally, the thermal comparison HEU versus LEU and the resulting conclusions are presented. (author)

Long core model apparatus for laboratory investigation of oil recovery processes

Energy Technology Data Exchange (ETDEWEB)

Milley, Gy; Bukta, B; Jonap, K; Lovei, J; Wagner, O

1982-01-01

In studying the secondary and Tertiary processes of intensifying oil output, an analysis was made of the following main factors: behavior of multiple-phase and multiple-component system in the porous medium, their stability, mobility, effect of viscosity and pressure differential in the system, configuration of the displacement front, influence of chemical additives on optimizing the surface energy, economic characteristics of the processes. All of these factors can be studied on a laboratory unit with core samples up to 120 cm long with temperatures to 120/sup 0/C, pressures to 30 MPa and consumption of reagents to 100 cm/sup 3//h. The unit contains feed vessels of high pressure for water, oil and gas in different reagents. There is a pumping unit of two-stage type, and in the first loop there is a piston pump which feeds the cylinder of the pump of the second stage of the plunger type. The outlet stage of this pump through the valve system is connected to the corresponding vessel for pressing the necessary reagent through the core sample. One can continually change the pressure to 100 MPa. The core is placed in a special core carrier and using special high temperature resins, it is packed in it in order to exclude side overflows. There is a technology of packing of comparatively soft rocks (clay). Sketches are presented of the sealing assemblies, and also the plans for inserting cables for the sensors. The sensor system is arranged over the entire length of the core and generates signals which are proportional to the magnitude of water saturation. The outlet of the core carrier has devices for resetting pressure, collecting filtrate, measurement of its composition and consumption. The core carrier is thermostatically controlled and contains two kW electrical heater for taking measurements at different temperatures. With a change in the system modes, the equilibrium is reached in 2-3 h.

Neutronic investigations of an equilibrium core for a tight-lattice light water reactor

International Nuclear Information System (INIS)

Broeders, C.H.M.

1992-01-01

Calculation procedures and first results concerning the neutronic design of an equilibrium core of an advanced pressurized water reactor (APWR) with mixed oxide fuel in a compact light water moderated triangular lattice are presented. Principle and qualification of the cell burnup calculations with the KARBUS program are briefly discussed. The fuel assembly design with single control rod positions filled with control rod material or coolant water requires special transport theory calculations, which are performed with a one-dimensional supercell model. The macroscopic fuel assembly cross section data is collected in a special library to be used in a new calculational procedure, ARCOSI, for multi-cycle reactor core simulations. Its first application for a reference design resulted in an equilibrium configuration with moderator density reactivity coefficients which are satisfactory as regards safety. (orig.) [de

Recent improvements in on-line core supervision at Loviisa NPP

International Nuclear Information System (INIS)

Antila, M.; Kuusisto, J.

1999-01-01

On-line core supervision system (RESU) based on monitoring of local fuel limits has been in use at the Loviisa VVER-440 reactors for more than twenty years. Minor modifications were made ten years ago when the computer hardware was upgraded. In April 1998 Loviisa got the licence for 1500 MW power. Power uprating and introduction of new fuel types gave rise to the latest improvements in the core supervision system, which is called RESU-98. In August 1999 the Finnish Safety Authority (STUK) has given approval for RESU-98, which is now in use at the Loviisa NPP. RESU-98 includes essentially the same computer codes, which are used in reload planning. The extensive in-core instrumentation is utilised to adjust the theoretical 3D-power distribution to get a best-estimate results. In this paper a general review of the RESU-98 system is given including instrumentation, methods, core monitoring, predictive functions and validation. Special attention is paid on the recent improvements. (Authors)

Recent improvements in on-line core supervision at Loviisa NPP

International Nuclear Information System (INIS)

Antila, M.; Kuusisto, J.

2000-01-01

AN on-line core supervision system (RESU) based on monitoring of local fuel limits has been in use at the Loviisa VVER-440 reactors for more than twenty years. Minor modifications were made ten years ago to upgrade the computer hardware. In April 1998 Loviisa obtained a licence for 1500 MW th power. Power up-rating and introduction of new fuel types gave rise to the latest improvements in the core supervision system, which is called RESU-98. In August 1999 the Finnish Safety Authority (STUCK) officially approved RESU-98, which is now in use at the Loviisa NPP. RESU-98 includes essentially the same computer codes, which are used in reload planning. The extensive in-core instrumentation is utilised to adjust the theoretical 3-D power distribution to get a best-estimate result. In this paper a general review of the RESU-98 system is given including instrumentation, methods, core monitoring, predictive functions and validation. Special attention is paid to recent improvements. (author)

An ultra-clean technique for accurately analysing Pb isotopes and heavy metals at high spatial resolution in ice cores with sub-pg g{sup -1} Pb concentrations

Energy Technology Data Exchange (ETDEWEB)

Burn, Laurie J. [Department of Imaging and Applied Physics, Curtin University of Technology, GPO Box U1987, Perth 6845, Western Australia (Australia); Rosman, Kevin J.R. [Department of Imaging and Applied Physics, Curtin University of Technology, GPO Box U1987, Perth 6845, Western Australia (Australia)], E-mail: K.Rosman@curtin.edu.au; Candelone, Jean-Pierre [Department of Imaging and Applied Physics, Curtin University of Technology, GPO Box U1987, Perth 6845, Western Australia (Australia); Vallelonga, Paul [Department of Imaging and Applied Physics, Curtin University of Technology, GPO Box U1987, Perth 6845, Western Australia (Australia); Istituto per la Dinamica dei Processi Ambientali (IDPA-CNR), Dorsoduro 2137, 30123 Venice (Italy); Burton, Graeme R. [Department of Imaging and Applied Physics, Curtin University of Technology, GPO Box U1987, Perth 6845, Western Australia (Australia); Smith, Andrew M. [Australian Nuclear Science and Technology Organisation (ANSTO), PMB 1, Menai, NSW 2234 (Australia); Morgan, Vin I. [Australian Antarctic Division and Antarctic Climate and Ecosystems CRC, Private Bag 80, Hobart, Tasmania 7001 (Australia); Barbante, Carlo [Istituto per la Dinamica dei Processi Ambientali (IDPA-CNR), Dorsoduro 2137, 30123 Venice (Italy); Hong, Sungmin [Korea Polar Research Institute, Songdo Techno Park, 7-50, Songdo-dong, Yeonsu-gu, Incheon 406-840 (Korea, Republic of); Boutron, Claude F. [Laboratoire de Glaciologie et Geophysique de l' Environnement du CNRS, 54, rue Moliere, B.P. 96, 3840.2 St Martin d' Heres Cedex (France)

2009-02-23

Measurements of Pb isotope ratios in ice containing sub-pg g{sup -1} concentrations are easily compromised by contamination, particularly where limited sample is available. Improved techniques are essential if Antarctic ice cores are to be analysed with sufficient spatial resolution to reveal seasonal variations due to climate. This was achieved here by using stainless steel chisels and saws and strict protocols in an ultra-clean cold room to decontaminate and section ice cores. Artificial ice cores, prepared from high purity water were used to develop and refine the procedures and quantify blanks. Ba and In, two other important elements present at pg g{sup -1} and fg g{sup -1} concentrations in Polar ice, were also measured. The final blank amounted to 0.2 {+-} 0.2 pg of Pb with {sup 206}Pb/{sup 207}Pb and {sup 208}Pb/{sup 207}Pb ratios of 1.16 {+-} 0.12 and 2.35 {+-} 0.16, respectively, 1.5 {+-} 0.4 pg of Ba and 0.6 {+-} 2.0 fg of In, most of which probably originates from abrasion of the steel saws by the ice. The procedure was demonstrated on a Holocene Antarctic ice core section and was shown to contribute blanks of only {approx}5%, {approx}14% and {approx}0.8% to monthly resolved samples with respective Pb, Ba and In concentrations of 0.12 pg g{sup -1}, 0.3 pg g{sup -1} and 2.3 fg g{sup -1}. Uncertainties in the Pb isotopic ratio measurements were degraded by only {approx}0.2%.

TMI-2 core-examination program: INEL facilities-readiness study

International Nuclear Information System (INIS)

McLaughlin, T.B.

1982-09-01

This document is a review of the Idaho National Engineering Laboratory's (INEL) remote handling facilities. Their availability and readiness to conduct examination and analyses of TMI-2 core samples was determined. Examination of these samples require that the facilities be capable of receiving commercial casks, unloading canisters from the casks, opening the canisters, handling the fuel debris and assemblies, and performing various examinations. The documentation that was necessary for the INEL to have before the receipt of the core material was identified. The core information was also required for input to these documents. The costs, schedules, and a preliminary-project plan are presented for the tasks which are identified as prerequisites to the receipt of the first core sample

Experimental and numerical analysis of fluid - structure interaction effects in a fast reactor core

International Nuclear Information System (INIS)

Martelli, A.; Forni, M.; Melloni, R.; Paoluzzi, R.; Bonacina, G.; Castoldi, A.; Zola, M.

1990-01-01

Dynamic experiments in air and water (simulating liquid sodium) were performed by ISMES, on behalf of ENEA, on various core element groups of the Italian PEC fast reactor. Bundles of one, seven and nineteen mock-ups reproducing fuel, reflecting and neutron shield elements in full scale were analysed on shaking tables. Tests concerned both groups of equal elements and mixed configurations which corresponded to real core parts. The effects of PEC core-restraint ring were also studied. Seismic excitations of up to 2.5 g were applied to core diagrid. Test results were analysed by use of the one-dimensional program CORALIE and the two-dimensional program CLASH. The study allowed the fluid effects in the PEC core to be evaluated; it also contributed to validation of the above mentioned programs for their general use for fast reactor core analysis. This paper presents the main features of the experimental and the numerical studies and reports comparisons between calculations and measurements. (author)

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

Seismic responses of N-Reactor core. Independent review of Phase II work

International Nuclear Information System (INIS)

Chen, J.C.; Lo, T.; Chinn, D.J.; Murray, R.C.; Johnson, J.J.; Maslenikov, O.R.

1985-08-01

Seismic response of the N-Reactor core was independently analyzed to validate the results of Impell's analysis. The analysis procedure consists of two major stages: linear soil-structure interaction (SSI) analysis of the overall N-Reactor structure complex and nonlinear dynamic analysis of the reactor core. In the SSI analysis, CLASSI computer codes were used to calculate the SSI response of the structures and to generate the input motions for the nonlinear reactor core analysis. In addition, the response was compared to the response from the SASSI analysis under review. The impact of foundation modeling techniques and the effect of soil stiffness variation on SSI response were also investigated. In the core analysis, a nonlinear dynamic analysis model was developed. The stiffness representation of the model was calculated through a finite element analysis of several local core geometries. Finite element analyses were also used to study the block to block interaction characteristics. Using this nonlinear dynamic model along with the basemat time histories generated from CLASSI and SASSI, several dynamic analyses of the core were performed. A series of sensitivity studies was performed to investigate the discretization of the core, the effect of vertical acceleration, the effect of basemat rocking, and modeling assumptions. In general, our independent analysis of core response validates the order of magnitude of the displacement calculated by Impell. 11 refs., 110 figs., 12 tabs

Correlations among FBR core characteristics for various fuel compositions

International Nuclear Information System (INIS)

Maruyama, Shuhei; Ohki, Shigeo; Okubo, Tsutomu; Kawashima, Katsuyuki; Mizuno, Tomoyasu

2012-01-01

In the design of a fast breeder reactor (FBR) core for the light water reactor (LWR) to FBR transition stage, it is indispensable to grasp the effect of a wide range of fuel composition variations on the core characteristics. This study finds good correlations between burnup reactivity and safety parameters, such as the sodium void reactivity and Doppler coefficient, for various fuel compositions and determines the mechanisms behind these correlations with the aid of sensitivity analyses. It is clarified that the Doppler coefficient is actually correlated with the other core characteristics by considering the constraint imposed by the requirement of sustaining criticality on the fuel composition variations. These correlations make it easy to specify the various properties ranges for core reactivity control and core safety, which are important for core design in determining the core specifications and performance. They provide significant information for FBR core design for the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. With the use of this index and these correlations, the core characteristic variations can be estimated for various fuel compositions without repeating the core calculations. (author)

SEDIMENTATION AND DEPOSITIONAL ENVIRONMENT BASED ON SEISMIC AND DRILLING CORE ANALYSES IN CIMANUK DELTA INDRAMAYU, WEST JAVA

Directory of Open Access Journals (Sweden)

I Nyoman Astawa

2017-07-01

Full Text Available Core drilling had been carried out in three locations such as in Brondong Village (BH-01, Pasekan Village (BH-02, and Karangsong Village (BH-03. Those three cores are similar in lithology consist of clay. They are correlated based on fragment content, such as fine sand lenses, mollusk shells, rock and carbonate materials which discovered from different depths. Single side band of shallow seismic reflection recorded paleochannels in E sequence at the north and the west of investigated area. It’s predicted the north paleo channels were part of Lawas River or Tegar River, while the west paleo channels were part of Rambatan Lama River. Microfauna content of all those three cores indicated that from the depth of 0.00 meter down to 25,00 meters are Holocene/Recent, from 25,00 meters to the bottom are Pleistocene which were deposited in the bay to middle neritic environment.

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)

Editorial: Special Issue on Experimental Vibration Analysis

Science.gov (United States)

Serra, Roger

2018-04-01

The vibratory analyses are particularly present today in the various fields of industry, from aeronautics to manufacturing, from machining and maintenance to civil engineering, to mention a few areas, which have made this special issue a true need. The International Journal of Mechanics & Industry compiles a Special Issue on Experimental Vibration Analysis. More than thirty manuscripts were received by the international scientific committee on the 6th congress AVE2016 and only eight papers have been selected after completing a careful and rigorous peer-review process for the Special Issue, which are briefly summarized below.

Deterministic analyses of severe accident issues

International Nuclear Information System (INIS)

Dua, S.S.; Moody, F.J.; Muralidharan, R.; Claassen, L.B.

2004-01-01

Severe accidents in light water reactors involve complex physical phenomena. In the past there has been a heavy reliance on simple assumptions regarding physical phenomena alongside of probability methods to evaluate risks associated with severe accidents. Recently GE has developed realistic methodologies that permit deterministic evaluations of severe accident progression and of some of the associated phenomena in the case of Boiling Water Reactors (BWRs). These deterministic analyses indicate that with appropriate system modifications, and operator actions, core damage can be prevented in most cases. Furthermore, in cases where core-melt is postulated, containment failure can either be prevented or significantly delayed to allow sufficient time for recovery actions to mitigate severe accidents

Special Operations Forces Reference Manual. Fourth Edition

Science.gov (United States)

2015-06-01

activities that support an adversary’s ability to negatively affect U.S. interests. CTF support can assist SOF in the execution of core activities in...the split team concept making up two six-man teams. Assistant Detachment Operations Sergeant Methods of Infiltration Special Forces soldiers possess...Twelve ODAs per SFG can infil- trate and exfiltrate by surface swim techniques. Unless specifically identified, the only teams with designated specialty

The influence of reactor core parameters on effective breeding coefficient Keff

Institute of Scientific and Technical Information of China (English)

Liu Li-Po; Liu Yi-Bao; Wang Juan; Yang Bo; Zhang Tao

2008-01-01

The values of effective breeding coefficient Keff in a reactor core of nuclear power plant are calculated for different values of parameters (core structure, fuel assembly component) by using the Monte Carlo method. The obtained values of Keff are compared and analysed, which can provide theoretical basis for reactor design.

The database 'EDUD Base' for validation of neutron-physics codes used to analyze the WWER-440 cores

International Nuclear Information System (INIS)

Rocek, J.; Belac, J.; Miasnikov, A.

2003-01-01

The program and data system EDUDBase for validation of reactor computing codes was developed at NRI. It is designed for validation and evaluation of the precision of different computer codes used for WWER core analyses. The main goal of this database is to provide data for comparison with calculation results of tested codes and tools for statistical analysis or differences between the calculation results and the test data. The benchmark data sets are based on in-core measurements performed on WWER-440 reactors of Dukovany NPP. The initial data from NPP are verified, errors and inaccuracies are eliminated and data are transferred to a form, which is suitable for comparison with results of calculations. A special reduced operating history data set is created for each operating cycle ('Benchmark Operation History') to be used as an input data for calculation. It contains values of some integral quantities for each time point: effective time, integral thermal power, boron concentration, position of working group control assemblies (group 6) and inlet coolant temperature. At present, sets are available for all completed cycles up to: (unit/cycle) 1/17, 2/16, 3/15, 4/15. Power distribution is described for approx. 40 time steps during each operating cycle. 2D-power distributions are transferred into 60-degree core symmetry sector of reactor core. At present, such data sets are available only for later cycles starting with: (unit/cycle) 1/7, 2/6, 3/5, 4/5 (in other words last II cycles for each unit) (Authors)

Modularized Functions of the Fanconi Anemia Core Complex

Directory of Open Access Journals (Sweden)

Yaling Huang

2014-06-01

Full Text Available The Fanconi anemia (FA core complex provides the essential E3 ligase function for spatially defined FANCD2 ubiquitination and FA pathway activation. Of the seven FA gene products forming the core complex, FANCL possesses a RING domain with demonstrated E3 ligase activity. The other six components do not have clearly defined roles. Through epistasis analyses, we identify three functional modules in the FA core complex: a catalytic module consisting of FANCL, FANCB, and FAAP100 is absolutely required for the E3 ligase function, and the FANCA-FANCG-FAAP20 and the FANCC-FANCE-FANCF modules provide nonredundant and ancillary functions that help the catalytic module bind chromatin or sites of DNA damage. Disruption of the catalytic module causes complete loss of the core complex function, whereas loss of any ancillary module component does not. Our work reveals the roles of several FA gene products with previously undefined functions and a modularized assembly of the FA core complex.

Analytical results of Tank 38H core samples -- Fall 1999

International Nuclear Information System (INIS)

Swingle, R.F.

2000-01-01

Two samples were pulled from Tank 38H in the Fall of 1999: a variable depth sample (VDS) of the supernate was pulled in October and a core sample from the salt layer was pulled in December. Analysis of the rinse from the outside of the core sample indicated no sign of volatile or semivolatile organics. Both supernate and solids from the VDS and the dried core sample solids were analyzed for isotopes which could pose a criticality concern and also for elements which could serve as neutron poisons, as well as other elements. Results of the elemental analyses of these samples show significant elements present to mitigate the potential for nuclear criticality. However, it should be noted the results given for the VDS solids elemental analyses may be higher than the actual concentration in the solids, since the filter paper was dissolved along with the sample solids

Towards sustainable innovation : analysing and dealing with systemic problems in innovation systems

NARCIS (Netherlands)

Wieczorek, Anna

2014-01-01

Technological Innovation System (TIS) perspective became a popular tool to analyse and understand the diffusion of particular, mostly renewable, technologies and their contribution to sustainability transitions. The core of the current TIS studies comprise of the analyses of the emergent structural

Angiographic core laboratory reproducibility analyses: implications for planning clinical trials using coronary angiography and left ventriculography end-points.

Science.gov (United States)

Steigen, Terje K; Claudio, Cheryl; Abbott, David; Schulzer, Michael; Burton, Jeff; Tymchak, Wayne; Buller, Christopher E; John Mancini, G B

2008-06-01

To assess reproducibility of core laboratory performance and impact on sample size calculations. Little information exists about overall reproducibility of core laboratories in contradistinction to performance of individual technicians. Also, qualitative parameters are being adjudicated increasingly as either primary or secondary end-points. The comparative impact of using diverse indexes on sample sizes has not been previously reported. We compared initial and repeat assessments of five quantitative parameters [e.g., minimum lumen diameter (MLD), ejection fraction (EF), etc.] and six qualitative parameters [e.g., TIMI myocardial perfusion grade (TMPG) or thrombus grade (TTG), etc.], as performed by differing technicians and separated by a year or more. Sample sizes were calculated from these results. TMPG and TTG were also adjudicated by a second core laboratory. MLD and EF were the most reproducible, yielding the smallest sample size calculations, whereas percent diameter stenosis and centerline wall motion require substantially larger trials. Of the qualitative parameters, all except TIMI flow grade gave reproducibility characteristics yielding sample sizes of many 100's of patients. Reproducibility of TMPG and TTG was only moderately good both within and between core laboratories, underscoring an intrinsic difficulty in assessing these. Core laboratories can be shown to provide reproducibility performance that is comparable to performance commonly ascribed to individual technicians. The differences in reproducibility yield huge differences in sample size when comparing quantitative and qualitative parameters. TMPG and TTG are intrinsically difficult to assess and conclusions based on these parameters should arise only from very large trials.

Special relativity - from particles to astrophysics

International Nuclear Information System (INIS)

Gourgoulhon, E.

2010-01-01

Special relativity is not a particular theory of physics but rather a theoretical framework through which various dynamical theories can be expressed. The main advantage of this book is to highlight the essential structures of special relativity before illustrating them with applications. One of these structures is the important Minkowski 4-dimensional space-time whose basic object is the quadri-vector. This mathematical framework is defined as early as the first chapter, which gives special relativity a more axiomatic approach than in other manuals. Another feature of this account of special relativity is to base the discussion of measurable physics effects on the point of view of a general observer who is no more restricted to be in a uniform motion, he can be accelerating or rotating. As a consequence the Lorentz transformation appears here less essential than in other presentations. In the second part of this book that begins with the 14. chapter, the main physical object is no more a particle but a field. The book ends with the issue of gravitation. The author highlights applications from particle physics (accelerators, particle collisions or quark-gluon plasmas), to astrophysics (relativistic jets or active cores of galaxies) via more practical applications such as Sagnac effect gyro-meters, synchrotron radiation or global positioning systems. (A.C.)

CORE: a phylogenetically-curated 16S rDNA database of the core oral microbiome.

Directory of Open Access Journals (Sweden)

Ann L Griffen

2011-04-01

Full Text Available Comparing bacterial 16S rDNA sequences to GenBank and other large public databases via BLAST often provides results of little use for identification and taxonomic assignment of the organisms of interest. The human microbiome, and in particular the oral microbiome, includes many taxa, and accurate identification of sequence data is essential for studies of these communities. For this purpose, a phylogenetically curated 16S rDNA database of the core oral microbiome, CORE, was developed. The goal was to include a comprehensive and minimally redundant representation of the bacteria that regularly reside in the human oral cavity with computationally robust classification at the level of species and genus. Clades of cultivated and uncultivated taxa were formed based on sequence analyses using multiple criteria, including maximum-likelihood-based topology and bootstrap support, genetic distance, and previous naming. A number of classification inconsistencies for previously named species, especially at the level of genus, were resolved. The performance of the CORE database for identifying clinical sequences was compared to that of three publicly available databases, GenBank nr/nt, RDP and HOMD, using a set of sequencing reads that had not been used in creation of the database. CORE offered improved performance compared to other public databases for identification of human oral bacterial 16S sequences by a number of criteria. In addition, the CORE database and phylogenetic tree provide a framework for measures of community divergence, and the focused size of the database offers advantages of efficiency for BLAST searching of large datasets. The CORE database is available as a searchable interface and for download at http://microbiome.osu.edu.

Core microbial functional activities in ocean environments revealed by global metagenomic profiling analyses.

KAUST Repository

Ferreira, Ari J S

2014-06-12

Metagenomics-based functional profiling analysis is an effective means of gaining deeper insight into the composition of marine microbial populations and developing a better understanding of the interplay between the functional genome content of microbial communities and abiotic factors. Here we present a comprehensive analysis of 24 datasets covering surface and depth-related environments at 11 sites around the world\\'s oceans. The complete datasets comprises approximately 12 million sequences, totaling 5,358 Mb. Based on profiling patterns of Clusters of Orthologous Groups (COGs) of proteins, a core set of reference photic and aphotic depth-related COGs, and a collection of COGs that are associated with extreme oxygen limitation were defined. Their inferred functions were utilized as indicators to characterize the distribution of light- and oxygen-related biological activities in marine environments. The results reveal that, while light level in the water column is a major determinant of phenotypic adaptation in marine microorganisms, oxygen concentration in the aphotic zone has a significant impact only in extremely hypoxic waters. Phylogenetic profiling of the reference photic/aphotic gene sets revealed a greater variety of source organisms in the aphotic zone, although the majority of individual photic and aphotic depth-related COGs are assigned to the same taxa across the different sites. This increase in phylogenetic and functional diversity of the core aphotic related COGs most probably reflects selection for the utilization of a broad range of alternate energy sources in the absence of light.

Core microbial functional activities in ocean environments revealed by global metagenomic profiling analyses.

KAUST Repository

Ferreira, Ari J S; Siam, Rania; Setubal, Joã o C; Moustafa, Ahmed; Sayed, Ahmed; Chambergo, Felipe S; Dawe, Adam S; Ghazy, Mohamed A; Sharaf, Hazem; Ouf, Amged; Alam, Intikhab; Abdel-Haleem, Alyaa M; Lehvä slaiho, Heikki; Ramadan, Eman; Antunes, André ; Stingl, Ulrich; Archer, John A.C.; Jankovic, Boris R; Sogin, Mitchell; Bajic, Vladimir B.; El-Dorry, Hamza

2014-01-01

Metagenomics-based functional profiling analysis is an effective means of gaining deeper insight into the composition of marine microbial populations and developing a better understanding of the interplay between the functional genome content of microbial communities and abiotic factors. Here we present a comprehensive analysis of 24 datasets covering surface and depth-related environments at 11 sites around the world's oceans. The complete datasets comprises approximately 12 million sequences, totaling 5,358 Mb. Based on profiling patterns of Clusters of Orthologous Groups (COGs) of proteins, a core set of reference photic and aphotic depth-related COGs, and a collection of COGs that are associated with extreme oxygen limitation were defined. Their inferred functions were utilized as indicators to characterize the distribution of light- and oxygen-related biological activities in marine environments. The results reveal that, while light level in the water column is a major determinant of phenotypic adaptation in marine microorganisms, oxygen concentration in the aphotic zone has a significant impact only in extremely hypoxic waters. Phylogenetic profiling of the reference photic/aphotic gene sets revealed a greater variety of source organisms in the aphotic zone, although the majority of individual photic and aphotic depth-related COGs are assigned to the same taxa across the different sites. This increase in phylogenetic and functional diversity of the core aphotic related COGs most probably reflects selection for the utilization of a broad range of alternate energy sources in the absence of light.

Core microbial functional activities in ocean environments revealed by global metagenomic profiling analyses.

Directory of Open Access Journals (Sweden)

Ari J S Ferreira

Full Text Available Metagenomics-based functional profiling analysis is an effective means of gaining deeper insight into the composition of marine microbial populations and developing a better understanding of the interplay between the functional genome content of microbial communities and abiotic factors. Here we present a comprehensive analysis of 24 datasets covering surface and depth-related environments at 11 sites around the world's oceans. The complete datasets comprises approximately 12 million sequences, totaling 5,358 Mb. Based on profiling patterns of Clusters of Orthologous Groups (COGs of proteins, a core set of reference photic and aphotic depth-related COGs, and a collection of COGs that are associated with extreme oxygen limitation were defined. Their inferred functions were utilized as indicators to characterize the distribution of light- and oxygen-related biological activities in marine environments. The results reveal that, while light level in the water column is a major determinant of phenotypic adaptation in marine microorganisms, oxygen concentration in the aphotic zone has a significant impact only in extremely hypoxic waters. Phylogenetic profiling of the reference photic/aphotic gene sets revealed a greater variety of source organisms in the aphotic zone, although the majority of individual photic and aphotic depth-related COGs are assigned to the same taxa across the different sites. This increase in phylogenetic and functional diversity of the core aphotic related COGs most probably reflects selection for the utilization of a broad range of alternate energy sources in the absence of light.

Energetics of LMFBR core disruptive accidents

International Nuclear Information System (INIS)

Marchaterre, J.F.

1979-01-01

In general, in the design of fast reactor systems, containment design margins are specified by investigating the response of the containment to core disruptive accidents. The results of these analyses are then translated into criteria which the designers must meet. Currently, uniform and agreed upon criteria are lacking, and in this time while they are being developed, the designer should be aware of the considerations which go into the particular criteria he must work with, and participate in their development. This paper gives an overview of the current state of the art in assessing core disruptive accidents and the design implications of this process. (orig.)

Evaluation of Features of Development of Sports Way of Swimming of Students of Various Sports Specialization

Science.gov (United States)

Yermahanova, Amina; Nurmakhambetova, Dinara; Bozhig, Zhanbolat; Imanbetov, Amanbek

2016-01-01

Bachelor educational program "Physical culture and sport" must master special, substantive and core competencies, not only in the chosen specialization, but also in the basic sports, including "Swimming." It is a necessity due to the fact that the graduate program in order to protect their health and life should own at least…

The Zero-Reject Policy in Special Education: A Moral Analysis

Science.gov (United States)

Ladenson, Robert F.

2005-01-01

This article analyzes the zero-reject policy at the core of American special education law from the standpoint of morality, by examining the policy in terms of the following three moral theories: utilitarianism, Rawlsian Kantianism (justice as fairness) and neo-Aristotelianism, as developed recently by Martha Nussbaum in her capabilities account…

Development of concept and neutronic calculation method for large LMFBR core

International Nuclear Information System (INIS)

Shirakata, K.; Ishikawa, M.; Ikegami, T.; Sanda, T.; Kaneto, K.; Kawashima, M.; Kaise, Y.; Shirakawa, M.; Hibi, K.

1991-01-01

Presented in this paper is the state of the art of reactor physics R and Ds for the development of concept and neutronic calculation method for large Liquid Metal Fast Breeder Reactor (LMFBR) core. Physics characteristics of concepts for mixed oxide (MOX) fueled large FBR core were investigated by a series of benchmark critical experiments. Next, an adequacy and accuracy of the current neutronic calculation method was assessed by the experiments analyses, and then neutronic prediction accuracies by the method were evaluated for physics characteristics of the large core. Concerns on core development were discussed in terms of neutronics. (author)

A 3.55 keV line from DM →a→γ: predictions for cool-core and non-cool-core clusters

Energy Technology Data Exchange (ETDEWEB)

Conlon, Joseph P.; Powell, Andrew J. [Rudolf Peierls Centre for Theoretical Physics, University of Oxford, 1 Keble Road, Oxford, OX1 3NP (United Kingdom)

2015-01-13

We further study a scenario in which a 3.55 keV X-ray line arises from decay of dark matter to an axion-like particle (ALP), that subsequently converts to a photon in astrophysical magnetic fields. We perform numerical simulations of Gaussian random magnetic fields with radial scaling of the magnetic field magnitude with the electron density, for both cool-core 'Perseus' and non-cool-core 'Coma' electron density profiles. Using these, we quantitatively study the resulting signal strength and morphology for cool-core and non-cool-core clusters. Our study includes the effects of fields of view that cover only the central part of the cluster, the effects of offset pointings on the radial decline of signal strength and the effects of dividing clusters into annuli. We find good agreement with current data and make predictions for future analyses and observations.

Nitrogen partitioning during core-mantle differentiation

Science.gov (United States)

Speelmanns, I. M.; Schmidt, M. W.; Liebske, C.

2016-12-01

This study investiagtes nitrogen partitioing between metal and silicate melts as relevant for core segregation during the accretion of planetesimals into the Earth. On present day Earth, N belongs to the most important elements, as it is one of the key constituents of our atmosphere and forms the basis of life. However, the geochemistry of N, i.e. its distribution and isotopic fractionation between Earth's deep reservoirs is not well constrained. In order to determine the partitioning behaviour of N, a centrifuging piston cylinder was used to euqilibrate and then gravitationally separate metal-silicate melt pairs at 1250 °C, 1 GPa over the range of oxygen fugacities thought to have prevailied druing core segreagtion (IW-4 to IW). Complete segregation of the two melts was reached within 3 hours at 1000 g, the interface showing a nice meniscus The applied double capsule technique, using an outer metallic and inner non-metallic (mostly graphite) capsule, minimizes volatile loss over the course of the experiment compared to single non-metallic capsules. The two quenched melts were cut apart, cleaned at the outside and N concentrations of the melts were analysed on bulk samples by an elemental analyser. Nevertheless, the low amount of sample material and the N yield in the high pressure experiments required the developement of new analytical routines. Despite these experimental and analytical difficulties, we were able to determine a DNmetal/silicateof 13±0.25 at IW-1, N partitioning into the core froming metal. The few availible literature data [1],[2] suggest that N changes its compatibility favoring the silicate melt or magma ocean at around IW-2.5. In order to asses how much N may effectively be contained in the core and the silicate Earth, experiments characterizing N behaviour over the entire range of core formation condtitions are well under way. [1] Kadik et al., (2011) Geochemistry International 49.5: 429-438. [2] Roskosz et al., (2013) GCA 121: 15-28.

Core reilforced braided composite armour as a substitute to steel in concrete reinforcement

OpenAIRE

Fangueiro, Raúl; Sousa, Guilherme José Miranda de; Araújo, Mário Duarte de; Pereira, C. Gonilho; Jalali, Said

2006-01-01

This paper presents the work that is being done at the University of Minho concerning the development of brainded rods concrete reinforcement. Several samples of core reinforced braided fabrics have been produced varying the type of braided fabric (core reinforced and hybrid), the linear density of the core reinforcing yarns and the type of braiding structure (with or without ribs). The tensile properties of braided fabrics has also been analysed. Core reinforced braided composites rods were ...

Tank 241-T-201, core 192 analytical results for the final report

Energy Technology Data Exchange (ETDEWEB)

Nuzum, J.L.

1997-08-07

This document is the final laboratory report for Tank 241-T-201. Push mode core segments were removed from Riser 3 between April 24, 1997, and April 25, 1997. Segments were received and extruded at 222-S Laboratory. Analyses were performed in accordance with Tank 241-T-201 Push Mode Core Sampling and Analysis Plan (TSAP) (Hu, 1997), Letter of Instruction for Core Sample Analysis of Tanks 241-T-201, 241-T-202, 241-T-203, and 241-T-204 (LOI) (Bell, 1997), Additional Core Composite Sample from Drainable Liquid Samples for Tank 241-T-2 01 (ACC) (Hall, 1997), and Safety Screening Data Quality Objective (DQO) (Dukelow, et al., 1995). None of the subsamples submitted for total alpha activity (AT) or differential scanning calorimetry (DSC) analyses exceeded the notification limits stated in DQO. The statistical results of the 95% confidence interval on the mean calculations are provided by the Tank Waste Remediation Systems Technical Basis Group, and are not considered in this report.

Tank 241-T-204, core 188 analytical results for the final report

Energy Technology Data Exchange (ETDEWEB)

Nuzum, J.L.

1997-07-24

TANK 241-T-204, CORE 188, ANALYTICAL RESULTS FOR THE FINAL REPORT. This document is the final laboratory report for Tank 241 -T-204. Push mode core segments were removed from Riser 3 between March 27, 1997, and April 11, 1997. Segments were received and extruded at 222-8 Laboratory. Analyses were performed in accordance with Tank 241-T-204 Push Mode Core Sampling and analysis Plan (TRAP) (Winkleman, 1997), Letter of instruction for Core Sample Analysis of Tanks 241-T-201, 241- T-202, 241-T-203, and 241-T-204 (LAY) (Bell, 1997), and Safety Screening Data Qual@ Objective (DO) ODukelow, et al., 1995). None of the subsamples submitted for total alpha activity (AT) or differential scanning calorimetry (DC) analyses exceeded the notification limits stated in DO. The statistical results of the 95% confidence interval on the mean calculations are provided by the Tank Waste Remediation Systems Technical Basis Group and are not considered in this report.

Tank 241-T-201, core 192 analytical results for the final report

International Nuclear Information System (INIS)

Nuzum, J.L.

1997-01-01

This document is the final laboratory report for Tank 241-T-201. Push mode core segments were removed from Riser 3 between April 24, 1997, and April 25, 1997. Segments were received and extruded at 222-S Laboratory. Analyses were performed in accordance with Tank 241-T-201 Push Mode Core Sampling and Analysis Plan (TSAP) (Hu, 1997), Letter of Instruction for Core Sample Analysis of Tanks 241-T-201, 241-T-202, 241-T-203, and 241-T-204 (LOI) (Bell, 1997), Additional Core Composite Sample from Drainable Liquid Samples for Tank 241-T-2 01 (ACC) (Hall, 1997), and Safety Screening Data Quality Objective (DQO) (Dukelow, et al., 1995). None of the subsamples submitted for total alpha activity (AT) or differential scanning calorimetry (DSC) analyses exceeded the notification limits stated in DQO. The statistical results of the 95% confidence interval on the mean calculations are provided by the Tank Waste Remediation Systems Technical Basis Group, and are not considered in this report

Structural failure analysis of reactor vessels due to molten core debris

International Nuclear Information System (INIS)

Pfeiffer, P.A.

1993-01-01

Maintaining structural integrity of the reactor vessel during a postulated core melt accident is an important safety consideration in the design of the vessel. This paper addresses the failure predictions of the vessel due to thermal and pressure loadings from the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on a wet or dry cavity and pressurization of the vessel based on operating pressure or atmospheric (pipe break). The analyses considered both short term (minutes) and long term (days) failure modes. Short term failure modes include creep at elevated temperatures and plastic instabilities of the structure. Long term failure modes are caused by creep rupture that lead to plastic instability of the structure. The analyses predict the reactor vessel will remain intact after the core melt has deposited on the lower vessel head

TMI-2 core damage: a summary of present knowledge

International Nuclear Information System (INIS)

Owen, D.E.; Mason, R.E.; Meininger, R.D.; Franz, W.A.

1983-01-01

Extensive fuel damage (oxidation and fragmentation) has occurred and the top approx. 1.5 m of the center portion of the TMI-2 core has relocated. The fuel fragmentation extends outward to slightly beyond one-half the core radius in the direction examined by the CCTV camera. While the radial extent of core fragmentation in other directions was not directly observed, control and spider drop data and in-core instrument data suggest that the core void is roughly symmetrical, although there are a few indications of severe fuel damage extending to the core periphery. The core material fragmented into a broad range of particle sizes, extending down to a few microns. APSR movement data, the observation of damaged fuel assemblies hanging unsupported from the bottom of the reactor upper plenum structure, and the observation of once-molten stainless steel immediately above the active core indicate high temperatures (up to at least 1720 K) extended to the very top of the core. The relative lack of damage to the underside of the plenum structure implies a sharp temperature demarcation at the core/plenum interface. Filter debris and leadscrew deposit analyses indicate extensive high temperature core materials interaction, melting of the Ag-In-Cd control material, and transport of particulate control material to the plenum and out of the vessel

Characterization of rapid climate changes through isotope analyses of ice and entrapped air in the NEEM ice core

DEFF Research Database (Denmark)

Guillevic, Myriam

Greenland ice core have revealed the occurrence of rapid climatic instabilities during the last glacial period, known as Dansgaard-Oeschger (DO) events, while marine cores from the North Atlantic have evidenced layers of ice rafted debris deposited by icebergs melt, caused by the collapse...... mechanisms at play. Recent analytical developments have made possible to measure new paleoclimate proxies in Greenland ice cores. In this thesis we first contribute to these analytical developments by measuring the new innovative parameter 17O-excess at LSCE (Laboratoire des Sciences du Climatet de l......'Environnement, France). At the Centre for Ice and Climate (CIC, Denmark) we contribute to the development of a protocol for absolute referencing of methane gas isotopes, and making full air standard with known concentration and isotopic composition of methane. Then, air (δ15N) and water stable isotope measurements from...

Quantitative Analyses of Core Promoters Enable Precise Engineering of Regulated Gene Expression in Mammalian Cells

Science.gov (United States)

Ede, Christopher; Chen, Ximin; Lin, Meng-Yin; Chen, Yvonne Y.

2016-01-01

Inducible transcription systems play a crucial role in a wide array of synthetic biology circuits. However, the majority of inducible promoters are constructed from a limited set of tried-and-true promoter parts, which are susceptible to common shortcomings such as high basal expression levels (i.e., leakiness). To expand the toolbox for regulated mammalian gene expression and facilitate the construction of mammalian genetic circuits with precise functionality, we quantitatively characterized a panel of eight core promoters, including sequences with mammalian, viral, and synthetic origins. We demonstrate that this selection of core promoters can provide a wide range of basal gene expression levels and achieve a gradient of fold-inductions spanning two orders of magnitude. Furthermore, commonly used parts such as minimal CMV and minimal SV40 promoters were shown to achieve robust gene expression upon induction, but also suffer from high levels of leakiness. In contrast, a synthetic promoter, YB_TATA, was shown to combine low basal expression with high transcription rate in the induced state to achieve significantly higher fold-induction ratios compared to all other promoters tested. These behaviors remain consistent when the promoters are coupled to different genetic outputs and different response elements, as well as across different host-cell types and DNA copy numbers. We apply this quantitative understanding of core promoter properties to the successful engineering of human T cells that respond to antigen stimulation via chimeric antigen receptor signaling specifically under hypoxic environments. Results presented in this study can facilitate the design and calibration of future mammalian synthetic biology systems capable of precisely programmed functionality. PMID:26883397

The Colorado Plateau Coring Project: A Continuous Cored Non-Marine Record of Early Mesozoic Environmental and Biotic Change

Science.gov (United States)

Irmis, Randall; Olsen, Paul; Geissman, John; Gehrels, George; Kent, Dennis; Mundil, Roland; Rasmussen, Cornelia; Giesler, Dominique; Schaller, Morgan; Kürschner, Wolfram; Parker, William; Buhedma, Hesham

2017-04-01

The early Mesozoic is a critical time in earth history that saw the origin of modern ecosystems set against the back-drop of mass extinction and sudden climate events in a greenhouse world. Non-marine sedimentary strata in western North America preserve a rich archive of low latitude terrestrial ecosystem and environmental change during this time. Unfortunately, frequent lateral facies changes, discontinuous outcrops, and a lack of robust geochronologic constraints make lithostratigraphic and chronostratigraphic correlation difficult, and thus prevent full integration of these paleoenvironmental and paleontologic data into a regional and global context. The Colorado Plateau Coring Project (CPCP) seeks to remedy this situation by recovering a continuous cored record of early Mesozoic sedimentary rocks from the Colorado Plateau of the western United States. CPCP Phase 1 was initiated in 2013, with NSF- and ICDP-funded drilling of Triassic units in Petrified Forest National Park, northern Arizona, U.S.A. This phase recovered a 520 m core (1A) from the northern part of the park, and a 240 m core (2B) from the southern end of the park, comprising the entire Lower-Middle Triassic Moenkopi Formation, and most of the Upper Triassic Chinle Formation. Since the conclusion of drilling, the cores have been CT scanned at the University of Texas - Austin, and split, imaged, and scanned (e.g., XRF, gamma, and magnetic susceptibility) at the University of Minnesota LacCore facility. Subsequently, at the Rutgers University Core Repository, core 1A was comprehensively sampled for paleomagnetism, zircon geochronology, petrography, palynology, and soil carbonate stable isotopes. LA-ICPMS U-Pb zircon analyses are largely complete, and CA-TIMS U-Pb zircon, paleomagnetic, petrographic, and stable isotope analyses are on-going. Initial results reveal numerous horizons with a high proportion of Late Triassic-aged primary volcanic zircons, the age of which appears to be a close

Simulation experiment on the flooding behaviour of core melts: KATS-9

International Nuclear Information System (INIS)

Fieg, G.; Massier, H.; Schuetz, W.; Stegmaier, U.; Stern, G.

2000-11-01

For future Light Water Reactors special devices (core catchers) are being developed to prevent containment failure by basement erosion after reactor pressure vessel meltthrough during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher devices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent water cooling. A KATS-experiment has been performed to investigate the flooding behaviour of high temperature melts using alumina-iron thermite melts as a simulant. The oxidic thermite melt is conditioned by adding other oxides to simulate a realistic corium melt as close as possible in terms of liquidus and solidus temperatures. Before flooding with water, spreading of the separate oxidic and metallic melts has been done in one-dimensional channels with a silicate concrete as the substrate. The flooding rate was, in relation to the melt surface, identical to the flooding rate in EPR. (orig.) [de

Neutronic analysis of a reference LEU core for Pakistan research reactor using oxide fuel

International Nuclear Information System (INIS)

Akhtar, K.M.; Qazi, M.K.; Bokhari, I.H.; Khan, L.A.; Pervez, S.

1988-07-01

Neutronic analysis of a 10 MW reference core for PARR, having 28 fresh LEU fuel elements arranged in a 6x5 configuration has been carried out using standard computer codes WIMS-D, EXTERMINATOR-II, and CITATION. Total nuclear power peaking of 3.2 has bee found to occur in the fuel plate adjacent to the water filled central flux trap at the depth of 43.8 cm from the top of the active core. Replacement of water in central flux trap with an aluminum block, having a 50 mm diameter water filled irradiation channel changes the flux profiles in fuel, core side flux trap and reflector. The thermal flux in the central flux trap decreases by about 53%. Therefore some of the fuel elements will have to be removed and the new configuration has to be analysed to determine the first operating core. However, after achieving some burn-up and confirmation from thermal hydraulic analysis, the core configuration analysed, will be the final working core. (orig./A.B.)

Lunar core formation: New constraints from metal-silicate partitioning of siderophile elements

NARCIS (Netherlands)

Rai, N.; van Westrenen, W.

2014-01-01

Analyses of Apollo era seismograms, lunar laser ranging data and the lunar moment of inertia suggest the presence of a small, at least partially molten Fe-rich metallic core in the Moon, but the chemical composition and formation conditions of this core are not well constrained. Here, we assess

Test of In-core Flux Detectors in KNK II

CERN Document Server

Hoppe, P

1979-01-01

The development of in-core detectors for Liquid Metal Fast Breeder Reactors (LMFBRs) is still in an early stage, and little operation experience is available. Therefore self-powered neutron and gamma detectors and neutron sensitive ionization chambers -especially developed for LMFBRs- have been tested in the Fast Sodium Cooled Test Reactor KNK II. Seven flux detectors have been installed in the core of KNK II by means of a special test rig. Five of them failed already within the first week during operation in the reactor. Due to measurements of electrical resistances and capacities, sodium penetrating into the detectors or cables probably seems to be the cause. As tests prior to the installation in the core proved the tightness of all detectors, it is suspected that small cracks have developed in the detector casings or in the outer cable sheaths during their exposure to the hot coolant. Two ionization chambers did not show these faults. However, one of them failed because the saturation current plateau disap...

Insights from Severe Accident Analyses for Verification of VVER SAMG

Energy Technology Data Exchange (ETDEWEB)

Gaikwad, A. J.; Rao, R. S.; Gupta, A.; Obaidurrahaman, K., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

2014-10-15

The severe accident analyses of simultaneous rupture of all four steam lines (case-a), simultaneous occurrence of LOCA with SBO (case-b) and Station blackout (case-c) were performed with the computer code ASTEC V2r2 for a typical VVER-1000. The results obtained will be used for verification of sever accident provisions and Severe Accident Management Guidelines (SAMG). Auxiliary feed water and emergency core cooling systems are modelled as boundary conditions. The ICARE module is used to simulate the reactor core, which is divided into five radial regions by grouping similarly powered fuel assemblies together. Initially, CESAR module computes thermal hydraulics in primary and secondary circuits. As soon as core uncovery begins, the ICARE module is actuated based on certain parameters, and after this, ICARE module computes the thermal hydraulics in the core, bypass, downcomer and the lower plenum. CESAR handles the remaining components in the primary and secondary loops. CPA module is used to simulate the containment and to predict the thermal-hydraulic and hydrogen behaviour in the containment. The accident sequences were selected in such a way that they cover low/high pressure and slow/fast core damage progression events. Events simulated included slow progression events with high pressure and fast accident progression with low primary pressure. Analysis was also carried out for the case of SBO with the opening of the PORVs when core exit temperature exceeds certain value as part of SAMG. Time step sensitivity study was carried out for LOCA with SBO. In general the trends and magnitude of the parameters are as expected. The key results of the above analyses are presented in this paper. (author)

Core design with respect to the safety concept

International Nuclear Information System (INIS)

Kollmar, W.

1981-01-01

In the present paper the following topics are dealt with: Principles of reactor core design and optimization, fuel management and safety concept for higher cycles and results of risk analyses (e.g. rod ejection, steam line break etc.) (RW)

Cores to the rescue: how old cores enable new science

Science.gov (United States)

Ito, E.; Noren, A. J.; Brady, K.

2016-12-01

The value of archiving scientific specimens and collections for the purpose of enabling further research using new analytical techniques, resolving conflicting results, or repurposing them for entirely new research, is often discussed in abstract terms. We all agree that samples with adequate metadata ought to be archived systematically for easy access, for a long time and stored under optimal conditions. And yet, as storage space fills, there is a temptation to cull the collection, or when a researcher retires, to discard the collection unless the researcher manages to make his or her own arrangement for the collection to be accessioned elsewhere. Nobody has done anything with these samples in over 20 years! Who would want them? It turns out that plenty of us do want them, if we know how to find them and if they have sufficient metadata to assess past work and suitability for new analyses. The LacCore collection holds over 33 km of core from >6700 sites in diverse geographic locations worldwide with samples collected as early as 1950s. From these materials, there are many examples to illustrate the scientific value of archiving geologic samples. One example that benefitted Ito personally were cores from Lakes Mirabad and Zeribar, Iran, acquired in 1963 by Herb Wright and his associates. Several doctoral and postdoctoral students generated and published paleoecological reconstructions based on cladocerans, diatoms, pollen or plant macrofossils, mostly between 1963 and 1967. The cores were resampled in 1990s by a student being jointly advised by Wright and Ito for oxygen isotope analysis of endogenic calcite. The results were profitably compared with pollen and the results published in 2001 and 2006. From 1979 until very recently, visiting Iran for fieldwork was not pallowed for US scientists. Other examples will be given to further illustrate the power of archived samples to advance science.

Improvement of open and semi-open core wall system in tall buildings by closing of the core section in the last story

Science.gov (United States)

Kheyroddin, A.; Abdollahzadeh, D.; Mastali, M.

2014-09-01

Increasing number of tall buildings in urban population caused development of tall building structures. One of the main lateral load resistant systems is core wall system in high-rise buildings. Core wall system has two important behavioral aspects where the first aspect is related to reduce the lateral displacement by the core bending resistance and the second is governed by increasing of the torsional resistance and core warping of buildings. In this study, the effects of closed section core in the last story have been considered on the behavior of models. Regarding this, all analyses were performed by ETABS 9.2.v software (Wilson and Habibullah). Considering (a) drift and rotation of the core over height of buildings, (b) total and warping stress in the core body, (c) shear in beams due to warping stress, (d) effect of closing last story on period of models in various modes, (e) relative displacement between walls in the core system and (f) site effects in far and near field of fault by UBC97 spectra on base shear coefficient showed that the bimoment in open core is negative in the last quarter of building and it is similar to wall-frame structures. Furthermore, analytical results revealed that closed section core in the last story improves behavior of the last quarter of structure height, since closing of core section in the last story does not have significant effect on reducing base shear value in near and far field of active faults.

Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review

International Nuclear Information System (INIS)

Wulff, W.

1990-01-01

A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes in the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs

Modeling of exchange bias in the antiferromagnetic (core)/ferromagnetic (shell) nanoparticles with specialized shapes

International Nuclear Information System (INIS)

Hu Yong; Liu Yan; Du An

2011-01-01

Zero-field-cooled (ZFC) and field-cooled (FC) hysteresis loops of egg- and ellipsoid-shaped nanoparticles with inverted ferromagnetic (FM)-antiferromagnetic (AFM) core-shell morphologies are simulated using a modified Monte Carlo method, which takes into account both the thermal fluctuations and energy barriers during the rotation of spin. Pronounced exchange bias (EB) fields and reduced coercivities are obtained in the FC hysteresis loops. The analysis of the microscopic spin configurations allows us to conclude that the magnetization reversal occurs by means of the nucleation process during both the ZFC and FC hysteresis branches. The nucleation takes place in the form of 'sparks' resulting from the energy competition and the morphology of the nanoparticle. The appearance of EB in the FC hysteresis loops is only dependent on that the movements of 'sparks' driven by magnetic field at both branches of hysteresis loops are not along the same axis, which is independent of the strength of AFM anisotropy. The tilt of 'spark' movement with respect to the symmetric axis implies the existence of additional unidirectional anisotropy at the AFM/FM interfaces as a consequence of the surplus magnetization in the AFM core, which is the commonly accepted origin of EB. Our simulations allow us to clarify the microscopic mechanisms of the observed EB behavior, not accessible in experiments. - Highlights: → A modified Monte Carlo method considers thermal fluctuations and energy barriers. → Egg and ellipsoid nanoparticles with inverted core-shell morphology are studied. → Pronounced exchange bias fields and reduced coercivities may be detected. → 'Sparks' representing nucleation sites due to energy competition are observed. → 'Sparks' can reflect or check directly and vividly the origin of exchange bias.

Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts

International Nuclear Information System (INIS)

Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji

2010-01-01

The Korea Atomic Energy Research Institute (KAERI) has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study

Candidate molten salt investigation for an accelerator driven subcritical core

Science.gov (United States)

Sooby, E.; Baty, A.; Beneš, O.; McIntyre, P.; Pogue, N.; Salanne, M.; Sattarov, A.

2013-09-01

We report a design for accelerator-driven subcritical fission in a molten salt core (ADSMS) that utilizes a fuel salt composed of NaCl and transuranic (TRU) chlorides. The ADSMS core is designed for fast neutronics (28% of neutrons >1 MeV) to optimize TRU destruction. The choice of a NaCl-based salt offers benefits for corrosion, operating temperature, and actinide solubility as compared with LiF-based fuel salts. A molecular dynamics (MD) code has been used to estimate properties of the molten salt system which are important for ADSMS design but have never been measured experimentally. Results from the MD studies are reported. Experimental measurements of fuel salt properties and studies of corrosion and radiation damage on candidate metals for the core vessel are anticipated. A special thanks is due to Prof. Paul Madden for introducing the ADSMS group to the concept of using the molten salt as the spallation target, rather than a conventional heavy metal spallation target. This feature helps to optimize this core as a Pu/TRU burner.

Recent enhancements of the INSIGHT integrated in-core fuel management tool

International Nuclear Information System (INIS)

Akio, Yamamoto

2001-01-01

Recent enhancements of the INSIGHT system are described in this paper. The INSIGHT system is an integrated in-core fuel management tool for pressurized water reactors (PWRs) runs on UNIX workstations. The INSIGHT system provides various capabilities which contribute to reduce fuel cycle cost and workload of in-core fuel management tasks, i.e. core follow calculations, interactive loading pattern design, automated multicycle analysis and interface between detailed core calculation codes. To minimize engineers' workload, most of input data for analysis modules are automatically generated by the INSIGHT system through specification of calculation conditions in the graphic user interface. Recent enhancements of the INSIGHT system are mainly focused to improve efficiency of loading pattern optimization and flexibility of multicycle analyses. To increase optimization efficiency, a parallel calculation capability, various optimization theories, extension of heuristic rules, screening by neural networks and so on were incorporated in the loading pattern optimization module. The multicycle analyses module was rewritten to increase flexibility such as cycle dependent specification of loading pattern search methods and so on. The INSIGHT system is currently used by Japanese utilities not only for regular in-core fuel management tasks but also for strategic fuel management studies to reduce fuel cycle cost

Financial Therapy and Planning for Families with Special Needs Children

Directory of Open Access Journals (Sweden)

Mitzi Lauderdale

2012-06-01

Full Text Available This study examines factors associated with the likelihood of having a plan that includes a special needs trust among families that have disabled minor children. Descriptive analyses indicate that the top two reasons families provide for not having a plan are the inability to save and no perceived need. Among families that do indicate having a plan, most do not include a special needs trust. Multivariate analyses reveal that professional involvement (financial, legal, and mental health professionals is a key factor to increasing the likelihood of having a plan with a special needs trust. Families that have met with a financial advisor are 23 times more likely, and families who are encouraged to create a plan by a mental health professional are almost three times more likely, to have a plan that includes a special needs trust. Results from this study suggest that financial therapists are uniquely positioned to educate and ensure that appropriate plans are in place to provide for the future of children with special needs.

AP1000 core design with 50% MOX loading

International Nuclear Information System (INIS)

Fetterman, Robert J.

2009-01-01

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core design and a mixed MOX/UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance

AP1000 core design with 50% MOX loading

International Nuclear Information System (INIS)

Fetterman, Robert J.

2008-01-01

The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core and a mixed MOX / UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

Quality assurance in the removal and transport of the TMI-2 [Three Mile Island Unit 2] core

International Nuclear Information System (INIS)

Hayes, G.R.; Marsden, J.F.

1988-01-01

The March 1979 accident at Three Mile Island Unit 2 (TMI-2) damaged the core of the reactor. One of the major cleanup activities involves removal of the damaged core from the reactor and transporting it from the TMI-2 site near Middletown, Pennsylvania, to the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho. Removal and transport of the damaged core necessitated the development of much specialized equipment. This paper focuses on the role quality assurance (QA) played in the design, fabrication, acceptance, and use of three important pieces of core debris removal and transportation equipment: (1) the core boring machine, (2) the fuel debris canisters, (3) the NuPac 125-B rail cask and handling equipment

KATS experiments to simulate corium spreading in the EPR core catcher concept

International Nuclear Information System (INIS)

Eppinger, B.; Fieg, G.; Schuetz, W.; Stegmaier, U.

2001-01-01

In future Light Water Reactors special devices (core catchers) might be required to prevent containment failure by basement erosion after reactor pressure vessel melt-through during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher de-vices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent cooling by flooding with water. Therefore a series of experiments to investigate high temperature melt spreading on flat surfaces has been carried out using alumina-iron thermite melts as a simulant. The oxidic thermite melt is conditioned by adding other oxides to simulate a realistic corium melt as close as possible. Spreading of oxidic and metallic melts have been performed in one- and two-dimensional geometry. Substrates were chemically inert ceramic layers, dry concrete and concrete with a shallow water layer on top. (authors)

Core OCD Symptoms: Exploration of Specificity and Relations with Psychopathology

Science.gov (United States)

Stasik, Sara M.; Naragon-Gainey, Kristin; Chmielewski, Michael; Watson, David

2012-01-01

Obsessive-compulsive disorder (OCD) is a heterogeneous condition, comprised of multiple symptom domains. This study used aggregate composite scales representing three core OCD dimensions (Checking, Cleaning, Rituals), as well as Hoarding, to examine the discriminant validity, diagnostic specificity, and predictive ability of OCD symptom scales. The core OCD scales demonstrated strong patterns of convergent and discriminant validity – suggesting that these dimensions are distinct from other self-reported symptoms – whereas hoarding symptoms correlated just as strongly with OCD and non-OCD symptoms in most analyses. Across analyses, our results indicated that Checking is a particularly strong, specific marker of OCD diagnosis, whereas the specificity of Cleaning and Hoarding to OCD was less strong. Finally, the OCD Checking scale was the only significant predictor of OCD diagnosis in logistic regression analyses. Results are discussed with regard to the importance of assessing OCD symptom dimensions separately and implications for classification. PMID:23026094

Multi-core events in cosmic-ray induced interactions with lead at around 10 TeV

International Nuclear Information System (INIS)

Amato, N.; Arata, N.

1989-01-01

The analysis is made on the cosmic-ray induced interactions with lead at around 10 TeV on the basis of emulsion chamber data at Chacaltaya. A special attention is paid to the events detected as multi-cores under the spatial resolution of a few tens of microns. The observation of six double-core events and two triple-core events with the average invariant mass of 1.8 GeV/c 2 leads to the estimation on production frequency of such multicores as about 5% at 10 TeV at the atmospheric depth 540 gr/cm 2 . (author)

Back up core designs for the experimental multi-purpose VHTR

International Nuclear Information System (INIS)

Aochi, Tetsuo; Yasuno, Takehiko; Miyamoto, Yoshiaki; Shindo, Ryuichi; Ikushima, Takeshi

1979-02-01

For the Experimental Multi-Purpose Very High Temperature Reactor (thermal power 50 MW and reactor outlet helium temperature 1000 0 C), design studies have been made of two backup cores loaded with new-type fuel elements. The purpose is to improve core operational characteristics, especially in thermohydraulics, of the reference design core consisting of pin-in-block type fuel elements having externally cooled hollow fuel rods. In this report are described the design principles and the analyses made of nuclear, thermal and hydraulic, fuel, and safety performances to determine the backup fuel and core design parameters. The first backup core (SP fuel core) is composed of fuel elements with internally cooled fuel rods (semi-pin), 36 rods in each standard element and 18 rods in each control element. The second backup core (MH fuel core) is composed of multihole fuel elements. 102 fuel and 54 coolant holes in each standard element and 30 fuel and 18 coolant holes in each control element. Either of the cores has 73 fuel columns 4 m high; the arrangement of active core and reactor internal structures is the same as that in the reference design. The backup cores meet nearly all design requirements of the VHTR, permitting the rated power operation with coolant Reynolds number of over 10,000 in the SP core and over 6,000 in the MH core. (author)

Critical experiments on enriched uranium graphite moderated cores

International Nuclear Information System (INIS)

Kaneko, Yoshihiko; Akino, Fujiyoshi; Kitadate, Kenji; Kurokawa, Ryosuke

1978-07-01

A variety of 20 % enriched uranium loaded and graphite-moderated cores consisting of the different lattice cells in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments systematically. In the present report, the experimental results for homogeneously or heterogeneously fuel loaded cores and for simulation core of the experimental reactor for a multi-purpose high temperature reactor are filed so as to be utilized for evaluating the accuracy of core design calculation for the experimental reactor. The filed experimental data are composed of critical masses of uranium, kinetic parameters, reactivity worths of the experimental control rods and power distributions in the cores with those rods. Theoretical analyses are made for the experimental data by adopting a simple ''homogenized cylindrical core model'' using the nuclear data of ENDF/B-III, which treats the neutron behaviour after smearing the lattice cell structure. It is made clear from a comparison between the measurement and the calculation that the group constants and fundamental methods of calculations, based on this theoretical model, are valid for the homogeneously fuel loaded cores, but not for both of the heterogeneously fuel loaded cores and the core for simulation of the experimental reactor. Then, it is pointed out that consideration to semi-homogeneous property of the lattice cells for reactor neutrons is essential for high temperature graphite-moderated reactors using dispersion fuel elements of graphite and uranium. (author)

Neutron dynamics of fast-spectrum dedicated cores for waste transmutation

International Nuclear Information System (INIS)

Massara, S.

2002-04-01

Among different scenarios achieving minor actinide transmutation, the possibility of double strata scenarios with critical, fast spectrum, dedicated cores must be checked and quantified. In these cores, the waste fraction has to be at the highest level compatible with safety requirements during normal operation and transient conditions. As reactivity coefficients are poor in such critical cores (low delayed neutron fraction and Doppler feed-back, high coolant void coefficient), their dynamic behaviour during transient conditions must be carefully analysed. Three nitride-fuel configurations have been analysed: two liquid metal-cooled (sodium and lead) and a particle-fuel helium-cooled one. A dynamic code, MAT4 DYN, has been developed during the PhD thesis, allowing the study of loss of flow, reactivity insertion and loss of coolant accidents, and taking into account two fuel geometries (cylindrical and spherical) and two thermal-hydraulics models for the coolant (incompressible for liquid metals and compressible for helium). Dynamics calculations have shown that if the fuel nature is appropriately chosen (letting a sufficient margin during transients), this can counterbalance the bad state of reactivity coefficients for liquid metal-cooled cores, thus proving the interest of this kind of concept. On the other side, the gas-cooled core dynamics is very badly affected by the high value of the helium void coefficient (which is a consequence of the choice of a hard spectrum), this effect being amplified by the very low thermal inertia of particle-fuel design. So, a new kind of concept should be considered for a helium-cooled fast-spectrum dedicated core. (authors)

Vocational Didactics: Core Assumptions and Approaches from Denmark, Germany, Norway, Spain and Sweden

Science.gov (United States)

Gessler, Michael; Moreno Herrera, Lázaro

2015-01-01

The design of vocational didactics has to meet special requirements. Six core assumptions are identified: outcome orientation, cultural-historical embedding, horizontal structure, vertical structure, temporal structure, and the changing nature of work. Different approaches and discussions from school-based systems (Spain and Sweden) and dual…

Probabilistic approach in treatment of deterministic analyses results of severe accidents

International Nuclear Information System (INIS)

Krajnc, B.; Mavko, B.

1996-01-01

Severe accidents sequences resulting in loss of the core geometric integrity have been found to have small probability of the occurrence. Because of their potential consequences to public health and safety, an evaluation of the core degradation progression and the resulting effects on the containment is necessary to determine the probability of a significant release of radioactive materials. This requires assessment of many interrelated phenomena including: steel and zircaloy oxidation, steam spikes, in-vessel debris cooling, potential vessel failure mechanisms, release of core material to the containment, containment pressurization from steam generation, or generation of non-condensable gases or hydrogen burn, and ultimately coolability of degraded core material. To asses the answer from the containment event trees in the sense of weather certain phenomenological event would happen or not the plant specific deterministic analyses should be performed. Due to the fact that there is a large uncertainty in the prediction of severe accidents phenomena in Level 2 analyses (containment event trees) the combination of probabilistic and deterministic approach should be used. In fact the result of the deterministic analyses of severe accidents are treated in probabilistic manner due to large uncertainty of results as a consequence of a lack of detailed knowledge. This paper discusses approach used in many IPEs, and which assures that the assigned probability for certain question in the event tree represent the probability that the event will or will not happen and that this probability also includes its uncertainty, which is mainly result of lack of knowledge. (author)

Thermal interaction of core melt debris with the TMI-2 baffle, core-former, and lower head structures

International Nuclear Information System (INIS)

Cronenberg, A.W.; Tolman, E.L.

1987-09-01

Recent inspection of the TMI-2 core-former baffle walls (vertical), former plates (horizontal), and lower plenum has been conducted to assess potential damage to these structures. Video observations show evidence of localized melt failure of the baffle walls, whereas fiberoptics data indicate the presence of resolidified debris on the former plates. Lower plenum inspection also confirms the presence of 20 tons or more of core debris in the lower plenum. These data indicate massive core melt relocation and the potential for melt attack on vessel structural components. This report presents analyses aimed at developing an understanding of melt relocation behavior and damage progression to TMI-2 vessel components. Thermal analysis indicates melt-through of the baffle plates, but maintenance of structural integrity of the former plates and lower head. Differences in the damage of these structures is attributed largely to differences in contact time with melt debris and pressure of water. 29 refs., 17 figs., 9 tabs

Safety analyses for sodium-cooled fast reactors with pelletized and sphere-pac oxide fuels within the FP-7 European project PELGRIMM - 15386

International Nuclear Information System (INIS)

Maschek, W.; Andriolo, L.; Matzerath-Boccaccini, C.; Delage, F.; Parisi, C.; Del Nevo, A.; Abbate, G.; Schmitt, D.

2015-01-01

The European FP-7 project PELGRIMM addresses the development of Minor-Actinide (MA) bearing oxide fuel for Sodium-cooled Fast Reactors. Optionally, both MA homogeneous recycling and heterogeneous recycling is investigated with pellet and sphere-pac fuel. A first safety assessment of sphere-pac fuelled cores should be given in the Work Package 4 of the project. This assessment is in continuity with the former FP-7 CP-ESFR project. Within the CP-ESFR project the CONF2 core design has been developed characterized by a core with a large upper sodium plenum to reduce the coolant void worth. This optimized core has been chosen for the safety analyses in PELGRIMM. The task within the PELGRIMM project is thus a safety assessment of the CONF2 core loaded either with pellets or with sphere-pac fuel. The investigations started with the design of the CONF2 core with sphere-pac fuel and the determination of core safety parameters and burn-up behavior. The neutronic analyses have been performed with the MCNPX code. Variants of the CONF2 core contain up to 4% Am in the fuel. The results revealed an extended void worth (core + upper plenum) for an Am free core of 1 up to 3 dollars for the 4% Am core. Thermal-hydraulic design analyses have been performed by RELAP5-3D. The accident simulations should be performed by different codes, some of which focus on the initiation phase of the accident, as SAS4A, BELLA and the MAT5DYN code, whereas the SIMMER-III code will also deal with the later accident phases and a potential whole core melting. The codes had to be adapted to the specifics of the sphere-pac fuel, in particular to the thermal conductivity and gap conditions. Analyses showed that the safety assessment has to take into account two main phases. Starting up the core, the green fuel shows a reduced fuel thermal conductivity. After restructuring within a couple of hours, the thermal conductivity recovers and the fuel temperature decreases. The main objective of the safety analyses

Performance analyses of naval ships based on engineering level of simulation at the initial design stage

Directory of Open Access Journals (Sweden)

Dong-Hoon Jeong

2017-07-01

Full Text Available Naval ships are assigned many and varied missions. Their performance is critical for mission success, and depends on the specifications of the components. This is why performance analyses of naval ships are required at the initial design stage. Since the design and construction of naval ships take a very long time and incurs a huge cost, Modeling and Simulation (M & S is an effective method for performance analyses. Thus in this study, a simulation core is proposed to analyze the performance of naval ships considering their specifications. This simulation core can perform the engineering level of simulations, considering the mathematical models for naval ships, such as maneuvering equations and passive sonar equations. Also, the simulation models of the simulation core follow Discrete EVent system Specification (DEVS and Discrete Time System Specification (DTSS formalisms, so that simulations can progress over discrete events and discrete times. In addition, applying DEVS and DTSS formalisms makes the structure of simulation models flexible and reusable. To verify the applicability of this simulation core, such a simulation core was applied to simulations for the performance analyses of a submarine in an Anti-SUrface Warfare (ASUW mission. These simulations were composed of two scenarios. The first scenario of submarine diving carried out maneuvering performance analysis by analyzing the pitch angle variation and depth variation of the submarine over time. The second scenario of submarine detection carried out detection performance analysis by analyzing how well the sonar of the submarine resolves adjacent targets. The results of these simulations ensure that the simulation core of this study could be applied to the performance analyses of naval ships considering their specifications.

Severe accident recriticality analyses (SARA)

DEFF Research Database (Denmark)

Frid, W.; Højerup, C.F.; Lindholm, I.

2001-01-01

with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality-both super-prompt power bursts and quasi steady-state power......Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies......, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g(-1), was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s(-1). In most cases, however, the predicted energy deposition was smaller, below...

Clustering of Pan- and Core-genome of Lactobacillus provides Novel Evolutionary Insights for Differentiation.

Science.gov (United States)

Inglin, Raffael C; Meile, Leo; Stevens, Marc J A

2018-04-24

Bacterial taxonomy aims to classify bacteria based on true evolutionary events and relies on a polyphasic approach that includes phenotypic, genotypic and chemotaxonomic analyses. Until now, complete genomes are largely ignored in taxonomy. The genus Lactobacillus consists of 173 species and many genomes are available to study taxonomy and evolutionary events. We analyzed and clustered 98 completely sequenced genomes of the genus Lactobacillus and 234 draft genomes of 5 different Lactobacillus species, i.e. L. reuteri, L. delbrueckii, L. plantarum, L. rhamnosus and L. helveticus. The core-genome of the genus Lactobacillus contains 266 genes and the pan-genome 20'800 genes. Clustering of the Lactobacillus pan- and core-genome resulted in two highly similar trees. This shows that evolutionary history is traceable in the core-genome and that clustering of the core-genome is sufficient to explore relationships. Clustering of core- and pan-genomes at species' level resulted in similar trees as well. Detailed analyses of the core-genomes showed that the functional class "genetic information processing" is conserved in the core-genome but that "signaling and cellular processes" is not. The latter class encodes functions that are involved in environmental interactions. Evolution of lactobacilli seems therefore directed by the environment. The type species L. delbrueckii was analyzed in detail and its pan-genome based tree contained two major clades whose members contained different genes yet identical functions. In addition, evidence for horizontal gene transfer between strains of L. delbrueckii, L. plantarum, and L. rhamnosus, and between species of the genus Lactobacillus is presented. Our data provide evidence for evolution of some lactobacilli according to a parapatric-like model for species differentiation. Core-genome trees are useful to detect evolutionary relationships in lactobacilli and might be useful in taxonomic analyses. Lactobacillus' evolution is directed

Linguistic explanation and domain specialization: a case study in bound variable anaphora.

Science.gov (United States)

Adger, David; Svenonius, Peter

2015-01-01

The core question behind this Frontiers research topic is whether explaining linguistic phenomena requires appeal to properties of human cognition that are specialized to language. We argue here that investigating this issue requires taking linguistic research results seriously, and evaluating these for domain-specificity. We present a particular empirical phenomenon, bound variable interpretations of pronouns dependent on a quantifier phrase, and argue for a particular theory of this empirical domain that is couched at a level of theoretical depth which allows its principles to be evaluated for domain-specialization. We argue that the relevant principles are specialized when they apply in the domain of language, even if analogs of them are plausibly at work elsewhere in cognition or the natural world more generally. So certain principles may be specialized to language, though not, ultimately, unique to it. Such specialization is underpinned by ultimately biological factors, hence part of UG.

Facile fabrication of siloxane @ poly (methylacrylic acid) core-shell microparticles with different functional groups

Energy Technology Data Exchange (ETDEWEB)

Zhao, Zheng-Bai; Tai, Li; Zhang, Da-Ming; Jiang, Yong, E-mail: yj@seu.edu.cn [Southeast University, School of Chemistry and Chemical Engineering (China)

2017-02-15

Siloxane @ poly (methylacrylic acid) core-shell microparticles with functional groups were prepared by a facile hydrolysis-condensation method in this work. Three different silane coupling agents 3-methacryloxypropyltrimethoxysilane (MPS), 3-triethoxysilylpropylamine (APTES), and 3-glycidoxypropyltrimethoxysilane (GPTMS) were added along with tetraethoxysilane (TEOS) into the polymethylacrylic acid (PMAA) microparticle ethanol dispersion to form the Si@PMAA core-shell microparticles with different functional groups. The core-shell structure and the surface special functional groups of the resulting microparticles were measured by transmission electron microscopy and FTIR. The sizes of these core-shell microparticles were about 350–400 nm. The corresponding preparation conditions and mechanism were discussed in detail. This hydrolysis-condensation method also could be used to functionalize other microparticles which contain active groups on the surface. Meanwhile, the Si@PMAA core-shell microparticles with carbon-carbon double bonds and amino groups have further been applied to prepare hydrophobic coatings.

Facile fabrication of siloxane @ poly (methylacrylic acid) core-shell microparticles with different functional groups

International Nuclear Information System (INIS)

Zhao, Zheng-Bai; Tai, Li; Zhang, Da-Ming; Jiang, Yong

2017-01-01

Siloxane @ poly (methylacrylic acid) core-shell microparticles with functional groups were prepared by a facile hydrolysis-condensation method in this work. Three different silane coupling agents 3-methacryloxypropyltrimethoxysilane (MPS), 3-triethoxysilylpropylamine (APTES), and 3-glycidoxypropyltrimethoxysilane (GPTMS) were added along with tetraethoxysilane (TEOS) into the polymethylacrylic acid (PMAA) microparticle ethanol dispersion to form the Si@PMAA core-shell microparticles with different functional groups. The core-shell structure and the surface special functional groups of the resulting microparticles were measured by transmission electron microscopy and FTIR. The sizes of these core-shell microparticles were about 350–400 nm. The corresponding preparation conditions and mechanism were discussed in detail. This hydrolysis-condensation method also could be used to functionalize other microparticles which contain active groups on the surface. Meanwhile, the Si@PMAA core-shell microparticles with carbon-carbon double bonds and amino groups have further been applied to prepare hydrophobic coatings.

Laser Heating of the Core-Shell Nanowires

Science.gov (United States)

Astefanoaei, Iordana; Dumitru, Ioan; Stancu, Alexandru

2016-12-01

The induced thermal stress in a heating process is an important parameter to be known and controlled in the magnetization process of core-shell nanowires. This paper analyses the stress produced by a laser heating source placed at one end of a core-shell type structure. The thermal field was computed with the non-Fourier heat transport equation using a finite element method (FEM) implemented in Comsol Multiphysics. The internal stresses are essentially due to thermal gradients and different expansion characteristics of core and shell materials. The stress values were computed using the thermo elastic formalism and are depending on the laser beam parameters (spot size, power etc.) and system characteristics (dimensions, thermal characteristics). Stresses in the GPa range were estimated and consequently we find that the magnetic state of the system can be influenced significantly. A shell material as the glass which is a good thermal insulator induces in the magnetic core, the smaller stresses and consequently the smaller magnetoelastic energy. These results lead to a better understanding of the switching process in the magnetic materials.

Nuclear power plants: Results of recent safety analyses

International Nuclear Information System (INIS)

Steinmetz, E.

1987-01-01

The contributions deal with the problems posed by low radiation doses, with the information currently available from analyses of the Chernobyl reactor accident, and with risk assessments in connection with nuclear power plant accidents. Other points of interest include latest results on fission product release from reactor core or reactor building, advanced atmospheric dispersion models for incident and accident analyses, reliability studies on safety systems, and assessment of fire hazard in nuclear installations. The various contributions are found as separate entries in the database. (DG) [de

Special Needs Adoption and Foster Exigencies (SAFE: A Model for Case Managers

Directory of Open Access Journals (Sweden)

Kristen Linton

2014-05-01

Full Text Available Children with special needs disproportionately receive child welfare services in out-of-home placements, such as foster and adoptive homes. This theoretical model has been developed to describe or explain exigencies of adoptive and foster families of children with special needs (n = 82. A web content analysis, including theme, feature, link, exchange, and language analyses, of online discussion forums of adoptive and foster parents of children with special needs using a phenomenological framework was conducted. Inductive and quantitative web content analyses were conducted on themes. Parenting concerns were clustered into two main themes, disability and placement issues, and focused on children’s pre and post placement needs. A phenomenological analysis resulted in the development of the Special Needs Adoption and Foster Exigencies (SAFE, which outlines exigencies of adoptive and foster parents of children with special needs during engagement, assessment, and intervention phases of case management.

A nodal Grean's function method of reactor core fuel management code, NGCFM2D

International Nuclear Information System (INIS)

Li Dongsheng; Yao Dong.

1987-01-01

This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

77 FR 36409 - ``Specially Designed'' Definition

Science.gov (United States)

2012-06-19

....''' The EAR defines ``development'' as ``related to all stages prior to serial production, such as: design, design research, design analyses, design concepts, assembly and testing of prototypes, pilot production... specific types of equipment for manufacturing semiconductor devices or materials, and specially designed...

Advanced Small-Safe Long-Life Lead Cooled Reactor Cores for Future Nuclear Energy

Energy Technology Data Exchange (ETDEWEB)

Kim, Jin Hyeong; Hong, Ser Gi [Kyung Hee University, Seoul (Korea, Republic of)

2014-10-15

One of the reasons for use of the lead or lead-bismuth alloy coolants is the high boiling temperature that avoids the possibility of coolant voiding. Also, these coolants are compatible with air, steam, and water. Therefore, intermediate coolant loop is not required as in the sodium cooled reactors 3. Lead is considered to be more attractive coolant than lead-bismuth alloy because of its higher availability, lower price, and much lower amount of polonium activity by factor of 104 relatively to lead. On the other hand, lead has higher melting temperature of 601K than that of lead-bismuth (398K), which narrows the operating temperature range and also leads to the possibility of freezing and blockage in fresh cores. Neutronically, the lead and lead-bismuth have very similar characteristics to each other. The lead-alloy coolants have lower moderating power and higher scattering without increasing moderation for neutrons below 0.5MeV, which reduces the leakage of the neutrons through the core and provides an excellent reflecting capability for neutrons. Due to the above features of lead or lead-alloy coolants, there have been lots of studies on the small lead cooled core designs. In this paper, small-safe long-life lead cooled reactor cores having high discharge burnup are designed and neutronically analyzed.. The cores considered in this work rates 110MWt (36.7MWe). In this work, the long-life with high discharge burnup was achieved by using thorium or depleted uranium blanket loaded in the central region of the core. Also, we considered a reference core having no blanket for the comparison. This paper provides the detailed neutronic analyses for these small long-life cores and the detailed analyses of the reactivity coefficients and the composition changes in blankets. The results of the core design and analyses show that our small long-life cores can be operated without refueling over their long-lives longer than 45EFPYs (Effective Full Power Year). In this work

3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

International Nuclear Information System (INIS)

Chvetsov, I.; Volkov, A.

2000-01-01

For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

Severe Accident Recriticality Analyses (SARA)

Energy Technology Data Exchange (ETDEWEB)

Frid, W. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hoejerup, F. [Risoe National Lab. (Denmark); Lindholm, I.; Miettinen, J.; Puska, E.K. [VTT Energy, Helsinki (Finland); Nilsson, Lars [Studsvik Eco and Safety AB, Nykoeping (Sweden); Sjoevall, H. [Teoliisuuden Voima Oy (Finland)

1999-11-01

Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B{sub 4}C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding

Severe Accident Recriticality Analyses (SARA)

International Nuclear Information System (INIS)

Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Puska, E.K.; Nilsson, Lars; Sjoevall, H.

1999-11-01

Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B 4 C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding

AP1000 core design with 50% MOX loading

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)

2008-07-01

The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core and a mixed MOX / UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

AP1000 core design with 50% MOX loading

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)], E-mail: fetterrj@westinghouse.com

2009-04-15

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core design and a mixed MOX/UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.

Analysis of reactivity accidents of the RSG-GAS core with silicide fuel

International Nuclear Information System (INIS)

Tukiran

2002-01-01

The fuels of RSG-GAS reactor is changed from uranium oxide to uranium silicide. For time being, the fuel of RSG-GAS core are mixed up between oxide and silicide fuels with 250 gr of loading and 2.96 g U/cm 3 of density, respectively. While, silicide fuel with 300 gr of loading is still under research. The advantages of silicide fuels are can be used in high density, so that, it can be stayed longer in the core at higher burn-up, therefore, the length of cycle is longer. The silicide fuel in RSG-GAS core is used in step-wise by using mixed up core. Firstly, it is used silicide fuel with 250 gr of loading and then, silicide fuel with 300 gr of loading (3.55 g U/cm 3 of density). In every step-wise of fuel loading must be analysed its safety margin. In this occasion, it is analysed the reactivity accident of RSG-GAS core with 300 gr of silicide fuel loading. The calculation was done by using POKDYN code which available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. From all cases which were have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 gr silicide fuel loading

PHAROS A pluri-detector, high-resolution, analyser of radiometric properties of soil

CERN Document Server

Rigollet, C

2002-01-01

PHAROS is a new type of core logger, designed to measure activity concentrations of sup 4 sup 0 K, sup 2 sup 3 sup 8 U, sup 2 sup 3 sup 2 Th and sup 1 sup 3 sup 7 Cs in sediment and rock cores with a spatial resolution of a few centimetres along the core. PHAROS has been developed as a non-destructive alternative to the traditional slicing of cores into sub-samples and their analysis on an HPGe detector. The core is scanned at fixed increments by three BGO scintillation detectors and the spectra analysed by the full spectrum analysis method. The core logger is also equipped with a collimated lead castle and a sup 1 sup 3 sup 7 Cs source for transmission measurements. In this paper, we report on the properties of the core logger and its detectors, and on the analysis techniques used for the determination of the radionuclides activity concentrations. Results from initial measurements are presented and discussed.

The Organisational Relationsships between Support Functions and Core Business

DEFF Research Database (Denmark)

Jensen, Per Anker

2007-01-01

The paper is based on research for a MBA thesis. The purpose is to clarify the organisational relationships between support functions and core business and how these relationships vary for strategic and operational support functions. The value chains for core businesses and support functions...... are analysed and related to empirical data from a case study on Danish Broadcasting Corporation. A particular support value chain is identified and a typology of archetypes of support functions is developed. The relationship between core business and strategic support is identified as primarily a general...... business orientation, while the relationship between core business and non-strategic functions is identified as mainly a specific customer orientation. It is concluded that a market relationship – internally or externally – is appropriate for non-strategic functions, while it is important to create a kind...

Mineral and chemical composition of rock core and surface gas composition in Horonobe Underground Research Laboratory project. Phase 1

International Nuclear Information System (INIS)

Hiraga, Naoto; Ishii, Eiichi

2008-02-01

The following three kinds of analyses were conducted for the 1st phase of the Horonobe Underground Research Laboratory Project. Mineral composition analysis of core sample. Whole rock chemical composition analysis of core sample. Surface gas composition analysis. This document summarizes the results of these analyses. (author)

Geochemical analysis of core from a geothermal anomaly

International Nuclear Information System (INIS)

Haverslew, B.; Tammemagi, H.Y.

1985-04-01

A mild geothermal area in western Montana, USA, has been studied, as a natural analog, to learn about the effects that long-term heat generated by a repository containing spent nuclear fuel might have on the surrounding rock mass. The results of previous geological, geophysical and hydrogeological studies are briefly summarized. Extensive petrological studies have been undertaken on core samples obtained from a 2 km deep borehole drilled into the Empire Creek Stock. These include a detailed petrographic study, x-ray diffraction analyses, scanning electron microscope and electron microprobe analyses, porosity and permeability measurements, oxygen isotope analyses, uranium disequilibrium analyses and K-Ar age determinations. The implications to deep burial of nuclear wastes are discussed. 40 refs

Numerical investigation on turbulent natural convection in partially connected cylindrical enclosures for analysing SFR safety under core meltdown scenario

International Nuclear Information System (INIS)

David, Dijo K.; Mangarjuna Rao, P.; Nashine, B.K.; Selvaraj, P.

2015-01-01

Under the unlikely event of severe core meltdown accident in pool type SFR, the molten core materials may rupture the grid plate which supports the fuel subassemblies and it can get relocated in to the lower pool. These debris may eventually settle on the debris collector (i.e., core catcher) installed above the bottom wall of the lower pool. The bed thus formed generates heat due to radioactive decay which has to be passively removed for maintaining the structural integrity of main vessel. By means of natural convection, the heat generated in the debris bed will be transferred to the top pool where the heat sink (i.e., Decay heat exchanger (DHX)) is installed. Heat transfer to the DHX (which is a part of safety grade decay heat removal system) can take place through the opening created in the grid plate which connects the two liquid pools (i.e., the top pool and the lower pool). Heat transfer can also take place through the lateral wall of the lower cylindrical pool to the side pool and eventually to the top pool, and thus to the DHX. This study numerically investigates the effectiveness of heat transfer between lower pool and top pool during PARR by considering them as partially connected cylindrical enclosures. The governing equations have been numerically solved using finite volume method in cylindrical co-ordinates using SIMPLE algorithm. Turbulence has been modeled using k-ω model and the model is validated against benchmark problems of natural convection found in literature. The effect of parameters such as the heat generation rate in the bed and the size of the grid plate opening are evaluated. Also PAHR in SFR pool is modeled using an axi-symmetric model to fund out the influence of grid plate opening on heat removal from core catcher. The results obtained are useful for improving the cooling capability of in-vessel tray type core catcher for handling the whole core meltdown scenarios in SFR. (author)

Development of core-shell coaxially electrospun composite PCL/chitosan scaffolds.

Science.gov (United States)

Surucu, Seda; Turkoglu Sasmazel, Hilal

2016-11-01

This study was related to combining of synthetic Poly (ε-caprolactone) (PCL) and natural chitosan polymers to develop three dimensional (3D) PCL/chitosan core-shell scaffolds for tissue engineering applications. The scaffolds were fabricated with coaxial electrospinning technique and the characterizations of the samples were done by thickness and contact angle (CA) measurements, scanning electron microscopy (SEM), transmission electron microscopy (TEM), X-Ray Photoelectron Spectroscopy (XPS) analyses, mechanical and PBS absorption and shrinkage tests. The average inter-fiber diameter values were calculated for PCL (0.717±0.001μm), chitosan (0.660±0.007μm) and PCL/chitosan core-shell scaffolds (0.412±0.003μm), also the average inter-fiber pore size values exhibited decreases of 66.91% and 61.90% for the PCL and chitosan scaffolds respectively, compared to PCL/chitosan core-shell ones. XPS analysis of the PCL/chitosan core-shell structures exhibited the characteristic peaks of PCL and chitosan polymers. The cell culture studies (MTT assay, Confocal Laser Scanning Microscope (CLSM) and SEM analyses) carried out with L929 ATCC CCL-1 mouse fibroblast cell line proved that the biocompatibility performance of the scaffolds. The obtained results showed that the created micro/nano fibrous structure of the PCL/chitosan core-shell scaffolds in this study increased the cell viability and proliferation on/within scaffolds. Copyright © 2016 Elsevier B.V. All rights reserved.

Study of plutonium multi-recycle in high moderation LWR cores

International Nuclear Information System (INIS)

Iwata, Yutaka; Yamamoto, Toru; Ueji, Masao; Hibi, Koki; Aoyama, Motoo; Sakurada, Koichi

2000-01-01

Nuclear Power Engineering Corporation (NUPEC) has been studying advanced cores that are dedicated to enhance the plutonium consumption per recycling for effective use of plutonium. In this study, a fissile plutonium consumption rate is adopted as an index of the effective use of plutonium, which is defined as a ratio of consumption to loading of fissile plutonium in a core. High moderation core concepts have been studied in order to increase this index based on full MOX cores in the latest designs of LWRs in Japan that are the Advanced Boiling Water Reactor (ABWR) and the Advanced Pressurized Water Reactor (APWR). As a part of this study, core performance in the case of plutonium multi-recycling has been surveyed with these higher moderation cores aiming further effective use of plutonium. The design and analyses for equilibrium cores show that nuclear and thermal hydraulics parameters satisfy design criteria, and a fissile plutonium consumption rate increases up to 20% for ABWRs and 30% for APWRs even in plutonium multi-recycling condition. It was confirmed that the high moderation cores are feasible from a viewpoint of nuclear and thermal hydraulics, safety and plutonium consumption in the condition of plutonium multi-recycling. (author)

Introduction to special section on Organizational Challenges in the Knowledge Society

Directory of Open Access Journals (Sweden)

Anne Murray

2016-04-01

Full Text Available The field of Organizational Behavior (OB has been built on human sciences of psychology, sociology, and anthropology. We know that people and their emotions, motivations, prejudices, skills, temperaments, experiences, attitudes, fears, etc. are the key components of our organizations. The articles in this special section each take a different approach to the topic of furthering our understanding of OB in a knowledge economy, but all address the core need to understand culture and behavioral principles of people. This common core of understanding organizations gets us back to attending to the people who work there.

Comment on a proposed ''crucial experiment'' to test Einstein's special theory of relativity

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Jr, W A [Universidade Estadual de Campinas (Brazil); Buonamano, V [Universidade Estadual de Campinas (Brazil). Instituto de Matematica

1976-08-11

A proposed ''crucial experiment'' to test Einstein's special theory of relativity is analysed and it is shown that it falls into the set of unsatisfactory proposals that attempt to make an experimental distinction between Einstein's special theory of relativity and a ''Lorentzian type'' special theory of relativity.

Thermal hydraulic analyses of LVR-15 research reactor with IRT-M fuel

International Nuclear Information System (INIS)

Macek, J.

1997-01-01

The LVR-15 pool-type research reactor has been in operation at the Nuclear Research Institute at Rez since 1955. Following a number of reconstructions and redesigning, the current reactor power is 15 MW. Thermal hydraulic analyses to demonstrate that the core heat will be safely removed during operation as well as in accident situations were performed based on methodology which had been specifically developed for the LVR-15 research reactor. This methodology was applied to stationary thermal hydraulic computations, as well as to transients, particularly with reactivity failure and loss of circulation pumps emergencies. The applied methodology and the core configuration as used in the Safety Report are described. The initial and boundary conditions are then considered and the summary of the calculated failures with regard to the defined safety limits is presented. The results of the core configuration analyses are also discussed with respect to meeting the safety limits and to the applicability of the methodology to this purpose

Core Hunter 3: flexible core subset selection.

Science.gov (United States)

De Beukelaer, Herman; Davenport, Guy F; Fack, Veerle

2018-05-31

Core collections provide genebank curators and plant breeders a way to reduce size of their collections and populations, while minimizing impact on genetic diversity and allele frequency. Many methods have been proposed to generate core collections, often using distance metrics to quantify the similarity of two accessions, based on genetic marker data or phenotypic traits. Core Hunter is a multi-purpose core subset selection tool that uses local search algorithms to generate subsets relying on one or more metrics, including several distance metrics and allelic richness. In version 3 of Core Hunter (CH3) we have incorporated two new, improved methods for summarizing distances to quantify diversity or representativeness of the core collection. A comparison of CH3 and Core Hunter 2 (CH2) showed that these new metrics can be effectively optimized with less complex algorithms, as compared to those used in CH2. CH3 is more effective at maximizing the improved diversity metric than CH2, still ensures a high average and minimum distance, and is faster for large datasets. Using CH3, a simple stochastic hill-climber is able to find highly diverse core collections, and the more advanced parallel tempering algorithm further increases the quality of the core and further reduces variability across independent samples. We also evaluate the ability of CH3 to simultaneously maximize diversity, and either representativeness or allelic richness, and compare the results with those of the GDOpt and SimEli methods. CH3 can sample equally representative cores as GDOpt, which was specifically designed for this purpose, and is able to construct cores that are simultaneously more diverse, and either are more representative or have higher allelic richness, than those obtained by SimEli. In version 3, Core Hunter has been updated to include two new core subset selection metrics that construct cores for representativeness or diversity, with improved performance. It combines and outperforms the

Assessment of the fracture toughness of irradiated stainless steel for BWR core shrouds

International Nuclear Information System (INIS)

Carter, R.G.; Gamble, R.M.

2002-01-01

Data from previously performed experiments were collected and evaluated to determine the relationship between fracture toughness and neutron fluence for conditions representative of BWR core shrouds. This relationship together with EPFM (elastic-plastic fracture mechanics) analysis methods similar to those in Appendix K of Section XI of the ASME Code were used to compute margin against failure as a function of neutron fluence for postulated cracks in BWR core shrouds. The results indicate that EPFM analyses can be used for flaw evaluation of core shrouds at fluence levels less than 3.10 21 n/cm 2 (E > 1 MeV). At fluence levels equal to or greater than 3.10 21 n/cm 2 , LEFM (linear-elastic fracture mechanics) analyses should be used with K Ic = 55 MPa-(m) 0.5 . (authors)

The in-core experimental program at the MIT Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Kohse, G.E.; Hu, L-W., E-mail: kohse@mit.edu [Massachusetts Inst. of Technology, Nuclear Reactor Lab., Cambridge, Massachusetts (United States)

2014-07-01

This paper describes the program of in-core experiments at the Massachusetts Institute of Technology Research Reactor (MITR), a 6 MW research reactor. The MITR has a neutron flux and spectrum similar to those in water-cooled power reactors and therefore provides a useful test environment for materials and fuels research. In-core facilities include: a water loop operating at pressurized water or boiling water reactor conditions, an inert gas irradiation facility operating at temperature up to 850 {sup o}C and special purpose facilities including fuel irradiation experiments. Recent and ongoing tests include: water loop investigations of corrosion and thermal and mechanical property evolution of SiC/SiC composites for fuel cladding, irradiation of advanced materials and in-core sensors at elevated temperatures, irradiation in molten fluoride salt at 700 {sup o}C of metal alloy, graphite and composite materials for power reactor applications and instrumented irradiations of metal-bonded hydride fuel. (author)

Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

International Nuclear Information System (INIS)

Yamamoto, Toru; Ando, Yoshihira; Hayashi, Yamato

2008-01-01

As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k eff s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k eff s of the both cores by 1.0 to 1.3 %dk and the k eff s of MVP are 1.001. The difference in k eff between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

The development of the specialism of emergency medicine: media and cultural influences.

Science.gov (United States)

Timmons, Stephen; Nairn, Stuart

2015-01-01

In this article we analyse, via a critical review of the literature, the development of a relatively new medical specialism in the United Kingdom, that of emergency medicine. Despite the high media profile of emergency care, it is a low-status specialism within UK medicine. The creation of a specialist College in 2008 means that, symbolically, recognition as a full specialism has now been achieved. In this article, we will show, using a sociology of professions approach, how emergency medicine defined itself as a specialism, and sought to carve out a distinctive jurisdiction. While, in the context of the UK National Health Service, the state was clearly an important factor in the development of this profession, we wish to develop the analysis further than is usual in the sociology of professions. We will analyse the wider cultural context for the development of this specialism, which has benefited from its high profile in the media, through both fictional and documentary sources. © The Author(s) 2014.

Overview of Automotive Core Tools: Applications and Benefits

Science.gov (United States)

Doshi, Jigar A.; Desai, Darshak

2017-08-01

Continuous improvement of product and process quality is always challenging and creative task in today's era of globalization. Various quality tools are available and used for the same. Some of them are successful and few of them are not. Considering the complexity in the continuous quality improvement (CQI) process various new techniques are being introduced by the industries, as well as proposed by researchers and academia. Lean Manufacturing, Six Sigma, Lean Six Sigma is some of the techniques. In recent years, there are new tools being opted by the industry, especially automotive, called as Automotive Core Tools (ACT). The intention of this paper is to review the applications and benefits along with existing research on Automotive Core Tools with special emphasis on continuous quality improvement. The methodology uses an extensive review of literature through reputed publications—journals, conference proceedings, research thesis, etc. This paper provides an overview of ACT, its enablers, and exertions, how it evolved into sophisticated methodologies and benefits used in organisations. It should be of value to practitioners of Automotive Core Tools and to academics who are interested in how CQI can be achieved using ACT. It needs to be stressed here that this paper is not intended to scorn Automotive Core Tools, rather, its purpose is limited only to provide a balance on the prevailing positive views toward ACT.

Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

Energy Technology Data Exchange (ETDEWEB)

Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

1999-07-01

The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

Random close packing in protein cores.

Science.gov (United States)

Gaines, Jennifer C; Smith, W Wendell; Regan, Lynne; O'Hern, Corey S

2016-03-01

Shortly after the determination of the first protein x-ray crystal structures, researchers analyzed their cores and reported packing fractions ϕ ≈ 0.75, a value that is similar to close packing of equal-sized spheres. A limitation of these analyses was the use of extended atom models, rather than the more physically accurate explicit hydrogen model. The validity of the explicit hydrogen model was proved in our previous studies by its ability to predict the side chain dihedral angle distributions observed in proteins. In contrast, the extended atom model is not able to recapitulate the side chain dihedral angle distributions, and gives rise to large atomic clashes at side chain dihedral angle combinations that are highly probable in protein crystal structures. Here, we employ the explicit hydrogen model to calculate the packing fraction of the cores of over 200 high-resolution protein structures. We find that these protein cores have ϕ ≈ 0.56, which is similar to results obtained from simulations of random packings of individual amino acids. This result provides a deeper understanding of the physical basis of protein structure that will enable predictions of the effects of amino acid mutations to protein cores and interfaces of known structure.

Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

International Nuclear Information System (INIS)

Ravnik, M.

1988-11-01

The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

Replacement of core components in the Advanced Test Reactor

International Nuclear Information System (INIS)

Durney, J.L.; Croucher, D.W.

1990-01-01

The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. The methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and is expected to continue operation for at least and additional 25 years. Aging evaluations are in progress to address additional replacements that may be needed during this period

Replacement of core components in the Advanced Test Reactor

International Nuclear Information System (INIS)

Durney, J.L.; Croucher, D.W.

1989-01-01

The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. Methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and will continue operation for perhaps another 20 years. Aging evaluations are in program to address additional replacements that may be needed during this extended time period. 3 figs

Correlation and flux tilt measurements of coupled-core reactor assemblies

International Nuclear Information System (INIS)

Harries, J.R.

1976-01-01

The systematics of coupling reactivity and time delay between cores have been investigated with a series of coupled-core assemblies on the AAEC Split-table Critical Facility. The assemblies were similar to the Universities' Training Reactor (UTR), but had graphite coupling region thickness of 450 mm, 600 mm and 800 mm. The coupling reactivity measured by both the cross-correlation of reactor noise and the flux tilt methods was stronger than for the UTRs, but showed a similar trend with core spacing. The cross-correlograms were analysed using the two-node model to derive the time delays between the cores. The time delays were compared with thermal neutron wave propagation, and found to be consistent when the time delays were added to the individual node response-function delays. (author)

Core and shell sizing of small silver-coated nanospheres by optical extinction spectroscopy

International Nuclear Information System (INIS)

Schinca, D C; Scaffardi, L B

2008-01-01

Silver metal nanoparticles (Nps) are extensively used in different areas of research and technology due to their interesting optical, thermal and electric properties, especially for bare core and core-shell nanostructures with sizes smaller than 10 nm. Since these properties are core-shell size-dependent, size measurement is important in manipulating their potential functionalization and applications. Bare and coated small silver Nps fabricated by physical and chemical methods present specific characteristics in their extinction spectra that are potentially useful for sizing purposes. This work presents a novel procedure to size mean core radius smaller than 10 nm and mean shell thickness of silver core-shell Nps based on a comparative study of the characteristics in their optical extinction spectra in different media as a function of core radii, shell thickness and coating refractive index. From the regularities derived from these relationships, it can be concluded that plasmon full width at half-maximum (FWHM) is sensitive to core size but not to coating thickness, while plasmon resonance wavelength (PRW) is related to shell thickness and mostly independent of core radius. These facts, which allow sizing simultaneously both mean core radius and shell thickness, can also be used to size bare silver Nps as a special case of core-shell Nps with zero shell thickness. The proposed method was applied to size experimental samples and the results show good agreement with conventional TEM microscopy.

Solid-phase data from cores at the proposed Dewey Burdock uranium in-situ recovery mine, near Edgemont, South Dakota

Science.gov (United States)

Johnson, Raymond H.; Diehl, Sharon F.; Benzel, William M.

2013-01-01

This report releases solid-phase data from cores at the proposed Dewey Burdock uranium in-situ recovery site near Edgemont, South Dakota. These cores were collected by Powertech Uranium Corporation, and material not used for their analyses were given to the U.S. Geological Survey for additional sampling and analyses. These additional analyses included total carbon and sulfur, whole rock acid digestion for major and trace elements, 234U/238U activity ratios, X-ray diffraction, thin sections, scanning electron microscopy analyses, and cathodoluminescence. This report provides the methods and data results from these analyses along with a short summary of observations.

Phylogenetic analysis of the core histone doublet and DNA topo II genes of Marseilleviridae: evidence of proto-eukaryotic provenance.

Science.gov (United States)

Erives, Albert J

2017-11-28

While the genomes of eukaryotes and Archaea both encode the histone-fold domain, only eukaryotes encode the core histone paralogs H2A, H2B, H3, and H4. With DNA, these core histones assemble into the nucleosomal octamer underlying eukaryotic chromatin. Importantly, core histones for H2A and H3 are maintained as neofunctionalized paralogs adapted for general bulk chromatin (canonical H2 and H3) or specialized chromatin (H2A.Z enriched at gene promoters and cenH3s enriched at centromeres). In this context, the identification of core histone-like "doublets" in the cytoplasmic replication factories of the Marseilleviridae (MV) is a novel finding with possible relevance to understanding the origin of eukaryotic chromatin. Here, we analyze and compare the core histone doublet genes from all known MV genomes as well as other MV genes relevant to the origin of the eukaryotic replisome. Using different phylogenetic approaches, we show that MV histone domains encode obligate H2B-H2A and H4-H3 dimers of possible proto-eukaryotic origin. MV core histone moieties form sister clades to each of the four eukaryotic clades of canonical and variant core histones. This suggests that MV core histone moieties diverged prior to eukaryotic neofunctionalizations associated with paired linear chromosomes and variant histone octamer assembly. We also show that MV genomes encode a proto-eukaryotic DNA topoisomerase II enzyme that forms a sister clade to eukaryotes. This is a relevant finding given that DNA topo II influences histone deposition and chromatin compaction and is the second most abundant nuclear protein after histones. The combined domain architecture and phylogenomic analyses presented here suggest that a primitive origin for MV histone genes is a more parsimonious explanation than horizontal gene transfers + gene fusions + sufficient divergence to eliminate relatedness to eukaryotic neofunctionalizations within the H2A and H3 clades without loss of relatedness to each of

Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

International Nuclear Information System (INIS)

Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

2001-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The 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 safety to a full-silicide core

Precursors to potential severe core damage accidents: 1992, A status report

International Nuclear Information System (INIS)

Cox, D.F.; Cletcher, J.W.; Copinger, D.A.; Cross-Dial, A.E.; Morris, R.H.; Vanden Heuvel, L.N.; Dolan, B.W.; Jansen, J.M.; Minarick, J.W.; Lau, W.; Salyer, W.D.

1993-12-01

Twenty-seven operational events with conditional probabilities of subsequent severe core damage of 1.0 x 10E-06 or higher occurring at commercial light-water reactors during 1992 are considered to be precursors to potential core damage. These are described along with associated significance estimates, categorization, and subsequent analyses. The report discusses (1) the general rationale for this study, (2) the selection and documentation of events as precursors, (3) the estimation and use of conditional probabilities of subsequent severe core damage to rank precursor events, and (4) the plant models used in the analysis process

Core-to-core uniformity improvement in multi-core fiber Bragg gratings

Science.gov (United States)

Lindley, Emma; Min, Seong-Sik; Leon-Saval, Sergio; Cvetojevic, Nick; Jovanovic, Nemanja; Bland-Hawthorn, Joss; Lawrence, Jon; Gris-Sanchez, Itandehui; Birks, Tim; Haynes, Roger; Haynes, Dionne

2014-07-01

Multi-core fiber Bragg gratings (MCFBGs) will be a valuable tool not only in communications but also various astronomical, sensing and industry applications. In this paper we address some of the technical challenges of fabricating effective multi-core gratings by simulating improvements to the writing method. These methods allow a system designed for inscribing single-core fibers to cope with MCFBG fabrication with only minor, passive changes to the writing process. Using a capillary tube that was polished on one side, the field entering the fiber was flattened which improved the coverage and uniformity of all cores.

General and special engineering materials science. Vol. 3

International Nuclear Information System (INIS)

Ondracek, G.; Hofmann, P.

1983-04-01

The report about general and special engineering materials science is the result of lectures given by the authors in two terms in 1982 at Instituto Balseiro, San Carlos de Bariloche, the graduated college of the Universidad de Cuyo and Comision Nacional de Energia Atomica, Republica Argentina. These lectures were organised in the frame of the project ''nuclear engineering'' (ARG/78/020) of the United Nations Development Program (UNDP) by the International Atomic Energy Agency (IAEA). Some chapters of the report are written in English, others in Spanish. The report is subdivided into three volumes. The present volume III concerns special engineering materials science and considers nuclear materials with respect to off-normal (''accident'') reactor operation conditions including nuclear materials in loss-of-coolant accident and nuclear materials in core melt accidents. (orig./IHOE) [de

Development of the core safety regulation technology for the SMART-P

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Zin; Kim, Do Sam; Lee, Kyeong Taek; Park, Young Ryoung; Lee, Gil Soo; Kim, Jong Woon; Yun, Sung Hwan; Lee, Jae Jun; Lee, Myung Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2003-06-15

As the SMART-P is different from existing general reactors, new regulation technology is required to understand and assess the SMART-P for its regulatory reviews. One of the these technologies is related to the core design analysis. Because the SMART-P used metallic fuels, this study also collects general metallic nuclear fuel data and SMART-P's metallic fuel data from the materials studied by KAERI. The core design methodologies of KWU, ABB-CE, Westinghouse, Studsvik, Scandpower, US NRC and domestic research centers were investigated. Specially, The Hellios lattice core was studied for hexagonal nuclear fuel assembly calculation. Also, the VVER-1000 benchmark problem was analyzed by the PARCS code which has been developed by U.S. NRC. In this study, a AFEN-based computing code KORDAX os developed for the regulatory review of the SMART-P. KORDAX which is a nodal code using AFEN method dose not use transverse integration and this it can give higher accuracy results. Also, Because KORDAX is useful for hexagonal core and uses a method different with the core design code of the SMART-P developed by KAERI, it is judged that KORDAX can be an independent and reliable regulation verification code. In the next year study, HELIOS will be further studied as a core lattice code, and a hexagonal kinetics code which is based on AFEN method will be developed more systematically.

Preface to Special Topic: Acoustic Metamaterials and Metasurfaces

Science.gov (United States)

Assouar, Badreddine

2018-03-01

The advent of acoustic metamaterials in the beginning of 2000s and very recently of acoustic metasurfaces has created tremendous excitement and efforts in the field of materials science and physics by introducing and building real transformative research and dealing with unprecedented physics and applications. The acoustic/elastic metamaterials and metasurfaces, which can simply be described as designed artificial materials with unusual physical properties, form the core of the present Special Topic published by the Journal of Applied Physics.

The wide range in-core neutron measurement system used in the Windscale AGR concluding experiments

International Nuclear Information System (INIS)

Goodings, A.; Budd, J.; Wilson, I.

1982-06-01

The Windscale AGR Concluding Experiments included a comparison of theoretical and experimental power transients and required measurements of neutron flux as a function of position and time within the reactor core. These measurements were specified to cover as wide as possible working range and had to be made against the in-core gamma background of up to 4 x 10 7 R(hr) - 1 . The detectors were required to operate in special, channels cooled by reactor inlet carbon dioxide and the overall system needed a response time such that it could follow transients with doubling times down to 2s with an accuracy of 2 or 3%. These problems were solved by the use of gas ion fission chambers operating in the current fluctuation or Campbelling mode with unusually low filling pressures and fitted with special trilaminax mineral insulated cables. Ten detectors were built and nine were installed in the reactor, three in each of three special stringers at different radial positions. The paper describes the specification against which this system was built, the design process for the detectors, and commissioning experiments together with some of the problems which were encountered. (U.K.)

Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

Energy Technology Data Exchange (ETDEWEB)

Fabbris, Olivier [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Dardour, Saied, E-mail: saied.dardour@cea.fr [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Blaise, Patrick [CEA DEN/DER/SPEX, 13108 Saint-Paul-Lez-Durance (France); Ferrasse, Jean-Henry [Aix-Marseille Université, CNRS, ECM, M2P2 UMR 7340, 13451 Marseille (France); Saez, Manuel [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France)

2016-08-15

Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

International Nuclear Information System (INIS)

Fabbris, Olivier; Dardour, Saied; Blaise, Patrick; Ferrasse, Jean-Henry; Saez, Manuel

2016-01-01

Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

Development of inherent core technologies for advanced reactor

International Nuclear Information System (INIS)

Kim, Keung Koo; Noh, J.M.; Hwang, D.H.

1999-03-01

Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

Development of inherent core technologies for advanced reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

1999-03-01

Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

Descriptions and preliminary report on sediment cores from the southwest coastal area, Everglades National Park, Florida

Science.gov (United States)

Wingard, G. Lynn; Cronin, Thomas M.; Holmes, Charles W.; Willard, Debra A.; Budet, Carlos A.; Ortiz, Ruth E.

2005-01-01

Sediment cores were collected from five locations in the southwest coastal area of Everglades National Park, Florida, in May 2004 for the purpose of determining the ecosystem history of the area and the impacts of changes in flow through the Shark River Slough. An understanding of natural cycles of change prior to significant human disturbance allows land managers to set realistic performance measures and targets for salinity and other water quality and quantity quality measures. Preliminary examination of the cores indicates significant changes have taken place over the last 1000-2000 years. The cores collected from the inner bays - the most landward bays - are distinctly different from other estuarine sediment cores examined in Florida Bay and Biscayne Bay. Peats in the inner-bay cores from Big Lostmans Bay, Broad River Bay, and Tarpon Bay were deposited at least 1000 years before present (BP) based on radiocarbon analyses. The peats are overlain by poorly sorted organic muds and sands containing species indicative of deposition in a freshwater to very low salinity environment. The Alligator Bay core, the most northern inner-bay core, is almost entirely sand; no detailed faunal analyses or radiometric dating has been completed on this core. The Roberts River core, taken from the mouth of the River where it empties into Whitewater Bay, is lithologically and faunally similar to previously examined cores from Biscayne and Florida Bays; however, the basal unit was deposited ~2000 years before the present based on radiocarbon analyses. A definite trend of increasing salinity over time is seen in the Roberts River core, from sediments representing a terrestrially dominated freshwater environment at the bottom of the core to those representing an estuarine environment with a strong freshwater influence at the top. The changes seen at Roberts River could represent a combination of factors including rising sea-level and changes in freshwater supply, but the timing and

A Fast CT Reconstruction Scheme for a General Multi-Core PC

Directory of Open Access Journals (Sweden)

Kai Zeng

2007-01-01

Full Text Available Expensive computational cost is a severe limitation in CT reconstruction for clinical applications that need real-time feedback. A primary example is bolus-chasing computed tomography (CT angiography (BCA that we have been developing for the past several years. To accelerate the reconstruction process using the filtered backprojection (FBP method, specialized hardware or graphics cards can be used. However, specialized hardware is expensive and not flexible. The graphics processing unit (GPU in a current graphic card can only reconstruct images in a reduced precision and is not easy to program. In this paper, an acceleration scheme is proposed based on a multi-core PC. In the proposed scheme, several techniques are integrated, including utilization of geometric symmetry, optimization of data structures, single-instruction multiple-data (SIMD processing, multithreaded computation, and an Intel C++ compilier. Our scheme maintains the original precision and involves no data exchange between the GPU and CPU. The merits of our scheme are demonstrated in numerical experiments against the traditional implementation. Our scheme achieves a speedup of about 40, which can be further improved by several folds using the latest quad-core processors.

A fast CT reconstruction scheme for a general multi-core PC.

Science.gov (United States)

Zeng, Kai; Bai, Erwei; Wang, Ge

2007-01-01

Expensive computational cost is a severe limitation in CT reconstruction for clinical applications that need real-time feedback. A primary example is bolus-chasing computed tomography (CT) angiography (BCA) that we have been developing for the past several years. To accelerate the reconstruction process using the filtered backprojection (FBP) method, specialized hardware or graphics cards can be used. However, specialized hardware is expensive and not flexible. The graphics processing unit (GPU) in a current graphic card can only reconstruct images in a reduced precision and is not easy to program. In this paper, an acceleration scheme is proposed based on a multi-core PC. In the proposed scheme, several techniques are integrated, including utilization of geometric symmetry, optimization of data structures, single-instruction multiple-data (SIMD) processing, multithreaded computation, and an Intel C++ compilier. Our scheme maintains the original precision and involves no data exchange between the GPU and CPU. The merits of our scheme are demonstrated in numerical experiments against the traditional implementation. Our scheme achieves a speedup of about 40, which can be further improved by several folds using the latest quad-core processors.

TMI-2 core debris grab samples: Examination and analysis: Part 1

International Nuclear Information System (INIS)

Akers, D.W.; Carlson, E.R.; Cook, B.A.; Ploger, S.A.; Carlson, J.O.

1986-09-01

Six samples of particulate debris were removed from the TMI-2 core rubble bed during September and October 1983, and five more samples were obtained in March 1984. The samples (up to 174 g each) were obtained at two locations in the core: H8 (center) and E9 (mid-radius). Ten of the eleven samples were examined at the Idaho National Engineering Laboratory to obtain data on the physical and chemical nature of the debris and the postaccident condition of the core. Portions of the samples also were subjected to differential thermal analysis at Rockwell Hanford Operations and metallurgical and chemical examinations at Argonne National Laboratories. This report presents results of the examination of the core debris grab samples, including physical, metallurgical, chemical, and radiochemical analyses. The results indicate that temperatures in the core reached at least 3100 K during the TMI-2 accident, fuel melting and significant mixing of core structural material occurred, and large fractions of some radionuclides (e.g., 90 Sr and 144 Ce) were retained in the core

Sodium, Iodine and Bromine in Polar Ice Cores

DEFF Research Database (Denmark)

Maffezzoli, Niccolo

Abstract: This research focuses on sodium, bromine and iodine in polar ice cores, with the aim of reviewing and advancing their current understanding with additional measurements and records, and investigating the connections of these tracers with sea ice and their feasibility as sea ice indicators...... with a description of the main analytic al techniques used to measure ionic and elemental species in ice cores. Chapter 4 introduces sodium, bromine and iodine with a theoretical perspective and a particular focus on their connections with sea ice. Some of the physical and chemical properties that are believed...... back trajectory analyses of the past 17 years. The results identify the aerosol source area influencing the Renland ice cap, a result necessary for the interpretation of impurity records obtained from the ice core. Chapter 6 reviews the published ice/snow measurements of bromine and iodine at polar...

Binding of ethidium to the nucleosome core particle. 2. Internal and external binding modes

International Nuclear Information System (INIS)

McMurray, C.T.; Small, E.W.; van Holde, K.E.

1991-01-01

The authors have previously reported that the binding of ethidium bromide to the nucleosome core particle results in a stepwise dissociation of the structure which involves the initial release of one copy each of H2A and H2B. In this report, they have examined the absorbance and fluorescence properties of intercalated and outside bound forms of ethidium bromide. From these properties, they have measured the extent of external, electrostatic binding of the dye versus internal, intercalation binding to the core particle, free from contribution by linker DNA. They have established that dissociation is induced by the intercalation mode of binding to DNA within the core particle DNA, and not by binding to the histones or by nonintercalative binding to DNA. The covalent binding of [ 3 H]-8-azidoethidium to the core particle clearly shows that < 1.0 adduct is formed per histone octamer over a wide range of input ratios. Simultaneously, analyses of steady-state fluorescence enhancement and fluorescence lifetime data from bound ethidium complexes demonstrate extensive intercalation binding. Combined analyses from steady-state fluorescence intensity with equilibrium dialysis or fluorescence lifetime data revealed that dissociation began when ∼14 ethidium molecules are bound by intercalation to each core particle and < 1.0 nonintercalated ion pair was formed per core particle

Coupling between core and cladding modes in a helical core fiber with large core offset

International Nuclear Information System (INIS)

Napiorkowski, Maciej; Urbanczyk, Waclaw

2016-01-01

We analyzed the effect of resonant coupling between core and cladding modes in a helical core fiber with large core offset using the fully vectorial method based on the transformation optics formalism. Our study revealed that the resonant couplings to lower order cladding modes predicted by perturbative methods and observed experimentally in fibers with small core offsets are in fact prohibited for larger core offsets. This effect is related to the lack of phase matching caused by elongation of the optical path of the fundamental modes in the helical core. Moreover, strong couplings to the cladding modes of the azimuthal modal number much higher than predicted by perturbative methods may be observed for large core offsets, as the core offset introduces higher order angular harmonics in the field distribution of the fundamental modes. Finally, in contrast to previous studies, we demonstrate the existence of spectrally broad polarization sensitive couplings to the cladding modes suggesting that helical core fibers with large core offsets may be used as broadband circular polarizers. (paper)

Assessing the Special Education Professional Development Needs of Northern Malawian Schoolteachers

Science.gov (United States)

Hughes, Elizabeth; Chitiyo, Morgan; Itimu-Phiri, Ambumulire; Montgomery, Kristen

2016-01-01

This research examines special needs education professional development needs among both general and special education schoolteachers in northern Malawi. A semi-structured questionnaire with open and close-ended questions was used for the research. Quantitative and thematic analyses were conducted to determine the extent to which teachers believe…

A comment on a proposed ''crucial experiment'' to test Einstein's special theory of relativity

International Nuclear Information System (INIS)

Rodrigues Jr, W.A.; Buonamano, V.

1976-01-01

A proposed ''crucial experiment'' to test Einstein's special theory of relativity is analysed and it is shown that it falls into the set of unsatisfactory proposals that attempt to make an experimental distinction between Einstein's special theory of relativity and a ''Lorentzian type'' special theory of relativity

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

Coupled neutronic/thermal-hydraulic analysis of the HPLWR three pass core

International Nuclear Information System (INIS)

Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

2008-01-01

The High Performance Light Water Reactor is an innovative Gen-IV reactor cooled and moderated with water at supercritical pressure. The three pass core concept has been proposed to reduce peaking factors, i.e. hot-channel effects, and it further increases the core heterogeneity, which is mainly due to pronounced water density reduction. For this kind of nuclear reactor, the significant feedbacks - which exist between the properties of the components and the power generation rate - can not be neglected and require a coupled Neutronic/Thermal-Hydraulic analysis even for steady state conditions. The main goal of this paper is to present the developed tool for coupled analyses of the HPLWR. Two state-of-the-art codes have been chosen for Thermal-Hydraulic and Neutronic core analyses, namely TRACE and ERANOS, and they have been coupled with in an iterative procedure in which they are run in series until a steady state condition has been reached. In the simplifying assumptions of uniform enrichment distribution, zero burn-up and ignoring the effect of the control rods, the obtained steady state condition will be discussed and a core power map, flow rate redistribution as well as water and fuel temperature variations will be presented. (author)

Oyster Creek fuel thermal margin during core thermal-hydraulic oscillations

International Nuclear Information System (INIS)

Dougher, J.D.

1990-01-01

The Oyster Creek nuclear facility, a boiling water reactor (BWR)-2 plant type, has never experienced core thermal-hydraulic instability. Power oscillations, however, have been observed in other BWR cores both domestically and internationally. Two modes of oscillations have been observed, core wide and regional half-core. During core wide oscillations, the neutron flux in the core oscillates in the radial fundamental mode. During regional half-core oscillations, higher order harmonics in the radial plane result in out-of-phase oscillations with the neutron flux in one half of the core oscillating 180 deg out-of-phase with the neutron flux in the other half of the core. General Design Criteria 12 requires either prevention or detection and suppression of power oscillations which could result in violations of fuel design limits. Analyses performed by General Electric have demonstrated that for large-magnitude oscillations the potential exists for violation of the safety limit minimum critical power ratio (MCPR). However, for plants with a flow-biased neutron flux scram automatic mitigation of oscillations may be provided at an oscillation magnitude below that at which the safety limit is challenged. Plant-specific analysis for Oyster Creek demonstrates that the existing average power range monitor (APRM) system will sense and suppress power oscillations prior to violation of any safety limits

Continuous methane measurements from a late Holocene Greenland ice core

DEFF Research Database (Denmark)

Rhodes, R.H.; Mitchell, L.E.; Brook, E.J.

2013-01-01

Ancient air trapped inside bubbles in ice cores can now be analysed for methane concentration utilising a laser spectrometer coupled to a continuous melter system. We present a new ultra-high resolution record of atmospheric methane variability over the last 1800yr obtained from continuous analysis...... of a shallow ice core from the North Greenland Eemian project (NEEM-2011-S1) during a 4-week laboratory-based measurement campaign. Our record faithfully replicates the form and amplitudes of multi-decadal oscillations previously observed in other ice cores and demonstrates the detailed depth resolution (5.3cm......), rapid acquisition time (30mday) and good long-term reproducibility (2.6%, 2s) of the continuous measurement technique.In addition, we report the detection of high frequency ice core methane signals of non-atmospheric origin. Firstly, measurements of air from the firn-ice transition region...

IEA-R1 reactor core simulation with RELAP5 code

International Nuclear Information System (INIS)

Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

2005-01-01

This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4

International Nuclear Information System (INIS)

Tombakoglu, M.; Cecen, Y.

2001-01-01

In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)

Preliminary analyses for HTTR's start-up physics tests by Monte Carlo code MVP

International Nuclear Information System (INIS)

Nojiri, Naoki; Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

1998-08-01

Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

Organic carbon-sulfur relationships in sediment cores from the western and eastern continental margins of India

Digital Repository Service at National Institute of Oceanography (India)

Rao, P.S.; Mascarenhas, A.; Paropkari, A.L.; Rao, Ch.M.

Two sediment cores from the western continental margin (WMI, WM2) and one core from the eastern continental margin (EM) of India have been analysed to determine the relativ importance of factors such as oxidizing/reducing environment, mass...

Evolution of feeding specialization in Tanganyikan scale-eating cichlids: a molecular phylogenetic approach

Directory of Open Access Journals (Sweden)

Nishida Mutsumi

2007-10-01

Full Text Available Abstract Background Cichlid fishes in Lake Tanganyika exhibit remarkable diversity in their feeding habits. Among them, seven species in the genus Perissodus are known for their unique feeding habit of scale eating with specialized feeding morphology and behaviour. Although the origin of the scale-eating habit has long been questioned, its evolutionary process is still unknown. In the present study, we conducted interspecific phylogenetic analyses for all nine known species in the tribe Perissodini (seven Perissodus and two Haplotaxodon species using amplified fragment length polymorphism (AFLP analyses of the nuclear DNA. On the basis of the resultant phylogenetic frameworks, the evolution of their feeding habits was traced using data from analyses of stomach contents, habitat depths, and observations of oral jaw tooth morphology. Results AFLP analyses resolved the phylogenetic relationships of the Perissodini, strongly supporting monophyly for each species. The character reconstruction of feeding ecology based on the AFLP tree suggested that scale eating evolved from general carnivorous feeding to highly specialized scale eating. Furthermore, scale eating is suggested to have evolved in deepwater habitats in the lake. Oral jaw tooth shape was also estimated to have diverged in step with specialization for scale eating. Conclusion The present evolutionary analyses of feeding ecology and morphology based on the obtained phylogenetic tree demonstrate for the first time the evolutionary process leading from generalised to highly specialized scale eating, with diversification in feeding morphology and behaviour among species.

Operational characteristics of hybrid-type SFCL with closed and open cores

International Nuclear Information System (INIS)

Cho, Y.S.; Lee, N.Y.; Choi, H.S.; Chung, D.C.; Lim, S.H.

2007-01-01

We investigated the operational characteristics of the hybrid-type superconducting fault current limiter (SFCL) with the closed and the open cores, which induced the variation of the magnetic flux between the primary and the secondary windings. The experimental set-up of the hybrid-type SFCL with the closed and the open cores were prepared and the experimental analyses for the current limiting characteristics were performed. The peak value of the fault current in the hybrid-type SFCL with the open core was higher than that of the closed core at the first cycle after fault occurrence. However, in the case of the hybrid-type SFCL with the open core, the limiting current level after fault occurrence was decreased less than that of the hybrid-type SFCL with the closed core, because the magnetic leakage reluctance of the open core was higher than that of the closed core. The quench time (T q ) and the arrival time (T a ) for the peak voltage (V SC ) in the hybrid-type SFCL with the closed core were faster than that of the hybrid-type SFCL with the open core due to the increase of the mutual flux. We verified that the consumption power in the hybrid-type SFCL with the open core was larger owing to the increase of leakage flux by the reduction of mutual inductance between primary and secondary windings

High moderation MOX cores for effective use of plutonium in LWRs

International Nuclear Information System (INIS)

Hamamoto, Kazuko; Kanagawa, Takashi; Hiraiwa, Koji; Sakurada, Koichi; Moriwaki, Masanao; Aoyama, Motoo; Yamamoto, Toru; Ueji, Masao

2001-01-01

Conceptual design studies have been performed for high moderation full MOX cores aiming at increasing fissile Pu consumption rate (ratio of the consumed to the loaded fissile Pu) and reducing residual Pu in discharged MOX fuel. The BWR cores studied have hydrogen to heavy metal ratio(H/HM) of 5.9 with increasing water rods and 7.0 with reducing a fuel rod diameter based on a reference 9x9 fuel (H/HM=4.9) of ABWR. The PWR cores studied have H/HM of 5.0 and 6.0 with reducing a fuel rod diameter based on a reference 17x17 fuel (H/HM=4.0) of APWR. Equilibrium core design and plant safety analyses showed that those high moderation cores have compatibility with ABWR and APWR. The fissile Pu consumption rate is 22% larger than the full MOX cores with reference fuel of ABWR and 50% for APWR. The core performance and compatibility has been also evaluated in the condition of multi-recycle of Pu in these high moderation cores. Study has been conducted to evaluate the effect of introducing these high moderation cores in the fuel cycle of Japan. It shows that the high moderation cores produce 26% more cumulative electricity and reduce 22% stock of the fissile Pu by 2050 than the reference cores. (author)

AMS analyses at ANSTO

Energy Technology Data Exchange (ETDEWEB)

Lawson, E.M. [Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia). Physics Division

1998-03-01

The major use of ANTARES is Accelerator Mass Spectrometry (AMS) with {sup 14}C being the most commonly analysed radioisotope - presently about 35 % of the available beam time on ANTARES is used for {sup 14}C measurements. The accelerator measurements are supported by, and dependent on, a strong sample preparation section. The ANTARES AMS facility supports a wide range of investigations into fields such as global climate change, ice cores, oceanography, dendrochronology, anthropology, and classical and Australian archaeology. Described here are some examples of the ways in which AMS has been applied to support research into the archaeology, prehistory and culture of this continent`s indigenous Aboriginal peoples. (author)

AMS analyses at ANSTO

International Nuclear Information System (INIS)

Lawson, E.M.

1998-01-01

The major use of ANTARES is Accelerator Mass Spectrometry (AMS) with 14 C being the most commonly analysed radioisotope - presently about 35 % of the available beam time on ANTARES is used for 14 C measurements. The accelerator measurements are supported by, and dependent on, a strong sample preparation section. The ANTARES AMS facility supports a wide range of investigations into fields such as global climate change, ice cores, oceanography, dendrochronology, anthropology, and classical and Australian archaeology. Described here are some examples of the ways in which AMS has been applied to support research into the archaeology, prehistory and culture of this continent's indigenous Aboriginal peoples. (author)

Reliability analyses of safety systems for WWER-440 nuclear power plants

International Nuclear Information System (INIS)

Dusek, J.; Hojny, V.

1985-01-01

The UJV in Rez near Prague studied the reliability of the system of emergency core cooling and of the system for suppressing pressure in the sealed area of the nuclear power plant in the occurrence of a loss-of-coolant accident. The reliability of the systems was evaluated by failure tree analysis. Simulation and analytical calculation programs were developed and used for the reliability analysis. The results are briefly presented of the reliability analyses of the passive system for the immediate short-term flooding of the reactor core, of the active low-pressure system of emergency core cooling, the spray system, the bubble-vacuum system and the system of emergency supply of the steam generators. (E.S.)

Preliminary safety analysis for key design features of KALIMER with breakeven core

Energy Technology Data Exchange (ETDEWEB)

Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

2001-06-01

KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

Energy Technology Data Exchange (ETDEWEB)

David W. Nigg; Sean R. Morrell

2012-09-01

Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

Chemical analyses of rocks, minerals, and detritus, Yucca Mountain--Preliminary report, special report No. 11

International Nuclear Information System (INIS)

Hill, C.A.; Livingston, D.E.

1993-09-01

This chemical analysis study is part of the research program of the Yucca Mountain Project intended to provide the State of Nevada with a detailed assessment of the geology and geochemistry of Yucca Mountain and adjacent regions. This report is preliminary in the sense that more chemical analyses may be needed in the future and also in the sense that these chemical analyses should be considered as a small part of a much larger geological data base. The interpretations discussed herein may be modified as that larger data base is examined and established. All of the chemical analyses performed to date are shown in Table 1. There are three parts to this table: (1) trace element analyses on rocks (limestone and tuff) and minerals (calcite/opal), (2) rare earth analyses on rocks (tuff) and minerals (calcite/opal), and (3) major element analyses + CO 2 on rocks (tuff) and detritus sand. In this report, for each of the three parts of the table, the data and its possible significance will be discussed first, then some overall conclusions will be made, and finally some recommendations for future work will be offered

Earth's inner core: Innermost inner core or hemispherical variations?

NARCIS (Netherlands)

Lythgoe, K. H.; Deuss, A.|info:eu-repo/dai/nl/412396610; Rudge, J. F.; Neufeld, J. A.

2014-01-01

The structure of Earth's deep inner core has important implications for core evolution, since it is thought to be related to the early stages of core formation. Previous studies have suggested that there exists an innermost inner core with distinct anisotropy relative to the rest of the inner core.

Dependence of Core and Extended Flux on Core Dominance ...

Indian Academy of Sciences (India)

Abstract. Based on two extragalactic radio source samples, the core dominance parameter is calculated, and the correlations between the core/extended flux density and core dominance parameter are investi- gated. When the core dominance parameter is lower than unity, it is linearly correlated with the core flux density, ...

The KALIMER-600 Reactor Core Design Concept with Varying Fuel Cladding Thickness

International Nuclear Information System (INIS)

Hong, Ser Gi; Jang, Jin Wook; Kim, Yeong Il

2006-01-01

Recently, Korea Atomic Energy Research Institute (KAERI) has developed a 600MWe sodium cooled fast reactor, the KALIMER-600 reactor core concept using single enrichment fuel. This reactor core concept is characterized by the following design targets : 1) Breakeven breeding (or fissile-self-sufficient) without any blanket, 2) Small burnup reactivity swing ( 23 n/cm 2 ). In the previous design, the single enrichment fuel concept was achieved by using the special fuel assembly designs where non-fuel rods (i.e., ZrH 1.8 , B 4 C, and dummy rods) were used. In particular, the moderator rods (ZrH 1.8 ) were used to reduce the sodium void worth and the fuel Doppler coefficient. But it has been known that this hydride moderator possesses relatively poor irradiation behavior at high temperature. In this paper, a new core design concept for use of single enrichment fuel is described. In this concept, the power flattening is achieved by using the core region wise cladding thicknesses but all non-fuel rods are removed to simplify the fuel assembly design

Dynamic Musical Communication of Core Affect

Directory of Open Access Journals (Sweden)

Nicole eFlaig

2014-03-01

Full Text Available Is there something special about the way music communicates feelings? Theorists since Meyer (1956 have attempted to explain how music could stimulate varied and subtle affective experiences by violating learned expectancies, or by mimicking other forms of social interaction. Our proposal is that music speaks to the brain in its own language; it need not imitate any other form of communication. We review recent theoretical and empirical literature, which suggests that all conscious processes consist of dynamic neural events, produced by spatially dispersed processes in the physical brain. Intentional thought and affective experience arise as dynamical aspects of neural events taking place in multiple brain areas simultaneously. At any given moment, this content comprises a unified scene that is integrated into a dynamic core through synchrony of neuronal oscillations. We propose that 1 neurodynamic synchrony with musical stimuli gives rise to musical qualia including tonal and temporal expectancies, and that 2 music-synchronous responses couple into core neurodynamics, enabling music to directly modulate core affect. Expressive music performance, for example, may recruit rhythm-synchronous neural responses to support affective communication. We suggest that the dynamic relationship between musical expression and the experience of affect presents a unique opportunity for the study of emotional experience. This may help elucidate the neural mechanisms underlying arousal and valence, and offer a new approach to exploring the complex dynamics of the how and why of emotional experience.

On-line core monitoring with CORE MASTER / PRESTO

International Nuclear Information System (INIS)

Lindahl, S.O.; Borresen, S.; Ovrum, S.

1986-01-01

Advanced calculational tools are instrumental in improving reactor plant capacity factors and fuel utilization. The computer code package CORE MASTER is an integrated system designed to achieve this objective. The system covers all main activities in the area of in-core fuel management for boiling water reactors; design, operation support, and on-line core monitoring. CORE MASTER operates on a common data base, which defines the reactor and documents the operating history of the core and of all fuel bundles ever used

Peculiarities of morphological indexes of the pushers of the nucleus at the stage of specialized basic preparation

Directory of Open Access Journals (Sweden)

Vladyslav Rozhkov

2016-04-01

Full Text Available Purpose: explore peculiarities morphological indexes of the pushers of the nucleus at the stage of specialized basic preparation. Material & Methods: the study was attended by 12 core pushers 15–17 years who were at the stage of specialized basic training. We used the following methods: analysis and synthesis of scientific and technical literature, the definition of anthropometric indicators index method. Results: presented morphological indexes of the pushers of the nucleus at the stage of specialized basic preparation. Conclusions: the figures obtained showed that at the stage of specialized basic preparation the somatotype of shot-putters corresponds the somatotype of the highly qualified shot-putters.

Protein kinases responsible for the phosphorylation of the nuclear egress core complex of human cytomegalovirus.

Science.gov (United States)

Sonntag, Eric; Milbradt, Jens; Svrlanska, Adriana; Strojan, Hanife; Häge, Sigrun; Kraut, Alexandra; Hesse, Anne-Marie; Amin, Bushra; Sonnewald, Uwe; Couté, Yohann; Marschall, Manfred

2017-10-01

Nuclear egress of herpesvirus capsids is mediated by a multi-component nuclear egress complex (NEC) assembled by a heterodimer of two essential viral core egress proteins. In the case of human cytomegalovirus (HCMV), this core NEC is defined by the interaction between the membrane-anchored pUL50 and its nuclear cofactor, pUL53. NEC protein phosphorylation is considered to be an important regulatory step, so this study focused on the respective role of viral and cellular protein kinases. Multiply phosphorylated pUL50 varieties were detected by Western blot and Phos-tag analyses as resulting from both viral and cellular kinase activities. In vitro kinase analyses demonstrated that pUL50 is a substrate of both PKCα and CDK1, while pUL53 can also be moderately phosphorylated by CDK1. The use of kinase inhibitors further illustrated the importance of distinct kinases for core NEC phosphorylation. Importantly, mass spectrometry-based proteomic analyses identified five major and nine minor sites of pUL50 phosphorylation. The functional relevance of core NEC phosphorylation was confirmed by various experimental settings, including kinase knock-down/knock-out and confocal imaging, in which it was found that (i) HCMV core NEC proteins are not phosphorylated solely by viral pUL97, but also by cellular kinases; (ii) both PKC and CDK1 phosphorylation are detectable for pUL50; (iii) no impact of PKC phosphorylation on NEC functionality has been identified so far; (iv) nonetheless, CDK1-specific phosphorylation appears to be required for functional core NEC interaction. In summary, our findings provide the first evidence that the HCMV core NEC is phosphorylated by cellular kinases, and that the complex pattern of NEC phosphorylation has functional relevance.

Analyses of containment loading by hydrogen burning during hypothetical core meltdown accidents

International Nuclear Information System (INIS)

Bracht, K.; Tiltmann, M.

1983-01-01

The possibility of occurance of violent hydrogen burning during a LWR meltdown accident and its consequences to containment atmosphere conditions are discussed. Two accident sequences with low and high system pressure during the in-vessel-melt phase of a meltdown accident are considered. In both sequences only deflagration, but no detonation may become possible, presuming homogeneity of the containment atmospheres. In a low pressure szenario the pressure increase due to deflagration will not reach the failure pressure of the containment, if combustion takes place when the flammability limit is reached. For the special situation of a rapid release of steam and hydrogen after a high-pressure failure of a reactor pressure vessel, calculations with a multicompartment code show that the possibility for hydrogen burning does not exist. Thus, an additional augmentation of the steam spike as a consequence of the failure of the pressure vessel cannot occur. (orig.)

Analysis of core damage frequency from internal events: Peach Bottom, Unit 2

International Nuclear Information System (INIS)

Kolaczkowski, A.M.; Lambright, J.A.; Ferrell, W.L.; Cathey, N.G.; Najafi, B.; Harper, F.T.

1986-10-01

This document contains the internal event initiated accident sequence analyses for Peach Bottom, Unit 2; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Peach Bottom, Unit 2, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provided additional insights regarding the dominant contributors to the Peach Bottom core damage frequency estimate. The mean core damage frequency at Peach Bottom was calculated to be 8.2E-6. Station blackout type accidents (loss of all ac power) were found to dominate the overall results. Anticipated Transient Without Scram accidents were also found to be non-negligible contributors. The numerical results are largely driven by common mode failure probability estimates and to some extent, human error. Because of significant data and analysis uncertainties in these two areas (important, for instance, to the most dominant scenario in this study), it is recommended that the results of the uncertainty and sensitivity analyses be considered before any actions are taken based on this analysis

Effect analysis of core barrel openings under CEFR normal condition

International Nuclear Information System (INIS)

Zhang Yabo; Yang Hongyi

2008-01-01

Openings on the bottom of core barrel are important part of the decay heat removal system of China Experimental Fast Reactor (CEFR), which are designed to discharge the decay heat from reactor under accident condition. This paper analyses the effect of the openings design on the normal operation condition using the famouse CFD code CFX. The result indicates that the decay heat can be discharged safely and at the same time the effect of core barrel openings on the normal operation condition is acceptable. (authors)

Improving core surgical training in a major trauma centre.

Science.gov (United States)

Morris, Daniel L J; Bryson, David J; Ollivere, Ben J; Forward, Daren P

2016-06-01

English Major Trauma Centres (MTCs) were established in April 2012. Increased case volume and complexity has influenced trauma and orthopaedic (T&O) core surgical training in these centres. To determine if T&O core surgical training in MTCs meets Joint Committee on Surgical Training (JCST) quality indicators including performance of T&O operative procedures and consultant supervised session attendance. An audit cycle assessing the impact of a weekly departmental core surgical trainee rota. The rota included allocated timetabled sessions that optimised clinical and surgical learning opportunities. Intercollegiate Surgical Curriculum Programme (ISCP) records for T&O core surgical trainees at a single MTC were analysed for 8 months pre and post rota introduction. Outcome measures were electronic surgical logbook evidence of leading T&O operative procedures and consultant validated work-based assessments (WBAs). Nine core surgical trainees completed a 4 month MTC placement pre and post introduction of the core surgical trainee rota. Introduction of core surgical trainee rota significantly increased the mean number of T&O operative procedures led by a core surgical trainee during a 4 month MTC placement from 20.2 to 34.0 (pcore surgical trainee during a 4 month MTC placement was significantly increased (0.3 vs 2.4 [p=0.04]). Those of dynamic hip screw fixation (2.3 vs 3.6) and ankle fracture fixation (0.7 vs 1.6) were not. Introduction of a core surgical trainee rota significantly increased the mean number of consultant validated WBAs completed by a core surgical trainee during a 4 month MTC placement from 1.7 to 6.6 (pcore surgical trainee rota utilising a 'problem-based' model can significantly improve T&O core surgical training in MTCs. Copyright © 2016 Elsevier Ltd. All rights reserved.

Health trends in a geriatric and special needs clinic patient population.

Science.gov (United States)

Lee, Katherine J; Ettinger, Ronald L; Cowen, Howard J; Caplan, Daniel J

2015-01-01

To quantify differences and recent changes in health status among patients attending the Geriatric and Special Needs Dentistry (GSND) and Family Dentistry (FAMD) clinics at the University of Iowa College of Dentistry. A total of 388 randomly selected records from patients attending the GSND or FAMD clinics from 1996-2000 or from 2006-2010 were reviewed. Univariate and bivariate analyses were conducted, followed by multivariable logistic regression analyses to compare characteristics of patients across clinics. Between the two GSND cohorts, the mean number of medications reported increased from 4.0 to 6.5 (p Special Care Dentistry Association and Wiley Periodicals, Inc.

Severe accident recriticality analyses (SARA)

Energy Technology Data Exchange (ETDEWEB)

Frid, W. E-mail: wiktor.frid@ski.se; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Nilsson, L.; Puska, E.K.; Sjoevall, H

2001-11-01

Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality--both super-prompt power bursts and quasi steady-state power generation--for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s{sup -1} injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g{sup -1}, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s{sup -1}. In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated

Severe accident recriticality analyses (SARA)

International Nuclear Information System (INIS)

Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Nilsson, L.; Puska, E.K.; Sjoevall, H.

2001-01-01

Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality--both super-prompt power bursts and quasi steady-state power generation--for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s -1 injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g -1 , was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s -1 . In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated quasi steady

Analysis of steam generator plugging on core thermohydraulic performance of NPP Krsko; Analiza vpliva cepljenja cevi v upravljaniku na termohidravliko sredice JE Krsko

Energy Technology Data Exchange (ETDEWEB)

Kostadinov, V; Petelin, S; Sarler, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)

1988-07-01

Nuclear safety analysis of NPP Krsko core operating at full power with 4% steam generator tubes plugged have been performed. Influence of individual parameters on core thermohydraulic performance have been evaluated. Using COBRA-III-C computer code we have analysed a core design (evaluation) model. The DNBR change was calculated as a consequence of 4% plugging. The influence of thermohydraulic parameters change on DNBR was analysed. (author)

A core syllabus for the teaching of neuroanatomy to medical students.

Science.gov (United States)

Moxham, Bernard; McHanwell, Stephen; Plaisant, Odile; Pais, Diogo

2015-09-01

There is increasingly a call for clinical relevance in the teaching of biomedical sciences within all health care courses. However, this presupposes that there is a clear understanding of what can be considered core material within the curricula. To date, the anatomical sciences have been relatively poorly served by the development of core syllabuses, particularly for specialized core syllabuses such as neuroanatomy. One of the aims of the International Federation of Associations of Anatomists (IFAA) and of the European Federation for Experimental Morphology (EFEM) is to formulate, on an international scale, core syllabuses for all branches of the anatomical sciences using Delphi Panels consisting of anatomists, scientists, and clinicians to initially evaluate syllabus content. In this article, the findings of a Delphi Panel for neuroanatomy are provided. These findings will subsequently be published on the IFAA website to enable anatomical (and other cognate learned) societies and individual anatomists, clinicians, and students to freely comment upon, and elaborate and amend, the syllabuses. The aim is to set internationally recognized standards and thus to provide guidelines concerning neuroanatomical knowledge when engaged in course development. © 2015 Wiley Periodicals, Inc.

Athabasca tar sand reservoir properties derived from cores and logs

International Nuclear Information System (INIS)

Woodhouse, R.

1976-01-01

Log interpretation parameters for the Athabasca Tar Sand Lease No. 24 have been determined by careful correlation with Dean and Stark core analysis data. Significant expansion of Athabasca cores occurs as overburden pressure is removed. In the more shaly sands the core analysis procedures remove adsorbed water from the clays leading to further overestimation of porosity and free water volume. Log interpretation parameters (R/sub w/ = 0.5 ohm . m and m = n = 1.5) were defined by correlation with the weight of tar as a fraction of the weight of rock solids (grain or dry weight fraction of tar). This quantity is independent of the water content of the cores, whereas porosity and the weight of tar as a fraction of the bulk weight of fluids plus solids (bulk weight fraction) are both dependent on water content. Charts are provided for the conversion of bulk weight fraction of fluids to porosity; grain weight fraction of fluids to porosity; log derived porosity and core grain weight tar to water saturation. Example results show that the core analysis grain weight fraction of tar is adequately matched by the log analyses. The log results provide a better representation of the reservoir fluid volumes than the core analysis data

Restraint system for core elements of a reactor core

International Nuclear Information System (INIS)

Class, G.

1975-01-01

In a nuclear reactor, a core element bundle formed of a plurality of side-by-side arranged core elements is surrounded by restraining elements that exert a radially inwardly directly restraining force generating friction forces between the core elements in a restraining plane that is transverse to the core element axes. The adjoining core elements are in rolling contact with one another in the restraining plane by virtue of rolling-type bearing elements supported in the core elements. (Official Gazette)

Development of the Northern European Ribes core collection based on a microsatellite (SSR) marker diversity analysis

DEFF Research Database (Denmark)

Antonius, Kristiina; Karhu, S.; Kaldmäe, H.

2012-01-01

The purpose of the study was to support the selection process of the most valuable currant and gooseberry accessions cultivated in Northern Europe, in order to establish a decentralized core collection and, following the selection, to ensure sufficient genetic diversity in the selected collection....... Molecular analyses of the material from nine project partners were run at seven different laboratories. The results were first analysed for each partner separately, and then combined to ensure sufficient genetic diversity in the core collection....

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)

Radiometric analyses of floodplain sediments at the Savannah River Plant

International Nuclear Information System (INIS)

Lower, M.W.

1987-09-01

A Comprehensive Cooling Water Study to assess the effects of reactor cooling water discharges and related reactor area liquid releases to onsite streams and the nearby Savannah River has been completed at the US Department of Energy's Savannah River Plant (SRP). Extensive radiometric analyses of man-made and naturally occurring gamma-emitting radionuclides were measured in floodplain sediment cores extracted from onsite surface streams at SRP and from the Savannah River. Gamma spectrometric analyses indicate that reactor operations contribute to floodplain radioactivity levels slightly higher than levels associated with global fallout. In locations historically unaffected by radioactive releases from SRP operations, Cs-137 concentrations were found at background and fallout levels of about 1 pCi/g. In onsite streams that provided a receptor for liquid radioactive releases from production reactor areas, volume-weighted Cs-137 concentrations ranged by core from background levels to 55 pCi/g. Savannah River sediments contained background and atmospheric fallout levels of Cs-137 only. 2 refs., 5 figs

Test of In-core Flux Detectors in KNK II

International Nuclear Information System (INIS)

Hoppe, P.; Mitzel, F.

1979-10-01

The development of in-core detectors for Liquid Metal Fast Breeder Reactors (LMFBRs) is still in an early stage, and little operation experience is available. Therefore self-powered neutron and gamma detectors and neutron sensitive ionization chambers -especially developed for LMFBRs- have been tested in the Fast Sodium Cooled Test Reactor KNK II. Seven flux detectors have been installed in the core of KNK II by means of a special test rig. Five of them failed already within the first week during operation in the reactor. Due to measurements of electrical resistances and capacities, sodium penetrating into the detectors or cables probably seems to be the cause. As tests prior to the installation in the core proved the tightness of all detectors, it is suspected that small cracks have developed in the detector casings or in the outer cable sheaths during their exposure to the hot coolant. Two ionization chambers did not show these faults. However, one of them failed because the saturation current plateau disappeared and the other one's sensitivity decreased by a factor of five during the test period. It is suspected that in both cases changes of the filling gas might be involved

Behavior of composite sandwich panels with several core designs at different impact velocities

Science.gov (United States)

Jiga, Gabriel; Stamin, Ştefan; Dinu, Gabriela

2018-02-01

A sandwich composite represents a special class of composite materials that is manufactured by bonding two thin but stiff faces to a low density and low strength but thick core. The distance between the skins given by the core increases the flexural modulus of the panel with a low mass increase, producing an efficient structure able to resist at flexural and buckling loads. The strength of sandwich panels depends on the size of the panel, skins material and number or density of the cells within it. Sandwich composites are used widely in several industries, such as aerospace, automotive, medical and leisure industries. The behavior of composite sandwich panels with different core designs under different impact velocities are analyzed in this paper by numerical simulations performed on sandwich panels. The modeling was done in ANSYS and the analysis was performed through LS-DYNA.

Lessons learned from radioactive/mixed waste analyses at EG ampersand G Idaho, Inc

International Nuclear Information System (INIS)

Murphy, R.J.; Sailer, S.J.; Bennett, J.T.; Arvizu, J.S.

1990-01-01

For the past 30 years extensive chemical characterizations of environmental and waste samples have been performed by numerous academic, commercial, and government analytical chemistry laboratories for the purposes of research, monitoring, and compliance with regulations. The vast majority of these analyses, however, has been conducted on samples containing natural concentrations of radioactive constituents. It is only within the last decade that a small number of laboratories have been conducting extensive chemical characterizations of highly radioactive samples and consequently have begun to identify many special requirements for the safe and accurate conduct of such analyses. Experience gained from chemical analyses of radioactively contaminated samples has indicated special requirements and actions needed in the following three general areas: Sample collection and preservation; chemical analysis protocols; disposal of waste from chemical analyses. In this paper we will summarize the experience and findings acquired from four years of radioactive sample analyses by the Environmental Chemistry Unit, an analytical chemistry laboratory of EG ampersand G Idaho, Inc. at the Idaho National Engineering Laboratory. 6 tabs

Integrating the older/special needs adoptive child into the family.

Science.gov (United States)

Clark, Pamela; Thigpen, Sally; Yates, Amy Moeller

2006-04-01

This qualitative, grounded theory study investigated 11 families who reported having successfully integrated into their family unit at least one older/special needs adoptee. The theory that emerged through the constant comparative methodology consisted of two categories (Decision to Adopt and Adjustment) and a core category (Developing a Sense of Family). The two categories and core category comprised a process that was informed by the Family Narrative Paradigm and culminated in the successful integration of the child or children into the existing family unit. Parental perceptions that appeared to facilitate this process included: (a) finding strengths in the children overlooked by previous caregivers, (b) viewing behavior in context, (c) reframing negative behavior, and (d) attributing improvement in behavior to parenting efforts.

Radiometric studies of box cores from the Ontong-Java plateau

International Nuclear Information System (INIS)

Krishnamurthy, R.V.; Lal, D.; Somayajulu, B.L.K.; Berger, W.H.

1979-01-01

Five box cores, 30-35 cm deep, from the Ontong-Java Plateau in the Pacific Ocean have been analysed for several radioisotopes, bringing the total number of cores thus studied to nine. All cores contain calcareous sediments; they are from water depths between 1600 and 4300 m. The studies were made with a view to measure CaCO 3 accumulation rates in the equatorial Pacific and to understand the nature of bioturbation and of erosion effects on carbonate accumulation. The 14 C based accumulation rates vary from between 0.7 and 3 cm/k yr. The upper-most disturbed layer in the cores showing distinct effects of surface processes, including benthic mixing, is between 4 and 10 cm thick, in good agreement with the mean values for mixed layer thicknesses of 8 +- 2 cm reported earlier. The nature of mixing is very complicated. It does not resemble diffusion; considerable mixing seems to have occurred due to discrete events. The effects of in-situ dissolution during Holocene accumulation are clearly seen in two cores taken below 4000m. (auth.)

PWR plant transient analyses using TRAC-PF1

International Nuclear Information System (INIS)

Ireland, J.R.; Boyack, B.E.

1984-01-01

This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

EL-Khawlani, Afrah [Physics Department, Sana' a (Yemen); Aziz, Moustafa [Nuclear and radiological regulatory authority, Cairo (Egypt); Ismail, Mahmud Yehia; Ellithi, Ali Yehia [Cairo Univ. (Egypt). Faculty of Science

2015-03-15

The MCNPX (Monte Carlo N-Particle Transport Code System) code has been used for modeling and simulation of a half core of CANDU (CANada Deuterium-Uranium) reactor, both homogenous and heterogeneous model for the reactor core are designed. The fuel is burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor for homogeneous and heterogeneous reactor core is calculated and compared during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the model. The flux and power distributions through channels are calculated.

Fuzzy logic based power-efficient real-time multi-core system

CERN Document Server

Ahmed, Jameel; Najam, Shaheryar; Najam, Zohaib

2017-01-01

This book focuses on identifying the performance challenges involved in computer architectures, optimal configuration settings and analysing their impact on the performance of multi-core architectures. Proposing a power and throughput-aware fuzzy-logic-based reconfiguration for Multi-Processor Systems on Chip (MPSoCs) in both simulation and real-time environments, it is divided into two major parts. The first part deals with the simulation-based power and throughput-aware fuzzy logic reconfiguration for multi-core architectures, presenting the results of a detailed analysis on the factors impacting the power consumption and performance of MPSoCs. In turn, the second part highlights the real-time implementation of fuzzy-logic-based power-efficient reconfigurable multi-core architectures for Intel and Leone3 processors. .

Robust Detection and Visualization of Jet-Stream Core Lines in Atmospheric Flow.

Science.gov (United States)

Kern, Michael; Hewson, Tim; Sadlo, Filip; Westermann, Rudiger; Rautenhaus, Marc

2018-01-01

Jet-streams, their core lines and their role in atmospheric dynamics have been subject to considerable meteorological research since the first half of the twentieth century. Yet, until today no consistent automated feature detection approach has been proposed to identify jet-stream core lines from 3D wind fields. Such 3D core lines can facilitate meteorological analyses previously not possible. Although jet-stream cores can be manually analyzed by meteorologists in 2D as height ridges in the wind speed field, to the best of our knowledge no automated ridge detection approach has been applied to jet-stream core detection. In this work, we -a team of visualization scientists and meteorologists-propose a method that exploits directional information in the wind field to extract core lines in a robust and numerically less involved manner than traditional 3D ridge detection. For the first time, we apply the extracted 3D core lines to meteorological analysis, considering real-world case studies and demonstrating our method's benefits for weather forecasting and meteorological research.

Seismic analysis methods for LMFBR core and verification with mock-up vibration tests

International Nuclear Information System (INIS)

Sasaki, Y.; Kobayashi, T.; Fujimoto, S.

1988-01-01

This paper deals with the vibration behaviors of a cluster of core elements with the hexagonal cross section in a barrel under the dynamic excitation due to seismic events. When a strong earthquake excitation is applied to the core support, the cluster of core elements displace to a geometrical limit determined by restraint rings in the barrel, and collisions could occur between adjacent elements as a result of their relative motion. For these reasons, seismic analysis on LMFBR core elements is a complicated non-linear vibration problem, which includes collisions and fluid interactions. In an actual core design, it is hard to include hundreds of elements in the numerical calculations. In order to study the seismic behaviors of core elements, experiments with single row 29 elements (17 core fuel assemblies, 4 radial blanket assemblies, and 8 neutron shield assemblies) simulated all elements in MONJU core central row, and experiments with 7 cluster rows of 37 core fuel assemblies in the core center were performed in a fluid filled tank, using a large-sized shaking table. Moreover, the numerical analyses of these experiments were performed for the validation of simplified and detailed analytical methods. 4 refs, 18 figs

Preliminary analyses for HTTR`s start-up physics tests by Monte Carlo code MVP

Energy Technology Data Exchange (ETDEWEB)

Nojiri, Naoki [Science and Technology Agency, Tokyo (Japan); Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

1998-08-01

Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

Special interest in decision making in entrepreneurship policy

DEFF Research Database (Denmark)

Bager, Torben; Klyver, Kim; Schou Nielsen, Pia

2015-01-01

The study investigates the role of the special interests of key decision makers in entrepreneurship policy formation at the national level. An ethnographic method is applied to analyse in depth the 2005 decision by the Danish Government to shift from volume oriented to growth oriented...... entrepreneurship policy. The theoretical value of this paper is its challenge to the widespread rationality view in the entrepreneurship field and a deepened understanding of how the pursuit of special interests is related to ambiguous evidence and system-level rationality....

analysis of reactivity accidents in MTR for various protection system parameters and core condition

International Nuclear Information System (INIS)

Mohamed, F.M.

2011-01-01

Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99 Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99 Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99 Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

Organisation of facilities management in relation to core business

DEFF Research Database (Denmark)

Jensen, Per Anker

2011-01-01

as mainly a specific customer orientation. It is concluded that a market relationship – internally or externally – is appropriate for non-strategic functions, while it is important to create a kind of coalition between strategic FM functions and the core business management. Originality/value: The paper......Purpose: The purpose of this article is to clarify the organisational relationships between Facilities Management (FM) and core business and how these relationships vary for strategic and operational support functions. Approach: The research takes a starting point in Michael Porter’s theory...... of value chains but also draws on theory of strategic FM, governance and forms of coordination. The value chains for core businesses and support functions are analysed and related to empirical data from a case study on a broadcasting corporation during a major relocation. Findings: A particular support...

Analysis of the Ford Nuclear Reactor LEU core

Energy Technology Data Exchange (ETDEWEB)

Rathkopf, J A; Drumm, C R; Martin, W R; Lee, J C [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

1983-09-01

This paper has summarized the current status of the effort to analyze the FNR HEU/LEU cores and to compare the calculated results with measurements. In general, calculated predictions of experimental results are quite good, especially for global parameters such as reactivity, as seen in the single HEU/LEU element substitution experiment and the LEU full core critical loading. Shim rod worths are predicted well for two of the rods but too high for a third rod possibly due to inaccurate thermal flux distribution calculation. The calculated thermal flux maps show excellent agreement with experiment throughout the FNR core. In the heavy water tank, however, experimental values for the thermal flux obtained by different methods are inconsistent among themselves as well as with the calculated finding. Work is under.way to use our computational tools to correct the discrepancies between the various measurement techniques and to improve the computational results for flux distribution and the rod worth experiment. Although uncertainties exist in our analysis, as evidenced by the discrepancies mentioned above, we consider our present calculational package to be a useful, reasonably accurate, and efficient system for performing analyses of MTR LEU/HEU core configurations.

VHTR core modeling: coupling between neutronic and thermal-hydraulics

International Nuclear Information System (INIS)

Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.

2005-01-01

Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)

Analysis of space-time core dynamics on reactor accident at Chernobyl

International Nuclear Information System (INIS)

Takano, Makoto; Shindo, Ryuichi; Yamashita, Kiyonobu; Sawa, Kazuhiro

1987-05-01

Regarding reactor accident at Chernobyl in USSR, core dynamics has been analyzed by COMIC code which solves space-time dependent diffusion equation in three-dimension taking spatial thermohydraulic effect into account. The code was originally developed for high temperature gas-cooled reactors (HTGR), however, has been modified to include light water as coolant, instead of helium, for analysis of the accident. In the analysis, emphasis is placed on spatial effects on core dynamics. The analyses are performed for the cases of modeling the core fully and partially where 6 fuel channels surround one control rod channel. The result shows that the speed of applying void reactivity averaged over the core depends on the power and coolant flow distributions. Therefore, these distributions have potential to influence on the value and the time of peak power estimated by calculation. (author)

Random close packing in protein cores

OpenAIRE

Gaines, Jennifer C.; Smith, W. Wendell; Regan, Lynne; O'Hern, Corey S.

2015-01-01

Shortly after the determination of the first protein x-ray crystal structures, researchers analyzed their cores and reported packing fractions $\\phi \\approx 0.75$, a value that is similar to close packing equal-sized spheres. A limitation of these analyses was the use of `extended atom' models, rather than the more physically accurate `explicit hydrogen' model. The validity of using the explicit hydrogen model is proved by its ability to predict the side chain dihedral angle distributions obs...

Neutron spectrum effects on TRU recycling in Pb-Bi cooled fast reactor core

International Nuclear Information System (INIS)

Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

2003-01-01

This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction

New Focus on the Tales of the Earth—Legacy Cores Redistribution Project Completed

Directory of Open Access Journals (Sweden)

Ursula Röhl

2009-03-01

Full Text Available Scientific drilling for marine cores began in 1968 under the auspices of the Deep Sea Drilling Project (DSDP, whose initial discoveries included salt domes on the sea floor and formation of oceanic crust by sea-floor spreading along the mid-ocean ridges rift zone. Analyses of cores in various laboratories all over the world provided key information toward a better understanding of Earth’s past, present, and future including the geology of the sea floor, evolution of the Earth, and past climatic changes. With an eye towards future development of analytical tools for core-based research, it was important to maintain cores in as close to their original condition as possible for the years to come. This led to the establishment of large repositories curating cores at 4ºC, conducting sub-sampling, and facilitating non-destructive observation of cores while following well-defined curation policies.

Natural convection cooling of LEU cores for Pakistan research reactor-1

International Nuclear Information System (INIS)

Khan, L.A.; Bokhari, I.H.; Akhtar, K.M.

1991-08-01

The first high power and equilibrium LEU cores of PARR-1 have been analysed to assess the maximum operating power based on natural convection cooling, need for forced cooling to remove the decay heat and to estimate safety margins that commensurate with the predetermined power limit. Computer code NATCON and standard correlations have been used for the analysis. The parameters studied includes coolant velocity, temperature distribution in the core, heat fluxes at onset of nucleate boiling, pulsed boiling and burnup. (author)

Investigation of Equilibrium Core by recycling MA and LLFP in fast reactor cycle (I)

International Nuclear Information System (INIS)

Mizutani, Akihiko; Shono, Akira; Ishikawa, Makoto

1999-05-01

Feasibility study on a self-consistent fuel cycle system is performed in the nuclear fuel recycle system with fast reactors. In this system, the self-generated MAs (Minor Actinides) and LLFPs (Long Lived Fission Products) are confined and incinerated in the fast reactor. Analyses of the nuclear properties for an 'Equilibrium Core', in which the self-generated MAs and LLFPs are confined, are investigated. A conventional sodium cooled oxide fuel fast reactor is selected as the core specifications for the 'Equilibrium Core'. This 600 MWe fast reactor does not have a radial blanket. In this study, the nuclear characteristics of the 'Equilibrium Core' are compared with those of a 'Standard Core' and '5 w/oMA Core'. The 'Standard Core' does not confine MAs and LLFPs in the core, and a 5 w/o-MA Rom LWR is loaded in the '5 w/oMA Core'. Through this comparison between 'Equilibrium Core' and the others, the specific characters of the 'Equilibrium Core' are investigated. In order to realize the 'Equilibrium Core' in the viewpoint of nuclear properties, whether the conventional design concept of fast reactors must be changed or not is also evaluated. The analyses for the nitride and metallic fuel cores are also performed because of their different nuclear characteristics compared with the oxide fuel core. Assuming the separation of REs (Rare Earth elements) from MAs and the isotope separation of LLFPs, most of the nuclear properties for the 'Equilibrium Core' are not beyond those for the '5 w/oMA Core'. It is, therefore, possible to bring the 'Equilibrium Core' into existence without any drastic modification for the design concept of the typical oxide fuel fast reactors. Although the 15.1[w/o] LLFPs are loading in the core of the oxide fuel 'Equilibrium Core', a breeding ratio is more than 1.0 and the difference in a amount of plutonium between a charging and discharging is only 0.04 [ton/year]. Without any drastic change for the design concept of the conventional oxide fuel

Introduction to special issue: moving forward in pediatric neuropsychology.

Science.gov (United States)

Daly, Brian P; Giovannetti, Tania; Zabel, T Andrew; Chute, Douglas L

2011-08-01

This special issue of The Clinical Neuropsychologist focuses on advances in the emerging subspecialty of pediatric neuropsychology. The national and international contributions in this issue cover a range of key clinical, research, training, and professional issues specific to pediatric neuropsychology. The genesis for this project developed out of a series of talks at the Philadelphia Pediatric Neuropsychology Symposium in 2010, hosted by the Stein Family Fellow, the Department of Psychology of the College of Arts and Sciences at Drexel University, and the Philadelphia Neuropsychology Society. Articles that explore clinical practice issue focus on the assessment of special medical populations with congenital and/or acquired central nervous system insults. Research articles investigate the core features of developmental conditions, the use of technology in neuropsychological research studies, and large sample size genomic, neuropsychological, and imaging studies of under-represented populations. The final series of articles examine new considerations in training, advocacy, and subspecialty board certification that have emerged in pediatric neuropsychology. This introductory article provides an overview of the articles in this special issue and concluding thoughts about the future of pediatric neuropsychology.

Procedures for the external event core damage frequency analyses for NUREG-1150

International Nuclear Information System (INIS)

Bohn, M.P.; Lambright, J.A.

1990-11-01

This report presents methods which can be used to perform the assessment of risk due to external events at nuclear power plants. These methods were used to perform the external events risk assessments for the Surry and Peach Bottom nuclear power plants as part of the NRC-sponsored NUREG-1150 risk assessments. These methods apply to the full range of hazards such as earthquakes, fires, floods, etc. which are collectively known as external events. The methods described in this report have been developed under NRC sponsorship and represent, in many cases, both advancements and simplifications over techniques that have been used in past years. They also include the most up-to-date data bases on equipment seismic fragilities, fire occurrence frequencies and fire damageability thresholds. The methods described here are based on making full utilization of the power plant systems logic models developed in the internal events analyses. By making full use of the internal events models one obtains an external event analysis that is consistent both in nomenclature and in level of detail with the internal events analyses, and in addition, automatically includes all the appropriate random and tests/maintenance unavailabilities as appropriate. 50 refs., 9 figs., 11 tabs

Core data from offshore Puerto Rico and the U.S. Virgin Islands

Science.gov (United States)

Hoy, Shannon K.; Chaytor, Jason D.; ten Brink, Uri S.

2014-01-01

In 2008, as a collaborative effort between Woods Hole Oceanographic Institution and the U.S. Geological Survey, 20 giant gravity cores were collected from areas surrounding Puerto Rico and the U.S. Virgin Islands. The regions sampled have had many large earthquake and landslide events, some of which are believed to have triggered tsunamis. The objective of this coring cruise, carried out aboard the National Oceanic and Atmospheric Administration research vessel Seward Johnson, was to determine the age of several substantial slope failures and seismite layers near Puerto Rico in an effort to map their temporal distribution. Data gathered from the cores collected in 2008 and 11 archive cores from the Lamont-Doherty Earth Observatory are included in this report. These data include lithologic logs, core summary sheets, x-ray fluorescence, wet-bulk density, magnetic susceptibility, grain-size analyses, radiographs, and radiocarbon age dates.

The KNK II instrumentation for global and local supervision of the reactor core

International Nuclear Information System (INIS)

Steiger, W.O.

1990-01-01

After an introduction into the KNK plant itself, their historical development and their present situation, the instrumentation of the global and local supervision of the KNK II-core as well as the main safety-related i- and c-systems are described. Special emphasis is laid on the instrumentation of the reactor protection systems and the shutdown systems. After that some practices are reported about instrumentation behavior and lessons learned from the operation and maintenance of the above mentioned systems. At last follows a short description of the special instrumentation for the detection of failed fuel subassemblies and of the plant data processing system. (orig.)

The KNK II instrumentation for global and local supervision of the reactor core

International Nuclear Information System (INIS)

Steiger, W.O.

1991-01-01

After an introduction into the KNK plant itself, their historical development and their present situation, the instrumentation of the global and local supervision of the KNK II-core as well as the main safety-related instrumentation and control systems is described. Special emphasis is laid on the instrumentation of the reactor protection systems and the shut down systems. After that some practices are reported about instrumentation behavior and lessons learned from the operation and maintenance of the above mentioned systems. At last follows a short description of the special instrumentation for the detection of failed fuel subassemblies and of the plant data processing system. (author). 4 refs, 18 tabs

Structural assessment of TAPS core shroud under accident loads

International Nuclear Information System (INIS)

Bhasin, Vivek; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1996-09-01

Over the last few years, the Core Shroud of Boiling Water Reactors (BWRs) operating in foreign countries, have developed cracks at weld locations. As a first step for assessment of structural safety of Tarapur Atomic Power Station (TAPS) core shroud, its detailed stress analysis was done for postulated accident loads. This report is concerned with structural assessment of core shroud, of BWR at TAPS, subjected to loads resulting from main steam line break (MSLB), recirculation line break (RLB) and safe shut down earthquake. The stress analysis was done for core shroud in healthy condition and without any crack since, visual examination conducted till now, do not indicate presence of any flaw. Dynamic structural analysis for MSLB and RLB events was done using dynamic load factor (DLF) method. The complete core shroud and its associated components were modelled and analysed using 3D plate/shell elements. Since, the components of core shroud are submerged in water, hence, hydrodynamic added mass was also considered for evaluation of natural frequencies. It was concluded that from structural point of view, adequate safety margin is available under all the accident loads. Nonlinear analysis was done to evaluate buckling/collapse load. The collapse/buckling load have sufficient margin against the allowable limits. The displacements are low hence, the insertion of control rod may not be affected. (author)

Thermohydraulic behaviour and heat transfer in the molten core

International Nuclear Information System (INIS)

Reineke, H.H.

1977-01-01

Increasing the application of nuclear reactors to produce electrical power extremely unprobable accidents should be investigated too. In the Federal Republic of Germany, a research program is performed for some years engaged in accidents at light water reactors in which the melting of the reactor core is presumed. A part of this program is to investigate the thermohydraulic and the heat transfer behavior in an accumulation of molten core material. The knowledge of these events is necessary to analyse the accident exactly. Further on the results of this work are of great importance to build a catcher for the molten core material. As a result of the decay heat the molten material is heated up and the density differences induce a free convection motion. In this work the thermohydraulic behavior and the distribution of the escaping heat fluxes for several accumulations of molten core material were determined. The numerical methods for solving the system of partial differential equation were used to develop computer codes, able to compute the average and local heat fluxes at the walls enclosing the molten core material and the inside increase of the temperature. The numerical computations were confirmed and verified by experimental investigations. In these investigations the molten core material was always assumed as a homogeneous fluid. In this case, the results could be reproduced by simple power laws

A soft-core processor architecture optimised for radar signal processing applications

CSIR Research Space (South Africa)

Broich, R

2013-12-01

Full Text Available -performance soft-core processing architecture is proposed. To develop such a processing architecture, data and signal-flow characteristics of common radar signal processing algorithms are analysed. Each algorithm is broken down into signal processing...

Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

International Nuclear Information System (INIS)

Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

2016-01-01

), Centre d'etude de l'energie Nucleaire (SCK.CEN), and University of Illinois (UIUC). The contributors applied different calculation methodologies as described in Chapter 3. For the four SFR cores modelled, the neutronics parameters such as the k-effective, beta effective, Doppler coefficient, sodium void worth, control rods worth, power map and isotopic content were obtained at the beginning and end of the equilibrium cycle. The results obtained by the various institutions are summarised in Chapter 4. The variations in the results obtained are analysed in Chapter 5 in order to understand the origin of differences and derive adequate recommendations. Finally, the conclusions of the study are provided in Chapter 6

Depressurization accident analyses for the Fort St. Vrain Reactor

International Nuclear Information System (INIS)

Paul, D.D.

1976-01-01

Design-basis depressurization accident analyses for the Fort St. Vrain reactor were performed using the FLODIS (Ref. 4) code. The FLODIS code models the active core, side reflector, gas annulus between the core barrel and the PCRV liner, and the PCRV cooling system. Results are presented for the Pelton circulators operating at 10,550, 8800, and 7000 rpm. Maximum temperatures of selected components are plotted as a function of time during the transient. None of the components studied exceeded the temperature at which failure or damage may occur. However, there must be sufficient mixing of the outlet gas in the lower plenum to insure the integrity of the steel liners of the steam generator inlet ducts

Environmental and individual determinants of core and non-core food and drink intake in preschool-aged children in the United Kingdom.

Science.gov (United States)

McGowan, L; Croker, H; Wardle, J; Cooke, L J

2012-03-01

Strategies to achieve healthier diets for children are likely to benefit from an understanding of the determinants. We examined environmental and individual predictors of children's intake of 'core' foods (fruit and vegetables) and 'non-core' foods (snacks and sweetened beverages). Predictors included parental intake, home availability, parental feeding styles (Encouragement and Monitoring) and children's food preferences. Based on research with older children, we expected intake of both food types to be associated with maternal intake, core foods to be more associated with children's preferences and non-core food intake more with the home environment. Primary caregivers (n=434) of children (2-5 years) from preschools and Children's Centres in London, UK, completed a self-report survey in 2008. Multiple regression analyses indicated children's fruit intake was associated with maternal fruit intake (B=0.29; P=0.000), children's liking for fruit (B=0.81; P=0.000) and a Monitoring style of parental feeding (B=0.13; P=0.021). Children's vegetable intake was similarly associated with maternal intake (B=0.39; P=0.000), children's liking for vegetables (B=0.77; P=0.000), Encouragement (B=0.19; P=0.021) and Monitoring (B=0.11; P=0.029). Non-core snack intake was associated with maternal intake (B=0.25; P=0.029), Monitoring (B=-0.16; P=0.010), home availability (B=0.10; P=0.022) and television viewing (TV) (B=0.28; P=0.012). Non-core drink intake was associated with maternal intake (B=0.32; P=0.000) and TV (B=0.20; P=0.019). Results indicate commonalities and differences in the predictors of core and non-core food intake, with only maternal intake being important across all types. Effective interventions to improve young children's diets may need to call on different strategies for different foods.

Development of seismic analysis model for HTGR core on commercial FEM code

International Nuclear Information System (INIS)

Tsuji, Nobumasa; Ohashi, Kazutaka

2015-01-01

The aftermath of the Great East Japan Earthquake prods to revise the design basis earthquake intensity severely. In aseismic design of block-type HTGR, the securement of structural integrity of core blocks and other structures which are made of graphite become more important. For the aseismic design of block-type HTGR, it is necessary to predict the motion of core blocks which are collided with adjacent blocks. Some seismic analysis codes have been developed in 1970s, but these codes are special purpose-built codes and have poor collaboration with other structural analysis code. We develop the vertical 2 dimensional analytical model on multi-purpose commercial FEM code, which take into account the multiple impacts and friction between block interfaces and rocking motion on contact with dowel pins of the HTGR core by using contact elements. This model is verified by comparison with the experimental results of 12 column vertical slice vibration test. (author)

Systematic investigation of the synthesis of core-shell poly(styrene-co-acrylic acid) colloids with varying shell thickness and core diameter

DEFF Research Database (Denmark)

Hinge, Mogens; Keiding, Kristian

2006-01-01

the morphology of the material for an specific application is going on. It is known from SFEP of styrene that the final colloidal size can be controlled by adjusting the ionic strength of the synthesis feed [1] and it is suggested that adding acrylic acid to the synthesis will result in a change...... in polymerization locus from the core to the surface [2]. There is at present not performed a systematically investigation in controlling the core size and shell thickness of poly(styrene-co-acrylic acid) core-shell colloids  (poly(ST-co-AA)).   Poly(ST-co-AA) colloids were synthesized by free-radical surfactant......-free emulsion co-polymerization (SFECP) at 70°C, using styrene as monomer and acrylic acid as co-monomer. Different batches of poly(ST-co-AA) colloids were synthesized with varying ionic strength and acrylic acid concentrations in the synthesis feed. The produced poly(ST-co-AA) colloids were analysed...

Advanced toroidal facility vaccuum vessel stress analyses

International Nuclear Information System (INIS)

Hammonds, C.J.; Mayhall, J.A.

1987-01-01

The complex geometry of the Advance Toroidal Facility (ATF) vacuum vessel required special analysis techniques in investigating the structural behavior of the design. The response of a large-scale finite element model was found for transportation and operational loading. Several computer codes and systems, including the National Magnetic Fusion Energy Computer Center Cray machines, were implemented in accomplishing these analyses. The work combined complex methods that taxed the limits of both the codes and the computer systems involved. Using MSC/NASTRAN cyclic-symmetry solutions permitted using only 1/12 of the vessel geometry to mathematically analyze the entire vessel. This allowed the greater detail and accuracy demanded by the complex geometry of the vessel. Critical buckling-pressure analyses were performed with the same model. The development, results, and problems encountered in performing these analyses are described. 5 refs., 3 figs

HTGR core seismic analysis using an array processor

International Nuclear Information System (INIS)

Shatoff, H.; Charman, C.M.

1983-01-01

A Floating Point Systems array processor performs nonlinear dynamic analysis of the high-temperature gas-cooled reactor (HTGR) core with significant time and cost savings. The graphite HTGR core consists of approximately 8000 blocks of various shapes which are subject to motion and impact during a seismic event. Two-dimensional computer programs (CRUNCH2D, MCOCO) can perform explicit step-by-step dynamic analyses of up to 600 blocks for time-history motions. However, use of two-dimensional codes was limited by the large cost and run times required. Three-dimensional analysis of the entire core, or even a large part of it, had been considered totally impractical. Because of the needs of the HTGR core seismic program, a Floating Point Systems array processor was used to enhance computer performance of the two-dimensional core seismic computer programs, MCOCO and CRUNCH2D. This effort began by converting the computational algorithms used in the codes to a form which takes maximum advantage of the parallel and pipeline processors offered by the architecture of the Floating Point Systems array processor. The subsequent conversion of the vectorized FORTRAN coding to the array processor required a significant programming effort to make the system work on the General Atomic (GA) UNIVAC 1100/82 host. These efforts were quite rewarding, however, since the cost of running the codes has been reduced approximately 50-fold and the time threefold. The core seismic analysis with large two-dimensional models has now become routine and extension to three-dimensional analysis is feasible. These codes simulate the one-fifth-scale full-array HTGR core model. This paper compares the analysis with the test results for sine-sweep motion

Activation and Shielding Analyses in Support of the GUINEVERE Project

International Nuclear Information System (INIS)

Serikov, A.; Fischer, U.; Mercatali, L.; Baeten, P.; Vittiglio, G.

2008-01-01

The GUINEVERE facility (Generator of Uninterrupted Intense Neutrons at the lead Venus Reactor) must satisfy the nuclear safety criteria required by the Belgian safety authority to be licensed. The radiation dose and activation analyses for the nuclear safety assessment of the GUINEVERE project were performed at FZK. The concerted efforts of several European institutions were concentrated on the development and construction of a subcritical fast lead core based on the Venus water moderated reactor at the SCK-CEN site in Mol, Belgium. A Monte Carlo (MC) MCNP5 model was developed in accordance with the current design of the GUINEVERE fast lead core. The analytical MC method does not work for shielding analysis of the GUINEVERE building because of the large size of the rooms and thick concrete walls and floors. MC variance reduction techniques, such as particles splitting, Russian roulette, and point detectors were therefore applied. The JEFF-3.1 nuclear data library was used for radiation transport calculations. The activation analyses for the lead core and building materials were performed with the FISPACT-2005 inventory code and the EAF-2005 library. The neutron and photon dose rate maps were produced using MCNP track-length estimations, point detectors, and a mesh tally superimposed over the GUINEVERE geometry. The effects of D-D and D-T fusion neutron sources were estimated. (authors)

Activation and Shielding Analyses in Support of the GUINEVERE Project

Energy Technology Data Exchange (ETDEWEB)

Serikov, A.; Fischer, U.; Mercatali, L. [Association FZK-EURATOM, KIT, Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Baeten, P.; Vittiglio, G. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

2008-07-01

The GUINEVERE facility (Generator of Uninterrupted Intense Neutrons at the lead Venus Reactor) must satisfy the nuclear safety criteria required by the Belgian safety authority to be licensed. The radiation dose and activation analyses for the nuclear safety assessment of the GUINEVERE project were performed at FZK. The concerted efforts of several European institutions were concentrated on the development and construction of a subcritical fast lead core based on the Venus water moderated reactor at the SCK-CEN site in Mol, Belgium. A Monte Carlo (MC) MCNP5 model was developed in accordance with the current design of the GUINEVERE fast lead core. The analytical MC method does not work for shielding analysis of the GUINEVERE building because of the large size of the rooms and thick concrete walls and floors. MC variance reduction techniques, such as particles splitting, Russian roulette, and point detectors were therefore applied. The JEFF-3.1 nuclear data library was used for radiation transport calculations. The activation analyses for the lead core and building materials were performed with the FISPACT-2005 inventory code and the EAF-2005 library. The neutron and photon dose rate maps were produced using MCNP track-length estimations, point detectors, and a mesh tally superimposed over the GUINEVERE geometry. The effects of D-D and D-T fusion neutron sources were estimated. (authors)

A comparative design study of PB-BI cooled reactor cores with forced and natural convection cooling

International Nuclear Information System (INIS)

Mizuno, Tomoyasu; Enuma, Yasuhiro; Tanji, Mikio

2003-01-01

A comparative core design study is performed on Pb-Bi cooled reactors with forced and natural convection (FC and NC) cooling. Major interests of the study are core performance and core safety features. The designed core concepts with nitride fuel achieve reasonable breeding capability. The results of unprotected event analyses such as UTOP and ULOF show that both of concepts have possible features to withstand unprotected events due to negative reactivity feedback by Doppler effect, control rod drive line expansion, etc. These results lead to a conclusion that both of concepts have possible capability as one of future promising core concepts. A FC cooling core concept has more advantage if fuel recycle viewpoint is emphasized. (author)

A core handling device for the Mars Sample Return Mission

Science.gov (United States)

Gwynne, Owen

1989-01-01

A core handling device for use on Mars is being designed. To provide a context for the design study, it was assumed that a Mars Rover/Sample Return (MRSR) Mission would have the following characteristics: a year or more in length; visits by the rover to 50 or more sites; 100 or more meter-long cores being drilled by the rover; and the capability of returning about 5 kg of Mars regolith to Earth. These characteristics lead to the belief that in order to bring back a variegated set of samples that can address the range of scientific objetives for a MRSR mission to Mars there needs to be considerable analysis done on board the rover. Furthermore, the discrepancy between the amount of sample gathered and the amount to be returned suggests that there needs to be some method of choosing the optimal set of samples. This type of analysis will require pristine material-unaltered by the drilling process. Since the core drill thermally and mechanically alters the outer diameter (about 10 pct) of the core sample, this outer area cannot be used. The primary function of the core handling device is to extract subsamples from the core and to position these subsamples, and the core itself if needed, with respect to the various analytical instruments that can be used to perform these analyses.

Neutronic feasibility of PWR core with mixed oxide fuels in the Republic of Korea

International Nuclear Information System (INIS)

Kim, Y.J.; Joo, H.K.; Jung, H.G.; Sohn, D.S.

1997-01-01

Neutronic feasibility of a PWR core with mixed oxide (MOX) fuels has been investigated as part of the feasibility study for recycling spent fuels in Korea. A typical 3-loop PWR with 900 MWe capacity is selected as reference plant to develop equilibrium core designs with low-leakage fuel management scheme, while incorporating various MOX loading. The fuel management analyses and limited safety analyses show that, safely stated, MOX recycling with 1/3 reload fraction can be accommodated for both annual and 18 month fuel cycle schemes in Korean PWRs, without major design modifications on the reactor systems. (author). 12 refs, 4 figs, 3 tabs

Siemens experience on linear and nonlinear analyses of out-of-phase BWR instabilities

International Nuclear Information System (INIS)

Kreuter, D.; Wehle, F.

1995-01-01

The Siemens design code STAIF has been applied extensively for linear analysis of BWR instabilities. The comparison between measurements and STAIF calculations for different plants under various conditions has shown good agreement for core-wide and regional instabilities. Based on the high quality of STAIF, the North German TUeV has decided to replace the licensing requirement of extensive stability measurements by predictive analyses with the code STAIF. Nonlinear stability analysis for beyond design boundary conditions with RAMONA has shown dryout during temporarily reversed flow at core inlet in case of core-wide oscillations. For large out-of-phase oscillations, dryout occurs already for small, still positive channel inlet flow. (orig.)

Development of a coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

International Nuclear Information System (INIS)

Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

2011-01-01

Highlights: → Advanced analysis and design techniques for innovative reactors are addressed. → Detailed investigation of a 3 pass core design with a multi-physics-scales tool. → Coupled 40-group neutron transport/equivalent channels TH core analyses methods. → Multi-scale capabilities: from equivalent channels to sub-channel pin-by-pin study. → High fidelity approach: reduction of conservatism involved in core simulations. - Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the

Analysis of core damage frequency from internal events: Surry, Unit 1

International Nuclear Information System (INIS)

Harper, F.T.

1986-11-01

This document contains the accident sequence analyses for Surry, Unit 1; one of the reference plants being examined as part of the NUREG-1150 effort by the Nuclear Regulatory Commission (NRC). NUREG-1150 will document the risk of a selected group of nuclear power plants. As part of that work, this report contains the overall core damage frequency estimate for Surry, Unit 1, and the accompanying plant damage state frequencies. Sensitivity and uncertainty analyses provide additional insights regarding the dominant contributors to the Surry core damage frequency estimate. The numerical results are driven to some degree by modeling assumptions and data selection for issues such as reactor coolant pump seal LOCAs, common cause failure probabilities, and plant response to station blackout and loss of electrical bust initiators. The sensitivity studies explore the impact of alternate theories and data on these issues

Operation and Licensing of Mixed Cores in Water Cooled Reactors

International Nuclear Information System (INIS)

2013-11-01

Nuclear fuel is a highly complex material that is subject to continuous development and is produced by a range of manufacturers. During operation of a nuclear power plant, the nuclear fuel is subject to extreme conditions of temperature, corroding environment and irradiation, and many different designs of fuel have been manufactured with differing fuel materials, cladding materials and assembly structure to ensure these conditions. The core of an operating power plant can contain hundreds of fuel assemblies, and where there is more than a single design of a fuel assembly in the core, whether through a change of fuel vendor, introduction of an improved design or for some other reason, the core is described as a mixed core. The task of ensuring that the different assembly types do not interact in a harmful manner, causing, for example, differing flow resistance resulting in under cooling, is an important part of ensuring nuclear safety. This report has compiled the latest information on the operational experience of mixed cores and the tools and techniques that are used to analyse the core operation and demonstrate that there are no safety related problems with its operation. This publication is a result of a technical meeting in 2011 and a series of consultants meetings

Detailed comparison between computed and measured FBR core seismic responses

International Nuclear Information System (INIS)

Forni, M.; Martelli, A.; Melloni, R.; Bonacina, G.

1988-01-01

This paper presents a detailed comparison between seismic calculations and measurements performed for various mock-ups consisting of groups of seven and nineteen simplified elements of the Italian PEC fast reactor core. Experimental tests had been performed on shaking tables in air and water (simulating sodium) with excitations increasing up to above Safe Shutdown Earthquake. The PEC core-restraint ring had been simulated in some tests. All the experimental tests have been analysed by use of both the one-dimensional computer program CORALIE and the two-dimensional program CLASH. Comparisons have been made for all the instrumented elements, in both the time and the frequency domains. The good agreement between calculations and measurements has confirmed adequacy of the fluid-structure interaction model used for PEC core seismic design verification

Core lifter

Energy Technology Data Exchange (ETDEWEB)

Pavlov, N G; Edel' man, Ya A

1981-02-15

A core lifter is suggested which contains a housing, core-clamping elements installed in the housing depressions in the form of semirings with projections on the outer surface restricting the rotation of the semirings in the housing depressions. In order to improve the strength and reliability of the core lifter, the semirings have a variable transverse section formed from the outside by the surface of the rotation body of the inner arc of the semiring aroung the rotation axis and from the inner a cylindrical surface which is concentric to the outer arc of the semiring. The core-clamping elements made in this manner have the possibility of freely rotating in the housing depressions under their own weight and from contact with the core sample. These semirings do not have weakened sections, have sufficient strength, are inserted into the limited ring section of the housing of the core lifter without reduction in its through opening and this improve the reliability of the core lifter in operation.

Side core lifter

Energy Technology Data Exchange (ETDEWEB)

Edelman, Ya A

1982-01-01

A side core lifter is proposed which contains a housing with guide slits and a removable core lifter with side projections on the support section connected to the core receiver. In order to preserve the structure of the rock in the core sample by means of guaranteeing rectilinear movement of the core lifter in the rock, the support and core receiver sections are hinged. The device is equipped with a spring for angular shift in the core-reception part.

Core mechanics and configuration behavior of advanced LMFBR core restraint concepts

International Nuclear Information System (INIS)

Fox, J.N.; Wei, B.C.

1978-02-01

Core restraint systems in LMFBRs maintain control of core mechanics and configuration behavior. Core restraint design is complex due to the close spacing between adjacent components, flux and temperature gradients, and irradiation-induced material property effects. Since the core assemblies interact with each other and transmit loads directly to the core restraint structural members, the core assemblies themselves are an integral part of the core restraint system. This paper presents an assessment of several advanced core restraint system and core assembly concepts relative to the expected performance of currently accepted designs. A recommended order for the development of the advanced concepts is also presented

Possibilities of utilizing used moulding and core sands by microwave treatment

Directory of Open Access Journals (Sweden)

K. Granat

2011-01-01

Full Text Available The paper presents a semi-industrial reactor designed for microwave utilization of waste moulds and cores made of moulding sandsprepared in furane resin technology. It was found that a possibility exists of effective incinerating this way prepared residues of coresseparated from moulding sands or waste moulds left after casting. The preliminary tests evidenced that microwave heating is an effectiveway of disposing waste moulding sands and the applied apparatus permits effective control of the microwave heating process. The special structure permitting rotations of charge material and proper selection of the generators working cycles guarantee significant speeding-up the process and its full stabilisation. Application of microwave heating for utilization of waste moulds and cores containing synthetic resins as binders ensures significant and measurable economical benefits resulting from shorter process time.

Experimental simulation of fragmentation and stratification of core debris on the core catcher of a fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Pillai, Dipin S.; Vignesh, R. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Sudha, A. Jasmin, E-mail: jasmin@igcar.gov.in [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Pushpavanam, S.; Sundararajan, T. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Nashine, B.K.; Selvaraj, P. [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

2016-05-15

Highlights: • Fragmentation of two simultaneous metals jets in a bulk coolant analysed. • Particle size from experiments compared with theoretical analysis. • Jet breakup modes explained using dimensionless numbers. • Settling aspects of aluminium and lead debris on collector plate studied. • Results analysed in light of core debris settling on core catcher in a FBR. - Abstract: The complex and coupled phenomena of two simultaneous molten metal jets fragmenting inside a quiescent liquid pool and settling on a collector plate are experimentally analysed in the context of safety analysis of a fast breeder reactor (FBR) in the post accident heat removal phase. Following a hypothetical core melt down accident in a FBR, a major portion of molten nuclear fuel and clad/structural material which are collectively termed as ‘corium’ undergoes fragmentation in the bulk coolant sodium in the lower plenum of the reactor main vessel and settles on the core catcher plate. The coolability of this decay heat generating debris bed is dependent on the particle size distribution and its layering i.e., stratification. Experiments have been conducted with two immiscible molten metals of different densities poured inside a coolant medium to understand their fragmentation behaviour and to assess the possibility of formation of a stratified debris bed. Molten aluminium and lead have been used as simulants in place of molten stainless steel and nuclear fuel to facilitate easy handling. This paper summarizes the major findings from these experiments. The fragmentation of the two molten metals are explained in the light of relevant dimensionless numbers such as Reynolds number and Weber Number. The mass median diameter of the fragmented debris is predicted from nonlinear stability analysis of slender jets for lead jet and using Rayleigh's classical theory of jet breakup for aluminium jet. The agreement of the predicted values with the experimental results is good. These

Refurbishment, core conversion and safety analysis of Apsara reactor

Energy Technology Data Exchange (ETDEWEB)

Raina, V.K.; Sasidharan, K.; Sengupta, S. [Bhabha Atomic Research Centre, Mumbai (India)]. E-mail: nram@@apsara.barc.ernet.in

1998-07-01

Apsara, a 1 MWt pool type reactor using HEU fuel has been in operation at the Bhabha Atomic Research Centre, Trombay since 1956. In view of the long service period seen by the reactor it is now planned to carry out extensive refurbishment of the reactor with a view to extend its useful life. It is also proposed to modify the design of the reactor wherein the core will be surrounded by a heavy water reflector tank to obtain a good thermal neutron flux over a large radial distance from the core. Beam holes and the majority of the irradiation facilities will be located inside the reflector tank. The coolant flow direction through the core will be changed from the existing upward flow to downward flow. A delay tank, located inside the pool, is provided to facilitate decay of short lived radioactivity in the coolant outlet from the core in order to bring down radiation field in the operating areas. Analysis of various anticipated operational occurrences and accident conditions like loss of normal power, core coolant flow bypass, fuel channel blockage and degradation of primary coolant pressure boundary have been performed for the proposed design. Details of the proposed design modifications and the safety analyses are given in the paper. (author)

Redefining the modular organization of the core Mediator complex.

Science.gov (United States)

Wang, Xuejuan; Sun, Qianqian; Ding, Zhenrui; Ji, Jinhua; Wang, Jianye; Kong, Xiao; Yang, Jianghong; Cai, Gang

2014-07-01

The Mediator complex plays an essential role in the regulation of eukaryotic transcription. The Saccharomyces cerevisiae core Mediator comprises 21 subunits, which are organized into Head, Middle and Tail modules. Previously, the Head module was assigned to a distinct dense domain at the base, and the Middle and Tail modules were identified to form a tight structure above the Head module, which apparently contradicted findings from many biochemical and functional studies. Here, we compared the structures of the core Mediator and its subcomplexes, especially the first 3D structure of the Head + Middle modules, which permitted an unambiguous assignment of the three modules. Furthermore, nanogold labeling pinpointing four Mediator subunits from different modules conclusively validated the modular assignment, in which the Head and Middle modules fold back on one another and form the upper portion of the core Mediator, while the Tail module forms a distinct dense domain at the base. The new modular model of the core Mediator has reconciled the previous inconsistencies between the structurally and functionally defined Mediator modules. Collectively, these analyses completely redefine the modular organization of the core Mediator, which allow us to integrate the structural and functional information into a coherent mechanism for the Mediator's modularity and regulation in transcription initiation.

Restructuring Vocational Special Needs Education through Interdisciplinary Team Effort: Local Motion in the Pacific Basin.

Science.gov (United States)

Smith, Garnett J.; Stodden, Robert A.

1994-01-01

The Restructuring through Interdisciplinary Team Effort project focuses on changing the culture and structure of vocational special needs education in the Pacific Basin. Its three dimensions are cognitive core (best practices, outcome-focused design, strategic planning); team network of stakeholders; and systemic renewal (school-to-work…

Core competencies for pharmaceutical physicians and drug development scientists

Science.gov (United States)

Silva, Honorio; Stonier, Peter; Buhler, Fritz; Deslypere, Jean-Paul; Criscuolo, Domenico; Nell, Gerfried; Massud, Joao; Geary, Stewart; Schenk, Johanna; Kerpel-Fronius, Sandor; Koski, Greg; Clemens, Norbert; Klingmann, Ingrid; Kesselring, Gustavo; van Olden, Rudolf; Dubois, Dominique

2013-01-01

Professional groups, such as IFAPP (International Federation of Pharmaceutical Physicians and Pharmaceutical Medicine), are expected to produce the defined core competencies to orient the discipline and the academic programs for the development of future competent professionals and to advance the profession. On the other hand, PharmaTrain, an Innovative Medicines Initiative project, has become the largest public-private partnership in biomedicine in the European Continent and aims to provide postgraduate courses that are designed to meet the needs of professionals working in medicines development. A working group was formed within IFAPP including representatives from PharmaTrain, academic institutions and national member associations, with special interest and experience on Quality Improvement through education. The objectives were: to define a set of core competencies for pharmaceutical physicians and drug development scientists, to be summarized in a Statement of Competence and to benchmark and align these identified core competencies with the Learning Outcomes (LO) of the PharmaTrain Base Course. The objectives were successfully achieved. Seven domains and 60 core competencies were identified and aligned accordingly. The effective implementation of training programs using the competencies or the PharmaTrain LO anywhere in the world may transform the drug development process to an efficient and integrated process for better and safer medicines. The PharmaTrain Base Course might provide the cognitive framework to achieve the desired Statement of Competence for Pharmaceutical Physicians and Drug Development Scientists worldwide. PMID:23986704

Core Competencies for Pharmaceutical Physicians and Drug Development Scientists

Directory of Open Access Journals (Sweden)

Honorio eSilva

2013-08-01

Full Text Available Professional groups, such as IFAPP (International Federation of Pharmaceutical Physicians and Pharmaceutical Medicine, are expected to produce the defined core competencies to orient the discipline and the academic programs for the development of future competent professionals and to advance the profession. On the other hand, PharmaTrain, an Innovative Medicines Initiative project, has become the largest public-private partnership in biomedicine in the European Continent and aims to provide postgraduate courses that are designed to meet the needs of professionals working in medicines development. A working group was formed within IFAPP including representatives from PharmaTrain, academic institutions and national member associations, with special interest and experience on Quality Improvement through education. The objectives were: to define a set of core competencies for pharmaceutical physicians and drug development scientists, to be summarized in a Statement of Competence and to benchmark and align these identified core competencies with the Learning Outcomes of the PharmaTrain Base Course. The objectives were successfully achieved. Seven domains and 60 core competencies were identified and aligned accordingly. The effective implementation of training programs using the competencies or the PharmaTrain Learning Outcomes anywhere in the world may transform the drug development process to an efficient and integrated process for better and safer medicines. The PharmaTrain Base Course might provide the cognitive framework to achieve the desired Statement of Competence for Pharmaceutical Physicians and Drug Development Scientists worldwide.

Localized Effects in the Nonlinear Behavior of Sandwich Panels with a Transversely Flexible Core

DEFF Research Database (Denmark)

Frostig, Y.; Thomsen, Ole Thybo

2005-01-01

This paper presents the results of an investigation of the role of localized effects within the geometrically nonlinear domain on structural sandwich panels with a "compliant" core. Special emphasis is focused on the nonlinear response near concentrated loads and stiffened core regions. The adopted...... nonlinear analysis approach incorporates the effects of the vertical flexibility of the core, and it is based on the approach of the High-order Sandwich Panel Theory (HSAPT). The results demonstrate that the effects of localized loads, when taken into the geometrically nonlinear domain, change the response...... of the panel from a strength problem controlled by stress constraints into a stability problem with unstable limit point behavior when force-controlled loads are applied. The stability problem emerge as the nonlinear response develops with the formation of a small number of buckling waves in the compressed...

Toroidal HTS transformer with cold magnetic core - analysis with FEM software

International Nuclear Information System (INIS)

Grzesik, B; Stepien, M; Jez, R

2010-01-01

The aim of this paper is to present a thorough characterization of the toroidal HTS transformer by means of FEM analysis. The analysis was a 2D/3D harmonic electromagnetic and thermal analysis. The toroidal transformer operated in LN2 by being immersed together with the magnetic core in it, for which its power losses were acceptable. Two extreme variants of windings were analysed. The first one called parallel and the second called perpendicular. Three variants of the magnetic core were considered. In the first one the core was put outside of the windings, in the second the core was inside of the windings and in the third variant the core was outside as well as inside of the windings. The windings were made of HTS tape BiSCCO-2223/Ag while the magnetic core was made of the nanocrystalline material Finemet. The two windings, with a 1:1 turn-to-turn ratio, were uniformly distributed along the whole torus circumference. The output power, efficiency and power density are in the results of the analysis. The temperature distribution was also calculated. In summary, the performance of the transformer is better than those currently known.

Core food of the French food supply: second Total Diet Study.

Science.gov (United States)

Sirot, V; Volatier, J L; Calamassi-Tran, G; Dubuisson, C; Menard, C; Dufour, A; Leblanc, J C

2009-05-01

As first described in the 1980s, the core food intake model allows a precise assessment of dietary nutrient intake and dietary exposure to contaminants insofar as it reflects the eating habits of a target population and covers the most important foods in terms of consumption, selected nutrient and contaminant contribution. This model has been used to set up the sampling strategy of the second French Total Diet Study (TDS) with the aim of obtaining a realistic panorama of nutrient intakes and contaminant exposure for the whole population, useful for quantitative risk assessment. Data on consumption trends and eating habits from the second French individual food consumption survey (INCA2) as well as data from a 2004 purchase panel of French households (SECODIP) were used to identify the core foods to be sampled. A total of 116 core foods on a national scale and 70 core foods on a regional scale were selected according to (1) the consumption data for adults and children, (2) their consumer rates, and (3) their high contribution to exposure to one or more contaminants of interest. Foods were collected in eight French regions (36 cities) and prepared 'as consumed' to be analysed for their nutritional composition and contamination levels. A total of 20 280 different food products were purchased to make up the 1352 composite samples of core foods to be analysed for additives, environmental contaminants, pesticide residues, trace elements and minerals, mycotoxins and acrylamide. The establishment of such a sampling plan is essential for effective, high-quality monitoring of dietary exposure from a public health point of view.

Determination of Educational Needs and Self-Efficacy Perceptions of Special Education Teachers

Science.gov (United States)

Ozcan, Deniz; Uzunboylu, Huseyin

2017-01-01

The purpose of this study is to analyse the need for curriculum development of special education teachers who work at special education centres and schools with resource rooms with regard to different variables and determine their perceptions of self-efficacy. In this study, a general survey model was employed that allows a general opinion about…

Social-Emotional Characteristics and Special Educational Needs

Science.gov (United States)

Meijer, Joost; Fossen, Miriam W. E. B.; van Putten, Cornelis M.; van der Leij, Aryan

2006-01-01

The aim of the research described in this article was the development of an instrument to measure social emotional characteristics and special educational and pedagogical needs of students in the last grade of primary education. Questionnaires were developed for teachers as well as for students. Exploratory factor analyses showed that the factors…

Non-linear Dynamic Analysis of Steel Hollow I-core Sandwich Panel under Air Blast Loading

Directory of Open Access Journals (Sweden)

Asghar Vatani Oskouei

2015-12-01

Full Text Available In this paper, the non-linear dynamic response of novel steel sandwich panel with hollow I-core subjected to blast loading was studied. Special emphasis is placed on the evaluation of midpoint displacements and energy dissipation of the models. Several parameters such as boundary conditions, strain rate, mesh dependency and asymmetrical loading are considered in this study. The material and geometric non-linearities are also considered in the numerical simulation. The results obtained are compared with available experimental data to verify the developed FE model. Modeling techniques are described in detail. According to the results, sandwich panels with hollow I-core allowed more plastic deformation and energy dissipation and less midpoint displacement than conventional I-core sandwich panels and also equivalent solid plate with the same weight and material.

Comparison of Histologic Core Portions Acquired from a Core Biopsy Needle and a Conventional Needle in Solid Mass Lesions: A Prospective Randomized Trial.

Science.gov (United States)

Lee, Ban Seok; Cho, Chang-Min; Jung, Min Kyu; Jang, Jung Sik; Bae, Han Ik

2017-07-15

The superiority of endoscopic ultrasound-guided fine needle biopsy (EUS-FNB) over EUS-guided fine needle aspiration (EUS-FNA) remains controversial. Given the lack of studies analyzing histologic specimens acquired from EUS-FNB or EUS-FNA, we compared the proportion of the histologic core obtained from both techniques. A total of 58 consecutive patients with solid mass lesions were enrolled and randomly assigned to the EUS-FNA or EUS-FNB groups. The opposite needle was used after the failure of core tissue acquisition using the initial needle with up to three passes. Using computerized analyses of the scanned histologic slide, the overall area and the area of the histologic core portion in specimens obtained by the two techniques were compared. No significant differences were identified between the two groups with respect to demographic and clinical characteristics. Fewer needle passes were required to obtain core specimens in the FNB group (pcore (11.8%±19.5% vs 8.0%±11.1%, p=0.376) or in the diagnostic accuracy (80.6% vs 81.5%, p=0.935) between two groups. The proportion of histologic core and the diagnostic accuracy were comparable between the FNB and FNA groups. However, fewer needle passes were required to establish an accurate diagnosis in EUS-FNB.

How cores grow by pebble accretion. I. Direct core growth

Science.gov (United States)

Brouwers, M. G.; Vazan, A.; Ormel, C. W.

2018-03-01

Context. Planet formation by pebble accretion is an alternative to planetesimal-driven core accretion. In this scenario, planets grow by the accretion of cm- to m-sized pebbles instead of km-sized planetesimals. One of the main differences with planetesimal-driven core accretion is the increased thermal ablation experienced by pebbles. This can provide early enrichment to the planet's envelope, which influences its subsequent evolution and changes the process of core growth. Aims: We aim to predict core masses and envelope compositions of planets that form by pebble accretion and compare mass deposition of pebbles to planetesimals. Specifically, we calculate the core mass where pebbles completely evaporate and are absorbed before reaching the core, which signifies the end of direct core growth. Methods: We model the early growth of a protoplanet by calculating the structure of its envelope, taking into account the fate of impacting pebbles or planetesimals. The region where high-Z material can exist in vapor form is determined by the temperature-dependent vapor pressure. We include enrichment effects by locally modifying the mean molecular weight of the envelope. Results: In the pebble case, three phases of core growth can be identified. In the first phase (Mcore mixes outwards, slowing core growth. In the third phase (Mcore > 0.5M⊕), the high-Z inner region expands outwards, absorbing an increasing fraction of the ablated material as vapor. Rainout ends before the core mass reaches 0.6 M⊕, terminating direct core growth. In the case of icy H2O pebbles, this happens before 0.1 M⊕. Conclusions: Our results indicate that pebble accretion can directly form rocky cores up to only 0.6 M⊕, and is unable to form similarly sized icy cores. Subsequent core growth can proceed indirectly when the planet cools, provided it is able to retain its high-Z material.

Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

International Nuclear Information System (INIS)

Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

2001-01-01

The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

Mechanical and thermo-mechanical response of a lead-core bearing device subjected to different loading conditions

Directory of Open Access Journals (Sweden)

Zhelyazov Todor

2018-01-01

Full Text Available The contribution is focused on the numerical modelling, simulation and analysis of a lead-core bearing device for passive seismic isolation. An accurate finite element model of a lead-core bearing device is presented. The model is designed to analyse both mechanical and thermo-mechanical responses of the seismic isolator to different loading conditions. Specifically, the mechanical behaviour in a typical identification test is simulated. The response of the lead-core bearing device to circular sinusoidal paths is analysed. The obtained shear displacement – shear force relationship is compared to experimental data found in literature sources. The hypothesis that heating of the lead-core during cyclic loading affects the degrading phenomena in the bearing device is taken into account. Constitutive laws are defined for each material: lead, rubber and steel. Both predefined constitutive laws (in the used general–purpose finite element code and semi-analytical procedures aimed at a more accurate modelling of the constitutive relations are tested. The results obtained by finite element analysis are to be further used to calibrate a macroscopic model of the lead-core bearing device seen as a single-degree-of-freedom mechanical system.

Neural evidence that human emotions share core affective properties.

Science.gov (United States)

Wilson-Mendenhall, Christine D; Barrett, Lisa Feldman; Barsalou, Lawrence W

2013-06-01

Research on the "emotional brain" remains centered around the idea that emotions like fear, happiness, and sadness result from specialized and distinct neural circuitry. Accumulating behavioral and physiological evidence suggests, instead, that emotions are grounded in core affect--a person's fluctuating level of pleasant or unpleasant arousal. A neuroimaging study revealed that participants' subjective ratings of valence (i.e., pleasure/displeasure) and of arousal evoked by various fear, happiness, and sadness experiences correlated with neural activity in specific brain regions (orbitofrontal cortex and amygdala, respectively). We observed these correlations across diverse instances within each emotion category, as well as across instances from all three categories. Consistent with a psychological construction approach to emotion, the results suggest that neural circuitry realizes more basic processes across discrete emotions. The implicated brain regions regulate the body to deal with the world, producing the affective changes at the core of emotions and many other psychological phenomena.

Magnetic collimation of fast electrons in specially engineered targets irradiated by ultraintense laser pulses

International Nuclear Information System (INIS)

Cai Hongbo; Zhu Shaoping; Wu Sizhong; Chen Mo; Zhou Cangtao; He, X. T.; Yu Wei; Nagatomo, Hideo

2011-01-01

The efficient magnetic collimation of fast electron flow transporting in overdense plasmas is investigated with two-dimensional collisional particle-in-cell numerical simulations. It is found that the specially engineered targets exhibiting either high-resistivity-core-low-resistivity-cladding structure or low-density-core-high-density-cladding structure can collimate fast electrons. Two main mechanisms to generate collimating magnetic fields are found. In high-resistivity-core-low-resistivity-cladding structure targets, the magnetic field at the interfaces is generated by the gradients of the resistivity and fast electron current, while in low-density-core-high-density-cladding structure targets, the magnetic field is generated by the rapid changing of the flow velocity of the background electrons in transverse direction (perpendicular to the flow velocity) caused by the density jump. The dependences of the maximal magnetic field on the incident laser intensity and plasma density, which are studied by numerical simulations, are supported by our analytical calculations.

A cost-benefit analysis on the specialization in departments of obstetrics and gynecology in Japan.

Science.gov (United States)

Shen, Junyi; Fukui, On; Hashimoto, Hiroyuki; Nakashima, Takako; Kimura, Tadashi; Morishige, Kenichiro; Saijo, Tatsuyoshi

2012-03-27

In April 2008, the specialization in departments of obstetrics and gynecology was conducted in Sennan area of Osaka prefecture in Japan, which aims at solving the problems of regional provision of obstetrical service. Under this specialization, the departments of obstetrics and gynecology in two city hospitals were combined as one medical center, whilst one hospital is in charge of the department of gynecology and the other one operates the department of obstetrics. In this paper, we implement a cost-benefit analysis to evaluate the validity of this specialization. The benefit-cost ratio is estimated at 1.367 under a basic scenario, indicating that the specialization can generate a net benefit. In addition, with a consideration of different kinds of uncertainty in the future, a number of sensitivity analyses are conducted. The results of these sensitivity analyses suggest that the specialization is valid in the sense that all the estimated benefit-cost ratios are above 1.0 in any case.

Analysis of core damage frequency, Surry, Unit 1 internal events appendices

International Nuclear Information System (INIS)

Bertucio, R.C.; Julius, J.A.; Cramond, W.R.

1990-04-01

This document contains the appendices for the accident sequence analyses of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 documents the risk of a selected group of nuclear power plants. The work performed is an extensive reanalysis of that published in November 1986 as NUREG/CR-4450, Volume 3. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA practitioners who need to know how the work was performed and the details for use in further studies. The mean core damage frequency at Surry was calculated to be 4.0E-5 per year, with a 95% upper bound of 1.3E-4 and 5% lower bound of 6.8E-6 per year. Station blackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were Loss of Coolant Accidents (LOCAs). These sequences account for 15% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency

Upper-bound fission product release assessment for large break LOCA in CANFLEX bundle reactor core

International Nuclear Information System (INIS)

Oh, Duk Ju; Lee, Kang Moon

1996-07-01

Quarter-core gap inventory assessment for CANDU-6 reactor core loaded with CANFLEX fuel bundles has been performed as one of the licensing safety analyses required for 24 natural uranium CANFLEX bundle irradiation in CANDU-6 reactor. The quarter-core gap inventory for the CANFLEX bundle core is 5 - 10 times lower than that for the standard bundle core, depending on the half-life of the isotope. The lower gap inventory of the CANFLEX bundle core is attributed to the lower linear power of the CANFLEX bundle compared with the standard bundle. However, the whole core total inventories for both the CANFLEX and standard bundle cores are nearly the same. The 6 - 8 times lower upper-bound fission product releases of the CANFLEX bundle core for large break LOCA than those of the standard bundle core imply that the loading of 24 natural uranium CANFLEX bundles would improve the predicted consequences of the postulated accident described in the Wolsung 2 safety report. 2 tabs., 6 figs., 3 refs. (Author)

IMPROVEMENT OF STRATEGIC MANIPULATED FEDERAL PROPERTY THE EXAMPLE NON-CORE ASSETS OF JSC «CENTER OF NUCLEAR INDUSTRY NONCORE ASSETS» STATE CORPORATION «ROSATOM»

OpenAIRE

Ilya I. Rodin

2015-01-01

The article describes the main measures to improve the management of assets, federally-owned or private of public corporations - an inventory of the property, the recognition of non-core assets, the organization of decision-making systems, the sale of non-core assets at market value. The article provides the rationale for the creation within the large state-owned corporations specialized management companies responsible for the restructuring of non-core assets and improve management of the pr...

PWR boron dilution transients. Thermal-hydraulic analyses of PKL-E experiments

International Nuclear Information System (INIS)

Pietro Alessandro Di Maio; Antonino Tomasello; Giuseppe Vella

2005-01-01

refilling. The reflux condenser phase could produce a different amount of un-borated water in the pump loop seals and a different boron concentration in the RPV according to the actual boron concentration in the primary coolant at incident start, that is a function of burnup. Moreover, since other parameters are directly correlated with core burnup (e.g.: the amount of burnable and permanent poisons) their effects have been investigated on a postulated SBLOCA starting from different initial core conditions. The analyses performed show that in the case of a SBLOCA, the break section area and the HPIS flow injection rate could affect the instant in which natural circulation stops, the reflux condenser time phase length and, consequently, the amount of low borated water that gathers in the pump loop seals. The analyses also show that during reflux condenser phase the condensate inside the loop seals is actual composed of low borated water and the boron concentration inside the reactor core can increase reaching very high values. Nevertheless the formation of un-borated water slugs is interfered by the injection of borated water which, partially, heads for the loop seal where it mixes with the un-borated water descending from the steam generator U tubes. The analyses show that after shut down of the system the core reactivity keep on going down because of the increase in core poisons, in particular Xe and Sm. When the primary system refilling allows natural circulation starting again an increase in the core reactivity is registered, due to the cold and low borated coolant that reaches the core from the pump seals. In all examined cases the total core reactivity never became positive and consequently it seems that boron dilution events during SBLOCA does not cause serious core damage. (authors)

PWR boron dilution transients. Thermal-hydraulic analyses of PKL-E experiments

Energy Technology Data Exchange (ETDEWEB)

Pietro Alessandro Di Maio; Antonino Tomasello; Giuseppe Vella [Dipartimento di Ingegneria Nucleare, Viale delle Scienze, 90128 Palermo (Italy)

2005-07-01

refilling. The reflux condenser phase could produce a different amount of un-borated water in the pump loop seals and a different boron concentration in the RPV according to the actual boron concentration in the primary coolant at incident start, that is a function of burnup. Moreover, since other parameters are directly correlated with core burnup (e.g.: the amount of burnable and permanent poisons) their effects have been investigated on a postulated SBLOCA starting from different initial core conditions. The analyses performed show that in the case of a SBLOCA, the break section area and the HPIS flow injection rate could affect the instant in which natural circulation stops, the reflux condenser time phase length and, consequently, the amount of low borated water that gathers in the pump loop seals. The analyses also show that during reflux condenser phase the condensate inside the loop seals is actual composed of low borated water and the boron concentration inside the reactor core can increase reaching very high values. Nevertheless the formation of un-borated water slugs is interfered by the injection of borated water which, partially, heads for the loop seal where it mixes with the un-borated water descending from the steam generator U tubes. The analyses show that after shut down of the system the core reactivity keep on going down because of the increase in core poisons, in particular Xe and Sm. When the primary system refilling allows natural circulation starting again an increase in the core reactivity is registered, due to the cold and low borated coolant that reaches the core from the pump seals. In all examined cases the total core reactivity never became positive and consequently it seems that boron dilution events during SBLOCA does not cause serious core damage. (authors)

Thermal hydraulic design of PFBR core

International Nuclear Information System (INIS)

Roychowdhury, D.G.; Vinayagam, P.P.; Ravichandar, S.C.

2000-01-01

The thermal-hydraulic design of core is important in respecting temperature limits while achieving higher outlet temperature. This paper deals with the analytical process developed and implemented for analysing steady state thermal-hydraulics of PFBR core. A computer code FLONE has been developed for optimisation of flow allocation through the subassemblies (SA). By calibrating β n (ratio between the maximum channel temperature rise and SA average temperature rise) values with SUPERENERGY code and using these values in FLONE code, prediction of average and maximum coolant temperature distribution is found to be reasonably accurate. Hence, FLONE code is very powerful design tool for core design. A computer code SAPD has been developed to calculate the pressure drop of fuel and blanket SA. Selection of spacer wire pitch depends on the pressure drop, flow-induced vibration and the mixing characteristics. A parametric study was made for optimisation of spacer wire pitch for the fuel SA. Experimental programme with 19 pin-bundle has been undertaken to find the flow-induced vibration characteristics of fuel SA. Also, experimental programme has been undertaken on a full-scale model to find the pressure drop characteristics in unorificed SA, orifices and the lifting force on the SA. (author)

Comparison of facility characteristics between SCTF Core-I and Core-II

International Nuclear Information System (INIS)

Adachi, Hiromichi; Iwamura, Takamichi; Sobajima, Makoto; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio.

1990-08-01

The Slab Core Test Facility (SCTF) was constructed to investigate two-dimensional thermal-hydraulics in the core and fluid behavior of carryover water out of the core including its feed-back effect to the core behavior mainly during the reflood phase of a large break loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). Since three simulated cores are used in the SCTF Test Program and the design of these three cores are slightly different one by one, repeatability test is required to justify a direct comparison of data obtained with different cores. In the present report, data of Test S2-13 (Run 618) obtained with SCTF Core-II were compared with those of Test S1-05 (Run 511) obtained with the Core-I, which were performed under the forced-flooding condition. Thermal-hydraulic behaviors in these two tests showed quite similar characteristics of both system behavior and two-dimensional core behaviors. Therefore, the test data obtained from the two cores can be compared directly with each other. After the turnaround of clad temperatures, however, some differences were found in upper plenum water accumulation and resultant two-dimensional core cooling behaviors such as quench front propagation from bottom to top of the core. (author)

Special Status of Budapest, the Capital of Hungary

OpenAIRE

Temesi, István

2012-01-01

Hungarian capital, Budapest, has always had a special legal status within the system of self-government, except between 1949 and 1990. It is organised in two-tiers: it functions a single local self-government unit (the City of Budapest); while at the same time, its 23 districts enjoy their self-government powers. The paper analyses the history of organisation of Budapest is analysed, as well as the current system of local self-government in Hungary, in order to identify historical and current...

An Introduction to the Special Issue. Politics of Precarity

DEFF Research Database (Denmark)

Schierup, Carl-Ulrik; Bak Jørgensen, Martin

2016-01-01

The current special issue examines the range and strength of analysing contemporary transformations and struggles through the lens of ‘precarity’. Rather than defining a single precariat, the interest is in exploring ‘varieties of precarity’. These take different forms in different parts of the w......The current special issue examines the range and strength of analysing contemporary transformations and struggles through the lens of ‘precarity’. Rather than defining a single precariat, the interest is in exploring ‘varieties of precarity’. These take different forms in different parts...... of the world, on different scales and in different socio-economic contexts, and yet they share certain characteristics in terms of conditions and capacity for agency. Contributions to this volume testify that precarity may be a political proposition as much as a sociological category that offers an analytical...

Synthesis method validation for Super-Phenix 1 start-up core studies