Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

Energy Technology Data Exchange (ETDEWEB)

Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Marie, N.; Prulhière, G.; Lecerf, J. [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France)

2016-02-15

Highlights: • Low void worth CFV and SFRv2 cores are compared for ASTRID pre-conceptual design. • Severe accident behaviour is assessed with a simplified calculation approach and tools. • Mitigation to limit reactivity inserted by core compaction is easier for CFV than for SFRv2 core. • When facing arbitrary reactivity ramps, CFV core would lead to lower energy release than SFRv2 core. • Time scale for core degradation is one order of magnitude larger for CFV than for SFRv2. - Abstract: The present paper is dedicated to the studies carried out during the first stage of the pre-conceptual design of the French demonstrator of fourth generation SFR reactors (ASTRID) in order to compare the behaviour of two envisaged core concepts under severe accident transients. Among the two studied core concepts, whose powers are 1500 MWth, the first one is a classical homogeneous core (called SFRv2) with large pin diameter whose the sodium overall voiding reactivity effect is 5 $. The second concept is an axially heterogeneous core (called CFV) whose global void reactivity effect is negative (−1.2 $ at the end of cycle at the equilibrium). The comparison of the cores relies on two typical accident families: a reactivity insertion (unprotected transient overpower, UTOP) and an overall loss of core cooling (unprotected loss of flow, ULOF). In the first part of the comparison, the primary phase of an UTOP is studied in order to assess typical features of the transient behaviour: power and reactivity evolutions, material heating and melting/vaporization and mechanical energy release due to fuel vapor expansion. The second part of the comparison deals with the calculation of the reactivity potential for degraded states (molten pools) representative of the secondary phase of a mild UTOP and of a strong UTOP (strong or mild qualifies the reactivity ramp inserted). According to the reactivity potential, the amount of fuel to extract from the core and the amount of absorber

Microscopic findings in EUS-guided fine needle (SharkCore) biopsies with type 1 and type 2 autoimmune pancreatitis

DEFF Research Database (Denmark)

Detlefsen, Sönke; Joergensen, Maiken Thyregod; Mortensen, Michael Bau

2017-01-01

The International Consensus Diagnostic Criteria (ICDC) for the diagnosis of autoimmune pancreatitis (AIP) include the histological criterion that is based on either pancreatic core needle biopsies (CNBs) or surgical specimens. However, CNBs are difficult to obtain by endoscopic ultrasound (EUS......). EUS fine-needle aspiration (EUS-FNA) cytology is usually not sufficient for the diagnosis of AIP, but may sometimes contain tissue microfragments. Another approach is EUS-guided histological fine-needle biopsy (EUS-FNB), using needles such as the SharkCore or ProCore needle. Published data regarding...... EUS-guided SharkCore FNB for the diagnosis of AIP are lacking. We aimed to describe our histological findings in one type 1 and two type 2 AIP patients who underwent EUS SharkCore FNB. The EUS-FNBs of two patients fulfilled the histological level 2 ICDC for type 1 AIP or type 2 AIP. The EUS-FNB of one...

Power distribution gradients in WWER type cores and fuel failure root causes

Energy Technology Data Exchange (ETDEWEB)

Mikuš, Ján M., E-mail: JanMikus.nrc@hotmail.com

2014-02-15

Highlights: • Power (fission rate) distribution gradients can represent fuel failure root causes. • Positions with above gradients were investigated in WWER type cores on reactor LR-0. • Above gradients were evaluated near core heterogeneities and construction materials. • Results can be used for code validation and fuel failure occurrence investigation. - Abstract: Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs): Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores, Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp., Neighboring baffle in WWER-1000 type cores, and Neighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part. An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation.

Three Types of Earth's Inner Core Boundary

Science.gov (United States)

Tian, D.; Wen, L.

2017-12-01

The Earth's inner core boundary (ICB) is the site where the liquid outer core solidifies and the solid inner core grows. Thus, the fine-scale structure of the ICB is important for our understanding of the thermo-compositional state of the Earth's core. In this study, we collect a large set of seismic records with high-quality pre-critical PKiKP and PcP phase pairs, recorded by two dense seismic arrays, Hi-net in Japan and USArray in US. This dataset samples the ICB regions beneath East Asia, Mexico and the Bering Sea. We use differential travel times, amplitude ratios and waveform differences between PKiKP and PcP phases to constrain fine-scale structure of the ICB. The sampled ICB can be grouped into three types based on their seismic characteristics: (1) a simple ICB with a flat and sharp boundary, (2) a bumpy ICB with topographic height changes of 10 km, and (3) a localized mushy ICB with laterally varying thicknesses of 4-8 km. The laterally varying fine-scale structure of the ICB indicates existence of complex small-scale forces at the surface and a laterally varying solidification process of the inner core due to lateral variation of thermo-compositional condition near the ICB.

Contribution to the build-up of a core calculation frame: comparison between ''diffusion'' and ''SPn'' operators on various configurations of the first N4 core

International Nuclear Information System (INIS)

Magat, Ph.

1997-04-01

The aim of this study is to compare two calculation methods implemented in the neutronic code CRONOS 2: the diffusion approximation and the SP n method. The APOLLO 2 code is used to build the multiparameter cross section libraries.The comparison is based on the first core of N4 type Chooz reactor. The rod worth and the power map have been calculated. Some recommendations about the SP n development order of flux are made and the results show that the diffusion calculations over-estimate the black rod efficiency up to 10%. (A.C.)

Uncertainty evaluatins of CASMO-3/MASTER system for PWR core neutronics calculations

International Nuclear Information System (INIS)

Song, Jae Seung; Kim, Kang Seog; Lee, Kibog; Park, Jin Ha; Zee, Sung Quun

1996-01-01

Uncertainties in core neutronic calculations of CASMO-3/MASTER, which is a KAERI developed core nuclear design code system, were evaluated via comparisons with measured data. Comparisons were performed with plant measurement data from one Westinghouse type and one ABB-CE type plant and two Korean standard type plants. The CASMO-3/MASTER capability and levels of accuracy are concluded to be sufficient for the neutronics design including safety related parameters related with reactivity, power distributions, temperature and power coefficients, inverse boron worth and control bank worth

Seismic responses of a pool-type fast reactor with different core support designs

International Nuclear Information System (INIS)

Wu, Ting-shu; Seidensticker, R.W.

1989-01-01

In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs

A study on Monte Carlo analysis of Pebble-type VHTR core for hydrogen production

International Nuclear Information System (INIS)

Kim, Hong Chul

2005-02-01

In order to pursue exact the core analysis for VHTR core which will be developed in future, a study on Monte Carol method was carried out. In Korea, pebble and prism type core are under investigation for VHTR core analysis. In this study, pebble-type core was investigated because it was known that it should not only maintain the nuclear fuel integrity but also have the advantage in economical efficiency and safety. The pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model. After the detailed MCNP modeling of the whole facility, calculations of nuclear characteristics were performed. The two core configurations, Core 4.3 and Core 5 (reference state no. 3), among the 10 configurations of the HTR-PROTEUS cores were chosen to be analyzed in order to treat different fuel loading pattern and modeled. The former is a random packing core and the latter deterministic packing core. Based on the experimental data and the benchmark result of other research groups for the two different cores, some nuclear characteristics were calculated. Firstly, keff was calculated for these cores. The effect for TRIO homogeneity model was investigated. Control rod and shutdown rod worths also were calculated and the sensitivity analysis on cross-section library and reflector thickness was pursued. Lastly, neutron flux profiles were investigated in reflector regions. It is noted that Monte Carlo analysis of pebble-type VHTR core was firstly carried out in Korea. Also, this study should not only provide the basic data for pebble-type VHTR core analysis for hydrogen production but also be utilized as the verified data to validate a computer code for VHTR core analysis which will be developed in future

Drilling and coring methods that minimize the disturbance of cuttings, core, and rock formation in the unsaturated zone, Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Hammermeister, D.P.; Blout, D.O.; McDaniel, J.C.

1985-01-01

A drilling-and-casing method (Odex 115 system) utilizing air as a drilling fluid was used successfully to drill through various rock types within the unsaturated zone at Yucca Mountain, Nevada. This paper describes this method and the equipment used to rapidly penetrate bouldery alluvial-colluvial deposits, poorly consolidated bedded and nonwelded tuff, and fractured, densely welded tuff to depths of about 130 meters. A comparison of water-content and water-potential data from drill cuttings with similar measurements on rock cores indicates that drill cuttings were only slightly disturbed for several of the rock types penetrated. Coring, sampling, and handling methods were devised to obtain minimally disturbed drive core from bouldery alluvial-colluvial deposits. Bulk-density values obtained from bulk samples dug from nearby trenches were compared to bulk-density values obtained from drive core to determine the effects of drive coring on the porosity of the core. Rotary coring methods utilizing a triple-tube core barrel and air as the drilling fluid were used to obtain core from welded and nonwelded tuff. Results indicate that the disturbance of the water content of the core was minimal. Water-content distributions in alluvium-colluvium were determined before drilling occurred by drive-core methods. After drilling, water-content distributions were determined by nuclear-logging methods. A comparison of the water-content distributions made before and after drilling indicates that Odex 115 drilling minimally disturbs the water content of the formation rock. 10 refs., 12 figs., 4 tabs

Design comparisons of TRU burner cores with similar sodium void worth

International Nuclear Information System (INIS)

Sang Ji, Kim; Young Il, Kim; Young Jin, Kim; Nam Zin, Cho

2001-01-01

This study summarizes the neutronic performance and fuel cycle behavior of five geometrically-different transuranic (TRU) burner cores with similar low sodium void reactivity. The conceptual cores encompass core geometries for annular, two-region homogeneous, dual pin type, pan-shaped and H-shaped cores. They have been designed with the same assembly specifications and managed to have similar end-of-cycle sodium void reactivities and beginning-of-cycle peak power densities through the changes in the core size and configuration. The requirement of low sodium void reactivity is shown to lead each design concept to characteristic neutronics performance and fuel cycle behavior. The H-/pan-shaped cores allow the core compaction as well as higher rate of TRU burning. (author)

WWER-440 type reactor core

International Nuclear Information System (INIS)

Mizov, J.; Svec, P.; Rajci, T.

1987-01-01

Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

Simulation of the Long period Core Design for WH type of KHNP

Energy Technology Data Exchange (ETDEWEB)

Jung, Ji-Eun; Moon, Sang-Rae [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

2016-10-15

The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve.

Simulation of the Long period Core Design for WH type of KHNP

International Nuclear Information System (INIS)

Jung, Ji-Eun; Moon, Sang-Rae

2016-01-01

The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve

Development of core technology for research reactors using plate type fuels

International Nuclear Information System (INIS)

Ha, Jae Joo; Lee, Doo Jeong; Park, Cheol

2009-12-01

Around 250 research reactors are under operation over the world. However, about 2/3 have been operated more than 30 years and demands for replacements are expected in the near future. The number of expected units is around 110, and around 55 units from 40 countries will be expected to be bid in the world market. In 2007, Netherlands started international bidding process to construct a new 80MW RR (named PALLAS) with the target of commercial operation in 2016, which will replace the existing HFR(45MW). KAERI consortium has been participated in that bid. Most of RRs use plate type fuels as a fuel assembly, Be and Graphite as a reflector. On the other hand, in Korea, the KAERI is operating the HANARO, which uses a rod type fuel assembly and heavy water as a reflector. Hence, core technologies for RRs using plate type fuels are in short. Therefore, core technologies should be secured for exporting a RR. In chapter 2, the conceptual design of PALLAS which use plate type fuels are described including core, cooling system and connected systems, layout of general components. Experimental verification tests for the plate type fuel and second shutdown system and the code verification for nuclear design are explained in Chapter 3 and 4, respectively

Core conversion from rod to plate type fuel elements in research reactors

International Nuclear Information System (INIS)

Khattab, M.S.; Mina, A.R.

1997-01-01

Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

Solitons in a hard-core bosonic system: Gross–Pitaevskii type and ...

Indian Academy of Sciences (India)

2015-10-20

Oct 20, 2015 ... Solitons in a hard-core bosonic system: Gross–Pitaevskii type and beyond ... the corresponding class of magnetic solitons in Heisenberg spin chains with different types of anisotropy. ... Pramana – Journal of Physics | News.

Emergency core cooling system for LMFBR type reactors

International Nuclear Information System (INIS)

Tamano, Toyomi; Fukutomi, Shigeki.

1980-01-01

Purpose: To enable elimination of decay heat in an LMFBR type reactor by securing natural cycling force in any state and securing reactor core cooling capacity even when both an external power supply and an emergency power supply are failed in emergency case. Method: Heat insulating material portion for surrounding a descent tube of a steam drum provided at high position for obtaining necessary flow rate for flowing resistance is removed from heat transmitting surface of a recycling type steam generator to provide a heat sink. That is, when both an external power supply and an emergency power supply are failed in emergency, the heat insulator at part of a steam generator recycling loop is removed to produce natural cycling force between it and the heat transmitting portion of the steam generator as a heat source for the heat sink so as to secure the flow rate of the recycling loop. When the power supply is failed in emergency, the heat removing capacity of the steam generator is secured so as to remove the decay heat produced in the reactor core. (Yoshihara, H.)

Defining the Core Genome of Salmonella enterica Serovar Typhimurium for Genomic Surveillance and Epidemiological Typing

Science.gov (United States)

Fu, Songzhe; Octavia, Sophie; Tanaka, Mark M.; Sintchenko, Vitali

2015-01-01

Salmonella enterica serovar Typhimurium is the most common Salmonella serovar causing foodborne infections in Australia and many other countries. Twenty-one S. Typhimurium strains from Salmonella reference collection A (SARA) were analyzed using Illumina high-throughput genome sequencing. Single nucleotide polymorphisms (SNPs) in 21 SARA strains ranged from 46 to 11,916 SNPs, with an average of 1,577 SNPs per strain. Together with 47 strains selected from publicly available S. Typhimurium genomes, the S. Typhimurium core genes (STCG) were determined. The STCG consist of 3,846 genes, a set that is much larger than that of the 2,882 Salmonella core genes (SCG) found previously. The STCG together with 1,576 core intergenic regions (IGRs) were defined as the S. Typhimurium core genome. Using 93 S. Typhimurium genomes from 13 epidemiologically confirmed community outbreaks, we demonstrated that typing based on the S. Typhimurium core genome (STCG plus core IGRs) provides superior resolution and higher discriminatory power than that based on SCG for outbreak investigation and molecular epidemiology of S. Typhimurium. STCG and STCG plus core IGR typing achieved 100% separation of all outbreaks compared to that of SCG typing, which failed to separate isolates from two outbreaks from background isolates. Defining the S. Typhimurium core genome allows standardization of genes/regions to be used for high-resolution epidemiological typing and genomic surveillance of S. Typhimurium. PMID:26019201

Some results of Krsko NPP core calculations and comparison with measurements

International Nuclear Information System (INIS)

Trkov, A.; Zefran, B.; Kromar, M.; Ravnik, M.; Slavic, S.

1996-01-01

Current status of the CORD-2 package is described. Results of the predictions of some important reactor core parameters are presented for the 12 th operation cycle of the Krsko NPP. Comparison with measurements is made to illustrate that the accuracy of the calculations is acceptable. Some comments are made on the enhancements, which are currently being implemented on the package. (author)

Power distribution investigation in the transition phase of the low moderation type MOX fueled LWR from the high conversion core to the breeding core

International Nuclear Information System (INIS)

Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

2011-01-01

The key concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a core transition from a high conversion (HC) type to a plutonium breeding (BR) type in a same reactor system only by replacing fuel assemblies. Consequently in this transition phase, there are two types of assemblies in the same core. Due to the differences of the two assembly types, region-wise soft to hard neutron spectra appears and result in a large power peaking. Therefore, power distribution of FLWR in the HC to BR transition phase was studied by performing assembly and core calculations. For the whole core calculation, a new 14-group energy structure is developed to better represent the power distribution obtained with the fine 107-group structure than the 9-group structure in the previous evaluations. Calculations on few assemblies geometries show large local power peakings can be effectively reduced by considering plutonium enrichment distribution in an assembly. In the whole core calculation, there is a power level mismatch between HC and BR assemblies, but overall power distribution flattening is possible by optimizing fuel assemblies loading. Although the fuel loading should be decided also taking into account the void coefficient, transition from HC to BR type FLWR seems feasible without difficulty. (author)

A new type of In-core sensor validation outline

International Nuclear Information System (INIS)

Figedy, S.

2006-01-01

In this a new type of in-core sensor validation method is outlined, which is based on combination of correlation coefficients and mutual information indices. First experience with this approach is described and further improvements to enhance the outcome reliability are proposed namely through computational intelligence techniques (Authors)

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Tunable Band Gap and Conductivity Type of ZnSe/Si Core-Shell Nanowire Heterostructures

Directory of Open Access Journals (Sweden)

Yijie Zeng

2014-10-01

Full Text Available The electronic properties of zincblende ZnSe/Si core-shell nanowires (NWs with a diameter of 1.1–2.8 nm are calculated by means of the first principle calculation. Band gaps of both ZnSe-core/Si-shell and Si-core/ZnSe-shell NWs are much smaller than those of pure ZnSe or Si NWs. Band alignment analysis reveals that the small band gaps of ZnSe/Si core-shell NWs are caused by the interface state. Fixing the ZnSe core size and enlarging the Si shell would turn the NWs from intrinsic to p-type, then to metallic. However, Fixing the Si core and enlarging the ZnSe shell would not change the band gap significantly. The partial charge distribution diagram shows that the conduction band maximum (CBM is confined in Si, while the valence band maximum (VBM is mainly distributed around the interface. Our findings also show that the band gap and conductivity type of ZnSe/Si core-shell NWs can be tuned by the concentration and diameter of the core-shell material, respectively.

Vibration Finite Element Analysis of SC10 Dry-type Transformer Core

Directory of Open Access Journals (Sweden)

Gao Sheng Wei

2014-06-01

Full Text Available As the popularization and application of dry-type power transformer, its work when the vibration noise problem widely concerned, on the basis of time-varying electromagnetic field and structural mechanics equation, this paper established a finite element analysis model of dry-type transformer, through the electromagnetic field – Structural mechanics field – sound field more than physical field coupling calculation analysis, obtained in no load and the vibration modes of the core under different load and frequency. According to the transformer vibration mechanism, compared with the experimental data, verified the accuracy of the calculation results, as the core of how to provide the theory foundation and to reduce the noise of the experiment.

Aseismic study of high temperature gas-cooled reactor core with block-type fuel, 3

International Nuclear Information System (INIS)

Ikushima, Takeshi; Honma, Toshiaki.

1985-01-01

A two-dimensional horizontal seismic experiment with single axis and simultaneous two-axes excitations was performed to obtain the core seismic design data on the block-type high temperature gas-cooled reactor. Effects of excitation directions and core side support stiffness on characteristics of core displacements and reaction forces of support were revealed. The values of the side reaction forces are the largest in the excitation of flat-to-flat of hexagonal block. Preload from the core periphery to the core center are effective to decrease core displacements and side reaction forces. (author)

A comparison of database systems for XML-type data.

Science.gov (United States)

Risse, Judith E; Leunissen, Jack A M

2010-01-01

In the field of bioinformatics interchangeable data formats based on XML are widely used. XML-type data is also at the core of most web services. With the increasing amount of data stored in XML comes the need for storing and accessing the data. In this paper we analyse the suitability of different database systems for storing and querying large datasets in general and Medline in particular. All reviewed database systems perform well when tested with small to medium sized datasets, however when the full Medline dataset is queried a large variation in query times is observed. There is not one system that is vastly superior to the others in this comparison and, depending on the database size and the query requirements, different systems are most suitable. The best all-round solution is the Oracle 11~g database system using the new binary storage option. Alias-i's Lingpipe is a more lightweight, customizable and sufficiently fast solution. It does however require more initial configuration steps. For data with a changing XML structure Sedna and BaseX as native XML database systems or MySQL with an XML-type column are suitable.

Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

International Nuclear Information System (INIS)

Hussein, H.M.; Sakr, A.M.; Amin, E.H.

2011-01-01

Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

Neutronic Design of KALIMER-600 Core with Moderator Rods

International Nuclear Information System (INIS)

Ser Gi Hong; Sang Ji Kim; Hoon Song; Yeong Il Kim

2004-01-01

Recently, the liquid-metal reactor research team of the Korea Atomic Energy Research Institute (KAERI) designed a 600 MWe sodium-cooled, metallic fueled fast reactor meeting the goals of Generation-IV, such as economics and proliferation resistance. In this paper, the core design analysis and its performance are reported. The core is designed to have a conversion ratio slightly larger than unity with no blanket assemblies in order not to produce an excess amount of high grade plutonium and to have no need for external feeds of fissile materials. To mitigate the sodium void reactivity of the fuel-self-sufficient core with no blanket assemblies, several design changes from a reference core are tried; reduction of the active core height, annular type cores with central dummy assemblies, and the use of moderator (BeO or ZrH 2 ) rods. As a result of the analysis, it is found that of the considered designs the use of moderator rods for the softening of the core neutron spectrum is the best choice for reducing the sodium void worth with the smallest changes from the reference fuel and assembly designs. The core analysis shows that the sodium void reactivity is reduced by ∼2$ in comparison with the reference core and the core has a much more negative fuel temperature reactivity feedback in comparison with the reference core. (authors)

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

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

Nonlinear Model of Tape Wound Core Transformers

Directory of Open Access Journals (Sweden)

A. Vahedi

2015-03-01

Full Text Available Recently, tape wound cores due to their excellent magnetic properties, are widely used in different types of transformers. Performance prediction of these transformers needs an accurate model with ability to determine flux distribution within the core and magnetic loss. Spiral structure of tape wound cores affects the flux distribution and always cause complication of analysis. In this paper, a model based on reluctance networks method is presented for analysis of magnetic flux in wound cores. Using this model, distribution of longitudinal and transverse fluxes within the core can be determined. To consider the nonlinearity of the core, a dynamic hysteresis model is included in the presented model. Having flux density in different points of the core, magnetic losses can be calculated. To evaluate the validity of the model, results are compared with 2-D FEM simulations. In addition, a transformer designed for series-resonant converter and simulation results are compared with experimental measurements. Comparisons show accuracy of the model besides simplicity and fast convergence

Provision of reliable core cooling in vessel-type boiling reactors

International Nuclear Information System (INIS)

Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

1987-01-01

Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

Preliminary Core Design Analysis of a 200MWth Pebble Bed-type VHTR

International Nuclear Information System (INIS)

Jo, Chang Keun; Noh, Jae Man

2007-01-01

This paper intends to suggest the preliminary core design analysis of a VHTR for a hydrogen production. The nuclear hydrogen system that utilizes the high temperature heat generated from the VHTR is a promising candidate for a cost effective, safe and clean supply of hydrogen in the age of hydrogen economy. Among two candidate VHTR cores, that is, a prismatic modular reactor (PMR) and a pebble bed-type reactor (PBR), we focus on the design of a 200MWth PBR (hereinafter PBR200) in this paper. Here, the 200MWth power is selected for a demonstration plant. The core configuration of the PBR200 is similar to the PBMR (Pebble Bed Modular Reactor, 400MWth) of South Africa, but the overall dimension of the reactor system is scaled-down. This paper is to suggest two candidate PBR200 cores. One is an annular core with an inner reflector (PBR200-CD1) which was presented at IWRES07, and the other is a cylindrical core without an inner reflector (PBR200-CD2)

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

Science.gov (United States)

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

2017-12-01

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

Development of Mycoplasma synoviae (MS) core genome multilocus sequence typing (cgMLST) scheme.

Science.gov (United States)

Ghanem, Mostafa; El-Gazzar, Mohamed

2018-05-01

Mycoplasma synoviae (MS) is a poultry pathogen with reported increased prevalence and virulence in recent years. MS strain identification is essential for prevention, control efforts and epidemiological outbreak investigations. Multiple multilocus based sequence typing schemes have been developed for MS, yet the resolution of these schemes could be limited for outbreak investigation. The cost of whole genome sequencing became close to that of sequencing the seven MLST targets; however, there is no standardized method for typing MS strains based on whole genome sequences. In this paper, we propose a core genome multilocus sequence typing (cgMLST) scheme as a standardized and reproducible method for typing MS based whole genome sequences. A diverse set of 25 MS whole genome sequences were used to identify 302 core genome genes as cgMLST targets (35.5% of MS genome) and 44 whole genome sequences of MS isolates from six countries in four continents were used for typing applying this scheme. cgMLST based phylogenetic trees displayed a high degree of agreement with core genome SNP based analysis and available epidemiological information. cgMLST allowed evaluation of two conventional MLST schemes of MS. The high discriminatory power of cgMLST allowed differentiation between samples of the same conventional MLST type. cgMLST represents a standardized, accurate, highly discriminatory, and reproducible method for differentiation between MS isolates. Like conventional MLST, it provides stable and expandable nomenclature, allowing for comparing and sharing the typing results between different laboratories worldwide. Copyright © 2018 The Authors. Published by Elsevier B.V. All rights reserved.

Design studies of back up cores for the experimental multi-purpose VHTR, (1)

International Nuclear Information System (INIS)

Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu

1982-09-01

For the Experimental Multi-Purpose Very High Temperature Reactor, design studies have been made of two backup cores loaded with new type fuel elements. The purpose is to improve core operational characteristics of the standard design core (Mark-III core) consisting of pin-in-block type fuel element having externally cooled hollow fuel rods. The first backup core (semi-pin fuel core) is composed of fuel elements with internally cooled fuel pins, and the second core (multihole fuel core) is composed of multihole fuel elements, which can be adopted for the experimental VHTR as the substitution of the standard Mark-III fuel element. Either of the cores has 73 fuel columns and 4 m height. The arrangement of active core and reactor internal structure is same as that in the standard design core. These backup cores meet almost all design requirements of the VHTR and increase the margins for some important design items in comparison with the standard core (Mark-III core). This report describes the overall characteristics of nuclear, thermal-hydraulic, fuel and safety, and structural consideration for these cores. (author)

Analysis of fuel management pattern of research reactor core of the MTR type design

International Nuclear Information System (INIS)

Lily Suparlina; Tukiran Surbakti

2014-01-01

Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x10 15 n/cm 2 s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 10 15 n/cm 2 s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

[Influence of different types of posts and cores on color of IPS-Empress 2 crown].

Science.gov (United States)

Li, Dong-fang; Yang, Jing-yuan; Yang, Xing-mei; Yang, Liu; Xu, Qiang; Guan, Hong-yu; Wan, Qian-bing

2007-10-01

To evaluate the influence of different types of posts and cores on the final color of the IPS-Emperss 2 crown. Five types of posts and cores (Cerapost with Empress cosmo, Cerapost with composite resin, gilded Ni-Cr alloy, gold alloy and Ni-Cr alloy) were made. The shifts in color of three points of IPS-Empress 2 crown surface (cervical, middle and incisal) with different posts and cores was measured with a spectroradiometer (PR-650). The L* a* b* values of zirconium oxide and gilded Ni-Cr alloy posts and cores with ceramic crown were the highest. The L* a* values of zirconium oxide posts composite cores were higher while the b* values were lower. The L* a* b* values of Ni-Cr alloy were lower than that of gold alloy and were the lowest. In combination with IPS-Empress 2 crown, zirconium oxide posts are suitable for routine use in the anterior dentition, and gilded Ni-Cr alloy and gold alloy posts and cores can be recommended for clinical practice. Ni-Cr alloy posts and cores can not be recommended for clinical practice.

Comparison study on in-core neutron detector for online neutron flux mapping of research and power reactor

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Mohd Idris Taib; Izhar Abu Husin; Nurfarhana Ayuni

2010-01-01

This paper presents the comparison study on In-Core neutron detector using for online flux mapping of Research and Power reactor. Technical description of in-core neutron also taken into consideration to identify the different characterization of neutron detector and describe on Self Power neutron detector (SPND) for online neutron flux mapping. Able to provide information on the neutron flux distribution and understand how in-core neutron detector are being used in nuclear power plant including to enable to state the principles of neutron detector. (author)

Interfacial charge separation and recombination in InP and quasi-type II InP/CdS core/shell quantum dot-molecular acceptor complexes.

Science.gov (United States)

Wu, Kaifeng; Song, Nianhui; Liu, Zheng; Zhu, Haiming; Rodríguez-Córdoba, William; Lian, Tianquan

2013-08-15

Recent studies of group II-VI colloidal semiconductor heterostuctures, such as CdSe/CdS core/shell quantum dots (QDs) or dot-in-rod nanorods, show that type II and quasi-type II band alignment can facilitate electron transfer and slow down charge recombination in QD-molecular electron acceptor complexes. To explore the general applicability of this wave function engineering approach for controlling charge transfer properties, we investigate exciton relaxation and dissociation dynamics in InP (a group III-V semiconductor) and InP/CdS core/shell (a heterostructure beween group III-V and II-VI semiconductors) QDs by transient absorption spectroscopy. We show that InP/CdS QDs exhibit a quasi-type II band alignment with the 1S electron delocalized throughout the core and shell and the 1S hole confined in the InP core. In InP-methylviologen (MV(2+)) complexes, excitons in the QD can be dissociated by ultrafast electron transfer to MV(2+) from the 1S electron level (with an average time constant of 11.4 ps) as well as 1P and higher electron levels (with a time constant of 0.39 ps), which is followed by charge recombination to regenerate the complex in its ground state (with an average time constant of 47.1 ns). In comparison, InP/CdS-MV(2+) complexes show similar ultrafast charge separation and 5-fold slower charge recombination rates, consistent with the quasi-type II band alignment in these heterostructures. This result demonstrates that wave function engineering in nanoheterostructures of group III-V and II-VI semiconductors provides a promising approach for optimizing their light harvesting and charge separation for solar energy conversion applications.

Comparison of the Reactor Core Characteristics of the AHR and the OPAL

International Nuclear Information System (INIS)

Seo, Chul Gyo; Lee, Byung Chul; Park, C.; Chae, Hee Taek

2008-09-01

The AHR (Advanced HANARO research Reactor) was designed using the experiences from the design, operation and utilization of HANARO. Its neutronic performance was compared to that of the OPAL with a 20 MW power which started its operation recently in Australia. As the OPAL does not have any in-core irradiation hole, a modified core model of the AHR, in which an in-core irradiation hole was changed into a fuel channel, was used for the comparison. For a clean, unperturbed core condition with all fresh fuels in the core and no irradiation holes in the reflector region, the maximum thermal neutron flux (E n 14 n/cm 2 s and that in the OPAL reaches 3.96x10 14 n/cm 2 s in the reflector region. The maximum flux in the AHR is 10.3% higher than that in the OPAL. The thermal flux region above 4.0x10 14 n/cm 2 s is widely distributed in the reflector of the AHR, but is not observed at all in the reflector of the OPAL. The uranium loading of the AHR core is 45.7 kgU, which is 16.3% higher than the 39.3 kgU of OPAL. For a clean core state, the excess reactivity of the AHR is higher than that of the OPAL. The assembly-average discharge burnup in the OPAL is estimated to be 49.1%U-235 whereas that in the AHR is 62.4%U-235. The difference for the discharge burnup is significant. For the conditions with the same cycle length of 30 days, the number of fuel assemblies consumed in the AHR is only 3/4 that of the OPAL

Comparison of the Reactor Core Characteristics of the AHR and the OPAL

Energy Technology Data Exchange (ETDEWEB)

Seo, Chul Gyo; Lee, Byung Chul; Park, C.; Chae, Hee Taek

2008-09-15

The AHR (Advanced HANARO research Reactor) was designed using the experiences from the design, operation and utilization of HANARO. Its neutronic performance was compared to that of the OPAL with a 20 MW power which started its operation recently in Australia. As the OPAL does not have any in-core irradiation hole, a modified core model of the AHR, in which an in-core irradiation hole was changed into a fuel channel, was used for the comparison. For a clean, unperturbed core condition with all fresh fuels in the core and no irradiation holes in the reflector region, the maximum thermal neutron flux (E{sub n}<0.625 eV) in the AHR reaches 4.41x10{sup 14} n/cm{sup 2} s and that in the OPAL reaches 3.96x10{sup 14} n/cm{sup 2}s in the reflector region. The maximum flux in the AHR is 10.3% higher than that in the OPAL. The thermal flux region above 4.0x10{sup 14} n/cm{sup 2}s is widely distributed in the reflector of the AHR, but is not observed at all in the reflector of the OPAL. The uranium loading of the AHR core is 45.7 kgU, which is 16.3% higher than the 39.3 kgU of OPAL. For a clean core state, the excess reactivity of the AHR is higher than that of the OPAL. The assembly-average discharge burnup in the OPAL is estimated to be 49.1%U-235 whereas that in the AHR is 62.4%U-235. The difference for the discharge burnup is significant. For the conditions with the same cycle length of 30 days, the number of fuel assemblies consumed in the AHR is only 3/4 that of the OPAL.

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

Directory of Open Access Journals (Sweden)

Gmyrek Zbigniew

2017-12-01

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

Comparison of serpent and triton generated FEW group constants for APR1400 nuclear reactor core

International Nuclear Information System (INIS)

Elsawi, Mohamed A.; Alnoamani, Zainab

2015-01-01

The accuracy of full-core reactor power calculations using diffusion codes is strongly dependent on the quality of the homogenized cross sections and other few-group constants generated by lattice codes. For many years, deterministic lattice codes have been used to generate these constants using different techniques: the discrete ordinates, collision probability or the method of characteristics, just to name a few. These codes, however, show some limitations, for example, on complex geometries or near heavy absorbers as in modern pressurized water reactor (PWR) designs like the Korean Advanced Power Reactor 1400 (APR1400) core. The use of continuous-energy Monte Carlo (MC) codes to produce nuclear constants can be seen as an attractive option when dealing with fuel or reactor types that lie beyond the capabilities of conventional deterministic lattice transport codes. In this paper, the few-group constants generated by two of the state-of-the-art reactor physics codes, SERPENT and SCALE/TRITON, will be critically studied and their reliability for being used in subsequent diffusion calculations will be evaluated. SERPENT is a 3D, continuous-energy, Monte Carlo reactor physics code which has a built-in burn-up calculation capability. It has been developed at the Technical Research Center of Finland (VTT) since 2004. SCALE/TRITON, on the other hand, is a control module developed within the framework of SCALE package that enables performing deterministic 2-D transport calculations on nuclear reactor core lattices. The approach followed in this paper is as follows. First, the few-group nuclear constants for the APR1400 reactor core were generated using SERPENT (version 2.1.22) and NEWT (in SCALE version 6.1.2) codes. For both codes, the critical spectrum, calculated using the B1 method, was used as a weighting function. Second, 2-D diffusion calculations were performed using the US NRC core simulator PARCS employing the two few-group constant sets generated in the first

Guiding and amplification properties of rod-type photonic crystal fibers with sectioned core doping

Science.gov (United States)

Selleri, S.; Poli, F.; Passaro, D.; Cucinotta, A.; Lægsgaard, J.; Broeng, J.

2009-05-01

Rod-type photonic crystal fibers are large mode area double-cladding fibers with an outer diameter of few millimeters which can provide important advantages for high-power lasers and amplifiers. Numerical studies have recently demonstrated the guidance of higher-order modes in these fibers, which can worsen the output beam quality of lasers and amplifiers. In the present analysis a sectioned core doping has been proposed for Ybdoped rod-type photonic crystal fibers, with the aim to improve the higher-order mode suppression. A full-vector modal solver based on the finite element method has been applied to properly design the low refractive index ring in the fiber core, which can provide an increase of the differential overlap between the fundamental and the higher-order mode. Then, the gain competition among the guided modes along the Yb-doped rod-type fibers has been investigated with a spatial and spectral amplifier model. Simulation results have shown the effectiveness of the sectioned core doping in worsening the higher-order mode overlap on the doped area, thus providing an effective single-mode behavior of the Yb-doped rod-type photonic crystal fibers.

Study on natural convection in core barrel. Experimental and numerical results for band type spacer pads

International Nuclear Information System (INIS)

Hayashi, Kenji; Kawamata, Nobuhiro; Kamide, Hideki

2003-03-01

In a fast reactor an Inter-Wrapper Flow (IWF) is one of significant phenomena for decay heat removal under natural circulation condition, when a direct reactor auxiliary cooling system (DRACS) is adopted for decay heat removal system. Cold coolant provided by dipped heat exchangers (DHX) of DRACS can penetrate into the core barrel (region between the subassemblies) and it makes natural convection int he core barrel. Such IWF will depend on a spacer pad geometry of subassemblies. Water experiment, TRIF (Test Rig for Inter-wrapper Flow), was carried out for IWF in a reactor core. The test section modeled a 1/12th sector of the core and upper plenum of reactor vessel. Experimental parameters were the spacer pad geometry and flow path geometries connecting the upper plenum and core barrel. Numerical simulation using AQUA code was also performed to confirm applicability of a simulation method. An experimental series using a button type spacer pad had been carried out. Here a band type spacer pad was examined. Temperatures at subassembly wall were measured with parameter of the flow path geometries; one was a connection pipe between the upper plenum and core barrel and the other was flow hole in core former plates between the outermost subassemblies and the core barrel. It was found that these flow paths were effective to remove heat in the core in case of the band type spacer pad. A general purpose three dimensional analysis code, AQUA, was applied to the experimental analysis. Each subassembly and inter wrapper gap region were modeled by slab mesh geometry. Pressure loss coefficient at the pacer pad was set based on the geometry. The numerical simulation results were in good agreement with measured temperature profiles in the core. (author)

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

Science.gov (United States)

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

2018-05-01

Which post-and-core combination will best improve the performance of extensively damaged endodontically treated incisors without a ferrule is still unclear. The purpose of this in vitro study was to investigate the restoration of extensively damaged endodontically treated incisors without a ferrule using glass-ceramic crowns bonded to various composite resin foundation restorations and 2 types of posts. Sixty decoronated endodontically treated bovine incisors without a ferrule were divided into 4 groups and restored with various post-and-core foundation restorations. NfPfB=no-ferrule (Nf) with glass-fiber post (Pf) and bulk-fill resin foundation restoration (B); NfPfP=no-ferrule (Nf) with glass-fiber post (Pf) and dual-polymerized composite resin core foundation restoration (P); NfPt=no-ferrule (Nf) with titanium post (Pt) and resin core foundation restoration; and NfPtB=no-ferrule (Nf) with titanium post (Pt) and bulk-fill resin core foundation restoration (B). Two additional groups from previously published data from the same authors (FPf=2mm of ferrule (F) and glass-fiber post (Pf) and composite resin core foundation restoration; and NfPf=no-ferrule (Nf) with glass-fiber post (Pf) and composite resin core foundation restoration), which were tested concomitantly and using the same experimental arrangement, were included for comparison. All teeth were prepared to receive bonded glass-ceramic crowns luted with dual-polymerized resin cement and were subjected to accelerated fatigue testing under submerged conditions at room temperature. Cyclic isometric loading was applied to the incisal edge at an angle of 30 degrees with a frequency of 5 Hz, beginning with a load of 100 N (5000 cycles). A 100-N load increase was applied every 15000 cycles. The specimens were loaded until failure or to a maximum of 1000 N (140000 cycles). The 6 groups (4 groups from the present study and 2 groups from the previously published study) were compared using the Kaplan-Meier survival

Quantitative Comparison of Dense-Core Amyloid Plaque Accumulation in Amyloid-β Precursor Protein Transgenic Mice

Science.gov (United States)

Liu, Peng; Reichl, John H.; Rao, Eshaan R.; McNellis, Brittany M.; Huang, Eric S.; Hemmy, Laura S.; Forster, Colleen L.; Kuskowski, Michael A.; Borchelt, David R.; Vassar, Robert; Ashe, Karen H.; Zahs, Kathleen R.

2016-01-01

There exist several dozen lines of transgenic mice that express human amyloid-β precursor protein (AβPP) with Alzheimer’s disease (AD)-linked mutations. AβPP transgenic mouse lines differ in the types and amounts of Aβ that they generate and in their spatiotemporal patterns of expression of Aβ assemblies, providing a toolkit to study Aβ amyloidosis and the influence of Aβ aggregation on brain function. More complete quantitative descriptions of the types of Aβ assemblies present in transgenic mice and in humans during disease progression should add to our understanding of how Aβ toxicity in mice relates to the pathogenesis of AD. Here, we provide a direct quantitative comparison of amyloid plaque burdens and plaque sizes in four lines of AβPP transgenic mice. We measured the fraction of cortex and hippocampus occupied by dense-core plaques, visualized by staining with Thioflavin S, in mice from young adulthood through advanced age. We found that the plaque burdens among the transgenic lines varied by an order of magnitude: at 15 months of age, the oldest age studied, the median cortical plaque burden in 5XFAD mice was already ~4.5 times that of 21-month Tg2576 mice and ~15 times that of 21–24-month rTg9191 mice. Plaque-size distributions changed across the lifespan in a line- and region-dependent manner. We also compared the dense-core plaque burdens in the mice to those measured in a set of pathologically-confirmed AD cases from the Nun Study. Cortical plaque burdens in Tg2576, APPSwePS1ΔE9, and 5XFAD mice eventually far exceeded those measured in the human cohort. PMID:28059792

Quantitative Comparison of Dense-Core Amyloid Plaque Accumulation in Amyloid-β Protein Precursor Transgenic Mice.

Science.gov (United States)

Liu, Peng; Reichl, John H; Rao, Eshaan R; McNellis, Brittany M; Huang, Eric S; Hemmy, Laura S; Forster, Colleen L; Kuskowski, Michael A; Borchelt, David R; Vassar, Robert; Ashe, Karen H; Zahs, Kathleen R

2017-01-01

There exist several dozen lines of transgenic mice that express human amyloid-β protein precursor (AβPP) with Alzheimer's disease (AD)-linked mutations. AβPP transgenic mouse lines differ in the types and amounts of Aβ that they generate and in their spatiotemporal patterns of expression of Aβ assemblies, providing a toolkit to study Aβ amyloidosis and the influence of Aβ aggregation on brain function. More complete quantitative descriptions of the types of Aβ assemblies present in transgenic mice and in humans during disease progression should add to our understanding of how Aβ toxicity in mice relates to the pathogenesis of AD. Here, we provide a direct quantitative comparison of amyloid plaque burdens and plaque sizes in four lines of AβPP transgenic mice. We measured the fraction of cortex and hippocampus occupied by dense-core plaques, visualized by staining with Thioflavin S, in mice from young adulthood through advanced age. We found that the plaque burdens among the transgenic lines varied by an order of magnitude: at 15 months of age, the oldest age studied, the median cortical plaque burden in 5XFAD mice was already ∼4.5 times that of 21-month-old Tg2576 mice and ∼15 times that of 21-24-month-old rTg9191 mice. Plaque-size distributions changed across the lifespan in a line- and region-dependent manner. We also compared the dense-core plaque burdens in the mice to those measured in a set of pathologically-confirmed AD cases from the Nun Study. Cortical plaque burdens in Tg2576, APPSwePS1ΔE9, and 5XFAD mice eventually far exceeded those measured in the human cohort.

Fission rate measurements in fuel plate type assembly reactor cores

International Nuclear Information System (INIS)

Rogers, J.W.

1988-01-01

The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

International Nuclear Information System (INIS)

Zheng, Meiyin; Tian, Wenxi; Zhang, Dalin; Qiu, Suizheng; Su, Guanghui

2015-01-01

Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

International Nuclear Information System (INIS)

Bandini, G.; Caira, M.; Naviglio, A.; Sorabella, L.

1995-01-01

The MARS nuclear plant is equipped with a 600 MWth PWR type nuclear steam supply system, with completely innovative engineered core safeguards. The most relevant innovative safety system of this plant is its Emergency Core Cooling System, which is completely passive (with only one non static component). The Emergency Core Cooling System (ECCS) of the MARS reactor is natural-circulation, passive-type, and its intervention follows a core flow decrease, whatever was the cause. The operation of the system is based on a cascade of three fluid systems, functionally interfacing through heat exchangers; the first fluid system is connected to the reactor vessel and the last one includes an atmospheric-pressure condenser, cooled by external air. The infinite thermal capacity of the final heat sink provides the system an unlimited autonomy. The capability and operability of the system are based on its integrity and on the integrity of the primary coolant boundary (both of them are permanently enclosed in a pressurized containment; 100% redundancy is also foreseen) and on the operation of only one non static component (a check valve), with 400% redundancy. In the paper, all main thermal hydraulic transients occurring as a consequence of postulated accidents are analysed, to verify the capability of the passive-type ECCS to intervene always in time, without causing undue conditions of reduced coolability of the core (DNB, etc.), and to verify its capability to guarantee a long-term (indefinite) coolability of the core without the need of any external intervention. (author)

Aspects of unconventional cores for large sodium cooled power reactors; evaluation of a literature survey

International Nuclear Information System (INIS)

Kiefhaber, E.

1978-10-01

The report gives an overview of a literature study on the application of unconventional cores for sodium cooled fast reactors. Different types of unconventional cores (heterogeneous cores, pancake cores, moderated cores and others) are compared with conventional cores, which are characterized by a cylindrical geometry with two or three fissile zones surrounded by an axial and a radial blanket. The main parameters of interest in this comparison are the neutronic parameters sodium void and Doppler effect, the breeding properties and the steel damage. Consequences for the core safety and the overall plant design are also mentioned

Post-Vacuum-Assisted Stereotactic Core Biopsy Clip Displacement: A Comparison Between Commercially Available Clips and Surgical Clip.

Science.gov (United States)

Yen, Peggy; Dumas, Sandra; Albert, Arianne; Gordon, Paula

2018-02-01

The placement of localization clips following percutaneous biopsy is a standard practice for a variety of situations. Subsequent clip displacement creates challenges for imaging surveillance and surgical planning, and may cause confusion amongst radiologists and between surgeons and radiologists. Many causes have been attributed for this phenomenon including the commonly accepted "accordion effect." Herein, we investigate the performance of a low cost surgical clip system against 4 commercially available clips. We retrospectively reviewed 2112 patients who underwent stereotactic vacuum-assisted core biopsy followed by clip placement between January 2013 and June 2016. The primary performance parameter compared was displacement >10 mm following vacuum-assisted stereotactic core biopsy. Within the group of clips that had displaced, the magnitude of displacement was compared. There was a significant difference in displacement among the clip types (P < .0001) with significant pairwise comparisons between pediatric surgical clips and SecureMark (38% vs 28%; P = .001) and SenoMark (38% vs 27%; P = .0001) in the proportion displaced. The surgical clips showed a significant magnitude of displacement of approximately 25% greater average distance displaced. As a whole, the commercial clips performed better than the surgical clip after stereotactic vacuum-assisted core biopsy suggesting the surrounding outer component acts to anchor the central clip and minimizes clip displacement. The same should apply to tomosynthesis-guided biopsy. Crown Copyright © 2017. Published by Elsevier Inc. All rights reserved.

Comparative analysis of core genome MLST and SNP typing within a European Salmonella serovar Enteritidis outbreak.

Science.gov (United States)

Pearce, Madison E; Alikhan, Nabil-Fareed; Dallman, Timothy J; Zhou, Zhemin; Grant, Kathie; Maiden, Martin C J

2018-06-02

Multi-country outbreaks of foodborne bacterial disease present challenges in their detection, tracking, and notification. As food is increasingly distributed across borders, such outbreaks are becoming more common. This increases the need for high-resolution, accessible, and replicable isolate typing schemes. Here we evaluate a core genome multilocus typing (cgMLST) scheme for the high-resolution reproducible typing of Salmonella enterica (S. enterica) isolates, by its application to a large European outbreak of S. enterica serovar Enteritidis. This outbreak had been extensively characterised using single nucleotide polymorphism (SNP)-based approaches. The cgMLST analysis was congruent with the original SNP-based analysis, the epidemiological data, and whole genome MLST (wgMLST) analysis. Combination of the cgMLST and epidemiological data confirmed that the genetic diversity among the isolates predated the outbreak, and was likely present at the infection source. There was consequently no link between country of isolation and genetic diversity, but the cgMLST clusters were congruent with date of isolation. Furthermore, comparison with publicly available Enteritidis isolate data demonstrated that the cgMLST scheme presented is highly scalable, enabling outbreaks to be contextualised within the Salmonella genus. The cgMLST scheme is therefore shown to be a standardised and scalable typing method, which allows Salmonella outbreaks to be analysed and compared across laboratories and jurisdictions. Copyright © 2018. Published by Elsevier B.V.

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

One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

Composition and strain effects in Type I and Type II heterostructure ZnSe/Cd(Zn)S and ZnSe/Cd1-xZnxS core/shell quantum dots

Science.gov (United States)

Gheshlaghi, Negar; Pisheh, Hadi Sedaghat; Ünlü, Hilmi

2017-11-01

We investigated the effect of ternary shell alloy composition on the bandgap and diameter of core of ZnSe / Cd1 - x Znx S heterostructure core/shell quantum dots, which were synthesized by using a simple colloidal technique. Characterization by using the x-ray diffraction (XRD), transmission electron microscopy (TEM), UV-Vis absorption and fluorescence emission spectroscopic techniques indicate that (i) there is a transition of ZnSe / Cd0.6 Zn0.4 S Type-I heterostructure (electrons and holes tend to localize in core) to ZnSe / Cd0.75 Zn0.25 S quasi-Type-II heterostructures (holes tend to localized in the core and electrons are delocalized) and (ii) then after large red shift and Stock-shift in PL emission spectra but not a distinct absorption peak in UV spectra become noticeable in ZnSe/Cd0.75Zn0.25 S quasi-Type II and ZnSe/CdS Type II heterostructures (electrons are localized in core and holes are localized in shell). Furthermore, the increase of Cd:S ratio in shell alloy composition shifts the XRD peaks to lower 2θ degrees which corresponds to tensile strain in the ZnSe core. Finally, the hydrostatic interfacial strain has effect on the squeezing or stretching the capped core: A decrease of compressive force on core from ZnSe/ZnS to tensile force in ZnSe/CdS with increase in Cd:S ratio indicates that transition of compressive strain to tensile strain takes place with the transition from Type-I to II heterostructure.

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.

Capital cost evaluation of liquid metal reactor by plant type - comparison of modular type with monolithic type -

International Nuclear Information System (INIS)

Mun, K. H.; Seok, S. D.; Song, K. D.; Kim, I. C.

1999-01-01

A preliminary economic comparison study was performed for KALIMER(Korea Advanced LIquid MEtal Reactor)between a modular plant type with 8 150MWe modules and a 1200MWe monolithic plant type. In both cases of FOAK (First-Of-A-Kind) Plant and NOAK (Nth-Of-A-Kind) Plant, the result says that the economics of monolithic plant is superior to its modular plant. In case of NOAK plant comparison, however, the cost difference is not significant. It means that modular plant can compete with monolithic plant in capital cost if it makes efforts of cost reduction and technical progress on the assumption that the same type of NOAK plant will be constructed continuously

Reactor core T-H characteristics determination in case of parallel operation of different fuel assembly types

International Nuclear Information System (INIS)

Hermansky, J.; Petenyi, V.; Zavodsky, M.

2009-01-01

The WWER-440 nuclear fuel vendor permanently improve the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. Therefore it is necessary to have the skilled methodology and computing code for analyzing factors which affecting the accuracy of flow redistributed determination through reactor on flows through separate parts of reactor core in case of parallel operation different assembly types. Whereas the geometric parameters of new manufactured assemblies were changed recently, the calculated flows through the fuel parts of different type of assemblies are depended also on their real position in reactor core. Therefore the computing code CORFLO was developed in VUJE Trnava for carrying out stationary analyses of T-H characteristics of reactor core within 60 deg symmetry. The CORFLO code deals the area of the active core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is calculated. Computing code is verified and validated at this time. Paper presents the short description of computing code CORFLO with some calculated results. (Authors)

Explaining Communication Displacement and Large-Scale Social Change in Core Networks: A Cross-National Comparison of Why Bigger is Not Better and Less Can Mean More

DEFF Research Database (Denmark)

Hampton, Keith; Ling, Richard

2013-01-01

The size and diversity of Americans’ core social networks has declined. Some suggest that the replacement of face-to-face contact with new media, and combined with more insular core networks is detrimental to both individual and societal well-being. Based on a cross-national comparison of the Uni......The size and diversity of Americans’ core social networks has declined. Some suggest that the replacement of face-to-face contact with new media, and combined with more insular core networks is detrimental to both individual and societal well-being. Based on a cross-national comparison...

EFFECTS OF FOSSIL MAGNETIC FIELDS ON CONVECTIVE CORE DYNAMOS IN A-TYPE STARS

International Nuclear Information System (INIS)

Featherstone, Nicholas A.; Toomre, Juri; Browning, Matthew K.; Brun, Allan Sacha

2009-01-01

The vigorous magnetic dynamo action achieved within the convective cores of A-type stars may be influenced by fossil magnetic fields within their radiative envelopes. We study such effects through three-dimensional simulations that model the inner 30% by radius of a 2 M sun A-type star, capturing the convective core and a portion of the overlying radiative envelope within our computational domain. We employ the three-dimensional anelastic spherical harmonic code to model turbulent dynamics within a deep rotating spherical shell. The interaction between a fossil field and the core dynamo is examined by introducing a large-scale magnetic field into the radiative envelope of a mature A star dynamo simulation. We find that the inclusion of a twisted toroidal fossil field can lead to a remarkable transition in the core dynamo behavior. Namely, a super-equipartition state can be realized in which the magnetic energy built by dynamo action is 10-fold greater than the kinetic energy of the convection itself. Such strong-field states may suggest that the resulting Lorentz forces should seek to quench the flows, yet we have achieved super-equipartition dynamo action that persists for multiple diffusion times. This is achieved by the relative co-alignment of the flows and magnetic fields in much of the domain, along with some lateral displacements of the fastest flows from the strongest fields. Convection in the presence of such strong magnetic fields typically manifests as 4-6 cylindrical rolls aligned with the rotation axis, each possessing central axial flows that imbue the rolls with a helical nature. The roll system also possesses core-crossing flows that couple distant regions of the core. We find that the magnetic fields exhibit a comparable global topology with broad, continuous swathes of magnetic field linking opposite sides of the convective core. We have explored several poloidal and toroidal fossil field geometries, finding that a poloidal component is essential

Phosphine-free synthesis and characterization of type-II ZnSe/CdS core-shell quantum dots

Science.gov (United States)

Ghasemzadeh, Roghayyeh; Armanmehr, Mohammad Hasan; Abedi, Mohammad; Fateh, Davood Sadeghi; Bahreini, Zaker

2018-01-01

A phosphine-free route for synthesis of type-II ZnSe/CdS core-shell quantum dots, using green, low cost and environmentally friendly reagents and phosphine-free solvents such as 1-octadecene (ODE) and liquid paraffin has been reported. Hot-injection technique has been used for the synthesis of ZnSe core quantum dots. The CdS shell quantum dots prepared by reaction of CdO precursor and S powder in 1-octadecene (ODE). The ZnSe/CdS core-shell quantum dots were synthesized via successive ion layer adsorption and reaction (SILAR) technique. The characterization of produced quantum dots were performed by absorption and fluorescence spectroscopy, X-ray diffraction (XRD) and energy dispersive X-ray spectroscopy (EDX) and transmission electron microscopy (TEM). The results showed the formation of type-II ZnSe/CdS core-shell quantum dots with FWHM 32 nm and uniform size distribution.

Simulation of single-phase rod bundle flow. Comparison between CFD-code ESTET, PWR core code THYC and experimental results

International Nuclear Information System (INIS)

Mur, J.; Larrauri, D.

1998-07-01

Computer simulation of flow in configurations close to pressurized water reactor (PWR) geometry is of great interest for Electricite de France (EDF). Although simulation of the flow through a whole PWR core with an all purpose CFD-code is not yet achievable, such a tool cna be quite useful to perform numerical experiments in order to try and improve the modeling introduced in computer codes devoted to reactor core thermal-hydraulic analysis. Further to simulation in small bare rod bundle configurations, the present study is focused on the simulation, with CFD-code ESTET and PWR core code THYC, of the flow in the experimental configuration VATICAN-1. ESTET simulation results are compared on the one hand to local velocity and concentration measurements, on the other hand with subchannel averaged values calculated by THYC. As far as the comparison with measurements is concerned, ESTET results are quite satisfactory relatively to available experimental data and their uncertainties. The effect of spacer grids and the prediction of the evolution of an unbalanced velocity profile seem to be correctly treated. As far as the comparison with THYC subchannel averaged values is concerned, the difficulty of a direct comparison between subchannel averaged and local values is pointed out. ESTET calculated local values are close to experimental local values. ESTET subchannel averaged values are also close to THYC calculation results. Thus, THYC results are satisfactory whereas their direct comparison to local measurements could show some disagreement. (author)

Activity of corrosion products in pool type reactors with ascending flow in the core

International Nuclear Information System (INIS)

Andrade e Silva, Graciete S. de; Queiroz Bogado Leite, Sergio de

1995-01-01

A model for the activity of corrosion products in the water of a pool type reactor with ascending flow is presented. The problem is described by a set of coupled differential equations relating the radioisotope concentrations in the core and pool circuits and taking into account two types of radioactive sources: i) those from radioactive species formed in the fuel cladding, control elements, reflector, etc, and afterwards released to the primary stream by corrosion (named reactor sources) and ii) those formed from non radioactive isotopes entering the primary stream by corrosion of the circuit components and being activated when passing through the core (named circuit sources). (author). 6 refs, 3 figs, 4 tabs

Polyaniline Coated Core-Shell Typed Stimuli-Responsive Microspheres and Their Electrorheology

Directory of Open Access Journals (Sweden)

Yu Zhen Dong

2018-03-01

Full Text Available Functional core-shell-structured particles have attracted considerable attention recently. This paper reviews the synthetic methods and morphologies of various electro-stimuli responsive polyaniline (PANI-coated core-shell-type microspheres, including PANI-coated Fe3O4, SiO2, Fe2O3, TiO2, poly(methyl methacrylate, poly(glycidyl methacrylate, and polystyrene along with their electrorheological (ER characteristics when prepared by dispersing these particles in an insulating medium. In addition to the various rheological characteristics and their analysis, such as shear stress and yield stress of their ER fluids, this paper summarizes some of the mechanisms proposed for ER fluids to further understand the responses of ER fluids to an externally applied electric field.

Fuel and core design study of the sodium-cooled fast reactors. Studies on metallic fuel cores in the JFY2002

International Nuclear Information System (INIS)

Sugino, Kazuteru; Mizuno, Tomoyasu

2003-06-01

Based on the results obtained in the former feasibility study, the metallic fueled core of ordinary-type, that is, 2-region homogeneous core, has been established aiming at the improvement in the core performance, and subsequent comparison has been performed with the mixed oxide fueled core. Further, the attractive concept of the metallic fueled core of high outlet temperature has been constructed which has good nuclear features as a metallic fueled core and has identical outlet temperature to mixed oxide fuelled core. Following items have been found as a result of the investigation on the ordinary-type core. The metallic fueled core whose maximum fast neutron fluence (En>0.1MeV) is set identical (5x10 23 n/cm 2 ) to the mixed oxide fueled cores with core discharge burnup 150GWd/t has sufficient core performances as a metallic fueled core, e.g. higher breeding ratio and longer operation period compared with mixed oxide fueled cores, but the core discharge burnup is limited up to 100GWd/t. However effective discharge burnup including the contribution of the blanket region is comparative to mixed oxide cores under the same breeding ratio condition. In order to enlarge the core discharge burnup to 150GWd/t keeping the core performance identical to above mentioned core's, the irradiation deformation of structural material should be reduced to that of mixed oxide fueled cores. Further the maximum fast neutron fluence reaches to 7-8x10 23 n/cm 2 (En>0.1MeV). The investigations on the core of high outlet temperature have clarified following items. Even in the change of core regions by pin-diameter form 3-region to 2-region and in the limited maximum fuel pin diameter 8.5 mm, realization of the identical outlet/inlet temperatures to the mixed oxide cores (550/395degC) is feasible under the criteria of the maximum temperature 650degC at the inner surface of the cladding. The constructed core accommodates the targets of breeding ratio from about 1.0 to 1.2 only by adjusting

Tunable luminescent emission characterization of type-I and type-II systems in CdS-ZnSe core-shell nanoparticles: Raman and photoluminescence study.

Science.gov (United States)

Ca, Nguyen Xuan; Lien, V T K; Nghia, N X; Chi, T T K; Phan, The-Long

2015-11-06

We used wet chemical methods to synthesize core-shell nanocrystalline samples CdS(d)/ZnSe N , where d = 3-6 nm and N = 1-5 are the size of CdS cores and the number of monolayers grown on the cores, respectively. By annealing typical CdS(d)/ZnSe N samples (with d = 3 and 6 nm and N = 2) at 300 °C for various times t an = 10-600 min, we created an intermediate layer composed of Zn1-x Cd x Se and Cd1-x Zn x S alloys with various thicknesses. The formation of core-shell structures and intermediate layers was monitored by Raman scattering and UV-vis absorption spectrometers. Careful photoluminescence studies revealed that the as-prepared CdS(d)/ZnSe N samples with d = 5 nm and N = 2-4, and the annealed samples CdS(3 nm)/ZnSe2 with t an ≤ 60 min and CdS(6 nm)/ZnSe2 with t an ≤ 180 min, show the emission characteristics of type-II systems. Meanwhile, the other samples show the emission characteristics of type-I systems. These results prove that the partial separation of photoexcited carriers between the core and shell is dependent strongly on the engineered core-shell nanostructures, meaning the sizes of the core, shell, and intermediate layers. With the tunable luminescence properties, CdS-ZnSe-based core-shell materials are considered as promising candidates for multiple-exciton generation and single-photon sources.

Feasibility study on thermal-hydraulic design of reduced-moderation PWR-type core

International Nuclear Information System (INIS)

Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

2000-03-01

At JAERI, a conceptual study on reduced-moderation water reactor (RMWR) has been performed as one of the advanced reactor system which is designed so as to realize the conversion ratio more than unity. In this reactor concept, the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated. Therefore, an evaluation of the core thermal margin becomes very important in the design of the RMWR. In this study, we have performed a feasibility evaluation on thermal-hydraulic design of RM-PWR type core (core thermal output: 2900 MWt, Rod gaps: 1 mm). In RM-PWR core, seed and blanket regions are exist. In the blanket region, power density is lower than that of the seed region. Then, evaluation was performed under setting a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because it is possible that the coolant boils in the seed region. In the feasibility evaluations, subchannel code COBRA-IV-I was used in combination with KfK DNB (departure nucleate boiling) correlation. When coolant mass flow rate to the blanket fuel assembly is reduced by 40%, and that to the seed fuel assembly is increased, coolant boiling is not occurred in the assembly region calculation. Provided that the channel boxes to the blanket fuel assembly are set up and coolant mass flow rate to the blanket fuel assembly is reduced by 40%, it is confirmed by the whole core calculation that the boiling of the coolant is not occurred and the RM-PWR core is feasible. (author)

Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

International Nuclear Information System (INIS)

Ehud Greenspan

2008-01-01

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

Xenon oscillation in a large PHWR core (Atucha II type): TRISIC code applicability

International Nuclear Information System (INIS)

Solanilla, Roberto

2000-01-01

A three dimensional nuclear reactor simulation code (TRISIC) was developed many years ago to design a PHWR (pressurizer heavy water reactors - Atucha type) based in the 'source-sink model' (heterogeneous theory). The limited processor computational performance available at that time was the constraint of the code when a detailed reactor description was necessary. A modern PC (pentium) code version with a full reactor core representation (461 fuel channels) including diagonal control rod banks and flux-reading detectors with theirs tube guide was used in the present paper for simulation of the Xenon transient when a local asymmetric perturbation was produced in a large core (Atucha II type). The results obtained and the computer time required for the 70 hour's simulation with an adequate time step, established the potential of the code to deal with this kind of transients. The paper shows that the method of TRISIC allows to detect and control azimuthal, radial and axial oscillation. This code is a proper way to elaborate a program of control rods movement from the flux reading detectors to damp the oscillation. TRISIC could also be a accurate tool to supervise the full core flux distribution in real time during the operation of the reactor. (author)

Ge/Si core/multi shell heterostructure FETs

Energy Technology Data Exchange (ETDEWEB)

Picraux, Samuel T [Los Alamos National Laboratory; Dayeh, Shadi A [Los Alamos National Laboratory

2010-01-01

Concentric heterostructured materials provide numerous design opportunities for engineering strain and interfaces, as well as tailoring energy band-edge combinations for optimal device performance. Key to the realization of such novel device concepts is the complete understanding and full control over their growth, crystal structure, and hetero-epitaxy. We report here on a new route for synthesizing Ge/Si core/multi-shell heterostructure nanowires that eliminate Au seed diffusion on the nanowire sidewalls by engineering the interface energy density difference. We show that such control over core/shell synthesis enable experimental realization of heterostructure FET devices beyond those available in the literature with enhanced transport characteristics. We provide a side-by-side comparison on the transport properties of Ge/Si core/multi-shell nanowires grown with and without Au diffusion and demonstrate heterostructure FETs with drive currents that are {approx} 2X higher than record results for p-type FETs.

Development of flow network analysis code for block type VHTR core by linear theory method

International Nuclear Information System (INIS)

Lee, J. H.; Yoon, S. J.; Park, J. W.; Park, G. C.

2012-01-01

VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. It requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)

Galois Corings and a Jacobson-Bourbaki type Correspondence

OpenAIRE

Cuadra, J.; Gomez-Torrecillas, J.

2005-01-01

The Jacobson-Bourbaki Theorem for division rings was formulated in terms of corings by Sweedler in 1975. Finiteness conditions hypotheses are not required in this new approach. In this paper we extend Sweedler's result to simple artinian rings using a particular class of corings, comatrix corings. A Jacobson-Bourbaki like correspondence for simple artinian rings is then obtained by duality.

Optimized core design and fuel management of a pebble-bed type nuclear reactor

NARCIS (Netherlands)

Boer, B.

2009-01-01

The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the

Alternating current dielectrophoresis of core-shell nanoparticles: Experiments and comparison with theory

Science.gov (United States)

Yang, Chungja

Nanoparticles are fascinating where physical and optical properties are related to size. Highly controllable synthesis methods and nanoparticle assembly are essential for highly innovative technological applications. Well-defined shaped and sized nanoparticles enable comparisons between experiments, theory and subsequent new models to explain experimentally observed phenomena. Among nanoparticles, nonhomogeneous core-shell nanoparticles (CSnp) have new properties that arise when varying the relative dimensions of the core and the shell. This CSnp structure enables various optical resonances, and engineered energy barriers, in addition to the high charge to surface ratio. Assembly of homogeneous nanoparticles into functional structures has become ubiquitous in biosensors (i.e. optical labeling), nanocoatings, and electrical circuits. Limited nonhomogenous nanoparticle assembly has only been explored. Many conventional nanoparticle assembly methods exist, but this work explores dielectrophoresis (DEP) as a new method. DEP is particle polarization via non-uniform electric fields while suspended in conductive fluids. Most prior DEP efforts involve microscale particles. Prior work on core-shell nanoparticle assemblies and separately, nanoparticle characterizations with dielectrophoresis and electrorotation, did not systematically explore particle size, dielectric properties (permittivity and electrical conductivity), shell thickness, particle concentration, medium conductivity, and frequency. This work is the first, to the best of our knowledge, to systematically examine these dielectrophoretic properties for core-shell nanoparticles. Further, we conduct a parametric fitting to traditional core-shell models. These biocompatible core-shell nanoparticles were studied to fill a knowledge gap in the DEP field. Experimental results (chapter 5) first examine medium conductivity, size and shell material dependencies of dielectrophoretic behaviors of spherical CSnp into 2D and

Design study on PWR-type reduced-moderation light water core. Investigation of core adopting seed-blanket fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu; Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

As a part of the design study on PWR-type Reduced-Moderation Water Reactors (RMWRs), a light water cooled core with the seed-blanket type fuel assemblies has been investigated. An assembly with seed of 13 layers and blanket of 5 layers was selected by optimization calculations. The core was composed with the 163 assemblies. The following results were obtained by burn-up calculations with the MVP-BURN code; The cycle length is 15 months by 3-batch refueling. The discharge burn-up including the inner blanket is about 25 GWd/t. The conversion ratio is about 1.0. The void reactivity coefficient is about-26.1 pcm/%void at BOC and -21.7pcm%void at EOC. About 10% of MA makes conversion ratio decrease about 0.05 to obtain the same burn-up. The void reactivity coefficient increased significantly and it is necessary to reduce it. FP amount corresponding to about 2 % of total plutonium weight makes reactivity decrease about 0.5 %{delta}k/k and void reactivity coefficient increase, however these changes are within the design margins. Capability of multi-recycling of plutonium was confirmed, using discharged plutonium for 4 cycles, if fissile plutonium of 15.5wt% is used. The conversion ratio increases by about 0.026 with recycling. However, void reactivity coefficient increases and some effort to obtain negative void reactivity coefficient is necessary. (author)

Fabrication of solenoid-type inductor with electroplated NiFe magnetic core

International Nuclear Information System (INIS)

Gao Xiaoyu; Cao Ying; Zhou Yong; Ding Wen; Lei Chong; Chen Jian

2006-01-01

Solenoid-type inductor with ultra-low profile was fabricated by MEMS (Microelectromechanical systems) technique. NiFe film was electroplated as the magnetic core, and polyimide with a low relative permittivity was used as the insulation material. In the fabrication process, UV-LIGA, dry etching, fine polishing and electroplating technique have been adopted to achieve high performance of the solenoid-type inductor. The inductor was in size of 1.5 mmx0.9 mmx0.1 mm with coil width of 20 μm and aspect ratio of 5:1. The inductance and the quality factor were 0.42-0.345 μH and 1.8-5.3 in the frequency range of 1-10 MHz, respectively

Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

Energy Technology Data Exchange (ETDEWEB)

El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

2017-11-15

In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

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)

Comparison of design margin for core shroud in between design and construction code and fitness-for-service code

International Nuclear Information System (INIS)

Dozaki, Koji

2007-01-01

Structural design methods for core shroud of BWR are specified in JSME Design and Construction Code, like ASME Boiler and Pressure Vessel Code Sec. III, as a part of core support structure. Design margins are defined according to combination of the structural design method selected and service limit considered. Basically, those margins in JSME Code were determined after ASME Sec. III. Designers can select so-called twice-slope method for core shroud design among those design methods. On the other hand, flaw evaluation rules have been established for core shroud in JSME Fitness-for-Service Code. Twice-slope method is also adopted for fracture evaluation in that code even when the core shroud contains a flaw. Design margin was determined as structural factors separately from Design and Construction Code. As a natural consequence, there is a difference in those design margins between the two codes. In this paper, it is shown that the design margin in Fitness-for-Service Code is conservative by experimental evidences. Comparison of design margins between the two codes is discussed. (author)

Somatization as a core symptom of melancholic type depression. Evidence from a cross-cultural study.

Science.gov (United States)

Ebert, D; Martus, P

1994-12-01

The study questions whether different types of somatization may be a core symptom of melancholia, thus, being invariable across cultures and being a candidate for neurobiological research and diagnostic criteria. 51 Turkish patients and 51 education-matched German patients with melancholic depression were compared for two types of somatization. Turkish patients had higher frequencies of somatic preoccupation and hypochondriasis but they were not different in the perception and experience of somatic symptoms. It is concluded that: (1) somatization has to be differentiated psychopathologically; (2) it may be a neurobiological core symptom of melancholia in the well-defined sense of 'perceiving abnormal somatic symptoms'; and (3) it may be a culture-bound symptom in the sense of 'being abnormally concerned with somatic symptoms or hypochondrial fears'.

Two-dimensional full-core transport theory Benchmarks for the WWER reactors

International Nuclear Information System (INIS)

Petkov, P.T.

2002-01-01

Several two-dimensional full-core real geometry many-group steady-state problems for the WWER-440 and WWER-1000 reactors have been solved by the MARIKO code, based on the method of characteristics. The reference transport theory solutions include assembly-wise and pin-wise power distributions. Homogenized two-group diffusion parameters and discontinuity factors have been calculated by MARIKO for each assembly type both for the whole assembly and for each cell in the smallest sector of symmetry, using the B1 method for calculation of the critical spectrum. Accurate albedo-type boundary conditions have been calculated by MARIKO for the core-reflector and core-absorber boundaries, both for each outer assembly face and for each outer cell face. Comparison with the reference solutions of the two-group nodal diffusion code SPPS-1.6 and the few-group fine-mesh diffusion codes HEX2DA and HEX2DB are presented (Authors)

Fabrication of a solenoid-type inductor with Fe-based soft magnetic core

International Nuclear Information System (INIS)

Lei Chong; Zhou Yong; Gao Xiaoyu; Ding Wen; Cao Ying; Choi, Hyung; Won, Jonghwa

2007-01-01

A solenoid-type inductor was fabricated by MEMS (Microelectromechanical systems) technique. The fabrication process uses UV-LIGA, dry etching, fine polishing, and electroplating technique to achieve high performance of the solenoid-type inductor. Fe-based soft magnetic thin film was sputtered as the magnetic core, and polyimide was used as the insulation materials. The inductor was in size of 4x4 mm with coil width of 20 μm and space of 35 μm. The inductance is 1.61 μH at a frequency of 5 MHz with the maximum quality factor of 1.42

k-core covers and the core

NARCIS (Netherlands)

Sanchez-Rodriguez, E.; Borm, Peter; Estevez-Fernandez, A.; Fiestras-Janeiro, G.; Mosquera, M.A.

This paper extends the notion of individual minimal rights for a transferable utility game (TU-game) to coalitional minimal rights using minimal balanced families of a specific type, thus defining a corresponding minimal rights game. It is shown that the core of a TU-game coincides with the core of

PWR core design calculations

International Nuclear Information System (INIS)

Trkov, A.; Ravnik, M.; Zeleznik, N.

1992-01-01

Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

A comparison of database systems for XML-type data

NARCIS (Netherlands)

Risse, J.E.; Leunissen, J.A.M.

2010-01-01

Background: In the field of bioinformatics interchangeable data formats based on XML are widely used. XML-type data is also at the core of most web services. With the increasing amount of data stored in XML comes the need for storing and accessing the data. In this paper we analyse the suitability

Method Of Bonding A Metal Connection To An Electrode Including A Core Having A Fiber Or Foam Type Structure For An Electrochemical Cell, An

Science.gov (United States)

Loustau, Marie-Therese; Verhoog, Roelof; Precigout, Claude

1996-09-24

A method of bonding a metal connection to an electrode including a core having a fiber or foam-type structure for an electrochemical cell, in which method at least one metal strip is pressed against one edge of the core and is welded thereto under compression, wherein, at least in line with the region in which said strip is welded to the core, which is referred to as the "main core", a retaining core of a type analogous to that of the main core is disposed prior to the welding.

A comparison between elemental logs and core data

International Nuclear Information System (INIS)

Kerr, S.A.; Grau, J.A.; Schweitzer, J.S.

1992-01-01

Neutron-induced gamma-ray spectroscopy, of prompt capture and delayed activation, together with natural gamma-ray measurements, provides a borehole elemental analysis to characterize rock matrix composition. This study involved extensive core and log data in two wells. One well was drilled with a barite-weighted oil-based mud through a shallow marine sand. The other was drilled with fresh water-based mud through a channel sand, mudstone sequence overlying limestone. The results illustrate the importance of a suitable core sampling strategy and the problems associated with matching core to log data. Possible inaccuracies from the modelling of Ca-, Fe- and S-bearing minerals have been determined. A method for correcting the total measured aluminium concentration for that due to the borehole mud has been successfully tested against aluminium concentrations measured in the cleaned core samples. Estimates of the overall accuracy and precision of the elemental logging concentrations are obtained by comparing the log results with those obtained from the laboratory core analysis. A comprehensive core elemental analysis can also provide useful insight into the way other logs, such as the photoelectric factor or formation thermal neutron macroscopic absorption cross section, are influenced by minor and trace elements. Differences between calculated values from elemental logs and measured macroscopic parameters provide additional data for a more detailed understanding of the rock properties. (Author)

Development of advanced nuclear core analysis system applicable to various reactor types (II)

International Nuclear Information System (INIS)

Kaneko, Kunio

2003-03-01

A 900 group cross section library based on the specification determined last year was produced for 27 nuclei of the fast reactor benchmark problem evaluated in nuclear data file JENDL-3.2. In addition, the new SLAROM code, which has been developed as an advanced detail analysis system, was revised so as to make cell calculations effectively with the above 900 group library. Furthermore, new functions were added to the SLAROM so that the SLAROM evaluates assembly parameters using effective cross sections derived by the SLAROM and produces any condensed effective cross section set for core performance analysis. With the 900 group cross section library and the revised SALROM, three cell calculations for fast and medium neutron speed reactors having different neutron spectrum were performed, and the results were compared with those calculated by the continuos energy Monte Carlo code MVP. By the comparisons, it is concluded that the newly revised SLAROM and a 900 group cross section library give accuracy comparable to MVP for predicting core performances. (author)

Shaking table testing of a HTGR reactor core, comparison with the results obtained using a nonlinear mathematical model

International Nuclear Information System (INIS)

Berriaud, C.; Cebe, E.; Livolant, M.; Buland, P.

1975-01-01

Two series of horizontal tests have been performed at Saclay on the shaking table VESUVE: sinusoidal test and time history response. Sinusoidal tests have shown the strongly nonlinear dynamic behavior of the core. The resonant frequency of the core is dependent on the level of the excitation. These phenomena have been explained by a computer code, which is a lumped mass nonlinear model. El Centro time history displacement at the level of PCRV was reproduced on the shaking table. The analytical model was applied to this excitation and good comparison was obtained for forces and velocities [fr

Synthesis of Magnetic Rattle-Type Silica with Controllable Magnetite and Tunable Size by Pre-Shell-Post-Core Method.

Science.gov (United States)

Chen, Xue; Tan, Longfei; Meng, Xianwei

2016-03-01

In this study, we have developed the pre-shell-post-core route to synthesize the magnetic rattle-type silica. This method has not only simplified the precursor's process and reduced the reacting time, but also ameliorated the loss of magnetite and made the magnetite content and the inner core size controllable and tunable. The magnetite contents and inner core size can be easily controlled by changing the type and concentration of alkali, reaction system and addition of water. The results show that alkali aqueous solution promotes the escape of the precursor iron ions from the inner space of rattle-type silica and results in the loss of magnetite. In this case, NaOH ethanol solution is better for the formation of magnetite than ammonia because it not only offers an appropriate alkalinity to facilitate the synthesis of. magnetic particles, but also avoids the escape of the iron ions from the mesopores of rattle-type silica. The synthesis process is very simple and efficient, and it takes no more than 2 hours to complete the total preparation and handling of the magnetic rattle-type silica. The end-product Fe3O4@SiO2 nanocomposites also have good magnetic properties which will perform potential application in biomedical science.

VVER-specific features regarding core degradation - Status Report

International Nuclear Information System (INIS)

Hozer, Z.; Trambauer, K.; Duspiva, J.

1999-01-01

The objective of this report is to compare VVER reactors to pressurised water reactors (PWRs) of Western design from the point of view of core degradation phenomena using the terminology which was applied to the systematisation of severe accident phenomena in earlier CSNI reports. In the following the acronym 'PWR' is used for a PWR of Western design. The basic design features are described and the most important parameters are summarised in order to identify the differences between the two reactor types. In some specific cases the comparison shows more similarities with boiling water reactors (BWRs) than with PWRs. The known VVER experimental support is also summarised. RBMKs are not included in this report, as this reactor type is not operated in OECD countries, furthermore its design is completely different from those of VVERs and PWRs. The scope of this report is limited to in-vessel severe fuel damage phenomena. Neither thermal hydraulic processes involving no core degradation, nor containment phenomena, are discussed in detail. The VVER (water-cooled water-moderated power reactor) is a pressurised light water reactor of Soviet design. It operates on the same principles as a Western PWR reactor and uses similar technological systems. The primary coolant is pressurised water, which heats up in the reactor core and steam is produced on the secondary side of steam generators. The comparison of basic geometrical and technological parameters pointed out some differences between a PWR and a VVER, but it should be noted that differences exist even between two Western PWRs of different design. The VVER reactors are special types of PWRs, the most important design features of which are the horizontal steam generators and the hexagonal core structure. Similarity between PWR and VVER reactors was found in the comparison of dominant accidents sequences leading to core melt. The accident progression sequence consists of the same steps for VVERs and PWRs. The larger water

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)

Transport-diffusion comparisons for small core LMFBR disruptive accidents

International Nuclear Information System (INIS)

Tomlinson, E.T.

1977-11-01

A number of numerical experiments were performed to assess the validity of diffusion theory for calculating the reactivity state of various small core LMFBR disrupted geometries. The disrupted configurations correspond, in general, to various configurations predicted by SAS3A for transient undercooling (TUC) and transient overpower (TOP) accidents for homogeneous cores and to the ZPPR-7 configurations for heterogeneous core. In all TUC cases diffusion theory was shown to be inadequate for the calculation of reactivity changes during core disassembly

CONSUL code package application for LMFR core calculations

Energy Technology Data Exchange (ETDEWEB)

Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)

2008-07-01

CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)

The new lattice code Paragon and its qualification for PWR core applications

International Nuclear Information System (INIS)

Ouisloumen, M.; Huria, H.C.; Mayhue, L.T.; Smith, R.M.; Kichty, M.J.; Matsumoto, H.; Tahara, Y.

2003-01-01

Paragon is a new two-dimensional transport code based on collision probability with interface current method and written entirely in Fortran 90/95. The qualification of Paragon has been completed and the results are very good. This qualification included a number of critical experiments. Comparisons to the Monte Carlo code MCNP for a wide variety of PWR assembly lattice types were also performed. In addition, Paragon-based core simulator models have been compared against PWR plant startup and operational data for a large number of plants. Some results of these calculations and also comparisons against models developed with a licensed Westinghouse lattice code, Phoenix-P, are presented. The qualification described in this paper provided the basis for the qualification of Paragon both as a validated transport code and as the nuclear data source for core simulator codes

Characterizing the Core via K-Core Covers

NARCIS (Netherlands)

Sanchez, S.M.; Borm, P.E.M.; Estevez, A.

2013-01-01

This paper extends the notion of individual minimal rights for a transferable utility game (TU-game) to coalitional minimal rights using minimal balanced families of a specific type, thus defining a corresponding minimal rights game. It is shown that the core of a TU-game coincides with the core of

Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model

International Nuclear Information System (INIS)

Oliveira, A.C.J.G. de.

1988-12-01

Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs

Comparison results on preconditioned SOR-type iterative method for Z-matrices linear systems

Science.gov (United States)

Wang, Xue-Zhong; Huang, Ting-Zhu; Fu, Ying-Ding

2007-09-01

In this paper, we present some comparison theorems on preconditioned iterative method for solving Z-matrices linear systems, Comparison results show that the rate of convergence of the Gauss-Seidel-type method is faster than the rate of convergence of the SOR-type iterative method.

Characteristics of Core Thermal-Hydraulic Design of SMART-P

International Nuclear Information System (INIS)

Hwang, Dae-Hyun; Seo, Kyong-Won; Kim, Tae-Wan; Lee, Chung-Chan

2006-01-01

The SMART (System-Integrated Modular Advanced ReacTor) is an integral-type advanced light water reactor which is purposed to be utilized as an energy source for sea water desalination as well as a small scale power generation. A prototype of this reactor, named SMART-P, has been studied at KAERI in order to demonstrate the relevant technologies incorporated in the SMART design. Due to the closed-channel type fuel assemblies and low mass velocity in the reactor core, the thermal hydraulic design features of SMART-P revealed fairly different characteristics in comparison with existing PWRs. The allowable operating region of the core, from the aspect of the thermal integrity of the fuel, should be primarily limited by two design parameters; critical heat flux (CHF) and fuel temperature. The occurrence of CHF may cause a sudden increase of the cladding temperature which eventually results in the fuel failure. The fuel temperature limit is relevant to a fuel failure mechanism such as a fuel centerline melting or a phase change of metallic fuels. Two phase flow instability is also an important design parameter since a flow oscillation may trigger a CHF or mechanical vibration of the channel. The characteristics of important thermal-hydraulic design parameters have been investigated for the SMART-P core with the closed-channel type fuel assemblies which contained non-square arrayed SSF (Self-sustained Square Finned) fuel rods

Preliminaries on core image analysis using fault drilling samples; Core image kaiseki kotohajime (danso kussaku core kaisekirei)

Energy Technology Data Exchange (ETDEWEB)

Miyazaki, T; Ito, H [Geological Survey of Japan, Tsukuba (Japan)

1996-05-01

This paper introduces examples of image data analysis on fault drilling samples. The paper describes the following matters: core samples used in the analysis are those obtained from wells drilled piercing the Nojima fault which has moved in the Hygoken-Nanbu Earthquake; the CORESCAN system made by DMT Corporation, Germany, used in acquiring the image data consists of a CCD camera, a light source and core rotation mechanism, and a personal computer, its resolution being about 5 pixels/mm in both axial and circumferential directions, and 24-bit full color; with respect to the opening fractures in core samples collected by using a constant azimuth coring, it was possible to derive values of the opening width, inclination angle, and travel from the image data by using a commercially available software for the personal computer; and comparison of this core image with the BHTV record and the hydrophone VSP record (travel and inclination obtained from the BHTV record agree well with those obtained from the core image). 4 refs., 4 figs.

Comparisons with measured data of the simulated local core parameters by the coupled code ATHLET-BIPR-VVER applying a new enhanced model of the reactor pressure vessel

International Nuclear Information System (INIS)

Nikonov, S.; Pasichnyk, I.; Velkov, K.; Pautz, A.

2011-01-01

The paper describes the performed comparisons of measured and simulated local core data based on the OECD/NEA Benchmark on Kalinin-3 NPP: 'Switching off of one of the four operating main circulation pumps at nominal reactor power'. The local measurements of in core self-powered neutron detectors (SPND) in 64 fuel assemblies on 7 axial levels are used for the comparisons of the assemblies axial power distributions and the thermocouples readings at 93 fuel assembly heads are applied for the fuel assembly coolant temperature comparisons. The analyses are done on the base of benchmark transient calculations performed with the coupled system code ATHLET/BIPR-VVER. In order to describe more realistically the fluid mixing phenomena in a reactor pressure vessel a new enhanced nodalization scheme is being developed. It could take into account asymmetric flow behaviour in the reactor pressure vessel structures like downcomer, reactor core inlet and outlet, control rods' guided tubes, support grids etc. For this purpose details of the core geometry are modelled. About 58000 control volumes and junctions are applied. Cross connection are used to describe the interaction between the fluid objects. The performed comparisons are of great interest because they show some advantages by performing coupled code production pseudo-3D analysis of NPPs applying the parallel thermo-hydraulic channel methodology (or 1D thermo-hydraulic system code modeling). (Authors)

Research on Shock Responses of Three Types of Honeycomb Cores

Science.gov (United States)

Peng, Fei; Yang, Zhiguang; Jiang, Liangliang; Ren, Yanting

2018-03-01

The shock responses of three kinds of honeycomb cores have been investigated and analyzed based on explicit dynamics analysis. According to the real geometric configuration and the current main manufacturing methods of aluminum alloy honeycomb cores, the finite element models of honeycomb cores with three different cellular configurations (conventional hexagon honeycomb core, rectangle honeycomb core and auxetic honeycomb core with negative Poisson’s ratio) have been established through FEM parametric modeling method based on Python and Abaqus. In order to highlight the impact response characteristics of the above three honeycomb cores, a 5 mm thick panel with the same mass and material was taken as contrast. The analysis results showed that the peak values of longitudinal acceleration history curves of the three honeycomb cores were lower than those of the aluminum alloy panel in all three reference points under the loading of a longitudinal pulse pressure load with the peak value of 1 MPa and the pulse width of 1 μs. It could be concluded that due to the complex reflection and diffraction of stress wave induced by shock in honeycomb structures, the impact energy was redistributed which led to a decrease in the peak values of the longitudinal acceleration at the measuring points of honeycomb cores relative to the panel.

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

Overview of homogeneous versus heterogeneous core configuration trade-off studies

International Nuclear Information System (INIS)

Chang, Y.I.

1982-01-01

The most significant development in core design trend in the U.S. LMFBR program has been the increased attention given to the heterogeneous core design concept. In recent years, numerous core configuration trade-off studies have been carried out to quantify advantages and disadvantages of the heterogeneous concept relative to the homogeneous concept, and a consensus is emerging among the U.S. core designers. It appears that the technical and economic performance differences between homogeneous and heterogeneous core designs are very small; however, the heterogeneous concept provides a definite safety/licensing advantage. he technical and economic performance comparison between homogeneous and heterogeneous core configurations is difficult to quantify. In fact, in most cases, the perceived advantages and/or disadvantages are dictated by the consistency in the comparison (optimized for one concept versus non-optimized for the other, etc.) rather than by any inherent differences. Some of the technical and economic issues relevant to the homogeneous versus heterogeneous comparison are summarized

Overview of homogeneous versus heterogeneous core configuration trade-off studies

Energy Technology Data Exchange (ETDEWEB)

Chang, Y I [Applied Physics Division, Argonne National Laboratory, Argonne, IL (United States)

1982-01-01

The most significant development in core design trend in the U.S. LMFBR program has been the increased attention given to the heterogeneous core design concept. In recent years, numerous core configuration trade-off studies have been carried out to quantify advantages and disadvantages of the heterogeneous concept relative to the homogeneous concept, and a consensus is emerging among the U.S. core designers. It appears that the technical and economic performance differences between homogeneous and heterogeneous core designs are very small; however, the heterogeneous concept provides a definite safety/licensing advantage. he technical and economic performance comparison between homogeneous and heterogeneous core configurations is difficult to quantify. In fact, in most cases, the perceived advantages and/or disadvantages are dictated by the consistency in the comparison (optimized for one concept versus non-optimized for the other, etc.) rather than by any inherent differences. Some of the technical and economic issues relevant to the homogeneous versus heterogeneous comparison are summarized.

Core losses of a permanent magnet synchronous motor with an amorphous stator core under inverter and sinusoidal excitations

Science.gov (United States)

Yao, Atsushi; Sugimoto, Takaya; Odawara, Shunya; Fujisaki, Keisuke

2018-05-01

We report core loss properties of permanent magnet synchronous motors (PMSM) with amorphous magnetic materials (AMM) core under inverter and sinusoidal excitations. To discuss the core loss properties of AMM core, a comparison with non-oriented (NO) core is also performed. In addition, based on both experiments and numerical simulations, we estimate higher (time and space) harmonic components of the core losses under inverter and sinusoidal excitations. The core losses of PMSM are reduced by about 59% using AMM stator core instead of NO core under sinusoidal excitation. We show that the average decrease obtained by using AMM instead of NO in the stator core is about 94% in time harmonic components.

Core losses of a permanent magnet synchronous motor with an amorphous stator core under inverter and sinusoidal excitations

Directory of Open Access Journals (Sweden)

Atsushi Yao

2018-05-01

Full Text Available We report core loss properties of permanent magnet synchronous motors (PMSM with amorphous magnetic materials (AMM core under inverter and sinusoidal excitations. To discuss the core loss properties of AMM core, a comparison with non-oriented (NO core is also performed. In addition, based on both experiments and numerical simulations, we estimate higher (time and space harmonic components of the core losses under inverter and sinusoidal excitations. The core losses of PMSM are reduced by about 59% using AMM stator core instead of NO core under sinusoidal excitation. We show that the average decrease obtained by using AMM instead of NO in the stator core is about 94% in time harmonic components.

Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

International Nuclear Information System (INIS)

Shemon, Emily R.

2016-01-01

not conservative and could be overestimating reactivity feedback effects that are closely tied to reactor safety. We conclude that there is indeed value in performing direct simulation of deformed meshes despite the increased computational expense. PROTEUS-SN is already part of the SHARP multi-physics toolkit where both thermal hydraulics and structural mechanical feedback modeling can be applied but this is the first comparison of direct simulation to legacy techniques for radial core expansion.

Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

Energy Technology Data Exchange (ETDEWEB)

Shemon, Emily R. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-10-10

not conservative and could be overestimating reactivity feedback effects that are closely tied to reactor safety. We conclude that there is indeed value in performing direct simulation of deformed meshes despite the increased computational expense. PROTEUS-SN is already part of the SHARP multi-physics toolkit where both thermal hydraulics and structural mechanical feedback modeling can be applied but this is the first comparison of direct simulation to legacy techniques for radial core expansion.

Comparison of steam-generator liquid holdup and core uncovery in two facilities of differing scale

International Nuclear Information System (INIS)

Motley, F.; Schultz, R.

1987-01-01

This paper reports on Run SB-CL-05, a test similar to Semiscale Run S-UT-8. The test results show that the core was uncovered briefly during the accident and that the rods overheated at certain core locations. Liquid holdup on the upflow side of the steam-generator tubes was observed. After the loop seal cleared, the core refilled and the rods cooled. These behaviors were similar to those observed in the Semiscale run. The Large-Scale Test Facility (LSTF) Run SB-CL-06 is a counterpart test to Semiscale Run S-LH-01. The comparison of the results of both tests shows similar phenomena. The similarity of phenomena in these two facilities build confidence that these results can be expected to occur in a PWR. Similar holdup has now been observed in the 6 tubes of Semiscale and in the 141 tubes of LSTF. It is now more believable that holdup may occur in a full-scale steam generator with 3000 or more tubes. These results confirm the scaling of these phenomena from Semiscale (1/1705) to LSTF (1/48). The TRAC results for SB-CL-05 are in reasonable agreement with the test data. TRAC predicted the core uncovery and resulting rod heatup. The liquid holdup on the upflow side of the steam-generator tubes was also correctly predicted. The clearing of the loop seal allowed core recovery and cooled the overheated rods just as it had in the data. The TRAC analysis results of Run SB-CL-05 are similar to those from Semiscale Run S-UT-8. The ability of the TRAC code to calculate the phenomena equally well in the two experiments of different scales confirms the scalability of the many models in the code that are important in calculating this small break

THE ANGULAR MOMENTUM OF MAGNETIZED MOLECULAR CLOUD CORES: A TWO-DIMENSIONAL-THREE-DIMENSIONAL COMPARISON

International Nuclear Information System (INIS)

Dib, Sami; Csengeri, Timea; Audit, Edouard; Hennebelle, Patrick; Pineda, Jaime E.; Goodman, Alyssa A.; Bontemps, Sylvain

2010-01-01

In this work, we present a detailed study of the rotational properties of magnetized and self-gravitating dense molecular cloud (MC) cores formed in a set of two very high resolution three-dimensional (3D) MC simulations with decaying turbulence. The simulations have been performed using the adaptative mesh refinement code RAMSES with an effective resolution of 4096 3 grid cells. One simulation represents a mildly magnetically supercritical cloud and the other a strongly magnetically supercritical cloud. We identify dense cores at a number of selected epochs in the simulations at two density thresholds which roughly mimic the excitation densities of the NH 3 (J - K) = (1,1) transition and the N 2 H + (1-0) emission line. A noticeable global difference between the two simulations is the core formation efficiency (CFE) of the high-density cores. In the strongly supercritical simulations, the CFE is 33% per unit free-fall time of the cloud (t ff,cl ), whereas in the mildly supercritical simulations this value goes down to ∼6 per unit t ff,cl . A comparison of the intrinsic specific angular momentum (j 3D ) distributions of the cores with the specific angular momentum derived using synthetic two-dimensional (2D) velocity maps of the cores (j 2D ) shows that the synthetic observations tend to overestimate the true value of the specific angular momentum by a factor of ∼8-10. We find that the distribution of the ratio j 3D /j 2D of the cores peaks at around ∼0.1. The origin of this discrepancy lies in the fact that contrary to the intrinsic determination of j which sums up the individual gas parcels' contributions to the angular momentum, the determination of the specific angular momentum using the standard observational procedure which is based on a measurement on the global velocity gradient under the hypothesis of uniform rotation smoothes out the complex fluctuations present in the 3D velocity field. Our results may well provide a natural explanation for the

Neutronic characteristics of FLWR in the transition phase changing from high conversion core to breeder core

International Nuclear Information System (INIS)

Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

2009-01-01

Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a low moderation type boiling water reactor which can realize plutonium multiple recycling and breeding. For the introduction stage of FLWR, a high conversion (HC) type FLWR is proposed to keep technical continuity from current light water reactors. The HC core of FLWR has a less tight fuel lattice with lower coolant void fraction than the breeder (BR) type core. The HC type FLWR core is to be shifted to the BR core by only replacing the fuel assemblies of the same outer shape and size in the same reactor system. In the HC to BR transition phase of FLWR, there exist both types of fuel assemblies in the same core configuration. In the HC assembly, neutron spectrum is softer than in the BR assembly, and the axial fuel and blanket arrangement is different from the BR assembly. Due to these differences, there might appear a power peaking in the adjacent region between HC and BR assemblies. The power distribution in the HC + BR assemblies mixed core configuration is studied by performing assembly calculations and core calculations on a few assemblies local geometry and the whole core geometry. As a result, although a power peaking can be locally very large in the HC and BR assemblies adjacent regions, such local power peakings are shown to be effectively reduced by considering a rod-wise fuel enrichment distribution. In the whole core calculation, it seems possible to optimize the fuel assembly loading and shuffling pattern to avoid large power level mismatch between the assemblies. It is expected that FLWR can be shifted from HC type to BR type without major neutronic difficulties. (author)

CALTECH CORE-COLLAPSE PROJECT (CCCP) OBSERVATIONS OF TYPE II SUPERNOVAE: EVIDENCE FOR THREE DISTINCT PHOTOMETRIC SUBTYPES

Energy Technology Data Exchange (ETDEWEB)

Arcavi, Iair; Gal-Yam, Avishay; Yaron, Ofer [Department of Particle Physics and Astrophysics, Weizmann Institute of Science, Rehovot 76100 (Israel); Cenko, S. Bradley; Becker, Adam B. [Department of Astronomy, University of California, Berkeley, CA 94720-3411 (United States); Fox, Derek B. [Department of Astronomy and Astrophysics, Pennsylvania State University, University Park, PA 16802 (United States); Leonard, Douglas C. [Department of Astronomy, San Diego State University, San Diego, CA 92182 (United States); Moon, Dae-Sik [Department of Astronomy and Astrophysics, University of Toronto, Toronto, ON M5S 3H4 (Canada); Sand, David J. [Las Cumbres Observatory Global Telescope Network, Santa Barbara, CA 93117 (United States); Soderberg, Alicia M. [Harvard-Smithsonian Center for Astrophysics, Cambridge, MA 02138 (United States); Kiewe, Michael [Department of Physics, University of Wisconsin, Madison, WI 53706 (United States); Scheps, Raphael [King' s College, University of Cambridge, Cambridge CB2 1ST (United Kingdom); Birenbaum, Gali [12 Amos St, Ramat Chen, Ramat Gan 52233 (Israel); Chamudot, Daniel [20 Chen St, Petach Tikvah 49520 (Israel); Zhou, Jonathan, E-mail: iair.arcavi@weizmann.ac.il [101 Dunster Street, Box 398, Cambridge, MA 02138 (United States)

2012-09-10

We present R-band light curves of Type II supernovae (SNe) from the Caltech Core-Collapse Project (CCCP). With the exception of interacting (Type IIn) SNe and rare events with long rise times, we find that most light curve shapes belong to one of three apparently distinct classes: plateau, slowly declining, and rapidly declining events. The last class is composed solely of Type IIb SNe which present similar light curve shapes to those of SNe Ib, suggesting, perhaps, similar progenitor channels. We do not find any intermediate light curves, implying that these subclasses are unlikely to reflect variance of continuous parameters, but rather might result from physically distinct progenitor systems, strengthening the suggestion of a binary origin for at least some stripped SNe. We find a large plateau luminosity range for SNe IIP, while the plateau lengths seem rather uniform at approximately 100 days. As analysis of additional CCCP data goes on and larger samples are collected, demographic studies of core-collapse SNe will likely continue to provide new constraints on progenitor scenarios.

Hydraulic Profiling of a Parallel Channel Type Reactor Core

International Nuclear Information System (INIS)

Seo, Kyong-Won; Hwang, Dae-Hyun; Lee, Chung-Chan

2006-01-01

An advanced reactor core which consisted of closed multiple parallel channels was optimized to maximize the thermal margin of the core. The closed multiple parallel channel configurations have different characteristics to the open channels of conventional PWRs. The channels, usually assemblies, are isolated hydraulically from each other and there is no cross flow between channels. The distribution of inlet flow rate between channels is a very important design parameter in the core because distribution of inlet flow is directly proportional to a margin for a certain hydraulic parameter. The thermal hydraulic parameter may be the boiling margin, maximum fuel temperature, and critical heat flux. The inlet flow distribution of the core was optimized for the boiling margins by grouping the inlet orifices by several hydraulic regions. The procedure is called a hydraulic profiling

RA-3 core with uranium silicide fuel elements

International Nuclear Information System (INIS)

Abbate, Maximo J.; Sbaffoni, Maria M.

2000-01-01

Following on with studies on uranium silicide fuel elements, this paper reports some comparisons between the use of standard ECN [U 3 O 8 ] fuel elements and type P-06 [from U 3 Si 2 ] fuel elements in the RA-3 core.The first results showed that the calculated overall mean burn up is in agreement with that reported for the facility, which gives more confidence to the successive ones. Comparing the mentioned cores, the silicide one presents several advantages such as: -) a mean burn up increase of 18 %; -) an extraction burn up increase of 20 %; -) 37.4 % increase in full power days, for mean burn up. All this is meritorious for this fuel. Moreover, grouped and homogenized libraries were prepared for CITVAP code that will be used for planning experiments and other bidimensional studies. Preliminary calculations were also performed. (author)

Effect of core configuration on the burnup calculations of MTR research reactors

International Nuclear Information System (INIS)

Hussein, H.M.; Amin, E.H.; Sakr, A.M.

2014-01-01

Highlights: • 3D burn-up calculations of MTR-type research reactor were performed. Examination of the effect of control rod pattern on power density and neutron flux distributions is presented. • The calculations are performed using the MTR P C package and the programs (WIMS and CITVAP). • An empirical formula was generated for every fuel element type, to correlate irradiation to burn-up. - Abstract: In the present paper, three-dimensional burn-up calculations were performed using different patterns of control rods, in order to examine their effect on power density and neutron flux distributions through out the entire core and hence on the local burn-up distribution. These different cores burn-up calculations are carried out for an operating cycle equivalent to 15 Full Power Days (FPDs), with a power rating of 22 MW. Calculations were performed using an example of a typical research reactor of MTR-type using the internationally known computer codes’ package “MTR P C system”, using the cell calculation transport code WIMS-D4 with 12 energy groups and the core calculation diffusion code CITVAP with 5 energy groups. A depletion study was done and the effects on the research reactor fuel (U-235) were performed. The burn-up percentage (B.U.%) curves for every fuel element type were drawn versus irradiation (MWD/TE). Then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Charts of power density and neutron flux distribution for each core were plotted at different sections of each fuel element of the reactor core. Then a complete discussion and analysis of these curves are performed with comparison between the different core configurations, illustrating the effect of insertion or extraction of either of the four control rods directly on the neutron flux and consequently on the power distribution and burn-up. A detailed study of fuel burn-up gives detailed insight on the different B.U.% calculations

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.

Seismic response of a block-type nuclear reactor core

International Nuclear Information System (INIS)

Dove, R.C.; Bennett, J.G.; Merson, J.L.

1976-05-01

An analytical model is developed to predict seismic response of large gas-cooled reactor cores. The model is used to investigate scaling laws involved in the design of physical models of such cores, and to make parameter studies

Fuel loading method to exchangeable reactor core of BWR type reactor and its core

International Nuclear Information System (INIS)

Koguchi, Kazushige.

1995-01-01

In a fuel loading method for an exchangeable reactor core of a BWR type reactor, at least two kinds of fresh fuel assemblies having different reactivities between axial upper and lower portions are preliminarily prepared, and upon taking out fuel assemblies of advanced combustion and loading the fresh fuel assemblies dispersingly, they are disposed so as to attain a predetermined axial power distribution in the reactor. At least two kinds of fresh fuel assemblies have a content of burnable poisons different between the axial upper portion and lower portions. In addition, reactivity characteristics are made different at a region higher than the central boundary and a region lower than the central boundary which is set within a range of about 6/24 to 16/24 from the lower portion of the fuel effective length. There can be attained axial power distribution as desired such as easy optimization of the axial power distribution, high flexibility, and flexible flattening of the power distribution, and it requires no special change in view of the design and has a good economical property. (N.H.)

Measurements of the isothermal temperature reactivity coefficient of KUCA C-Core with a D{sub 2}O tank

Energy Technology Data Exchange (ETDEWEB)

Pyeon, Cheol Ho [Research Reactor Institute, Kyoto Univ., Osaka (Japan); Shim, Hyung Jin; Choi, Sung Hoon; Jeon, Byoung Kyu [Seoul National Univ., Seoul (Korea, Republic of); Ryu, Eun Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

The Kyoto University Critical Assembly (KUCA) is a multi-core type critical assembly consisting of three independent cores in the Kyoto University Research Reactor Institute. The light-water-moderated core (Ccore) is a tank type reactor, and the experiments of the isothermal temperature reactivity coefficient (ITRC) of C-core with a D{sub 2}O tank were carried out with the use of six 10 kW heaters and a radiator system in a dump tank, one 10 kW heater in a core tank, and one 5 kW heater in the D{sub 2}O tank. The ITRCs of the C-core with the D{sub 2}O tank immersed in the core tank are considered important to investigate the mechanism of moderation and reflection effects of H{sub 2}O and D{sub 2}O in the core on the evaluation by numerical simulations. The objectives of this paper are to report the ITRC measurements for C-core with D{sub 2}O tank ranging between 26.7 .deg. C and 58.5 .deg. C, and to examine the accuracy of the numerical simulations by the Seoul National University Monte Carlo code, McCARD, through the comparison between measured and calculated results.

Study of core characteristics on fuel and coolant type. Results of F/S phase-I

International Nuclear Information System (INIS)

Ikegami, Tetsuo; Hayashi, Hideyuki; Sasaki, Makoto; Mizuno, Tomoyasu; Yamadate, Megumi; Takaki, Naoyuki; Kurosawa, Norifumi; Sakashita, Yoshiaki; Naganuma, Masayuki

2001-03-01

The phase-I of the Feasibility Study of Commercialized Fast Reactor Cycle Systems (F/S) were started from July, 1999 and terminated at the end of FY2000 in order to executed examination about technology alternatives of various commercialized fast reactor (FR) recycle concepts, in response to the JNC middle long term enterprise plan. In the phase-I of this F/S, a number of conceptual candidates have been selected from the following 5 viewpoints: a) ensuring safety, b) economic competitiveness to future LWRs, c) efficient utilization of resources, d) reduction of environmental burden, e) enhancement of nuclear non-proliferation. As for this study from the above viewpoints, core characteristics of many kinds of reactors have been investigated, analyzed and examined a core / a fuel characteristic in the combinations of fuel and coolant types and power output scales. Based on these results, R and D plans of the phase-II to be performed have been proposed, and a database to select candidate reactor concepts has been prepared. The conclusions have been obtained in the phase-I are as follows: (1) Evaluation of a fuel form in every each coolant was compared. A promising fuel form was extracted as follows: an oxide and a metal fuel for sodium coolant cores, a metal and a nitride fuel for heavy metal coolant cores, an oxide and a nitride fuel for carbon dioxide coolant cores and a nitride fuel for He gas coolant cores. (2) As the general idea that performance of a core nucleus can be compatible with re-criticality evasion in sodium coolant large-sized oxide fuel cores, a axial blanket particle elimination radial heterogeneous core is one influential candidate. (3) In case of Pb-Bi coolant nature circulation medium size core with an oxide fuel, it is difficult to simultaneously achieve higher discharged burn-up and higher breeding ratio according to the viewpoints of the phase-I. (4) Core characteristics of a carbon dioxide coolant core shows to be almost equivalent to that of

Guiding and amplification properties of rod-type photonic crystal fibers with sectioned core doping

DEFF Research Database (Denmark)

Selleri, Stefano; Poli, Federica; Passaro, Davide

2009-01-01

Rod-type photonic crystal fibers are large mode area double-cladding fibers with an outer diameter of few millimeters which can provide important advantages for high-power lasers and amplifiers. Numerical studies have recently demonstrated the guidance of higher-order modes in these fibers, which...... can worsen the output beam quality of lasers and amplifiers. In the present analysis a sectioned core doping has been proposed for Yb-doped rod-type photonic crystal fibers, with the aim to improve the higher-order mode suppression. A full-vector modal solver based on the finite element method has...

Core schemas across the continuum of psychosis: a comparison of clinical and non-clinical groups.

Science.gov (United States)

Taylor, Hannah E; Stewart, Suzanne L K; Dunn, Graham; Parker, Sophie; Fowler, David; Morrison, Anthony P

2014-11-01

Research suggests that core schemas are important in both the development and maintenance of psychosis. The aim of the study was to investigate and compare core schemas in four groups along the continuum of psychosis and examine the relationships between schemas and positive psychotic symptomatology. A measure of core schemas was distributed to 20 individuals experiencing first-episode psychosis (FEP), 113 individuals with "at risk mental states" (ARMS), 28 participants forming a help-seeking clinical group (HSC), and 30 non-help-seeking individuals who endorse some psychotic-like experiences (NH). The clinical groups scored significantly higher than the NH group for negative beliefs about self and about others. No significant effects of group on positive beliefs about others were found. For positive beliefs about the self, the NH group scored significantly higher than the clinical groups. Furthermore, negative beliefs about self and others were related to positive psychotic symptomatology and to distress related to those experiences. Negative evaluations of the self and others appear to be characteristic of the appraisals of people seeking help for psychosis and psychosis-like experiences. The results support the literature that suggests that self-esteem should be a target for intervention. Future research would benefit from including comparison groups of people experiencing chronic psychosis and people who do not have any psychotic-like experiences.

NSGA-II Algorithm with a Local Search Strategy for Multiobjective Optimal Design of Dry-Type Air-Core Reactor

Directory of Open Access Journals (Sweden)

Chengfen Zhang

2015-01-01

Full Text Available Dry-type air-core reactor is now widely applied in electrical power distribution systems, for which the optimization design is a crucial issue. In the optimization design problem of dry-type air-core reactor, the objectives of minimizing the production cost and minimizing the operation cost are both important. In this paper, a multiobjective optimal model is established considering simultaneously the two objectives of minimizing the production cost and minimizing the operation cost. To solve the multi-objective optimization problem, a memetic evolutionary algorithm is proposed, which combines elitist nondominated sorting genetic algorithm version II (NSGA-II with a local search strategy based on the covariance matrix adaptation evolution strategy (CMA-ES. NSGA-II can provide decision maker with flexible choices among the different trade-off solutions, while the local-search strategy, which is applied to nondominated individuals randomly selected from the current population in a given generation and quantity, can accelerate the convergence speed. Furthermore, another modification is that an external archive is set in the proposed algorithm for increasing the evolutionary efficiency. The proposed algorithm is tested on a dry-type air-core reactor made of rectangular cross-section litz-wire. Simulation results show that the proposed algorithm has high efficiency and it converges to a better Pareto front.

Do we really know Mup (i.e. the transition mass between Type Ia and core-collapse supernova progenitors)?

International Nuclear Information System (INIS)

Straniero, O; Piersanti, L; Cristallo, S

2016-01-01

M up is the minimum stellar mass that, after the core-helium burning, develops temperature and density conditions for the occurrence of a hydrostatic carbon burning. Stars whose mass is lower than this limit are the progenitors of C-O white dwarfs and, when belong to a close binary system, may give rise to explosive phenomena, such as novae or type Ia supernovae. Stars whose mass is only slightly larger than M up ignite C in a degenerate core and, in turn, experience a thermonuclear runaway. Their final fate may be a massive O-Ne WDs or, if the core mass approaches the Chandrasekhar limit, an e-capture SNe. More massive objects ignite C in non-degenerate conditions. These “massive “ stars are the progenitors of various kind of core-collapse supernovae (type IIp. IIL, IIN, Ib, Ic). It goes without saying that M up is a fundamental astrophysical parameter. From its knowledge depends our understanding of the SNe progenitors, of their rates, of the chemical evolution, of the WD luminosity functions and much more. A precise evaluation of M up relies on our knowledge of various input physics used in stellar modeling, such as the plasma neutrino rate, responsible of the cooling of the core, the equation of state of high density plasma, which affects the heating of the contracting core and its compressibility, and some key nuclear reaction rates, such as, in particular, the 12 C+ 12 C and the 12 C+α. In this paper we review the efforts made to determine this important parameter and we provide an up-to-date evaluation of the uncertainties due to the relevant nuclear physics inputs. (paper)

Determination of 210Pb in marine sediment core- A comparison between alpha and gamma techniques

International Nuclear Information System (INIS)

Zal Uyun Wan Mahmood; Yii Mei Wo; Zaharudin Ahmad

2010-01-01

The most commonly used techniques of alpha and gamma spectrometry were performed to measure 210 Pb activity in marine sediment core as a comparison. Alpha analytical technique measured the activity of 210 Pb from its in-grow grand-daughter 210 Po after a chemical separation, assuming radioactive equilibrium between the two radionuclides. Meanwhile, gamma analysis technique allows direct measurement, non-destructive and no preliminary chemical separation. Through the comparison, it is found that both alpha and gamma analysis techniques were slightly difference. Overall, the results from alpha analytical technique were basically higher than those from the gamma analytical techniques. Some logical argument had been discussed to explain this situation. In routine analysis, the analytical technique used should be chosen carefully based on advantages and disadvantages of the each technique and analysis requirements. Therefore, it is recommended to determine exactly the needs and purpose of analysis and to know the sample history before decide the appropriate analytical technique. (author)

Analysis of RA-8 critical facility core in some configurations

International Nuclear Information System (INIS)

Abbate, Maximo J.; Sbaffoni, Maria M.

2000-01-01

The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

Benchmarking and qualification of the NUFREQ-NPW code for best estimate prediction of multi-channel core stability margins

International Nuclear Information System (INIS)

Taleyarkhan, R.; Lahey, R.T. Jr.; McFarlane, A.F.; Podowski, M.Z.

1988-01-01

The NUFREQ-NPW code was modified and set up at Westinghouse, USA for mixed fuel type multi-channel core-wide stability analysis. The resulting code, NUFREQ-NPW, allows for variable axial power profiles between channel groups and can handle mixed fuel types. Various models incorporated into NUFREQ-NPW were systematically compared against the Westinghouse channel stability analysis code MAZDA-NF, for which the mathematical model was developed, in an entirely different manner. Excellent agreement was obtained which verified the thermal-hydraulic modeling and coding aspects. Detailed comparisons were also performed against nuclear-coupled reactor core stability data. All thirteen Peach Bottom-2 EOC-2/3 low flow stability tests were simulated. A key aspect for code qualification involved the development of a physically based empirical algorithm to correct for the effect of core inlet flow development on subcooled boiling. Various other modeling assumptions were tested and sensitivity studies performed. Good agreement was obtained between NUFREQ-NPW predictions and data. Moreover, predictions were generally on the conservative side. The results of detailed direct comparisons with experimental data using the NUFREQ-NPW code; have demonstrated that BWR core stability margins are conservatively predicted, and all data trends are captured with good accuracy. The methodology is thus suitable for BWR design and licensing purposes. 11 refs., 12 figs., 2 tabs

A core design study for 'zero-sodium-void-worth' cores

International Nuclear Information System (INIS)

Kawashima, Masatoshi; Suzuki, Masao; Hill, R.N.

1992-01-01

Recently, a number of low sodium-void-worth metal-fueled core design concepts have been proposed; to provide for flexibility in transuranic nuclide management strategy, core designs which exhibit a wide range of breeding characteristics have been developed. Two core concepts, a flat annular (transuranic burning) core and an absorber-type parfait (transuranic self-sufficient) core, are selected for this study. In this paper, the excess reactivity management schemes applied in the two designs are investigated in detail. In addition, the transient effect of reactivity insertions on the parfait core design is assessed. The upper and lower core regions in the parfait design are neutronically decoupled; however, the common coolant channel creates thermalhydraulic coupling. This combination of neutronic and thermalhydraulic characteristics leads to unique behavior in anticipated transient overpower events. (author)

Digital data storage of core image using high resolution full color core scanner; Kokaizodo full color scanner wo mochiita core image no digital ka

Energy Technology Data Exchange (ETDEWEB)

Takahashi, W; Ujo, S; Osato, K; Takasugi, S [Geothermal Energy Research and Development Co. Ltd., Tokyo (Japan)

1996-05-01

This paper reports on digitization of core images by using a new type core scanner system. This system consists of a core scanner unit (equipped with a CCD camera), a personal computer and ancillary devices. This is a modification of the old type system, with measurable core length made to 100 cm/3 scans, and resolution enhanced to 5100 pixels/m (1024 pixels/m in the old type). The camera was changed to that of a color specification, and the A/D conversion was improved to 24-bit full color. As a result of carrying out a detail reproduction test on digital images of this core scanner, it was found that objects can be identified at a level of about the size of pixels constituting the image in the case when the best contrast is obtained between the objects and the background, and that in an evaluation test on visibility of concaves and convexes on core surface, reproducibility is not very good in large concaves and convexes. 2 refs., 6 figs.

Seismic response of high temperature gas-cooled reactor core with block-type fuel, (2)

International Nuclear Information System (INIS)

Ikushima, Takeshi; Honma, Toshiaki.

1980-01-01

For the aseismic design of a high temperature gas-cooled reactor (HTGR) with block-type fuel, it is necessary to predict the motion and force of core columns and blocks. To reveal column vibration characteristics in three-dimensional space and impact response, column vibration tests were carried out with a scale model of a one-region section (seven columns) of the HTGR core. The results are as follows: (1) the column has a soft spring characteristic based on stacked blocks connected with loose pins, (2) the column has whirling phenomena, (3) the compression spring force simulating the gas pressure has the effect of raising the column resonance frequency, and (4) the vibration behavior of the stacked block column and impact response of the surrounding columns show agreement between experiment and analysis. (author)

Pressure Tube and Pressure Vessel Reactors; certain comparisons

Energy Technology Data Exchange (ETDEWEB)

Margen, P H; Ahlstroem, P E; Pershagen, B

1961-04-15

In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

Pressure Tube and Pressure Vessel Reactors; certain comparisons

International Nuclear Information System (INIS)

Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

1961-04-01

In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

Proposal for a advanced PWR core with adequate characteristics for passive safety concept

International Nuclear Information System (INIS)

Perrotta, Jose Augusto

1999-01-01

This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

Overview of core designs and requirements/criteria for core restraint systems

International Nuclear Information System (INIS)

Sutherland, W.H.

1984-09-01

The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented

Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

International Nuclear Information System (INIS)

Setiyanto; Tukiran Surbakti

2016-01-01

In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

Overview of core designs and requirements/criteria for core restraint systems

International Nuclear Information System (INIS)

Sutherland, W.H.

1984-01-01

The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System is presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented. (author)

A core performance study on an actinide recycling 'zero-sodium-void worth' core

International Nuclear Information System (INIS)

Kawashima, M.; Nakagawa, M.; Yamaoka, M.; Kasahara, F.

1994-01-01

A core performance study was made for an absorber-type parfait core (A-APC) as one of 'Zero-sodium-void-worth' core concepts. This evaluation study pursued different two aspects; one for transuranic (TRU) management strategy, and another for a loss-of-coolant anticipated transient behavior considering the unique core configuration. The results indicated that this core has a large flexibility for actinide recycling in terms of self-sufficiency and minor actinide burning. The result also showed that this core has kept a large mitigation potential for ULOF events as well as a simple flat core concept, reflecting detailed three dimensional core bowing behavior for the A-APC configuration. (author)

Comparison of the electrical activity of trunk core muscles and knee muscles in subjects with and without patellofemoral pain syndrome during gait

Directory of Open Access Journals (Sweden)

Raheleh Dorosti

2017-10-01

Conclusion: It seems that electromyographic activities of some of core muscles in patients with patellofemoral pain syndrome in comparison with healthy subjects are different. However, there was no differences in electromyographic activities in some of the muscles around the knee between patients and healthy subjects.

Undermoderated spectrum MOX core study. Pressurized water-type breeder

International Nuclear Information System (INIS)

Tochihara, Hiroshi; Komano, Yasuo

1998-01-01

The purpose of this development is the advance of the PWR core. The conversion ratio (the breeding ratio) are examined. In the case of heavy water as the coolant, the breeding ratio can be achieved about 1.1 with using a hexagonal lattice and space about 1 mm assembly fuel. In the case of light water coolant, the breeding ratio becomes about 1.0, using a hexagonal lattice and fuel space about 0.5 mm fuel assembly. Here, it reports on the situation of the examination such as the nucleus design of core, the design of fuel assembly, the heat hydraulics design of the core, the structure design and so on. (author)

Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

Energy Technology Data Exchange (ETDEWEB)

Morrey, E.V.; Tingey, J.M.; Elliott, M.L.

1996-10-01

A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

Comparison of simulants to actual neutralized current acid waste: Process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

Energy Technology Data Exchange (ETDEWEB)

Morrey, E.V.; Tingey, J.M.

1996-04-01

A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs have been established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste is being performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

International Nuclear Information System (INIS)

Morrey, E.V.; Tingey, J.M.; Elliott, M.L.

1996-10-01

A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions

Comparison of simulants to actual neutralized current acid waste: Process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

International Nuclear Information System (INIS)

Morrey, E.V.; Tingey, J.M.

1996-04-01

A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs have been established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste is being performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions

Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core

Energy Technology Data Exchange (ETDEWEB)

Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)

2014-10-15

The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.

Studies of mixed HEU-LEU-MTR cores using 3D models

Energy Technology Data Exchange (ETDEWEB)

Haenggi, P.; Lehmann, E.; Hammer, J.; Christen, R. [Paul Scherrer Institute, Villigen (Switzerland)

1997-08-01

Several different core loadings were assembled at the SAPHIR research reactor in Switzerland combining the available types of MTR-type fuel elements, consisting mainly of both HEU and LEU fuel. Bearing in mind the well known problems which can occur in such configurations (especially power peaking), investigations have been carried out for each new loading with a 2D neutron transport code (BOXER). The axial effects were approximated by a global buckling value and therefore the radial effects could be studied in considerably detail. Some of the results were reported at earlier RERTR meetings and were compared to those obtained by other methods and with experimental values. For the explicit study of the third dimension of the core, another code (SILWER), which has been developed in PSI for LWR power plant cores, has been selected. With the help of an adapted model for the MTR-core of SAPHIR, several important questions have been addressed. Among other aspects, the estimation of the axial contribution to the hot channel factors, the influence of the control rod position and of the Xe-poisoning on the power distribution were studied. Special attention was given to a core position where a new element was assumed placed near a empty, water filled position. The comparison of elements of low and high enrichments at this position was made in terms of the induced power peaks, with explicit consideration of axial effects. The program SILWER has proven to be applicable to MTR-cores for the investigation of axial effects. For routine use as for the support of reactor operation, this 3D code is a good supplement to the standard 2D model.

Hepatitis B Core Antigen in Hepatocytes of Chronic Hepatitis B: Comparison between Indirect Immunofluorescence and Immunoperoxidase Method

Science.gov (United States)

Tabassum, Shahina; Al-Mahtab, Mamun; Nessa, Afzalun; Jahan, Munira; Shamim Kabir, Chowdhury Mohammad; Kamal, Mohammad; Cesar Aguilar, Julio

2015-01-01

Background Hepatitis B virus (HBV) infection has many faces. Precore and core promoter mutants resemble inactive carrier status. The identification of hepatitis B core antigen (HBcAg) in hepatocytes may have variable clinical significance. The present study was undertaken to detect HBcAg in chronic hepatitis B (CHB) patients and to assess the efficacy of detection system by indirect immunofluorescence (IIF) and indirect immunoperoxidase (IIP). Materials and methods The study was done in 70 chronic HBV-infected patients. Out of 70 patients, eight (11.4%) were hepatitis B e antigen (HBeAg) positive and 62 (88.57%) were HBeAg negative. Hepatitis B core antigen was detected by indirect immunofluorescence (IIF) and indirect immunoperoxidase (IIP) methods in liver tissue. Results All HBeAg positive patients expressed HBcAg by both IIF and IIP methods. Out of 62 patients with HBeAg-negative CHB, HBcAg was detected by IIF in 55 (88.7%) patients and by IIP in 51 (82.26%) patients. A positive relation among viral load and HBcAg detection was also found. This was more evident in the case of HBeAg negative patients and showed a positive relation with HBV DNA levels. Conclusion Hepatitis B core antigen can be detected using the IIF from formalin fixed paraffin block preparation and also by IIP method. This seems to reflect the magnitudes of HBV replication in CHB. How to cite this article Raihan R, Tabassum S, Al-Mahtab M, Nessa A, Jahan M, Kabir CMS, Kamal M, Aguilar JC. Hepatitis B Core Antigen in Hepatocytes of Chronic Hepatitis B: Comparison between Indirect Immunofluorescence and Immunoperoxidase Method. Euroasian J Hepato-Gastroenterol 2015;5(1):7-10. PMID:29201677

GrowYourIC: an open access Python code to facilitate comparison between kinematic models of inner core evolution and seismic observations

Science.gov (United States)

Lasbleis, M.; Day, E. A.; Waszek, L.

2017-12-01

The complex nature of inner core structure has been well-established from seismic studies, with heterogeneities at various length scales, both radially and laterally. Despite this, no geodynamic model has successfully explained all of the observed seismic features. To facilitate comparisons between seismic observations and geodynamic models of inner core growth we have developed a new, open access Python tool - GrowYourIC - that allows users to compare models of inner core structure. The code allows users to simulate different evolution models of the inner core, with user-defined rates of inner core growth, translation and rotation. Once the user has "grown" an inner core with their preferred parameters they can then explore the effect of "their" inner core's evolution on the relative age and growth rate in different regions of the inner core. The code will convert these parameters into seismic properties using either built-in mineral physics models, or user-supplied ones that calculate these seismic properties with users' own preferred mineralogical models. The 3D model of isotropic inner core properties can then be used to calculate the predicted seismic travel time anomalies for a random, or user-specified, set of seismic ray paths through the inner core. A real dataset of inner core body-wave differential travel times is included for the purpose of comparing user-generated models of inner core growth to actual observed travel time anomalies in the top 100km of the inner core. Here, we explore some of the possibilities of our code. We investigate the effect of the limited illumination of the inner core by seismic waves on the robustness of kinematic model interpretation. We test the impact on seismic differential travel time observations of several kinematic models of inner core growth: fast lateral translation; slow differential growth; and inner core super-rotation. We find that a model of inner core evolution incorporating both differential growth and slow

Refinement Types for TypeScript

OpenAIRE

Vekris, Panagiotis; Cosman, Benjamin; Jhala, Ranjit

2016-01-01

We present Refined TypeScript (RSC), a lightweight refinement type system for TypeScript, that enables static verification of higher-order, imperative programs. We develop a formal core of RSC that delineates the interaction between refinement types and mutability. Next, we extend the core to account for the imperative and dynamic features of TypeScript. Finally, we evaluate RSC on a set of real world benchmarks, including parts of the Octane benchmarks, D3, Transducers, and the TypeScript co...

An analysis of the uniform core experiment

Energy Technology Data Exchange (ETDEWEB)

Waterson, R H

1973-10-15

This report describes an analysis of the Uniform Core of HITREX using the WIMS E codes, and presents the results of theory/experiment comparisons. The overall picture is one of good agreement for core reaction rate distributions, but theory umderestimating k{sub eff} by about 1.5% {delta}k/k.

Effect of the shell material and confinement type on the conversion efficiency of core/shell quantum dot nanocrystal solar cells.

Science.gov (United States)

Sahin, Mehmet

2018-05-23

In this study, the effects of the shell material and confinement type on the conversion efficiency of core/shell quantum dot nanocrystal (QDNC) solar cells have been investigated in detail. For this purpose, the conventional, i.e. original, detailed balance model, developed by Shockley and Queisser to calculate an upper limit for the conversion efficiency of silicon p-n junction solar cells, is modified in a simple and effective way to calculate the conversion efficiency of core/shell QDNC solar cells. Since the existing model relies on the gap energy ([Formula: see text]) of the solar cell, it does not make an estimation about the effect of QDNC materials on the efficiency of the solar cells, and gives the same efficiency values for several QDNC solar cells with the same [Formula: see text]. The proposed modification, however, estimates a conversion efficiency in relation to the material properties and also the confinement type of the QDNCs. The results of the modified model show that, in contrast to the original one, the conversion efficiencies of different QDNC solar cells, even if they have the same [Formula: see text], become different depending upon the confinement type and shell material of the core/shell QDNCs, and this is crucial in the design and fabrication of the new generation solar cells to predict the confinement type and also appropriate QDNC materials for better efficiency.

Effect of the shell material and confinement type on the conversion efficiency of core/shell quantum dot nanocrystal solar cells

Science.gov (United States)

Sahin, Mehmet

2018-05-01

In this study, the effects of the shell material and confinement type on the conversion efficiency of core/shell quantum dot nanocrystal (QDNC) solar cells have been investigated in detail. For this purpose, the conventional, i.e. original, detailed balance model, developed by Shockley and Queisser to calculate an upper limit for the conversion efficiency of silicon p–n junction solar cells, is modified in a simple and effective way to calculate the conversion efficiency of core/shell QDNC solar cells. Since the existing model relies on the gap energy () of the solar cell, it does not make an estimation about the effect of QDNC materials on the efficiency of the solar cells, and gives the same efficiency values for several QDNC solar cells with the same . The proposed modification, however, estimates a conversion efficiency in relation to the material properties and also the confinement type of the QDNCs. The results of the modified model show that, in contrast to the original one, the conversion efficiencies of different QDNC solar cells, even if they have the same , become different depending upon the confinement type and shell material of the core/shell QDNCs, and this is crucial in the design and fabrication of the new generation solar cells to predict the confinement type and also appropriate QDNC materials for better efficiency.

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

Advantages of iron core in a tokamak

International Nuclear Information System (INIS)

Bettis, E.S.; Ballou, J.K.; Becraft, W.R.; Peng, Y.K.M.; Watts, H.L.

1977-01-01

A quantitative comparison of the iron core vs air core concepts was carried out on a preliminary basis by using a representative tokamak reactor design with the following self-consistent reference parameters. In the area of plasma engineering, poloidal field and MHD equilibrium considerations with an unsaturated iron core is discussed. The question of proper poloidal field coils to maintain D-shaped plasmas of relatively high anti β (7%) with a saturated iron core is also discussed. Estimates of the required iron core size, volt seconds, magnetic flux and its influence on force loading on the superconducting toroidal field coils are shown. Conceptual designs of the mechanical structure of an iron core device are presented. Favorable impacts on the OH power supply cost and complexity are indicated

Different timing features in brain processing of core and moral disgust pictures: an event-related potentials study.

Directory of Open Access Journals (Sweden)

Xiangyi Zhang

Full Text Available Disgust, an emotion motivating withdrawal from offensive stimuli, protects us from the risk of biological pathogens and sociomoral violations. Homogeneity of its two types, namely, core and moral disgust has been under intensive debate. To examine the dynamic relationship between them, we recorded event-related potentials (ERPs for core disgust, moral disgust and neutral pictures while participants performed a modified oddball task. ERP analysis revealed that N1 and P2 amplitudes were largest for the core disgust pictures, indicating automatic processing of the core disgust-evoking pictures. N2 amplitudes were higher for pictures evoking moral disgust relative to core disgust and neutral pictures, reflecting a violation of social norms. The core disgust pictures elicited larger P3 and late positive potential (LPP amplitudes in comparison with the moral disgust pictures which, in turn, elicited larger P3 and LPP amplitudes when compared to the neutral pictures. Taken together, these findings indicated that core and moral disgust pictures elicited different neural activities at various stages of information processing, which provided supporting evidence for the heterogeneity of disgust.

Different timing features in brain processing of core and moral disgust pictures: an event-related potentials study.

Science.gov (United States)

Zhang, Xiangyi; Guo, Qi; Zhang, Youxue; Lou, Liandi; Ding, Daoqun

2015-01-01

Disgust, an emotion motivating withdrawal from offensive stimuli, protects us from the risk of biological pathogens and sociomoral violations. Homogeneity of its two types, namely, core and moral disgust has been under intensive debate. To examine the dynamic relationship between them, we recorded event-related potentials (ERPs) for core disgust, moral disgust and neutral pictures while participants performed a modified oddball task. ERP analysis revealed that N1 and P2 amplitudes were largest for the core disgust pictures, indicating automatic processing of the core disgust-evoking pictures. N2 amplitudes were higher for pictures evoking moral disgust relative to core disgust and neutral pictures, reflecting a violation of social norms. The core disgust pictures elicited larger P3 and late positive potential (LPP) amplitudes in comparison with the moral disgust pictures which, in turn, elicited larger P3 and LPP amplitudes when compared to the neutral pictures. Taken together, these findings indicated that core and moral disgust pictures elicited different neural activities at various stages of information processing, which provided supporting evidence for the heterogeneity of disgust.

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Reactor core file

International Nuclear Information System (INIS)

1983-03-01

The first neutronic studies were based on the MSBR project for comparisons with ORNL results. Specific effects depending on the different options are evaluated, such as cross sections in the thorium-U233 cycle, replacement of lithium by sodium, delayed neutron balance, reactivity, kinetics, residual power and recent studies concerning the core with fertile exchange zones. Fuel evolution computation taking into account chemical reprocessing is exposed. Finally different type of lattices are examined. [fr

Cell type-specific sorting of neuropeptides : a mechanism to modulate peptide composition of large dense core vesicles

NARCIS (Netherlands)

Klumperman, J.; Spijker, S.; Minnen, J. van; Sharp-Baker, H.; Smit, A.B.; Geraerts, W.P.M.

1996-01-01

The CNS of Lymnaea stagnalis contains two populations of egg-laying hormone (ELH)-producing neurons that differ in size and topology. In type I neurons, all peptides located C-terminally from the cleavage site Arg-Ser-Arg-Arg180-183 are sorted into secretory large dense-core vesicles (LDCV), whereas

Comparison of ompP5 sequence-based typing and pulsed-filed gel ...

African Journals Online (AJOL)

In this study, comparison of the outer membrane protein P5 gene (ompP5) sequence-based typing with pulsed-field gel electrophoresis (PFGE) for the genotyping of Haemophilus parasuis, the 15 serovar reference strains and 43 isolates were investigated. When comparing the two methods, 31 ompP5 sequence types ...

Design Report for the core design of the first core Mark-Ia of the SNR-300

International Nuclear Information System (INIS)

Stanculescu, A.

1984-05-01

The report describes the first core Mark-Ia of the SNR-300 reactor and its different assembly types with their operational strategy. Methods, criteria and results of the neutron physical, thermal hydraulic and core mechanical design of the whole core and its assemblies are presented

Biogeographic Comparison of Lophelia-Associated Bacterial Communities in the Western Atlantic Reveals Conserved Core Microbiome

Directory of Open Access Journals (Sweden)

Christina A. Kellogg

2017-05-01

Full Text Available Over the last decade, publications on deep-sea corals have tripled. Most attention has been paid to Lophelia pertusa, a globally distributed scleractinian coral that creates critical three-dimensional habitat in the deep ocean. The bacterial community associated with L. pertusa has been previously described by a number of studies at sites in the Mediterranean Sea, Norwegian fjords, off Great Britain, and in the Gulf of Mexico (GOM. However, use of different methodologies prevents direct comparisons in most cases. Our objectives were to address intra-regional variation and to identify any conserved bacterial core community. We collected samples from three distinct colonies of L. pertusa at each of four locations within the western Atlantic: three sites within the GOM and one off the east coast of the United States. Amplicon libraries of 16S rRNA genes were generated using primers targeting the V4–V5 hypervariable region and 454 pyrosequencing. The dominant phylum was Proteobacteria (75–96%. At the family level, 80–95% of each sample was comprised of five groups: Pirellulaceae, Pseudonocardiaceae, Rhodobacteraceae, Sphingomonadaceae, and unclassified Oceanospirillales. Principal coordinate analysis based on weighted UniFrac distances showed a clear distinction between the GOM and Atlantic samples. Interestingly, the replicate samples from each location did not always cluster together, indicating there is not a strong site-specific influence. The core bacterial community, conserved in 100% of the samples, was dominated by the operational taxonomic units of genera Novosphingobium and Pseudonocardia, both known degraders of aromatic hydrocarbons. The sequence of another core member, Propionibacterium, was also found in prior studies of L. pertusa from Norway and Great Britain, suggesting a role as a conserved symbiont. By examining more than 40,000 sequences per sample, we found that GOM samples were dominated by the identified conserved core

Core antigen and circulating anti-core antibody in hepatitis B infection

Energy Technology Data Exchange (ETDEWEB)

Howard, C R; Zuckerman, A J [London School of Hygiene and Tropical Medicine (UK)

1977-02-01

Core antigen was obtained from the sera of persistent chronic carriers of hepatitis B virus by centrifugation and treatment with Nonidet P40 and 2-mercaptoethanol. The separated core antigen was radiolabelled and identified as a nucleoprotein structure of buoyant density 1.36 g/cm/sup 3/ and possessing an isoelectric point of 4.4. This material was employed in a radioimmnoassay procedure of high sensitivity for the detection of core antibody. In a series of sera from patients with acute type B hepatitis, core antibody was demonstrated 2 to 3 weeks after the onset of jaundice during the period of surface antigenaemia. The presence of core antibody may therefore provide an accurate serological marker for the detection of active or recent virus replication in future epidemiological studies of hepatitis B infection.

PWR core design calculations

Energy Technology Data Exchange (ETDEWEB)

Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

1992-07-01

Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

WIMS/PANTHER analysis of UO2/MOX cores using embedded super-cells

International Nuclear Information System (INIS)

Knight, M.; Bryce, P.; Hall, S.

2012-01-01

This paper describes a method of analysing PWR UO 2 MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

A comparison of measured radionuclide release rates from Three Mile Island Unit-2 core debris for different oxygen chemical potentials

International Nuclear Information System (INIS)

Baston, V.F.; Hofstetter, K.J.; Ryan, R.F.

1987-01-01

Chemical and radiochemical analyses of reactor coolant samples taken during defueling of the Three Mile Island Unit-2 (TMI-2) reactor provide relevant data to assist in understanding the solution chemistry of the radionuclides retained within the TMI-2 reactor coolant system. Hydrogen peroxide was added to various plant systems to provide disinfection for microbial contamination and has provided the opportunity to observe radionuclide release under different oxygen chemical potentials. A comparison of the radionuclide release rates with and without hydrogen peroxide has been made for these separate but related cases, i.e., the fuel transfer canal and connecting spent-fuel pool A with the TMI-2 reactor plenum in the fuel transfer canal, core debris grab sample laboratory experiments, and the reactor vessel fluid and associated core debris. Correlation and comparison of these data indicate a physical parameter dependence (surface-to-volume ratio) affecting all radionuclide release; however, selected radionuclides also demonstrate a chemical dependence release under the different oxygen chemical potentials. Chemical and radiochemical analyses of reactor coolant samples taken during defueling of the Three Mile Island Unit-2 (TMI-2) reactor provide relevant data to assist in understanding the solution chemistry of the radionuclides retained within the TMI-2 reactor coolant system

Comparison of computational performance of GA and PSO optimization techniques when designing similar systems - Typical PWR core case

Energy Technology Data Exchange (ETDEWEB)

Souza Lima, Carlos A. [Instituto de Engenharia Nuclear - Divisao de Reatores/PPGIEN, Rua Helio de Almeida 75, Cidade Universitaria - Ilha do Fundao, P.O. Box: 68550 - Zip Code: 21941-972, Rio de Janeiro (Brazil); Instituto Politecnico, Universidade do Estado do Rio de Janeiro, Pos-Graduacao em Modelagem Computacional, Rua Alberto Rangel - s/n, Vila Nova, Nova Friburgo, Zip Code: 28630-050, Nova Friburgo (Brazil); Lapa, Celso Marcelo F.; Pereira, Claudio Marcio do N.A. [Instituto de Engenharia Nuclear - Divisao de Reatores/PPGIEN, Rua Helio de Almeida 75, Cidade Universitaria - Ilha do Fundao, P.O. Box: 68550 - Zip Code: 21941-972, Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (INCT) (Brazil); Cunha, Joao J. da [Eletronuclear Eletrobras Termonuclear - Gerencia de Analise de Seguranca Nuclear, Rua da Candelaria, 65, 7 andar. Centro, Zip Code: 20091-906, Rio de Janeiro (Brazil); Alvim, Antonio Carlos M. [Universidade Federal do Rio de Janeiro, COPPE/Nuclear, Cidade Universitaria - Ilha do Fundao s/n, P.O.Box 68509 - Zip Code: 21945-970, Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (INCT) (Brazil)

2011-06-15

Research highlights: > Performance of PSO and GA techniques applied to similar system design. > This work uses ANGRA1 (two loop PWR) core as a prototype. > Results indicate that PSO technique is more adequate than GA to solve this kind of problem. - Abstract: This paper compares the performance of two optimization techniques, particle swarm optimization (PSO) and genetic algorithm (GA) applied to the design a typical reduced scale two loop Pressurized Water Reactor (PWR) core, at full power in single phase forced circulation flow. This comparison aims at analyzing the performance in reaching the global optimum, considering that both heuristics are based on population search methods, that is, methods whose population (candidate solution set) evolve from one generation to the next using a combination of deterministic and probabilistic rules. The simulated PWR, similar to ANGRA 1 power plant, was used as a case example to compare the performance of PSO and GA. Results from simulations indicated that PSO is more adequate to solve this kind of problem.

Manufacture of the second core of Rapsodie - Type 2 A - Fortissimo version

International Nuclear Information System (INIS)

Mandicourt, Louis

1977-11-01

This report concerns the fuel manufacturing of the second core of RAPSODIE reactor 2 A type, FORTISSIMO version. This fuel is a mixed oxide of Plutonium and highly enriched Uranium. The manufacturing facility which is used, set starting from 1969, is essentially designed for the production of fuels with high percentage of fissile material. It includes devices having been used to produce the first RAPSODIE fuel and new machines resulting from technological improvements obtained during previous manufactures. The manufacturing process is described after a recall fuel physical characteristics. A final quantitative balance for the whole fabrication is presented. The results obtained on fuel pellets and pins are discussed with starting from the fabrication process and the changes which may have occured or arised from the use of different devices. A more theoretical approach of some particular problems is presented: the multiphasage, the amount of oxygen and the porosities. A general conclusion is given on the main theoretical and technical problems met up to now on the manufacture of nuclear fuel of this type. [fr

Contribution to the build-up of a core calculation frame: comparison between ``diffusion`` and ``SP{sub n}`` operators on various configurations of the first N4 core; Contribution a l`elaboration d`un schema de coeur en transport: comparaison des operateurs ``diffusion`` et ``SP{sub n}`` sur differentes configurations du premier coeur N4

Energy Technology Data Exchange (ETDEWEB)

Magat, Ph

1997-04-01

The aim of this study is to compare two calculation methods implemented in the neutronic code CRONOS 2: the diffusion approximation and the SP{sub n} method. The APOLLO 2 code is used to build the multiparameter cross section libraries.The comparison is based on the first core of N4 type Chooz reactor. The rod worth and the power map have been calculated. Some recommendations about the SP{sub n} development order of flux are made and the results show that the diffusion calculations over-estimate the black rod efficiency up to 10%. (A.C.) 12 refs.

A common and optimized age scale for Antarctic ice cores

Science.gov (United States)

Parrenin, F.; Veres, D.; Landais, A.; Bazin, L.; Lemieux-Dudon, B.; Toye Mahamadou Kele, H.; Wolff, E.; Martinerie, P.

2012-04-01

Dating ice cores is a complex problem because 1) there is a age shift between the gas bubbles and the surrounding ice 2) there are many different ice cores which can be synchronized with various proxies and 3) there are many methods to date the ice and the gas bubbles, each with advantages and drawbacks. These methods fall into the following categories: 1) Ice flow (for the ice) and firn densification modelling (for the gas bubbles); 2) Comparison of ice core proxies with insolation variations (so-called orbital tuning methods); 3) Comparison of ice core proxies with other well dated archives; 4) Identification of well-dated horizons, such as tephra layers or geomagnetic anomalies. Recently, an new dating tool has been developped (DATICE, Lemieux-Dudon et al., 2010), to take into account all the different dating information into account and produce a common and optimal chronology for ice cores with estimated confidence intervals. In this talk we will review the different dating information for Antarctic ice cores and show how the DATICE tool can be applied.

A calculation model for a HTR core seismic response

International Nuclear Information System (INIS)

Buland, P.; Berriaud, C.; Cebe, E.; Livolant, M.

1975-01-01

The paper presents the experimental results obtained at Saclay on a HTGR core model and comparisons with analytical results. Two series of horizontal tests have been performed on the shaking table VESUVE: sinusoidal test and time history response. Acceleration of graphite blocks, forces on the boundaries, relative displacement of the core and PCRB model, impact velocity of the blocks on the boundaries were recorded. These tests have shown the strongly non-linear dynamic behaviour of the core. The resonant frequency of the core is dependent on the level of the excitation. These phenomena have been explained by a computer code, which is a lumped mass non-linear model. Good correlation between experimental and analytical results was obtained for impact velocities and forces on the boundaries. This comparison has shown that the damping of the core is a critical parameter for the estimation of forces and velocities. Time history displacement at the level of PCRV was reproduced on the shaking table. The analytical model was applied to this excitation and good agreement was obtained for forces and velocities. (orig./HP) [de

Matérn's hard core models of types I and II with arbitrary compact grains

DEFF Research Database (Denmark)

Kiderlen, Markus; Hörig, Mario

Matérn's classical hard core models can be interpreted as models obtained from a stationary marked Poisson process by dependent thinning. The marks are balls of fixed radius, and a point is retained when its associated ball does not hit any other balls (type I) or when its random birth time is st...... of this model with the process of intact grains of the dead leaves model and the Stienen model leads to analogous results for the latter....

Evaluating nuclear physics inputs in core-collapse supernova models

Science.gov (United States)

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

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

A tetrapyridine ligand with a rigid tetrahedral core forms metal-organic frameworks with PtS type architecture.

Science.gov (United States)

Caputo, Christopher B; Vukotic, V Nicholas; Sirizzotti, Natalie M; Loeb, Stephen J

2011-08-14

A new tetradentate, pyridine ligand with a rigid tetrahedral core can be prepared in good yield by a cross-coupling methodology. Two metal organic framework structures of Cu(II) with PtS-type topology having a carbon atom as the tetrahedral node have been characterized utilising this ligand. This journal is © The Royal Society of Chemistry 2011

Nonlinear core deflection in injection molding

Science.gov (United States)

Poungthong, P.; Giacomin, A. J.; Saengow, C.; Kolitawong, C.; Liao, H.-C.; Tseng, S.-C.

2018-05-01

Injection molding of thin slender parts is often complicated by core deflection. This deflection is caused by molten plastics race tracking through the slit between the core and the rigid cavity wall. The pressure of this liquid exerts a lateral force of the slender core causing the core to bend, and this bending is governed by a nonlinear fifth order ordinary differential equation for the deflection that is not directly in the position along the core. Here we subject this differential equation to 6 sets of boundary conditions, corresponding to 6 commercial core constraints. For each such set of boundary conditions, we develop an explicit approximate analytical solution, including both a linear term and a nonlinear term. By comparison with finite difference solutions, we find our new analytical solutions to be accurate. We then use these solutions to derive explicit analytical approximations for maximum deflections and for the core position of these maximum deflections. Our experiments on the base-gated free-tip boundary condition agree closely with our new explicit approximate analytical solution.

Transmission properties of hollow-core photonic bandgap fibers

DEFF Research Database (Denmark)

Falk, Charlotte Ijeoma; Hald, Jan; Petersen, Jan C.

2010-01-01

Variations in optical transmission of four types of hollow-core photonic bandgap fibers are measured as a function of laser frequency. These variations influence the potential accuracy of gas sensors based on molecular spectroscopy in hollow-core fibers.......Variations in optical transmission of four types of hollow-core photonic bandgap fibers are measured as a function of laser frequency. These variations influence the potential accuracy of gas sensors based on molecular spectroscopy in hollow-core fibers....

On reactor type comparisons for the next generation of reactors

International Nuclear Information System (INIS)

Alesso, H.P.; Majumdar, K.C.

1991-01-01

In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs

Effect of welding process, type of electrode and electrode core diameter on the tensile property of 304L austenitic stainless steel

Directory of Open Access Journals (Sweden)

Akinlabi OYETUNJI

2014-11-01

Full Text Available The effect of welding process, type of electrode and electrode core diameter on the tensile property of AISI 304L Austenitic Stainless Steel (ASS was studied. The tensile strength property of ASS welded samples was evaluated. Prepared samples of the ASS were welded under these three various variables. Tensile test was then carried out on the welded samples. It was found that the reduction in ultimate tensile strength (UTS of the butt joint samples increases with increase in core diameter of the electrode. Also, the best electrode for welding 304L ASS is 308L stainless steel-core electrode of 3.2 mm core diameter. It is recommended that the findings of this work can be applied in the chemical, food and oil industries where 304L ASS are predominantly used.

Inventories of radioactive fission products in the core of thermal nuclear reactor

International Nuclear Information System (INIS)

Marinkovic, N.

1977-01-01

As a part of the analysis concerning radiological consequences of a major LWR accident, inventories of the most significant radioactive nuclides and stable fission gases in the core of a PWR type reactor have been calculated. Calculations were performed by the DELFIN code using nuclide data and neutron flux data earlier obtained by the METHUSELAH code. Comparison with simplified calculation method show that it is quite rough for certain nuclides but the accuracy may be sufficient for safety analysis purposes recalling the inaccuracies in the later parts of fission product transport process (author)

Evidence for core-coupled states in 87Y from a 89Y(p, t)87Y and 88Sr(p, t)86Sr comparison

International Nuclear Information System (INIS)

Oelrich, I.C.; Krien, K.; DelVecchio, R.M.; Naumann, R.A.

1976-01-01

The 89 Y(p, t) 87 Y and 88 Sr(p, t) 86 Sr reactions were studied at 42 MeV proton energy, using a quadrupole-dipole-dipole-dipole spectograph. Comparison of excitation energies, (p, t) cross section strengths and angular distribution shapes indicates that basis features of the core-coupling model apply to these nuclei. However, mixing of single particle states with the core-coupled states is evident. The (p, t) cross-section strength summed over the 87 Y multiplet is found with few exceptions to be nearly a constant multiple of the (p, t) strength of the associated 86 Sr state

Core-shell architectures as nano-size transporters

International Nuclear Information System (INIS)

Adeli, M.; Zarnegar, Z.; Kabiri, R.; Salimi, F.; Dadkah, A.

2006-01-01

Core-shell architectures containing poly (ethylene imine) (PEI) as a core and poly (lactide) (PLA) as arms were prepared. PEI was used as macro initiator for ring opening polymerization of lactide. PEI-PLA core-shell architectures were able to encapsulate guest molecules. Size of the core-shell architectures was between 10- 100 nm, hence they can be considered as nano carriers to transport the guest molecules. Transport capacity of nano carriers depends on their nano-environments and type of self-assembly in solvent. In solid state nano carriers self-assemble as long structures with nano-size diameter or they form network structures. Aggregations type depends on the concentration of nano carriers in solution. Effect of the shell thickness and aggregation type on the release rate are also investigated

Dislocation core structures in (0001) InGaN

International Nuclear Information System (INIS)

Rhode, S. L.; Sahonta, S.-L.; Kappers, M. J.; McAleese, C.; Humphreys, C. J.; Horton, M. K.; Haigh, S. J.; Pennycook, T. J.; Dusane, R. O.; Moram, M. A.

2016-01-01

Threading dislocation core structures in c-plane GaN and In x Ga 1−x N (0.057 ≤ x ≤ 0.20) films were investigated by aberration-corrected scanning transmission electron microscopy. a-type dislocations are unaffected by alloying with indium and have a 5/7-atom ring core structure in both GaN and In x Ga 1−x N. In contrast, the dissociation lengths of (a + c)-type dislocations are reduced, and new 7/4/9-atom ring and 7/4/8/5-atom ring core structures were observed for the dissociated (a + c)-type dislocations in In x Ga 1−x N, which is associated with the segregation of indium near (a + c)-type and c-type dislocation cores in In x Ga 1−x N, consistent with predictions from atomistic Monte Carlo simulations.

Does core mobility of lumbar total disc arthroplasty influence sagittal and frontal intervertebral displacement? Radiologic comparison with fixed-core prosthesis

Science.gov (United States)

Delécrin, Joël; Allain, Jérôme; Beaurain, Jacques; Steib, Jean-Paul; Chataigner, Hervé; Aubourg, Lucie; Huppert, Jean; Ameil, Marc; Nguyen, Jean-Michel

2009-01-01

Background An artificial disc prosthesis is thought to restore segmental motion in the lumbar spine. However, it is reported that disc prosthesis can increase the intervertebral translation (VT). The concept of the mobile-core prosthesis is to mimic the kinematic effects of the migration of the natural nucleus and therefore core mobility should minimize the VT. This study explored the hypothesis that core translation should influence VT and that a mobile core prosthesis may facilitate physiological motion. Methods Vertebral translation (measured with a new method presented here), core translation, range of motion (ROM), and distribution of flexion-extension were measured on flexion-extension, neutral standing, and lateral bending films in 89 patients (63 mobile-core [M]; 33 fixed-core [F]). Results At L4-5 levels the VT with M was lower than with F and similar to the VT of untreated levels. At L5-S1 levels the VT with M was lower than with F but was significantly different compared to untreated levels. At M levels a strong correlation was found between VT and core translation; the VT decreases as the core translation increases. At F levels the VT increases as the ROM increases. No significant difference was found between the ROM of untreated levels and levels implanted with either M or F. Regarding the mobility distribution with M and F we observed a deficit in extension at L5-S1 levels and a similar distribution at L4-5 levels compared to untreated levels. Conclusion The intervertebral mobility was different between M and F. The M at L4-5 levels succeeded to replicate mobility similar to L4-5 untreated levels. The M at L5-S1 succeeded in ROM, but failed regarding VT and mobility distribution. Nevertheless M minimized VT at L5-S1 levels. The F increased VT at both L4-5 and L5-S1. Clinical Relevance This study validates the concept that the core translation of an artificial lumbar disc prosthesis minimizes the VT. PMID:25802632

Reactor core

International Nuclear Information System (INIS)

Azekura, Kazuo; Kurihara, Kunitoshi.

1992-01-01

In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

Humidity effects on soluble core mechanical and thermal properties (polyvinyl alcohol/microballoon composite) type CG extendospheres, volume 2

Science.gov (United States)

1993-01-01

This document constitutes the final report for the study of humidity effects and loading rate on soluble core (PVA/MB composite material) mechanical and thermal properties under Contract No. 100345. This report describes test results procedures employed, and any unusual occurrences or specific observations associated with this test program. The primary objective of this work was to determine if cured soluble core filler material regains its tensile and compressive strength after exposure to high humidity conditions and following a drying cycle. Secondary objectives include measurements of tensile and compressive modulus, and Poisson's ratio, and coefficient of thermal expansion (CTE) for various moisture exposure states. A third objective was to compare the mechanical and thermal properties of the composite using 'SG' and 'CG' type extendospheres. The proposed facility for the manufacture of soluble cores at the Yellow Creek site incorporates no capability for the control of humidity. Recent physical property tests performed with the soluble core filler material showed that prolonged exposure to high humidity significantly degradates in strength. The purpose of these tests is to determine if the product, process or facility designs require modification to avoid imparting a high risk condition to the ASRM.

Performance comparisons of dish type solar concentrator with mirror arrangements and receiver shapes

Energy Technology Data Exchange (ETDEWEB)

Seo, Joo Hyun; Kim, Yong; Ma, Dae Sung; Seo, Tae Beom [Graduate School, Dept. of Mechanical Engineering, Inha Univ., Inchon (Korea, Republic of); Kang, Yong Heack [Korea Inst. of Energy Research, Daejeon (Korea, Republic of)

2008-07-01

The performance comparisons of dish type solar concentrators are numerically investigated. The dish type solar concentrator considered in this paper consists of a receiver and multi-faceted mirrors. In order to investigate the performance comparisons of dish type solar concentrators, six different mirror arrangements and four different receivers are considered. A parabolic-shaped perfect mirror of which diameter is 1.40 m is considered as the reference for the mirror arrangements. The other mirror arrangements consist of twelve identical parabolic-shaped mirror facets of which diameter are 0.405 m. Their total collecting areas, which are 1.545 m{sup 2}, are the same. Four different solar receiver shapes are a conical, a dome, a cylindrical, and a unicorn type. In order to investigate the thermal performance of the dish type solar concentrator, the radiative heat loss in the receiver should be calculated. For calculation, the net radiation method and the Monte-Carlo method are used. Also, because the thermal performance of the dish type solar concentrator can vary as the receiver surface temperature, the various surface temperatures are considered. Based on the calculation, the unicorn type has the best performance in receiver shapes and the STAR has the best performance in mirror arrangements except the perfect mirror. (orig.)

SUPERNOVA 2003ie WAS LIKELY A FAINT TYPE IIP EVENT

Energy Technology Data Exchange (ETDEWEB)

Arcavi, Iair; Gal-Yam, Avishay [Department of Particle Physics and Astrophysics, The Weizmann Institute of Science, Rehovot 76100 (Israel); Sergeev, Sergey G., E-mail: iair.arcavi@weizmann.ac.il [Crimean Astrophysical Observatory, P/O Nauchny, Crimea 98409 (Ukraine)

2013-04-15

We present new photometric observations of supernova (SN) 2003ie starting one month before discovery, obtained serendipitously while observing its host galaxy. With only a weak upper limit derived on the mass of its progenitor (<25 M{sub Sun }) from previous pre-explosion studies, this event could be a potential exception to the ''red supergiant (RSG) problem'' (the lack of high-mass RSGs exploding as Type IIP SNe). However, this is true only if SN2003ie was a Type IIP event, something which has never been determined. Using recently derived core-collapse SN light-curve templates, as well as by comparison to other known SNe, we find that SN2003ie was indeed a likely Type IIP event. However, with a plateau magnitude of {approx} - 15.5 mag, it is found to be a member of the faint Type IIP class. Previous members of this class have been shown to arise from relatively low-mass progenitors (<12 M{sub Sun }). It therefore seems unlikely that this SN had a massive RSG progenitor. The use of core-collapse SN light-curve templates is shown to be helpful in classifying SNe with sparse coverage. These templates are likely to become more robust as large homogeneous samples of core-collapse events are collected.

Study on high conversion type core of innovative water reactor for flexible fuel cycle (FLWR) for minor actinide (MA) recycling

International Nuclear Information System (INIS)

Fukaya, Yuji; Nakano, Yoshihiro; Okubo, Tsutomu

2009-01-01

In order to ensure sustainable energy supplies in the future based on the well-established light water reactor (LWR) technologies, conceptual design studies have been performed on the innovative water reactor for flexible fuel cycle (FLWR) with the high conversion ratio core. For early introduction of FLWR without a serious technical gap from the LWR technologies, the conceptual design of the high conversion type one (HC-FLWR) was constructed to recycle reprocessed plutonium. Furthermore, an investigation of minor actinide (MA) recycling based on the HC-FLWR core concept has been performed and is presented in this paper. Because HC-FLWR is a near-term technology, it would be a good option in the future if HC-FLWR can recycle MAs. In order to recycle MAs in HC-FLWR, it has been found that the core design should be changed, because the loaded MA makes the void reactivity coefficient worse and decreases the discharge burn-up. To find a promising core design specification, the investigation on the core characteristics were performed using the results from parameter surveys with core burn-up calculations. The final core designs were established by coupled three dimensional neutronics and thermal-hydraulics core calculations. The major core specifications are as follows. The plutonium fissile (Puf) content is 13 wt%. The discharge burn-up is about 55 GWd/t. Around 2 wt% of Np or Am can be recycled. The MA conversion ratios are around unity. In particular, it has been found that loaded Np can be transmuted effectively in this core concept. Therefore, these concepts would be a good option to reduce environmental burdens.

Dislocation core structures in Si-doped GaN

International Nuclear Information System (INIS)

Rhode, S. L.; Fu, W. Y.; Sahonta, S.-L.; Kappers, M. J.; Humphreys, C. J.; Horton, M. K.; Pennycook, T. J.; Dusane, R. O.; Moram, M. A.

2015-01-01

Aberration-corrected scanning transmission electron microscopy was used to investigate the core structures of threading dislocations in plan-view geometry of GaN films with a range of Si-doping levels and dislocation densities ranging between (5 ± 1) × 10 8  and (10 ± 1) × 10 9  cm −2 . All a-type (edge) dislocation core structures in all samples formed 5/7-atom ring core structures, whereas all (a + c)-type (mixed) dislocations formed either double 5/6-atom, dissociated 7/4/8/4/9-atom, or dissociated 7/4/8/4/8/4/9-atom core structures. This shows that Si-doping does not affect threading dislocation core structures in GaN. However, electron beam damage at 300 keV produces 4-atom ring structures for (a + c)-type cores in Si-doped GaN

Dislocation core structures in Si-doped GaN

Energy Technology Data Exchange (ETDEWEB)

Rhode, S. L., E-mail: srhode@imperial.ac.uk; Fu, W. Y.; Sahonta, S.-L.; Kappers, M. J.; Humphreys, C. J. [Department of Materials Science and Metallurgy, University of Cambridge, Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Horton, M. K. [Department of Materials, Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom); Pennycook, T. J. [SuperSTEM, STFC Daresbury Laboratories, Warrington WA4 4AD (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Dusane, R. O. [Department of Metallurgical Engineering and Materials Science, Indian Institute of Technology Bombay, Mumbai 400076 (India); Moram, M. A. [Department of Materials Science and Metallurgy, University of Cambridge, Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Department of Materials, Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom)

2015-12-14

Aberration-corrected scanning transmission electron microscopy was used to investigate the core structures of threading dislocations in plan-view geometry of GaN films with a range of Si-doping levels and dislocation densities ranging between (5 ± 1) × 10{sup 8} and (10 ± 1) × 10{sup 9} cm{sup −2}. All a-type (edge) dislocation core structures in all samples formed 5/7-atom ring core structures, whereas all (a + c)-type (mixed) dislocations formed either double 5/6-atom, dissociated 7/4/8/4/9-atom, or dissociated 7/4/8/4/8/4/9-atom core structures. This shows that Si-doping does not affect threading dislocation core structures in GaN. However, electron beam damage at 300 keV produces 4-atom ring structures for (a + c)-type cores in Si-doped GaN.

WIMS/PANTHER analysis of UO{sub 2}/MOX cores using embedded super-cells

Energy Technology Data Exchange (ETDEWEB)

Knight, M.; Bryce, P. [EDF Energy, Barnett Way, Barnwood, Gloucester (United Kingdom); Hall, S. [Advanced Modelling and Computation Group, Imperial College, London (United Kingdom)

2012-07-01

This paper describes a method of analysing PWR UO{sub 2}MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

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

Directory of Open Access Journals (Sweden)

Bogna Stawarczyk

2016-06-01

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

Dislocation core structures in (0001) InGaN

Energy Technology Data Exchange (ETDEWEB)

Rhode, S. L.; Sahonta, S.-L.; Kappers, M. J.; McAleese, C.; Humphreys, C. J. [Department of Materials Science and Metallurgy, University of Cambridge, Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Horton, M. K. [Department of Materials, Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom); Haigh, S. J. [School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); SuperSTEM, STFC Daresbury Laboratories, Warrington WA4 4AD (United Kingdom); Pennycook, T. J. [SuperSTEM, STFC Daresbury Laboratories, Warrington WA4 4AD (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Dusane, R. O. [Department of Metallurgical Engineering and Materials Science, Indian Institute of Technology Bombay, Mumbai 400076 (India); Moram, M. A. [Department of Materials Science and Metallurgy, University of Cambridge, Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Department of Materials, Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom)

2016-03-14

Threading dislocation core structures in c-plane GaN and In{sub x}Ga{sub 1−x}N (0.057 ≤ x ≤ 0.20) films were investigated by aberration-corrected scanning transmission electron microscopy. a-type dislocations are unaffected by alloying with indium and have a 5/7-atom ring core structure in both GaN and In{sub x}Ga{sub 1−x}N. In contrast, the dissociation lengths of (a + c)-type dislocations are reduced, and new 7/4/9-atom ring and 7/4/8/5-atom ring core structures were observed for the dissociated (a + c)-type dislocations in In{sub x}Ga{sub 1−x}N, which is associated with the segregation of indium near (a + c)-type and c-type dislocation cores in In{sub x}Ga{sub 1−x}N, consistent with predictions from atomistic Monte Carlo simulations.

Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

International Nuclear Information System (INIS)

Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

2009-01-01

In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

Core physics design calculation of mini-type fast reactor based on Monte Carlo method

International Nuclear Information System (INIS)

He Keyu; Han Weishi

2007-01-01

An accurate physics calculation model has been set up for the mini-type sodium-cooled fast reactor (MFR) based on MCNP-4C code, then a detailed calculation of its critical physics characteristics, neutron flux distribution, power distribution and reactivity control has been carried out. The results indicate that the basic physics characteristics of MFR can satisfy the requirement and objectives of the core design. The power density and neutron flux distribution are symmetrical and reasonable. The control system is able to make a reliable reactivity balance efficiently and meets the request for long-playing operation. (authors)

Natural vibrations of a core banel of a PWR type reactor by elements of revolution shell

International Nuclear Information System (INIS)

Barcellos, C.S. de.

1980-01-01

Aim to estimate the behavior of the cove barrel of PWR type reactors, submitted to several load conditions, their dynamic characteristic, were determined. In order to obtain the natural modes and frequencies of the core barrel, the CYLDYFE comprete code based in the finite element method, was developed. The obtained results are compared with results obtained by other programs such as SAP, ASKA and STRUDL/DYNAL and by other analytical methods. (M.C.K.) [pt

Core/shell-type nanorods of Tb{sup 3+}-doped LaPO{sub 4}, modified with amine groups, revealing reduced cytotoxicity

Energy Technology Data Exchange (ETDEWEB)

Runowski, Marcin [Adam Mickiewicz University, Department of Rare Earths, Faculty of Chemistry (Poland); Dąbrowska, Krystyna [Polish Academy of Sciences, Bacteriophage Laboratory, Institute of Immunology and Experimental Therapy (Poland); Grzyb, Tomasz [Adam Mickiewicz University, Department of Rare Earths, Faculty of Chemistry (Poland); Miernikiewicz, Paulina [Polish Academy of Sciences, Bacteriophage Laboratory, Institute of Immunology and Experimental Therapy (Poland); Lis, Stefan, E-mail: blis@amu.edu.pl [Adam Mickiewicz University, Department of Rare Earths, Faculty of Chemistry (Poland)

2013-11-15

A simple co-precipitation reaction between Ln{sup 3+} cations (Ln = lanthanide) and phosphate ions in the presence of polyethylene glycol (PEG), including post-treatment under hydrothermal conditions, leads to the formation of Tb{sup 3+}-doped LaPO{sub 4} crystalline nanorods. The nanoparticles obtained can be successfully coated with amorphous and porous silica, forming core/shell-type nanorods. Both products reveal intensive green luminescence under UV lamp irradiation. The surface of the core/shell-type product can also be modified with –NH{sub 2} groups via silylation procedure, using 3-aminopropyltriethoxysilane as a modifier. Powder X-ray diffraction, transmission electron microscopy, and scanning electron microscopy confirm the desired structure and needle-like shape of the products synthesized. Fourier transform infrared spectroscopy and specific surface area measurements by Brunauer–Emmett–Teller method reveal a successful surface modification with amine groups of the core/shell-type nanoparticles prepared. The nanomaterials synthesized exhibit green luminescence characteristic of Tb{sup 3+} ions, as solid powders and aqueous colloids, examined by spectrofluorometry. The in vitro cytotoxicity studies reveal different degree toxicity of the products. LaPO{sub 4}:Tb{sup 3+}@SiO{sub 2}@NH{sub 2} exhibits the smallest toxicity against B16F0 mouse melanoma cancer cells and human skin microvascular endothelial cell lines, in contrast to the most toxic LaPO{sub 4}:Tb{sup 3+}@SiO{sub 2}.Graphical Abstract.

DENSE CORES IN THE PIPE NEBULA: AN IMPROVED CORE MASS FUNCTION

International Nuclear Information System (INIS)

Rathborne, J. M.; Lada, C. J.; Muench, A. A.; Alves, J. F.; Kainulainen, J.; Lombardi, M.

2009-01-01

In this paper, we derive an improved core mass function (CMF) for the Pipe Nebula from a detailed comparison between measurements of visual extinction and molecular-line emission. We have compiled a refined sample of 201 dense cores toward the Pipe Nebula using a two-dimensional threshold identification algorithm informed by recent simulations of dense core populations. Measurements of radial velocities using complimentary C 18 O (1-0) observations enable us to cull out from this sample those 43 extinction peaks that are either not associated with dense gas or are not physically associated with the Pipe Nebula. Moreover, we use the derived C 18 O central velocities to differentiate between single cores with internal structure and blends of two or more physically distinct cores, superposed along the same line of sight. We then are able to produce a more robust dense core sample for future follow-up studies and a more reliable CMF than was possible previously. We confirm earlier indications that the CMF for the Pipe Nebula departs from a single power-law-like form with a break or knee at M ∼ 2.7 ± 1.3 M sun . Moreover, we also confirm that the CMF exhibits a similar shape to the stellar initial mass function (IMF), but is scaled to higher masses by a factor of ∼4.5. We interpret this difference in scaling to be a measure of the star formation efficiency (22% ± 8%). This supports earlier suggestions that the stellar IMF may originate more or less directly from the CMF.

F.B.R. Core mock-up RAPSODIE - II - numerical models

International Nuclear Information System (INIS)

Brochard, D.; Hammami, L.; Gantenbein, F.

1990-01-01

To study the behaviour of LMFBR cores excited by a seism, tests have been performed on the RAPSODIE core mock-up. The aim of this paper is to present the numerical models used to interprete these tests and the comparisons between calculations and experimental results

Electrical conductivity measurements from the GISP2 and GRIP Greenland ice cores

DEFF Research Database (Denmark)

Dahl-Jensen, Dorthe; Clausen, Henrik Brink; Taylor, K. C.

1993-01-01

. Here we present electrical conductivity records for the Greenland Ice Sheet Project 2 (GISP2) and Greenland Ice-core Project (GRIP) ice cores, drilled 28 km apart to enable direct comparison of the results. The upper parts of both records are consistent with previous evidence from other Greenland cores...

Biophysical investigation of type A PutAs reveals a conserved core oligomeric structure

Energy Technology Data Exchange (ETDEWEB)

Korasick, David A. [Department of Biochemistry, University of Missouri, Columbia MO USA; Singh, Harkewal [Department of Chemistry, University of Missouri, Columbia MO USA; Pemberton, Travis A. [Department of Chemistry, University of Missouri, Columbia MO USA; Luo, Min [Department of Chemistry, University of Missouri, Columbia MO USA; Dhatwalia, Richa [Department of Chemistry, University of Missouri, Columbia MO USA; Tanner, John J. [Department of Biochemistry, University of Missouri, Columbia MO USA; Department of Chemistry, University of Missouri, Columbia MO USA

2017-08-01

Many enzymes form homooligomers, yet the functional significance of self-association is seldom obvious. Herein, we examine the connection between oligomerization and catalytic function for proline utilization A (PutA) enzymes. PutAs are bifunctional enzymes that catalyze both reactions of proline catabolism. Type A PutAs are the smallest members of the family, possessing a minimal domain architecture consisting of N-terminal proline dehydrogenase and C-terminal l-glutamate-γ-semialdehyde dehydrogenase modules. Type A PutAs form domain-swapped dimers, and in one case (Bradyrhizobium japonicum PutA), two of the dimers assemble into a ring-shaped tetramer. Whereas the dimer has a clear role in substrate channeling, the functional significance of the tetramer is unknown. To address this question, we performed structural studies of four-type A PutAs from two clades of the PutA tree. The crystal structure of Bdellovibrio bacteriovorus PutA covalently inactivated by N-propargylglycine revealed a fold and substrate-channeling tunnel similar to other PutAs. Small-angle X-ray scattering (SAXS) and analytical ultracentrifugation indicated that Bdellovibrio PutA is dimeric in solution, in contrast to the prediction from crystal packing of a stable tetrameric assembly. SAXS studies of two other type A PutAs from separate clades also suggested that the dimer predominates in solution. To assess whether the tetramer of B. japonicum PutA is necessary for catalytic function, a hot spot disruption mutant that cleanly produces dimeric protein was generated. The dimeric variant exhibited kinetic parameters similar to the wild-type enzyme. These results implicate the domain-swapped dimer as the core structural and functional unit of type A PutAs.

Current limiting behavior in three-phase transformer-type SFCLs using an iron core according to variety of fault

Science.gov (United States)

Cho, Yong-Sun; Jung, Byung-Ik; Ha, Kyoung-Hun; Choi, Soo-Geun; Park, Hyoung-Min; Choi, Hyo-Sang

To apply the superconducting fault current limiter (SFCL) to the power system, the reliability of the fault-current-limiting operation must be ensured in diverse fault conditions. The SFCL must also be linked to the operation of the high-speed recloser in the power system. In this study, a three-phase transformer-type SFCL, which has a neutral line to improve the simultaneous quench characteristics of superconducting elements, was manufactured to analyze the fault-current-limiting characteristic according to the single, double, and triple line-to-ground faults. The transformer-type SFCL, wherein three-phase windings are connected to one iron core, reduced the burden on the superconducting element as the superconducting element on the sound phase was also quenched in the case of the single line-to-ground fault. In the case of double or triple line-to-ground faults, the flux from the faulted phase winding was interlinked with other faulted or sound phase windings, and the fault-current-limiting rate decreased because the windings of three phases were inductively connected by one iron core.

Comparison of computational performance of GA and PSO optimization techniques when designing similar systems - Typical PWR core case

International Nuclear Information System (INIS)

Souza Lima, Carlos A.; Lapa, Celso Marcelo F.; Pereira, Claudio Marcio do N.A.; Cunha, Joao J. da; Alvim, Antonio Carlos M.

2011-01-01

Research highlights: → Performance of PSO and GA techniques applied to similar system design. → This work uses ANGRA1 (two loop PWR) core as a prototype. → Results indicate that PSO technique is more adequate than GA to solve this kind of problem. - Abstract: This paper compares the performance of two optimization techniques, particle swarm optimization (PSO) and genetic algorithm (GA) applied to the design a typical reduced scale two loop Pressurized Water Reactor (PWR) core, at full power in single phase forced circulation flow. This comparison aims at analyzing the performance in reaching the global optimum, considering that both heuristics are based on population search methods, that is, methods whose population (candidate solution set) evolve from one generation to the next using a combination of deterministic and probabilistic rules. The simulated PWR, similar to ANGRA 1 power plant, was used as a case example to compare the performance of PSO and GA. Results from simulations indicated that PSO is more adequate to solve this kind of problem.

Core Emergence in a Massive Infrared Dark Cloud: A Comparison between Mid-IR Extinction and 1.3 mm Emission

Science.gov (United States)

Kong, Shuo; Tan, Jonathan C.; Arce, Héctor G.; Caselli, Paola; Fontani, Francesco; Butler, Michael J.

2018-03-01

Stars are born from dense cores in molecular clouds. Observationally, it is crucial to capture the formation of cores in order to understand the necessary conditions and rate of the star formation process. The Atacama Large Millimeter/submillimeter Array (ALMA) is extremely powerful for identifying dense gas structures, including cores, at millimeter wavelengths via their dust continuum emission. Here, we use ALMA to carry out a survey of dense gas and cores in the central region of the massive (∼105 M ⊙) infrared dark cloud (IRDC) G28.37+0.07. The observation consists of a mosaic of 86 pointings of the 12 m array and produces an unprecedented view of the densest structures of this IRDC. In this first Letter about this data set, we focus on a comparison between the 1.3 mm continuum emission and a mid-infrared (MIR) extinction map of the IRDC. This allows estimation of the “dense gas” detection probability function (DPF), i.e., as a function of the local mass surface density, Σ, for various choices of thresholds of millimeter continuum emission to define “dense gas.” We then estimate the dense gas mass fraction, f dg, in the central region of the IRDC and, via extrapolation with the DPF and the known Σ probability distribution function, to the larger-scale surrounding regions, finding values of about 5% to 15% for the fiducial choice of threshold. We argue that this observed dense gas is a good tracer of the protostellar core population and, in this context, estimate a star formation efficiency per free-fall time in the central IRDC region of ɛ ff ∼ 10%, with approximately a factor of two systematic uncertainties.

Study on Turn-to-turn Short Circuit On一line Monitoring System for Dry一type Ai r一core Reactor

Directory of Open Access Journals (Sweden)

GAO Zi-wei

2017-04-01

Full Text Available The change of current value caused by turn-to-turn short circuit of dry-type air-core reactor is so little that failure detection is difficult to be carried out. In order to solve this problem，a new on-line monitoring system based on impedance variation of turn-to-turn short circuit is proposed. The numerical method is applied to analyze the variation of equivalent resistance and equivalent reactance when dry-type air-core reactor winding short circuit happens in different places，and the monitoring method based on harmonic analysis method and quasi- synchronization sampling method is analyzed by theory. The hardware system，which takes single-chip microcomputer as the core of data processing and logic control，completes data acquisition of voltage signal and current signal of the reactor. In the respect of software design，the impedance variation will be uploaded to the PC after it has been calculated by using the above monitoring method，and then monitoring of turn-to-turn short circuit fault will be realized. Finally，the design of on-line monitoring system is studied by testing. The research result shows that，the equivalent resistance increases and the equivalent reactance decreases when turn-to-turn short circuit occurs，and the variation of equivalent resistance is more obvious than equivalent reactance. The experiment results prove that this monitoring method is true and the on-line monitoring system is feasible.

Reactivity accident analysis in MTR cores

International Nuclear Information System (INIS)

Waldman, R.M.; Vertullo, A.C.

1987-01-01

The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

Type I band alignment in GaAs{sub 81}Sb{sub 19}/GaAs core-shell nanowires

Energy Technology Data Exchange (ETDEWEB)

Xu, T. [Institut d' Electronique, de Microélectronique et de Nanotechnologies (IEMN), CNRS, UMR 8520, Département ISEN, 41 bd Vauban, 59046 Lille Cedex (France); Key Laboratory of Advanced Display and System Application, Shanghai University, 149 Yanchang Road, Shanghai 200072 (China); Wei, M. J. [Key Laboratory of Advanced Display and System Application, Shanghai University, 149 Yanchang Road, Shanghai 200072 (China); Capiod, P.; Díaz Álvarez, A.; Han, X. L.; Troadec, D.; Nys, J. P.; Berthe, M.; Lefebvre, I.; Grandidier, B., E-mail: bruno.grandidier@isen.iemn.univ-lille1.fr [Institut d' Electronique, de Microélectronique et de Nanotechnologies (IEMN), CNRS, UMR 8520, Département ISEN, 41 bd Vauban, 59046 Lille Cedex (France); Patriarche, G. [CNRS-Laboratoire de Photonique et de Nanostructures (LPN), Route de Nozay, 91460 Marcoussis (France); Plissard, S. R. [Institut d' Electronique, de Microélectronique et de Nanotechnologies (IEMN), CNRS, UMR 8520, Département ISEN, 41 bd Vauban, 59046 Lille Cedex (France); CNRS-Laboratoire d' Analyse et d' Architecture des Systèmes (LAAS), Univ. de Toulouse, 7 Avenue du Colonel Roche, F-31400 Toulouse (France); Caroff, P. [Institut d' Electronique, de Microélectronique et de Nanotechnologies (IEMN), CNRS, UMR 8520, Département ISEN, 41 bd Vauban, 59046 Lille Cedex (France); Department of Electronic Materials Engineering, Research School of Physics and Engineering, The Australian National University, Canberra, Australian Capital Territory 0200 (Australia); and others

2015-09-14

The composition and band gap of the shell that formed during the growth of axial GaAs/GaAs{sub 81}Sb{sub 19}/ GaAs heterostructure nanowires have been investigated by transmission electron microscopy combined with energy dispersion spectroscopy, scanning tunneling spectroscopy, and density functional theory calculations. On the GaAs{sub 81}Sb{sub 19} intermediate segment, the shell is found to be free of Sb (pure GaAs shell) and transparent to the tunneling electrons, despite the (110) biaxial strain that affects its band gap. As a result, a direct measurement of the core band gap allows the quantitative determination of the band offset between the GaAs{sub 81}Sb{sub 19} core and the GaAs shell and identifies it as a type I band alignment.

Experimental Study and Mathematical Modeling of Asphaltene Deposition Mechanism in Core Samples

Directory of Open Access Journals (Sweden)

Jafari Behbahani T.

2015-11-01

Full Text Available In this work, experimental studies were conducted to determine the effect of asphaltene deposition on the permeability reduction and porosity reduction of carbonate, sandstone and dolomite rock samples using an Iranian bottom hole live oil sample which is close to reservoir conditions, whereas in the majority of previous work, a mixture of recombined oil (a mixture of dead oil and associated gas was injected into a core sample which is far from reservoir conditions. The effect of the oil injection rate on asphaltene deposition and permeability reduction was studied. The experimental results showed that an increase in the oil injection flow rate can result in an increase in asphaltene deposition and permeability reduction. Also, it can be observed that at lower injection flow rates, a monotonic decrease in permeability of the rock samples can be attained upon increasing the injection flow rate, while at higher injection rates, after a decrease in rock permeability, an increasing trend is observed before a steady-state condition can be reached. The experimental results also showed that the rock type can affect the amount of asphaltene deposition, and the asphaltene deposition has different mechanisms in sandstone and carbonate core samples. It can be seen that the adsorption and plugging mechanisms have a more important role in asphaltene deposition in carbonate core samples than sandstone core samples. From the results, it can be observed that the pore volumes of the injected crude oil are higher for sandstone cores compared with the carbonate cores. Also, it can be inferred that three depositional types may take place during the crude oil injection, i.e., continuous deposition for low-permeability cores, slow, steady plugging for high-permeability cores and steady deposition for medium-permeability cores. It can be seen from the experimental results that damage to the core samples was found to increase when the production pressures were

HPLC-CUPRAC post-column derivatization method for the determination of antioxidants: a performance comparison between porous silica and core-shell column packing.

Science.gov (United States)

Haque, Syed A; Cañete, Socrates Jose P

2018-01-01

An HPLC method employing a post-column derivatization strategy using the cupric reducing antioxidant capacity reagent (CUPRAC reagent) for the determining antioxidants in plant-based materials leverages the separation capability of regular HPLC approaches while allowing for detection specificity for antioxidants. Three different column types, namely core-shell and porous silica including two chemically different core-shell materials (namely phenyl-hexyl and C18), were evaluated to assess potential improvements that could be attained by changing from a porous silica matrix to a core-shell matrix. Tea extracts were used as sample matrices for the evaluation specifically looking at catechin and epigallocatechin gallate (EGCG). Both the C18 and phenyl-hexyl core-shell columns showed better performance compared to the C18 porous silica one in terms of separation, peak shape, and retention time. Among the two core-shell materials, the phenyl-hexyl column showed better resolving power compared to the C18 column. The CUPRAC post-column derivatization method can be improved using core-shell columns and suitable for quantifying antioxidants, exemplified by catechin and EGCG, in tea samples.

Torque Characteristic Analysis of a Transverse Flux Motor Using a Combined-Type Stator Core

Directory of Open Access Journals (Sweden)

Xiaobao Yang

2016-11-01

Full Text Available An external rotor transverse flux motor using a combined-type stator core is proposed for a direct drive application in this paper. The stator core is combined by two kinds of components that can both be manufactured conveniently by generic laminated silicon steel used in traditional motors. The motor benefits from the predominance of low manufacturing cost and low iron loss by using a silicon-steel sheet. Firstly, the basic structure and operation principles of the proposed motor are introduced. Secondly, the expressions of the electromagnetic torque and the cogging torque are deduced by theoretical analysis. Thirdly, the basic characteristics such as permanent magnet flux linkage, no-load back electromotive force, cogging torque and electromagnetic torque are analyzed by a three-dimensional finite element method (3D FEM. Then, the influence of structure parameters on the torque density is investigated, which provides a useful foundation for optimum design of the novel motor. Finally, the torque density of the proposed motor is calculated and discussed, and the result shows that the proposed motor in this paper can provide considerable torque density by using few permanent magnets.

The design and installation of a core discharge monitor for CANDU-type reactors

International Nuclear Information System (INIS)

Halbig, J.K.; Monticone, A.C.; Ksiezak, L.; Smiltnieks, V.

1990-01-01

A new type of surveillance systems that monitors neutron and gamma radiation in a reactor containment is being installed at the Ontario Hydro Darlington Nuclear Generating Station A, Unit 2. Unlike video or film surveillance that monitors mechanical motion, this system measures fuel-specific radiation emanating from irradiated fuel as it is pushed from the core of CANDU-type reactors. Proof-of-principle measurements have been carried out at Bruce Nuclear Generating Station A, Unit 3. The system uses (γ,n) threshold detectors and ionization detectors. A microprocessor-based electronics package, GRAND-II (Gamma Ray and Neutron Detector electronics package), provides detector bias, preamplifier power, and signal processing. Firmware in the GRAND-2 controls the surveillance activities, including data acquisition and a level of detector authentication, and it handles authenticated communication with a central data logging computer. Data from the GRAND-II are transferred to an MS-DOS-compatible computer and stored. These data are collected and reviewed for fuel-specific radiation signatures from the primary detector and proper ratios of signals from secondary detectors. 5 figs

Development of a core outcome set for clinical trials in facial aging: study protocol for a systematic review of the literature and identification of a core outcome set using a Delphi survey.

Science.gov (United States)

Schlessinger, Daniel I; Iyengar, Sanjana; Yanes, Arianna F; Henley, Jill K; Ashchyan, Hovik J; Kurta, Anastasia O; Patel, Payal M; Sheikh, Umar A; Franklin, Matthew J; Hanna, Courtney C; Chen, Brian R; Chiren, Sarah G; Schmitt, Jochen; Deckert, Stefanie; Furlan, Karina C; Poon, Emily; Maher, Ian A; Cartee, Todd V; Sobanko, Joseph F; Alam, Murad

2017-08-01

Facial aging is a concern for many patients. Wrinkles, loss of volume, and discoloration are common physical manifestations of aging skin. Genetic heritage, prior ultraviolet light exposure, and Fitzpatrick skin type may be associated with the rate and type of facial aging. Although many clinical trials assess the correlates of skin aging, there is heterogeneity in the outcomes assessed, which limits the quality of evaluation and comparison of treatment modalities. To address the inconsistency in outcomes, in this project we will develop a core set of outcomes that are to be evaluated in all clinical trials relevant to facial aging. A long list of measureable outcomes will be created from four sources: (1) systematic medical literature review, (2) patient interviews, (3) other published sources, and (4) stakeholder involvement. Two rounds of Delphi processes with homogeneous groups of physicians and patients will be performed to prioritize and condense the list. At a consensus meeting attended by physicians, patients, and stakeholders, outcomes will be further condensed on the basis of participant scores. By the end of the meeting, members will vote and decide on a final recommended set of core outcomes. Subsequent to this, specific measures will be selected or created to assess these outcomes. The aim of this study is to develop a core outcome set and relevant measures for clinical trials relevant to facial aging. We hope to improve the reliability and consistency of outcome reporting of skin aging, thereby enabling improved evaluation of treatment efficacy and patient satisfaction. Core Outcome Measures in Effectiveness Trials (COMET) Initiative, accessible at http://www.comet-initiative.org/studies/details/737 . Core Outcomes Set Initiative, (CSG-COUSIN) accessible at https://www.uniklinikum-dresden.de/de/das-klinikum/universitaetscentren/zegv/cousin/meet-the-teams/project-groups/core-outcome-set-for-the-appearance-of-facial-aging . Protocol version date is 28

Improvement of SSR core design for ABWR-II

International Nuclear Information System (INIS)

Moriwaki, Masanao; Aoyama, Motoo; Okada, Hiroyuki; Kitamura, Hideya; Sakurada, Koichi; Tanabe, Akira

2003-01-01

In order to enhance the spectral shift effect in the ABWR-II reactor, a novel core design to bring out better performance of spectral shift rods (SSRs) is studied. The SSR is a new type of water rod, in which the water level develops naturally during operation and changes according to the coolant flow rate through the channel. By using the SSR, the average moderator density, which is directly related to core reactivity, can be controlled over a wide range by the core flow rate. In the new SSR core design, two types of SSR bundles, in which settings for the SSR water levels are different, are utilized and loaded according to flow distribution in the core. This two-region SSR core design allows wide variation in the average SSR water level, thus improving fuel economy. Enhancement of SSR function in the two-region SSR core increases the uranium saving factor by about 25%, from the 6% of the conventional uniform SSR core to about 8%. (author)

Glacial–interglacial dynamics of Antarctic firn columns: comparison between simulations and ice core air-δ15N measurements

Directory of Open Access Journals (Sweden)

E. Capron

2013-05-01

Full Text Available Correct estimation of the firn lock-in depth is essential for correctly linking gas and ice chronologies in ice core studies. Here, two approaches to constrain the firn depth evolution in Antarctica are presented over the last deglaciation: outputs of a firn densification model, and measurements of δ15N of N2 in air trapped in ice core, assuming that δ15N is only affected by gravitational fractionation in the firn column. Since the firn densification process is largely governed by surface temperature and accumulation rate, we have investigated four ice cores drilled in coastal (Berkner Island, BI, and James Ross Island, JRI and semi-coastal (TALDICE and EPICA Dronning Maud Land, EDML Antarctic regions. Combined with available ice core air-δ15N measurements from the EPICA Dome C (EDC site, the studied regions encompass a large range of surface accumulation rates and temperature conditions. Our δ15N profiles reveal a heterogeneous response of the firn structure to glacial–interglacial climatic changes. While firn densification simulations correctly predict TALDICE δ15N variations, they systematically fail to capture the large millennial-scale δ15N variations measured at BI and the δ15N glacial levels measured at JRI and EDML – a mismatch previously reported for central East Antarctic ice cores. New constraints of the EDML gas–ice depth offset during the Laschamp event (~41 ka and the last deglaciation do not favour the hypothesis of a large convective zone within the firn as the explanation of the glacial firn model–δ15N data mismatch for this site. While we could not conduct an in-depth study of the influence of impurities in snow for firnification from the existing datasets, our detailed comparison between the δ15N profiles and firn model simulations under different temperature and accumulation rate scenarios suggests that the role of accumulation rate may have been underestimated in the current description of firnification

In-core fuel management for nuclear reactor

International Nuclear Information System (INIS)

Ross, M.F.; Visner, S.

1986-01-01

This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

Emergency core cooling system in BWR type reactors

International Nuclear Information System (INIS)

Takizawa, Yoji

1981-01-01

Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

Comment on the in-core measurement in the WWER nuclear power plant

International Nuclear Information System (INIS)

Krett, V.; Dach, K.; Erben, O.

1985-01-01

The activity of the Nuclear Research Institute (NRI) Rez in the field of in-core measurement sensors is described in the paper. The results of comparison and calibration experiments realized on the WWR-S research reactor at the NRI are presented. Measurements with fission calorimeters and SPN detectors carried out in the framework of diagnostic fuel assembly program of WWER NPP reactors are described. Noise measurements with detectors of in-core instrumentation of diagnostic fuel assemblies are also mentioned. Comparison experiments on the WWER-440 NPP reactor are described and the method of function verification of neutron sensors of the in-core control system of these reactors is given. (author)

Piezo-phototronic Effect Enhanced UV/Visible Photodetector Based on Fully Wide Band Gap Type-II ZnO/ZnS Core/Shell Nanowire Array.

Science.gov (United States)

Rai, Satish C; Wang, Kai; Ding, Yong; Marmon, Jason K; Bhatt, Manish; Zhang, Yong; Zhou, Weilie; Wang, Zhong Lin

2015-06-23

A high-performance broad band UV/visible photodetector has been successfully fabricated on a fully wide bandgap ZnO/ZnS type-II heterojunction core/shell nanowire array. The device can detect photons with energies significantly smaller (2.2 eV) than the band gap of ZnO (3.2 eV) and ZnS (3.7 eV), which is mainly attributed to spatially indirect type-II transition facilitated by the abrupt interface between the ZnO core and ZnS shell. The performance of the device was further enhanced through the piezo-phototronic effect induced lowering of the barrier height to allow charge carrier transport across the ZnO/ZnS interface, resulting in three orders of relative responsivity change measured at three different excitation wavelengths (385, 465, and 520 nm). This work demonstrates a prototype UV/visible photodetector based on the truly wide band gap semiconducting 3D core/shell nanowire array with enhanced performance through the piezo-phototronic effect.

A simple reactivity feedback model accounting for radial core expansion effects in the liquid metal fast reactor

International Nuclear Information System (INIS)

Kwon, Young Min; Lee, Yong Bum; Chang, Won Pyo; Haha, Do Hee

2002-01-01

The radial core expansion due to the structure temperature rise is one of major negative reactivity insertion mechanisms in metallic fueled reactor. Thermal expansion is a result of both the laws of nature and the particular core design and it causes negative reactivity feedback by the combination of increased core volume captures and increased core surface leakage. The simple radial core expansion reactivity feedback model developed for the SSC-K code was evaluated by the code-to-code comparison analysis. From the comparison results, it can be stated that the radial core expansion reactivity feedback model employed into the SSC-K code may be reasonably accurate in the UTOP analysis

A simple reactivity feedback model accounting for radial core expansion effects in the liquid metal fast reactor

Energy Technology Data Exchange (ETDEWEB)

Kwon, Young Min; Lee, Yong Bum; Chang, Won Pyo; Haha, Do Hee [KAERI, Taejon (Korea, Republic of)

2002-10-01

The radial core expansion due to the structure temperature rise is one of major negative reactivity insertion mechanisms in metallic fueled reactor. Thermal expansion is a result of both the laws of nature and the particular core design and it causes negative reactivity feedback by the combination of increased core volume captures and increased core surface leakage. The simple radial core expansion reactivity feedback model developed for the SSC-K code was evaluated by the code-to-code comparison analysis. From the comparison results, it can be stated that the radial core expansion reactivity feedback model employed into the SSC-K code may be reasonably accurate in the UTOP analysis.

Family and infant characteristics associated with timing of core and non-core food introduction in early childhood

OpenAIRE

Schrempft, Stephanie; van Jaarsveld, Cornelia H.M.; Fisher, Abigail; Wardle, Jane

2013-01-01

Objective To identify family and infant characteristics associated with timing of introduction of two food types: core foods (nutrient-dense) and non-core foods (nutrient-poor) in a population-based sample of mothers and infants. Method Participants were 1861 mothers and infants from the Gemini twin birth cohort (one child per family). Family and infant characteristics were assessed when the infants were around 8 months old. Timing of introducing core and non-core foods was assessed at 8 and ...

Preparation of Cotton-Wool-Like Poly(lactic acid-Based Composites Consisting of Core-Shell-Type Fibers

Directory of Open Access Journals (Sweden)

Jian Wang

2015-11-01

Full Text Available In previous works, we reported the fabrication of cotton-wool-like composites consisting of siloxane-doped vaterite and poly(l-lactic acid (SiVPCs. Various irregularly shaped bone voids can be filled with the composite, which effectively supplies calcium and silicate ions, enhancing the bone formation by stimulating the cells. The composites, however, were brittle and showed an initial burst release of ions. In the present work, to improve the mechanical flexibility and ion release, the composite fiber was coated with a soft, thin layer consisting of poly(d,l-lactic-co-glycolic acid (PLGA. A coaxial electrospinning technique was used to prepare a cotton-wool-like material comprising “core-shell”-type fibers with a diameter of ~12 µm. The fibers, which consisted of SiVPC coated with a ~2-µm-thick PLGA layer, were mechanically flexible; even under a uniaxial compressive load of 1.5 kPa, the cotton-wool-like material did not exhibit fracture of the fibers and, after removing the load, showed a ~60% recovery. In Tris buffer solution, the initial burst release of calcium and silicate ions from the “core-shell”-type fibers was effectively controlled, and the ions were slowly released after one day. Thus, the mechanical flexibility and ion-release behavior of the composites were drastically improved by the thin PLGA coating.

Core size effects on safety performances of LMRs

Energy Technology Data Exchange (ETDEWEB)

Na, Byung Chan; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

An oxide fuel small size core (1200 MWt) was analyzed in comparison with a large size core (3600 MWt) in order to evaluate the size effects on transient safety performances of liquid-metal reactors (LMRs). In the first part of the study, main static safety parameters (i.e., Doppler coefficient, sodium void effect, etc.) of the two cores were characterized, and the second part of the study was focused on the dynamic behavior of the cores in two representative transient events: the unprotected loss-of-flow (ULOF) and the unprotected transient overpower (UTOP). Margins to fuel melting and sodium boiling have been evaluated for these representative transients. Results show that the small core has a generally better or equivalent level of safety performances during these events. 6 refs., 4 figs., 2 tabs. (Author)

Core size effects on safety performances of LMRs

Energy Technology Data Exchange (ETDEWEB)

Na, Byung Chan; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

An oxide fuel small size core (1200 MWt) was analyzed in comparison with a large size core (3600 MWt) in order to evaluate the size effects on transient safety performances of liquid-metal reactors (LMRs). In the first part of the study, main static safety parameters (i.e., Doppler coefficient, sodium void effect, etc.) of the two cores were characterized, and the second part of the study was focused on the dynamic behavior of the cores in two representative transient events: the unprotected loss-of-flow (ULOF) and the unprotected transient overpower (UTOP). Margins to fuel melting and sodium boiling have been evaluated for these representative transients. Results show that the small core has a generally better or equivalent level of safety performances during these events. 6 refs., 4 figs., 2 tabs. (Author)

Core monitoring with analytical model adaption

International Nuclear Information System (INIS)

Linford, R.B.; Martin, C.L.; Parkos, G.R.; Rahnema, F.; Williams, R.D.

1992-01-01

The monitoring of BWR cores has evolved rapidly due to more capable computer systems, improved analytical models and new types of core instrumentation. Coupling of first principles diffusion theory models such as applied to design to the core instrumentation has been achieved by GE with an adaptive methodology in the 3D Minicore system. The adaptive methods allow definition of 'leakage parameters' which are incorporated directly into the diffusion models to enhance monitoring accuracy and predictions. These improved models for core monitoring allow for substitution of traversing in-core probe (TIP) and local power range monitor (LPRM) with calculations to continue monitoring with no loss of accuracy or reduction of thermal limits. Experience in small BWR cores has shown that with one out of three TIP machines failed there was no operating limitation or impact from the substitute calculations. Other capabilities exist in 3D Monicore to align TIPs more accurately and accommodate other types of system measurements or anomalies. 3D Monicore also includes an accurate predictive capability which uses the adaptive results from previous monitoring calculations and is used to plan and optimize reactor maneuvers/operations to improve operating efficiency and reduce support requirements

Defining and Evaluating a Core Genome Multilocus Sequence Typing Scheme for Genome-Wide Typing of Clostridium difficile.

Science.gov (United States)

Bletz, Stefan; Janezic, Sandra; Harmsen, Dag; Rupnik, Maja; Mellmann, Alexander

2018-06-01

Clostridium difficile , recently renamed Clostridioides difficile , is the most common cause of antibiotic-associated nosocomial gastrointestinal infections worldwide. To differentiate endogenous infections and transmission events, highly discriminatory subtyping is necessary. Today, methods based on whole-genome sequencing data are increasingly used to subtype bacterial pathogens; however, frequently a standardized methodology and typing nomenclature are missing. Here we report a core genome multilocus sequence typing (cgMLST) approach developed for C. difficile Initially, we determined the breadth of the C. difficile population based on all available MLST sequence types with Bayesian inference (BAPS). The resulting BAPS partitions were used in combination with C. difficile clade information to select representative isolates that were subsequently used to define cgMLST target genes. Finally, we evaluated the novel cgMLST scheme with genomes from 3,025 isolates. BAPS grouping ( n = 6 groups) together with the clade information led to a total of 11 representative isolates that were included for cgMLST definition and resulted in 2,270 cgMLST genes that were present in all isolates. Overall, 2,184 to 2,268 cgMLST targets were detected in the genome sequences of 70 outbreak-associated and reference strains, and on average 99.3% cgMLST targets (1,116 to 2,270 targets) were present in 2,954 genomes downloaded from the NCBI database, underlining the representativeness of the cgMLST scheme. Moreover, reanalyzing different cluster scenarios with cgMLST were concordant to published single nucleotide variant analyses. In conclusion, the novel cgMLST is representative for the whole C. difficile population, is highly discriminatory in outbreak situations, and provides a unique nomenclature facilitating interlaboratory exchange. Copyright © 2018 American Society for Microbiology.

Accelerating the SCE-UA Global Optimization Method Based on Multi-Core CPU and Many-Core GPU

Directory of Open Access Journals (Sweden)

Guangyuan Kan

2016-01-01

Full Text Available The famous global optimization SCE-UA method, which has been widely used in the field of environmental model parameter calibration, is an effective and robust method. However, the SCE-UA method has a high computational load which prohibits the application of SCE-UA to high dimensional and complex problems. In recent years, the hardware of computer, such as multi-core CPUs and many-core GPUs, improves significantly. These much more powerful new hardware and their software ecosystems provide an opportunity to accelerate the SCE-UA method. In this paper, we proposed two parallel SCE-UA methods and implemented them on Intel multi-core CPU and NVIDIA many-core GPU by OpenMP and CUDA Fortran, respectively. The Griewank benchmark function was adopted in this paper to test and compare the performances of the serial and parallel SCE-UA methods. According to the results of the comparison, some useful advises were given to direct how to properly use the parallel SCE-UA methods.

Flexural Behavior of Aluminum Honeycomb Core Sandwich Structure

Science.gov (United States)

Matta, Vidyasagar; Kumar, J. Suresh; Venkataraviteja, Duddu; Reddy, Guggulla Bharath Kumar

2017-05-01

This project is concerned with the fabrication and flexural testing of aluminium honey comb sandwich structure which is a special case of composite materials that is fabricated by attaching two thin but stiff skins to a light weight but thick core. The core material is normally low density material but its high thickness provide the sandwich composite with high bonding stiffness. Honeycomb core are classified into two types based on the materials and structures. Hexagonal shape has a unique properties i.e has more bonding strength and less formation time based on the cell size and sheet thickness. Sandwich structure exhibit different properties such as high load bearing capacity at low weight and has excellent thermal insulation. By considering the above properties it has tendency to minimize the structural problem. So honey comb sandwich structure is choosed. The core structure has a different applications such as aircraft, ship interiors, construction industries. As there is no proper research on strength characteristics of sandwich structure. So, we use light weight material to desire the strength. There are different parameters involved in this structure i.e cell size, sheet thickness and core height. In this project we considered 3 level of comparison among the 3 different parameters cell size of 4, 6 and 8 mm, sheet thickness of 0.3, 0.5 and 0.7 mm, and core height of 20,25 and 30 mm. In order to reduce the number of experiment we use taguchi design of experiment, and we select the L8 orthogonal array is the best array for this type of situation, which clearly identifies the parameters by independent of material weight to support this we add the minitab software, to identify the main effective plots and regression equation which involves the individual response and corresponding parameters. Aluminium material is used for the fabrication of Honeycomb sandwich structure among the various grades of aluminium we consider the AL6061 which is light weight material

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

Energy Technology Data Exchange (ETDEWEB)

Ehud Greenspan

2008-09-30

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

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)

The role of the core in degeneracy of chiral candidate band doubling

International Nuclear Information System (INIS)

Timar, J.; Sohler, D.; Vaman, C.; SUNY, Stony Brook, NY; Starosta, K.; Fossan, D.B.; Koike, T.; Tohoku Univ., Sendai; Lee, I.Y.; Macchiavelli, A.O.

2005-01-01

Complete text of publication follows. Nearly degenerate ΔI=1 rotational bands have been observed recently in several odd-odd nuclei in the A ∼ 130 and A ∼ 100 mass regions. The properties of these doublet bands have been found to agree with the scenario of spontaneous formation of chirality and disagree with other possible scenarios. However, the most recent results obtained from life-time experiments for some chiral candidate nuclei in the A ∼ 130 mass region seem to contradict the chiral interpretation of the doublet bands in these nuclei based on the observed differences in the absolute electromagnetic transition rates; the transition rates expected for chiral doublets are predicted to be very similar. Therefore it is interesting to search for new types of experimental data that may provide further possibilities to distinguish between alternative interpretations, and may uncover new properties of the mechanism that is responsible for the band doubling in these nuclei. Such a new type of experimental data was found by studying the chiral candidate bands in neighboring Rh nuclei. High-spin states of 103 Rh were studied using the 96 Zr( 11 B,4n) reaction at 40 MeV beam energy and chiral partner candidate bands have been found in it. As a result of this observation a special quartet of neighboring chiral candidate nuclei can be investigated for the first time. With this quartet identified a comparison between the behavior of the nearly degenerate doublet bands belonging to the same core but to different valence quasiparticle configurations, as well as belonging to different cores but to the same valence quasiparticle configuration, becomes possible. The comparison shows that the energy separation of these doublet band structures depends mainly on the core properties and only at less extent on the valence quasiparticle coupling. This observation sets up new criteria for the explanations of the band doublings, restricting the possible scenarios and providing

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.

Methodological aspects of core meltdown accidents frequency estimates

International Nuclear Information System (INIS)

Matthis, P.

1984-01-01

A survey is given of the work of the ecological institute relating to models and methods used in the German Risk Study for the assessment of core meltdown accident frequency. A statistical model used by the ecological institute for the estimation of the outage behaviour of components is taken as a comparison, which leads to the conclusion that no appropriate methods for the assessment of component reliability are available to date. Furthermore, there are no secured methods for error propagation computation. The lower limits for the ranges of reliability of components are calculated by approximation. As a result of imperfect modelling and of a number of methodical inaccuracies and neglects, the German Risk Study underestimates the ranges of component reliability by a factor of 3 to 70 (depending on the type of component). (RF) [de

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

additional consideration should be required in nuclear design and fuel treating facilities due to reactivity coefficient being shifted to the plus side, larger neutron yield and increased heat source caused by MA loading. (2) Confirmation of TRU burning reactor core concepts. The core specification of sodium cooled-nitride fueled TRU burning large reactor was designed based on commercial type fast reactor (sodium cooled nitride fueled large fast reactor, 38000 MWt) which was designed in the feasibility studies on commercialized fast reactor cycle system. The composition of MAs from LWR's spent fuel was supposed. MA content in the core fuel is settled to 60 wt% based on the JAERI's design in order to maximize the MA transmutation amount. We need to exchange 25% of core fuel with zirconium hydride (ZrH 1.6 ) to attain Doppler coefficient being equivalent to that of the conventional type commercial fast reactor loaded 5 wt% MA. Furthermore, this reactor could transmute MAs produced in forty-eight sodium cooled nitride fueled large fast reactors generating the same output. In order to investigate the dependency of MA transmutation characteristics on the reactor output, 1200 MWt TRU burning middle or small reactor core concept was designed. This core was settled by reducing the number of core fuel assemblies from that of TRU burning large reactor designed above. MA transmutation rate in this core is smaller than that in the TRU burning large reactor core because the neutron flux of this core becomes smaller than that of the TRU burning large reactor core due to the higher Pu enrichment. (3) Comparison between TRU burning reactor and conventional type commercial fast reactor. MA transmutation and nuclear characteristics of the sodium cooled nitride fuel commercial type fast reactor loaded 5 wt%MA were evaluated and compared with those of TRU burning large reactor designed in (2). The commercial type fast reactor could only transmute MAs produced in seven sodium cooled nitride

Core damage frequency estimation using accident sequence precursor data: 1990-1993

International Nuclear Information System (INIS)

Martz, H.F.

1998-01-01

The Nuclear Regulatory Commission's (NRC's) ongoing Accident Sequence Precursor (ASP) program uses probabilistic risk assessment (PRA) techniques to assess the potential for severe core damage (henceforth referred to simply as core damage) based on operating events. The types of operating events considered include accident sequence initiators, safety equipment failures, and degradation of plant conditions that could increase the probability that various postulated accident sequences occur. Such operating events potentially reduce the margin of safety available for prevention of core damage an thus can be considered as precursors to core damage. The current process for identifying, analyzing, and documenting ASP events is described in detail in Vanden Heuval et al. The significance of a Licensee Event Report (LER) event (or events) is measured by means of the conditional probability that the event leads to core damage, the so-called conditional core damage probability or, simply, CCDP. When the first ASP study results were published in 1982, it covered the period 1969--1979. In addition to identification and ranking of precursors, the original study attempted to estimate core damage frequency (CDF) based on the precursor events. The purpose of this paper is to compare the average annual CDF estimates calculated using the CCDP sum, Cooke-Goossens, Bier, and Abramson estimators for various reactor classes using the combined ASP data for the four years, 1990--1993. An important outcome of this comparison is an answer to the persistent question regarding the degree and effect of the positive bias of the CCDP sum method in practice. Note that this paper only compares the estimators with each other. Because the true average CDF is unknown, the estimation error is also unknown. Therefore, any observations or characterizations of bias are based on purely theoretical considerations

Reinventing the Platform Core Through Acquisition

DEFF Research Database (Denmark)

Toppenberg, Gustav; Henningsson, Stefan; Eaton, Ben

2016-01-01

the acquisition and integration of companies presenting innovative technologies of relevance to the platform core. Using a revelatory case study of Cisco Systems, we develop the explanatory notion of ‘coring acquisition’. In this type of acquisition value is created through the acquisition of companies...

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)

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

Estimation of reactor core calculation by HELIOS/MASTER at power generating condition through DeCART, whole-core transport code

International Nuclear Information System (INIS)

Kim, H. Y.; Joo, H. G.; Kim, K. S.; Kim, G. Y.; Jang, M. H.

2003-01-01

The reactivity and power distribution errors of the HELIOS/MASTER core calculation under power generating conditions are assessed using a whole core transport code DeCART. For this work, the cross section tablesets were generated for a medium sized PWR following the standard procedure and two group nodal core calculations were performed. The test cases include the HELIOS calculations for 2-D assemblies at constant thermal conditions, MASTER 3D assembly calculations at power generating conditions, and the core calculations at HZP, HFP, and an abnormal power conditions. In all these cases, the results of the DeCART code in which pinwise thermal feedback effects are incorporated are used as the reference. The core reactivity, assemblywise power distribution, axial power distribution, peaking factor, and thermal feedback effects are then compared. The comparison shows that the error of the HELIOS/MASTER system in the core reactivity, assembly wise power distribution, pin peaking factor are only 100∼300 pcm, 3%, and 2%, respectively. As far as the detailed pinwise power distribution is concerned, however, errors greater than 15% are observed

Alteration of properties of rock during their selection by shooting core lifter

Energy Technology Data Exchange (ETDEWEB)

Malinin, V F

1969-01-01

During the process of intrusion of the core lifter into rock, splitting and dislocation of the granules and crystals which compose it occur. In the core lifters, single small nondisintegrated fragments are sometimes encountered. Data on comparison of porosity of crushed cores and rock from which they were selected indicate increase in porosity and penetration of the filtrate of the drilling solution during the process of coring. The determined residual oil saturation of the core is different from the residual oil saturation of the rock from which they were selected. The permeability of cores of rock with high porosity is altered.

Comparison of Several Thermal Conductivity Constants for Thermal Hydraulic Calculation of Pebble Bed Reactor

Science.gov (United States)

Irwanto, Dwi; Setiadipura, Topan; Pramutadi, Asril

2017-07-01

There are two type of High Temperature Gas Reactor (HTGR), prismatic and pebble bed. Pebble Bed type has unique configuration because the fuels are randomly distributed inside the reactor core. In term of safety features, Pebble Bed Reactor (PBR) is one of the most promising reactor type in avoiding severe nuclear accidents. In order to analyze heat transfer and safety of this reactor type, a computer code is now under development. As a first step, calculation method proposed by Stroh [1] is adopted. An approach has been made to treat randomly distributed pebble balls contains fissile material inside the reactor core as a porous medium. Helium gas act as coolant on the reactor system are carrying heat flowing in the area between the pebble balls. Several parameters and constants are taken into account in the new developed code. Progress of the development of the code especially comparison of several thermal conductivity constants for a certain PBR-case are reported in the present study.

RA-3 reactor core with uranium silicide fuel elements P-07 type

International Nuclear Information System (INIS)

Abbate, Maximo J.; Sbaffoni, Maria M.

2003-01-01

Following the studies on the utilization of fuel elements (FE) containing uranium silicide, core of the RA-3 was analyzed with several calculation models. At first, the present situation, i.e. the core charged with normal FE (U 3 O 8 ), has been analyzed to validate the simulation methodology comparing with experimental results and to establish reference data to 5 and 10 MW able to be compared with future new situations. Also, CITVAP's nuclear data libraries to be used in irradiation experiment planning were completed. The results were satisfactory and were applied to the study of the core containing P-07 FE [U 3 Si 2 ], in face of a future core change. Comparing with the performance of the U 3 O 8 FE, the silicides ones show the following advantages: - average burnup: 45 % greater; -extraction burnup increase 12 %; and, -the residence time [in full power days] could be a 117 % greater. (author)

Thermohydraulic model of WWER-1000 core

International Nuclear Information System (INIS)

Maroti, L.; Szabados, L.

1987-11-01

Safe and economic operation of the WWER-1000 type reactor requires more accurate calculation of the thermohydraulic processes than the one satisfactory for the 440 type cores. The high degree of accuracy is needed both for reactor physics calculations and for the determination of the operational safety limits of the core. The paper illustrates the most important differences between the 1000 and 440 type reactors and presents the main fields of the development work necessary to reach the required accuracy. A prediction for the capability of the computer programs after the proposed development is also given and some suggestions for the further improvement is outlined. (author) 7 refs

Core calculations of JMTR

Energy Technology Data Exchange (ETDEWEB)

Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

1998-03-01

In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)

Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

International Nuclear Information System (INIS)

Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

1979-01-01

Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

APROS 3-D core models for simulators and plant analyzers

International Nuclear Information System (INIS)

Puska, E.K.

1999-01-01

The 3-D core models of APROS simulation environment can be used in simulator and plant analyzer applications, as well as in safety analysis. The key feature of APROS models is that the same physical models can be used in all applications. For three-dimensional reactor cores the APROS models cover both quadratic BWR and PWR cores and the hexagonal lattice VVER-type cores. In APROS environment the user can select the number of flow channels in the core and either five- or six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the channel description have a decisive effect on the calculation time of the 3-D core model and thus just these selection make at present the major difference between a safety analysis model and a training simulator model. The paper presents examples of various types of 3-D LWR-type core descriptions for simulator and plant analyzer use and discusses the differences of calculation speed and physical results between a typical safety analysis model description and a real-time simulator model description in transients. (author)

Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

International Nuclear Information System (INIS)

Boer, Brian

2007-01-01

The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

Axial heterogeneous core concept applied for super phoenix reactor

International Nuclear Information System (INIS)

Batista, J.L.; Renke, C.A.C.; Waintraub, M.; Santos Bastos, W. dos; Brito Aghina, L.O. de.

1991-11-01

Always maintaining the current design rules, this paper presents a parametric study on the type of axial heterogeneous core concept (CHA), utilizing a core of fast reactor Super Phenix type, reaching a maximum thermal burnup rate of 150000 M W d/t and being managed in single batch. (author)

Nuclear reactor core servicing apparatus

International Nuclear Information System (INIS)

Andrea, C.

1977-01-01

Disclosed is an improved core servicing apparatus for a nuclear reactor of the type having a reactor vessel, a vessel head having a head penetration therethrough, a removable plug adapted to fit in the head penetration, and a core of the type having an array of elongated assemblies. The improved core servicing apparatus comprises a plurality of support columns suspended from the removable plug and extending downward toward the nuclear core, rigid support means carried by each of the support columns, and a plurality of servicing means for each of the support columns for servicing a plurality of assemblies. Each of the plurality of servicing means for each of the support columns is fixedly supported in a fixed array from the rigid support means. Means are provided for rotating the rigid support means and servicing means between condensed and expanded positions. When in the condensed position, the rigid support means and servicing means lie completely within the coextensive boundaries of the plug, and when in the expanded position, some of the rigid support means and servicing means lie without the coextensive boundaries of the plug

Core barrel motion calibration factor calculation

International Nuclear Information System (INIS)

Shahrokhi, F.; Robinson, J.C.

1976-01-01

Neutron transport theory calculations were performed to obtain a calibration factor for inferring core-barrel motion from spectral density data using excore ionization chambers in PWRs. The analysis of core-barrel movement was based on the postulate that the movement is a cantilevered type, with the preferred direction x-x'

Corrosion of MTR type fuel plates containing U3O8-Al cermet cores

International Nuclear Information System (INIS)

Durazzo, M.

1985-01-01

The fuel plate samples containing U 3 O 8 -Al cermet cores with concentrations from 10 to 90% of U 3 O 8 weight were fabricated. Samples with 58% of U 3 O 8 eight were fabricated using compacts with densities from 75 to 95% of theoretical density. The influences of U 3 O 8 concentration and porosity of compacted core on porosity and uniformity of core thickness are discussed. The U 3 O 8 -Al cores were submitted to corrosion tests and exposed to deionized water at temperatures of 30, 50, 70 and 90 0 C by cladding deffect produced artificially. The results shown that core corrosion is accompanied by hydrogen release. The total volum of released hydrogen and the time interval to observe the initiation of hydrogen releasing (incubation time) are depending on core pososity and absolute temperature. A mechanism for U 3 O 8 -Al core corrosion process is proposed and discussed. The cladding of fuel plate samples was submitted to corrosion tests under similar conditons of the IAE-R1 reactor operating at 2, 5 and 10 MW. (Author) [pt

Effect of ultraviolet light irradiation period on bond strengths between fiber-reinforced composite post and core build-up composite resin.

Science.gov (United States)

Asakawa, Yuya; Takahashi, Hidekazu; Iwasaki, Naohiko; Kobayashi, Masahiro

2014-01-01

The aim of the present study was to characterize the effects of the ultraviolet light (UV) irradiation period on the bond strength of fiber-reinforced composite (FRC) posts to core build-up resin. Three types of FRC posts were prepared using polymethyl methacrylate, urethane dimethacrylate, and epoxy resin. The surfaces of these posts were treated using UV irradiation at a distance of 15 mm for 0 to 600 s. The pull-out bond strength was measured and analyzed with the Dunnett's comparison test (α=0.05). The bond strengths of the post surfaces without irradiation were 6.9 to 7.4 MPa; those after irradiation were 4.2 to 26.1 MPa. The bond strengths significantly increased after 15 to 120-s irradiation. UV irradiation on the FRC posts improved the bond strengths between the FRC posts and core build-up resin regardless of the type of matrix resin.

Muon spin rotation studies of electronic excitations and magnetism in the vortex cores of superconductors

International Nuclear Information System (INIS)

Sonier, J E

2007-01-01

The focus of this paper is on recent progress in muon spin rotation (μSR) studies of the vortex cores in type-II superconductors. By comparison of μSR measurements of the vortex core size in a variety of materials with results from techniques that directly probe electronic states, the effect of delocalized quasiparticles on the spatial variation of field in a lattice of interacting vortices has been determined for both single-band and multi-band superconductors. These studies demonstrate the remarkable accuracy of what some still consider an exotic technique. In recent years μSR has also been used to search for magnetism in and around the vortex cores of high-temperature superconductors. As a local probe μSR is specially suited for detecting static or quasistatic magnetism having short-range or random spatial correlations. As discussed in this review, μSR experiments support a generic phase diagram of competing superconducting and magnetic order parameters, characterized by a quantum phase transition to a state where the competing order is spatially nonuniform

A study of the Eocene S-type granites of Chapedony metamorphic core complex (northeast of Yazd province, Central Iran)

International Nuclear Information System (INIS)

Zakipour, A.; Torabi, Gh.

2016-01-01

The Eocene Chapedony metamorphic core complex, is located in western part of the Posht-e-Badam block. This complex is consisting of migmatite, gneiss, amphibolite, marble, micaschist and various types of granitoids. In middle part of this complex (Kalut-e-Chapedony), an Eocene granitic rock unit cross cuts the other rocks. The minerals of this granite are plagioclase (An 9 Ab 8 7O r 4), potassium feldspars (orthoclase), quartz, euhedral garnet (Alm 7 7Sps 1 3Prp 9 Grs 1 ), zircon, apatite, fibrolitic sillimanite and muscovite. Petrology and geochemical studies reveal calc-alkaline, peraluminous and S-type nature of the studied granites. Chondrite-normalized REE patterns represent evident negative anomaly of Eu and low values of the REEs. Continental crust and North American shale composite (NASC) - normalized multi-elements spider diagrams indicate trace elements depletion. These granites are formed by melting of continental crust sedimentary rocks, resulted by emplacement of mantle-derived magma at the bottom of continental crust which formed the Chapedony metamorphic core complex. The source rock of these granites should be a clay-rich sedimentary rock with low amount of plagioclase and CaO/Na 2 O ratio.

Efficient Multicriteria Protein Structure Comparison on Modern Processor Architectures

Science.gov (United States)

Manolakos, Elias S.

2015-01-01

Fast increasing computational demand for all-to-all protein structures comparison (PSC) is a result of three confounding factors: rapidly expanding structural proteomics databases, high computational complexity of pairwise protein comparison algorithms, and the trend in the domain towards using multiple criteria for protein structures comparison (MCPSC) and combining results. We have developed a software framework that exploits many-core and multicore CPUs to implement efficient parallel MCPSC in modern processors based on three popular PSC methods, namely, TMalign, CE, and USM. We evaluate and compare the performance and efficiency of the two parallel MCPSC implementations using Intel's experimental many-core Single-Chip Cloud Computer (SCC) as well as Intel's Core i7 multicore processor. We show that the 48-core SCC is more efficient than the latest generation Core i7, achieving a speedup factor of 42 (efficiency of 0.9), making many-core processors an exciting emerging technology for large-scale structural proteomics. We compare and contrast the performance of the two processors on several datasets and also show that MCPSC outperforms its component methods in grouping related domains, achieving a high F-measure of 0.91 on the benchmark CK34 dataset. The software implementation for protein structure comparison using the three methods and combined MCPSC, along with the developed underlying rckskel algorithmic skeletons library, is available via GitHub. PMID:26605332

Efficient Multicriteria Protein Structure Comparison on Modern Processor Architectures.

Science.gov (United States)

Sharma, Anuj; Manolakos, Elias S

2015-01-01

Fast increasing computational demand for all-to-all protein structures comparison (PSC) is a result of three confounding factors: rapidly expanding structural proteomics databases, high computational complexity of pairwise protein comparison algorithms, and the trend in the domain towards using multiple criteria for protein structures comparison (MCPSC) and combining results. We have developed a software framework that exploits many-core and multicore CPUs to implement efficient parallel MCPSC in modern processors based on three popular PSC methods, namely, TMalign, CE, and USM. We evaluate and compare the performance and efficiency of the two parallel MCPSC implementations using Intel's experimental many-core Single-Chip Cloud Computer (SCC) as well as Intel's Core i7 multicore processor. We show that the 48-core SCC is more efficient than the latest generation Core i7, achieving a speedup factor of 42 (efficiency of 0.9), making many-core processors an exciting emerging technology for large-scale structural proteomics. We compare and contrast the performance of the two processors on several datasets and also show that MCPSC outperforms its component methods in grouping related domains, achieving a high F-measure of 0.91 on the benchmark CK34 dataset. The software implementation for protein structure comparison using the three methods and combined MCPSC, along with the developed underlying rckskel algorithmic skeletons library, is available via GitHub.

Comparison of phoswich and ARSA-type detectors for radioxenon measurements

International Nuclear Information System (INIS)

Ward, R.M.; Biegalski, S.R.F.; Hennig, W.

2009-01-01

The monitoring of atmospheric radioxenon to ensure compliance with the Comprehensive Nuclear Test Ban Treaty (CTBT) has driven the development of improved detectors for measuring xenon, including the development of a phoswich detector. This detector uses only one PMT to detect β-γ coincidence, thus greatly reducing the bulk and electronics of the detector in comparison to the ARSA-type detector. In this experiment, 135 Xe was produced through neutron activation and a phoswich detector was used to attain spectra from the gas. These results were compared to similar results from an ARSA-type β-γ coincidence spectrum. The spectral characteristics and resolution were compared for the coincidence and beta spectra. Using these metrics, the overall performance of the phoswich detector for β-γ coincidence of radioxenon was evaluated. (author)

First order study for an iron core OH system for TNS

International Nuclear Information System (INIS)

Ballou, J.K.; Schultz, J.

1977-01-01

A simple comparison has been made between an air core and an iron core ohmic heating system for a particular device, and it was shown that the peak power requirements can be substantially reduced by the use of an iron core to power levels handled by industry today. It was also shown that for an ohmic heating system initiated plasma that the cost of the iron core ohmic heating power system (iron core, dual rectifier, and DC switch) is less than the cost for a subset of the power system for an air core system (dual rectifier and DC switch). There is considerable work being done on other methods of initiating the plasma none of which seem to be incompatible with the use of an iron core system

Comparative study of heterogeneous and homogeneous LMFBR cores in some accident conditions

International Nuclear Information System (INIS)

Renard, A.; Evrard, G.

1978-01-01

An heterogeneous design and a homogeneous one of a LMFBR core with the same power and similar dimensions are compared from the safety point-of-view. The comparison is performed for several accident conditions, such as Loss-of-Flow and Transient Overpower, with the same failure criteria and model assumptions for both cores. Qualitative trends are deduced from the behaviour of the core designs in the investigated transient conditions. (author)

Comparison of Thrust Characteristics in Pencil Sized Cylinder-type Linear Motors with Different Magnet Arrays

OpenAIRE

Nakaiwa, K; Yamada, A; Tashiro, K; Wakiwaka, H

2009-01-01

From a strong demand on the miniaturization of a chip mounter or a semiconductor device, the thrust improvement considering the magnets arrangement is studied. We accept a core stator with a Halbach type magnet array for a current linear motor. The thrust characteristics are compared with two kinds of mover, a NS magnet array and a Halbach magnet array.

Benign papilloma diagnosed on image-guided 14 G core biopsy of the breast: Effect of lesion type on likelihood of malignancy at excision

International Nuclear Information System (INIS)

Maxwell, A.J.; Mataka, G.; Pearson, J.M.

2013-01-01

Aim: To ascertain the negative predictive value (NPV) for atypia and malignancy of 14 G core biopsy of papillomas and to determine whether lesion type influences the likelihood of malignancy at lesion excision. Materials and methods: Ninety-six lesions with a 14 G core biopsy diagnosis of benign papilloma without atypia in 95 women were included. The imaging features (mass or microcalcification), biopsy mode, and number of core samples taken were documented. All patients subsequently underwent lesion excision with either extensive vacuum-assisted biopsy (VAB; 72 lesions) or surgery (24 lesions). Mammographic follow-up of at least 2 years was available for 32 lesions that were benign at VAB. Results: Atypia or malignancy was found more commonly in association with microcalcification (six of 29 lesions: 21%; median number of nine 14 G cores) than a mass (five of 67 lesions: 7%; median number of three 14 G cores), although the difference does not reach statistical significance (p = 0.088). The NPV of a 14 G core biopsy diagnosis of papilloma for atypia or malignancy is 89% (85/96). Disease underestimation may be more common in microcalcification lesions despite the greater number of cores obtained. Conclusion: Excision (using VAB or surgically) of all papillomas diagnosed as benign on 14 G needle core biopsy is recommended. Surgery may be more appropriate than VAB for some microcalcification lesions unless they are small and can be confidently removed in their entirety using VAB

Performance evaluation of CPF shredder type mechanical crusher with simulated core fuel pin

International Nuclear Information System (INIS)

Nakahara, Masaumi; Sano, Yuichi; Aose, Shin-ichi

2006-12-01

In the advanced aqueous reprocessing system, powder fuel dissolution has been investigated, which is quite effective on the dissolution for highly concentrated solution. As one of the effective means that powder the irradiated MOX fuel, we have been developing shredder type mechanical crusher. This apparatus can automatically crush the sheared fuel pieces by twin-shaft disk blades, powder the crushed fragments by disk blades and screen blade, and recover the powdered fuel. The shredder type mechanical crusher was developed for using in a hot cell in Chemical Processing Facility, and the first crush experiment with this crusher was carried out at July 2004 using the simulated core fuel pin. This experiment showed that the crushed fragments could not be grinded efficiency because screen blade vibrated up and down during the operation. Additionally, the strength of screen blade block was insufficient to crush the sheared fuel pieces stably. Therefore, about 70% of fuel was recovered in maximum. Based on the results of the first experiment, screen blade was fixed up mainly and the second experiment was carried out with improved apparatus at September 2005. In this experiment, about 96% of fuel could be recovered in maximum because screen blade was stable during the operation. (J.P.N.)

Reference Monte Carlo calculations of Maria reactor core

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.

2002-01-01

The reference Monte Carlo calculations of MARIA reactor core have been carried to evaluate accuracy of the calculations at each stage of its neutron-physics analysis using deterministic codes. The elementary cell has been calculated with two main goals; evaluation of effects of simplifications introduced in deterministic lattice spectrum calculations by the WIMS code and evaluation of library data in recently developed WIMS libraries. In particular the beryllium data of those libraries needed evaluation. The whole core calculations mainly the first MARIA critical experiment and the first critical core after the 8-year break in operation. Both cores contained only fresh fuel elements but only in the first critical core the beryllium blocks were not poisoned by Li-6 and He-3. Thus the MCNP k-eff results could be compared with the experiment. The MCNP calculations for the cores with beryllium poisoned suffered the deficiency of uncertainty in the poison concentration, but a comparison of power distribution shows that realistic poison levels have been carried out for the operating reactor MARIA configurations. (author)

Continuous-flow leaching studies of crushed and cored SYNROC

International Nuclear Information System (INIS)

Coles, D.G.; Bazan, F.

1982-01-01

Both crushed (150- to 300- μm) and cored (1,8- mm-diam) samples of SYNROC have been leached with single-pass continuous-flow leaching equipment. Crushed samples of cesium-hollandite were also leached in a similar experiment. Temperatures used were 25 0 and 75 0 C and leachates were 0.03 N NaHCO 3 and distilled water. Leaching rates from SYNROC-C were ranked cesium > strontium greater than or equal to calcium > barium > zirconium. A comparison of leaching rates is made between crushed SYNROC, cored SYNROC, and Pacific Northwest Laboratory 76-68 glass beads. This comparison depends on how the surface areas are determined for each sample. Based on geometric surface areas for SYNROC cores and glass beads, cesium leach rates from SYNROC compare well with both sodium and neptunium leached from the glass. The other elements leached from SYNROC are lower than sodium and neptunium leached from glass. They also vary for each element, while glass shows nearly the same leach rate for both sodium and neptunium

Continuous-flow leaching studies of crushed and cored SYNROC

International Nuclear Information System (INIS)

Coles, D.G.; Bazan, F.

1981-01-01

Both crushed (150 to 300 μm) and cored (1.8 mm diameter) samples of SYNROC have been leached with the single-pass continuous-flow leaching equipment. Crushed samples of Cs-hollandite were also leached in a similar experiment. Temperatures used were 25 and 75 0 C and leachates were 0.03 N NaHCO 3 and distilled water. Leaching rates from SYNROC C were ranked Cs > Sr greater than or equal to Ca > Ba > Zr. A comparison of leaching rates is made between crushed SYNROC, cored SYNROC, and PNL 76-68 glass beads. This comparison depends on how the surface areas are determined for each sample. Based on geometric surface areas for SYNROC cores and glass beads Cs leach rates from SYNROC compare well with both Na and Np leached from the glass. The other elements leached from SYNROC are lower than Na and Np leached from glass. They also vary for each element while glass shows nearly the same leach rate for both Na and Np

Development and verification of a three-dimensional core model for WWR type reactors and its coupling with the accident code ATHLET. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Lucas, D.; Mittag, S.; Rohde, U.

1995-04-01

The main goal of the project was the coupling of the 3D core model DYN3D for Russian VVER-type reactors, which has been developed in the RCR, with the thermohydraulic code ATHLET. The coupling has been realized on two basically different ways: - The implementation of only the neutron kinetics model of DYN3D into ATHLET (internal coupling), - the connection of the complete DYN3D core model including neutron kinetics, thermohydraulics and fuel rod model via data interfaces at the core top and bottom (external coupling). For the test of the coupling, comparative calculations between internal and external coupling versions have been carried out for a LOCA and a reactivity transient. Complementary goals of the project were: - The development of a DYN3D version for burn-up calculations, - the verification of DYN3D on benchmark tasks and experimental data on fuel rod behaviour, - a study on the extension of the neutron-physical data base. The project contributed to the development of advanced tools for the safety analysis of VVER-type reactors. Future work is aimed to the verification of the coupled code complex DYN3D-ATHLET. (orig.) [de

Wavelet Types Comparison for Extracting Iris Feature Based on Energy Compaction

Science.gov (United States)

Rizal Isnanto, R.

2015-06-01

Human iris has a very unique pattern which is possible to be used as a biometric recognition. To identify texture in an image, texture analysis method can be used. One of method is wavelet that extract the image feature based on energy. Wavelet transforms used are Haar, Daubechies, Coiflets, Symlets, and Biorthogonal. In the research, iris recognition based on five mentioned wavelets was done and then comparison analysis was conducted for which some conclusions taken. Some steps have to be done in the research. First, the iris image is segmented from eye image then enhanced with histogram equalization. The features obtained is energy value. The next step is recognition using normalized Euclidean distance. Comparison analysis is done based on recognition rate percentage with two samples stored in database for reference images. After finding the recognition rate, some tests are conducted using Energy Compaction for all five types of wavelets above. As the result, the highest recognition rate is achieved using Haar, whereas for coefficients cutting for C(i) < 0.1, Haar wavelet has a highest percentage, therefore the retention rate or significan coefficient retained for Haaris lower than other wavelet types (db5, coif3, sym4, and bior2.4)

Fuel pin failure root causes and power distribution gradients in WWER cores

International Nuclear Information System (INIS)

Mikus, J.

2008-01-01

The purpose of this work is to investigate the influence of some core heterogeneities and reactor construction materials on space power distribution in WWER type cores, especially from viewpoint of the values and gradient occurrence that could result in static loads with some consequences, e.g., fuel pin (FP) or fuel assembly (FA) bowing and possible contribution to the FP failure root causes. Presented information were obtained by means of experiments on research reactor LR-0 concerning the: 1) Power distribution estimation on pellet surface of the FPs neighbouring a FP containing gadolinium (Gd 2 O 3 ) burnable absorber integrated into fuel in WWER-440 and -1000 type cores; 2) Power distribution measurement in periphery FAs neighbouring the baffle in WWER-1000 type cores and 3) Power distribution in FAs neighbouring the control rod absorbing part in a WWER-440 type core. (author)

Kinematically Decoupled Cores in Dwarf (Elliptical) Galaxies

NARCIS (Netherlands)

Toloba, E.; Peletier, R. F.; Guhathakurta, P.; van de Ven, G.; Boissier, S.; Boselli, A.; Brok, M. d.; Falcón-Barroso, J.; Hensler, G.; Janz, J.; Laurikainen, E.; Lisker, T.; Paudel, S.; Ryś, A.; Salo, H.

An overview is given of what we know about the frequency of kinematically decoupled cores in dwarf elliptical galaxies. New observations show that kinematically decoupled cores happen just as often in dwarf elliptical as in ordinary early-type galaxies. This has important consequences for the

SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

International Nuclear Information System (INIS)

Molnar, J.; Vocka, R.

2010-01-01

In this work authors present 12 years of operation experience of core monitoring and surveillance system with advanced capabilities on nuclear power plants on 6 unit of VVER-440 type of reactors at two different NPPs. The original version of the SCORPIO (Surveillance of reactor CORe by PIcture On-line display) system was developed for the western type of PWR reactors. The first version of the SCORPIO-VVER Core Monitoring System for Dukovany NPP (VVER-440 type of reactor, Czech Republic) was developed in 1998. For SCORPIO-VVER implementation at Bohunice NPP in Slovakia (2001) the system was enhanced with startup module KRITEX.

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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...

Development of conceptual nuclear design of 10MWt research reactor core

International Nuclear Information System (INIS)

Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

2008-03-01

KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

Development of adaptive core emulator for PMS-XRBP of CE type plant

International Nuclear Information System (INIS)

Song, Jae Seung; Zee, Sung Quun; Lee, Chung Chan; Lee, Ki Bog; Rhy, Hyo Sang; Chang, Jong Hwa; Lee, Young Ouk; Baek, Seung Min; Seo, Ho Joon.

1996-12-01

The purpose of this report is to develop ONED-based adaptive core emulator (ACE) for Korean Standard Nuclear Power Plant. This report is first year report and includes 1) augmentation of ONED94 I/O system 2) non-equilibrium xenon initialization for core transient simulation 3) ONED94 verification via plant measurements 4) automatic data link system from PMS and personal computer. (author). 4 tabs., 4 figs., 8 refs

Development of adaptive core emulator for PMS-XRBP of CE type plant

Energy Technology Data Exchange (ETDEWEB)

Song, Jae Seung; Zee, Sung Quun; Lee, Chung Chan; Lee, Ki Bog; Rhy, Hyo Sang; Chang, Jong Hwa; Lee, Young Ouk; Baek, Seung Min; Seo, Ho Joon

1996-12-01

The purpose of this report is to develop ONED-based adaptive core emulator (ACE) for Korean Standard Nuclear Power Plant. This report is first year report and includes (1) augmentation of ONED94 I/O system (2) non-equilibrium xenon initialization for core transient simulation (3) ONED94 verification via plant measurements (4) automatic data link system from PMS and personal computer. (author). 4 tabs., 4 figs., 8 refs.

Neutronic evolution of SENA reactor during the first and second cycles. Comparison between the experimental power distributions obtained from the in-core instrumentation evaluation code CIRCE and the theoretical power values computed with the two-dimensional diffusion-evolution code EVOE

International Nuclear Information System (INIS)

Andrieux, Chantal

1976-03-01

The neutronic evolution of the reacteur Sena during the first and second cycles is presented. The experimental power distributions, obtained from the in-core instrumentation evaluation code CIRCE are compared with the theoretical powers calculated with the two-dimensional diffusion-evolution code EVOE. The CIRCE code allows: the study of the evolution of the principal parameters of the core, the comparison of the results of measured and theoretical estimates. Therefore this study has a great interest for the knowledge of the neutronic evolution of the core, as well as the validation of the refinement of theoretical estimation methods. The core calculation methods and requisite data for the evaluation of the measurements are presented after a brief description of the SENA core and its inner instrumentation. The principle of the in-core instrumentation evaluation code CIRCE, and calculation of the experimental power distributions and nuclear core parameters are then exposed. The results of the evaluation are discussed, with a comparison of the theoretical and experimental results. Taking account of the approximations used, these results, as far as the first and second cycles at SENA are concerned, are satisfactory, the deviations between theoretical and experimental power distributions being lower than 3% at the middle of the reactor and 9% at the periphery [fr

Protein homology network families reveal step-wise diversification of Type III and Type IV secretion systems.

Directory of Open Access Journals (Sweden)

Duccio Medini

2006-12-01

Full Text Available From the analysis of 251 prokaryotic genomes stored in public databases, the 761,260 deduced proteins were used to reconstruct a complete set of bacterial proteic families. Using the new Overlap algorithm, we have partitioned the Protein Homology Network (PHN, where the proteins are the nodes and the links represent homology relationships. The algorithm identifies the densely connected regions of the PHN that define the families of homologous proteins, here called PHN-Families, recognizing the phylogenetic relationships embedded in the network. By direct comparison with a manually curated dataset, we assessed that this classification algorithm generates data of quality similar to a human expert. Then, we explored the network to identify families involved in the assembly of Type III and Type IV secretion systems (T3SS and T4SS. We noticed that, beside a core of conserved functions (eight proteins for T3SS, seven for T4SS, a variable set of accessory components is always present (one to nine for T3SS, one to five for T4SS. Each member of the core corresponds to a single PHN-Family, while accessory proteins are distributed among different pure families. The PHN-Family classification suggests that T3SS and T4SS have been assembled through a step-wise, discontinuous process, by complementing the conserved core with subgroups of nonconserved proteins. Such genetic modules, independently recruited and probably tuned on specific effectors, contribute to the functional specialization of these organelles to different microenvironments.

Intercomparison of liquid metal fast reactor seismic analysis codes. V. 3: Comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Proceedings of a final research co-ordination meeting

International Nuclear Information System (INIS)

1996-05-01

This publication contains the final papers summarizing the validation of the codes on the basis of comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Refs, figs tabs

Experimental constraints on light elements in the Earth’s outer core

OpenAIRE

Youjun Zhang; Toshimori Sekine; Hongliang He; Yin Yu; Fusheng Liu; Mingjian Zhang

2016-01-01

Earth?s outer core is liquid and dominantly composed of iron and nickel (~5?10?wt%). Its density, however, is ~8% lower than that of liquid iron, and requires the presence of a significant amount of light element(s). A good way to specify the light element(s) is a direct comparison of density and sound velocity measurements between seismological data and those of possible candidate compositions at the core conditions. We report the sound velocity measurements of a model core composition in th...

The Inhibitory Core of the Myostatin Prodomain: Its Interaction with Both Type I and II Membrane Receptors, and Potential to Treat Muscle Atrophy.

Science.gov (United States)

Ohsawa, Yutaka; Takayama, Kentaro; Nishimatsu, Shin-ichiro; Okada, Tadashi; Fujino, Masahiro; Fukai, Yuta; Murakami, Tatsufumi; Hagiwara, Hiroki; Itoh, Fumiko; Tsuchida, Kunihiro; Hayashi, Yoshio; Sunada, Yoshihide

2015-01-01

Myostatin, a muscle-specific transforming growth factor-β (TGF-β), negatively regulates skeletal muscle mass. The N-terminal prodomain of myostatin noncovalently binds to and suppresses the C-terminal mature domain (ligand) as an inactive circulating complex. However, which region of the myostatin prodomain is required to inhibit the biological activity of myostatin has remained unknown. We identified a 29-amino acid region that inhibited myostatin-induced transcriptional activity by 79% compared with the full-length prodomain. This inhibitory core resides near the N-terminus of the prodomain and includes an α-helix that is evolutionarily conserved among other TGF-β family members, but suppresses activation of myostatin and growth and differentiation factor 11 (GDF11) that share identical membrane receptors. Interestingly, the inhibitory core co-localized and co-immunoprecipitated with not only the ligand, but also its type I and type II membrane receptors. Deletion of the inhibitory core in the full-length prodomain removed all capacity for suppression of myostatin. A synthetic peptide corresponding to the inhibitory core (p29) ameliorates impaired myoblast differentiation induced by myostatin and GDF11, but not activin or TGF-β1. Moreover, intramuscular injection of p29 alleviated muscle atrophy and decreased the absolute force in caveolin 3-deficient limb-girdle muscular dystrophy 1C model mice. The injection suppressed activation of myostatin signaling and restored the decreased numbers of muscle precursor cells caused by caveolin 3 deficiency. Our findings indicate a novel concept for this newly identified inhibitory core of the prodomain of myostatin: that it not only suppresses the ligand, but also prevents two distinct membrane receptors from binding to the ligand. This study provides a strong rationale for the use of p29 in the amelioration of skeletal muscle atrophy in various clinical settings.

Radioimmunoassay of surface antigen and core antibody of hepatitis B virus. Comparison of kits AUSRIA/CORE; AUSRIA II-125 and CORAB

Energy Technology Data Exchange (ETDEWEB)

Kselikova, M; Urbankova, J [Ustav Hematologie a Krevni Transfuze, Prague (Czechoslovakia)

1984-08-01

The sensitivity is compared of determination of surface antigen HBsAg and nuclear antibody HBcAb of the hepatitis B virus using kits for separate (AUSRIA II-125, CORAB) and simultaneous (AUSRIAsup(R)/CORE) determinations of third generation tests in selected samples of medical personnel, HBsAg carriers, patients at a dialysis centre, blood donors and in sera of HBsAg carriers diluted in steps from 4x10/sup -2/ to 4x10/sup -7/. HBsAg is always determined using the RIA technique, HBcAb is determined using the technique of radioimmunoassay with the CORAB kit and with the AUSRIA sup(R)/CORE kit using enzymeimmunoassay. The sensitivity of determination using the AUSRIA sup(R)/CORE kit is at least as good for both investigated indicators of the hepatitis B virus and that obtained using separate determination of HBsAg (AUSRIA II-125) and HBcAb (CORAB), this also using modified photocolorimetric determination. Only one AUSRIA/sup R//CORE kit was available for the investigation and the informative character of the report is emphasized.

Neutron flux measurement in the central channel (XC-1) of TRIGA 14 MW LEU core

International Nuclear Information System (INIS)

BARBOS, D.; BUSUIOC, P.; ROTH, Cs.; PAUNOIU, C.

2008-01-01

The TRIGA 14 MW reactor, operated by Institute for Nuclear Research Pitesti, Romania, is a pool type reactor, and has a rectangular shape which holds fuel bundles and is surrounded with beryllium reflectors. Each fuel bundle is composed of 25 nuclear fuel rods. The TRIGA 14 MW reactor was commissioned 28 years ago with HEU fuel rods. The conversion was gradually achieved, starting in February 1992 and completed in March 2006. The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel, replaced by LEU fuel, accompanied by a large set of theoretical evaluation and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC in the XC-1 water 1 water-filled channel were performed using multi multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral neutron flux value for LEU core is greater of 13% than for the standard HEU core. Also thermal neutron flux value for converted LEU core is smaller by 0.38% than for the standard HEU core. These differences appear because the foil activation detectors have been irradiated using a pneumatic rabbit having a diameter of 32 mm, whereas foil irradiations in standard HEU core has been performed with a pneumatic rabbit having a diameter of 14 mm, and therefore the neutron spectra in LEU core is less thermalized and the weight of fast neutron is greater

Synthesis of parallel and antiparallel core-shell triangular nanoparticles

Science.gov (United States)

Bhattacharjee, Gourab; Satpati, Biswarup

2018-04-01

Core-shell triangular nanoparticles were synthesized by seed mediated growth. Using triangular gold (Au) nanoparticle as template, we have grown silver (Ag) shellto get core-shell nanoparticle. Here by changing the chemistry we have grown two types of core-shell structures where core and shell is having same symmetry and also having opposite symmetry. Both core and core-shell nanoparticles were characterized using transmission electron microscopy (TEM) and energy dispersive X-ray spectroscopy (EDX) to know the crystal structure and composition of these synthesized core-shell nanoparticles. From diffraction pattern analysis and energy filtered TEM (EFTEM) we have confirmed the crystal facet in core is responsible for such two dimensional growth of core-shell nanostructures.

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Science.gov (United States)

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

2017-11-01

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

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)

Thermal neutron flux distribution in the ET R R-1 reactor core as experimentally measured and theoretically calculated by the code triton

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

1995-10-01

Thermal neutron flux distributions that were measured earlier at the ET-R R-1 reactor are compared with those calculated by the three dimensional diffusion code Triton. This comparison was made for the horizontal and vertical flux distributions. The horizontal thermal flux distributions considered in this comparison were along the core diagonals at two planes of different heights from core bottom, where one at a level passing through the control rod at core center and the other at a level below this control rod. In the meantime all the control rods were taken into consideration. The effect of the existence of a water cavity inside the core as well as the influence of the control rods on the thermal flux are illustrated in this work. The vertical thermal flux distributions considered in the comparison were at two positions in core namely; one along the core height the horizontal reactor power distribution along the core height and the horizontal reactor power distribution along the core diagonal as calculated by the code Triton are also given this work. 8 figs., 1 tab.

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)

First identification and characterization of Borrobol-type tephra in the Greenland ice cores

DEFF Research Database (Denmark)

Cook, Eliza; Davies, Siwan M.; Guðmundsdóttir, Esther R.

2018-01-01

in the ice-cores or that it relates to just one of the ice-core events. A firm correlation cannot be established at present due to their strong geochemical similarities. The older tephra horizon, found within all three ice-cores and dated to 17326 ± 319 a b2k, can be correlated to a known layer within marine....... The older deposit is consistent with BT age estimates derived from Scottish sites, while the younger deposit overlaps with both BT and PT age estimates. We suggest that either the BT in Northern European terrestrial sequences represents an amalgamation of tephra from both of the GI-1e events identified...

Interrelating the breakage and composition of mined and drill core coal

Science.gov (United States)

Wilson, Terril Edward

Particle size distribution of coal is important if the coal is to be beneficiated, or if a coal sales contract includes particle size specifications. An exploration bore core sample of coal ought to be reduced from its original cylindrical form to a particle size distribution and particle composition that reflects, insofar as possible, a process stream of raw coal it represents. Often, coal cores are reduced with a laboratory crushing machine, the product of which does not match the raw coal size distribution. This study proceeds from work in coal bore core reduction by Australian investigators. In this study, as differentiated from the Australian work, drop-shatter impact breakage followed by dry batch tumbling in steel cylinder rotated about its transverse axis are employed to characterize the core material in terms of first-order and zeroth-order breakage rate constants, which are indices of the propensity of the coal to degrade during excavation and handling. Initial drop-shatter and dry tumbling calibrations were done with synthetic cores composed of controlled low-strength concrete incorporating fly ash (as a partial substitute for Portland cement) in order to reduce material variables and conserve difficult-to-obtain coal cores. Cores of three different coalbeds--Illinois No. 6, Upper Freeport, and Pocahontas No. 5 were subjected to drop-shatter and dry batch tumbling tests to determine breakage response. First-order breakage, characterized by a first-order breakage index for each coal, occurred in the drop-shatter tests. First- and zeroth-order breakage occurred in dry batch tumbling; disappearance of coarse particles and creation of fine particles occurred in a systematic way that could be represented mathematically. Certain of the coal cores available for testing were dry and friable. Comparison of coal preparation plant feed with a crushed bore core and a bore core prepared by drop-shatter and tumbling (all from the same Illinois No.6 coal mining

Finite element program ARKAS: verification for IAEA benchmark problem analysis on core-wide mechanical analysis of LMFBR cores

International Nuclear Information System (INIS)

Nakagawa, M.; Tsuboi, Y.

1990-01-01

''ARKAS'' code verification, with the problems set in the International Working Group on Fast Reactors (IWGFR) Coordinated Research Programme (CRP) on the inter-comparison between liquid metal cooled fast breeder reactor (LMFBR) Core Mechanics Codes, is discussed. The CRP was co-ordinated by the IWGFR around problems set by Dr. R.G. Anderson (UKAEA) and arose from the IWGFR specialists' meeting on The Predictions and Experience of Core Distortion Behaviour (ref. 2). The problems for the verification (''code against code'') and validation (''code against experiment'') were set and calculated by eleven core mechanics codes from nine countries. All the problems have been completed and were solved with the core structural mechanics code ARKAS. Predictions by ARKAS agreed very well with other solutions for the well-defined verification problems. For the validation problems based on Japanese ex-reactor 2-D thermo-elastic experiments, the agreements between measured and calculated values were fairly good. This paper briefly describes the numerical model of the ARKAS code, and discusses some typical results. (author)

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Characteristics of core sampling from crumbing Paleozoic rock

Energy Technology Data Exchange (ETDEWEB)

Barabashkin, I I; Edelman, Y A; Filippov, V N; Lychev, V N

1981-01-01

The results of analysis of core sampling using standard core sampling tools with small and medium inside diameter are cited. It is demonstrated that when using these tools loss of core in Paleozoic deposits promising with regard to oil and gas content does not exceed 25 - 30%. The use of a new core sampling tool with a large inside diameter which includes drill bits of different types and a core lifter ''Krembriy'' SKU-172/100 made it possible to increase core removal approximately 52%. A representative core from a highly crumbling and vesicular rock belinging to groups III - IV in terms of difficulty of core sampling was obtained first. A description of a new core sampling tool is given. The characteristics of the technology of its use which promote preservation of the core are cited. Means of continued improvement of this tool are noted.

Performance of single wire earth return transformers with amorphous alloy core in a rural electric energy distribution system

Directory of Open Access Journals (Sweden)

Benedito Antonio Luciano

2012-10-01

Full Text Available In this paper are presented some considerations about the performance of single wire earth return amorphous alloy core transformers in comparison with conventional silicon steel sheets cores transformers used in rural electric energy distribution network. It has been recognized that amorphous metal core transformers improve electrical power distribution efficiency by reducing transformer core losses. This reduction is due to some electromagnetic properties of the amorphous alloys such as: high magnetic permeability, high resistivity, and low coercivity. Experimental results obtained with some single-phase, 60 Hz, 5 kVA amorphous core transformers installed in a rural area electric distribution system in Northern Brazil have been confirming their superior performance in comparison to identical nominal rated transformers built with conventional silicon steel cores, particularly with regard to the excitation power and to the no-load losses.

Post-processor for simulations of the ORIGEN program and calculation of the composition of the activity of a burnt fuel core by a BWR type reactor

International Nuclear Information System (INIS)

Sandoval V, S.

2006-01-01

The composition calculation and the activity of nuclear materials subject to processes of burnt, irradiation and decay periods are of utility for diverse activities inside the nuclear industry, as they are it: the processes design and operations that manage radioactive material, the calculation of the inventory and activity of a core of burnt nuclear fuel, for studies of type Probabilistic Safety Analysis (APS), as well as for regulation processes and licensing of nuclear facilities. ORIGEN is a program for computer that calculates the composition and the activity of nuclear materials subject to periods of burnt, irradiation and decay. ORIGEN generates a great quantity of information whose processing and analysis are laborious, and it requires thoroughness to avoid errors. The automation of the extraction, conditioning and classification of that information is of great utility for the analyst. By means of the use of the post-processor presented in this work it is facilitated, it speeds up and wide the capacity of analysis of results, since diverse consultations with several classification options and filtrate of results can be made. As illustration of the utility of the post-processor, and as an analysis of interest for itself, it is also presented in this work the composition of the activity of a burned core in a BWR type reactor according to the following classification criteria: by type of radioisotope (fission products, activation products and actinides), by specie type (gassy, volatile, semi-volatile and not volatile), by element and by chemical group. The results show that the total activity of the studied core is dominated by the fission products and for the actinides, in proportion four to one, and that the gassy and volatile species conform a fifth part of the total activity of the core. (Author)

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.

SIMULATION OF CHARACTERISTICS OF DUAL-CORE PHASE SHIFTING TRANSFORMER

Directory of Open Access Journals (Sweden)

Kalinin L.P.

2014-04-01

Full Text Available The role and importance of phase shifting transformers are increased as a result of the further development of integrated power systems. This gives the rise to new technical solutions which entails the necessity of comparison of new developments with existing. The article consider the technical characteristics of dual-core phase shifting transformer which later will be used as a basis for comparison with other competing options and assess of their technical efficiency.

Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

Energy Technology Data Exchange (ETDEWEB)

Stauff, N.E.; Klim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States); Fiorina, C. [Politecnico di Milano, Milan (Italy); Franceschini, F. [Westinghouse Electric Company LLC., Cranberry Township, Pennsylvania (United States)

2013-07-01

A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic

Assessment of assembly homogenized two-steps core dynamic calculations using direct whole core transport solutions

International Nuclear Information System (INIS)

Hursin, Mathieu; Downar, Thomas J.; Yoon, Joo Il; Joo, Han Gyu

2016-01-01

Highlights: • Reactivity initiated accident analysis with direct whole core transient transport code. • Comparison with usual “two steps” procedure. • Effect of effective delayed neutron fraction definition on energy deposition in the fuel. • Effect of homogenized few-group cross sections generation at the assembly level on energy deposition in the fuel. • Effect of effective fuel temperature definition on energy deposition in the fuel. - Abstract: The impact of the approximations in the “two-steps” procedure used in the current generation of nodal simulators for core transient calculations is assessed by using a higher order solution obtained from a direct, whole core, transient transport calculation. A control rod ejection accident in an idealized minicore is analyzed with PARCS, which uses the two-steps procedure and DeCART which provides the higher order solution. DeCART is used as lattice code to provide the homogenized cross sections and kinetics parameters to PARCS. The approximations made by using (1) the homogenized few-group cross sections and kinetic parameters generated at the assembly level, (2) an effective delayed neutrons fraction, (3) an effective fuel temperature and (4) the few-group formulation are investigated in terms of global and local core power behavior. The results presented in the paper show that the current two-steps procedure produces sufficiently accurate transient results with respect to the direct whole core calculation solution, provided that its parameters are carefully generated using the prescriptions described in the present article.

Overview of the design of core restraint systems

International Nuclear Information System (INIS)

Heinecke, J.

1984-01-01

The optimization of the core restraint system is an important condition for the safe and reliable operation of a fast breeder reactor. For KNK II which is under successful operation and SNR 300 all requirements for safety and operation have been met with help of a ring type system. For SNR 2 the decision between the ring type system and the free standing core has to be done in the near future. Within these considerations the advantages of a ring type restraint system of limiting deflections during operation and limiting of possible movements under seismic conditions have to be balanced against the somewhat more complicated structure of the ring type restraint system

Nuclear characteristic simulation device for reactor core

International Nuclear Information System (INIS)

Arakawa, Akio; Kobayashi, Yuji.

1994-01-01

In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

Rhometal interface in pseudo-core shell powders like Permalloy/Rhometal type

Energy Technology Data Exchange (ETDEWEB)

Chicinaş, I.; Marinca, T.F.; Popa, F.; Neamţu, B.V.

2015-12-15

Highlights: • Pseudo-core shell powders like Permalloy/Rhometal type obtained by microalloying. • During annealing, by interdiffusion, Rhometal phase is formed at the interface. • Both bcc and fcc structures of the Rhometal have been evidenced in interface. - Abstract: The nanocrystalline Ni{sub 3}Fe (around Permalloy composition) powders were prepared by dry mechanical alloying. The nanocrystalline Ni{sub 3}Fe and carbonyl Fe mixture powders and green compacts have been subjected to heat treatments in an argon atmosphere in order to obtain pseudo-core-shell like particles by micro-alloying in the temperature range of 400–900 °C. The large Permalloy particles are partially covered by very small Fe particles and at the interface a layer of Rhometal is formed by micro-alloying. The Permalloy particles remain in the nanocrystalline/nanostructured state after the annealing independent on the annealing temperature up to 900 °C. Structural, microstructural characterisation and local elemental chemical analysis have been performed by X-ray diffraction, scanning electron microscopy and X-ray microanalysis. The Rhometal interface was studied and evidenced by Fe and Ni concentration profile (EDX microanalysis) and X-ray diffraction. It was found that by a heat treatment up to 900 °C the interface is in the iron zone and at the heat treatment temperature of 900 °C the interface is in both Permalloy and iron zones. By XRD the bcc and fcc structures of the Rhometal have been evidenced. The interface length in iron zones is about 0.8 μm for the heat treatment at 400 °C and reaches 2.5 μm for heat treatment at 900 °C. The interface reaches 10 μm for a temperature of a heat treatment of 900 °C. The iron zones welded by Permalloy zones is evidenced the presence of the Ni atoms up to 4.7 at%. Also, in Permalloy particles are evidenced two zones with a different amount of Ni: one around Ni{sub 0.6}Fe{sub 0.4} composition close to the interface and the second in the

Hypernuclear properties derived from the Nijmegen soft-core OBE potential

International Nuclear Information System (INIS)

Yamamoto, Yasuo; Bando, Hiroharu.

1990-01-01

The Nijmegen soft-core YN potential is applied to the G-matrix calculation in nuclear matter, characteristics of which are investigated in comparison with the hard-core models D and F. The ΛN G-matrix interaction is simulated in a three-range Gaussian form and applied to various hypernuclear calculations. Λ binding energies in ground and excited states are wholly reproduced from light to medium heavy hypernuclei observed in experiments. (author)

Medullary neurons in the core white matter of the olfactory bulb: a new cell type.

Science.gov (United States)

Paredes, Raúl G; Larriva-Sahd, Jorge

2010-02-01

The structure of a new cell type, termed the medullary neuron (MN) because of its intimate association with the rostral migratory stream (RMS) in the bulbar core, is described in the adult rat olfactory bulb. The MN is a triangular or polygonal interneuron whose soma lies between the cellular clusters of the RMS or, less frequently, among the neuron progenitors therein. MNs are easily distinguished from adjacent cells by their large size and differentiated structure. Two MN subtypes have been categorized by the Golgi technique: spiny pyramidal neurons and aspiny neurons. Both MN subtypes bear a large dendritic field impinged upon by axons in the core bulbar white matter. A set of collaterals from the adjacent axons appears to terminate on the MN dendrites. The MN axon passes in close apposition to adjacent neuron progenitors in the RMS. MNs are immunoreactive with antisera raised against gamma-aminobutyric acid and glutamate decarboxylase 65/67. Electron-microscopic observations confirm that MNs correspond to fully differentiated, mature neurons. MNs seem to be highly conserved among macrosmatic species as they occur in Nissl-stained brain sections from mouse, guinea pig, and hedgehog. Although the functional role of MNs remains to be determined, we suggest that MNs represent a cellular interface between endogenous olfactory activity and the differentiation of new neurons generated during adulthood.

Reload core safety verification

International Nuclear Information System (INIS)

Svetlik, M.; Minarcin, M.

2003-01-01

This paper presents a brief look at the process of reload core safety evaluation and verification in Slovak Republic. It gives an overview of experimental verification of selected nuclear parameters in the course of physics testing during reactor start-up. The comparison of IAEA recommendations and testing procedures at Slovak and European nuclear power plants of similar design is included. An introduction of two level criteria for evaluation of tests represents an effort to formulate the relation between safety evaluation and measured values (Authors)

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

Determining the magnetically nonlinear characteristics of a three phase core-type power transformer

International Nuclear Information System (INIS)

Dolinar, Matjaz; Stumberger, Gorazd; Polajzer, Bostjan; Dolinar, Drago

2006-01-01

This paper presents nonlinear iron core model of a three-phase, three-limb power transformer which is given by the current-dependant characteristics of flux linkages. The magnetically nonlinear characteristics are determined by controlled magnetic excitation of all three limbs which allows to take into account the variable magnetic-cross couplings between different coils placed on limbs, caused by saturation. The corresponding partial derivatives of measured flux linkage characteristics are used in the transformer circuit model as a magnetically nonlinear iron core model in order to analyze the behaviour of a nonsymmetrically excited transformer. Numerical results using transformer model with the determined iron core model agree very well with the measured results

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)

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)

An iterative homogenization technique that preserves assembly core exchanges

International Nuclear Information System (INIS)

Mondot, Ph.; Sanchez, R.

2003-01-01

A new interactive homogenization procedure for reactor core calculations is proposed that requires iterative transport assembly and diffusion core calculations. At each iteration the transport solution of every assembly type is used to produce homogenized cross sections for the core calculation. The converged solution gives assembly fine multigroup transport fluxes that preserve macro-group assembly exchanges in the core. This homogenization avoids the periodic lattice-leakage model approximation and gives detailed assembly transport fluxes without need of an approximated flux reconstruction. Preliminary results are given for a one-dimensional core model. (authors)

Enhanced Core Noise Modeling for Turbofan Engines

Science.gov (United States)

Stone, James R.; Krejsa, Eugene A.; Clark, Bruce J.

2011-01-01

This report describes work performed by MTC Technologies (MTCT) for NASA Glenn Research Center (GRC) under Contract NAS3-00178, Task Order No. 15. MTCT previously developed a first-generation empirical model that correlates the core/combustion noise of four GE engines, the CF6, CF34, CFM56, and GE90 for General Electric (GE) under Contract No. 200-1X-14W53048, in support of GRC Contract NAS3-01135. MTCT has demonstrated in earlier noise modeling efforts that the improvement of predictive modeling is greatly enhanced by an iterative approach, so in support of NASA's Quiet Aircraft Technology Project, GRC sponsored this effort to improve the model. Since the noise data available for correlation are total engine noise spectra, it is total engine noise that must be predicted. Since the scope of this effort was not sufficient to explore fan and turbine noise, the most meaningful comparisons must be restricted to frequencies below the blade passage frequency. Below the blade passage frequency and at relatively high power settings jet noise is expected to be the dominant source, and comparisons are shown that demonstrate the accuracy of the jet noise model recently developed by MTCT for NASA under Contract NAS3-00178, Task Order No. 10. At lower power settings the core noise became most apparent, and these data corrected for the contribution of jet noise were then used to establish the characteristics of core noise. There is clearly more than one spectral range where core noise is evident, so the spectral approach developed by von Glahn and Krejsa in 1982 wherein four spectral regions overlap, was used in the GE effort. Further analysis indicates that the two higher frequency components, which are often somewhat masked by turbomachinery noise, can be treated as one component, and it is on that basis that the current model is formulated. The frequency scaling relationships are improved and are now based on combustor and core nozzle geometries. In conjunction with the Task

An Overnight Comparison of Core Temperature Using a Rectal Probe and a Radio Pill

National Research Council Canada - National Science Library

Paul, Michel

1999-01-01

Previous efforts to record core temperature with radio pills produced consistent results showing that core temperature provided by radio pill tended to be lower than that provided by rectal probe by about 0.5c to o...

Review and role of plyometrics and core rehabilitation in competitive sport.

Science.gov (United States)

Hill, John; Leiszler, Matthew

2011-01-01

Core stability and plyometric training have become common elements of training programs in competitive athletes. Core stability allows stabilization of the spine and trunk of the body in order to allow maximal translation of force to the extremities. Plyometric training is more dynamic and involves explosive-strength training. Integration of these exercises theoretically begins with core stabilization using more static exercises, allowing safe and effective transition to plyometric exercises. Both core strengthening and plyometric training have demonstrated mixed but generally positive results on injury prevention rehabilitation of certain types of injuries. Improvement in performance compared to other types of exercise is unclear at this time. This article discusses the theory and strategy behind core stability and plyometric training; reviews the literature on injury prevention, rehabilitation of injury, and performance enhancement with these modalities; and discusses the evaluation and rehabilitation of core stability.

GFR fuel and core pre-conceptual design studies

International Nuclear Information System (INIS)

Chauvin, N.; Ravenet, A.; Lorenzo, D.; Pelletier, M.; Escleine, J.M.; Munoz, I.; Bonnerot, J.M.; Malo, J.Y.; Garnier, J.C.; Bertrand, F.; Bosq, J.C.

2007-01-01

The revision of the GFR core design - plate type - has been undertaken since previous core presented at Global'05. The self-breeding searched for has been achieved with an optimized design ('12/06 E'). The higher core pressure drop was a matter of concern. First of all, the core coolability in natural circulation for pressurized conditions has been studied and preliminary plant transient calculations have been performed. The design and safety criteria are met but no more margin remains. The project is also addressing the feasibility and the design of the fuel S/A. The hexagonal shape together with the principle of closed S/A (wrapper tube) is kept. Ceramic plate type fuel element combines a high enough core power density (minimization of the Pu inventory) and plutonium and minor actinides recycling capabilities. Innovative for many aspects, the fuel element is central to the GFR feasibility. It is supported already by a significant R and D effort also applicable to a pin concept that is considered as the other fuel element of interest. This combination of fuel/core feasibility and performance analysis, safety dispositions and performances analysis will compose the 'GFR preliminary feasibility' which is a project milestone at the end of the year 2007. (authors)

Comparison of DNBR estimation methods in the Westinghouse and KWU reactor cores

International Nuclear Information System (INIS)

Camargo, C.T.M.; Pontedeiro, A.C.

1984-11-01

A method for foreseeing departure from nucleate boiling phenomenon in Westinghouse reator cores (OTΔT- signal for reator shut down) is described. The results from investigations done with the OTΔT system and in the efficiency of different methods used in the Westinghouse and KWU nuclear power plants to estimate thermohydraulic conditions of the PWR reactor cores, are presented. The investigations were done, by support of computer codes. The modifications, purposed by Westinghouse, in the original project of Angra-1 OTΔT system are analysed. (M.C.K.) [pt

POEM a core instrument to measure symptoms in clinical trials: a HOME statement

OpenAIRE

Spuls, Ph.I.; Gerbens, L.A.A.; Simpson, E.; Apfelbacher, C.J.; Chalmers, J.R.; Thomas, K.S.; Prinsen, C.A.C.; Kobyletzki, L.B. von; Singh, J.A.; Williams, Hywel C.; Schmitt, J.

2016-01-01

Background: The Harmonising Outcome Measures for Eczema (HOME) initiative has defined four core outcome domains for a core outcome set (COS) to be measured in all atopic eczema (AE) trials to ensure cross-trial comparison: clinical signs, symptoms, quality of life and longterm control. Objectives: The aim of this paper is to report on the consensus process that was used to select the core instrument to consistently assess symptoms in all future AE trials. Methods: Following the HOME roa...

Electron momentum spectroscopy of the core state of solid carbon

International Nuclear Information System (INIS)

Caprari, R.S.; Clark, S.A.C.; McCarthy, I.E.; Storer, P.J.; Vos, M.; Weigold, E.

1994-08-01

Electron momentum spectroscopy (binary encounter (e,2e)) experimental results are presented for the core state of an amorphous carbon allotrope. The (e,2e) cross section has two identifiable regions. One is a narrow energy width 'core band peak' that does not disperse with momentum. At higher binding energies there is an energy diffuse 'multiple scattering continuum', which is a consequence of (e,2e) collisions with core electrons that are accompanied by inelastic scattering of one or more of the incoming or outgoing electrons. Comparisons of experimental momentum distributions with the Hartree-Fock atomic carbon ls orbital are presented for both regions. 16 refs., 4 figs

Reflood behavior at low initial clad temperature in Slab Core Test Facility Core-II

International Nuclear Information System (INIS)

Akimoto, Hajime; Sobajima, Makoto; Abe, Yutaka; Iwamura, Takamichi; Ohnuki, Akira; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Adachi, Hiromichi.

1990-07-01

In order to study the reflood behavior with low initial clad temperature, a reflood test was performed using the Slab Core Test Facility (SCTF) with initial clad temperature of 573 K. The test conditions of the test are identical with those of SCTF base case test S2-SH1 (initial clad temperature 1073 K) except the initial clad temperature. Through the comparison of results from these two tests, the following conclusions were obtained. (1) The low initial clad temperature resulted in the low differential pressures through the primary loops due to smaller steam generation in the core. (2) The low initial clad temperature caused the accumulated mass in the core to be increased and the accumulated mass in the downcomer to be decreased in the period of the lower plenum injection with accumulator (before 50s). In the later period of the cold leg injection with LPCI (after 100s), the water accumulation rates in the core and the downcomer were almost the same between both tests. (3) The low initial clad temperature resulted in the increase of the core inlet mass flow rate in the lower plenum injection period. However, the core inlet mass flow rate was almost the same regardless of the initial clad temperature in the later period of the cold leg injection period. (4) The low initial clad temperature resulted in the low turnaround temperature, high temperature rise and fast bottom quench front propagation. (5) In the region apart from the quench front, low initial clad temperature resulted in the lower heat transfer. In the region near the quench front, almost the same heat transfer coefficient was observed between both tests. (6) No flow oscillation with a long period was observed in the SCTF test with low initial clad temperature of 573 K, while it was remarkable in the Cylindrical Core Test Facility (CCTF) test which was performed with the same initial clad temperature. (J.P.N.)

Development of the Monju core safety analysis numerical models by super-COPD code

International Nuclear Information System (INIS)

Yamada, Fumiaki; Minami, Masaki

2010-12-01

Japan Atomic Energy Agency constructed a computational model for safety analysis of Monju reactor core to be built into a modularized plant dynamics analysis code Super-COPD code, for the purpose of heat removal capability evaluation at the in total 21 defined transients in the annex to the construction permit application. The applicability of this model to core heat removal capability evaluation has been estimated by back to back result comparisons of the constituent models with conventionally applied codes and by application of the unified model. The numerical model for core safety analysis has been built based on the best estimate model validated by the actually measured plant behavior up to 40% rated power conditions, taking over safety analysis models of conventionally applied COPD and HARHO-IN codes, to be capable of overall calculations of the entire plant with the safety protection and control systems. Among the constituents of the analytical model, neutronic-thermal model, heat transfer and hydraulic models of PHTS, SHTS, and water/steam system are individually verified by comparisons with the conventional calculations. Comparisons are also made with the actually measured plant behavior up to 40% rated power conditions to confirm the calculation adequacy and conservativeness of the input data. The unified analytical model was applied to analyses of in total 8 anomaly events; reactivity insertion, abnormal power distribution, decrease and increase of coolant flow rate in PHTS, SHTS and water/steam systems. The resulting maximum values and temporal variations of the key parameters in safety evaluation; temperatures of fuel, cladding, in core sodium coolant and RV inlet and outlet coolant have negligible discrepancies against the existing analysis result in the annex to the construction permit application, verifying the unified analytical model. These works have enabled analytical evaluation of Monju core heat removal capability by Super-COPD utilizing the

Paleoceanographic records in the sedimentary cores from the middle Okinawa Trough

Institute of Scientific and Technical Information of China (English)

无

2003-01-01

Two gravity piston cores (Cores 155 and 180) involved in this study were collected from the middle Okinawa Trough. Stratigraphy of the two cores was divided and classified based on the features of planktonic foraminifera oxygen isotope changes together with depositional sequence, millennium-scale climatic event comparison, carbonate cycles and AMS14C dating. Some paleoclimatic information contained in sediments of these cores was extracted to discuss the paleoclimatic change rules and the short-time scale events presented in interglacial period. Analysis on the variation of oxygen isotope values in stage two shows that the middle part of the Okinawa Trough may have been affected by fresh water from the Yellow River and the Yangtze River during the Last Glacial Maximum (LGM). The oxygen isotope value oscillating ranges of the cores have verified that the marginal sea has an amplifying effect on climate changes. The Δ13c of benthic foraminifera Uvigerina was lighter in the glacial period than that in the interglacial period, which indicates that the Paleo-Kuroshio's main stream moved eastward and its influence area decreased. According to the temperature difference during the "YD" period existing in Core 180 and other data, we can reach the conclusion that the climatic changes in the middle Okinawa Trough area were controlled by global climatic changes, but some regional factors had also considerable influence on the climate changes. Some results in this paper support Fairbanks's point that the "YD" event was a brief stagnation of sea level rising during the global warming up procession. Moreover, the falling of sea level in the glacial period weakened the exchange between the bottom water of the Okinawa Trough and the deep water of the northwestern Pacific Ocean and resulted in low oxygen state of bottom water in this area. These procedures are the reasons for carbonate cycle in the Okinawa Trough area being consistent with the "Atlantic type " carbonate cycle.

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 ...

Core arrangement in BWR type reactors

International Nuclear Information System (INIS)

Asano, Masayuki.

1981-01-01

Purpose: To decrease the number of fuel assemblies whose locations are to be changed upon fuel exchange, as well as unify the power distribution in the core by arranging, in a chess board configuration, a plurality pattern of unit reactor lattices each containing fuel assemblies of different burnup degrees in orthogonal positions to each other. Constitution: A first pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 1 and fuel assemblies of burnup degree 3 at orthogonal positions to each other. A second pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 2 and fuel assemblies of burnup degree 1 at orthogonal positions to each other. The unit lattices each in such a dispositions are arranged in a chess board arrangement. Since, the fuel assemblies of the burnup degree 1 in the first pattern unit lattices proceed to the burnup degree 2 and the fuel assemblies of the burnup degree 2 in the second pattern unit lattices proceed to the burnup degree 3 up to the fuel exchange stage, fuel exchange and movement have only to be made, not for those fuel assemblies, but for another half of the fuel assemblies. (Kawakami, Y.)

Comparison of high temperature, high frequency core loss and dynamic B-H loops of two 50 Ni-Fe crystalline alloys and an iron-based amorphous alloy

International Nuclear Information System (INIS)

Wieserman, W.R.; Schwarze, G.E.; Niedra, J.M.

1994-01-01

The availability of experimental data that characterizes the performance of soft magnetic materials for the combined conditions of high temperature and high frequency is almost non-existent. An experimental investigation was conducted over the temperature range of 23 to 300 C and frequency range of 1 to 50 kHz to determine the effects of temperature and frequency on the core loss and dynamic B-H loops of three different soft magnetic materials; an oriented-grain 50Ni-50Fe alloy, a nonoriented-grain 50Ni-50Fe alloy, and an iron-based amorphous material (Metglas 2605SC). A comparison of these materials show that the nonoriented-grain 50Ni-50Fe alloy tends to have either the lowest or next lowest core loss for all temperatures and frequencies investigated

ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

Directory of Open Access Journals (Sweden)

Setiyanto Setiyanto

2016-10-01

Full Text Available ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities and central irradiation position (CIP, especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g, but very low value for Lazy Susan position (lest then 0,11 W/g. Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung

Subharmonic excitation in an HTGR core

International Nuclear Information System (INIS)

Bezler, P.; Curreri, J.R.

1977-01-01

The occurrence of subharmonic resonance in a series of blocks with clearance between blocks and with springs on the outer most ends is the subject of this paper. This represents an HTGR core response to an earthquake input. An analytical model of the cross section of this type of core is a series of blocks arranged horizontally between outer walls. Each block represents many graphite hexagonal core elements acting in unison as a single mass. The blocks are of unequal size to model the true mass distribution through the core. Core element elasticity and damping characteristics are modeled with linear spring and viscous damping units affixed to each block. The walls and base represent the core barell or core element containment structure. For forced response calculations, these boundaries are given prescribed motions. The clearance between each block could be the same or different with the total clearance duplicating that of the entire core. Spring packs installed between the first and last block and the boundaries model the boundary elasticity. The system non-linearity is due to the severe discontinuity in the interblock elastic forces when adjacent blocks collide. A computer program using a numerical integration scheme was developed to solve for the response of the system to arbitrary inputs

Core Genome Multilocus Sequence Typing Scheme for Stable, Comparative Analyses of Campylobacter jejuni and C. coli Human Disease Isolates.

Science.gov (United States)

Cody, Alison J; Bray, James E; Jolley, Keith A; McCarthy, Noel D; Maiden, Martin C J

2017-07-01

Human campylobacteriosis, caused by Campylobacter jejuni and C. coli , remains a leading cause of bacterial gastroenteritis in many countries, but the epidemiology of campylobacteriosis outbreaks remains poorly defined, largely due to limitations in the resolution and comparability of isolate characterization methods. Whole-genome sequencing (WGS) data enable the improvement of sequence-based typing approaches, such as multilocus sequence typing (MLST), by substantially increasing the number of loci examined. A core genome MLST (cgMLST) scheme defines a comprehensive set of those loci present in most members of a bacterial group, balancing very high resolution with comparability across the diversity of the group. Here we propose a set of 1,343 loci as a human campylobacteriosis cgMLST scheme (v1.0), the allelic profiles of which can be assigned to core genome sequence types. The 1,343 loci chosen were a subset of the 1,643 loci identified in the reannotation of the genome sequence of C. jejuni isolate NCTC 11168, chosen as being present in >95% of draft genomes of 2,472 representative United Kingdom campylobacteriosis isolates, comprising 2,207 (89.3%) C. jejuni isolates and 265 (10.7%) C. coli isolates. Validation of the cgMLST scheme was undertaken with 1,478 further high-quality draft genomes, containing 150 or fewer contiguous sequences, from disease isolate collections: 99.5% of these isolates contained ≥95% of the 1,343 cgMLST loci. In addition to the rapid and effective high-resolution analysis of large numbers of diverse isolates, the cgMLST scheme enabled the efficient identification of very closely related isolates from a well-defined single-source campylobacteriosis outbreak. Copyright © 2017 Cody et al.

Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents. (Modelling of steam condensation on the particles)

International Nuclear Information System (INIS)

Bunz, H.; Dunbar, L.H.; Fermandjian, J.; Lhiaubet, G.

1987-11-01

An aerosol code comparison exercise was performed within the framework of the Commission of European Communities (Division of Safety of Nuclear Installations). This exercise, focused on the process of steam condensation onto the aerosols occurring in PWR containment buildings during severe core damage accidents, has allowed to understand the discrepancies between the results obtained. These discrepancies are due, in particular, to whether the curvature effect is modelled or not in the codes

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)

75 FR 52596 - Financial Education Core Competencies; Comment Request

Science.gov (United States)

2010-08-26

... impacts of your spending decisions. Saving Saved money grows... Start saving early. Pay yourself first... Subcommittee, identified five core concept areas: (1) Earning, (2) spending, (3) saving, (4) borrowing, and (5... financial system. Know about financial Comparison shop. assets (savings Balance risk and accounts, bonds...

Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

International Nuclear Information System (INIS)

Hong, S. G.; Kim, Y. H.; Venneri, F.

2010-01-01

In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

High-resolution OSL chronology of a sediment core from Lake Nam Co on the southern Tibetan Plateau: Comparison with radiocarbon dating

Science.gov (United States)

Long, Hao; Shen, Ji; Haberzettl, Torsten; Fuchs, Markus; Frechen, Manfred; Wang, Junbo

2013-04-01

Numerous studies on lake sediment cores from the Tibetan Plateau aimed to reconstruct spatial and temporal changes of the late Glacial and Holocene monsoon variations, which are characterized by fluctuations in precipitation and expansion of monsoonal air masses across the Plateau. Accurate and reliable dating of lacustrine deposits is of crucial importance in both the reconstruction of palaeolake and palaeoclimate evolution and the understanding of the mechanisms for climate changes, especially abrupt changes of regional-hemispheric hydrological circulation. Radiocarbon dating is the most commonly used method for establishing chronologies of lake sediments. However, 14C dating of such sediments could be problematic due to the lack of organic matter or a reservoir effect, which appears common in radiocarbon dating of lacustrine sediments from the Tibetan Plateau. In this study, ca. 10.5 m long core (NC core) was retrieved from the water depth of 93 m at Nam Co Lake, which is the second largest saline lake in China, located on the southern Tibetan Plateau. For this core, high-resolution samples (23 samples) were obtained for optically stimulated luminescence (OSL) dating. Quartz of fine-grain (4-11 μm) fraction were extracted from this samples for OSL dating, which is compared with the 14C chronology of NC core based on accelerator mass spectrometry (AMS) dating of bulk organic matter of 22 samples. The sample from the top of NC core was dated to approximate zero age, indicating the sediments bleached well before deposition. Luminescence behaviors of the study samples further confirmed the robustness of OSL ages. Comparison between OSL and 14C ages suggested that: (1) two kinds of ages are in agreement within error from top to 1.8 m, (2) from 1.8 m to the base of NC core, the 14C ages show a general trend to ca. 4 ka higher ages than the OSL ages. Although 14C dating overestimate the ages from 1.8 m to base, they still show the general trend in sedimentation rate

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.

Core Genome Multilocus Sequence Typing for Identification of Globally Distributed Clonal Groups and Differentiation of Outbreak Strains of Listeria monocytogenes

OpenAIRE

Chen, Yi; Gonzalez-Escalona, Narjol; Hammack, Thomas S.; Allard, Marc W.; Strain, Errol A.; Brown, Eric W.

2016-01-01

ABSTRACT Many listeriosis outbreaks are caused by a few globally distributed clonal groups, designated clonal complexes or epidemic clones, of Listeria monocytogenes, several of which have been defined by classic multilocus sequence typing (MLST) schemes targeting 6 to 8 housekeeping or virulence genes. We have developed and evaluated core genome MLST (cgMLST) schemes and applied them to isolates from multiple clonal groups, including those associated with 39 listeriosis outbreaks. The cgMLST...

Synthesis and Plasmonic Understanding of Core/Satellite and Core Shell Nanostructures

Science.gov (United States)

Ruan, Qifeng

Au nanospheres with molecular linkers. The plasmon resonances of the core/satellite nanostructures undergo red shifts in comparison to those of the sole Au cores, which is consistent with Mie theory analysis. As predicted by finite-difference time-domain simulations, the assembled core/satellite nanostructures exhibit large enhancements for Raman scattering. The facile growth of Au nanospheres and assembly of core/satellite nanostructures blaze a new way to the design of nanoarchitectures with desired plasmonic properties and functions. Coating semiconductors onto Au nanocrystals to form core shell configurations can increase the interactions between the two materials, benefiting from their large active interfacial area. The shell can also protect the Au nanocrystal core from aggregation, reshaping, and chemical corrosion. In this thesis, (Au nanocrystal core) (titania shell) nanostructures with tunable shell thicknesses were prepared by a facile wetchemistry method. Au nanocrystals with strong and tunable plasmon resonances in the visible and near-infrared regions can enhance and broaden the light utilization of TiO2 through the scattering/absorption enhancement, sensitization, and hot-electron injection. The integration of Au nanocrystals therefore hold the prospect of breaking the light-harvesting limit of TiO2 arising from its wide band gap. The resultant (Au core) (TiO2 shell) nanostructures were examined to be capable of efficiently generating reactive oxygen species under near-infrared resonant excitation. On the other hand, the transverse plasmon modes of Au nanorods, which are often too weak to be observed on scattering spectra, are enhanced by the TiO2 shell through energy transfer. With the increment of the shell thickness, the intensity of the transverse plasmon mode increases significantly and even becomes comparable with the longitudinal plasmon mode. Interestingly, both the transverse and longitudinal modes of the (Au core) (TiO2 shell) nanostructures

A Practical Core Loss Model for Filter Inductors of Power Electronic Converters

DEFF Research Database (Denmark)

Matsumori, Hiroaki; Shimizu, Toshihisa; Wang, Xiongfei

2018-01-01

This paper proposes a core loss model for filter inductors of power electronic converters. The model allows a computationally efficient analysis on the core loss of the inductor under the square voltage excitation and the premagnetization condition. First, the core loss of the filter inductor under...... buck chopper excitation is evaluated with the proposed model and compared with the conventional methods. The comparison shows that the proposed method results in a better core loss prediction under the premagnetized condition than that of conventional alternatives. Then, the core loss of the filter...... inductor with the pulsewidth modulated inverter excitation is evaluated, which shows that the proposed model not only accurately predicts the core loss but also identifies the hysteresis loss part. These results demonstrate that the approach can further be used for the development of magnetic materials...

Efforts for optimization of BWR core internals replacement

International Nuclear Information System (INIS)

Iizuka, N.

2000-01-01

The core internal components replacement of a BWR was successfully completed at Fukushima-Daiichi Unit 3 (1F3) of the Tokyo Electric Power Company (TEPCO) in 1998. The core shroud and the majority of the internal components made by type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. Although this core internals replacement project was completed, several factors combined to result in a longer-than-expected period for the outage. It was partly because the removal work of the internal components was delayed. Learning a lesson from whole experience in this project, some methods were adopted for the next replacement project at Fukushima-Daiichi Unit 2 (1F2) to shorten the outage and reduce the total radiation exposure. Those are new removal processes and new welding machine and so on. The core internals replacement work was ended at 1F2 in 1999, and both the period of outage and the total radiation exposure were the same degree as expected previous to starting of this project. This result shows that the methods adopted in this project are basically applicable for the core internals replacement work and the whole works about the BWR core internals replacement were optimized. The outline of the core internals replacement project and applied technologies at 1F3 and 1F2 are discussed in this paper. (author)

Efficacy of flagellin gene typing for epidemiological studies of Campylobacter jejuni in poultry estimated by comparison with macrorestriction profiling

DEFF Research Database (Denmark)

Petersen, L.; On, S.L.W.

2000-01-01

isolates were obtained. The PCR-Fla type of one isolate changed spontaneously after five subcultures, illustrating the relative plasticity of the gene locus. Comparison of MRPs within and between Fla-types support the view that some PCR-Fla types may be conserved within clonal lines. It is concluded...

Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR

Energy Technology Data Exchange (ETDEWEB)

Galushin, Sergey, E-mail: galushin@kth.se; Kudinov, Pavel, E-mail: pkudinov@kth.se

2016-12-15

Highlights: • A data base of the debris properties in lower plenum generated using MELCOR code. • The timing of safety systems has significant effect on the relocated debris properties. • Loose coupling between core relocation and vessel failure analyses was established. - Abstract: Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario

Perceptions of similarity and response to selected comparison targets in type 2 diabetes.

Science.gov (United States)

Arigo, Danielle; Smyth, Joshua M; Suls, Jerry M

2015-01-01

Social comparisons (i.e. self-evaluations relative to others) may affect motivation for diabetes self-care behaviours. Comparisons can have either positive or negative effects, but it is not clear what differentiates these responses. This study tested the effect of a patient's perceived similarity to a comparison target on motivation for self-care. Individuals with type 2 diabetes (n = 180, MA1c = 7.59%) selected to read one of four brief descriptions of a patient with diabetes. Participants rated their motivation for self-care behaviours prior and subsequent to reading and reported the extent to which they focused on similarities between the self and the selected patient while reading. Perceived similarity moderated the effect of selection on motivation for self-care (p = .01, η2 = .06). Increased motivation was observed if participants focused on similarities with patients 'doing better' (i.e. high coping effectiveness/low symptom severity) and decreased motivation if they focused on similarities with patients 'doing worse' (low coping effectiveness/high symptom severity). Providing social comparison information in diabetes management (and perhaps other chronic diseases) may improve motivation for self-care among some patients. A subset of patients, however, may benefit from guidance to focus on similarities with certain targets.

Comparison of tree coring and soil gas sampling for screening of contaminated sites

DEFF Research Database (Denmark)

Nielsen, Mette Algreen; Stalder, Marcel; Riis, Charlotte

and then identify high risk areas. The uptake of BTEX into trees varies to a greater extent with the tree species and the site conditions than chlorinated solvents, which lead to greater uncertainty. Both methods have their advantages and disadvantages. Hence, the methods supplement each other. Based on results......Site characterization is often time consuming and a financial burden for the site owners, which raises a demand for rapid and inexpensive (pre)screening methods. Phytoscreening by tree coring has shown to be a useful tool to detect subsurface contamination, especially of chlorinated solvents...... suitable as initial screening methods for site characterization. The aim of this study is to compare tree coring and soil gas sampling to evaluate to which extent tree coring may supplement or substitute soil gas sampling as a site contaminant screening tool. And where both methods are feasible, evaluate...

SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

International Nuclear Information System (INIS)

Molnar, Jozef; Vocka, Radim

2010-01-01

The SCORPIO (SCORPIO-VVER) core monitoring system, its basic features and history of implementation at Czech NPPs are described. The most important improvements in the area of neutron physics, core thermal analysis and operation support are as follows: Moving to the 42 axial nodes across the whole system (2004); Implementation of new cross section library to support mixed reactor core with differences in axial geometry of used fuel types and enhancement of Core Simulator boundary conditions model, to properly address the 'wild' geometry in axial direction; Adjusting the thermohydraulic and neutron-physical models regarding to the Gd2 fuel needs; Support up to 5 types of FAs and 2 types of SPND (Posit, IST); Extension of form functions for pin-wise reconstruction to improve pin-power prediction in control rod coupler region; System adaptation to the new upgraded digital I and C unit system; Integration of the SCORPIO-VVER system and its workstation into the plant redundant in-core system; Implementation of new On-Line form function generation to module RECON; New design of the Strategy Generator with advanced predictions; Adaptation of the system to support the new up-rated reactor thermal power; Adding new online SDM calculation function into to system; Implementation of the new 3D power reconstruction with SPND interpretation; Extending the limit checking to the 'full core' checking. The control of margins to the technical specification: Extended to full core - all FA is controlled individually in core; The limits are definable up to 59 FA (1/6 symmetry); 4 limited parameters are controlled - Kr, qlin, Tout-fa, dTfa; 2 additional parameters are monitored - dTsat, DNBR; New MMI are developed to present the limited and controlled parameters in core. Upgrade 3 is planned for the Slovak Bohunice NPP in 2011-2012. (P.A.)

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

Energy Technology Data Exchange (ETDEWEB)

None

1979-04-15

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

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

Energy Technology Data Exchange (ETDEWEB)

Lell, R. M.; Morman, J. A.; Schaefer, R.W.; McKnight, R.D.; Nuclear Engineering Division

2009-02-23

component to construct a central core zone with a composition closer to that in an LMFBR core with high burnup. The high {sup 240}Pu configuration was constructed for two reasons. First, the composition of the high {sup 240}Pu zone more closely matched the composition of LMFBR cores anticipated in design work in 1970. Second, comparison of measurements in the ZPR-6/7 uniform core with corresponding measurements in the high {sup 240}Pu zone provided an assessment of some of the effects of long-term {sup 240}Pu buildup in LMFBR cores. The uniform core version of ZPR-6/7 is evaluated in ZPR-LMFR-EXP-001. This document only addresses measurements in the high {sup 240}Pu core version of ZPR-6/7. Many types of measurements were performed as part of the ZPR-6/7 program. Measurements of criticality, sodium void worth, control rod worth and reaction rate distributions in the high {sup 240}Pu core configuration are evaluated here. For each category of measurements, the uncertainties are evaluated, and benchmark model data are provided.

Pasta phases in core-collapse supernova matter

International Nuclear Information System (INIS)

Pais, Helena; Chiacchiera, Silvia; Providência, Constança

2016-01-01

The pasta phase in core-collapse supernova matter (finite temperatures and fixed proton fractions) is studied within relativistic mean field models. Three different calculations are used for comparison, the Thomas-Fermi (TF), the Coexisting Phases (CP) and the Compressible Liquid Drop (CLD) approximations. The effects of including light clusters in nuclear matter and the densities at which the transitions between pasta configurations and to uniform matter occur are also investigated. The free energy and pressure, in the space of particle number densities and temperatures expected to cover the pasta region, are calculated. Finally, a comparison with a finite temperature Skyrme-Hartree-Fock calculation is drawn. (paper)

Tuned Chamber Core Panel Acoustic Test Results

Science.gov (United States)

Schiller, Noah H.; Allen, Albert R.

2016-01-01

This report documents acoustic testing of tuned chamber core panels, which can be used to supplement the low-frequency performance of conventional acoustic treatment. The tuned chamber core concept incorporates low-frequency noise control directly within the primary structure and is applicable to sandwich constructions with a directional core, including corrugated-, truss-, and fluted-core designs. These types of sandwich structures have long, hollow channels (or chambers) in the core. By adding small holes through one of the facesheets, the hollow chambers can be utilized as an array of low-frequency acoustic resonators. These resonators can then be used to attenuate low-frequency noise (below 400 Hz) inside a vehicle compartment without increasing the weight or size of the structure. The results of this test program demonstrate that the tuned chamber core concept is effective when used in isolation or combined with acoustic foam treatments. Specifically, an array of acoustic resonators integrated within the core of the panels was shown to improve both the low-frequency absorption and transmission loss of the structure in targeted one-third octave bands.

Characterizing Mobile/Less-Mobile Porosity and Solute Exchange in Dual-Domain Media Using Tracer Experiments and Electrical Measurements in a Hassler-Type Core Holder

Science.gov (United States)

Falzone, S.; Slater, L. D.; Day-Lewis, F. D.; Parker, B. L.; Keating, K.; Robinson, J.

2017-12-01

Mass transfer is the process by which solute is retained in less-mobile porosity domains, and later released into the mobile porosity domain. This process is often responsible for the slow arrival and gradual release of contaminants and solute tracers. Recent studies have outlined methods using dual-domain mass transfer (DDMT) models for characterizing this phenomenon. These models use the non-linear relationship of bulk (σb) and fluid (σf) conductivity, collected from electrical methods during tracer experiments, to characterize the less-mobile/mobile porosity ratio (β) and the mass-transfer rate coefficient (α). DDMT models use the hysteretic σb-σf relationship observed while solute tracers are injected and then flushed from a sample media. Due to limitations in observing the hysteretic σb-σf relationship, this method has not been used to characterize low permeability samples. We have developed an experimental method for testing porous rock cores that allows us to develop a fundamental understanding of contaminant storage and release in consolidated rock. We test the approach on cores from sedimentary rock sites where mass transfer is expected to occur between hydraulically connected fractures and the adjacent low permeability rock matrix. Our method uses a Hassler-type core holder, designed to apply confining pressure around the outside of a sample core, which hydraulically isolates the sample core, allowing water to be injected into it at increased pressures. The experimental apparatus was also designed to measure σb with spectral induced polarization (SIP) measurements, and σf from a sampling port located at the center of the core. Cores were initially saturated with a solution with high electrical conductivity ( 80000 μS/cm). DI water was then injected into the cores at elevated pressures (>60 psi) and the saturating solution was flushed from the cores, in order to generate flow rates fast enough to capture the non-linear σb-σf relationship

Emergency core cooling system

International Nuclear Information System (INIS)

Kato, Ken.

1989-01-01

In PWR type reactors, a cooling water spray portion of emergency core cooling pipelines incorporated into pipelines on high temperature side is protruded to the inside of an upper plenum. Upon rupture of primary pipelines, pressure in a pressure vessel is abruptly reduced to generate a great amount of steams in the reactor core, which are discharged at a high flow rate into the primary pipelines on high temperature side. However, since the inside of the upper plenum has a larger area and the steam flow is slow, as compared with that of the pipelines on the high temperature side, ECCS water can surely be supplied into the reactor core to promote the re-flooding of the reactor core and effectively cool the reactor. Since the nuclear reactor can effectively be cooled to enable the promotion of pressure reduction and effective supply of coolants during the period of pressure reduction upon LOCA, the capacity of the pressure accumulation vessel can be decreased. Further, the re-flooding time for the reactor is shortened to provide an effect contributing to the improvement of the safety and the reduction of the cost. (N.H.)

Evolution of the Internet and its cores

Energy Technology Data Exchange (ETDEWEB)

Zhang Guoqing; Zhang Guoqiang; Yang Qingfeng; Cheng Suqi [Institute of Computing Technology, Chinese Academy of Sciences, Beijing 100190 (China); Zhou Tao [Department of Modern Physics, University of Science and Technology of China, Hefei 230026 (China)], E-mail: gqzhang@ict.ac.cn, E-mail: zhutou@ustc.edu

2008-12-15

In this paper, we empirically study the evolution of large scale Internet topology at the autonomous system (AS) level. The network size grows in an exponential form, obeying the famous Moore's law. We theoretically predict that the size of the AS-level Internet will double every 5.32 years. We apply the k-core decomposition method on the real Internet, and find that the size of a k-core with larger k is nearly stable over time. In addition, the maximal coreness is very stable after 2003. In contrast to the predictions of most previous models, the maximal degree of the Internet is also relatively stable versus time. We use the edge-exchange operation to obtain the randomized networks with the same degree sequence. A systematical comparison is drawn, indicating that the real Internet is more loosely connected, and both the full Internet and the nucleus are more disassortative than their randomized versions.

Evolution of the Internet and its cores

International Nuclear Information System (INIS)

Zhang Guoqing; Zhang Guoqiang; Yang Qingfeng; Cheng Suqi; Zhou Tao

2008-01-01

In this paper, we empirically study the evolution of large scale Internet topology at the autonomous system (AS) level. The network size grows in an exponential form, obeying the famous Moore's law. We theoretically predict that the size of the AS-level Internet will double every 5.32 years. We apply the k-core decomposition method on the real Internet, and find that the size of a k-core with larger k is nearly stable over time. In addition, the maximal coreness is very stable after 2003. In contrast to the predictions of most previous models, the maximal degree of the Internet is also relatively stable versus time. We use the edge-exchange operation to obtain the randomized networks with the same degree sequence. A systematical comparison is drawn, indicating that the real Internet is more loosely connected, and both the full Internet and the nucleus are more disassortative than their randomized versions.

The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

International Nuclear Information System (INIS)

Mursid Djokolelono.

1976-01-01

Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

DotU and VgrG, core components of type VI secretion systems, are essential for Francisella LVS pathogenicity.

Directory of Open Access Journals (Sweden)

Jeanette E Bröms

Full Text Available The Gram-negative bacterium Francisella tularensis causes tularemia, a disease which requires bacterial escape from phagosomes of infected macrophages. Once in the cytosol, the bacterium rapidly multiplies, inhibits activation of the inflammasome and ultimately causes death of the host cell. Of importance for these processes is a 33-kb gene cluster, the Francisella pathogenicity island (FPI, which is believed to encode a type VI secretion system (T6SS. In this study, we analyzed the role of the FPI-encoded proteins VgrG and DotU, which are conserved components of type VI secretion (T6S clusters. We demonstrate that in F. tularensis LVS, VgrG was shown to form multimers, consistent with its suggested role as a trimeric membrane puncturing device in T6SSs, while the inner membrane protein DotU was shown to stabilize PdpB/IcmF, another T6SS core component. Upon infection of J774 cells, both ΔvgrG and ΔdotU mutants did not escape from phagosomes, and subsequently, did not multiply or cause cytopathogenicity. They also showed impaired activation of the inflammasome and marked attenuation in the mouse model. Moreover, all of the DotU-dependent functions investigated here required the presence of three residues that are essentially conserved among all DotU homologues. Thus, in agreement with a core function in T6S clusters, VgrG and DotU play key roles for modulation of the intracellular host response as well as for the virulence of F. tularensis.

Metal-core@metal oxide-shell nanomaterials for gas-sensing applications: a review

Energy Technology Data Exchange (ETDEWEB)

Mirzaei, A.; Janghorban, K.; Hashemi, B. [Shiraz University, Department of Materials Science and Engineering (Iran, Islamic Republic of); Neri, G., E-mail: gneri@unime.it [University of Messina, Department of Electronic Engineering, Chemistry and Industrial Engineering (Italy)

2015-09-15

With an ever-increasing number of applications in many advanced fields, gas sensors are becoming indispensable devices in our daily life. Among different types of gas sensors, conductometric metal oxide semiconductor (MOS) gas sensors are found to be the most appealing for advanced applications in the automotive, biomedical, environmental, and safety sectors because of the their high sensitivity, reduced size, and low cost. To improve their sensing characteristics, new metal oxide-based nanostructures have thus been proposed in recent years as sensing materials. In this review, we extensively review gas-sensing properties of core@ shell nanocomposites in which metals as the core and metal oxides as the shell structure, both of nanometer sizes, are assembled into a single metal@metal oxide core–shell. These nanostructures not only combine the properties of both noble metals and metal oxides, but also bring unique synergetic functions in comparison with single-component materials. Up-dated achievements in the synthesis and characterization of metal@metal oxide core–shell nanostructures as well as their use in MOS sensors are here reported with the main objective of providing an overview about their gas-sensing properties.

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)

Analysis of RA-8 critical facility core in some configurations; Analisis del nucleo de la facilidad critica RA-8 en distintas configuraciones

Energy Technology Data Exchange (ETDEWEB)

Abbate, Maximo J [Instituto de Investigaciones Cientificas y Tecnicas de las Fuerzas Armadas (CITEFA), Villa Martelli (Argentina); Sbaffoni, Maria M [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Gerencia de Tecnologia

2000-07-01

The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

Comparison of the WSA-ENLIL model with three CME cone types

Science.gov (United States)

Jang, Soojeong; Moon, Y.; Na, H.

2013-07-01

We have made a comparison of the CME-associated shock propagation based on the WSA-ENLIL model with three cone types using 29 halo CMEs from 2001 to 2002. These halo CMEs have cone model parameters as well as their associated interplanetary (IP) shocks. For this study we consider three different cone types (an asymmetric cone model, an ice-cream cone model and an elliptical cone model) to determine 3-D CME parameters (radial velocity, angular width and source location), which are the input values of the WSA-ENLIL model. The mean absolute error (MAE) of the arrival times for the asymmetric cone model is 10.6 hours, which is about 1 hour smaller than those of the other models. Their ensemble average of MAE is 9.5 hours. However, this value is still larger than that (8.7 hours) of the empirical model of Kim et al. (2007). We will compare their IP shock velocities and densities with those from ACE in-situ measurements and discuss them in terms of the prediction of geomagnetic storms.Abstract (2,250 Maximum Characters): We have made a comparison of the CME-associated shock propagation based on the WSA-ENLIL model with three cone types using 29 halo CMEs from 2001 to 2002. These halo CMEs have cone model parameters as well as their associated interplanetary (IP) shocks. For this study we consider three different cone types (an asymmetric cone model, an ice-cream cone model and an elliptical cone model) to determine 3-D CME parameters (radial velocity, angular width and source location), which are the input values of the WSA-ENLIL model. The mean absolute error (MAE) of the arrival times for the asymmetric cone model is 10.6 hours, which is about 1 hour smaller than those of the other models. Their ensemble average of MAE is 9.5 hours. However, this value is still larger than that (8.7 hours) of the empirical model of Kim et al. (2007). We will compare their IP shock velocities and densities with those from ACE in-situ measurements and discuss them in terms of the

Rational synthesis of AB-type N-substituted core-functionalized naphthalene diimides (cNDIs).

Science.gov (United States)

Berezin, Andrey A; Sciutto, Andrea; Demitri, Nicola; Bonifazi, Davide

2015-04-17

Acid-mediated transformation of tetraethyl 2,6-diethoxynaphthalene-1,4,5,8-tetracarboxylate selectively affords the core-substituted naphthalene-anhydride-ester (cNAE) in quantitative yield. This anhydride can be selectively converted into hetero-N-substituted core-functionalized naphthalene diimides (cNDIs) through sequential condensation reactions in the presence of the precursor amine with very high isolated yields over four steps. The approach can be applied to prepare a large variety of heterocyclic, aromatic, and aliphatic heterodiimides.

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)

Vibration and acoustic noise emitted by dry-type air-core reactors under PWM voltage excitation

Science.gov (United States)

Li, Jingsong; Wang, Shanming; Hong, Jianfeng; Yang, Zhanlu; Jiang, Shengqian; Xia, Shichong

2018-05-01

According to coupling way between the magnetic field and the structural order, structure mode is discussed by engaging finite element (FE) method and both natural frequency and modal shape for a dry-type air-core reactor (DAR) are obtained in this paper. On the basis of harmonic response analysis, electromagnetic force under PWM (Pulse Width Modulation) voltage excitation is mapped with the structure mesh, the vibration spectrum is gained and the consequences represents that the whole structure vibration predominates in the radial direction, with less axial vibration. Referring to the test standard of reactor noise, the rules of emitted noise of the DAR are measured and analyzed at chosen switching frequency matches the sample resonant frequency and the methods of active vibration and noise reduction are put forward. Finally, the low acoustic noise emission of a prototype DAR is verified by measurement.

Study of the effect of varying core diameter, shell thickness and strain velocity on the tensile properties of single crystals of Cu-Ag core-shell nanowire using molecular dynamics simulations

Science.gov (United States)

Sarkar, Jit; Das, D. K.

2018-01-01

Core-shell type nanostructures show exceptional properties due to their unique structure having a central solid core of one type and an outer thin shell of another type which draw immense attention among researchers. In this study, molecular dynamics simulations are carried out on single crystals of copper-silver core-shell nanowires having wire diameter ranging from 9 to 30 nm with varying core diameter, shell thickness, and strain velocity. The tensile properties like yield strength, ultimate tensile strength, and Young's modulus are studied and correlated by varying one parameter at a time and keeping the other two parameters constant. The results obtained for a fixed wire size and different strain velocities were extrapolated to calculate the tensile properties like yield strength and Young's modulus at standard strain rate of 1 mm/min. The results show ultra-high tensile properties of copper-silver core-shell nanowires, several times than that of bulk copper and silver. These copper-silver core-shell nanowires can be used as a reinforcing agent in bulk metal matrix for developing ultra-high strength nanocomposites.

Core status computing system

International Nuclear Information System (INIS)

Yoshida, Hiroyuki.

1982-01-01

Purpose: To calculate power distribution, flow rate and the like in the reactor core with high accuracy in a BWR type reactor. Constitution: Total flow rate signals, traverse incore probe (TIP) signals as the neutron detector signals, thermal power signals and pressure signals are inputted into a process computer, where the power distribution and the flow rate distribution in the reactor core are calculated. A function generator connected to the process computer calculates the absolute flow rate passing through optional fuel assemblies using, as variables, flow rate signals from the introduction part for fuel assembly flow rate signals, data signals from the introduction part for the geometrical configuration data at the flow rate measuring site of fuel assemblies, total flow rate signals for the reactor core and the signals from the process computer. Numerical values thus obtained are given to the process computer as correction signals to perform correction for the experimental data. (Moriyama, K.)

Modified ferrite core-shell nanoparticles magneto-structural characterization

Science.gov (United States)

Klekotka, Urszula; Piotrowska, Beata; Satuła, Dariusz; Kalska-Szostko, Beata

2018-06-01

In this study, ferrite nanoparticles with core-shell structures and different chemical compositions of both the core and shell were prepared with success. Proposed nanoparticles have in the first and second series magnetite core, and the shell is composed of a mixture of ferrites with Fe3+, Fe2+ and M ions (where M = Co2+, Mn2+ or Ni2+) with a general composition of M0.5Fe2.5O4. In the third series, the composition is inverted, the core is composed of a mixture of ferrites and as a shell magnetite is placed. Morphology and structural characterization of nanoparticles were done using Transmission Electron Microscopy (TEM), X-ray diffraction (XRD), and Infrared spectroscopy (IR). While room temperature magnetic properties were measured using Mössbauer spectroscopy (MS). It is seen from Mössbauer measurements that Co always increases hyperfine magnetic field on Fe atoms at RT, while Ni and Mn have opposite influences in comparison to pure Fe ferrite, regardless of the nanoparticles structure.

LMFBR core flowering response to an impulse load

International Nuclear Information System (INIS)

Brochard, D.; Petret, J.C.; Queval, J.C.; Gibert, R.J.

1993-01-01

Some incidental situations like MFCI (Meeting Fuel Coolant Incident) may induce a core flowering and lead to consider impulse loans applied to LMFBR core. These highly dynamic loads are very different considering their spatial repartition and their frequency content from the seismic loads which have been deeply studied. Recently, tests have been performed on the LMFBR core mock-up RAPSODIE in order to validate the calculation methods for centered impulse load. These tests consist in injecting water quickly in the mock-up through a specific device replacing the core central assembly. The influence of the injection pressure and the influence of the injection axial position have been investigate. During the tests, the top displacements of some assemblies have been measured. The aim of this paper is first to present the experimental device and the test results. Then a non linear numerical model is described; this model includes the impact between subassemblies and is based on an homogenization method allowing to take into account with accuracy the fluid structure interaction.The comparisons between calculation results an test results will finally be presented

Reactor core and control rod assembly in FBR type reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

1993-01-01

Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

Simple analytical expression for crosstalk estimation in homogeneous trench-assisted multi-core fibers

DEFF Research Database (Denmark)

Ye, Feihong; Tu, Jiajing; Saitoh, Kunimasa

2014-01-01

An analytical expression for the mode coupling coe cient in homogeneous trench-assisted multi-core fibers is derived, which has a sim- ple relationship with the one in normal step-index structures. The amount of inter-core crosstalk reduction (in dB) with trench-assisted structures compared...... to the one with normal step-index structures can then be written by a simple expression. Comparison with numerical simulations confirms that the obtained analytical expression has very good accuracy for crosstalk estimation. The crosstalk properties in trench-assisted multi-core fibers, such as crosstalk...... dependence on core pitch and wavelength-dependent crosstalk, can be obtained by this simple analytical expression....

Ge/Si core/shell quantum dots in alumina: tuning the optical absorption by the core and shell size

Directory of Open Access Journals (Sweden)

Nekić Nikolina

2017-03-01

Full Text Available Ge/Si core/shell quantum dots (QDs recently received extensive attention due to their specific properties induced by the confinement effects of the core and shell structure. They have a type II confinement resulting in spatially separated charge carriers, the electronic structure strongly dependent on the core and shell size. Herein, the experimental realization of Ge/Si core/shell QDs with strongly tunable optical properties is demonstrated. QDs embedded in an amorphous alumina glass matrix are produced by simple magnetron sputtering deposition. In addition, they are regularly arranged within the matrix due to their self-assembled growth regime. QDs with different Ge core and Si shell sizes are made. These core/shell structures have a significantly stronger absorption compared to pure Ge QDs and a highly tunable absorption peak dependent on the size of the core and shell. The optical properties are in agreement with recent theoretical predictions showing the dramatic influence of the shell size on optical gap, resulting in 0.7 eV blue shift for only 0.4 nm decrease at the shell thickness. Therefore, these materials are very promising for light-harvesting applications.

Pre-cometary ice composition from hot core chemistry.

Science.gov (United States)

Tornow, Carmen; Kührt, Ekkehard; Motschmann, Uwe

2005-10-01

Pre-cometary ice located around star-forming regions contains molecules that are pre-biotic compounds or pre-biotic precursors. Molecular line surveys of hot cores provide information on the composition of the ice since it sublimates near these sites. We have combined a hydrostatic hot core model with a complex network of chemical reactions to calculate the time-dependent abundances of molecules, ions, and radicals. The model considers the interaction between the ice and gas phase. It is applied to the Orion hot core where high-mass star formation occurs, and to the solar-mass binary protostar system IRAS 16293-2422. Our calculations show that at the end of the hot core phase both star-forming sites produce the same prebiotic CN-bearing molecules. However, in the Orion hot core these molecules are formed in larger abundances. A comparison of the calculated values with the abundances derived from the observed line data requires a chemically unprocessed molecular cloud as the initial state of hot core evolution. Thus, it appears that these objects are formed at a much younger cloud stage than previously thought. This implies that the ice phase of the young clouds does not contain CN-bearing molecules in large abundances before the hot core has been formed. The pre-biotic molecules synthesized in hot cores cause a chemical enrichment in the gas phase and in the pre-cometary ice. This enrichment is thought to be an important extraterrestrial aspect of the formation of life on Earth and elsewhere.

Computational Software to Fit Seismic Data Using Epidemic-Type Aftershock Sequence Models and Modeling Performance Comparisons

Science.gov (United States)

Chu, A.

2016-12-01

Modern earthquake catalogs are often analyzed using spatial-temporal point process models such as the epidemic-type aftershock sequence (ETAS) models of Ogata (1998). My work implements three of the homogeneous ETAS models described in Ogata (1998). With a model's log-likelihood function, my software finds the Maximum-Likelihood Estimates (MLEs) of the model's parameters to estimate the homogeneous background rate and the temporal and spatial parameters that govern triggering effects. EM-algorithm is employed for its advantages of stability and robustness (Veen and Schoenberg, 2008). My work also presents comparisons among the three models in robustness, convergence speed, and implementations from theory to computing practice. Up-to-date regional seismic data of seismic active areas such as Southern California and Japan are used to demonstrate the comparisons. Data analysis has been done using computer languages Java and R. Java has the advantages of being strong-typed and easiness of controlling memory resources, while R has the advantages of having numerous available functions in statistical computing. Comparisons are also made between the two programming languages in convergence and stability, computational speed, and easiness of implementation. Issues that may affect convergence such as spatial shapes are discussed.

Comparison of type 2 diabetes mellitus incidence in different phases of hepatitis B virus infection: A meta-analysis.

Science.gov (United States)

Shen, Yi; Zhang, Sheng; Wang, Xulin; Wang, Yuanyuan; Zhang, Jian; Qin, Gang; Li, Wenchao; Ding, Kun; Zhang, Lei; Liang, Feng

2017-10-01

Because whether hepatitis B virus infection increases the risk of type 2 diabetes mellitus has been a controversial topic, pair-wise and network meta-analyses of published literature were carried out to accurately evaluate the association between different phases of hepatitis B virus infection and the risk of type 2 diabetes mellitus. A comprehensive literature retrieval was conducted from the PubMed, Embase, Cochrane Library and Chinese Database to identify epidemiological studies on the association between hepatitis B virus infection and the risk of type 2 diabetes mellitus that were published from 1999 to 2015. A pair-wise meta-analysis of direct evidence was performed to estimate the pooled odds ratios and 95% confidence intervals. A network meta-analysis was conducted, including the construction of a network plot, inconsistency plot, predictive interval plot, comparison-adjusted funnel plot and rank diagram, to graphically link the direct and indirect comparisons between different hepatitis B virus infective phases. Eighteen publications (n=113 639) describing 32 studies were included in this meta-analysis. In the pair-wise meta-analysis, the pooled odds ratio for type 2 diabetes mellitus in chronic hepatitis B cirrhosis patients was 1.76 (95% confidence interval: 1.44-2.14) when compared with non-cirrhotic chronic hepatitis B patients. In the network meta-analysis, six comparisons of four hepatitis B virus infectious states indicated the following descending order for the risk of type 2 diabetes mellitus: hepatitis B cirrhosis patients, non-cirrhotic chronic hepatitis B patients, hepatitis B virus carriers and non-hepatitis B virus controls. This study suggests that hepatitis B virus infection is not an independent risk factor for type 2 diabetes mellitus, but the development of cirrhosis may increase the incidence of type 2 diabetes mellitus cirrhosis. © 2016 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

The Kinematics of Molecular Cloud Cores in the Presence of Driven and Decaying Turbulence: Comparisons with Observations

Energy Technology Data Exchange (ETDEWEB)

Offner, S R; Krumholz, M R; Klein, R I; McKee, C F

2007-12-17

In this study we investigate the formation and properties of prestellar and protostellar cores using hydrodynamic, self-gravitating Adaptive Mesh Refinement simulations, comparing the cases where turbulence is continually driven and where it is allowed to decay. We model observations of these cores in the C{sup 18}O(2 {yields} 1), NH{sub 3}(1, 1), and N{sub 2}H{sup +}(1 {yields} 0) lines, and from the simulated observations we measure the linewidths of individual cores, the linewidths of the surrounding gas, and the motions of the cores relative to one another. Some of these distributions are significantly different in the driven and decaying runs, making them potential diagnostics for determining whether the turbulence in observed star-forming clouds is driven or decaying. Comparing our simulations with observed cores in the Perseus and {rho} Ophiuchus clouds shows reasonably good agreement between the observed and simulated core-to-core velocity dispersions for both the driven and decaying cases. However, we find that the linewidths through protostellar cores in both simulations are too large compared to the observations. The disagreement is noticeably worse for the decaying simulation, in which cores show highly supersonic in fall signatures in their centers that decrease toward their edges, a pattern not seen in the observed the regions.

Type Classes for Lightweight Substructural Types

Directory of Open Access Journals (Sweden)

Edward Gan

2015-02-01

Full Text Available Linear and substructural types are powerful tools, but adding them to standard functional programming languages often means introducing extra annotations and typing machinery. We propose a lightweight substructural type system design that recasts the structural rules of weakening and contraction as type classes; we demonstrate this design in a prototype language, Clamp. Clamp supports polymorphic substructural types as well as an expressive system of mutable references. At the same time, it adds little additional overhead to a standard Damas-Hindley-Milner type system enriched with type classes. We have established type safety for the core model and implemented a type checker with type inference in Haskell.

Comparison of US/FRG accident condition models for HTGR fuel failure and radionuclide release

International Nuclear Information System (INIS)

Verfondern, K.

1991-03-01

The objective was to compare calculation models used in safety analyses in the US and FRG which describe fission product release behavior from TRISO coated fuel particles under core heatup accident conditions. The frist step performed is the qualitative comparison of both sides' fuel failure and release models in order to identify differences and similarities in modeling assumptions and inputs. Assumptions of possible particle failure mechanisms under accident conditions (SiC degradation, pressure vessel) are principally the same on both sides though they are used in different modeling approaches. The characterization of a standard (= intact) coated particle to be of non-releasing (GA) or possibly releasing (KFA/ISF) type is one of the major qualitative differences. Similar models are used regarding radionuclide release from exposed particle kernels. In a second step, a quantitative comparison of the calculation models was made by assessing a benchmark problem predicting particle failure and radionuclide release under MHTGR conduction cooldown accident conditions. Calculations with each side's reference method have come to almost the same failure fractions after 250 hours for the core region with maximum core heatup temperature despite the different modeling approaches of SORS and PANAMA-I. The comparison of the results of particle failure obtained with the Integrated Failure and Release Model for Standard Particles and its revision provides a 'verification' of these models in this sense that the codes (SORS and PANAMA-II, and -III, respectively) which were independently developed lead to very good agreement in the predictions. (orig./HP) [de

Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

Energy Technology Data Exchange (ETDEWEB)

Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

2017-05-15

This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

International Nuclear Information System (INIS)

Murao, Y.; Iguchi, T.; Sugimoto, J.

1988-09-01

The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

Intrinsically secure fast reactors with dense cores

International Nuclear Information System (INIS)

Slessarev, Igor

2007-01-01

Secure safety, resistance to weapons material proliferation and problems of long-lived wastes remain the most important 'painful points' of nuclear power. Many innovative reactor concepts have been developed aimed at a radical enhancement of safety. The promising potential of innovative nuclear reactors allows for shifting accents in current reactor safety 'strategy' to reveal this worth. Such strategy is elaborated focusing on the priority for intrinsically secure safety features as well as on sure protection being provided by the first barrier of defence. Concerning the potential of fast reactors (i.e. sodium cooled, lead-cooled, etc.), there are no doubts that they are able to possess many favourable intrinsically secure safety features and to lay the proper foundation for a new reactor generation. However, some of their neutronic characteristics have to be radically improved. Among intrinsically secure safety properties, the following core parameters are significantly important: reactivity margin values, reactivity feed-back and coolant void effects. Ways of designing intrinsically secure safety features in fast reactors (titled hereafter as Intrinsically Secure Fast Reactors - ISFR) can be found in the frame of current reactor technologies by radical enhancement of core neutron economy and by optimization of core compositions. Simultaneously, respecting resistance to proliferation, by using non-enriched fuel feed as well as a core breeding gain close to zero, are considered as the important features (long-lived waste problems will be considered in a separate paper). This implies using the following reactor design options as well as closed fuel cycles with natural U as the reactor feed: ·Ultra-plate 'dense cores' of the ordinary (monolithic) type with negative total coolant void effects. ·Modular type cores. Multiple dense modules can be embedded in the common reflector for achieving the desired NPP total power. The modules can be used also independently (as

Multiregional coupled conduction--convection model for heat transfer in an HTGR core

International Nuclear Information System (INIS)

Giles, G.E. Jr.; Childs, K.W.; Sanders, J.P.

1978-01-01

HEXEREI is a three-dimensional, coupled conduction-convection heat transfer and multichannel fluid dynamic analysis computer code with both steady-state and transient capabilities. The program was developed to provide thermal-fluid dynamic analysis of a core following the general design for high-temperature gas-cooled reactors (HTGRs); its purpose was to provide licensing evaluations for the U.S. Nuclear Regulatory Commission. In order to efficiently model the HTGR core, the nodal geometry of HEXEREI was chosen as a regular hexagonal array perpendicular to the axis of and bounded by a right circular cylinder. The cylindrical nodal geometry surrounds the hexagonal center portion of the mesh; these two different types of nodal geometries must be connected by interface nodes to complete the accurate modeling of the HTGR core. HEXEREI will automatically generate a nodal geometry that will accurately model a complex assembly of hexagonal and irregular prisms. The accuracy of the model was proven by a comparison of computed values with analytical results for steady-state and transient heat transfer problems. HEXEREI incorporates convective heat transfer to the coolant in many parallel axial flow channels. Forced and natural convection (which permits different flow directions in parallel channels) is included in the heat transfer and fluid dynamic models. HEXEREI incorporates a variety of steady-state and transient solution techniques that can be matched with a particular problem to minimize the computational time. HEXEREI was compared with a code of similar capabilities that was based on a Cartesian mesh. This code modeled only one specific core design, and the mesh spacing was closer than that generated by HEXEREI. Good agreement was obtained with the detail provided by the representations

Zooplankton trophic niches respond to different water types of the western Tasman Sea: A stable isotope analysis

Science.gov (United States)

Henschke, Natasha; Everett, Jason D.; Suthers, Iain M.; Smith, James A.; Hunt, Brian P. V.; Doblin, Martina A.; Taylor, Matthew D.

2015-10-01

The trophic relationships of 21 species from an oceanic zooplankton community were studied using stable isotopes of carbon and nitrogen. Zooplankton and suspended particulate organic matter (POM) were sampled in three different water types in the western Tasman Sea: inner shelf (IS), a cold core eddy (CCE) and a warm core eddy (WCE). δ15N values ranged from 3.9‰ for the parasitic copepod Sapphirina augusta to 10.2‰ for the euphausiid, Euphausia spinifera. δ13C varied from -22.6 to -19.4‰ as a result of the copepod Euchirella curticauda and E. spinifera. The isotopic composition of POM varied significantly among water types; as did the trophic enrichment of zooplankton over POM, with the lowest enrichment in the recently upwelled IS water type (0.5‰) compared to the warm core eddy (1.6‰) and cold core eddy (2.7‰). The WCE was an oligotrophic environment and was associated with an increased trophic level for omnivorous zooplankton (copepods and euphausiids) to a similar level as carnivorous zooplankton (chaetognaths). Therefore carnivory in zooplankton can increase in response to lower abundance and reduced diversity in their phytoplankton and protozoan prey. Trophic niche width comparisons across three zooplankton species: the salp Thalia democratica, the copepod Eucalanus elongatus and the euphausiid Thysanoessa gregaria, indicated that both niche partitioning and competition can occur within the zooplankton community. We have shown that trophic relationships among the zooplankton are dynamic and respond to different water types. The changes to the zooplankton isotopic niche, however, were still highly variable as result of oceanographic variation within water types.

Plant Core Environmental Stress Response Genes Are Systemically Coordinated during Abiotic Stresses

Directory of Open Access Journals (Sweden)

Kenneth W. Berendzen

2013-04-01

Full Text Available Studying plant stress responses is an important issue in a world threatened by global warming. Unfortunately, comparative analyses are hampered by varying experimental setups. In contrast, the AtGenExpress abiotic stress experiment displays intercomparability. Importantly, six of the nine stresses (wounding, genotoxic, oxidative, UV-B light, osmotic and salt can be examined for their capacity to generate systemic signals between the shoot and root, which might be essential to regain homeostasis in Arabidopsis thaliana. We classified the systemic responses into two groups: genes that are regulated in the non-treated tissue only are defined as type I responsive and, accordingly, genes that react in both tissues are termed type II responsive. Analysis of type I and II systemic responses suggest distinct functionalities, but also significant overlap between different stresses. Comparison with salicylic acid (SA and methyl-jasmonate (MeJA responsive genes implies that MeJA is involved in the systemic stress response. Certain genes are predominantly responding in only one of the categories, e.g., WRKY genes respond mainly non-systemically. Instead, genes of the plant core environmental stress response (PCESR, e.g., ZAT10, ZAT12, ERD9 or MES9, are part of different response types. Moreover, several PCESR genes switch between the categories in a stress-specific manner.

Engineered inorganic core/shell nanoparticles

Energy Technology Data Exchange (ETDEWEB)

Mélinon, Patrice, E-mail: patrice.melinon@univ-lyon1.fr [Institut Lumière matière Université Claude Bernard Lyon 1 et CNRS et OMNT, Domaine Scientifique de la Doua, Bâtiment Léon Brillouin, 43 Boulevard du 11 Novembre 1918, F 69622 Villeurbanne (France); Begin-Colin, Sylvie [IPCMS et OMNT, 23 rue du Loess BP 43, 67034 STRASBOURG Cedex 2 (France); Duvail, Jean Luc [IMN UMR 6502 et OMNT Campus Sciences : 2 rue de la Houssinire, BP32229, 44322 Nantes Cedex3 (France); Gauffre, Fabienne [SPM et OMNT : Institut des sciences chimiques de Rennes - UMR 6226, 263 Avenue du General Leclerc, CS 74205, 35042 RENNES Cedex (France); Boime, Nathalie Herlin [IRAMIS-NIMBE, Laboratoire Francis Perrin (CEA CNRS URA 2453) et OMNT, Bat 522, CEA Saclay, 91191 Gif sur Yvette Cedex (France); Ledoux, Gilles [Institut Lumière Matière Université Claude Bernard Lyon 1 et CNRS et OMNT, Domaine Scientifique de la Doua, Bâtiment Alfred Kastler 43 Boulevard du 11 Novembre 1918 F 69622 Villeurbanne (France); Plain, Jérôme [Universit de technologie de Troyes LNIO-ICD, CNRS et OMNT 12 rue Marie Curie - CS 42060 - 10004 Troyes cedex (France); Reiss, Peter [CEA Grenoble, INAC-SPrAM, UMR 5819 CEA-CNRS-UJF et OMNT, Grenoble cedex 9 (France); Silly, Fabien [CEA, IRAMIS, SPEC, TITANS, CNRS 2464 et OMNT, F-91191 Gif sur Yvette (France); Warot-Fonrose, Bénédicte [CEMES-CNRS, Université de Toulouse et OMNT, 29 rue Jeanne Marvig F 31055 Toulouse (France)

2014-10-20

It has been for a long time recognized that nanoparticles are of great scientific interest as they are effectively a bridge between bulk materials and atomic structures. At first, size effects occurring in single elements have been studied. More recently, progress in chemical and physical synthesis routes permitted the preparation of more complex structures. Such structures take advantages of new adjustable parameters including stoichiometry, chemical ordering, shape and segregation opening new fields with tailored materials for biology, mechanics, optics magnetism, chemistry catalysis, solar cells and microelectronics. Among them, core/shell structures are a particular class of nanoparticles made with an inorganic core and one or several inorganic shell layer(s). In earlier work, the shell was merely used as a protective coating for the core. More recently, it has been shown that it is possible to tune the physical properties in a larger range than that of each material taken separately. The goal of the present review is to discuss the basic properties of the different types of core/shell nanoparticles including a large variety of heterostructures. We restrict ourselves on all inorganic (on inorganic/inorganic) core/shell structures. In the light of recent developments, the applications of inorganic core/shell particles are found in many fields including biology, chemistry, physics and engineering. In addition to a representative overview of the properties, general concepts based on solid state physics are considered for material selection and for identifying criteria linking the core/shell structure and its resulting properties. Chemical and physical routes for the synthesis and specific methods for the study of core/shell nanoparticle are briefly discussed.

Cross-section imaging and p-type doping assessment of ZnO/ZnO:Sb core-shell nanowires by scanning capacitance microscopy and scanning spreading resistance microscopy

Energy Technology Data Exchange (ETDEWEB)

Wang, Lin, E-mail: lin.wang@insa-lyon.fr; Brémond, Georges [Institut des Nanotechnologies de Lyon (INL), Université de Lyon, CNRS UMR 5270, INSA Lyon, Bat. Blaise Pascal, 7 Avenue, Jean Capelle, 69621 Villeurbanne (France); Sallet, Vincent; Sartel, Corinne [Groupe d' étude de la Matière Condensée (GEMaC), CNRS - Université de Versailles St Quentin en Yvelines, Université Paris-Saclay, 45 Avenue des Etats-Unis, 78035 Versailles (France)

2016-08-29

ZnO/ZnO:Sb core-shell structured nanowires (NWs) were grown by the metal organic chemical vapor deposition method where the shell was doped with antimony (Sb) in an attempt to achieve ZnO p-type conduction. To directly investigate the Sb doping effect in ZnO, scanning capacitance microscopy (SCM) and scanning spreading resistance microscopy (SSRM) were performed on the NWs' cross-sections mapping their two dimensional (2D) local electrical properties. Although no direct p-type inversion in ZnO was revealed, a lower net electron concentration was pointed out for the Sb-doped ZnO shell layer with respect to the non-intentionally doped ZnO core, indicating an evident compensating effect as a result of the Sb incorporation, which can be ascribed to the formation of Sb-related acceptors. The results demonstrate SCM/SSRM investigation being a direct and effective approach for characterizing radial semiconductor one-dimensional (1D) structures and, particularly, for the doping study on the ZnO nanomaterial towards its p-type realization.

New long-cycle small modular PWR cores using particle type burnable poisons for low boron operation

Energy Technology Data Exchange (ETDEWEB)

Yoo, Hoseong; Hwang, Dae Hee [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Hong, Ser Gi, E-mail: sergihong@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Shin, Ho Choel [Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI), Daejon 305-343 (Korea, Republic of)

2017-04-01

Highlights: • New advanced burnable poison rods (BPR) are suggested for low boron operation in PWR. • The new SMR cores have long cycle length of ∼4.5 EFPYs with low boron concentration. • The SMR core satisfies all the design targets and constraints. - Abstract: In this paper, new small long-cycle PWR (Pressurized Water Reactor) cores for low boron concentration operation are designed by employing advanced burnable poison rods (BPRs) in which the BISO (Bi-Isotropic) particles of burnable poison are distributed in a SiC matrix. The BPRs are designed by adjusting the kernel diameter, the kernel material and the packing fraction to effectively reduce the excess reactivity in order to reduce the boron concentration in the coolant and achieve a flat change in excess reactivity over a long operational cycle. In addition, axial zoning of the BPRs was suggested to improve the core performances, and it was shown that the suggested axial zoning of BPRs considerably extends the cycle length compared to a core with no BPR axial zoning. The results of the core physics analyses showed that the cores using BPRs with a B{sub 4}C kernel have long cycle lengths of ∼4.5 EFPYs (Effective Full Power Years), small maximum CBCs (Critical Boron Concentration) lower than 370 ppm, low power peaking factors, and large shutdown margins of control element assemblies.

Spatial scaling of core and dominant forest cover in the Upper Mississippi and Illinois River floodplains, USA

Science.gov (United States)

De Jager, Nathan R.; Rohweder, Jason J.

2011-01-01

Different organisms respond to spatial structure in different terms and across different spatial scales. As a consequence, efforts to reverse habitat loss and fragmentation through strategic habitat restoration ought to account for the different habitat density and scale requirements of various taxonomic groups. Here, we estimated the local density of floodplain forest surrounding each of ~20 million 10-m forested pixels of the Upper Mississippi and Illinois River floodplains by using moving windows of multiple sizes (1–100 ha). We further identified forest pixels that met two local density thresholds: 'core' forest pixels were nested in a 100% (unfragmented) forested window and 'dominant' forest pixels were those nested in a >60% forested window. Finally, we fit two scaling functions to declines in the proportion of forest cover meeting these criteria with increasing window length for 107 management-relevant focal areas: a power function (i.e. self-similar, fractal-like scaling) and an exponential decay function (fractal dimension depends on scale). The exponential decay function consistently explained more variation in changes to the proportion of forest meeting both the 'core' and 'dominant' criteria with increasing window length than did the power function, suggesting that elevation, soil type, hydrology, and human land use constrain these forest types to a limited range of scales. To examine these scales, we transformed the decay constants to measures of the distance at which the probability of forest meeting the 'core' and 'dominant' criteria was cut in half (S 1/2, m). S 1/2 for core forest was typically between ~55 and ~95 m depending on location along the river, indicating that core forest cover is restricted to extremely fine scales. In contrast, half of all dominant forest cover was lost at scales that were typically between ~525 and 750 m, but S 1/2 was as long as 1,800 m. S 1/2 is a simple measure that (1) condenses information derived from multi

APROS couplings from core to containment

International Nuclear Information System (INIS)

Puska, E.K.; Ylijoki, J.

2005-01-01

APROS simulation environment is able to describe the 1-D and 3-D neutronics of the reactor core. It is also able to describe the thermal hydraulics of the core and circuits either with 5- equation or 6-equation thermal hydraulics. It can also describe the plant automation and electrical systems, as well as the behaviour of the containment. The peculiar feature of APROS in comparison to other coupled systems is that all parts in the coupled system are described with the same code instead of coupling two or three separate codes together with information exchange between the separate codes. The most recent possibility is the coupled calculation of the process and the containment. The more traditional coupling, the coupling of the process containing both the process description and the automation description with more or less detailed description of the 3-D core either for safety analysis or real-time simulation purposes has been discussed in previous work. The paper presents and discusses the capabilities of the code in coupling the plant process and automation description with the plant containment description with two example transient cases. An improved boron concentration solution with second order upwind discretization has been recently included in APROS. An example on the increased accuracy acquired in the 3-D core model has been included. (authors)

Practical electron dosimetry: a comparison of different types of ionization chambers

International Nuclear Information System (INIS)

Dohm, O.S.; Christ, G.

2002-01-01

Since Markus chambers are no longer recommended in the 1997 DIN 6800-2 version there are uncertainties as to the use of alternative chamber types for electron dosimetry. Therefore, we performed a comparison between different types of ionization chambers. In particular, the widespread Farmer and Roos chambers were compared with the Markus chamber for polarity effect, chamber-to-chamber variation, and deviations of the measured absorbed dose relative to the value obtained with the Roos chamber (which is regarded as an ideal Bragg-Gray-chamber). The perturbation correction factor at 60 Co radiation was determined experimentally as 1,029 ± 0,5% (Roos chamber) and 1,018 ± 0,5% (Markus chamber) for the investigated plane-parallel chambers. In addition, we could show that the Roos chambers do not have a larger chamber-to-chamber variation than the Farmer chambers. Likewise, our results suggest that Farmer chambers could be used for electron energies above 6 MeV. (orig.) [de

Comparison of 210Pb level in Kuala Muda tsunami affected marine sediment core measured using two different techniques

International Nuclear Information System (INIS)

Yii Mei Wo; Zal Uyun Wan Mahmood; Jalal Sharib; Zaharudin Ahmad

2010-01-01

The most commonly used techniques of alpha and gamma spectrometry were compared to measure 210 Pb activity in marine sediment core. Alpha analytical technique measured the activity of 210 Pb from its in-grow grand-daughter 210 Po after a chemical separation by assuming that there is radioactive equilibrium between the two radionuclides. Meanwhile, gamma analysis technique allows direct measurement, non-destructive and no preliminary chemical separation. Through the comparison, it was found that both alpha and gamma analysis techniques were slightly different. Overall, the results from gamma analytical technique were higher than those from the alpha analytical techniques. Some logical argument had been discussed to explain this situation. In routine analysis, the analytical technique used should be chosen carefully based on advantages and disadvantages of the each technique and analysis requirements. Therefore, it is recommended to determine the exact needs and purpose of analysis and to know the sample history before deciding the appropriate analytical technique. (author)

Numerical Analysis of Heat Storage and Heat Conductivity in the Concrete Hollow Core Deck Element

DEFF Research Database (Denmark)

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

2011-01-01

extent these simplified models estimate the heat storage potential of precast hollow-core concrete decks correctly. This study investigates various approaches on how to model the heat transfer within the air void in the deck. Furthermore, it is analysed how different heat transfer models influence...... the overall heat transfer and heat storage in the hollow-core decks. The presented results allow comparison between detailed results from 2D-COMSOL simulations and simple 1D calculations from the whole building simulation tool such as BSim program and moreover, it is possible to validate the calculation...... method in BSim for the concrete deck element with air voids. Finally, this paper presents a comparison of the calculated heat conductivity of the hollow-core concrete deck and the measured heat conductivity for the same deck by using hot box apparatus....

A new type of sandwich panel with periodic cellular metal cores and its mechanical performances

International Nuclear Information System (INIS)

Lim, Chae-Hong; Jeon, Insu; Kang, Ki-Ju

2009-01-01

Many studies have been performed on the mechanical properties and optimization of truss PCMs (periodic cellular metals), but those on the fabrication process, which is one of key factors determining the survivability of PCMs in the market, have been relatively limited. This study introduces a new idea on the fabrication of quasi Kagome truss cored sandwich panels, which is based on the expanded-metal process. And the mechanical behavior of the sandwich panels is to be evaluated. The mechanical strengths and failure mechanisms under compression and bending load are estimated based on elementary mechanics of materials, and the optimal design is derived. Its validity is proved by comparison with the results of experiments. The results showed that the new idea is promising with respect to all three requirements, i.e., the morphology, fabrication cost, and raw materials. The simple mechanical analysis was sufficiently effective and accurate for estimating the performance and failure mode of the sandwich panels. In the experiments, sandwich panel specimens of three different designs were compared in their bending behaviors to demonstrate sensitivity of geometric parameters. Namely, although all the designs had little difference in their load capacity-per-weight, the failure mechanisms and the behaviors after a peak load were totally different.

Solitons in a hard-core bosonic system: Gross–Pitaevskii type and ...

Indian Academy of Sciences (India)

-core .... smoothly varying function with a length scale greater than the lattice spacing a. Using ..... φ(x,τ) must be of the form ωτ +φ(x −vτ) (recall eq. (3)), i.e. .... solutions for the density in the HCB system, providing a comprehensive picture of its.

Applicability of the diffusion and simplified P3 theories for BWR pin-by-pin core analysis

International Nuclear Information System (INIS)

Tada, Kenichi; Yamamoto, Akio; Kitamura, Yasunori; Yamane, Yoshihiro; Watanabe, Masato; Noda, Hiroshi

2007-01-01

The pin-by-pin fine mesh core calculation method is considered as a candidate of next-generation core calculation method for BWR. In this study, the diffusion and the simplified P 3 (SP 3 ) theories are applied to the pin-by-pin core analysis of BWR. Performances of the diffusion and the SP 3 theories for cell-homogeneous pin-by-pin fine mesh BWR core analysis are evaluated through comparison with cell-heterogeneous detailed transport calculation by the method of characteristics (MOC). In this study, two-dimensional, 2x2 multi-assemblies geometry is used to compare the prediction accuracies of the diffusion and the SP 3 theories. The 2x2 multi- assemblies geometry consists of two types of 9x9 UO 2 assembly that have two different enrichment splittings. To mitigate the cell-homogenization error, the SPH method is applied for the pin-by-pin fine mesh calculation. The SPH method is a technique that reproduces a result of heterogeneous calculation by that of homogeneous calculation. The calculation results indicated that diffusion theory shows larger discrepancy than that of SP 3 theory on pin-wise fission rates. Furthermore, the accuracy of the diffusion theory would not be sufficient for the pin-by-pin fine mesh calculation. In contrast to the diffusion theory, the SP 3 theory shows much better accuracy on pin wise fission rates. Therefore, if the SP 3 theory is applied, the accuracy of the pin-by-pin fine mesh BWR core analysis will be higher and will be sufficient for production calculation. (author)

Behind the Final Grade in Hybrid v. Traditional Courses: Comparing Student Performance by Assessment Type, Core Competency, and Course Objective

Science.gov (United States)

Bain, Lisa Z.

2012-01-01

There are many different delivery methods used by institutions of higher education. These include traditional, hybrid, and online course offerings. The comparisons of these typically use final grade as the measure of student performance. This research study looks behind the final grade and compares student performance by assessment type, core…

Theoretical models for Type I and Type II supernova

International Nuclear Information System (INIS)

Woosley, S.E.; Weaver, T.A.

1985-01-01

Recent theoretical progress in understanding the origin and nature of Type I and Type II supernovae is discussed. New Type II presupernova models characterized by a variety of iron core masses at the time of collapse are presented and the sensitivity to the reaction rate 12 C(α,γ) 16 O explained. Stars heavier than about 20 M/sub solar/ must explode by a ''delayed'' mechanism not directly related to the hydrodynamical core bounce and a subset is likely to leave black hole remnants. The isotopic nucleosynthesis expected from these massive stellar explosions is in striking agreement with the sun. Type I supernovae result when an accreting white dwarf undergoes a thermonuclear explosion. The critical role of the velocity of the deflagration front in determining the light curve, spectrum, and, especially, isotopic nucleosynthesis in these models is explored. 76 refs., 8 figs

Functional requirements for core surveillance systems

International Nuclear Information System (INIS)

Andersson, T.

2000-01-01

Operating experience at Ringhals-2 has demonstrated the feasibility of a mixed core surveillance system comprised of fixed in-core detectors combined with the original movable detector system. A small number of fixed in-core detectors provide continuous measurement of the thermal margins while the movable detectors are used mainly at start-up to verify the expected power distribution. Reactor noise diagnostics and neural networks can further improve the monitoring system. The reliability of the movable detector system can be improved by mechanical simplification. Wear and maintenance costs are lowered if the required flux-mapping frequency is reduced. Improved computer codes make the measurement uncertainties less dependent on the number of instrumented positions. A mixed system requires new types of technical specifications. (author)

Adaptive Core Simulation Employing Discrete Inverse Theory - Part I: Theory

International Nuclear Information System (INIS)

Abdel-Khalik, Hany S.; Turinsky, Paul J.

2005-01-01

Use of adaptive simulation is intended to improve the fidelity and robustness of important core attribute predictions such as core power distribution, thermal margins, and core reactivity. Adaptive simulation utilizes a selected set of past and current reactor measurements of reactor observables, i.e., in-core instrumentation readings, to adapt the simulation in a meaningful way. A meaningful adaption will result in high-fidelity and robust adapted core simulator models. To perform adaption, we propose an inverse theory approach in which the multitudes of input data to core simulators, i.e., reactor physics and thermal-hydraulic data, are to be adjusted to improve agreement with measured observables while keeping core simulator models unadapted. At first glance, devising such adaption for typical core simulators with millions of input and observables data would spawn not only several prohibitive challenges but also numerous disparaging concerns. The challenges include the computational burdens of the sensitivity-type calculations required to construct Jacobian operators for the core simulator models. Also, the computational burdens of the uncertainty-type calculations required to estimate the uncertainty information of core simulator input data present a demanding challenge. The concerns however are mainly related to the reliability of the adjusted input data. The methodologies of adaptive simulation are well established in the literature of data adjustment. We adopt the same general framework for data adjustment; however, we refrain from solving the fundamental adjustment equations in a conventional manner. We demonstrate the use of our so-called Efficient Subspace Methods (ESMs) to overcome the computational and storage burdens associated with the core adaption problem. We illustrate the successful use of ESM-based adaptive techniques for a typical boiling water reactor core simulator adaption problem

Social comparisons in adults with type 2 diabetes: Patients' reasons for target selection.

Science.gov (United States)

Arigo, Danielle; Cornell, Max; Smyth, Joshua M

2018-07-01

To examine reasons for selecting a social comparison target (i.e. a specific other for relative self-evaluation), and their influence on affect and motivation for self-care, in type 2 diabetes (T2DM). Adults with T2DM (n = 180, M A1c  = 7.6%) chose to read about one of four targets. Participants rated five reasons for their choice (strongly disagree - strongly agree), and rated affect and self-care motivation before and after reading. To boost confidence in my ability to manage diabetes was rated highest overall (ps motivation (p motivation only among those who chose better-off targets (p = 0.01). Patients' reasons for a particular comparison are associated with short-term changes in affect and self-care motivation, and warrant greater empirical and clinical attention.

Determination of the hydrogen source term during the reflooding of an overheated core: Calculation results of the integral reflood test QUENCH-03 with PWR-type bundle

International Nuclear Information System (INIS)

Chikhi, Nourdine; Nguyen, Nam Giang; Fleurot, Joelle

2012-01-01

Highlights: ► Calculation of QUENCH-03 experiment with ASTEC/CATHARE. ► Validation of reflooding model in severe accidents conditions. ► Demonstration of a minimum flow rate for a successful reflood by using a system code. ► Effect of injection flow rate on hydrogen production. ► Effect of initial core temperature on hydrogen production. - Abstract: During a severe accident, one of the main accident management procedure consists of injecting water in the reactor core by means of various safety injection devices. Nevertheless, the success of a core reflood is not guaranteed because of possible negative effects: temperature escalation, enhanced hydrogen production, enhanced release of fission products, core degradation due to thermal shock, shattering, debris and melt formation. The QUENCH-03 experiment was carried out to investigate the behavior on reflooding at high temperature of LWR fuel rods with little oxidation. Posttest calculations with the ASTEC-CATHARE V2 code were made for code assessment and validation of the new reflooding model. This thermal–hydraulic model is used to detect the quench front position and to calculate the heat transfer between fuel and fluid in the transition boiling region. Comparisons between the calculational and experimental results are presented. Emphasis has been placed on clad temperature, hydrogen production and melt relocation. The effects of core state damage (initial temperature at reflooding onset) and the reflood mass flow rate on the hydrogen source term were investigated using the QUENCH-03 test as a base case. Calculations were made by varying both parameters in the input data deck. The results demonstrate (and confirm) the existence of a minimum flow rate for a successful reflood.