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

International Nuclear Information System (INIS)

McLaughlin, T.B.

1982-09-01

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

Comparison of facility characteristics between SCTF Core-I and Core-II

International Nuclear Information System (INIS)

Adachi, Hiromichi; Iwamura, Takamichi; Sobajima, Makoto; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio.

1990-08-01

The Slab Core Test Facility (SCTF) was constructed to investigate two-dimensional thermal-hydraulics in the core and fluid behavior of carryover water out of the core including its feed-back effect to the core behavior mainly during the reflood phase of a large break loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). Since three simulated cores are used in the SCTF Test Program and the design of these three cores are slightly different one by one, repeatability test is required to justify a direct comparison of data obtained with different cores. In the present report, data of Test S2-13 (Run 618) obtained with SCTF Core-II were compared with those of Test S1-05 (Run 511) obtained with the Core-I, which were performed under the forced-flooding condition. Thermal-hydraulic behaviors in these two tests showed quite similar characteristics of both system behavior and two-dimensional core behaviors. Therefore, the test data obtained from the two cores can be compared directly with each other. After the turnaround of clad temperatures, however, some differences were found in upper plenum water accumulation and resultant two-dimensional core cooling behaviors such as quench front propagation from bottom to top of the core. (author)

Gravitational Core-Mantle Coupling and the Acceleration of the Earth

Science.gov (United States)

Rubincam, David Parry; Smith, David E. (Technical Monitor)

2001-01-01

Gravitational core-mantle coupling may be the cause of the observed variable acceleration of the Earth's rotation on the 1000 year timescale. The idea is that density inhomogeneities which randomly come and go in the liquid outer core gravitationally attract density inhomogeneities in the mantle and crust, torquing the mantle and changing its rotation state. The corresponding torque by the mantle on the core may also explain the westward drift of the magnetic field of 0.2 deg per year. Gravitational core-mantle coupling would stochastically affect the rate of change of the Earth's obliquity by just a few per cent. Its contribution to polar wander would only be about 0.5% the presently observed rate. Tidal friction is slowing down the rotation of the Earth, overwhelming a smaller positive acceleration from postglacial rebound. Coupling between the liquid outer core of the Earth and the mantle has long been a suspected reason for changes in the length-of-day. The present investigation focuses on the gravitational coupling between the density anomalies in the convecting liquid outer core and those in the mantle and crust as a possible cause for the observed nonsecular acceleration on the millenial timescale. The basic idea is as follows. There are density inhomogeneities caused by blobs circulating in the outer core like the blobs in a lava lamp; thus the outer core's gravitational field is not featureless. Moreover, these blobs will form and dissipate somewhat randomly. Thus there will be a time variability to the fields. These density inhomogeneities will gravitationally attract the density anomalies in the mantle.

Fast Flux Test Facility core system

International Nuclear Information System (INIS)

Ethridge, J.L.; Baker, R.B.; Leggett, R.D.; Pitner, A.L.; Waltar, A.E.

1990-11-01

A review of Liquid Metal Reactor (LMR) core system accomplishments provides an excellent road map through the maze of issues that faced reactor designers 10 years ago. At that time relatively large uncertainties were associated with fuel pin and fuel assembly performance, irradiation of structural materials, and performance of absorber assemblies. The extensive core systems irradiation program at the US Department of Energy's Fast Flux Test Facility (FFTF) has addressed each of these principal issues. As a result of the progress made, the attention of long-range LMR planners and designers can shift away from improving core systems and focus on reducing capital costs to ensure the LMR can compete economically in the 21st century with other nuclear reactor concepts. 3 refs., 6 figs., 1 tab

MIMI: Multimodality, Multiresource, Information Integration Environment for Biomedical Core Facilities

OpenAIRE

Szymanski, Jacek; Wilson, David L.; Zhang, Guo-Qiang

2007-01-01

The rapid expansion of biomedical research has brought substantial scientific and administrative data management challenges to modern core facilities. Scientifically, a core facility must be able to manage experimental workflow and the corresponding set of large and complex scientific data. It must also disseminate experimental data to relevant researchers in a secure and expedient manner that facilitates collaboration and provides support for data interpretation and analysis. Administrativel...

VHTR core modeling: coupling between neutronic and thermal-hydraulics

International Nuclear Information System (INIS)

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

2005-01-01

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

Optimal power and distribution control for weakly-coupled-core reactor

International Nuclear Information System (INIS)

Oohori, Takahumi; Kaji, Ikuo

1977-01-01

A numerical procedure has been devised for obtaining the optimal power and distribution control for a weakly-coupled-core reactor. Several difficulties were encountered in solving this optimization problem: (1) nonlinearity of the reactor kinetics equations; (2) neutron-leakage interaction between the cores; (3) localized power changes occurring in addition to the total power changes; (4) constraints imposed on the states - e.g. reactivity, reactor period. To obviate these difficulties, use is made of the generalized Newton method to convert the problem into an iterative sequence of linear programming problems, after approximating the differential equations and the integral performance criterion by a set of discrete algebraic equations. In this procedure, a heuristic but effective method is used for deriving an initial approximation, which is then made to converge toward the optimal solution. Delayed-neutron one-group point reactor models embodying transient temperature feed-back to the reactivity are used in obtaining the kinetics equations for the weakly-coupled-core reactor. The criterion adopted for determining the optimality is a norm relevant to the deviations of neutron density from the desired trajectories or else to the time derivatives of the neutron density; uniform control intervals are prescribed. Examples are given of two coupled-core reactors with typical parameters to illustrate the results obtained with this procedure. A comparison is also made between the coupled-core reactor and the one-point reactor. (auth.)

Modulational Instability in Linearly Coupled Asymmetric Dual-Core Fibers

Directory of Open Access Journals (Sweden)

Arjunan Govindarajan

2017-06-01

Full Text Available We investigate modulational instability (MI in asymmetric dual-core nonlinear directional couplers incorporating the effects of the differences in effective mode areas and group velocity dispersions, as well as phase- and group-velocity mismatches. Using coupled-mode equations for this system, we identify MI conditions from the linearization with respect to small perturbations. First, we compare the MI spectra of the asymmetric system and its symmetric counterpart in the case of the anomalous group-velocity dispersion (GVD. In particular, it is demonstrated that the increase of the inter-core linear-coupling coefficient leads to a reduction of the MI gain spectrum in the asymmetric coupler. The analysis is extended for the asymmetric system in the normal-GVD regime, where the coupling induces and controls the MI, as well as for the system with opposite GVD signs in the two cores. Following the analytical consideration of the MI, numerical simulations are carried out to explore nonlinear development of the MI, revealing the generation of periodic chains of localized peaks with growing amplitudes, which may transform into arrays of solitons.

The calculation of the MEU-HEU coupled core in the KUCA

International Nuclear Information System (INIS)

Hayashi, M.; Shiroya, S.; Kanda, K.; Shibata, T.

1984-01-01

The KUCA has a plan for critical experiments of the MEU-HEU coupled core in 1984. The neutronics calculation has been performed for the MEU-HEU coupled core in the KUCA. The GGC-4 and THERMOS were used to generate the four-group constants and the 2D-FEM-KUR, based on the finite-element method, was used for the diffusion calculation. The calculations with four-group constants agreed with experiments within 1.8% for the both single-cores with the MEU and the HEU. (author)

Estimation of subcriticality with the computed values analysis using MCNP of experiment on coupled cores

International Nuclear Information System (INIS)

Sakurai, Kiyoshi; Yamamoto, Toshihiro; Arakawa, Takuya; Naito, Yoshitaka

1998-01-01

Experiments on coupled cores performed at TCA were analysed using continuous energy Monte Carlo calculation code MCNP 4A. Errors of neutron multiplication factors are evaluated using Indirect Bias Estimation Method proposed by authors. Calculation for simulation of pulsed neutron method was performed for 17 X 17 + 5G + 17 x 17 core system and its of exponential experiment method was also performed for 16 x 9 + 3G + 16 x 9 and 16 x 9 + 5G + 16 x 9 core systems. Errors of neutron multiplication factors are estimated to be (-1.5) - (-0.6)% evaluated by Indirect Bias Estimation Method. Its errors evaluated by conventional pulsed neutron method and exponential experiment method are estimated to be 7%, but it is below 1% for estimation of subcriticality with the computed values by applying Indirect Bias Estimation Method. Feasibility of subcriticality management is higher by application of the method to full scale fuel strage facility. (author)

Coupled full core neutron transport/CFD simulations of pressurized water reactors

International Nuclear Information System (INIS)

Kochunas, B.; Stimpson, S.; Collins, B.; Downar, T.; Brewster, R.; Baglietto, E.; Yan, J.

2012-01-01

Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)

ORNL instrumentation performance for Slab Core Test Facility (SCTF)-Core I Reflood Test Facility

International Nuclear Information System (INIS)

Hardy, J.E.; Hess, R.A.; Hylton, J.O.

1983-11-01

Instrumentation was developed for making measurements in experimental refill-reflood test facilities. These unique instrumentation systems were designed to survive the severe environmental conditions that exist during a simulated pressurized water reactor loss-of-coolant accident (LOCA). Measurement of in-vessel fluid phenomena such as two-phase flow velocity and void fraction and film thickness and film velocity are required for better understanding of reactor behavior during LOCAs. The Advanced Instrumentation for Reflood Studies (AIRS) Program fabricated and delivered instrumentation systems and data reduction software algorithms that allowed the above measurements to be made. Data produced by AIRS sensors during three experimental runs in the Japanese Slab Core Test Facility are presented. Although many of the sensors failed before any useful data could be obtained, the remaining probes gave encouraging and useful results. These results are the first of their kind produced during simulated refill-reflood stage of a LOCA near actual thermohydrodynamic conditions

Appendix BB: long coring facility (LCF)

International Nuclear Information System (INIS)

Driscoll, A.H.; Silva, A.J.

1981-01-01

During the 1979 performance period the Engineering Design of the Long Coring Facility has addressed a variety of tasks relating to the establishment of a series of operating parameters for a conceptual 50 meter long coring system. The results of these efforts have indicated that an operational system capable of the recovery of 50 meter long cores, from oceanic depths in sediments of 400 gm cm 2 is wholly possible given existing technology. Specific tasks included in the 1979 Engineering Design are as follows: (1) Hydrodynamic Stability; (2) Corer Structural Stability; (3) Corer Penetration Mechanics; (4) Anticipated Corer Pullout forces; (5) LCF Cable Dynamics; and (6) Core Head Instrumentation. Within the realm of Subseabed Disposal Programs Master Plan the LCF, as a part of the Instrumentation Development activity, is currently on schedule. Delays in receiving funding during 1979 have reduced, some what, the latitude enjoyed by the LCF project and have limited our progress to a point where any future delay can result in the LCF's falling behind the program schedule. However, at this time the LCF is considered to be on schedule, but lacking in flexibility to respond to any major contingency that may arise in the future

Fast Flux Test Facility core restraint system performance

International Nuclear Information System (INIS)

Hecht, S.L.; Trenchard, R.G.

1990-02-01

Characterizing Fast Flux Test Facility (FFTF) core restraint system performance has been ongoing since the first operating cycle. Characterization consists of prerun analysis for each core load, in-reactor and postirradiation measurements of subassembly withdrawal loads and deformations, and using measurement data to fine tune predictive models. Monitoring FFTF operations and performing trend analysis has made it possible to gain insight into core restraint system performance and head off refueling difficulties while maximizing component lifetimes. Additionally, valuable information for improved designs and operating methods has been obtained. Focus is on past operating experience, emphasizing performance improvements and avoidance of potential problems. 4 refs., 12 figs., 2 tabs

Gravity driven emergency core cooling experiments with the PACTEL facility

International Nuclear Information System (INIS)

Munther, R.; Kalli, H.; Kouhia, J.

1996-01-01

PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs

Gravity driven emergency core cooling experiments with the PACTEL facility

Energy Technology Data Exchange (ETDEWEB)

Munther, R; Kalli, H [University of Technology, Lappeenranta (Finland); Kouhia, J [Technical Research Centre of Finland, Lappeenranta (Finland)

1996-12-01

PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs.

On the Dynamics of Edge-core Coupling

Energy Technology Data Exchange (ETDEWEB)

Hahm,T.S.; Diamond, P.H.; Lin, Z.; Rewoldt, G.; Gurcan, O.; Ethier, S.

2005-08-26

One of the nagging, unresolved questions in fusion theory is concerned with the extent of the edge. Gyrokinetic particle simulations of toroidal ion temperature gradient (ITG) turbulence spreading using the Gyrokinetic Toroidal Code (GTC) [Z. Lin et al., Science 281, 1835 (1998)] and its related dynamical model have been extended to a system with radially varying ion temperature gradient, in order to study the inward spreading of edge turbulence toward the core plasma. Due to such spreading, the turbulence intensity in the core region is significantly enhanced over the value obtained from simulations of the core region only, and the precise boundary of the edge region is blurred. Even when the core gradient is within the Dimits shift regime (i.e., dominated by self-generated zonal flows which reduce the transport to a negligible value), a significant level of turbulence can penetrate to the core due to spreading from the edge. The scaling of the turbulent front propagation speed is closer to the prediction from a nonlinear diffusion model than from one based on linear toroidal coupling.

Development of whole core thermal-hydraulic analysis program ACT. 3. Coupling core module with primary heat transport system module

International Nuclear Information System (INIS)

Ohtaka, Masahiko; Ohshima, Hiroyuki

1998-10-01

A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including inter-wrapper flow under various reactor operation conditions. In this work, the core module as a main part of the ACT developed last year, which simulates thermal-hydraulics in the subassemblies and the inter-subassembly gaps, was coupled with an one dimensional plant system thermal-hydraulic analysis code LEDHER to simulate transients in the primary heat transport system and to give appropriate boundary conditions to the core model. The effective algorithm to couple these two calculation modules was developed, which required minimum modification of them. In order to couple these two calculation modules on the computing system, parallel computing technique using PVM (Parallel Virtual Machine) programming environment was applied. The code system was applied to analyze an out-of-pile sodium experiment simulating core with 7 subassemblies under transient condition for code verification. It was confirmed that the analytical results show a similar tendency of experimental results. (author)

Code Coupling for Multi-Dimensional Core Transient Analysis

International Nuclear Information System (INIS)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

2015-01-01

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

Code Coupling for Multi-Dimensional Core Transient Analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

2015-05-15

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

Demountable toroidal fusion core facility for physics optimization and fusion engineering

International Nuclear Information System (INIS)

Bogart, S.L.; Wagner, C.E.; Krall, N.A.; Dalessandro, J.A.; Weggel, C.F.; Lund, K.O.; Sedehi, S.

1986-01-01

Following a successful compact ignition tokamak (CIT) experiment, a fusion facility will be required for physics optimization (POF) and fusion engineering research (FERF). The POF will address issues such as high-beta operation, current drive, impurity control, and will test geometric and configurational variations such as the spherical torus or the reversed-field pinch (RFP). The FERF will be designed to accumulate rapidly a large neutron dose in prototypical fusion subsystems exposed to radiation. Both facilities will require low-cost replacement cores and rapid replacement times. The Demountable Toroidal Fusion Core (DTFC) facility is designed to fulfill these requirements. It would be a cost-effective stepping stone between the CIT and a demonstration fusion reactor

76 FR 1213 - Core Principles and Other Requirements for Swap Execution Facilities

Science.gov (United States)

2011-01-07

... Part II Commodity Futures Trading Commission 17 CFR Part 37 Core Principles and Other Requirements... RIN Number 3038-AD18 Core Principles and Other Requirements for Swap Execution Facilities AGENCY... Compliance With the Core Principles III. Effective Date and Transition Period IV. Related Matters A...

Single-mode annular chirally-coupled core fibers for fiber lasers

Science.gov (United States)

Zhang, Haitao; Hao, He; He, Linlu; Gong, Mali

2018-03-01

Chirally-coupled core (CCC) fiber can transmit single fundamental mode and effectively suppresses higher-order mode (HOM) propagation, thus improve the beam quality. However, the manufacture of CCC fiber is complicated due to its small side core. To decrease the manufacture difficulty in China, a novel fiber structure is presented, defined as annular chirally-coupled core (ACCC) fiber, replacing the small side core by a larger side annulus. In this paper, we designed the fiber parameters of this new structure, and demonstrated that the new structure has a similar property of single mode with traditional CCC fiber. Helical coordinate system was introduced into the finite element method (FEM) to analyze the mode field in the fiber, and the beam propagation method (BPM) was employed to analyze the influence of the fiber parameters on the mode loss. Based on the result above, the fiber structure was optimized for efficient single-mode transmission, in which the core diameter is 35 μm with beam quality M2 value of 1.04 and an optical to optical conversion efficiency of 84%. In this fiber, fundamental mode propagates in an acceptable loss, while the HOMs decay rapidly.

An approach for coupled-code multiphysics core simulations from a common input

International Nuclear Information System (INIS)

Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger; Clarno, Kevin; Simunovic, Srdjan; Slattery, Stuart; Turner, John; Palmtag, Scott

2015-01-01

Highlights: • We describe an approach for coupled-code multiphysics reactor core simulations. • The approach can enable tight coupling of distinct physics codes with a common input. • Multi-code multiphysics coupling and parallel data transfer issues are explained. • The common input approach and how the information is processed is described. • Capabilities are demonstrated on an eigenvalue and power distribution calculation. - Abstract: This paper describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak

Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

International Nuclear Information System (INIS)

Israr, M.; Shami, Qamar-ud-din; Pervez, S.

1997-11-01

In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

Stability region for a prompt power variation of a coupled-core system with positive prompt feedback

International Nuclear Information System (INIS)

Watanabe, S.; Nishina, K.

1984-01-01

A stability analysis using a one-group model is presented for a coupled-core system. Positive prompt feedback of a γp /SUB j/ form is assumed, where p /SUB j/ is the fractional power variation of core j. Prompt power variations over a range of a few milliseconds after a disturbance are analyzed. The analysis combines Lapunov's method, prompt jump approximation, and the eigenfunction expansion of coupling region response flux. The last is treated as a pseudo-delayed neutron precursor. An asymptotic stability region is found for p /SUB j/. For an asymmetric flux variation over a system of two coupled cores, either p /SUB I/ or p /SUB II/ can slightly exceed, by virtue of the coupling effect, the critical value (β/γ-1) of a single-core case. Such a stability region is increased by additional inclusion of the coupling region fundamental mode in the treatment. The coupling region contributes to stability through its delayed response and coupling. An optimum core separation distance for stability is found

A study on criticality of coupled fast-thermal core HERBE at RB reactor

Energy Technology Data Exchange (ETDEWEB)

Pesic, M; Zavaljevski, M; Milosevic, M; Stefanovic, D; Nikolic, D; Avdic, S [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia); Popovic, D; Marinkovic, P [Faculty of Electrical Engineering, Beograd (Yugoslavia)

1991-07-01

The coupled fast-thermal core HERBE at the RB zero power heavy water reactor in Vinca was designed with the aim of improving the experimental possibilities in fast neutron fields. The requirements for minimum modifications in the RB construction and the use available fuel, restricted design flexibility of the coupled system. The following core is considered optimal in the light of the foregoing constraints: the central fast core of natural uranium is surrounded by a neutron filter zone (cadmium and natural uranium) and a converter zone (enriched uranium fuel, without moderator). The coupling region is heavy water. The thermal core in the form of the RB heavy water 80% enriched uranium lattice with 12 cm pitch. The criticality of the system is obtained by adjusting the moderator level. The critical heavy water levels were measured for normal reactor operation and some simulated accidental conditions. These data were analyzed by a computer code for the design of thermal and coupled fast-thermal reactor recently developed in IBK Nuclear Engineering Laboratory. Good agreement between the computations and experimental data was achieved. (author)

A study on criticality of coupled fast-thermal core HERBE at RB reactor

International Nuclear Information System (INIS)

Pesic, M.; Zavaljevski, M.; Milosevic, M.; Stefanovic, D.; Nikolic, D.; Avdic, S.; Popovic, D.; Marinkovic, P.

1991-01-01

The coupled fast-thermal core HERBE at the RB zero power heavy water reactor in Vinca was designed with the aim of improving the experimental possibilities in fast neutron fields. The requirements for minimum modifications in the RB construction and the use available fuel, restricted design flexibility of the coupled system. The following core is considered optimal in the light of the foregoing constraints: the central fast core of natural uranium is surrounded by a neutron filter zone (cadmium and natural uranium) and a converter zone (enriched uranium fuel, without moderator). The coupling region is heavy water. The thermal core in the form of the RB heavy water 80% enriched uranium lattice with 12 cm pitch. The criticality of the system is obtained by adjusting the moderator level. The critical heavy water levels were measured for normal reactor operation and some simulated accidental conditions. These data were analyzed by a computer code for the design of thermal and coupled fast-thermal reactor recently developed in IBK Nuclear Engineering Laboratory. Good agreement between the computations and experimental data was achieved. (author)

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.)

TRIGA out of core gamma irradiation facility

International Nuclear Information System (INIS)

Rant, J.; Pregl, G.

1988-01-01

A possibility to irradiate extended objects in a gamma field inside the shielding water tank and above the core of operating TRIGA Mark II Reactor has been investigated. The irradiation cask is shielded with Cd cover to filter out thermal neutrons. The dose rate of the gamma field strongly depends on the distance of the irradiation position above the core. At 25 cm above the core, the gamma dose rate is 2.2 Gy/s and epithermal neutron flux is ∼ 8.10 6 ncm -2 s -1 ∼ 3 as measured by TLD (CaF 2 : Mn) dosimeters and Au foils respectively. Tentative applications of the gamma irradiation facility are in the studies of radiation induced accelerated aging and within the Nuclear Power Plant Equipment Qualification Program (EQP). A complete characterization of the neutron spectrum and optimization of the 7 radiation field within the cask has still to be performed. (author)

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

International Nuclear Information System (INIS)

Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

2008-01-01

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

Coupled Tort-TD/CTF Capability for high-fidelity LWR core calculations - 321

International Nuclear Information System (INIS)

Christienne, M.; Avramova, M.; Perin, Y.; Seubert, A.

2010-01-01

This paper describes the developed coupling scheme between TORT-TD and CTF. TORT-TD is a time-dependent 3D discrete ordinates neutron transport code. TORT-TD is utilized for high-fidelity reactor core neutronics calculations while CTF is providing the thermal-hydraulics feedback information. CTF is an improved version of the advanced thermal-hydraulic sub-channel code COBRA-TF, which is widely used for best-estimate evaluations of LWR safety margins. CTF is a transient code based on a separated flow representation of the two-phase flow. The coupled code TORT-TD/CTF allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. Steady-state and transient test cases, based on the OECD/NRC PWR MOX/UO 2 Core Transient Benchmark, have been calculated. The steady state cases are based on a quarter core model while the transient test case models a control rod ejection transient in a small PWR mini-core fuel assembly arrangement. The obtained results with TORT-TD/CTF are verified by a code-to-code comparison with the previously developed NEM/CTF and TORT-TD/ATHLET coupled code systems. The performed comparative analysis indicates the applicability and high-fidelity potential of the TORT-TD/CTF coupling. (authors)

A benchmark for coupled thermohydraulics system/three-dimensional neutron kinetics core models

International Nuclear Information System (INIS)

Kliem, S.

1999-01-01

During the last years 3D neutron kinetics core models have been coupled to advanced thermohydraulics system codes. These coupled codes can be used for the analysis of the whole reactor system. Although the stand-alone versions of the 3D neutron kinetics core models and of the thermohydraulics system codes generally have a good verification and validation basis, there is a need for additional validation work. This especially concerns the interaction between the reactor core and the other components of a nuclear power plant (NPP). In the framework of the international 'Atomic Energy Research' (AER) association on VVER Reactor Physics and Reactor Safety, a benchmark for these code systems was defined. (orig.)

A complete fuel development facility utilizing a dual core TRIGA reactor system

Energy Technology Data Exchange (ETDEWEB)

Middleton, A; Law, G C [General Atomic Co., San Diego, CA (United States)

1974-07-01

A TRIGA Dual Core Reactor System has been chosen by the Romanian Government as the heart of a new fuel development facility which will be operated by the Romanian Institute for Nuclear Technologies. The Facility, which will be operational in 1976, is an integral part of the Romanian National Program for Power Reactor Development, with particular emphasis being placed on fuel development. The unique combination of a new 14 MW steady state TRIGA reactor, and the well-proven TRIGA Annular Core Pulsing Reactor (ACPR) in one below-ground reactor pool resulted in a substantial construction cost savings and gives the facility remarkable experimental flexibility. The inherent safety of the TRIGA fuel elements in both reactor cores means that a secondary containment building is not necessary, resulting in further construction cost savings. The 14 MW steady state reactor gives acceptably high neutron fluxes for long- term testing of various prototype fuel-cladding-coolant combinations; and the TRIGA ACPR high pulse capability allows transient testing of fuel specimens, which is so important for accurate prediction of the performance of power reactor fuel elements under postulated failure conditions. The 14 MW steady state reactor has one large and three small in-core irradiation loop positions, two large irradiation loop positions adjacent to the core face, and twenty small holes in the beryllium reflector for small capsule irradiation. The power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position approaching 3.0 x 10{sup 14} n/cm{sup 2}-sec. The ACPR has one large dry central experimental cavity which can be loaded at pool level through a shielded offset loading tube; a small diameter in-core flux trap; and an in-core pneumatically-operated capsule irradiation position. A peak pulse of 15,000 MW will yield a peak fast neutron flux in the central experimental cavity of about 1.5 x 10{sup 17} n/cm{sup 2}-sec. The pulse width at

Coupled MCNP - SAS-SFR calculations for sodium fast reactor core at steady-state - 15460

International Nuclear Information System (INIS)

Ponomarev, A.; Travleev, A.; Pfrang, W.; Sanchez, V.

2015-01-01

The prediction of core parameters at steady state is the first step when studying core accident transient behaviour. At this step thermal hydraulics (TH) and core geometry parameters are calculated corresponding to initial operating conditions. In this study we present the coupling of the SAS-SFR code to the Monte-Carlo neutron transport code MCNP at steady state together with application to the European Sodium Fast Reactor (ESFR). The SAS-SFR code employs a multi-channel core representation where each channel represents subassemblies with similar power, thermal-hydraulics and pin mechanics conditions. For every axial node of every channel the individual geometry and material compositions parameters are calculated in accord with power and cooling conditions. This requires supplying the SAS-SFR-code with nodal power values which should be calculated by neutron physics code with given realistic core parameters. In the conventional approach the neutron physics model employs some core averaged TH and geometry data (fuel temperature, coolant density, core axial and radial expansion). In this study we organize a new approach coupling the MCNP neutron physics models and the SAS-SFR models, so that calculations of power can be improved by using distributed core parameters (TH and geometry) taken from SAS-SFR. The MCNP code is capable to describe cores with distributed TH parameters and even to model non-uniform axial expansion of fuel subassemblies. In this way, core TH and geometrical data calculated by SAS-SFR are taken into account accurately in the neutronics model. The coupling implementation is done by data exchange between two codes with help of processing routines managed by driver routine. Currently it is model-specific and realized for the ESFR 'Reference Oxide' core. The Beginning-Of-Life core state is considered with 10 channel representation for fuel subassemblies. For this model several sets of coupled calculations are performed, in which different

Analysis of criticality safety of coupled fast-thermal core 'HERBE'

International Nuclear Information System (INIS)

Pesic, M.

1991-01-01

Power excursion during possible fast core flooding is analyzed as serious accident. Model gives short filling time of fast zone with moderator after break of fast core tank. Reactivity increase is determined by computer codes and verified in specific experiments. Measurements of safety rods drop time and reactivity worth are performed. Coupled core kinetics parameters are determined according to model of Avery. Power excursion study, depending on power level threshold and safety instrumentation response time is performed. It was shown that safety system can shut-down reactor safely even in case of highly set power thresholds and partially failure of safety chain. (author)

Electromagnetic core-mantle coupling associated with changes in the geomagnetic dipole field

International Nuclear Information System (INIS)

Watanabe, Hidehumi; Yukutake, Takesi.

1975-01-01

On a shelluar earth model electromagnetic coupling between the mantle and the core is investigated when the geomagnetic dipole field changes its intensity. Besides electromagnetic interaction between the dipole change and the relative slip of the mantle to the core, coupling of the dipole change with shear motions within the core is considered. If, in the core, the dipole change is limited within a surface layer shallower than a few hundred kilometers, the electromagnetic interaction gives the same order of magnitudes and phases of mantle oscillation as suggested from observation for three different periods, 8000, 400 and 65 years, provided that the electrical conductivity of the bottom part of the mantle is 10 -9 to 10 -8 emu. It is shown that mean motion of the surface shells of the core thus calculated is compatible with the observed variations in the drift velocity of the geomagnetic secular change. Except for surface shells, those in the deep interior is confirmed to oscillate almost with the same angular velocity, like a rigid rotation, for all the periods. (auth.)

The coupling one quasi-particle to a Bohr core

International Nuclear Information System (INIS)

Lewenkopf, C.H.

1988-01-01

Odd nuclei are studied coupling one quasi-particle to a Bohr's core, solved by Kumar Baranger's method. Calculations are performed for energies and transition rates for the following isotopes: 133 Xe, 183 W, 99 Tc and 101 Rh. Limitations of the model are discussed. (author) [pt

Supermodes in Coupled Multi-Core Waveguide Structures

Science.gov (United States)

2016-04-01

this section, we begin the study of higher-order supermodes from the simplest two-core structure by demon - strating how angle-dependent coupling...Communication Conf., Los Angeles , CA, USA, 2011, Paper PDPB10. 4401212 IEEE JOURNAL OF SELECTED TOPICS IN QUANTUM ELECTRONICS, VOL. 22, NO. 2, MARCH/APRIL...microstructured fiber,” presented at the IEEE Optical Fiber Com- munication Conf., Los Angeles , CA, USA, 2012, Paper PDP5C.2. [12] S. G. Leon-Saval, A

Mechanical core coupling and reactors stability

International Nuclear Information System (INIS)

Suarez Antola, R.

2006-01-01

Structural parts of nuclear reactors are complex mechanical systems, able to vibrate with a set of proper frequencies when suitably excited. Cyclical variations in the strain state of the materials, including density perturbations, are produced. This periodic changes may affect reactor reactivity. But a variation in reactivity affects reactor thermal power, thus modifying the temperature field of the abovementiones materials. If the variation in temperature fields is fast enough, thermal-mechanical coupling may produce fast variations in strain states, and this, at its turn, modifies the reactivity, and so on. This coupling between mechanical vibrations of the structure and the materials of the core, with power oscillations of the reactor, not only may not be excluded a priori, but it seems that it has been present in some stage of the incidents or accidents that happened during the development of nuclear reactor technology. The purpose of the present communication is: (a) To review and generalize some mathematical models that were proposed in order to describe thermal-mechanical coupling in nuclear reactors. (b) To discuss some conditions in which significant instabilities could arise, including large amplitude power oscillations coupled with mechanical vibrations whose amplitudes are too small to be excluded by conventional criteria of mechanical design. Enough Certain aspects of thr physical safety of nuclear power reactors, that are objected by people that opposes to the renaissance of nucleoelectric generation, are discussed in the framework of the mathematical model proposed in this paper [es

Exploiting nonlinear dynamics in a coupled-core fluxgate magnetometer

International Nuclear Information System (INIS)

Bulsara, Adi R; In, Visarath; Kho, Andy; Longhini, Patrick; Neff, Joe; Anderson, Gregory; Obra, Christopher; Palacios, Antonio; Baglio, Salvatore; Ando, Bruno

2008-01-01

Unforced bistable dynamical systems having dynamics of the general form τ F x-dot (t)=-∇ x U(x) cannot oscillate (i.e. switch between their stable attractors). However, a number of such systems subject to carefully crafted coupling schemes have been shown to exhibit oscillatory behavior under carefully chosen operating conditions. This behavior, in turn, affords a new mechanism for the detection and quantification of target signals having magnitude far smaller than the energy barrier height in the potential energy function U(x) for a single (uncoupled) element. The coupling-induced oscillations are a feature that appears to be universal in systems described by bi- or multi-stable potential energy functions U(x), and are being exploited in a new class of dynamical sensors being developed by us. In this work we describe one of these devices, a coupled-core fluxgate magnetometer (CCFM), whose operation is underpinned by this dynamic behavior. We provide an overview of the underlying dynamics and, also, quantify the performance of our test device; in particular, we provide a quantitative performance comparison to a conventional (single-core) fluxgate magnetometer via a 'resolution' parameter that embodies the device sensitivity (the slope of its input–output transfer characteristic) as well as the noise floor

Cylindrical core reflood test facility (CCTF) and slab core reflood test facility (SCTF) for Japan Atomic Energy Research Institute (JAERI)

International Nuclear Information System (INIS)

1981-01-01

IHI has designed and constructed the CCTF at JAERI to be used in the safety analysis research on the loss of coolant accident in a PWR plant. This test facility is planned so that reflood phenomenon in the PWR plant (a phenomenon is that the bared and overheated core is reflooded by the emergency core cooling system when the coolant loss accident occurred) is simulated under various test conditions. The CCTF is the largest-scale test plant in the world, composed of approximately 2000 simulated fuel rods (electric heaters), 1 simulated pressure vessel, 4 primary cooling loops, 2 simulated steam generators, emergency core cooling system, and so on. The test conditions are controlled, and the test steps are sequentially progressed by the computing system, and test data are collected by the data acquisition system. Furthermore, IHI is now designing and constructing the SCTF in accordance with the JAERI research plan. The SCTF is similar to the CCTF in scale. Main feature of the SCTF is the form of the simulated core and the simulated pressure vessel, which is of slab construction to be representative of the radial section of the PWR reactor. Reliable and various data for safety analysis are expected by the CCTF and the SCTF. (author)

CANDU reactor core simulations using fully coupled DRAGON and DONJON calculations

International Nuclear Information System (INIS)

Varin, E.; Marleau, G.

2006-01-01

The operating CANDU-6 reactors are refueled on-power to compensate for the reactivity loss due to fuel burnup. In order to predict the core behavior, fuel bundle burnups and local parameter information need to be tracked. The history-based approach has been developed to follow local parameter as well as history effect in CANDU reactors. The finite reactor diffusion code DONJON and the lattice code DRAGON have been coupled to perform reactor follow-up calculations using a history-based approach. A coupled methodology that manages the transfer of information between standard DONJON and DRAGON data structures has been developed. Push-through refueling can be taken into account directly in cell calculations. Using actual on-site information, an isotopic core content database has been generated with coupled DONJON and DRAGON calculations. Moreover calculations have been performed for different local parameters. Results are compared with those obtained using standard cross section generation approaches

Feasibility study for a postaccident heat removal facility

International Nuclear Information System (INIS)

Barts, E.W.; Apperson, C.E. Jr.; Dunwoody, W.E.; Bennett, J.G.

1978-01-01

An initial feasibility investigation for PAHRTEF, a Postaccident Heat Removal Test Facility, is presented. The facility would provide an experimental capability for PAHR experiments beyond that available in any currently existing or proposed U.S. safety test facility. The facility design presented in this report is based upon the technology developed for the ROVER nuclear rocket propulsion program. The core is a graphite-moderated, helium-cooled, epithermal core with radial reflector control. The PAHR experiments are located just below the reactor containment vessel, very near the bottom of the core. The experiments (up to 55% enriched) are driven and controlled by neutrons leaking axially from the core such that the PAHRTEF core and the experiment form a coupled reactor system. The experiment can be designed so that it is extremely unlikely that the test fuel by itself could form a critical system. The investigation indicates that adequate fission heating of large PAHR experiments could be provided at low driver core power levels. Both the reactor and the experiment handling and examination equipment can use available technology and, whenever possible, existing equipment and buildings

Validation and applicability of the 3D core kinetics and thermal hydraulics coupled code SPARKLE

International Nuclear Information System (INIS)

Miyata, Manabu; Maruyama, Manabu; Ogawa, Junto; Otake, Yukihiko; Miyake, Shuhei; Tabuse, Shigehiko; Tanaka, Hirohisa

2009-01-01

The SPARKLE code is a coupled code system based on three individual codes whose physical models have already been verified and validated. Mitsubishi Heavy Industries (MHI) confirmed the coupling calculation, including data transfer and the total reactor coolant system (RCS) behavior of the SPARKLE code. The confirmation uses the OECD/NEA MSLB benchmark problem, which is based on Three Mile Island Unit 1 (TMI-1) nuclear power plant data. This benchmark problem has been used to verify coupled codes developed and used by many organizations. Objectives of the benchmark program are as follows. Phase 1 is to compare the results of the system transient code using point kinetics. Phase 2 is to compare the results of the coupled three-dimensional (3D) core kinetics code and 3D core thermal-hydraulics (T/H) code, and Phase 3 is to compare the results of the combined coupled system transient code, 3D core kinetics code, and 3D core T/H code as a total validation of the coupled calculation. The calculation results of the SPARKLE code indicate good agreement with other benchmark participants' results. Therefore, the SPARKLE code is validated through these benchmark problems. In anticipation of applying the SPARKLE code to licensing analyses, MHI and Japanese PWR utilities have established a safety analysis method regarding the calculation conditions such as power distributions, reactivity coefficients, and event-specific features. (author)

Survey of current trends in DNA synthesis core facilities.

Science.gov (United States)

Hager, K M; Fox, J W; Gunthorpe, M; Lilley, K S; Yeung, A

1999-12-01

The Nucleic Acids Research Group of the Association of Biomolecular Resource Facilities (ABRF) last surveyed DNA synthesis core facilities in April 1995. Because of the introduction of new technologies and dramatic changes in the market, we sought to update survey information and to determine how academic facilities responded to the challenge presented by commercial counterparts. The online survey was opened in January 1999 by notifying members and subscribers to the ABRF electronic discussion group. The survey consisted of five parts: general facility information, oligonucleotide production profile, oligonucleotide charges, synthesis protocols, and trends in DNA synthesis (including individual comments). All submitted data were anonymously coded. Respondents from DNA synthesis facilities were primarily from the academic category and were established between 1984 and 1991. Typically, a facility provides additional services such as DNA sequencing and has upgraded to electronic ordering. There is stability in staffing profiles for these facilities in that the total number of employees is relatively unchanged, the tenure for staff averages 5.9 years, and experience is extensive. On average, academic facilities annually produce approximately 1/16 the number of oligonucleotides produced by the average commercial facilities, but all facilities report an increase in demand. Charges for standard oligonucleotides from academic facilities are relatively higher than from commercial companies; however, the opposite is true for modified phosphoramidites. Subsidized facilities charge less than nonsubsidized facilities. Synthesis protocols and reagents are standard across the categories. Most facilities offer typical modifications such as biotinylation. Despite the competition by large commercial facilities that have reduced costs dramatically, academic facilities remain a stable entity. Academic facilities enhance the quality of service by focusing on nonstandard

Organisation of facilities management in relation to core business

DEFF Research Database (Denmark)

Jensen, Per Anker

2011-01-01

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

A coupling model for the two-stage core calculation method with subchannel analysis for boiling water reactors

International Nuclear Information System (INIS)

Mitsuyasu, Takeshi; Aoyama, Motoo; Yamamoto, Akio

2017-01-01

Highlights: • A coupling model of the two-stage core calculation with subchannel analysis. • BWR fuel assembly parameters are assumed and verified. • The model was evaluated for heterogeneous problems. - Abstract: The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop a core analysis model coupled with subchannel analysis within the two-stage calculation scheme using an assembly-based thermal-hydraulics calculation in the core analysis. The model changes the 2D lattice physics scheme, and couples with 3D subchannel analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. In order to couple with these two analyses, some BWR fuel assembly parameters are assumed and verified. The developed model is evaluated for the heterogeneous problem with and without a control rod. The present model is especially effective for the control rod inserted condition. The present model can incorporate the subchannel effect into the current two-stage core calculation method.

Core Polarization and Tensor Coupling Effects on Magnetic Moments of Hypernuclei

International Nuclear Information System (INIS)

Jiang-Ming, Yao; Jie, Meng; Hong-Feng, Lü; Greg, Hillhouse

2008-01-01

Effects of core polarization and tensor coupling on the magnetic moments in Λ 13 C, Λ 17 O, and Λ 41 Ca Λ-hypernuclei are studied by employing the Dirac equation with scalar, vector and tensor potentials. It is found that the effect of core polarization on the magnetic moments is suppressed by Λ tensor coupling. The Λ tensor potential reduces the spin-orbit splitting of p Λ states considerably. However, almost the same magnetic moments are obtained using the hyperon wavefunction obtained via the Dirac equation either with or without the A tensor potential in the electromagnetic current vertex. The deviations of magnetic moments for p Λ states from the Schmidt values are found to increase with nuclear mass number. (nuclear physics)

78 FR 47154 - Core Principles and Other Requirements for Swap Execution Facilities; Correction

Science.gov (United States)

2013-08-05

... COMMODITY FUTURES TRADING COMMISSION 17 CFR Part 37 RIN 3038-AD18 Core Principles and Other Requirements for Swap Execution Facilities; Correction AGENCY: Commodity Futures Trading Commission. ACTION... Principles [Corrected] 2. On page 33600, in the second column, under the heading Core Principle 3 of Section...

Coupled fast-thermal core 'HERBE', as the benchmark experiment at the RB reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-10-01

Validation of the well-known Monte Carlo code MCNP TM against measured criticality data for the coupled fast-thermal HERBE. System at the RB research reactor is shown in this paper. Experimental data are obtained for regular HERBE core and for the cases of controlled flooding of the neutron converter zone by heavy water. Earlier calculations of these criticality parameters, done by combination of transport and diffusion codes using 2D geometry model are also compared to new calculations carried out by the MCNP code in 3D geometry, applying new detailed 3D model of the HEU fuel slug, developed recently. Satisfactory agreements in comparison of the HERBE criticality calculation results with experimental data, in spite complex heterogeneous composition of the HERBE core, are obtained and confirmed that HERBE core could be used as a criticality benchmark for coupled fast-thermal core. (author)

Challenges and Opportunities for Biological Mass Spectrometry Core Facilities in the Developing World.

Science.gov (United States)

Bell, Liam; Calder, Bridget; Hiller, Reinhard; Klein, Ashwil; Soares, Nelson C; Stoychev, Stoyan H; Vorster, Barend C; Tabb, David L

2018-04-01

The developing world is seeing rapid growth in the availability of biological mass spectrometry (MS), particularly through core facilities. As proteomics and metabolomics becomes locally feasible for investigators in these nations, application areas associated with high burden in these nations, such as infectious disease, will see greatly increased research output. This article evaluates the rapid growth of MS in South Africa (currently approaching 20 laboratories) as a model for establishing MS core facilities in other nations of the developing world. Facilities should emphasize new services rather than new instruments. The reduction of the delays associated with reagent and other supply acquisition would benefit both facilities and the users who make use of their services. Instrument maintenance and repair, often mediated by an in-country business for an international vendor, is also likely to operate on a slower schedule than in the wealthiest nations. A key challenge to facilities in the developing world is educating potential facility users in how best to design experiments for proteomics and metabolomics, what reagents are most likely to introduce problematic artifacts, and how to interpret results from the facility. Here, we summarize the experience of 6 different institutions to raise the level of biological MS available to researchers in South Africa.

Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

Energy Technology Data Exchange (ETDEWEB)

Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta [Lappeenranta University of Technology, P.O. 20 53851 Lappeenranta (Finland); Sairanen, Risto [Radiation and Nuclear Safety Authority, P.O. 14 00881 Helsinki (Finland)

2008-07-01

The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

International Nuclear Information System (INIS)

Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta; Sairanen, Risto

2008-01-01

The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

Fiber up-tapering and down-tapering for low-loss coupling between anti-resonant hollow-core fiber and solid-core fiber

Science.gov (United States)

Zhang, Naiqian; Wang, Zefeng; Xi, Xiaoming

2017-10-01

In this paper, we demonstrate a novel method for the low-loss coupling between solid-core multi-mode fibers (MMFs) and anti-resonant hollow-core fibers (AR-HCFs). The core/cladding diameter of the MMF is 50/125μm and the mode field diameter of the AR-HCFs are 33.3μm and 71.2μm of the ice-cream type AR-HCFs and the non-node type ARHCFs, respectively. In order to match the mode field diameters of these two specific AR-HCFs, the mode field diameter of the MMFs is increased or decreased by up-tapering or down-tapering the MMFs. Then, according to the principle of coupled fiber mode matching, the optimal diameter of tapered fiber for low-loss coupling is calculated. Based on beam propagation method, the calculated coupling losses without tapering process are 0.31dB and 0.89dB, respectively for a MMF-HCF-MMF structure of the ice-cream type AR-HCFs and the non-node type AR-HCFs. These values can be reduced to 0.096dB and 0.047dB when the outer diameters of the MMF are down-tapered to 116μm and up-tapered to 269μm, respectively. What's more, these results can also be verified by existing experiments.

Printed freeform lens arrays on multi-core fibers for highly efficient coupling in astrophotonic systems.

Science.gov (United States)

Dietrich, Philipp-Immanuel; Harris, Robert J; Blaicher, Matthias; Corrigan, Mark K; Morris, Tim M; Freude, Wolfgang; Quirrenbach, Andreas; Koos, Christian

2017-07-24

Coupling of light into multi-core fibers (MCF) for spatially resolved spectroscopy is of great importance to astronomical instrumentation. To achieve high coupling efficiencies along with fill-fractions close to unity, micro-optical elements are required to concentrate the incoming light to the individual cores of the MCF. In this paper we demonstrate facet-attached lens arrays (LA) fabricated by two-photon polymerization. The LA provide close to 100% fill-fraction along with efficiencies of up to 73% (down to 1.4 dB loss) for coupling of light from free space into an MCF core. We show the viability of the concept for astrophotonic applications by integrating an MCF-LA assembly in an adaptive-optics test bed and by assessing its performance as a tip/tilt sensor.

Bringing the Pieces Together – Placing Core Facilities at the Core of Universities and Institutions: Lessons from Mergers, Acquisitions and Consolidations

Science.gov (United States)

Mundoma, Claudius

2013-01-01

As organizations expand and grow, the core facilities have become more dispersed disconnected. This is happening at a time when collaborations within the organization is a driver to increased productivity. Stakeholders are looking at the best way to bring the pieces together. It is inevitable that core facilities at universities and research institutes have to be integrated in order to streamline services and facilitate ease of collaboration. The path to integration often goes through consolidation, merging and shedding of redundant services. Managing this process requires a delicate coordination of two critical factors: the human (lab managers) factor and the physical assets factor. Traditionally more emphasis has been placed on reorganizing the physical assets without paying enough attention to the professionals who have been managing the assets for years, if not decades. The presentation focuses on how a systems approach can be used to effect a smooth core facility integration process. Managing the human element requires strengthening existing channels of communication and if necessary, creating new ones throughout the organization to break cultural and structural barriers. Managing the physical assets requires a complete asset audit and this requires direct input from the administration as well as the facility managers. Organizations can harness the power of IT to create asset visibility. Successfully managing the physical assets and the human assets increases productivity and efficiency within the organization.

Criticality experiment for No.2 core of DF-VI fast neutron criticality facility

International Nuclear Information System (INIS)

Yang Lijun; Liu Zhenhua; Yan Fengwen; Luo Zhiwen; Chu Chun; Liang Shuhong

2007-01-01

At the completion of the DF-VI fast neutron criticality facility, its core changed, and it was restarted and a series of experiments and measurements were made. According to the data from 29 criticality experiments, the criticality element number and mass were calculated, the control rod reactivity worth were measured by period method and rod compensate method, reactivity worth of safety rod and safety block were measured using reactivity instrument; the reactivity worth of outer elements and radial distribution of elements were measured too. Based on all the measurements mentioned above, safety operation parameters for core 2 in DF-VI fast neutron criticality facility were conformed. (authors)

An assessment of coupling algorithms for nuclear reactor core physics simulations

Energy Technology Data Exchange (ETDEWEB)

Hamilton, Steven, E-mail: hamiltonsp@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Berrill, Mark, E-mail: berrillma@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Clarno, Kevin, E-mail: clarnokt@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Pawlowski, Roger, E-mail: rppawlo@sandia.gov [Sandia National Laboratories, MS 0316, P.O. Box 5800, Albuquerque, NM 87185 (United States); Toth, Alex, E-mail: artoth@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Kelley, C.T., E-mail: tim_kelley@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Evans, Thomas, E-mail: evanstm@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States)

2016-04-15

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

Characterization Of Core Sample Collected From The Saltstone Disposal Facility

International Nuclear Information System (INIS)

Cozzi, A.; Duncan, A.

2010-01-01

During the month of September 2008, grout core samples were collected from the Saltstone Disposal Facility, Vault 4, cell E. This grout was placed during processing campaigns in December 2007 from Deliquification, Dissolution and Adjustment Batch 2 salt solution. The 4QCY07 Waste Acceptance Criteria sample collected on 11/16/07 represents the salt solution in the core samples. Core samples were retrieved to initiate the historical database of properties of emplaced Saltstone and to demonstrate the correlation between field collected and laboratory prepared samples. Three samples were collected from three different locations. Samples were collected using a two-inch diameter concrete coring bit. In April 2009, the core samples were removed from the evacuated sample container, inspected, transferred to PVC containers, and backfilled with nitrogen. Samples furthest from the wall were the most intact cylindrically shaped cored samples. The shade of the core samples darkened as the depth of coring increased. Based on the visual inspection, sample 3-3 was selected for all subsequent analysis. The density and porosity of the Vault 4 core sample, 1.90 g/cm 3 and 59.90% respectively, were comparable to values achieved for laboratory prepared samples. X-ray diffraction analysis identified phases consistent with the expectations for hydrated Saltstone. Microscopic analysis revealed morphology features characteristic of cementitious materials with fly ash and calcium silicate hydrate gel. When taken together, the results of the density, porosity, x-ray diffraction analysis and microscopic analysis support the conclusion that the Vault 4, Cell E core sample is representative of the expected waste form.

Effect of particle-core-vibration coupling near the double closed $^{132}$Sn nucleus from precise magnetic moment measurements

CERN Multimedia

Postma, H; Heyde, K; Walker, P; Grant, I; Veskovic, M; Stone, N; Stone, J

2002-01-01

% IS301 \\\\ \\\\ Low temperature nuclear orientation of isotope-separator implanted short-lived radio-isotopes makes possible the measurements of nuclear magnetic dipole moments of oriented ground and excited states with half-lives longer than a few seconds. Coupling schemes characterizing the odd nucleons and ground-state deformations can be extracted from the nuclear moments. \\\\ We thus propose to measure the magnetic dipole moments of $^{127-133}$Sb to high precision using NMR/ON at the NICOLE facility. With (double magic +1) $^{133}$Sb as the reference, the main aim of this experiment is to examine whether the collective component in the 7/2$^+$ Sb ground state magnetic dipole moment varies as expected according to particle-core coupling calculations carried out for the Sb (Z=51) isotopes. Comparison of the 1-proton-particle excitations in Sb to 1-proton-hole states in In nuclei will shed light on differences between particle and hole excitations as understood within the present model. Comparison of ...

Dosimetry Characteristics of Coupled Fast-Thermal Core 'HERBE'

International Nuclear Information System (INIS)

Pesic, M.; Milosevic, M.; Milovanovic, S.

1996-01-01

The 'HERBE' is new coupled fast-thermal core, designed in 1991, at the 'RB' reactor in the 'Vinca' Institute. It is used for verification of designed oriented computer codes developed in the Institute, training and sample irradiation in fast neutron field. For the last purpose a vertical experimental channel (VCH) is placed in the central axis of the fast core. Neutron spectrum in the centre of the VCR is calculated in 44 energy groups. Space distributions of two energy group neutron flux in the 'HERBE' are measured using gold foils and converted into the neutron absorbed dose (in air and tissue) using group flux-dose conversion factors. Gamma absorption doses in the air in the centre of the VCH are measured using calibrated small ionisation chamber filled with air. Determined dose rates are related to the reactor power. The first preliminary irradiations of silicon diodes (designed for production of the neutron dosemeters) in the centre of the VCH of the 'HERBE' fast core are carried out in 1994 and 1995. This paper describes calculation methods and measurement techniques applied to determination of the irradiation performance and dosimetry characteristics of the 'HERBE' system. (author)

Coupling-induced cooperative behaviour in dynamic ferromagnetic cores in the presence of a noise floor

International Nuclear Information System (INIS)

Bulsara, Adi R.; Lindner, John F.; In, Visarath; Kho, Andy; Baglio, Salvatore; Sacco, Vincenzo; Ando, Bruno; Longhini, Patrick; Palacios, Antonio; Rappel, Wouter-Jan

2006-01-01

Recently, we have shown the emergence of oscillations in overdamped undriven nonlinear dynamic systems subject to carefully crafted coupling schemes and operating conditions. Here, we summarize experimental results obtained on a system of N=3 coupled ferromagnetic cores, the underpinning of a 'coupled-core fluxgate magnetometer' (CCFM); the oscillatory behaviour is triggered when the coupling constant exceeds a threshold value (bifurcation point), and the oscillation frequency exhibits a characteristic scaling behaviour with the 'separation' of the coupling constant from its threshold value, as well as with an external target DC magnetic flux signal. The oscillations, which can be induced at frequencies ranging from a few Hz to high-kHz, afford a new detection scheme for weak target magnetic signals. We also present the first (numerical) results on the effects of a (Gaussian, exponentially correlated) noise floor on the spectral properties of the system response

Coupling-induced cooperative behaviour in dynamic ferromagnetic cores in the presence of a noise floor

Energy Technology Data Exchange (ETDEWEB)

Bulsara, Adi R. [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull Street, San Diego, CA 92152-5001 (United States)]. E-mail: bulsara@spawar.navy.mil; Lindner, John F. [Physics Department, College of Wooster, Wooster, OH 44691 (United States); In, Visarath [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull Street, San Diego, CA 92152-5001 (United States); Kho, Andy [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull Street, San Diego, CA 92152-5001 (United States); Baglio, Salvatore [Dipartimento di Ingegneria Elettrica Elettronica e dei Sistemi, Universita degli Studi di Catania, Viale A. Doria 6, 95125 Catania (Italy); Sacco, Vincenzo [Dipartimento di Ingegneria Elettrica Elettronica e dei Sistemi, Universita degli Studi di Catania, Viale A. Doria 6, 95125 Catania (Italy); Ando, Bruno [Dipartimento di Ingegneria Elettrica Elettronica e dei Sistemi, Universita degli Studi di Catania, Viale A. Doria 6, 95125 Catania (Italy); Longhini, Patrick [Nonlinear Dynamics Group, Department of Mathematics, San Diego State University, San Diego, CA 92182 (United States); Palacios, Antonio [Nonlinear Dynamics Group, Department of Mathematics, San Diego State University, San Diego, CA 92182 (United States); Rappel, Wouter-Jan [Physics Department, University of California at San Diego, La Jolla, CA 929093 (United States)

2006-04-17

Recently, we have shown the emergence of oscillations in overdamped undriven nonlinear dynamic systems subject to carefully crafted coupling schemes and operating conditions. Here, we summarize experimental results obtained on a system of N=3 coupled ferromagnetic cores, the underpinning of a 'coupled-core fluxgate magnetometer' (CCFM); the oscillatory behaviour is triggered when the coupling constant exceeds a threshold value (bifurcation point), and the oscillation frequency exhibits a characteristic scaling behaviour with the 'separation' of the coupling constant from its threshold value, as well as with an external target DC magnetic flux signal. The oscillations, which can be induced at frequencies ranging from a few Hz to high-kHz, afford a new detection scheme for weak target magnetic signals. We also present the first (numerical) results on the effects of a (Gaussian, exponentially correlated) noise floor on the spectral properties of the system response.

Steady-State Core Temperature Prediction Based on GAMMA+/CAPP Coupling

International Nuclear Information System (INIS)

Tak, Nam-il; Lee, Hyun-Chul; Lim, Hong-Sik

2015-01-01

In spite of sizable applications of the GAMMA+ code for the thermo-fluid analysis and design of a prismatic VHTR, the existing works are limited to stand-alone calculations. In the stand-alone calculations, information from the neutronic analysis (e.g., reactor power density profile) was considered only once i.e., when the calculations get started. For the neutronic analysis and design of a VHTR, the CAPP code, which is also under development at KAERI, is used. The main objective of this paper is to investigate the capability of GAMMA+ and CAPP coupling and to examine the results of the coupled analysis. Based on the coupling of GAMMA+ and CAPP, the steady-state core temperature was investigated in this work. It is found that the communication of data was successful. And the results of the GAMMA+ and CAPP coupling are found to be reasonable. The design modification of PMR200 is required to satisfy the design limit for the hot spot fuel temperature

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

Science.gov (United States)

Ruan, Qifeng

Localized surface plasmon resonance, which stems from the collective oscillations of conduction-band electrons, endows Au nanocrystals with unique optical properties. Au nanocrystals possess extremely large scattering/absorption cross-sections and enhanced local electromagnetic field, both of which are synthetically tunable. Moreover, when Au nanocrystals are closely placed or hybridized with semiconductors, the coupling and interaction between the individual components bring about more fascinating phenomena and promising applications, including plasmon-enhanced spectroscopies, solar energy harvesting, and cancer therapy. The continuous development in the field of plasmonics calls for further advancements in the preparation of high-quality plasmonic nanocrystals, the facile construction of hybrid plasmonic nanostructures with desired functionalities, as well as deeper understanding and efficient utilization of the interaction between plasmonic nanocrystals and semiconductor components. In this thesis, I developed a seed-mediated growth method for producing size-controlled Au nanospheres with high monodispersity and assembled Au nanospheres of different sizes into core/satellite nanostructures for enhancing Raman signals. For investigating the interactions between Au nanocrystals and semiconductors, I first prepared (Au core) (TiO2 shell) nanostructures, and then studied their synthetically controlled plasmonic properties and light-harvesting applications. Au nanocrystals with spherical shapes are desirable in plasmon-coupled systems owing to their high geometrical symmetry, which facilitates the analysis of electrodynamic responses in a classical electromagnetic framework and the investigation of quantum tunneling and nonlocal effects. I prepared remarkably uniform Au nanospheres with diameters ranging from 20 nm to 220 nm using a simple seed-mediated growth method associated with mild oxidation. Core/satellite nanostructures were assembled out of differently sized

Archive of Geosample Information from the British Ocean Sediment Core Research Facility (BOSCORF)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The British Ocean Sediment Core Research Facility (BOSCORF), National Oceanography Centre, is a contributor to the Index to Marine and Lacustrine Geological Samples...

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)

Interface engineered ferrite@ferroelectric core-shell nanostructures: A facile approach to impart superior magneto-electric coupling

Science.gov (United States)

Abraham, Ann Rose; Raneesh, B.; Das, Dipankar; Oluwafemi, Oluwatobi Samuel; Thomas, Sabu; Kalarikkal, Nandakumar

2018-04-01

The electric field control of magnetism in multiferroics is attractive for the realization of ultra-fast and miniaturized low power device applications like nonvolatile memories. Room temperature hybrid multiferroic heterostructures with core-shell (0-0) architecture (ferrite core and ferroelectric shell) were developed via a two-step method. High-Resolution Transmission Electron Microscopy (HRTEM) images confirm the core-shell structure. The temperature dependant magnetization measurements and Mossbauer spectra reveal superparamagnetic nature of the core-shell sample. The ferroelectric hysteresis loops reveal leaky nature of the samples. The results indicate the promising applications of the samples for magneto-electric memories and spintronics.

Coupled-core fluxgate magnetometer: Novel configuration scheme and the effects of a noise-contaminated external signal

International Nuclear Information System (INIS)

Palacios, Antonio; Aven, John; In, Visarath; Longhini, Patrick; Kho, Andy; Neff, Joseph D.; Bulsara, Adi

2007-01-01

Recent theoretical and experimental work has shown that unidirectional coupling can induce oscillations in overdamped and undriven nonlinear dynamical systems that are non-oscillatory when uncoupled; in turn, this has been shown to lead to new mechanisms for weak (compared to the energy barrier height) signal detection and amplification. The potential applications include fluxgate magnetometers, electric field sensors, and arrays of Superconducting Quantum Interference Device (SQUID) rings. In the particular case of the fluxgate magnetometer, we have developed a ''coupled-core fluxgate magnetometer'' (CCFM); this device has been realized in the laboratory and its dynamics used to quantify many properties that are generic to this class of systems and coupling. The CCFM operation is underpinned by the emergent oscillatory behavior in a unidirectionally coupled ring of wound ferromagnetic cores, each of which can be treated as an overdamped bistable dynamic system when uncoupled. In particular, one can determine the regimes of existence and stability of the (coupling-induced) oscillations, and the scaling behavior of the oscillation frequency. More recently, we studied the effects of a (Gaussian) magnetic noise floor on a CCFM system realized with N=3 coupled ferromagnetic cores. In this Letter, we first introduce a variation on the basic CCFM configuration that affords a path to enhanced device sensitivity, particularly for N>=3 coupled elements. We then analyze the response of the basic CCFM configuration as well as the new setup to a dc target signal that has a small noisy component (or ''contamination'')

Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

Energy Technology Data Exchange (ETDEWEB)

Travis, Adam R [ORNL

2014-05-01

A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

Simulation of the first step of the coupling of the PARCS/RELAP5 codes to ANGRA 2 facility

International Nuclear Information System (INIS)

Del Pozzo, Andrea Sanchez; Andrade, Delvonei A. de; Sabundjian, Gaiane

2015-01-01

Since the Three Mile Island (1979) and Chernobyl (1986) accidents, the International Agency of Energy Atomic (IAEA) has worked with the authorities of other countries that use nuclear power plants in order to guarantee the safe of those facilities. The utilities have simulated design basic accidents to verify the integrity of the nuclear power plant to these events. However, after Fukushima accident in Japan (2011), the people have felt insecure and been afraid in relation to nuclear power plants. Today, the international and national organizations, such as the International Agency of Energy Atomic (IAEA) and Comissao Nacional de Energia Nuclear (CNEN), respectively, have worked very hard to prevent some accidents and transients in nuclear power plants in order to ensure the security of the general population. In case of accidents, as the Rod Ejection Accident (REA), it is very important to do the coupling between neutronic and thermal hydraulic areas of nuclear reactors. To solve this type of problem there is the coupling between PARCS/RELAP5 codes. However, to perform this analysis it is necessary to simulate three steps. The first step is simulating the steady state of one nuclear power plant by using RELAP5 code. The second step is to run the steady state of this reactor using the coupling PARCS/RELAP5, and the final step is simulating the REA of this facility with PARCS/RELAP5 coupling. The aim of this work is to show the results of the first step of this analysis, i.e., by means of simulation the steady state of Angra 2 nuclear power plant using RELAP5 version 3.3. In this case, the modeling from the core was more detailed than in the original version developed some years ago for Angra 2. The results obtained in this work were satisfactory. (author)

Simulation of the first step of the coupling of the PARCS/RELAP5 codes to ANGRA 2 facility

Energy Technology Data Exchange (ETDEWEB)

Del Pozzo, Andrea Sanchez; Andrade, Delvonei A. de; Sabundjian, Gaiane, E-mail: delvonei@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

Since the Three Mile Island (1979) and Chernobyl (1986) accidents, the International Agency of Energy Atomic (IAEA) has worked with the authorities of other countries that use nuclear power plants in order to guarantee the safe of those facilities. The utilities have simulated design basic accidents to verify the integrity of the nuclear power plant to these events. However, after Fukushima accident in Japan (2011), the people have felt insecure and been afraid in relation to nuclear power plants. Today, the international and national organizations, such as the International Agency of Energy Atomic (IAEA) and Comissao Nacional de Energia Nuclear (CNEN), respectively, have worked very hard to prevent some accidents and transients in nuclear power plants in order to ensure the security of the general population. In case of accidents, as the Rod Ejection Accident (REA), it is very important to do the coupling between neutronic and thermal hydraulic areas of nuclear reactors. To solve this type of problem there is the coupling between PARCS/RELAP5 codes. However, to perform this analysis it is necessary to simulate three steps. The first step is simulating the steady state of one nuclear power plant by using RELAP5 code. The second step is to run the steady state of this reactor using the coupling PARCS/RELAP5, and the final step is simulating the REA of this facility with PARCS/RELAP5 coupling. The aim of this work is to show the results of the first step of this analysis, i.e., by means of simulation the steady state of Angra 2 nuclear power plant using RELAP5 version 3.3. In this case, the modeling from the core was more detailed than in the original version developed some years ago for Angra 2. The results obtained in this work were satisfactory. (author)

Coupling of the 3D neutron kinetic core model DYN3D with the CFD software ANSYS-CFX

International Nuclear Information System (INIS)

Grahn, Alexander; Kliem, Sören; Rohde, Ulrich

2015-01-01

Highlights: • Improved thermal hydraulic description of nuclear reactor cores. • Possibility of three-dimensional flow phenomena in the core, such as cross flow, flow reversal, flow around obstacles. • Simulation at higher spatial resolution as compared to system codes. - Abstract: This article presents the implementation of a coupling between the 3D neutron kinetic core model DYN3D and the commercial, general purpose computational fluid dynamics (CFD) software ANSYS-CFX. In the coupling approach, parts of the thermal hydraulic calculation are transferred to CFX for its better ability to simulate the three-dimensional coolant redistribution in the reactor core region. The calculation of the heat transfer from the fuel into the coolant remains with DYN3D, which incorporates well tested and validated heat transfer models for rod-type fuel elements. On the CFX side, the core region is modeled based on the porous body approach. The implementation of the code coupling is verified by comparing test case results with reference solutions of the DYN3D standalone version. Test cases cover mini and full core geometries, control rod movement and partial overcooling transients

Coupled-core fluxgate magnetometer: Novel configuration scheme and the effects of a noise-contaminated external signal

Energy Technology Data Exchange (ETDEWEB)

Palacios, Antonio [San Diego State University, Nonlinear Dynamical Systems Group, Department of Mathematics and Statistics, San Diego, CA 92182-7720 (United States)]. E-mail: palacios@euler.sdsu.edu; Aven, John [San Diego State University, Nonlinear Dynamical Systems Group, Department of Mathematics and Statistics, San Diego, CA 92182-7720 (United States); In, Visarath [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull St, San Diego, CA 92152-5001 (United States)]. E-mail: visarath@spawar.navy.mil; Longhini, Patrick [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull St, San Diego, CA 92152-5001 (United States); Kho, Andy [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull St, San Diego, CA 92152-5001 (United States); Neff, Joseph D. [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull St, San Diego, CA 92152-5001 (United States); Bulsara, Adi [Space and Naval Warfare Systems Center San Diego, Code 2363, 53560 Hull St, San Diego, CA 92152-5001 (United States)]. E-mail: bulsara@spawar.navy.mil

2007-07-16

Recent theoretical and experimental work has shown that unidirectional coupling can induce oscillations in overdamped and undriven nonlinear dynamical systems that are non-oscillatory when uncoupled; in turn, this has been shown to lead to new mechanisms for weak (compared to the energy barrier height) signal detection and amplification. The potential applications include fluxgate magnetometers, electric field sensors, and arrays of Superconducting Quantum Interference Device (SQUID) rings. In the particular case of the fluxgate magnetometer, we have developed a ''coupled-core fluxgate magnetometer'' (CCFM); this device has been realized in the laboratory and its dynamics used to quantify many properties that are generic to this class of systems and coupling. The CCFM operation is underpinned by the emergent oscillatory behavior in a unidirectionally coupled ring of wound ferromagnetic cores, each of which can be treated as an overdamped bistable dynamic system when uncoupled. In particular, one can determine the regimes of existence and stability of the (coupling-induced) oscillations, and the scaling behavior of the oscillation frequency. More recently, we studied the effects of a (Gaussian) magnetic noise floor on a CCFM system realized with N=3 coupled ferromagnetic cores. In this Letter, we first introduce a variation on the basic CCFM configuration that affords a path to enhanced device sensitivity, particularly for N>=3 coupled elements. We then analyze the response of the basic CCFM configuration as well as the new setup to a dc target signal that has a small noisy component (or ''contamination'')

Coupling and quantifying resilience and sustainability in facilities management

DEFF Research Database (Denmark)

Cox, Rimante Andrasiunaite; Nielsen, Susanne Balslev; Rode, Carsten

2015-01-01

Purpose – The purpose of this paper is to consider how to couple and quantify resilience and sustainability, where sustainability refers to not only environmental impact, but also economic and social impacts. The way a particular function of a building is provisioned may have significant repercus......Purpose – The purpose of this paper is to consider how to couple and quantify resilience and sustainability, where sustainability refers to not only environmental impact, but also economic and social impacts. The way a particular function of a building is provisioned may have significant...... repercussions beyond just resilience. The goal is to develop a decision support tool for facilities managers. Design/methodology/approach – A risk framework is used to quantify both resilience and sustainability in monetary terms. The risk framework allows to couple resilience and sustainability, so...... that the provisioning of a particular building can be investigated with consideration of functional, environmental, economic and, possibly, social dimensions. Findings – The method of coupling and quantifying resilience and sustainability (CQRS) is illustrated with a simple example that highlights how very different...

Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

Directory of Open Access Journals (Sweden)

Dong Hun Lee

2017-01-01

Full Text Available A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique because (i it has an inclined cooling channel with downwards-facing heating surface, of which flow processes are not fully exploited, (ii it is usually exposed to a low-pressure condition, where phase change causes dramatic changes in the flow, and (iii the effects of a multidimensional flow are very large in the upper part of the core catcher. These features make computational analysis more difficult. In this study, the MARS code is assessed using the two-phase natural circulation experiments that had been conducted at the CE-PECS facility to verify the cooling performance of a core catcher. The code is a system-scale thermal-hydraulic (TH code and has a multidimensional TH component. The facility was modeled by using both one- and three-dimensional components. Six experiments at the facility were selected to investigate the parametric effects of heat flux, pressure, and form loss. The results show that MARS can predict the two-phase flow at the facility reasonably well. However, some limitations are obviously revealed.

Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion

International Nuclear Information System (INIS)

Marques, J.G.; Sousa, M.; Santos, J.P.; Fernandes, A.C.

2011-01-01

The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1 MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

Low coupling loss core-strengthened Bi 2212\\/Ag Rutherford cables

CERN Document Server

Collings, E W; Scanlan, R M; Dietderich, D R; Motowidlo, L R

1999-01-01

In a comprehensive "vertically integrated" program multifilamentary (MF) high temperature superconducting (HTSC) Bi:2212/Ag strand was fabricated using the powder-in-tube process and heat treated in oxygen by a modified standard $9 procedure. The reaction-heat-treatment (HT) was adjusted to maximize critical current (density), I/sub c/ (J /sub c/), as measured in various magnetic fields, B. A series of Rutherford cables was designed, each of which included a $9 metallic (Nichrome-80) core for strengthening and reduction of coupling loss. Prior to cable winding a series of tests examined the possibility of strand "poisoning" by the core during HT. Small model Rutherford cables were wound, $9 and after HT were prepared for I/sub c/(B) measurement and calorimetric measurement of AC loss and hence interstrand contact resistance I/sub c/(B). It was deduced that, if in direct contact with the strand during HT, the core $9 material can degrade the I/sub c/ of the cable; but steps can be taken to eliminate this probl...

Fluid-structure coupled dynamic response of PWR core barrel during LOCA

International Nuclear Information System (INIS)

Lu, M.W.; Zhang, Y.G.; Shi, F.

1991-01-01

This paper is engaged in the Fluid-Structure Interaction LOCA analysis of the core barrel of PWR. The analysis is performed by a multipurpose computer code SANES. The FSI inside the pressure vessel is treated by a FEM code including some structural and acoustic elements. The transient in the primary loop is solved by a two-phase flow code. Both codes are coupled one another. Some interesting conclusions are drawn. (author)

A fiber optic temperature sensor based on multi-core microstructured fiber with coupled cores for a high temperature environment

Science.gov (United States)

Makowska, A.; Markiewicz, K.; Szostkiewicz, L.; Kolakowska, A.; Fidelus, J.; Stanczyk, T.; Wysokinski, K.; Budnicki, D.; Ostrowski, L.; Szymanski, M.; Makara, M.; Poturaj, K.; Tenderenda, T.; Mergo, P.; Nasilowski, T.

2018-02-01

Sensors based on fiber optics are irreplaceable wherever immunity to strong electro-magnetic fields or safe operation in explosive atmospheres is needed. Furthermore, it is often essential to be able to monitor high temperatures of over 500°C in such environments (e.g. in cooling systems or equipment monitoring in power plants). In order to meet this demand, we have designed and manufactured a fiber optic sensor with which temperatures up to 900°C can be measured. The sensor utilizes multi-core fibers which are recognized as the dedicated medium for telecommunication or shape sensing, but as we show may be also deployed advantageously in new types of fiber optic temperature sensors. The sensor presented in this paper is based on a dual-core microstructured fiber Michelson interferometer. The fiber is characterized by strongly coupled cores, hence it acts as an all-fiber coupler, but with an outer diameter significantly wider than a standard fused biconical taper coupler, which significantly increases the coupling region's mechanical reliability. Owing to the proposed interferometer imbalance, effective operation and high-sensitivity can be achieved. The presented sensor is designed to be used at high temperatures as a result of the developed low temperature chemical process of metal (copper or gold) coating. The hermetic metal coating can be applied directly to the silica cladding of the fiber or the fiber component. This operation significantly reduces the degradation of sensors due to hydrolysis in uncontrolled atmospheres and high temperatures.

Verification of kinetic parameters of coupled fast-thermal core HERBE

International Nuclear Information System (INIS)

Pesic, M.; Marinkovic, P.; Milosevic, M.; Nikolic, D.; Zavaljevski, N.; Milovanovic, S.; Ljubenov, V.

1997-03-01

The HERBE system is a new coupled fast-thermal core constructed in 1989 at the RB critical heavy water assembly at the VINCA Institute. It was designed with the aim to improve experimental possibilities in fast neutron fields and for experimental verification of reactor design-oriented methods. This paper overviews experiments for kinetic parameters verification carried out at HERBE system. Their short description and comparison of experimental and calculation results are included. A brief introduction to the computer codes used in the calculations is presented too. (author)

Obliquity histories of Earth and Mars: Influence of inertial and dissipative core-mantle coupling

International Nuclear Information System (INIS)

Bills, B.G.

1990-01-01

For both the Earth and Mars, secular variations in the angular separation of the spin axis from the orbit normal are suspected of driving major climatic changes. There is considerable interest in determining the amplitude and timing of these obliquity variations. If the orientation of the orbital plane were inertially fixed, the spin axis would simply precess around the orbit at a fixed obliquity and at a uniform angular rate. The precession rate parameter depends on the principal moments of inertia and rotation rate of the perturbed body, and on the gravitational masses and semiminor axes of the perturbing bodies. For Mars, the precession rate is not well known, but probably lies in the interval 8 to 10 arcsec/year. In the rigid body case, the spin axis still attempts to precess about the instantaneous orbit normal, but now the obliquity varies. The hydrostatic figure of a planet represents a compromise between gravitation, which attempts to attain spherical symmetry, and rotation, which prefers cylindrical symmetry. Due to their higher mean densities the cores of the Earth and Mars will be more nearly spherical than the outer layers of these planets. On short time scales it is appropriate to consider the core to be an inviscid fluid. The inertial coupling provided by this mechanism is effective whenever the ellipticicy of the container exceeds the ratio of precessional to rotational rates. If the mantle were actually rigid, this would be an extremely effective type of coupling. However, on sufficiently long time scales, the mantle will deform viscously and can accommodate the motions of the core fluid. A fundamentally different type of coupling is provided by electromagnetic or viscous torques. This type of coupling is likely to be most important on longer time scales

Measured lifetimes of states in 197Au and a critical comparison with the weak-coupling core-excitation model

International Nuclear Information System (INIS)

Bolotin, H.H.; Kennedy, D.L.; Linard, B.J.; Stuchbery, A.E.

1979-01-01

The lifetimes of five excited states in 197 Au up to an excitation energy of 885 keV were measured by the recoil-distance method (RDM). These levels were populated by Coulomb excitation using both 90 MeV 20 Ne and 120 MeV 35 Cl ion beams. The experimentally determined spectroscopy of the low-lying levels 3/2 + (ground state) and 1/2 + , (3/2) + 2 , 5/2 + and 7/2 + at 77.3, 268.8, 278.9, and 547.5 keV excitation energy, respectively, has been critically compared with the detailed predictions of the de-Shalit weak-coupling core-excitation model. When the model is taken to represent the case of a dsub(3/2) proton hole coupled to a 198 Hg core, the model parameters obtained are in accord with the criteria implicit for weak core coupling and, at the same time, are in remarkably good agreement with virtually all measured E2 and M1 transition rates. (Auth.)

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

International Nuclear Information System (INIS)

Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

2011-01-01

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

Obliquity histories of Earth and Mars: Influence of inertial and dissipative core-mantle coupling

Science.gov (United States)

Bills, Bruce G.

1990-01-01

For both the Earth and Mars, secular variations in the angular separation of the spin axis from the orbit normal are suspected of driving major climatic changes. There is considerable interest in determining the amplitude and timing of these obliquity variations. If the orientation of the orbital plane were inertially fixed, and the planet were to act as a rigid body in it response to precessional torques, the spin axis would simply precess around the orbit at a fixed obliquity and at a uniform angular rate. The precession rate parameter depends on the principal moments of inertia and rotation rate of the perturbed body, and on the gravitational masses and semiminor axes of the perturbing bodies. For Mars, the precession rate is not well known, but probably lies in the interval 8 to 10 arcsec/year. Gravitational interactions between the planets lead to secular motions of the orbit planes. In the rigid body case, the spin axis still attempts to precess about the instantaneous orbit normal, but now the obliquity varies. The hydrostatic figure of a planet represents a compromise between gravitation, which attempts to attain spherical symmetry, and rotation, which prefers cylindrical symmetry. Due to their higher mean densities the cores of the Earth and Mars will be more nearly spherical than the outer layers of these planets. On short time scales it is appropriate to consider the core to be an inviscid fluid constrained to move with the ellipsoidal region bounded by the rigid mantle. The inertial coupling provided by this mechanism is effective whenever the ellipticicy of the container exceeds the ratio of precessional to rotational rates. If the mantle were actually rigid, this would be an extremely effective type of coupling. However, on sufficiently long time scales, the mantle will deform viscously and can accommodate the motions of the core fluid. A fundamentally different type of coupling is provided by electromagnetic or viscous torques. This type of coupling

Altered intrinsic functional coupling between core neurocognitive networks in Parkinson's disease

Directory of Open Access Journals (Sweden)

Deepti Putcha

2015-01-01

Full Text Available Parkinson's disease (PD is largely attributed to disruptions in the nigrostriatal dopamine system. These neurodegenerative changes may also have a more global effect on intrinsic brain organization at the cortical level. Functional brain connectivity between neurocognitive systems related to cognitive processing is critical for effective neural communication, and is disrupted across neurological disorders. Three core neurocognitive networks have been established as playing a critical role in the pathophysiology of many neurological disorders: the default-mode network (DMN, the salience network (SN, and the central executive network (CEN. In healthy adults, DMN–CEN interactions are anti-correlated while SN–CEN interactions are strongly positively correlated even at rest, when individuals are not engaging in any task. These intrinsic between-network interactions at rest are necessary for efficient suppression of the DMN and activation of the CEN during a range of cognitive tasks. To identify whether these network interactions are disrupted in individuals with PD, we used resting state functional magnetic resonance imaging (rsfMRI to compare between-network connectivity between 24 PD participants and 20 age-matched controls (MC. In comparison to the MC, individuals with PD showed significantly less SN–CEN coupling and greater DMN–CEN coupling during rest. Disease severity, an index of striatal dysfunction, was related to reduced functional coupling between the striatum and SN. These results demonstrate that individuals with PD have a dysfunctional pattern of interaction between core neurocognitive networks compared to what is found in healthy individuals, and that interaction between the SN and the striatum is even more profoundly disrupted in those with greater disease severity.

Model of coupling with core in the Green function method

International Nuclear Information System (INIS)

Kamerdzhiev, S.P.; Tselyaev, V.I.

1983-01-01

Models of coupling with core in the method of the Green functions, presenting generalization of conventional method of chaotic phases, i.e. account of configurations of more complex than monoparticle-monohole (1p1h) configurations, have been considered. Odd nuclei are studied only to the extent when the task of odd nucleus is solved for even-even nucleus. Microscopic model of the account of delay effects in mass operator M=M(epsilon), which corresponds to the account of the effects influence only on the change of quasiparticle behaviour in magic nucleus as compared with their behaviour, described by pure model of cores, has been considered. The change results in fragmentation of monoparticle levels, which is the main effect, and in the necessity to use new basis as compared with the shell one, corresponding to inoculative quasiparticles. When formulas have been devived concrete type of mass operator M(epsilon) is not used

Tapered silicon core fibers with nano-spikes for optical coupling via spliced silica fibers

OpenAIRE

Ren, Haonan; Aktas, Ozan; Franz, Yohann; Runge, Antoine; Hawkins, Thomas A.; Ballato, John; Gibson, Ursula; Peacock, Anna

2017-01-01

Abstract: Reported here is the fabrication of tapered silicon core fibers possessing a nanospike input that facilitates their seamless splicing to conventional single mode fibers. A proofof-concept 30 µm cladding diameter fiber-based device is demonstrated with nano-spike coupling and propagation losses below 4 dB and 2 dB/cm, respectively. Finite-elementmethod-based simulations show that the nano-spike coupling losses could be reduced to below 1 dB by decreasing the cladding diameters down t...

Coupling of the computational fluid dynamics code ANSYS CFX with the 3D neutron kinetic core model DYN3D

International Nuclear Information System (INIS)

Kliem, S.; Grahn, A.; Rohde, U.; Schuetze, J.; Frank, Th.

2010-01-01

The computational fluid dynamics code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactors coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for two small-size test problems confirm the correctness of the implementation of the prototype coupling. The first test problem was a mini-core consisting of nine real-size fuel assemblies with quadratic cross section. Comparison was performed with the DYN3D stand-alone code. In the steady state, the effective multiplication factor obtained by the DYN3D/ANSYS CFX codes hows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power in the same mini-core. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. The same calculations were carried for a mini-core with seven real-size fuel assemblies with hexagonal cross section in

Development of the test facilities for the measurement of core flow and pressure distribution of SMART reactor

International Nuclear Information System (INIS)

Ko, Y.J.; Euh, D.J.; Youn, Y.J.; Chu, I.C.; Kwon, T.S.

2011-01-01

A design of SMART reactor has been developed, of which the primary system is composed of four internal circulation pumps, a core of 57 fuel assemblies, eight cassettes of steam generators, flow mixing head assemblies, and other internal structures. Since primary design features are very different from conventional reactors, the characteristics of flow and pressure distribution are expected to be different accordingly. In order to analyze the thermal margin and hydraulic design characteristics of SMART reactor, design quantification tests for flow and pressure distribution with a preservation of flow geometry are necessary. In the present study, the design feature of the test facility in order to investigate flow and pressure distribution, named “SCOP” is described. In order to preserve the flow distribution characteristics, the SCOP is linearly reduced with a scaling ratio of 1/5. The core flow rate of each fuel assembly is measured by a venturi meter attached in the lower part of the core simulator having a similarity of pressure drop for nominally scaled flow conditions. All the 57 core simulators and 8 S/G simulators are precisely calibrated in advance of assembling in test facilities. The major parameters in tests are pressures, differential pressures, and core flow distribution. (author)

Applications for coupled core neutronics and thermal-hydraulic models

International Nuclear Information System (INIS)

Eller, J.

1996-01-01

The unprecedented increases in computing capacity that have occurred during the last decade have affected our sciences, and thus our lives, to an extent that is difficult to overstate. All indications are that this trend will continue for years to come. Nuclear reactor systems analysis is one of many areas of engineering that has changed dramatically as a result of this evolution. Our ability to model the various mechanical and physical systems in greater and greater detail has allowed significant improvements in operational efficiency in spite of increasing regulatory requirements. Many of these efficiencies result from the use of more complex and geometrically detailed computer modeling, which is used to justify a reduction or elimination of some of the conservatisms required by earlier, less sophisticated analyses. And more recently, as our industries open-quotes downsize,close quotes efforts are being made to find ways to use the ever-increasing computing capacity to design systems that accomplish more work, in less time, and with fewer people. The balance of this paper discusses some of the visions that Duke Power Company feels would most benefit their particular methodologies. One of the concepts receiving a lot of attention involves an automated coupling of a thermal-hydraulic plant systems analysis model to a three-dimensional core neutronics program. The thermal-hydraulic analysis of several postulated system transients incorporates large conservatisms because of limited ability to model complex time-dependent asymmetric heat sources in adequate geometric detail. For these transients, the core behavior is closely coupled with the thermal-hydraulic behavior of the total plant system and vice versa. Steam-line break, uncontrolled rod withdrawal, and rod drop anayses are likely to benefit most from this type of linked process

Direct measurement of the transition from edge to core power coupling in a light-ion helicon source

Science.gov (United States)

Piotrowicz, P. A.; Caneses, J. F.; Showers, M. A.; Green, D. L.; Goulding, R. H.; Caughman, J. B. O.; Biewer, T. M.; Rapp, J.; Ruzic, D. N.

2018-05-01

We present time-resolved measurements of an edge-to-core power transition in a light-ion (deuterium) helicon discharge in the form of infra-red camera imaging of a thin stainless steel target plate on the Proto-Material Exposure eXperiment device. The time-resolved images measure the two-dimensional distribution of power deposition in the helicon discharge. The discharge displays a mode transition characterized by a significant increase in the on-axis electron density and core power coupling, suppression of edge power coupling, and the formation of a fast-wave radial eigenmode. Although the self-consistent mechanism that drives this transition is not yet understood, the edge-to-core power transition displays characteristics that are consistent with the discharge entering a slow-wave anti-resonant regime. RF magnetic field measurements made across the plasma column, together with the power deposition results, provide direct evidence to support the suppression of the slow-wave in favor of core plasma production by the fast-wave in a light-ion helicon source.

Automatized facility for express gamma-spectrometric studies of full-size core and muds of oil-gas wells

International Nuclear Information System (INIS)

Antropov, S.Yu.; Ermilov, A.P.; Ermilov, S.A.; Komarov, N.A.; Krokhin, I.I.

2005-01-01

In the automatized facility 'Sputnik-Geo' for automated core feeding of the conveyor belt with 5 m length is using. Activity measurement has being conducted by four scintillation gamma radiation detectors. Symmetrical location of the detectors relatively the core allows to exclude of the geometrical uncertainty of component. The spectrum processing and the control by conveyor engine are carried out by computer, which preserves the results of measurements in the database. For density measurement the scintillation detector with collimated source on the 137 Cs radionuclide base is applied. The facility is provided with the light-diode coefficient stabilization of detector amplification, that permits to operation time increase without spectrometers requalification by energy up to 1 working day

Core-edge coupling and the effect of the edge on overall plasma performance

International Nuclear Information System (INIS)

Fichtmueller, M.; Corrigan, G.; Lauro-Taroni, L.

1999-01-01

Several attempts to model the entire plasma cross section have been reported in the last few years. Two possibilities are to either couple a core code to a scrape-off layer (SOL) code at a specified interface or to extend the computational region of an SOL-code all the way to the plasma centre. The most advanced global code is the code COCONUT which is based on the former principle and comprises the Monte-Carlo code NIMBUS, the 2D scrape-off layer code EDGE2D, the core transport code JETTO and the core impurity transport code SANCO. A main feature of COCONUT is its modular structure which ensures a high degree of flexibility and the capability to cover a large range of time-scales. The influence of the SOL on the core is illustrated with a range of global simulations carried out with COCONUT. The simulations show that the primary effect of the SOL is the control of the particle sources and sinks with a secondary effect on plasma dilution, radiation and perhaps pedestal temperatures. (author)

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR

International Nuclear Information System (INIS)

Kurosawa, M.

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54 Mn and 60 Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. (authors)

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

Science.gov (United States)

Kurosawa, Masahiko

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

Study on development of virtual reactor core laboratory (1). Development of prototype coupled neutronic, thermal-hydraulic and structural analysis system

International Nuclear Information System (INIS)

Uto, Nariaki; Sugaya, Toshio; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sakai, Takaaki

1999-09-01

A study on development of virtual reactor core laboratory, which is to conduct numerical experiments representative of complicated physical phenomena in practical reactor core systems on a computational environment, has progressed at Japan Nuclear Cycle Development Institute (JNC). The study aims at systematic evaluation of these phenomena into which nuclear reactions, thermal-hydraulic characteristics, structural responses and fuel behaviors combine, and effective utilization of the obtained comprehension for core design. This report presents a production of a prototype computational system which is required to construct the virtual reactor core laboratory. This system is to evaluate reactor core performance under the coupled neutronic, thermal-hydraulic and structural phenomena, and is composed of two analysis tools connected by a newly developed interface program; 1) an existing space-dependent coupled neutronic and thermal-hydraulic analysis system arranged at JNC and 2) a core deformation analysis code. It acts on a cluster of several DEC/Alpha workstations. A specific library called MPI1 (Message Passing Interface 1) is incorporated as a tool for communicating among the analysis modules consisting of the system. A series of calculations for simulating a sequence of Unprotected Loss Of Heat Sink (ULOHS) coupled with rapid drop of some neutron absorber devices in a prototype fast reactor is tried to investigate how the system works. The obtained results show the core deformation behavior followed by the reactivity change that can be properly evaluated. The results of this report show that the system is expected to be useful for analyzing sensitivity of reactor core performance with respect to uncertainties of various design parameters and establishing a concept of passive safety reactor system, taking into account space distortion of neutron flux distribution during abnormal events as well as reactivity feedback from core deformation. (author)

A High-Throughput Biological Calorimetry Core: Steps to Startup, Run, and Maintain a Multiuser Facility.

Science.gov (United States)

Yennawar, Neela H; Fecko, Julia A; Showalter, Scott A; Bevilacqua, Philip C

2016-01-01

Many labs have conventional calorimeters where denaturation and binding experiments are setup and run one at a time. While these systems are highly informative to biopolymer folding and ligand interaction, they require considerable manual intervention for cleaning and setup. As such, the throughput for such setups is limited typically to a few runs a day. With a large number of experimental parameters to explore including different buffers, macromolecule concentrations, temperatures, ligands, mutants, controls, replicates, and instrument tests, the need for high-throughput automated calorimeters is on the rise. Lower sample volume requirements and reduced user intervention time compared to the manual instruments have improved turnover of calorimetry experiments in a high-throughput format where 25 or more runs can be conducted per day. The cost and efforts to maintain high-throughput equipment typically demands that these instruments be housed in a multiuser core facility. We describe here the steps taken to successfully start and run an automated biological calorimetry facility at Pennsylvania State University. Scientists from various departments at Penn State including Chemistry, Biochemistry and Molecular Biology, Bioengineering, Biology, Food Science, and Chemical Engineering are benefiting from this core facility. Samples studied include proteins, nucleic acids, sugars, lipids, synthetic polymers, small molecules, natural products, and virus capsids. This facility has led to higher throughput of data, which has been leveraged into grant support, attracting new faculty hire and has led to some exciting publications. © 2016 Elsevier Inc. All rights reserved.

Modular synthesis of the pyrimidine core of the manzacidins by divergent Tsuji–Trost coupling

Directory of Open Access Journals (Sweden)

Sebastian Bretzke

2016-06-01

Full Text Available The design, development and application of an efficient procedure for the concise synthesis of the 1,3-syn- and anti-tetrahydropyrimidine cores of manzacidins are reported. The intramolecular allylic substitution reaction of a readily available joint urea-type substrate enables the facile preparation of both diastereomers in high yields. The practical application of this approach is demonstrated in the efficient and modular preparation of the authentic heterocyclic cores of manzacidins, structurally unique bromopyrrole alkaloids of marine origin. Additional features of this route include the stereoselective generation of the central amine core with an appending quaternary center by an asymmetric addition of a Grignard reagent to a chiral tert-butanesulfinyl ketimine following an optimized Ellman protocol and a cross-metathesis of a challenging homoallylic urea substrate, which proceeds in good yields in the presence of an organic phosphoric acid.

Facile synthesis and excellent microwave absorption properties of FeCo-C core-shell nanoparticles

Science.gov (United States)

Liang, Bingbing; Wang, Shiliang; Kuang, Daitao; Hou, Lizhen; Yu, Bowen; Lin, Liangwu; Deng, Lianwen; Huang, Han; He, Jun

2018-02-01

FeCo-C core-shell nanoparticles (NPs) with diameters of 10-50 nm have been fabricated on a large scale by one-step metal-organic chemical vapor deposition using the mixture of cobalt acetylacetonate and iron acetylacetonate as the precursor. The Fe/Co molar ratio of the alloy nanocores and graphitization degree of C shells, and thus the magnetic and electric properties of the core-shell NPs, can be tuned by the deposition temperature ranging from 700 °C to 900 °C. Comparative tests reveal that a relatively high Fe/Co molar ratio and low graphitization degree benefit the microwave absorption (MA) performance of the core-shell NPs. The composite with 20 wt% core-shell NP obtained at 800 °C and 80 wt% paraffin exhibits an optimal reflection loss ({{R}}{{L}}) of -60.4 dB at 7.5 GHz with a thickness of 3.3 mm, and an effective absorption bandwidth (frequency range for RL ≤10 dB) of 9.2 GHz (8.8-18.0 GHz) under an absorber thickness of 2.5 mm. Our study provides a facile route for the fabrication of alloy-C core-shell nanostructures with high MA performance.

Magnetic field reversals, polar wander, and core-mantle coupling.

Science.gov (United States)

Courtillot, V; Besse, J

1987-09-04

True polar wander, the shifting of the entire mantle relative to the earth's spin axis, has been reanalyzed. Over the last 200 million years, true polar wander has been fast (approximately 5 centimeters per year) most of the time, except for a remarkable standstill from 170 to 110 million years ago. This standstill correlates with a decrease in the reversal frequency of the geomagnetic field and episodes of continental breakup. Conversely, true polar wander is high when reversal frequency increases. It is proposed that intermittent convection modulates the thickness of a thermal boundary layer at the base of the mantle and consequently the core-to-mantle heat flux. Emission of hot thermals from the boundary layer leads to increases in mantle convection and true polar wander. In conjunction, cold thermals released from a boundary layer at the top of the liquid core eventually lead to reversals. Changes in the locations of subduction zones may also affect true polar wander. Exceptional volcanism and mass extinctions at the Cretaceous-Tertiary and Permo-Triassic boundaries may be related to thermals released after two unusually long periods with no magnetic reversals. These environmental catastrophes may therefore be a consequence of thermal and chemical couplings in the earth's multilayer heat engine rather than have an extraterrestrial cause.

Extension of the analytic nodal diffusion solver ANDES to triangular-Z geometry and coupling with COBRA-IIIc for hexagonal core analysis

International Nuclear Information System (INIS)

Lozano, Juan-Andres; Jimenez, Javier; Garcia-Herranz, Nuria; Aragones, Jose-Maria

2010-01-01

In this paper the extension of the multigroup nodal diffusion code ANDES, based on the Analytic Coarse Mesh Finite Difference (ACMFD) method, from Cartesian to hexagonal geometry is presented, as well as its coupling with the thermal-hydraulic (TH) code COBRA-IIIc for hexagonal core analysis. In extending the ACMFD method to hexagonal assemblies, triangular-Z nodes are used. In the radial plane, a direct transverse integration procedure is applied along the three directions that are orthogonal to the triangle interfaces. The triangular nodalization avoids the singularities, that appear when applying transverse integration to hexagonal nodes, and allows the advantage of the mesh subdivision capabilities implicit within that geometry. As for the thermal-hydraulics, the extension of the coupling scheme to hexagonal geometry has been performed with the capability to model the core using either assembly-wise channels (hexagonal mesh) or a higher refinement with six channels per fuel assembly (triangular mesh). Achieving this level of TH mesh refinement with COBRA-IIIc code provides a better estimation of the in-core 3D flow distribution, improving the TH core modelling. The neutronics and thermal-hydraulics coupled code, ANDES/COBRA-IIIc, previously verified in Cartesian geometry core analysis, can also be applied now to full three-dimensional VVER core problems, as well as to other thermal and fast hexagonal core designs. Verification results are provided, corresponding to the different cases of the OECD/NEA-NSC VVER-1000 Coolant Transient Benchmarks.

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.

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

Updated procedures for using drill cores and cuttings at the Lithologic Core Storage Library, Idaho National Laboratory, Idaho

Science.gov (United States)

Hodges, Mary K.V.; Davis, Linda C.; Bartholomay, Roy C.

2018-01-30

In 1990, the U.S. Geological Survey, in cooperation with the U.S. Department of Energy Idaho Operations Office, established the Lithologic Core Storage Library at the Idaho National Laboratory (INL). The facility was established to consolidate, catalog, and permanently store nonradioactive drill cores and cuttings from subsurface investigations conducted at the INL, and to provide a location for researchers to examine, sample, and test these materials.The facility is open by appointment to researchers for examination, sampling, and testing of cores and cuttings. This report describes the facility and cores and cuttings stored at the facility. Descriptions of cores and cuttings include the corehole names, corehole locations, and depth intervals available.Most cores and cuttings stored at the facility were drilled at or near the INL, on the eastern Snake River Plain; however, two cores drilled on the western Snake River Plain are stored for comparative studies. Basalt, rhyolite, sedimentary interbeds, and surficial sediments compose most cores and cuttings, most of which are continuous from land surface to their total depth. The deepest continuously drilled core stored at the facility was drilled to 5,000 feet below land surface. This report describes procedures and researchers' responsibilities for access to the facility and for examination, sampling, and return of materials.

Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

International Nuclear Information System (INIS)

Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

1982-01-01

The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling

Assessment of uncertainties in core melt phenomenology and their impact on risk at the Z/IP facilities

International Nuclear Information System (INIS)

Pratt, W.T.; Ludewig, H.; Bari, R.A.; Meyer, J.F.

1983-01-01

An evaluation of core meltdown accidents in the Z/IP facilities has been performed. Containment event trees have been developed to relate the progression of a given accident to various potential containment building failure modes. An extensive uncertainty analysis related to core melt phenomenology has been performed. A major conclusion of the study is that large variations in parameters associated with major phenomenological uncertainties have a relatively minor impact on risk when external initiators are considered. This is due to the inherent capability fo the Z/IP containment buildings to contain a wide range of core meltdown accidents. 12 references, 2 tables

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

A facile way to realize exchange coupling interaction in hard/soft magnetic composites

Energy Technology Data Exchange (ETDEWEB)

Li, Dongyun, E-mail: lidongyun@cjlu.edu.cn [College of Materials Science and Engineering, China Jiliang University, Hangzhou 310018 (China); Wang, Fan [College of Materials Science and Engineering, China Jiliang University, Hangzhou 310018 (China); Xia, Ailin, E-mail: alxia@126.com [Anhui Key Laboratory of Metal Materials and Processing, School of Materials Science and Engineering, Anhui University of Technology, Maanshan 243032 (China); Zhang, Lijiao [School of Science, Hebei University of Science and Technology, Shijiazhuang 050018 (China); Li, Tingting; Jin, Chuangui; Liu, Xianguo [Anhui Key Laboratory of Metal Materials and Processing, School of Materials Science and Engineering, Anhui University of Technology, Maanshan 243032 (China)

2016-11-01

SrFe{sub 12}O{sub 19}/CoFe{sub 2}O{sub 4} and SrFe{sub 12}O{sub 19}/Fe–B hard/soft magnetic composites were obtained by using powders synthesized via a hydrothermal and a molten salt method, respectively. The exchange coupling interaction was found to exist in the composites after a facile grinding according to the results of magnetic hysteresis loops and irreversible sloping recoil loops. It can be found that different grinding time affects their magnetic properties slightly. Our study proves that the conditions of realizing exchange coupling interaction may not be so stringent. - Highlights: • SrM/CFO and SrM/Fe–B with exchange coupling were obtained via a grinding way. • Different grinding time affects their magnetic properties slightly. • The conditions of realizing exchange coupling may not be so stringent.

Whole core pin-by-pin coupled neutronic-thermal-hydraulic steady state and transient calculations using COBAYA3 code

International Nuclear Information System (INIS)

Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.

2010-10-01

Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)

Nanoscale semiconductor-insulator-metal core/shell heterostructures: facile synthesis and light emission

Science.gov (United States)

Li, Gong Ping; Chen, Rui; Guo, Dong Lai; Wong, Lai Mun; Wang, Shi Jie; Sun, Han Dong; Wu, Tom

2011-08-01

Controllably constructing hierarchical nanostructures with distinct components and designed architectures is an important theme of research in nanoscience, entailing novel but reliable approaches of bottom-up synthesis. Here, we report a facile method to reproducibly create semiconductor-insulator-metal core/shell nanostructures, which involves first coating uniform MgO shells onto metal oxide nanostructures in solution and then decorating them with Au nanoparticles. The semiconductor nanowire core can be almost any material and, herein, ZnO, SnO2 and In2O3 are used as examples. We also show that linear chains of short ZnO nanorods embedded in MgO nanotubes and porous MgO nanotubes can be obtained by taking advantage of the reduced thermal stability of the ZnO core. Furthermore, after MgO shell-coating and the appropriate annealing treatment, the intensity of the ZnO near-band-edge UV emission becomes much stronger, showing a 25-fold enhancement. The intensity ratio of the UV/visible emission can be increased further by decorating the surface of the ZnO/MgO nanowires with high-density plasmonic Au nanoparticles. These heterostructured semiconductor-insulator-metal nanowires with tailored morphologies and enhanced functionalities have great potential for use as nanoscale building blocks in photonic and electronic applications.Controllably constructing hierarchical nanostructures with distinct components and designed architectures is an important theme of research in nanoscience, entailing novel but reliable approaches of bottom-up synthesis. Here, we report a facile method to reproducibly create semiconductor-insulator-metal core/shell nanostructures, which involves first coating uniform MgO shells onto metal oxide nanostructures in solution and then decorating them with Au nanoparticles. The semiconductor nanowire core can be almost any material and, herein, ZnO, SnO2 and In2O3 are used as examples. We also show that linear chains of short ZnO nanorods embedded in

2-µm wavelength-range low-loss inhibited-coupling hollow-core PCF

Science.gov (United States)

Maurel, M.; Chafer, M.; Delahaye, F.; Amrani, F.; Debord, B.; Gerome, F.; Benabid, F.

2018-02-01

We report on the design and fabrication of inhibited-coupling guiding hollow-core photonic crystal fiber with a transmission band optimized for low loss guidance around 2 μm. Two fibers design based on a Kagome-lattice cladding have been studied to demonstrate a minimum loss figure of 25 dB/km at 2 μm associated to an ultra-broad transmission band spanning from the visible to our detection limit of 3.4 μm. Such fibers could be an excellent tool to deliver and compress ultra-short pulse laser systems, especially for the emerging 2-3 μm spectral region.

Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

International Nuclear Information System (INIS)

Conklin, J.C.

1986-08-01

The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

Coupled study of the Molten Salt Fast Reactor core physics and its associated reprocessing unit

International Nuclear Information System (INIS)

Doligez, X.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Ghetta, V.

2014-01-01

Highlights: • The limit on the reprocessing is due to the redox potential control. • Alkali and Earth-alkaline elements do not have to be extracted. • Criticality risks have to be studied in the reprocessing unit. • The neutronics properties are not sensitive to chemical data. • The reprocessing chemistry, from a pure numerical point of view, is an issue. - Abstract: Molten Salt Reactors (MSRs) are liquid-fuel reactors, in which the fuel is also the coolant and flows through the core. A particular configuration presented in this paper called the Molten Salt Fast Reactor consists in a Molten Salt Reactor with no moderator inside the core and a salt composition that leads to a fast neutron spectrum. Previous studies showed that this concept (previously called Thorium Molten Salt Reactor – Nonmoderated) has very promising characteristics. The liquid fuel implies a special reprocessing. Each day a small amount of the fuel salt is extracted from the core for on-site reprocessing. To study such a reactor, the materials evolution within the core has to be coupled to the reprocessing unit, since the latter cleans the salt quasi continuously and feeds the reactor. This paper details the issues associated to the numerical coupling of the core and the reprocessing. It presents how the chemistry is introduced inside the classical Bateman equation (evolution of nuclei within a neutron flux) in order to carry a numerical coupled study. To achieve this goal, the chemistry has to be modeled numerically and integrated to the equations of evolution. This paper presents how is it possible to describe the whole concept (reactor + reprocessing unit) by a system of equations that can be numerically solved. Our program is a connection between MCNP and a homemade evolution code called REM. Thanks to this tool; constraints on the fuel reprocessing were identified. Limits are specified to preserve the good neutronics properties of the MSFR. In this paper, we show that the limit

Phase locking of vortex cores in two coupled magnetic nanopillars

Directory of Open Access Journals (Sweden)

Qiyuan Zhu

2014-11-01

Full Text Available Phase locking dynamics of the coupled vortex cores in two identical magnetic spin valves induced by spin-polarized current are studied by means of micromagnetic simulations. Our results show that the available current range of phase locking can be expanded significantly by the use of constrained polarizer, and the vortices undergo large orbit motions outside the polarization areas. The effects of polarization areas and dipolar interaction on the phase locking dynamics are studied systematically. Phase locking parameters extracted from simulations are discussed by theoreticians. The dynamics of vortices influenced by spin valve geometry and vortex chirality are discussed at last. This work provides deeper insights into the dynamics of phase locking and the results are important for the design of spin-torque nano-oscillators.

Licensing procedures and safety criteria for core conversion in Japan

International Nuclear Information System (INIS)

Kanda, K.; Nakagome, Y.; Hayashi, M.

1983-01-01

In Japan, the establishment and operation of nuclear installations are governed mainly by the Law for Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors. This law lays down the regulations and conditions for licensing of the various installations involved in the nuclear fuel cycle, namely licensing of installations for refining, fabricating and reprocessing; and reactors, as well as licensing of the use of nuclear fuels in research facilities. Although procedures for the installations listed above vary depending on the installation concerned, only those relating to construction and operation of reactor facilities will be analysed in this study, as the conditions and principles applying to licensing and control of other installations are, to a large extent, similar to those concerning reactor facilities. The second part of this presentation describes the safety review of the KUCA reactor core conversion form HEU to MEU. For the safety review of the core conversion, the Committee on Examination of Reactor Safety of Japanese Government examined mainly the the nuclear characteristics and the integrity of aluminide fuel plates, which was very severe because we had no experience to use aluminide fuel plates in Japan. The integrity of fuel plates and the results of the worst accident analysis for the MEU core are shown with the comparison between the HEU and MEU cores. The significant difference was not observed between them. All the regulatory procedures were completed in September 1980. Fabrication of MEU fuel elements for the KUCA experiments by CERCA in France was started in September 1980, and will be completed in March 1981. The critical experiments in the KUCA with MEU fuel will be started on a single-core in May 1981 as a first step. Those on a coupled-core will follow

Little Earth Experiment: An instrument to model planetary cores.

Science.gov (United States)

Aujogue, Kélig; Pothérat, Alban; Bates, Ian; Debray, François; Sreenivasan, Binod

2016-08-01

In this paper, we present a new experimental facility, Little Earth Experiment, designed to study the hydrodynamics of liquid planetary cores. The main novelty of this apparatus is that a transparent electrically conducting electrolyte is subject to extremely high magnetic fields (up to 10 T) to produce electromagnetic effects comparable to those produced by moderate magnetic fields in planetary cores. This technique makes it possible to visualise for the first time the coupling between the principal forces in a convection-driven dynamo by means of Particle Image Velocimetry (PIV) in a geometry relevant to planets. We first present the technology that enables us to generate these forces and implement PIV in a high magnetic field environment. We then show that the magnetic field drastically changes the structure of convective plumes in a configuration relevant to the tangent cylinder region of the Earth's core.

A high-resolution detector based on liquid-core scintillating fibres with readout via an electron-bombarded charge-coupled device

International Nuclear Information System (INIS)

Cianfarani, C.; Duane, A.; Fabre, J.P.; Frenkel, A.; Golovkin, S.V.; Gorin, A.M.; Harrison, K.; Kozarenko, E.N.; Kushnirenko, A.E.; Ladygin, E.A.; Martellotti, G.; Medvedkov, A.M.; Nass, P.A.; Obudovski, V.P.; Penso, G.; Petukhov, Yu.P.; Siegmund, W.P.; Tyukov, V.E.; Vasilchenko, V.G.

1994-01-01

This paper is a presentation of results from tests in a 5 GeV/c hadron beam of detectors based on liquid-core scintillating fibres, each fibre consisting of a glass capillary filled with organic liquid scintillator. Fibre readout was performed via an Electron-Bombarded Charge-Coupled Device (EBCCD) image tube, a novel instrument that combines the functions of a high-gain, gated image intensifier and a Charge-Coupled Device. Using 1-methylnaphthalene doped with 3 g/l of R45 as liquid scintillator, the attenuation lengths obtained for light propagation over distances greater than 16 cm were 1.5 m in fibres of 20 μm core and 1.0 m in fibres of 16 μm core. For particles that crossed the fibres of 20 μm core at distances of ∼1.8 cm and ∼95 cm from the fibres' readout ends, the recorded hit densities were 5.3 mm -1 and 2.5 mm -1 respectively. Using 1-methylnaphthalene doped with 3.6 g/l of R39 as liquid scintillator and fibres of 75 μm core, the hit density obtained for particles that crossed the fibres at a distance of ∼1.8 cm from their readout ends was 8.5 mm -1 . With a specially designed bundle of tapered fibres, having core diameters that smoothly increase from 16 μm to 75 μm, a spatial precision of 6 μm was measured. (orig.)

On the coupling of fluid dynamics and electromagnetism at the top of the earth's core

Science.gov (United States)

Benton, E. R.

1985-01-01

A kinematic approach to short-term geomagnetism has recently been based upon pre-Maxwell frozen-flux electromagnetism. A complete dynamic theory requires coupling fluid dynamics to electromagnetism. A geophysically plausible simplifying assumption for the vertical vorticity balance, namely that the vertical Lorentz torque is negligible, is introduced and its consequences are developed. The simplified coupled magnetohydrodynamic system is shown to conserve a variety of magnetic and vorticity flux integrals. These provide constraints on eligible models for the geomagnetic main field, its secular variation, and the horizontal fluid motions at the top of the core, and so permit a number of tests of the underlying assumptions.

STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

Energy Technology Data Exchange (ETDEWEB)

Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

2016-05-01

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

Trichloroethylene (TCE) in tree cores to complement a subsurface investigation on residential property near a former electroplating facility.

Science.gov (United States)

Wilcox, Jeffrey D; Johnson, Kathy M

2016-10-01

Tree cores were collected and analyzed for trichloroethylene (TCE) on a private property between a former electroplating facility in Asheville, North Carolina (USA), and a contaminated wetland/spring complex. TCE was detected in 16 of 31 trees, the locations of which were largely consistent with a "plume core" delineated by a more detailed subsurface investigation nearly 2 years later. Concentrations in tree cores and nearby soil borings were not correlated, perhaps due to heterogeneities in both geologic and tree root structure, spatial and temporal variability in transpiration rates, or interferences caused by other contaminants at the site. Several tree cores without TCE provided evidence for significantly lower TCE concentrations in shallow groundwater along the margins of the contaminated spring complex in an area with limited accessibility. This study demonstrates that tree core analyses can complement a more extensive subsurface investigation, particularly in residential or ecologically sensitive areas.

Transport-diffusion coupling for Candu reactor core follow-Up

International Nuclear Information System (INIS)

Varin, E.; Marleau, G.; Chambon, R.

2003-01-01

We couple the finite reactor diffusion code DONJON and the lattice code DRAGON, called for simplicity DD, to perform reactor follow-up calculations using a history-based approach. In order to do this, a new DD module is developed. This module manages the transfer of information between standard DONJON and DRAGON data structures. Moreover, it stores in a history data structure the global and local parameters required for cell calculations as well as the isotopic composition of the various materials present in each cell of the reactor. We then implement in DD a parallel algorithm to perform history-based Candu reactor calculations. Here, we assign to each processor a specific number of fuel channels to be analyzed. The DRAGON cell calculations for each of the fuel bundles associated with the specified channels are performed on the same processor in order to minimize communication time. Only the macroscopic cross section libraries are exchanged between the processor. Since the amount of data exchanged is relatively small, we expect to obtain an ideal speed-up. The coupling is tested for the analysis of a simplified Candu reactor model with 4 x 4 channels each containing 4 bundles. A 100 full-power days core tracking sequence with 16 refueling steps is studied. Results are coherent with those obtained using more approximate approaches. Parallel speed-up is near optimal indicating that the use of this approach for more realistic reactor calculations should be pursued. (authors)

Partnership between CTSI and business schools can promote best practices for core facilities and resources.

Science.gov (United States)

Reeves, Lilith; Dunn-Jensen, Linda M; Baldwin, Timothy T; Tatikonda, Mohan V; Cornetta, Kenneth

2013-08-01

Biomedical research enterprises require a large number of core facilities and resources to supply the infrastructure necessary for translational research. Maintaining the financial viability and promoting efficiency in an academic environment can be particularly challenging for medical schools and universities. The Indiana Clinical and Translational Sciences Institute sought to improve core and service programs through a partnership with the Indiana University Kelley School of Business. The program paired teams of Masters of Business Administration students with cores and programs that self-identified the need for assistance in project management, financial management, marketing, or resource efficiency. The projects were developed by CTSI project managers and business school faculty using service-learning principles to ensure learning for students who also received course credit for their participation. With three years of experience, the program demonstrates a successful partnership that improves clinical research infrastructure by promoting business best practices and providing a valued learning experience for business students. © 2013 Wiley Periodicals, Inc.

Facile synthesis of hierarchical Co3O4@MnO2 core-shell arrays on Ni foam for asymmetric supercapacitors

Science.gov (United States)

Huang, Ming; Zhang, Yuxin; Li, Fei; Zhang, Lili; Wen, Zhiyu; Liu, Qing

2014-04-01

Hierarchical Co3O4@MnO2 core-shell arrays on Ni foam have been fabricated by a facile hydrothermal approach and further investigated as the electrode for high-performance supercapacitors. Owing to the high conductivity of the well-defined mesoporous Co3O4 nanowire arrays in combination with the large surface area provided by the ultrathin MnO2 nanosheets, the unique designed Co3O4@MnO2 core-shell arrays on Ni foam have exhibited a high specific capacitance (560 F g-1 at a current density of 0.2 A g-1), good rate capability, and excellent cycling stability (95% capacitance retention after 5000 cycles). An asymmetric supercapacitor with Co3O4@MnO2 core-shell nanostructure as the positive electrode and activated microwave exfoliated graphite oxide activated graphene (MEGO) as the negative electrode yielded an energy density of 17.7 Wh kg-1 and a maximum power density of 158 kW kg-1. The rational design of the unique core-shell array architectures demonstrated in this work provides a new and facile approach to fabricate high-performance electrode for supercapacitors.

A scaling study of the natural circulation flow of the ex-vessel core catcher cooling system of a 1400MW PWR for designing a scale-down test facility

International Nuclear Information System (INIS)

Rhee, Bo. W.; Ha, K. S.; Park, R. J.; Song, J. H.

2012-01-01

A scaling study on the steady state natural circulation flow along the flow path of the ex-vessel core catcher cooling system of 1400MWe PWR is described. The scaling criteria for reproducing the same thermalhydraulic characteristics of the natural circulation flow as the prototype core catcher cooling system in the scale-down test facility is derived and the resulting natural circulation flow characteristics of the prototype and scale-down facility analyzed and compared. The purpose of this study is to apply the similarity law to the prototype EU-APR1400 core catcher cooling system and the model test facility of this prototype system and derive a relationship between the heating channel characteristics and the down-comer piping characteristics so as to determine the down-comer pipe size and the orifice size of the model test facility. As the geometry and the heating wall heat flux of the heating channel of the model test facility will be the same as those of the prototype core catcher cooling system except the width of the heating channel is reduced, the axial distribution of the coolant quality (or void fraction) is expected to resemble each other between the prototype and model facility. Thus using this fact, the down-comer piping design characteristics of the model facility can be determined from the relationship derived from the similarity law

Facile Fabrication of a Hierarchical Superhydrophobic Coating with Aluminate Coupling Agent Modified Kaolin

Directory of Open Access Journals (Sweden)

Hui Li

2013-01-01

Full Text Available A superhydrophobic coating was fabricated from the dispersion of unmodified kaolin particles and aluminate coupling agent in anhydrous ethanol. Through surface modification, water contact angle of the coating prepared by modified kaolin particles increased dramatically from 0° to 152°, and the sliding angle decreased from 90° to 3°. Scanning electron microscopy was used to examine the surface morphology. A structure composed of micro-nano hierarchical component, combined with the surface modification by aluminate coupling agent which reduced the surface energy greatly, was found to be responsible for the superhydrophobicity. The method adopted is relatively simple, facile, and cost-effective and can potentially be applied to large water-repellent surface coatings.

Procedures for use of, and drill cores and cuttings available for study at, the Lithologic Core Storage Library, Idaho National Engineering Laboratory, Idaho

International Nuclear Information System (INIS)

Davis, L.C.; Hannula, S.R.; Bowers, B.

1997-03-01

In 1990, the US Geological Survey, in cooperation with the US Department of Energy, Idaho Operations Office, established the Lithologic Core Storage Library at the Idaho National Engineering Laboratory (INEL). The facility was established to consolidate, catalog, and permanently store nonradioactive drill cores and cuttings from investigations of the subsurface conducted at the INEL, and to provide a location for researchers to examine, sample, and test these materials. The facility is open by appointment to researchers for examination, sampling, and testing of cores and cuttings. This report describes the facility and cores and cuttings stored at the facility. Descriptions of cores and cuttings include the well names, well locations, and depth intervals available. Most cores and cuttings stored at the facility were drilled at or near the INEL, on the eastern Snake River Plain; however, two cores drilled on the western Snake River Plain are stored for comparative studies. Basalt, rhyolite, sedimentary interbeds, and surficial sediments compose the majority of cores and cuttings, most of which are continuous from land surface to their total depth. The deepest core stored at the facility was drilled to 5,000 feet below land surface. This report describes procedures and researchers' responsibilities for access to the facility, and examination, sampling, and return of materials

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

The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes

Energy Technology Data Exchange (ETDEWEB)

Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Jeltsov, Marti, E-mail: marti@safety.sci.kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Karbojian, Aram, E-mail: karbojan@kth.se; Villanueva, Walter, E-mail: walter@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

2015-08-15

Highlights: • Design of a heavy liquid thermal-hydraulic loop for CFD/STH code validation. • Description of the loop instrumentation and assessment of measurement error. • Experimental data from forced to natural circulation transient. - Abstract: Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, and (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper.

TRAC code assessment using data from SCTF Core-III, a large-scale 2D/3D facility

International Nuclear Information System (INIS)

Boyack, B.E.; Shire, P.R.; Harmony, S.C.; Rhee, G.

1988-01-01

Nine tests from the SCTF Core-III configuration have been analyzed using TRAC-PF1/MOD1. The objectives of these assessment activities were to obtain a better understanding of the phenomena occurring during the refill and reflood phases of a large-break loss-of-coolant accident, to determine the accuracy to which key parameters are calculated, and to identify deficiencies in key code correlations and models that provide closure for the differential equations defining thermal-hydraulic phenomena in pressurized water reactors. Overall, the agreement between calculated and measured values of peak cladding temperature is reasonable. In addition, TRAC adequately predicts many of the trends observed in both the integral effect and separate effect tests conducted in SCTF Core-III. The importance of assessment activities that consider potential contributors to discrepancies between the measured and calculated results arising from three sources are described as those related to (1) knowledge about the facility configuration and operation, (2) facility modeling for code input, and (3) deficiencies in code correlations and models. An example is provided. 8 refs., 7 figs., 2 tabs

A new coupling kernel for the three-dimensional simulation of a boiling water reactor core by the nodal coupling method

International Nuclear Information System (INIS)

Gupta, N.K.

1981-01-01

A new coupling kernel is developed for the three-dimensional (3-D) simulation of Boiling Water Reactors (BWR's) by the nodal coupling method. The new kernel depends not only on the properties of the node under consideration but also on the properties of its neighbouring nodes. This makes the kernel more useful in particular for fuel bundles lying in a surrounding of different nuclear characteristics, e.g. for a controlled bundle in the surrounding of uncontrolled bundles or vice-versa. The main parameter in the new kernel is a space-dependent factor obtained from the ratio of thermal-to-fast flux. The average value of the above ratio for each node is evaluated analytically. The kernel is incorporated in a 3-D BWR core simulation program MOGS. As an experimental verification of the model, the cycle-6 operations of the two units of the Tarapur Atomic Power Station (TAPS) are simulated and the result of the simulation are compared with Travelling Incore Probe (TIP) data. (orig.)

Prototype coupling of the CFD software ansys CFX with the 3D neutron kinetic core model DYN3D - 249

International Nuclear Information System (INIS)

Kliem, S.; Rohde, U.; Schutze, J.; Frank, Th.

2010-01-01

The CFD code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactor's coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for a small-size test problem confirm the correctness of the implementation of the prototype coupling. This test problem was a mini-core consisting of nine real-size fuel assemblies. Comparison was performed with the DYN3D standalone code. In the steady state, the effective multiplication factor obtained by the ANSYS CFX/DYN3D codes shows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. (authors)

Facile Synthesis of Yolk/Core-Shell Structured TS-1@Mesosilica Composites for Enhanced Hydroxylation of Phenol

KAUST Repository

Zou, Houbing

2015-12-14

© 2015 by the authors. In the current work, we developed a facile synthesis of yolk/core-shell structured TS-1@mesosilica composites and studied their catalytic performances in the hydroxylation of phenol with H2O2 as the oxidant. The core-shell TS-1@mesosilica composites were prepared via a uniform coating process, while the yolk-shell TS-1@mesosilica composite was prepared using a resorcinol-formaldehyde resin (RF) middle-layer as the sacrificial template. The obtained materials were characterized by X-ray diffraction (XRD), N2 sorption, Fourier transform infrared spectoscopy (FT-IR) UV-Visible spectroscopy, scanning electron microscopy (SEM) and transmission electron microscopy (TEM). The characterization results showed that these samples possessed highly uniform yolk/core-shell structures, high surface area (560–700 m2 g−1) and hierarchical pore structures from oriented mesochannels to zeolite micropores. Importantly, owing to their unique structural properties, these composites exhibited enhanced activity, and also selectivity in the phenol hydroxylation reaction.

Framework Application for Core Edge Transport Simulation (FACETS)

Energy Technology Data Exchange (ETDEWEB)

Krasheninnikov, Sergei; Pigarov, Alexander

2011-10-15

The FACETS (Framework Application for Core-Edge Transport Simulations) project of Scientific Discovery through Advanced Computing (SciDAC) Program was aimed at providing a high-fidelity whole-tokamak modeling for the U.S. magnetic fusion energy program and ITER through coupling separate components for each of the core region, edge region, and wall, with realistic plasma particles and power sources and turbulent transport simulation. The project also aimed at developing advanced numerical algorithms, efficient implicit coupling methods, and software tools utilizing the leadership class computing facilities under Advanced Scientific Computing Research (ASCR). The FACETS project was conducted by a multi-discipline, multi-institutional teams, the Lead PI was J.R. Cary (Tech-X Corp.). In the FACETS project, the Applied Plasma Theory Group at the MAE Department of UCSD developed the Wall and Plasma-Surface Interaction (WALLPSI) module, performed its validation against experimental data, and integrated it into the developed framework. WALLPSI is a one-dimensional, coarse grained, reaction/advection/diffusion code applied to each material boundary cell in the common modeling domain for a tokamak. It incorporates an advanced model for plasma particle transport and retention in the solid matter of plasma facing components, simulation of plasma heat power load handling, calculation of erosion/deposition, and simulation of synergistic effects in strong plasma-wall coupling.

ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

International Nuclear Information System (INIS)

Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien

2007-01-01

Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA R and concludes on customer benefits. ARCADIA R is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA R system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)

Initial design for an experimental investigation of strongly coupled plasma behavior in the ATLAS facility

CERN Document Server

Munson, C P; Taylor, A J; Trainor, R J; Wood, B P; Wysocki, F J

1999-01-01

Summary form only given. Atlas is a high current (~30 MA peak, with a current risetime ~4.5 mu sec), high energy (E/sub stored/=24 MJ, E /sub load/=3-6 MJ), pulsed power facility which is being constructed at Los Alamos National Laboratory with a scheduled completion date in the year 2000. When operational, this facility will provide a platform for experiments in high pressure shocks (>20 Mbar), adiabatic compression ( rho / rho /sub 0/>5, P>10 Mbar), high magnetic fields (~2000 T), high strain and strain rates ( epsilon >200, d epsilon /dt~10/sup 4/ to 10/sup 6/ s/sup -1/), hydrodynamic instabilities of materials in turbulent regimes, magnetized target fusion, equation of state, and strongly coupled plasmas. For the strongly coupled plasma experiments, an auxiliary capacitor bank will be used to generate a moderate density (<0.1 solid), relatively cold (~1 eV) plasma by ohmic heating of a conducting material of interest such as titanium. This target plasma will be compressed against a central column conta...

Neutronics analysis of the TRIGA Mark II reactor core and its experimental facilities

International Nuclear Information System (INIS)

Khan, R.

2010-01-01

core into a complete mixed core. To analyze the current core, a good knowledge of burned fuel material composition is essential. Because of the complications of experimental methods for measuring each FE, the ORIGEN2 computer code is selected for burn up and relevant material composition calculation. These calculations are verified by measuring the Cesium isotope (Cs-137) for six spent FE(s). Modifying the confirmed ORIGEN2 model for 104 and 110 (FLIP) FE(s), the burn up calculations of all 83 FE(s) of the current core are completed and applied to the already developed MCNP model. The detailed MCNP model of the burned core is verified by three local consistent experiments performed in June 2009. The criticality experiment confirms the model that the current core achieves its criticality on addition of 78th FE. The five FE(s) from different ring positions are measured to confirm the theoretical results. The percent deviation between MCNP predictions and experimental observations ranges from 3 to 19 %. The radial and axial neutron flux density distribution experiment verifies the MCNP theoretical results in the core. The theoretical and experimental perturbation study in the Central Irradiation Channel (CIR) of the core is performed. The reactivity effect of three small cylindrical samples (void, Cadmium and heavy water) are measured and compared with the MCNP predictions for verification. Applying the current core MCNP model, the void coefficient of reactivity is calculated as 11 cents per %-void. To perform the calculation in the experimental facilities outside the reactor core, the MCNP model is extended to the thermal column, radiographic collimator, four beam tubes and biological shielding. The MCNP results are verified in the thermal column and the beam tube A region. The percent difference between the simulated and experimental neutron diffusion length is 13 %. (author) [de

Distribution Coeficients (Kd) Generated From A Core Sample Collected From The Saltstone Disposal Facility

International Nuclear Information System (INIS)

Almond, P.; Kaplan, D.

2011-01-01

Core samples originating from Vault 4, Cell E of the Saltstone Disposal Facility (SDF) were collected in September of 2008 (Hansen and Crawford 2009, Smith 2008) and sent to SRNL to measure chemical and physical properties of the material including visual uniformity, mineralogy, microstructure, density, porosity, distribution coefficients (K d ), and chemical composition. Some data from these experiments have been reported (Cozzi and Duncan 2010). In this study, leaching experiments were conducted with a single core sample under conditions that are representative of saltstone performance. In separate experiments, reducing and oxidizing environments were targeted to obtain solubility and Kd values from the measurable species identified in the solid and aqueous leachate. This study was designed to provide insight into how readily species immobilized in saltstone will leach from the saltstone under oxidizing conditions simulating the edge of a saltstone monolith and under reducing conditions, targeting conditions within the saltstone monolith. Core samples were taken from saltstone poured in December of 2007 giving a cure time of nine months in the cell and a total of thirty months before leaching experiments began in June 2010. The saltstone from Vault 4, Cell E is comprised of blast furnace slag, class F fly ash, portland cement, and Deliquification, Dissolution, and Adjustment (DDA) Batch 2 salt solution. The salt solution was previously analyzed from a sample of Tank 50 salt solution and characterized in the 4QCY07 Waste Acceptance Criteria (WAC) report (Zeigler and Bibler 2009). Subsequent to Tank 50 analysis, additional solution was added to the tank solution from the Effluent Treatment Project as well as from inleakage from Tank 50 pump bearings (Cozzi and Duncan 2010). Core samples were taken from three locations and at three depths at each location using a two-inch diameter concrete coring bit (1-1, 1-2, 1-3; 2-1, 2-2, 2-3; 3-1, 3-2, 3-3) (Hansen and Crawford

DISTRIBUTION COEFICIENTS (KD) GENERATED FROM A CORE SAMPLE COLLECTED FROM THE SALTSTONE DISPOSAL FACILITY

Energy Technology Data Exchange (ETDEWEB)

Almond, P.; Kaplan, D.

2011-04-25

Core samples originating from Vault 4, Cell E of the Saltstone Disposal Facility (SDF) were collected in September of 2008 (Hansen and Crawford 2009, Smith 2008) and sent to SRNL to measure chemical and physical properties of the material including visual uniformity, mineralogy, microstructure, density, porosity, distribution coefficients (K{sub d}), and chemical composition. Some data from these experiments have been reported (Cozzi and Duncan 2010). In this study, leaching experiments were conducted with a single core sample under conditions that are representative of saltstone performance. In separate experiments, reducing and oxidizing environments were targeted to obtain solubility and Kd values from the measurable species identified in the solid and aqueous leachate. This study was designed to provide insight into how readily species immobilized in saltstone will leach from the saltstone under oxidizing conditions simulating the edge of a saltstone monolith and under reducing conditions, targeting conditions within the saltstone monolith. Core samples were taken from saltstone poured in December of 2007 giving a cure time of nine months in the cell and a total of thirty months before leaching experiments began in June 2010. The saltstone from Vault 4, Cell E is comprised of blast furnace slag, class F fly ash, portland cement, and Deliquification, Dissolution, and Adjustment (DDA) Batch 2 salt solution. The salt solution was previously analyzed from a sample of Tank 50 salt solution and characterized in the 4QCY07 Waste Acceptance Criteria (WAC) report (Zeigler and Bibler 2009). Subsequent to Tank 50 analysis, additional solution was added to the tank solution from the Effluent Treatment Project as well as from inleakage from Tank 50 pump bearings (Cozzi and Duncan 2010). Core samples were taken from three locations and at three depths at each location using a two-inch diameter concrete coring bit (1-1, 1-2, 1-3; 2-1, 2-2, 2-3; 3-1, 3-2, 3-3) (Hansen and

Development of a fiber-coupled laser-induced breakdown spectroscopy instrument for analysis of underwater debris in a nuclear reactor core

International Nuclear Information System (INIS)

Saeki, Morihisa; Iwanade, Akio; Ohba, Hironori; Ito, Chikara; Wakaida, Ikuo; Thornton, Blair; Sakka, Tetsuo

2014-01-01

To inspect the post-accident nuclear core reactor of the TEPCO Fukushima Daiichi nuclear power plant (F1-NPP), a transportable fiber-coupled laser-induced breakdown spectroscopy (LIBS) instrument has been developed. The developed LIBS instrument was designed to analyze underwater samples in a high-radiation field by single-pulse breakdown with gas flow or double-pulse breakdown. To check the feasibility of the assembled fiber-coupled LIBS instrument for the analysis of debris material (mixture of the fuel core, fuel cladding, construction material and so on) in the F1-NPP, we investigated the influence of the radiation dose on the optical transmittance of the laser delivery fiber, compared data quality among various LIBS techniques for an underwater sample and studied the feasibility of the fiber-coupled LIBS system in an analysis of the underwater sample of the simulated debris in F1-NPP. In a feasible study conducted by using simulated debris, which was a mixture of CeO 2 (surrogate of UO 2 ), ZrO 2 and Fe, we selected atomic lines suitable for the analysis of materials, and prepared calibration curves for the component elements. The feasible study has guaranteed that the developed fiber-coupled LIBS system is applicable for analyzing the debris materials in the F1-NPP. (author)

Boundary perturbations coupled to core 3/2 tearing modes on the DIII-D tokamak

International Nuclear Information System (INIS)

Tobias, B; Yu, L; Domier, C W; Luhmann, N C Jr; Austin, M E; Paz-Soldan, C; Turnbull, A D; Classen, I G J

2013-01-01

High confinement (H-mode) discharges on the DIII-D tokamak are routinely subject to the formation of long-lived, non-disruptive magnetic islands that degrade confinement and limit fusion performance. Simultaneous, 2D measurement of electron temperature fluctuations in the core and edge regions allows for reconstruction of the radially resolved poloidal mode number spectrum and phase of the global plasma response associated with these modes. Coherent, n = 2 excursions of the plasma boundary are found to be the result of coupling to an n = 2, kink-like mode which arises locked in phase to the 3/2 island chain. This coupling dictates the relative phase of the displacement at the boundary with respect to the tearing mode. This unambiguous phase relationship, for which no counter-examples are observed, is presented as a test for modeling of the perturbed fields to be expected outside the confined plasma. (paper)

Investigation of reflood models by coupling REFLA-1D and multi-loop system model

International Nuclear Information System (INIS)

Sugimoto, Jun; Murao, Yoshio

1983-09-01

A system analysis code REFLA-1DS was developed by coupling reflood analysis code REFLA-1D and a multi-loop primary system model. The reflood models in the code were investigated for the development of the integral system analysis code. The REFLA-1D, which was developed with the small scale reflood experiment at JAERI, consists of one-dimensional core model and a primary system model with a constant loop resistance. The multi-loop primary system model was developed with the Cylindrical Core Test Facility of JAERI's large scale reflood tests. The components modeled in the code are the upper plenum, the steam generator, the coolant pump, the ECC injection port, the downcomer and the broken cold leg nozzle. The coupling between the two models in REFLA-1DS is accomplished by applying the equivalent flow resistance calculated with the multiloop model to the REFLA-1D. The characteristics of the code is its simplicity of the system model and the solution method which enables the fast running and the easy reflood analysis for the further model development. A fairly good agreement was obtained with the results of the Cylindrical Core Test Facility for the calculated water levels in the downcomer, the core and the upper plenum. A qualitatively good agreement was obtained concerning the parametric effects of the system pressure, the ECC flow rate and the initial clad temperature. Needs for further code improvements of the models, however, were pointed out. These include the problem concerning the generation rate of the steam and water droplets in the core in an early period, the effect of the flow oscillation on the core cooling, the heat release from the downcomer wall, and the stable system calculation. (author)

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

International Nuclear Information System (INIS)

Okubo, Tsutomu; Murao, Yoshio

1984-03-01

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

Coupled neutronics and thermal hydraulics of high density cores for FRM II

Energy Technology Data Exchange (ETDEWEB)

Breitkreutz, Harald

2011-03-04

According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X{sup 2}', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50

Coupled neutronics and thermal hydraulics of high density cores for FRM II

International Nuclear Information System (INIS)

Breitkreutz, Harald

2011-01-01

According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X 2 ', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50% enriched disperse UMo core with

A sensitive DNA biosensor based on a facile sulfamide coupling reaction for capture probe immobilization

International Nuclear Information System (INIS)

Wang, Qingxiang; Ding, Yingtao; Gao, Feng; Jiang, Shulian; Zhang, Bin; Ni, Jiancong; Gao, Fei

2013-01-01

Graphical abstract: A novel DNA biosensor was fabricated through a facile sulfamide coupling reaction between probe DNA and the sulfonic dye of 1-amino-2-naphthol-4-sulfonic acid that electrodeposited on a glassy carbon electrode. -- Highlights: •A versatile sulfonic dye of ANS was electrodeposited on a GCE. •A DNA biosensor was fabricated based on a facile sulfamide coupling reaction. •High probe DNA density of 3.18 × 10 13 strands cm −2 was determined. •A wide linear range and a low detection limit were obtained. -- Abstract: A novel DNA biosensor was fabricated through a facile sulfamide coupling reaction. First, the versatile sulfonic dye molecule of 1-amino-2-naphthol-4-sulfonate (AN-SO 3 − ) was electrodeposited on the surface of a glassy carbon electrode (GCE) to form a steady and ordered AN-SO 3 − layer. Then the amino-terminated capture probe was covalently grafted to the surface of SO 3 − -AN deposited GCE through the sulfamide coupling reaction between the amino groups in the probe DNA and the sulfonic groups in the AN-SO 3 − . The step-by-step modification process was characterized by electrochemistry and attenuated total reflectance Fourier transform infrared (ATR-FTIR) spectroscopy. Using Ru(NH 3 ) 6 3+ as probe, the probe density and the hybridization efficiency of the biosensor were determined to be 3.18 × 10 13 strands cm −2 and 86.5%, respectively. The hybridization performance of the biosensor was examined by differential pulse voltammetry using Co(phen) 3 3+/2+ (phen = 1,10-phenanthroline) as the indicator. The selectivity experiments showed that the biosensor presented distinguishable response after hybridization with the three-base mismatched, non-complementary and complementary sequences. Under the optimal conditions, the oxidation peak currents of Co(phen) 3 3+/2+ increased linearly with the logarithm values of the concentration of the complementary sequences in the range from 1.0 × 10 −13 M to 1.0 × 10 −8 M with

Development of inherent core technologies for advanced reactor

International Nuclear Information System (INIS)

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

1999-03-01

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

Development of inherent core technologies for advanced reactor

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

Facile synthesis of silver immobilized-poly(methyl methacrylate)/polyethyleneimine core-shell particle composites

Energy Technology Data Exchange (ETDEWEB)

Jenjob, Somkieath [Department of Chemistry, Faculty of Science, Mahidol University, 999 Phuttamonthon 4 Road, Salaya, Nakhon Pathom 73170 (Thailand); Center of Excellence for Innovation in Chemistry (PERCH-CIC), Department of Chemistry, Faculty of Science, Mahidol University, Rama 6 Road, Ratchathewi, Bangkok 10400 (Thailand); Tharawut, Teeralak [Department of Chemistry, Faculty of Science, Mahidol University, 999 Phuttamonthon 4 Road, Salaya, Nakhon Pathom 73170 (Thailand); Sunintaboon, Panya, E-mail: panya.sun@mahidol.ac.th [Department of Chemistry, Faculty of Science, Mahidol University, 999 Phuttamonthon 4 Road, Salaya, Nakhon Pathom 73170 (Thailand); Center of Excellence for Innovation in Chemistry (PERCH-CIC), Department of Chemistry, Faculty of Science, Mahidol University, Rama 6 Road, Ratchathewi, Bangkok 10400 (Thailand); Center for Alternative Energy, Faculty of Science, Mahidol University, 999 Phuttamonthon 4 Road, Salaya, Nakhon Pathom 73170 (Thailand)

2012-10-01

A facile route to synthesize silver-embedded-poly(methyl methacrylate)/polyethyleneimine (PMMA/PEI-Ag) core-shell particle composites was illustrated in this present work. PMMA/PEI core-shell particle templates were first prepared by a surfactant-free emulsion polymerization. PEI on the templates' surface was further used to complex and reduce Ag{sup +} ions (from silver nitrate solution) to silver nanoparticles (AgNPs) at ambient temperature, resulting in the PMMA/PEI-Ag particle composites. The formation of AgNPs was affected by the pHs of the reaction medium. The pH of reaction medium at 6.5 was optimal for the formation of PMMA/PEI-Ag with good colloidal stability, which was confirmed by size and size distribution, FTIR spectroscopy, UV-vis spectroscopy and X-ray diffraction. Moreover, the amount of AgNO{sub 3} solution (4.17-12.50 g) was found to affect the formation of AgNPs. Transmission electron microscopy (TEM) indicated that the AgNPs were incorporated in the PMMA/PEI core-shell matrix, and had 6-10 nm in diameter. AgNPs immobilized on PMMA/PEI core-shell particles were also investigated by energy dispersive X-ray spectroscopy analysis mode extended from scanning electron microscopy (SEM/EDS). Furthermore, the presence of AgNPs was found to influence the thermal degradation behavior of PMMA/PEI particle composites as observed through thermogravimetric analysis (TGA). Highlights: Black-Right-Pointing-Pointer A 2-step synthesis of Ag immobilized-PMMA/PEI particle composites was shown. Black-Right-Pointing-Pointer PMMA/PEI core-shell templates were first formed and PEI assisted AgNP formation. Black-Right-Pointing-Pointer Formation of PMMA/PEI-Ag was affected by pH of medium and amount of AgNO{sub 3}. Black-Right-Pointing-Pointer PMMA/PEI-Ag can be confirmed by color change, UV-vis, TEM, SEM with EDS, and X-ray. Black-Right-Pointing-Pointer Effect of AgNPs on thermal degradation of PMMA/PEI-Ag can be observed through TGA.

Experimental subcritical facility driven by D-D/D-T neutron generator at BARC, India

Energy Technology Data Exchange (ETDEWEB)

Sinha, Amar, E-mail: image@barc.gov.in; Roy, Tushar; Kashyap, Yogesh; Ray, Nirmal; Shukla, Mayank; Patel, Tarun; Bajpai, Shefali; Sarkar, P.S.; Bishnoi, Saroj

2015-05-01

Highlights: •Experimental subcritical facility BRAHMMA coupled to D-D/D-T neutron generator. •Preliminary results of PNS experiments reported. •Feynman-alpha noise measurements explored with continuous source. -- Abstract: The paper presents design of an experimental subcritical assembly driven by D-D/D-T neutron and preliminary experimental measurements. The system has been developed for investigating the static and dynamic neutronic properties of accelerator driven sub-critical systems. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The fuel is embedded in high density polyethylene moderator matrix. Estimated k{sub eff} of the system is ∼0.89. One of the unique features of subcritical core is the use of Beryllium oxide (BeO) as reflector and HDPE as moderator making the assembly a compact modular system. The subcritical core is coupled to Purnima Neutron Generator which works in D-D and D-T mode with both DC and pulsed operation. It has facility for online source strength monitoring using neutron tagging and programmable source modulation. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques with D-D neutrons. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor k{sub s} and external neutron source efficiency φ{sup ∗} in great details. Experiments with D-T neutrons are also underway.

Coupling of 3-D core computational codes and a reactor simulation software for the computation of PWR reactivity accidents induced by thermal-hydraulic transients

International Nuclear Information System (INIS)

Raymond, P.; Caruge, D.; Paik, H.J.

1994-01-01

The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs

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

Thermal and Mechanical Performance of the First MICE Coupling Coil and the Fermilab Solenoid Test Facility

Energy Technology Data Exchange (ETDEWEB)

Rabehl, Roger [Fermilab; Carcagno, Ruben [Fermilab; Caspi, Shlomo [LBNL, Berkeley; DeMello, Allan [LBNL, Berkeley; Kokoska, Lidija [Fermilab; Orris, D. [Fermilab; Pan, Heng [LBNL, Berkeley; Sylvester, Cosmore [Fermilab; Tartaglia, Michael

2014-11-06

The first coupling coil for the Muon Ionization Cooling Experiment (MICE) has been tested in a conduction-cooled environment at the Solenoid Test Facility at Fermilab. An overview of the thermal and mechanical performance of the magnet and the test stand during cool-down and power testing of the magnet is presented.

Experimental identification of nonlinear coupling between (intermediate, small)-scale microturbulence and an MHD mode in the core of a superconducting tokamak

Science.gov (United States)

Sun, P. J.; Li, Y. D.; Ren, Y.; Zhang, X. D.; Wu, G. J.; Xu, L. Q.; Chen, R.; Li, Q.; Zhao, H. L.; Zhang, J. Z.; Shi, T. H.; Wang, Y. M.; Lyu, B.; Hu, L. Q.; Li, J.; The EAST Team

2018-01-01

In this paper, we present clear experimental evidence of core region nonlinear coupling between (intermediate, small)-scale microturbulence and an magnetohydrodynamics (MHD) mode during the current ramp-down phase in a set of L-mode plasma discharges in the experimental advanced superconducting tokamak (EAST, Wan et al (2006 Plasma Sci. Technol. 8 253)). Density fluctuations of broadband microturbulence (k\\perpρi˜2{-}5.2 ) and the MHD mode (toroidal mode number m = -1 , poloidal mode number n = 1 ) are measured simultaneously, using a four-channel tangential CO2 laser collective scattering diagnostic in core plasmas. The nonlinear coupling between the broadband microturbulence and the MHD mode is directly demonstrated by showing a statistically significant bicoherence and modulation of turbulent density fluctuation amplitude by the MHD mode.

Comparison of TRAC-PF1/MOD1 to a no-failure UPI test in the Cylindrical Core Test Facility

International Nuclear Information System (INIS)

Cappiello, M.; Spore, J.

1986-01-01

TRAC-PF1/MOD1 is compared to a no-failure upper plenum injection reflood test in the Cylindrical Core Test Facility. The results show that TRAC can accurately predict the asymmetric channeling of fluid from upper plenum into the core and that a multidimensional modeling capability is required to do so. The rod temperature behavior is accurately predicted for both the peak cladding temperature and the quench time in the high- and low-power zones. Excessive downflow of liquid at the tie plate is predicted as a result of the interfacial drag model used in TRAC. 10 figs

Simulation of an MSLB scenario using the 3D neutron kinetic core model DYN3D coupled with the CFD software Trio-U

Energy Technology Data Exchange (ETDEWEB)

Grahn, Alexander, E-mail: a.grahn@hzdr.de; Gommlich, André; Kliem, Sören; Bilodid, Yurii; Kozmenkov, Yaroslav

2017-04-15

Highlights: • Improved thermal-hydraulic description of nuclear reactor cores. • Providing reactor dynamics code with realistic thermal-hydraulic boundary conditions. • Possibility of three-dimensional flow phenomena in the core, such as cross flow, flow reversal. • Simulation at higher spatial resolution as compared to system codes. - Abstract: In the framework of the European project NURESAFE, the reactor dynamics code DYN3D, developed at Helmholtz-Zentrum Dresden-Rossendorf (HZDR), was coupled with the Computational Fluid Dynamics (CFD) solver Trio-U, developed at CEA France, in order to replace DYN3D’s one-dimensional hydraulic part with a full three-dimensional description of the coolant flow in the reactor core at higher spatial resolution. The present document gives an introduction into the coupling method and shows results of its application to the simulation of a Main Steamline Break (MSLB) accident of a Pressurised Water Reactor (PWR).

Weakly and strongly coupled Belousov-Zhabotinsky patterns

Science.gov (United States)

Weiss, Stephan; Deegan, Robert D.

2017-02-01

We investigate experimentally and numerically the synchronization of two-dimensional spiral wave patterns in the Belousov-Zhabotinsky reaction due to point-to-point coupling of two separate domains. Different synchronization modalities appear depending on the coupling strength and the initial patterns in each domain. The behavior as a function of the coupling strength falls into two qualitatively different regimes. The weakly coupled regime is characterized by inter-domain interactions that distorted but do not break wave fronts. Under weak coupling, spiral cores are pushed around by wave fronts in the other domain, resulting in an effective interaction between cores in opposite domains. In the case where each domain initially contains a single spiral, the cores form a bound pair and orbit each other at quantized distances. When the starting patterns consist of multiple randomly positioned spiral cores, the number of cores decreases with time until all that remains are a few cores that are synchronized with a partner in the other domain. The strongly coupled regime is characterized by interdomain interactions that break wave fronts. As a result, the wave patterns in both domains become identical.

PUMA Version 6 Multiplatform with Facilities to be coupled with other Simulation Models

International Nuclear Information System (INIS)

Grant, Carlos

2013-01-01

PUMA is a code for nuclear reactor calculation used in all nuclear installations in Argentina for simulation of fuel management, power cycles and transient events by means of spatial kinetic diffusion theory in 3D. For the versions used up to now the WINDOWS platform was used with very good results. Nowadays PUMA must work in different operative systems, LINUX among others, and must also have facilities to be coupled with other models. For this reason this new version was reprogrammed in ADA, language oriented to a safe programming and be found in any operative system. In former versions PUMA was executed through macro instructions written in LOGO. For this version it is possible to use also PYTHON, which makes also possible the access in execution time to internal data of PUMA. The use of PYTHON allows a easy way to couple PUMA with other codes. The possibilities of this new version of PUMA are shown by means of examples of input data and process control using PYTHON and LOGO. It is discussed the implementation of this methodology in other codes to be coupled with PUMA for versions run in WINDOWS and LINUX. (author)

Simulation of a hypothetical core disruptive accident in the mars test-facility

International Nuclear Information System (INIS)

Robbe, M.F.; Lepareux, M.

2001-01-01

In France, a large experimental programme MARA/MARS was undertaken in the 80's to estimate the mechanical consequences of an HCDA (Hypothetical Core Disruptive Accident) and to validate the SIRIUS computer code used at that time for the numerical simulations. At the end of the 80's, it was preferred to add a HCDA sodium-bubble-argon tri-component constitutive law to the general ALE fast dynamics finite element CASTEM-PLEXUS code rather than going on developing and using the specialized SIRIUS code. The experimental results of the MARA programme were used in the 90's to validate and qualify the CASTEM-PLEXUS code. A first series of computations of the tests MARA 8, MARA 10 and MARS was realised. The simulations showed a rather good agreement between the experimental and computed results for the MARA 8 and MARA 10 tests - even if there were some discrepancies - but the prediction of the MARS structure displacements and strains was overestimated. This conservatism was supposed to come from the fact that several MARS non axisymmetric structures like core elements, pumps and heat exchangers were not represented in the CASTEM-PLEXUS model. These structures, acting as porous barriers, had a protective effect on the mock-up containment by absorbing energy and slowing down the fluid impacting the containment. For these reasons, we developed in CASTEM-PLEXUS a new HCDA constitutive law taking into account the presence of the internal structures (without meshing them) by means of an equivalent porosity method. In other respects, the process used for dealing with the fluid-structure coupling in CASTEM-PLEXUS was improved. Thus a second series of simulations of the tests MARA8 and MARA10 was realised. A simulation of the test MARS was carried out too with the same simplified representation of the peripheral structures as in order to estimate the improvement provided by the new fluid-structure coupling. This paper presents a third numerical simulation of the MARS test with the

Platinum-nanoparticle-supported core-shell polymer nanospheres with unexpected water stability and facile further modification

Science.gov (United States)

Yuan, Conghui; Xu, Yiting; Luo, Weiang; Zeng, Birong; Qiu, Wuhui; Liu, Jie; Huang, Huiling; Dai, Lizong

2012-05-01

Core-shell nanospheres (CSNSs) with hydrophobic cores and hydrophilic shells were fabricated via a simple mini-emulsion polymerization for the stabilization of platinum nanoparticles (Pt-NPs). The CSNSs showed extremely high loading capacity of Pt-NPs (the largest loading amount of the Pt-NPs was about 49.2 wt%). Importantly, the Pt-NPs/CSNSs nanocomposites had unexpected stability in aqueous solution. DLS results revealed that the CSNSs loaded with Pt-NPs exhibited almost no aggregation after standing for a long time . However, the Pt-NPs immobilized on the CSNSs were not straitlaced: they could transport and redistribute between CSNSs freely when the environmental temperature was higher than the melting point of the CSNS shell. Owing to their excellent stability in aqueous solution, the surface of the Pt-NPs/CSNSs nanocomposites could be further decorated easily. For example, polyaniline (PANI)-coated Pt-NPs/CSNSs, nickel (Ni)-coated Pt-NPs/CSNSs and PANI/Pt-NPs dual-layer hollow nanospheres were facilely fabricated from the Pt-NPs/CSNS nanocomposites.

French experimental facilities for measurements of transverse flows and assessment of the corresponding risk of vibrations in heterogeneous cores

International Nuclear Information System (INIS)

Le Borgne, E.; Mattei, A.; Oceraies, Y.; Fardeau, P.

1994-01-01

Due to insertion of a limited number of new assemblies at each cycle, the cores in Pressurized Water Reactors are not homogeneous. Referring only to the impact on coolant flow, these differences can range from variable hydraulic resistances in the assembly, which depend on the geometric changes occurring during preceding cycles, to coexistence of assemblies with new design structures. Deviations in resistance between neighboring fuel assemblies causes the flow rates to be distributed differently between the assembly rods. This results in development of transverse flows from the main axial flow, and changes in the axial velocity gradients. These particularities of coolant flow have an effect on both vibration levels and cooling of the fuel rods, and also on the axial forces exerted on the assemblies in the core cavity. Since 1985, French Atomic Energy Commission (CEA) has gradually acquired experimental and measuring facilities that have allowed it to engage in research and development programs in these areas, in cooperation with industry partners in the nuclear field. Two complementary test loops have been constructed, called ARIANE and HERMES T. Use of these experimental facilities allows to obtain complete and detailed information on the hydraulic and vibratory phenomena specific to heterogeneous cores. In particular it is possible to establish a direct assessment of the actual compatibility between two different assemblies. By making a few specific changes, these facilities can also be used as a unique tool for assembly behaviour studies under seismic conditions with simulation of the flow effects. Also, a source of information in thus made available for qualification of computation codes for vibratory mechanics and multidimensional fluid mechanics under development at CEA and also used in the field of nuclear fuel. (authors). 6 figs., 1 ref

Initial design for an experimental investigation of strongly coupled plasma behavior in the Atlas facility

Energy Technology Data Exchange (ETDEWEB)

Munson, C.P.; Benage, J.F. Jr.; Taylor, A.J.; Trainor, R.J. Jr.; Wood, B.P.; Wysocki, F.J.

1999-07-01

Atlas is a high current ({approximately} 30 MA peak, with a current risetime {approximately} 4.5 {micro}sec), high energy (E{sub stored} = 24 MJ, E{sub load} = 3--6 MJ), pulsed power facility which is being constructed at Los Alamos National Laboratory with a scheduled completion date in the year 2000. When operational, this facility will provide a platform for experiments in high pressure shocks (> 20 Mbar), adiabatic compression ({rho}/{rho}{sub 0} > 5, P > 10 Mbar), high magnetic fields ({approximately} 2,000 T), high strain and strain rates ({var_epsilon} > 200%, d{var_epsilon}/dt {approximately} 10{sup 4} to 10{sup 6} s{sup {minus}1}), hydrodynamic instabilities of materials in turbulent regimes, magnetized target fusion, equation of state, and strongly coupled plasmas. For the strongly coupled plasma experiments, an auxiliary capacitor bank will be used to generate a moderate density (< 0.1 solid), relatively cold ({approximately} 1 eV) plasma by ohmic heating of a conducting material of interest such as titanium. This stargate plasma will be compressed against a central column containing diagnostic instrumentation by a cylindrical conducting liner that is driven radially inward by current from the main Atlas capacitor bank. The plasma is predicted to reach densities of {approximately} 1.1 times solid, achieve ion and electron temperatures of {approximately} 10 eV, and pressures of {approximately} 4--5 Mbar. This is a density/temperature regime which is expected to experience strong coupling, but only partial degeneracy. X-ray radiography is planned for measurements of the material density at discrete times during the experiments; diamond Raman measurements are anticipated for determination of the pressure. In addition, a neutron resonance spectroscopic technique is being evaluated for possible determination of the temperature (through low percentage doping of the titanium with a suitable resonant material). Initial target plasma formation experiments are

Initial design for an experimental investigation of strongly coupled plasma behavior in the Atlas facility

International Nuclear Information System (INIS)

Munson, C.P.; Benage, J.F. Jr.; Taylor, A.J.; Trainor, R.J. Jr.; Wood, B.P.; Wysocki, F.J.

1999-01-01

Atlas is a high current (approximately 30 MA peak, with a current risetime approximately 4.5 microsec), high energy (E stored = 24 MJ, E load = 3--6 MJ), pulsed power facility which is being constructed at Los Alamos National Laboratory with a scheduled completion date in the year 2000. When operational, this facility will provide a platform for experiments in high pressure shocks (> 20 Mbar), adiabatic compression (ρ/ρ 0 > 5, P > 10 Mbar), high magnetic fields (approximately 2,000 T), high strain and strain rates (var e psilon > 200%, dvar e psilon/dt approximately 10 4 to 10 6 s -1 ), hydrodynamic instabilities of materials in turbulent regimes, magnetized target fusion, equation of state, and strongly coupled plasmas. For the strongly coupled plasma experiments, an auxiliary capacitor bank will be used to generate a moderate density (< 0.1 solid), relatively cold (approximately 1 eV) plasma by ohmic heating of a conducting material of interest such as titanium. This stargate plasma will be compressed against a central column containing diagnostic instrumentation by a cylindrical conducting liner that is driven radially inward by current from the main Atlas capacitor bank. The plasma is predicted to reach densities of approximately 1.1 times solid, achieve ion and electron temperatures of approximately 10 eV, and pressures of approximately 4--5 Mbar. This is a density/temperature regime which is expected to experience strong coupling, but only partial degeneracy. X-ray radiography is planned for measurements of the material density at discrete times during the experiments; diamond Raman measurements are anticipated for determination of the pressure. In addition, a neutron resonance spectroscopic technique is being evaluated for possible determination of the temperature (through low percentage doping of the titanium with a suitable resonant material). Initial target plasma formation experiments are being planned on an existing pulsed power facility at LANL and

Facile in situ synthesis of wurtzite ZnS/ZnO core/shell heterostructure with highly efficient visible-light photocatalytic activity and photostability

Science.gov (United States)

Xiao, Jian-Hua; Huang, Wei-Qing; Hu, Yong-sheng; Zeng, Fan; Huang, Qin-Yi; Zhou, Bing-Xin; Pan, Anlian; Li, Kai; Huang, Gui-Fang

2018-02-01

High photocatalytic activity and photostability are the pursuit of the goal for designing promising photocatalysts. Herein, using ZnO to encapsulate ZnS nanoparticles is proposed as an effective strategy to enhance photocatalytic activity and anti-photocorrosion. The ZnS/ZnO core/shell heterostructures are obtained via an annealing treatment of ZnS nanoparticles produced by a facile wet chemical approach. Due to its small size, the nascent cubic sphalerite ZnS (s-ZnS) converts into a hexagonal wurtzite ZnS (w-ZnS)/ZnO core/shell structure after annealing treatment. In situ oxidation leads to increasing ZnO, simultaneously decreasing the w-ZnS content in the resultant w-ZnS/ZnO with thermal annealing time. The w-ZnS/ZnO core/shell heterostructures show high photocatalytic activity, demonstrated by the photodegradation rate of methylene blue being up to ten-fold and seven-fold higher than that of s-ZnS under UV and visible light irradiation, respectively, and the high capability of degrading rhodamine B. The enhanced photocatalytic activity may be attributed to the large specific surface and improved charge carrier separation at the core/shell interface. Moreover, it displays high photostability owing to the protection of the ZnO shell, greatly inhibiting the photocorrosion of ZnS. This facile in situ oxidation is effective and easily scalable, providing opportunities for developing novel core/shell structure photocatalysts with high activity and photostability.

Transient simulation of an endothermic chemical process facility coupled to a high temperature reactor: Model development and validation

International Nuclear Information System (INIS)

Brown, Nicholas R.; Seker, Volkan; Revankar, Shripad T.; Downar, Thomas J.

2012-01-01

Highlights: ► Models for PBMR and thermochemical sulfur cycle based hydrogen plant are developed. ► Models are validated against available data in literature. ► Transient in coupled reactor and hydrogen plant system is studied. ► For loss-of-heat sink accident, temperature feedback within the reactor core enables shut down of the reactor. - Abstract: A high temperature reactor (HTR) is a candidate to drive high temperature water-splitting using process heat. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Three thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. The models and coupling scheme are presented here, as well as a transient test case initiated within the chemical plant. The 50% feed flow failure within the chemical plant results in a slow loss-of-heat sink (LOHS) accident in the nuclear reactor. Due to the temperature feedback within the reactor core the nuclear reactor partially shuts down over 1500 s. Two distinct regions are identified within the coupled plant response: (1) immediate LOHS due to the loss of the sulfuric

Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

Directory of Open Access Journals (Sweden)

Po Hu

2014-01-01

Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

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

International Nuclear Information System (INIS)

Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

2001-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

Coupling of the core simulator DYN3D with the CFD solver TrioU and its application to a Main Steamline Break scenario

Energy Technology Data Exchange (ETDEWEB)

Grahn, Alexander; Gommlich, Andre; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

2017-06-01

In the framework of the European project NURESAFE, the reactor dynamics code DYN3D developed at HZDR was coupled with the CFD solver TrioU from CEA France. This coupling was used to simulate the coolant mixing in the reactor pressure vessel and in the core during a Main Steamline Break (MSLB) accident and to study its effect on the reactor power.

Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

Science.gov (United States)

Kazeminezhad, F.; Anghaie, S.

2008-01-01

Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

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)

Advanced light microscopy core facilities: Balancing service, science and career

Science.gov (United States)

Hartmann, Hella; Reymann, Jürgen; Ansari, Nariman; Utz, Nadine; Fried, Hans‐Ulrich; Kukat, Christian; Peychl, Jan; Liebig, Christian; Terjung, Stefan; Laketa, Vibor; Sporbert, Anje; Weidtkamp‐Peters, Stefanie; Schauss, Astrid; Zuschratter, Werner; Avilov, Sergiy

2016-01-01

ABSTRACT Core Facilities (CF) for advanced light microscopy (ALM) have become indispensable support units for research in the life sciences. Their organizational structure and technical characteristics are quite diverse, although the tasks they pursue and the services they offer are similar. Therefore, throughout Europe, scientists from ALM‐CFs are forming networks to promote interactions and discuss best practice models. Here, we present recommendations for ALM‐CF operations elaborated by the workgroups of the German network of ALM‐CFs, German Bio‐Imaging (GerBI). We address technical aspects of CF planning and instrument maintainance, give advice on the organization and management of an ALM‐CF, propose a scheme for the training of CF users, and provide an overview of current resources for image processing and analysis. Further, we elaborate on the new challenges and opportunities for professional development and careers created by CFs. While some information specifically refers to the German academic system, most of the content of this article is of general interest for CFs in the life sciences. Microsc. Res. Tech. 79:463–479, 2016. © 2016 THE AUTHORS MICROSCOPY RESEARCH AND TECHNIQUE PUBLISHED BY WILEY PERIODICALS, INC. PMID:27040755

Radiological Control Technician: Phase 4, Facility practical training attachment

International Nuclear Information System (INIS)

1992-10-01

At DOE sites with more than one facility, and where RCT tasks at each facility may differ, site and facility tasks should be separated. The tasks that are common to all the facilities on the site should be included in Phase II training with the core tasks. Tasks unique to a facility should be added to the training program qualification standard, as an attachment, as Phase IV training. Not all the DOE sites will include Phase IV training in their programs. Phase IV training allows each site to qualify technicians to a select facility. Since the core training for the technicians is standardized, the transfer of technicians between facilities requires that only facility tasks be taught, provided the core qualification is current

Three-dimensional coupled Monte Carlo-discrete ordinates computational scheme for shielding calculations of large and complex nuclear facilities

International Nuclear Information System (INIS)

Chen, Y.; Fischer, U.

2005-01-01

Shielding calculations of advanced nuclear facilities such as accelerator based neutron sources or fusion devices of the tokamak type are complicated due to their complex geometries and their large dimensions, including bulk shields of several meters thickness. While the complexity of the geometry in the shielding calculation can be hardly handled by the discrete ordinates method, the deep penetration of radiation through bulk shields is a severe challenge for the Monte Carlo particle transport technique. This work proposes a dedicated computational scheme for coupled Monte Carlo-Discrete Ordinates transport calculations to handle this kind of shielding problems. The Monte Carlo technique is used to simulate the particle generation and transport in the target region with both complex geometry and reaction physics, and the discrete ordinates method is used to treat the deep penetration problem in the bulk shield. The coupling scheme has been implemented in a program system by loosely integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and a newly developed coupling interface program for mapping process. Test calculations were performed with comparison to MCNP solutions. Satisfactory agreements were obtained between these two approaches. The program system has been chosen to treat the complicated shielding problem of the accelerator-based IFMIF neutron source. The successful application demonstrates that coupling scheme with the program system is a useful computational tool for the shielding analysis of complex and large nuclear facilities. (authors)

TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

International Nuclear Information System (INIS)

Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

1984-06-01

The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

Hot Hydrogen Test Facility

International Nuclear Information System (INIS)

W. David Swank

2007-01-01

The core in a nuclear thermal rocket will operate at high temperatures and in hydrogen. One of the important parameters in evaluating the performance of a nuclear thermal rocket is specific impulse, ISp. This quantity is proportional to the square root of the propellant's absolute temperature and inversely proportional to square root of its molecular weight. Therefore, high temperature hydrogen is a favored propellant of nuclear thermal rocket designers. Previous work has shown that one of the life-limiting phenomena for thermal rocket nuclear cores is mass loss of fuel to flowing hydrogen at high temperatures. The hot hydrogen test facility located at the Idaho National Lab (INL) is designed to test suitability of different core materials in 2500 C hydrogen flowing at 1500 liters per minute. The facility is intended to test non-uranium containing materials and therefore is particularly suited for testing potential cladding and coating materials. In this first installment the facility is described. Automated Data acquisition, flow and temperature control, vessel compatibility with various core geometries and overall capabilities are discussed

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

International Nuclear Information System (INIS)

Heard, F.J.

1997-01-01

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

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

International Nuclear Information System (INIS)

S, Tukiran; MS, Tagor; P, Surian

2003-01-01

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

"Cycling around an emotional core of sadness": emotion regulation in a couple after the loss of a child.

Science.gov (United States)

Hooghe, An; Neimeyer, Robert A; Rober, Peter

2012-09-01

In contrast to the traditional view of working through grief by confronting it, recent theories have emphasized an oscillating process of confronting and avoiding the pain of loss. In this qualitative study, we sought a better understanding of this process by conducting a detailed case study of a bereaved couple after the loss of their infant daughter. We employed multiple data collection methods (using interviews and written feedback) and an intensive auditing process in our thematic analysis, with special attention to a recurrent metaphor used by this bereaved couple in describing their personal and relational experience. The findings suggest the presence of a dialectic tension between the need to be close to the deceased child and the need for distance from the pain of the loss, which was evidenced on both individual and relational levels. For this couple, the image of "cycling around an emotional core of sadness" captured their dynamic way of dealing with this dialectic of closeness and distance.

Geodynamo Modeling of Core-Mantle Interactions

Science.gov (United States)

Kuang, Wei-Jia; Chao, Benjamin F.; Smith, David E. (Technical Monitor)

2001-01-01

Angular momentum exchange between the Earth's mantle and core influences the Earth's rotation on time scales of decades and longer, in particular in the length of day (LOD) which have been measured with progressively increasing accuracy for the last two centuries. There are four possible coupling mechanisms for transferring the axial angular momentum across the core-mantle boundary (CMB): viscous, magnetic, topography, and gravitational torques. Here we use our scalable, modularized, fully dynamic geodynamo model for the core to assess the importance of these torques. This numerical model, as an extension of the Kuang-Bloxham model that has successfully simulated the generation of the Earth's magnetic field, is used to obtain numerical results in various physical conditions in terms of specific parameterization consistent with the dynamical processes in the fluid outer core. The results show that depending on the electrical conductivity of the lower mantle and the amplitude of the boundary topography at CMB, both magnetic and topographic couplings can contribute significantly to the angular momentum exchange. This implies that the core-mantle interactions are far more complex than has been assumed and that there is unlikely a single dominant coupling mechanism for the observed decadal LOD variation.

FFTF [Fast Flux Test Facility] management

International Nuclear Information System (INIS)

Bennett, C.L.

1986-11-01

Fuel Management at the Fast Flux Test Facility (FFTF) involves more than just the usual ex-core and in-core management of standard fuel and non-fuel components between storage locations and within the core since it is primarily an irradiation test facility. This mission involves testing an ever increasing variety of fueled and non-fueled experiments, each having unique requirements on the reactor core as well as having its own individual impact on the reload design. This paper describes the fuel management process used by the Westinghouse Hanford Company Core Engineering group that has led to the successful reload design of nine operating cycles and the irradiation of over 120 tests

Flow analysis of HANARO flow simulated test facility

International Nuclear Information System (INIS)

Park, Yong-Chul; Cho, Yeong-Garp; Wu, Jong-Sub; Jun, Byung-Jin

2002-01-01

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial critical in February, 1995. Many experiments should be safely performed to activate the utilization of the NANARO. A flow simulated test facility is being developed for the endurance test of reactivity control units for extended life times and the verification of structural integrity of those experimental facilities prior to loading in the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The half-core structure assembly is composed of plenum, grid plate, core channel with flow tubes, chimney and dummy pool. The flow channels are to be filled with flow orifices to simulate core channels. This test facility must simulate similar flow characteristics to the HANARO. This paper, therefore, describes an analytical analysis to study the flow behavior of the test facility. The computational flow analysis has been performed for the verification of flow structure and similarity of this test facility assuming that flow rates and pressure differences of the core channel are constant. The shapes of flow orifices were determined by the trial and error method based on the design requirements of core channel. The computer analysis program with standard k - ε turbulence model was applied to three-dimensional analysis. The results of flow simulation showed a similar flow characteristic with that of the HANARO and satisfied the design requirements of this test facility. The shape of flow orifices used in this numerical simulation can be adapted for manufacturing requirements. The flow rate and the pressure difference through core channel proved by this simulation can be used as the design requirements of the flow system. The analysis results will be verified with the results of the flow test after construction of the flow system. (author)

Introducing FACETS, the Framework Application for Core-Edge Transport Simulations

International Nuclear Information System (INIS)

Cary, John R.; Candy, Jeff; Cohen, Ronald H.; Krasheninnikov, Sergei I.; McCune, Douglas C.; Estep, Donald J.; Larson, Jay W.; Malony, Allen; Worley, Patrick H.; Carlsson, Johann Anders; Hakim, A.H.; Hamill, P.; Kruger, Scott E.; Muzsala, S.; Pletzer, Alexander; Shasharina, Svetlana; Wade-Stein, D.; Wang, N.; McInnes, Lois C.; Wildey, T.; Casper, T.A.; Diachin, Lori A.; Epperly, Thomas; Rognlien, T.D.; Fahey, Mark R.; Kuehn, Jeffery A.; Morris, A.; Shende, Sameer; Feibush, E.; Hammett, Gregory W.; Indireshkumar, K.; Ludescher, C.; Randerson, L.; Stotler, D.; Pigarov, A.; Bonoli, P.; Chang, C.S.; D'Ippolito, D.A.; Colella, Philip; Keyes, David E.; Bramley, R.

2007-01-01

The FACETS (Framework Application for Core-Edge Transport Simulations) project began in January 2007 with the goal of providing core to wall transport modeling of a tokamak fusion reactor. This involves coupling previously separate computations for the core, edge, and wall regions. Such a coupling is primarily through connection regions of lower dimensionality. The project has started developing a component-based coupling framework to bring together models for each of these regions. In the first year, the core model will be a 1 dimensional model (1D transport across flux surfaces coupled to a 2D equilibrium) with fixed equilibrium. The initial edge model will be the fluid model, UEDGE, but inclusion of kinetic models is planned for the out years. The project also has an embedded Scientific Application Partnership that is examining embedding a full-scale turbulence model for obtaining the crosssurface fluxes into a core transport code.

Coupling of Large Eddy Simulations with Meteorological Models to simulate Methane Leaks from Natural Gas Storage Facilities

Science.gov (United States)

Prasad, K.

2017-12-01

Atmospheric transport is usually performed with weather models, e.g., the Weather Research and Forecasting (WRF) model that employs a parameterized turbulence model and does not resolve the fine scale dynamics generated by the flow around buildings and features comprising a large city. The NIST Fire Dynamics Simulator (FDS) is a computational fluid dynamics model that utilizes large eddy simulation methods to model flow around buildings at length scales much smaller than is practical with models like WRF. FDS has the potential to evaluate the impact of complex topography on near-field dispersion and mixing that is difficult to simulate with a mesoscale atmospheric model. A methodology has been developed to couple the FDS model with WRF mesoscale transport models. The coupling is based on nudging the FDS flow field towards that computed by WRF, and is currently limited to one way coupling performed in an off-line mode. This approach allows the FDS model to operate as a sub-grid scale model with in a WRF simulation. To test and validate the coupled FDS - WRF model, the methane leak from the Aliso Canyon underground storage facility was simulated. Large eddy simulations were performed over the complex topography of various natural gas storage facilities including Aliso Canyon, Honor Rancho and MacDonald Island at 10 m horizontal and vertical resolution. The goal of these simulations included improving and validating transport models as well as testing leak hypotheses. Forward simulation results were compared with aircraft and tower based in-situ measurements as well as methane plumes observed using the NASA Airborne Visible InfraRed Imaging Spectrometer (AVIRIS) and the next generation instrument AVIRIS-NG. Comparison of simulation results with measurement data demonstrate the capability of the coupled FDS-WRF models to accurately simulate the transport and dispersion of methane plumes over urban domains. Simulated integrated methane enhancements will be presented and

Mode-coupling in photonic crystal fibers with multiple cores

DEFF Research Database (Denmark)

Kristensen, Martin

2000-01-01

Summary form only given. We have fabricated a photonic crystal fiber (PCF) with multiple cores by drawing a fiber preform from stacked glass tubes. Transmission is high through each core despite many unintentional defects in the cladding indicating that the guidance is determined by the holes near...

Off-resonance frequency operation for power transfer in a loosely coupled air core transformer

Science.gov (United States)

Scudiere, Matthew B

2012-11-13

A power transmission system includes a loosely coupled air core transformer having a resonance frequency determined by a product of inductance and capacitance of a primary circuit including a primary coil. A secondary circuit is configured to have a substantially same product of inductance and capacitance. A back EMF generating device (e.g., a battery), which generates a back EMF with power transfer, is attached to the secondary circuit. Once the load power of the back EMF generating device exceeds a certain threshold level, which depends on the system parameters, the power transfer can be achieved at higher transfer efficiency if performed at an operating frequency less than the resonance frequency, which can be from 50% to 95% of the resonance frequency.

Simulation of natural circulation on an integral type experimental facility, MASLWR

Energy Technology Data Exchange (ETDEWEB)

Chung, Youngjong; Lim, Sungwon; Ha, Jaejoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

The OSU MASLWR test facility was reconfigured to eliminate a recurring grounding problem and improve facility reliability in anticipation of conducting an IAEA International Collaborative Standard Problem (ICSP). The purpose of ICSP is to provide experimental data on flow instability phenomena under natural circulation conditions and coupled containment/reactor vessel behavior in integral-type reactors, and to evaluate system code capabilities to predict natural circulation phenomena for integral type PWR, by simulating an integrated experiment. A natural circulation in the primary side during various core powers is analyzed using TASS/SMR code for the integral type experimental facility. The calculation results show higher steady state primary flow than experiment. If it matches the initial flow with experiment, it shows lower primary flow than experiment according to the increase of power. The code predictions may be improved by applying a Reynolds number dependent form loss coefficient to accurately account for unrecoverable pressure losses.

Facile measurement of {sup 1}H-{sup 15}N residual dipolar couplings in larger perdeuterated proteins

Energy Technology Data Exchange (ETDEWEB)

Fitzkee, Nicholas C.; Bax, Ad, E-mail: bax@nih.go [National Institutes of Health, Laboratory of Chemical Physics, National Institute of Diabetes and Digestive and Kidney Diseases (United States)

2010-10-15

We present a simple method, ARTSY, for extracting {sup 1}J{sub NH} couplings and {sup 1}H-{sup 15}N RDCs from an interleaved set of two-dimensional {sup 1}H-{sup 15}N TROSY-HSQC spectra, based on the principle of quantitative J correlation. The primary advantage of the ARTSY method over other methods is the ability to measure couplings without scaling peak positions or altering the narrow line widths characteristic of TROSY spectra. Accuracy of the method is demonstrated for the model system GB3. Application to the catalytic core domain of HIV integrase, a 36 kDa homodimer with unfavorable spectral characteristics, demonstrates its practical utility. Precision of the RDC measurement is limited by the signal-to-noise ratio, S/N, achievable in the 2D TROSY-HSQC spectrum, and is approximately given by 30/(S/N) Hz.

Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

International Nuclear Information System (INIS)

Sevon, T.

2005-11-01

In a hypothetical severe accident in a nuclear power plant, the molten core of the reactor may flow onto the concrete floor of containment building. This would cause a molten core . concrete interaction (MCCI), in which the heat transfer from the hot melt to the concrete would cause melting of the concrete. In assessing the safety of nuclear reactors, it is important to know the consequences of such an interaction. As background to the subject, this publication includes a description of the core melt stabilization concept of the European Pressurized water Reactor (EPR), which is being built in Olkiluoto in Finland. The publication includes a description of the basic theory of the interaction and the process of spalling or cracking of concrete when it is heated rapidly. A literature survey and some calculations of the physical properties of concrete and corium. concrete mixtures at high temperatures have been conducted. In addition, an equation is derived for conservative calculation of the maximum possible concrete ablation depth. The publication also includes a literature survey of experimental research on the subject of the MCCI and discussion of the results and deficiencies of the experiments. The main result of this work is the general design of an experimental facility to examine the interaction of molten metals and concrete. The main objective of the experiments is to assess the probability of spalling, or cracking, of concrete under pouring of molten material. A program of five experiments has been designed, and pre-test calculations of the experiments have been conducted with MELCOR 1.8.5 accident analysis program and conservative analytic calculations. (orig.)

Magnetohydrodynamic Convection in the Outer Core and its Geodynamic Consequences

Science.gov (United States)

Kuang, Weijia; Chao, Benjamin F.; Fang, Ming

2004-01-01

The Earth's fluid outer core is in vigorous convection through much of the Earth's history. In addition to generating and maintaining Earth s time-varying magnetic field (geodynamo), the core convection also generates mass redistribution in the core and a dynamical pressure field on the core-mantle boundary (CMB). All these shall result in various core-mantle interactions, and contribute to surface geodynamic observables. For example, electromagnetic core-mantle coupling arises from finite electrically conducting lower mantle; gravitational interaction occurs between the cores and the heterogeneous mantle; mechanical coupling may also occur when the CMB topography is aspherical. Besides changing the mantle rotation via the coupling torques, the mass-redistribution in the core shall produce a spatial-temporal gravity anomaly. Numerical modeling of the core dynamical processes contributes in several geophysical disciplines. It helps explain the physical causes of surface geodynamic observables via space geodetic techniques and other means, e.g. Earth's rotation variation on decadal time scales, and secular time-variable gravity. Conversely, identification of the sources of the observables can provide additional insights on the dynamics of the fluid core, leading to better constraints on the physics in the numerical modeling. In the past few years, our core dynamics modeling efforts, with respect to our MoSST model, have made significant progress in understanding individual geophysical consequences. However, integrated studies are desirable, not only because of more mature numerical core dynamics models, but also because of inter-correlation among the geophysical phenomena, e.g. mass redistribution in the outer core produces not only time-variable gravity, but also gravitational core-mantle coupling and thus the Earth's rotation variation. They are expected to further facilitate multidisciplinary studies of core dynamics and interactions of the core with other

Understanding the effects of the core on the nutation of the Earth

Directory of Open Access Journals (Sweden)

Véronique Dehant

2017-11-01

Full Text Available In this review paper, we examine the changes in the Earth orientation in space and focus on the nutation (shorter-term periodic variations, which is superimposed on precession (long-term trend on a timescale of years. We review the nutation modelling involving several coupling mechanisms at the core-mantle boundary using the Liouville angular momentum equations for a two-layered Earth with a liquid flattened core. The classical approach considers a Poincaré fluid for the core with an inertial pressure coupling mechanism at the core-mantle boundary. We examine possible additional coupling mechanisms to explain the observations. In particular, we examine how we can determine the flattening of the core as well as information on the magnetic field and the core flow from the nutation observations. The precision of the observations is shown to be high enough to increase our understanding on the coupling mechanisms at the core-mantle boundary.

Determination of BEACON Coupling Coefficients using data from Xenon transient

International Nuclear Information System (INIS)

Bozic, M.; Kurincic, B.

2007-01-01

NEK uses BEACO TM code (BEACO TM - Westinghouse Best Estimate Analyzer for Core Operating Nuclear) for core monitoring, analysis and core behaviour prediction. Coupling Coefficients determine relationship between core response and excore instrumentation. Measured power distribution using incore moveable detectors during Xenon transient with sufficient power axial offset change is the most important data for further analysis. Classic methodology and BEACO TM Conservative methodology using established Coupling Coefficients are compared on NPP Krsko case. BEACON TM Conservative methodology with predefined Coupling Coefficients is used as a surveillance tool for verification of relationship between core and excore instrumentation during power operation. (author)

TMI-2 core examination plan

International Nuclear Information System (INIS)

Owen, D.E.; MacDonald, P.E.; Hobbins, R.R.; Ploggr, S.A.

1982-01-01

The Three Mile Island (TMI-2) core examination is divided into four stages: (1) before removing the head; (2) before removing the plenum; (3) during defueling; and (4) offsite examinations. Core examinations recommended during the first three stages are primarily devoted to documenting the post-accident condition of the core. The detailed analysis of core damage structures will be performed during offsite examinations at government and commercial hot cell facilities. The primary objectives of these examinations are to enhance the understanding of the degraded core accident sequence, to develop the technical bases for reactor regulations, and to improve LWR design and operation

Concurrent, parallel, multiphysics coupling in the FACETS project

Energy Technology Data Exchange (ETDEWEB)

Cary, J R; Carlsson, J A; Hakim, A H; Kruger, S E; Miah, M; Pletzer, A; Shasharina, S [Tech-X Corporation, 5621 Arapahoe Avenue, Suite A, Boulder, CO 80303 (United States); Candy, J; Groebner, R J [General Atomics (United States); Cobb, J; Fahey, M R [Oak Ridge National Laboratory (United States); Cohen, R H; Epperly, T [Lawrence Livermore National Laboratory (United States); Estep, D J [Colorado State University (United States); Krasheninnikov, S [University of California at San Diego (United States); Malony, A D [ParaTools, Inc (United States); McCune, D C [Princeton Plasma Physics Laboratory (United States); McInnes, L; Balay, S [Argonne National Laboratory (United States); Pankin, A, E-mail: cary@txcorp.co [Lehigh University (United States)

2009-07-01

FACETS (Framework Application for Core-Edge Transport Simulations), is now in its third year. The FACETS team has developed a framework for concurrent coupling of parallel computational physics for use on Leadership Class Facilities (LCFs). In the course of the last year, FACETS has tackled many of the difficult problems of moving to parallel, integrated modeling by developing algorithms for coupled systems, extracting legacy applications as components, modifying them to run on LCFs, and improving the performance of all components. The development of FACETS abides by rigorous engineering standards, including cross platform build and test systems, with the latter covering regression, performance, and visualization. In addition, FACETS has demonstrated the ability to incorporate full turbulence computations for the highest fidelity transport computations. Early indications are that the framework, using such computations, scales to multiple tens of thousands of processors. These accomplishments were a result of an interdisciplinary collaboration among computational physics, computer scientists and applied mathematicians on the team.

ARTEMIS: The core simulator of AREVA NP's next generation coupled neutronics/thermal-hydraulics code system ARCADIAR

International Nuclear Information System (INIS)

Hobson, Greg; Merk, Stephan; Bolloni, Hans-Wilhelm; Breith, Karl-Albert; Curca-Tivig, Florin; Van Geemert, Rene; Heinecke, Jochen; Hartmann, Bettina; Porsch, Dieter; Tiles, Viatcheslav; Dall'Osso, Aldo; Pothet, Baptiste

2008-01-01

AREVA NP has developed a next-generation coupled neutronics/thermal-hydraulics code system, ARCADIA R , to fulfil customer's current demands and even anticipate their future demands in terms of accuracy and performance. The new code system will be implemented world-wide and will replace several code systems currently used in various global regions. An extensive phase of verification and validation of the new code system is currently in progress. One of the principal components of this new system is the core simulator, ARTEMIS. Besides the stand-alone tests on the individual computational modules, integrated tests on the overall code are being performed in order to check for non-regression as well as for verification of the code. Several benchmark problems have been successfully calculated. Full-core depletion cycles of different plant types from AREVA's French, American and German regions (e.g. N4 and KONVOI types) have been performed with ARTEMIS (using APOLLO2-A cross sections) and compared directly with current production codes, e.g. with SCIENCE and CASCADE-3D, and additionally with measurements. (authors)

Superrotation of Earth’s Inner Core, Extraterrestrial Impacts, and the Effective Viscosity of Outer Core

OpenAIRE

Pirooz Mohazzabi; John D. Skalbeck

2015-01-01

The recently verified superrotation of Earth’s inner core is examined and a new model is presented which is based on the tidal despinning of the mantle and the viscosity of the outer core. The model also takes into account other damping mechanisms arising from the inner core superrotation such as magnetic and gravitational coupling as well as contribution from eddy viscosity in the outer core. The effective viscosity obtained in this model confirms a previously well constrained value of about...

A facile approach for cupric ion detection in aqueous media using polyethyleneimine/PMMA core-shell fluorescent nanoparticles

International Nuclear Information System (INIS)

Chen Jian; Zeng Fang; Wu Shuizhu; Su Junhua; Zhao Jianqing; Tong Zhen

2009-01-01

A facile approach was developed to produce a dye-doped core-shell nanoparticle chemosensor for detecting Cu 2+ in aqueous media. The core-shell nanoparticle sensor was prepared by a one-step emulsifier-free polymerization, followed by the doping of the fluorescent dye Nile red (9-diethylamino- 5H-benzo[alpha] phenoxazine-5-one, NR) into the particles. For the nanoparticles, the hydrophilic polyethyleneimine (PEI) chain segments serve as the shell and the hydrophobic polymethyl methacrylate (PMMA) constitutes the core of the nanoparticles. The non-toxic and biocompatible PEI chain segments on the nanoparticle surface exhibit a high affinity for Cu 2+ ions in aqueous media, and the quenching of the NR fluorescence is observed upon binding of Cu 2+ ions. This makes the core-shell nanoparticle system a water-dispersible chemosensor for Cu 2+ ion detection. The quenching of fluorescence arises through intraparticle energy transfer (FRET) from the dye in the hydrophobic PMMA core to the Cu 2+ /PEI complexes on the nanoparticle surface. The energy transfer efficiency for PEI/PMMA particles with different diameters was determined, and it is found that the smaller nanoparticle sample exhibits higher quenching efficiency, and the limit for Cu 2+ detection is 1 μM for a nanoparticle sample with a diameter of ∼30 nm. The response of the fluorescent nanoparticle towards different metal ions was investigated and the nanoparticle chemosensor displays high selectivity and antidisturbance for the Cu 2+ ion among the metal ions examined (Na + , K + , Mg 2+ , Ca 2+ , Zn 2+ , Hg 2+ , Mn 2+ , Fe 2+ , Ni 2+ , Co 2+ and Pb 2+ ). This emulsifier-free, biocompatible and sensitive fluorescent nanoparticle sensor may find applications in cupric ion detection in the biological and environmental areas.

A facile approach for cupric ion detection in aqueous media using polyethyleneimine/PMMA core-shell fluorescent nanoparticles.

Science.gov (United States)

Chen, Jian; Zeng, Fang; Wu, Shuizhu; Su, Junhua; Zhao, Jianqing; Tong, Zhen

2009-09-09

A facile approach was developed to produce a dye-doped core-shell nanoparticle chemosensor for detecting Cu(2+) in aqueous media. The core-shell nanoparticle sensor was prepared by a one-step emulsifier-free polymerization, followed by the doping of the fluorescent dye Nile red (9-diethylamino- 5H-benzo[alpha] phenoxazine-5-one, NR) into the particles. For the nanoparticles, the hydrophilic polyethyleneimine (PEI) chain segments serve as the shell and the hydrophobic polymethyl methacrylate (PMMA) constitutes the core of the nanoparticles. The non-toxic and biocompatible PEI chain segments on the nanoparticle surface exhibit a high affinity for Cu(2+) ions in aqueous media, and the quenching of the NR fluorescence is observed upon binding of Cu(2+) ions. This makes the core-shell nanoparticle system a water-dispersible chemosensor for Cu(2+) ion detection. The quenching of fluorescence arises through intraparticle energy transfer (FRET) from the dye in the hydrophobic PMMA core to the Cu(2+)/PEI complexes on the nanoparticle surface. The energy transfer efficiency for PEI/PMMA particles with different diameters was determined, and it is found that the smaller nanoparticle sample exhibits higher quenching efficiency, and the limit for Cu(2+) detection is 1 microM for a nanoparticle sample with a diameter of approximately 30 nm. The response of the fluorescent nanoparticle towards different metal ions was investigated and the nanoparticle chemosensor displays high selectivity and antidisturbance for the Cu(2+) ion among the metal ions examined (Na(+), K(+), Mg(2+), Ca(2+), Zn(2+), Hg(2+), Mn(2+), Fe(2+), Ni(2+), Co(2+) and Pb(2+)). This emulsifier-free, biocompatible and sensitive fluorescent nanoparticle sensor may find applications in cupric ion detection in the biological and environmental areas.

A facile approach for cupric ion detection in aqueous media using polyethyleneimine/PMMA core-shell fluorescent nanoparticles

Science.gov (United States)

Chen, Jian; Zeng, Fang; Wu, Shuizhu; Su, Junhua; Zhao, Jianqing; Tong, Zhen

2009-09-01

A facile approach was developed to produce a dye-doped core-shell nanoparticle chemosensor for detecting Cu2+ in aqueous media. The core-shell nanoparticle sensor was prepared by a one-step emulsifier-free polymerization, followed by the doping of the fluorescent dye Nile red (9-diethylamino- 5H-benzo[alpha] phenoxazine-5-one, NR) into the particles. For the nanoparticles, the hydrophilic polyethyleneimine (PEI) chain segments serve as the shell and the hydrophobic polymethyl methacrylate (PMMA) constitutes the core of the nanoparticles. The non-toxic and biocompatible PEI chain segments on the nanoparticle surface exhibit a high affinity for Cu2+ ions in aqueous media, and the quenching of the NR fluorescence is observed upon binding of Cu2+ ions. This makes the core-shell nanoparticle system a water-dispersible chemosensor for Cu2+ ion detection. The quenching of fluorescence arises through intraparticle energy transfer (FRET) from the dye in the hydrophobic PMMA core to the Cu2+/PEI complexes on the nanoparticle surface. The energy transfer efficiency for PEI/PMMA particles with different diameters was determined, and it is found that the smaller nanoparticle sample exhibits higher quenching efficiency, and the limit for Cu2+ detection is 1 µM for a nanoparticle sample with a diameter of ~30 nm. The response of the fluorescent nanoparticle towards different metal ions was investigated and the nanoparticle chemosensor displays high selectivity and antidisturbance for the Cu2+ ion among the metal ions examined (Na+, K+, Mg2+, Ca2+, Zn2+, Hg2+, Mn2+, Fe2+, Ni2+, Co2+ and Pb2+). This emulsifier-free, biocompatible and sensitive fluorescent nanoparticle sensor may find applications in cupric ion detection in the biological and environmental areas.

Environmentally Regulated Facilities in Iowa

Data.gov (United States)

Iowa State University GIS Support and Research Facility — A unique record for each facility site with an environmental interest by DNR (such as permits). This brings together core environmental information in one place for...

Dinuclear Tetrapyrazolyl Palladium Complexes Exhibiting Facile Tandem Transfer Hydrogenation/Suzuki Coupling Reaction of Fluoroarylketone

KAUST Repository

Dehury, Niranjan

2016-07-18

Herein, we report an unprecedented example of dinuclear pyrazolyl-based Pd complexes exhibiting facile tandem catalysis for fluoroarylketone: Tetrapyrazolyl di-palladium complexes with varying Pd-Pd distances efficiently catalyze the tandem reaction involving transfer hydrogenation of fluoroarylketone to the corresponding alcohol and Suzuki-Miyaura cross coupling reaction of the resulting fluoroarylalcohol under moderate reaction conditions, to biaryl alcohol. The complex with the shortest Pd-Pd distance exhibits the highest tandem activity among its di-metallic analogues, and exceeds in terms of activity and selectivity the analogous mononuclear compound. The kinetics of the reaction indicates clearly that reductive transformation of haloarylketone into haloaryalcohol is the rate determining step in the tandem reaction. Interestingly while fluoroarylketone undergoes the multistep tandem catalysis, the chloro- and bromo-arylketones undergo only a single step C-C coupling reaction resulting in biarylketone as the final product. Unlike the pyrazole based Pd compounds, the precursor PdCl2 and the phosphine based relevant complexes (PPh3)2PdCl2 and (PPh3)4Pd are found to be unable to exhibit the tandem catalysis.

Test of the core design methods for the THTR 300 with experimental results from the critical facility KAHTER

Energy Technology Data Exchange (ETDEWEB)

Hofmann, K.; Huebner, A.; Brandes, S.; Krings, F.

1974-10-15

At the Kernforschungsanlage Juelich, core physics experiments with core 1 and core 2 of the critical facility for high temperature reactors KAHTER were carried out in 1973. Core 2 corresponds to the THTR initial core in its moderation ratio S = 7500. Selected experimental results on the critical mass, on control rod worths, and reaction rate distributions were used for testing the most important procedures for the THTR core physics design. The zero-dimensional spectrum program MUPO with its cross section library and the and neutron flux calculations in two-dimensional diffusion approximation by CRAM are of central importance. It proved to be important to introduce modifications specific to the KAHTER plant into the standard models. Thus the void effect (void above the pebble bed) was investigated with DOT-2 by transport theory and a correction was introduced for the critical masses calculated by diffusion theory. Another feature already contained in the standard procedure, the increase of the diffusion constants for the hollow spaces between the spheres, results in a correction of 3.8% < delta-k for KAHTER, whereas in the THTR 300 it only amounts to several tenths % delta-k. Critical masses are predicted with accuracies of < 1.5 % or with regard to reactivity < 0.65 % delta-k. The calculated values for the radial neutron flux distributions deviate from the measured values in the core area by approximately lo %. In the case of the axial profiles, deviations are observed at the pebble bed surface which can be explained by the upper void, which cannot be satisfactorily represented by the diffusion theory. Control rod worths are predicted quite well, i.e., to within +/- 5%. An exception is the bank of 4 reflector rods, where the applied model of the "grey curtain" is not accurate because of the large distances between rods. The calculated control rod worths for that case were found to be too low, which does, however, not result in a safety problem.

Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

International Nuclear Information System (INIS)

Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

2011-01-01

Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

Development of three dimensional transient analysis code STTA for SCWR core

International Nuclear Information System (INIS)

Wang, Lianjie; Zhao, Wenbo; Chen, Bingde; Yao, Dong; Yang, Ping

2015-01-01

Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation

Core map generation for the ITU TRIGA Mark II research reactor using Genetic Algorithm coupled with Monte Carlo method

Energy Technology Data Exchange (ETDEWEB)

Türkmen, Mehmet, E-mail: tm@hacettepe.edu.tr [Nuclear Engineering Department, Hacettepe University, Beytepe Campus, Ankara (Turkey); Çolak, Üner [Energy Institute, Istanbul Technical University, Ayazağa Campus, Maslak, Istanbul (Turkey); Ergün, Şule [Nuclear Engineering Department, Hacettepe University, Beytepe Campus, Ankara (Turkey)

2015-12-15

Highlights: • Optimum core maps were generated for the ITU TRIGA Mark II Research Reactor. • Calculations were performed using a Monte Carlo based reactor physics code, MCNP. • Single-Objective and Multi-Objective Genetic Algorithms were used for the optimization. • k{sub eff} and ppf{sub max} were considered as the optimization objectives. • The generated core maps were compared with the fresh core map. - Abstract: The main purpose of this study is to present the results of Core Map (CM) generation calculations for the İstanbul Technical University TRIGA Mark II Research Reactor by using Genetic Algorithms (GA) coupled with a Monte Carlo (MC) based-particle transport code. Optimization problems under consideration are: (i) maximization of the core excess reactivity (ρ{sub ex}) using Single-Objective GA when the burned fuel elements with no fresh fuel elements are used, (ii) maximization of the ρ{sub ex} and minimization of maximum power peaking factor (ppf{sub max}) using Multi-Objective GA when the burned fuels with fresh fuels are used. The results were obtained when all the control rods are fully withdrawn. ρ{sub ex} and ppf{sub max} values of the produced best CMs were provided. Core-averaged neutron spectrum, and variation of neutron fluxes with respect to radial distance were presented for the best CMs. The results show that it is possible to find an optimum CM with an excess reactivity of 1.17 when the burned fuels are used. In the case of a mix of burned fuels and fresh fuels, the best pattern has an excess reactivity of 1.19 with a maximum peaking factor of 1.4843. In addition, when compared with the fresh CM, the thermal fluxes of the generated CMs decrease by about 2% while change in the fast fluxes is about 1%.Classification: J. Core physics.

Full Core TREAT Kinetics Demonstration Using Rattlesnake/BISON Coupling Within MAMMOTH

Energy Technology Data Exchange (ETDEWEB)

Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alberti, Anthony L. [Oregon State Univ., Corvallis, OR (United States); Palmer, Todd S. [Oregon State Univ., Corvallis, OR (United States)

2015-08-01

This report summarizes key aspects of research in evaluation of modeling needs for TREAT transient simulation. Using a measured TREAT critical measurement and a transient for a small, experimentally simplified core, Rattlesnake and MAMMOTH simulations are performed building from simple infinite media to a full core model. Cross sections processing methods are evaluated, various homogenization approaches are assessed and the neutronic behavior of the core studied to determine key modeling aspects. The simulation of the minimum critical core with the diffusion solver shows very good agreement with the reference Monte Carlo simulation and the experiment. The full core transient simulation with thermal feedback shows a significantly lower power peak compared to the documented experimental measurement, which is not unexpected in the early stages of model development.

Application of consistent fluid added mass matrix to core seismic

International Nuclear Information System (INIS)

Koo, K. H.; Lee, J. H.

2003-01-01

In this paper, the application algorithm of a consistent fluid added mass matrix including the coupling terms to the core seismic analysis is developed and installed at SAC-CORE3.0 code. As an example, we assumed the 7-hexagon system of the LMR core and carried out the vibration modal analysis and the nonlinear time history seismic response analysis using SAC-CORE3.0. Used consistent fluid added mass matrix is obtained by using the finite element program of the FAMD(Fluid Added Mass and Damping) code. From the results of the vibration modal analysis, the core duct assemblies reveal strongly coupled vibration modes, which are so different from the case of in-air condition. From the results of the time history seismic analysis, it was verified that the effects of the coupled terms of the consistent fluid added mass matrix are significant in impact responses and the dynamic responses

Facile consecutive solvothermal growth of highly fluorescent InP/ZnS core/shell quantum dots using a safer phosphorus source.

Science.gov (United States)

Byun, Ho-June; Song, Woo-Seuk; Yang, Heesun

2011-06-10

The work presents a facile, stepwise synthetic approach for the production of highly fluorescent InP/ZnS core/shell quantum dots (QDs) by using a safer phosphorus (P) precursor. First, InP quantum dots (QDs) were solvothermally prepared at 180 °C for 24 h by using a P source of P(N(CH(3))(2))(3). The as-grown InP QDs were consecutively placed in another solvothermal condition for ZnS shell overcoating. In contrast to the almost non-fluorescent InP QDs, due to their highly defective surface states, the ZnS-coated InP QDs were highly fluorescent as a result of effective surface passivation. After the shell growth, the resulting InP/ZnS core/shell QDs were subjected to a size-sorting processing, by which red- to green-emitting QDs with quantum yields (QYs) of 24-60% were produced. Solvothermal shell growth parameters such as the reaction time and Zn/In solution concentration ratio were varied and optimized toward the highest QYs of core/shell QDs.

Facile consecutive solvothermal growth of highly fluorescent InP/ZnS core/shell quantum dots using a safer phosphorus source

International Nuclear Information System (INIS)

Byun, Ho-June; Song, Woo-Seuk; Yang, Heesun

2011-01-01

The work presents a facile, stepwise synthetic approach for the production of highly fluorescent InP/ZnS core/shell quantum dots (QDs) by using a safer phosphorus (P) precursor. First, InP quantum dots (QDs) were solvothermally prepared at 180 deg. C for 24 h by using a P source of P(N(CH 3 ) 2 ) 3 . The as-grown InP QDs were consecutively placed in another solvothermal condition for ZnS shell overcoating. In contrast to the almost non-fluorescent InP QDs, due to their highly defective surface states, the ZnS-coated InP QDs were highly fluorescent as a result of effective surface passivation. After the shell growth, the resulting InP/ZnS core/shell QDs were subjected to a size-sorting processing, by which red- to green-emitting QDs with quantum yields (QYs) of 24-60% were produced. Solvothermal shell growth parameters such as the reaction time and Zn/In solution concentration ratio were varied and optimized toward the highest QYs of core/shell QDs.

Facile consecutive solvothermal growth of highly fluorescent InP/ZnS core/shell quantum dots using a safer phosphorus source

Science.gov (United States)

Byun, Ho-June; Song, Woo-Seuk; Yang, Heesun

2011-06-01

The work presents a facile, stepwise synthetic approach for the production of highly fluorescent InP/ZnS core/shell quantum dots (QDs) by using a safer phosphorus (P) precursor. First, InP quantum dots (QDs) were solvothermally prepared at 180 °C for 24 h by using a P source of P(N(CH3)2)3. The as-grown InP QDs were consecutively placed in another solvothermal condition for ZnS shell overcoating. In contrast to the almost non-fluorescent InP QDs, due to their highly defective surface states, the ZnS-coated InP QDs were highly fluorescent as a result of effective surface passivation. After the shell growth, the resulting InP/ZnS core/shell QDs were subjected to a size-sorting processing, by which red- to green-emitting QDs with quantum yields (QYs) of 24-60% were produced. Solvothermal shell growth parameters such as the reaction time and Zn/In solution concentration ratio were varied and optimized toward the highest QYs of core/shell QDs.

Facile consecutive solvothermal growth of highly fluorescent InP/ZnS core/shell quantum dots using a safer phosphorus source

Energy Technology Data Exchange (ETDEWEB)

Byun, Ho-June; Song, Woo-Seuk; Yang, Heesun, E-mail: hyang@hongik.ac.kr [Department of Materials Science and Engineering, Hongik University, Seoul 121-791 (Korea, Republic of)

2011-06-10

The work presents a facile, stepwise synthetic approach for the production of highly fluorescent InP/ZnS core/shell quantum dots (QDs) by using a safer phosphorus (P) precursor. First, InP quantum dots (QDs) were solvothermally prepared at 180 deg. C for 24 h by using a P source of P(N(CH{sub 3}){sub 2}){sub 3}. The as-grown InP QDs were consecutively placed in another solvothermal condition for ZnS shell overcoating. In contrast to the almost non-fluorescent InP QDs, due to their highly defective surface states, the ZnS-coated InP QDs were highly fluorescent as a result of effective surface passivation. After the shell growth, the resulting InP/ZnS core/shell QDs were subjected to a size-sorting processing, by which red- to green-emitting QDs with quantum yields (QYs) of 24-60% were produced. Solvothermal shell growth parameters such as the reaction time and Zn/In solution concentration ratio were varied and optimized toward the highest QYs of core/shell QDs.

Core Facility of the Juelich Observatory for Cloud Evolution (JOYCE - CF)

Science.gov (United States)

Beer, J.; Troemel, S.

2017-12-01

A multiple and holistic multi-sensor monitoring of clouds and precipitation processes is a challenging but promising task in the meteorological community. Instrument synergies offer detailed views in microphysical and dynamical developments of clouds. Since 2017 The the Juelich Observatory for Cloud Evolution (JOYCE) is transformed into a Core Facility (JOYCE - CF). JOYCE - CF offers multiple long-term remote sensing observations of the atmosphere, develops an easy access to all observations and invites scientists word wide to exploit the existing data base for their research but also to complement JOYCE-CF with additional long-term or campaign instrumentation. The major instrumentation contains a twin set of two polarimetric X-band radars, a microwave profiler, two cloud radars, an infrared spectrometer, a Doppler lidar and two ceilometers. JOYCE - CF offers easy and open access to database and high quality calibrated observations of all instruments. E.g. the two polarimetric X-band radars which are located in 50 km distance are calibrated using the self-consistency method, frequently repeated vertical pointing measurements as well as instrument synergy with co-located micro-rain radar and distrometer measurements. The presentation gives insights into calibration procedures, the standardized operation procedures and recent synergistic research exploiting our radars operating at three different frequencies.

CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

International Nuclear Information System (INIS)

Cleveland, J.C.

1977-01-01

CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP

Core design methods for advanced LMFBRs

International Nuclear Information System (INIS)

Chandler, J.C.; Marr, D.R.; McCurry, D.C.; Cantley, D.A.

1977-05-01

The multidiscipline approach to advanced LMFBR core design requires an iterative design procedure to obtain a closely-coupled design. HEDL's philosophy requires that the designs should be coupled to the extent that the design limiting fuel pin, the design limiting duct and the core reactivity lifetime should all be equal and should equal the fuel residence time. The design procedure consists of an iterative loop involving three stages of the design sequence. Stage 1 consists of general mechanical design and reactor physics scoping calculations to arrive at an initial core layout. Stage 2 consists of detailed reactor physics calculations for the core configuration arrived at in Stage 1. Based upon the detailed reactor physics results, a decision is made either to alter the design (Stage 1) or go to Stage 3. Stage 3 consists of core orificing and detailed component mechanical design calculations. At this point, an assessment is made regarding design adequacy. If the design is inadequate the entire procedure is repeated until the design is acceptable

Biochemistry Instrumentation Core Technology Center

Data.gov (United States)

Federal Laboratory Consortium — The UCLA-DOE Biochemistry Instrumentation Core Facility provides the UCLA biochemistry community with easy access to sophisticated instrumentation for a wide variety...

Synthesis of Au@Ag core-shell nanocubes containing varying shaped cores and their localized surface plasmon resonances.

Science.gov (United States)

Gong, Jianxiao; Zhou, Fei; Li, Zhiyuan; Tang, Zhiyong

2012-06-19

We have synthesized Au@Ag core-shell nanocubes containing Au cores with varying shapes and sizes through modified seed-mediated methods. Bromide ions are found to be crucial in the epitaxial growth of Ag atoms onto Au cores and in the formation of the shell's cubic shape. The Au@Ag core-shell nanocubes exhibit very abundant and distinct localized surface plasmon resonance (LSPR) properties, which are core-shape and size-dependent. With the help of theoretical calculation, the physical origin and the resonance mode profile of each LSPR peak are identified and studied. The core-shell nanocrystals with varying shaped cores offer a new rich category for LSPR control through the plasmonic coupling effect between core and shell materials.

Study of long-range orders of hard-core bosons coupled to cooperative normal modes in two-dimensional lattices

Science.gov (United States)

Ghosh, A.; Yarlagadda, S.

2017-09-01

Understanding the microscopic mechanism of coexisting long-range orders (such as lattice supersolidity) in strongly correlated systems is a subject of immense interest. We study the possible manifestations of long-range orders, including lattice-supersolid phases with differently broken symmetry, in a two-dimensional square lattice system of hard-core bosons (HCBs) coupled to archetypal cooperative/coherent normal-mode distortions such as those in perovskites. At strong HCB-phonon coupling, using a duality transformation to map the strong-coupling problem to a weak-coupling one, we obtain an effective Hamiltonian involving nearest-neighbor, next-nearest-neighbor, and next-to-next-nearest-neighbor hoppings and repulsions. Using stochastic series expansion quantum Monte Carlo, we construct the phase diagram of the system. As coupling strength is increased, we find that the system undergoes a first-order quantum phase transition from a superfluid to a checkerboard solid at half-filling and from a superfluid to a diagonal striped solid [with crystalline ordering wave vector Q ⃗=(2 π /3 ,2 π /3 ) or (2 π /3 ,4 π /3 )] at one-third filling without showing any evidence of supersolidity. On tuning the system away from these commensurate fillings, checkerboard supersolid is generated near half-filling whereas a rare diagonal striped supersolid is realized near one-third filling. Interestingly, there is an asymmetry in the extent of supersolidity about one-third filling. Within our framework, we also provide an explanation for the observed checkerboard and stripe formations in La2 -xSrxNiO4 at x =1 /2 and x =1 /3 .

Experiments utilizing two coupled TRIGA-type reactors

Energy Technology Data Exchange (ETDEWEB)

Thayer, G [Southern California Edison Co., Rosemead, CA (United States); Jones, B G; Miley, G H [University of Illinois (United States)

1974-07-01

An experimental study has been performed on a coupled-core system consisting of two reactors each of which can be made critical by itself, coupled neutronically by a graphite thermal column. Both steady-state and transient measurements were performed on the system. The steady-state measurement consisted of measuring the coupling coefficient between the two reactors. Also, series of measurements were performed while one of the cores was far subcritical and the coupling between the two cores was varied between 1.6 x 10{sup -2} and 1.6 x 10{sup -5} cents by the insertion of a water gap and from 1.6 x 10{sup -2} cents to 6.0 x 10{sup -4} cents by the insertion of a cadmium sheet between the cores. The transient portion of the study was performed by pulsing one of the reactors (the Illinois Advanced TRIGA) and following the pulse into the passive core (the Low Power Reactor Assembly). The first pulse series measured the pulse as it emerged from the thermal column and propagated through the water, where no fuel was present. This provided an analysis of the neutron source to the passive core. The second pulse series was performed with the passive core far subcritical (k{sub eff} {approx_equal} 0.94) and investigated the effects on the transient coupling of the insertion of water gaps of up to 9 inches or a cadmium sheet ({sigma}T = 3.2) between the two cores. Spatial measurements of the pulse in the far subcritical assembly also were performed. The third series of pulses investigated the characteristics of the pulse in the passive core when it was subcritical, just critical, and supercritical, The effects on the FWHM of the pulse in the passive core and on the delay time between the peak of the pulse in the TRIGA and the passive core were measured for the passive core having a k{sub eff} from 0.936 to 1.0015 and the initial period of the pulse in TRIGA varying from 15.6 {+-} .7 ms to 3.58 {+-} .05 ms. The FWHM increased from 13.5 {+-} 0.5 ms to 18.8 {+-} 0.5 ms and delay

Quantized orbits in weakly coupled Belousov-Zhabotinsky reactors

Science.gov (United States)

Weiss, S.; Deegan, R. D.

2015-06-01

Using numerical and experimental tools, we study the motion of two coupled spiral cores in a light-sensitive variant of the Belousov-Zhabotinsky reaction. Each core resides on a separate two-dimensional domain, and is coupled to the other by light. When both spirals have the same sense of rotation, the cores are attracted to a circular trajectory with a diameter quantized in integer units of the spiral wavelength λ. When the spirals have opposite senses of rotation, the cores are attracted towards different but parallel straight trajectories, separated by an integer multiple of λ/2. We present a model that explains this behavior as the result of a spiral wavefront-core interaction that produces a deterministic displacement of the core and a retardation of its phase.

Facile one-step hydrothermal synthesis toward strongly coupled TiO2/graphene quantum dots photocatalysts for efficient hydrogen evolution

International Nuclear Information System (INIS)

Min, Shixiong; Hou, Jianhua; Lei, Yonggang; Ma, Xiaohua; Lu, Gongxuan

2017-01-01

Highlights: • TiO 2 /GQDs composites were prepared by a facile one-step hydrothermal method. • GQDs were strongly coupled onto the surface of TiO 2 nanoparticles by this method. • The TiO 2 /GQDs showed enhanced light absorption and charge separation efficiency. • The TiO 2 /GQDs exhibited higher photocatalytic H 2 evolution activity than pure TiO 2 . • GQDs play synergistic roles by acting as both photosensitizer and electron acceptor. - Abstract: The coupling of semiconductor photocatalysts with graphene quantum dots (GQDs) has been proven to be an effective strategy to enhance the photocatalytic and photoelectrical conversion performances of the resulted composites; however, the preparation of semiconductor/GQDs composites usually involves several time-inefficient and tedious post-treatment steps. Herein, we present a facile one-step hydrothermal route for the preparation of GQDs coupled TiO 2 (TiO 2 /GQDs) photocatalysts using 1,3,6-trinitropyrene (TNP) as the sole precursor of GQDs. During the hydrothermal process, TNP molecules undergo an intramolecular fusion to form GQDs, which simultaneously decorate on the surface of TiO 2 nanoparticles, leading to a strong surface interaction between the two components. The effective coupling of GQDs on TiO 2 can effectively extend the light absorption of the TiO 2 to visible region and enhance the charge separation efficiency of TiO 2 /GQDs composites as a result of GQDs acting as a photosensitizer and an excellent electron acceptor. These key advances make the TiO 2 /GQDs photocatalyst highly active towards the H 2 evolution reaction, resulting in 7 and 3 times higher H 2 evolution rate and photocurrent response at optimal GQDs content than TiO 2 alone, respectively. This study provides a new methodology for the development of high-performance GQDs modified semiconductor photocatalysts for energy conversion applications.

SIMULATE-3 K coupled code applications

Energy Technology Data Exchange (ETDEWEB)

Joensson, Christian [Studsvik Scandpower AB, Vaesteraas (Sweden); Grandi, Gerardo; Judd, Jerry [Studsvik Scandpower Inc., Idaho Falls, ID (United States)

2017-07-15

This paper describes the coupled code system TRACE/SIMULATE-3 K/VIPRE and the application of this code system to the OECD PWR Main Steam Line Break. A short description is given for the application of the coupled system to analyze DNBR and the flexibility the system creates for the user. This includes the possibility to compare and evaluate the result with the TRACE/SIMULATE-3K (S3K) coupled code, the S3K standalone code (core calculation) as well as performing single-channel calculations with S3K and VIPRE. This is the typical separate-effect-analyses required for advanced calculations in order to develop methodologies to be used for safety analyses in general. The models and methods of the code systems are presented. The outline represents the analysis approach starting with the coupled code system, reactor and core model calculation (TRACE/S3K). This is followed by a more detailed core evaluation (S3K standalone) and finally a very detailed thermal-hydraulic investigation of the hot pin condition (VIPRE).

LMFBR post accident heat removal testing needs and conceptual design of a test facility

International Nuclear Information System (INIS)

Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

1977-03-01

A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

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)

Antiferromagnetic exchange coupling measurements on single Co clusters

Science.gov (United States)

Wernsdorfer, W.; Leroy, D.; Portemont, C.; Brenac, A.; Morel, R.; Notin, L.; Mailly, D.

2009-03-01

We report on single-cluster measurements of the angular dependence of the low-temperature ferromagnetic core magnetization switching field in exchange-coupled Co/CoO core-shell clusters (4 nm) using a micro-bridge DC superconducting quantum interference device (μ-SQUID). It is observed that the coupling with the antiferromagnetic shell induces modification in the switching field for clusters with intrinsic uniaxial anisotropy depending on the direction of the magnetic field applied during the cooling. Using a modified Stoner-Wohlfarth model, it is shown that the core interacts with two weakly coupled and asymmetrical antiferromagnetic sublattices. Ref.: C. Portemont, R. Morel, W. Wernsdorfer, D. Mailly, A. Brenac, and L. Notin, Phys. Rev. B 78, 144415 (2008)

SCTF Core-I test results

International Nuclear Information System (INIS)

Adachi, Hiromichi; Sudo, Yukio; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Hirano, Kemmei

1982-07-01

The Slab Core Test Facility (SCTF) of Japan Atomic Energy Research Institute (JAERI) was constructed to investigate two-dimensional thermohydrodynamics in the core and the communication in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a posturated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). In the present report, effects of system pressure on reflooding phenomena shall be discussed based on the data of Tests S1-SH2, S1-01 and S1-02 which are the parameteris tests for system pressure effects belonging to the SCTF Core-I forced flooding test series. Major items discussed in this report are (1) hydrodynamic behavior in the system, (2) core thermal behavior, (3) core heat transfer and (4) two-dimensional hydrodynamic behavior in the pressure vessel including the core. (author)

Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

International Nuclear Information System (INIS)

Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.

2012-01-01

The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

Energy Technology Data Exchange (ETDEWEB)

Sun, K. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Chenu, A. [Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Mikityuk, K.; Krepel, J. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland)

2012-07-01

The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

Directory of Open Access Journals (Sweden)

Jorge Pérez Mañes

2014-01-01

Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

Rearrangement of a polar core provides a conserved mechanism for constitutive activation of class B G protein-coupled receptors

Science.gov (United States)

Yin, Yanting; de Waal, Parker W.; He, Yuanzheng; Zhao, Li-Hua; Yang, Dehua; Cai, Xiaoqing; Jiang, Yi; Melcher, Karsten; Wang, Ming-Wei; Xu, H. Eric

2017-01-01

The glucagon receptor (GCGR) belongs to the secretin-like (class B) family of G protein-coupled receptors (GPCRs) and is activated by the peptide hormone glucagon. The structures of an activated class B GPCR have remained unsolved, preventing a mechanistic understanding of how these receptors are activated. Using a combination of structural modeling and mutagenesis studies, we present here two modes of ligand-independent activation of GCGR. First, we identified a GCGR-specific hydrophobic lock comprising Met-338 and Phe-345 within the IC3 loop and transmembrane helix 6 (TM6) and found that this lock stabilizes the TM6 helix in the inactive conformation. Disruption of this hydrophobic lock led to constitutive G protein and arrestin signaling. Second, we discovered a polar core comprising conserved residues in TM2, TM3, TM6, and TM7, and mutations that disrupt this polar core led to constitutive GCGR activity. On the basis of these results, we propose a mechanistic model of GCGR activation in which TM6 is held in an inactive conformation by the conserved polar core and the hydrophobic lock. Mutations that disrupt these inhibitory elements allow TM6 to swing outward to adopt an active TM6 conformation similar to that of the canonical β2-adrenergic receptor complexed with G protein and to that of rhodopsin complexed with arrestin. Importantly, mutations in the corresponding polar core of several other members of class B GPCRs, including PTH1R, PAC1R, VIP1R, and CRFR1, also induce constitutive G protein signaling, suggesting that the rearrangement of the polar core is a conserved mechanism for class B GPCR activation. PMID:28356352

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

Science.gov (United States)

Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

2017-01-01

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one

The OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark - Steady-state results and status

International Nuclear Information System (INIS)

Reitsma, F.; Han, J.; Ivanov, K.; Sartori, E.

2008-01-01

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated. Since only a few pebble-bed HTR experimental facilities or plant data are available the use of code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MW design and a representative set of transient cases is defined as an OECD benchmark. The scope of the benchmark is to establish a series of well-defined multi-dimensional computational benchmark problems with a common given set of cross-sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. The OECD benchmark includes steady-state and transients cases. Although the focus of the benchmark is on the modelling of the transient behaviour of the PBMR core, it was also necessary to define some steady-state cases to ensure consistency between the different approaches before results of transient cases could be compared. This paper describes the status of the benchmark project and shows the results for the three steady state exercises defined as a standalone neutronics calculation, a standalone thermal-hydraulic core calculation, and a coupled neutronics/thermal-hydraulic simulation. (authors)

Measurements of neutron fluxes and cadmium ratio at equilibrium core in JRR-3M

International Nuclear Information System (INIS)

Ohtomo, Akitoshi; Sasajima, Fumio; Ishida, Takuya; Shigemoto, Masamitsu; Takahashi, Hidetake; Maejima, Takeshi; Sekine, Katsunori.

1993-08-01

Construction and characteristics tests of JRR-3M (Modified JRR-3) had been completed on October 1990, and the reactor reached to equilibrium core in July 1991. Measurements of neutron flux and cadmium ratio in Hydraulic irradiation facility (HR) and Pneumatic irradiation facility (PN) at 20 MW reactor power were carried out for the equilibrium core from May to August 1991 and for the latest core in April 1993. The results at the equilibrium core and the latest core are described in this paper. (author)

Facile synthesis of NaYF4:Yb, Ln/NaYF4:Yb core/shell upconversion nanoparticles via successive ion layer adsorption and one-pot reaction technique

NARCIS (Netherlands)

Zeng, Q.; Xue, B.; Zhang, Y.; Wang, D.; Liu, X.; Tu, L.; Zhao, H.; Kong, X.; Zhang, H.

2013-01-01

The facile one-pot synthesis of NaYF4:Yb, Ln/NaYF4:Yb core/shell (CS) upconversion nanoparticles (UCNPs) was firstly developed through the successive ion layer adsorption and reaction (SILAR) technique, which represents an attractive alternative to conventional synthesis utilizing the chloride of Ln

First results from core-edge parallel composition in the FACETS project.

Energy Technology Data Exchange (ETDEWEB)

Cary, J. R.; Candy, J.; Cohen, R. H.; Krasheninnikov, S.; McCune, D. C.; Estep, D. J.; Larson, J.; Malony, A. D.; Pankin, A.; Worley, P. H.; Carlsson, J. A.; Hakim, A. H.; Hamill, P.; Kruger, S.; Miah, M.; Muzsala, S.; Pletzer, A.; Shasharina, S.; Wade-Stein, D.; Wang, N.; Balay, S.; McInnes, L.; Zhang, H.; Casper, T.; Diachin, L. (Mathematics and Computer Science); (Tech-X Corp.); (General Atomics); (LLNL); (Univ. of California at San Diego); (Princeton Plasma Physics Lab.); (Colorado State Univ.); (ParaTools Inc.); (Lehigh Univ.); (ORNL)

2008-01-01

FACETS (Framework Application for Core-Edge Transport Simulations), now in its second year, has achieved its first coupled core-edge transport simulations. In the process, a number of accompanying accomplishments were achieved. These include a new parallel core component, a new wall component, improvements in edge and source components, and the framework for coupling all of this together. These accomplishments were a result of an interdisciplinary collaboration among computational physics, computer scientists, and applied mathematicians on the team.

First results from core-edge parallel composition in the FACETS project

Energy Technology Data Exchange (ETDEWEB)

Cary, J R; Carlsson, J A; Hakim, A H; Hamill, P; Kruger, S; Miah, M; Muzsala, S; Pletzer, A; Shasharina, S; Wade-Stein, D; Wang, N [Tech-X Corporation, Boulder, CO 80303 (United States); Candy, J [General Atomics, San Diego, CA 92186 (United States); Cohen, R H [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Krasheninnikov, S [University of California at San Diego, San Diego, CA 92093 (United States); McCune, D C [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Estep, D J [Colorado State University, Fort Collins, CO 80523 (United States); Larson, J [Argonne National Laboratory, Argonne, IL 60439 (United States); Malony, A D [ParaTools, Inc., Eugene, OR 97405 (United States); Pankin, A [Lehigh University, Bethlehem, PA 18015 (United States); Worley, P H [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)], E-mail: cary@txcorp.com (and others)

2008-07-15

FACETS (Framework Application for Core-Edge Transport Simulations), now in its second year, has achieved its first coupled core-edge transport simulations. In the process, a number of accompanying accomplishments were achieved. These include a new parallel core component, a new wall component, improvements in edge and source components, and the framework for coupling all of this together. These accomplishments were a result of an interdisciplinary collaboration among computational physics, computer scientists, and applied mathematicians on the team.

First results from core-edge parallel composition in the FACETS project

Energy Technology Data Exchange (ETDEWEB)

Cary, John R. [Tech-X Corporation; Candy, Jeff [General Atomics; Cohen, Ronald H. [Lawrence Livermore National Laboratory (LLNL); Krasheninnikov, Sergei [University of California, San Diego; McCune, Douglas [Princeton Plasma Physics Laboratory (PPPL); Estep, Donald J [Colorado State University, Fort Collins; Larson, Jay [Argonne National Laboratory (ANL); Malony, Allen [University of Oregon; Pankin, A. [Lehigh University, Bethlehem, PA; Worley, Patrick H [ORNL; Carlsson, Johann [Tech-X Corporation; Hakim, A H [Tech-X Corporation; Hamill, P [Tech-X Corporation; Kruger, Scott [Tech-X Corporation; Miah, Mahmood [Tech-X Corporation; Muzsala, S [Tech-X Corporation; Pletzer, Alexander [Tech-X Corporation; Shasharina, Svetlana [Tech-X Corporation; Wade-Stein, D [Tech-X Corporation; Wang, N [Tech-X Corporation; Balay, Satish [Argonne National Laboratory (ANL); McInnes, Lois [Argonne National Laboratory (ANL); Zhang, Hong [Argonne National Laboratory (ANL); Casper, T. A. [Lawrence Livermore National Laboratory (LLNL); Diachin, Lori [Lawrence Livermore National Laboratory (LLNL); Epperly, Thomas [Lawrence Livermore National Laboratory (LLNL); Rognlien, T. D. [Lawrence Livermore National Laboratory (LLNL); Fahey, Mark R [ORNL; Cobb, John W [ORNL; Morris, A [University of Oregon; Shende, Sameer [University of Oregon; Hammett, Greg [Princeton Plasma Physics Laboratory (PPPL); Indireshkumar, K [Tech-X Corporation; Stotler, D. [Princeton Plasma Physics Laboratory (PPPL); Pigarov, A [University of California, San Diego

2008-01-01

FACETS (Framework Application for Core-Edge Transport Simulations), now in its second year, has achieved its first coupled core-edge transport simulations. In the process, a number of accompanying accomplishments were achieved. These include a new parallel core component, a new wall component, improvements in edge and source components, and the framework for coupling all of this together. These accomplishments were a result of an interdisciplinary collaboration among computational physics, computer scientists, and applied mathematicians on the team.

ZZ-PBMR-400, OECD/NEA PBMR Coupled Neutronics/Thermal Hydraulics Transient Benchmark - The PBMR-400 Core Design

International Nuclear Information System (INIS)

Reitsma, Frederik

2007-01-01

Description of benchmark: This international benchmark, concerns Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transients based on the PBMR-400 MW design. The deterministic neutronics, thermal-hydraulics and transient analysis tools and methods available to design and analyse PBMRs lag, in many cases, behind the state of the art compared to other reactor technologies. This has motivated the testing of existing methods for HTGRs but also the development of more accurate and efficient tools to analyse the neutronics and thermal-hydraulic behaviour for the design and safety evaluations of the PBMR. In addition to the development of new methods, this includes defining appropriate benchmarks to verify and validate the new methods in computer codes. The scope of the benchmark is to establish well-defined problems, based on a common given set of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark exercise has the following objectives: - Establish a standard benchmark for coupled codes (neutronics/thermal-hydraulics) for PBMR design; - Code-to-code comparison using a common cross section library ; - Obtain a detailed understanding of the events and the processes; - Benefit from different approaches, understanding limitations and approximations. Major Design and Operating Characteristics of the PBMR (PBMR Characteristic and Value): Installed thermal capacity: 400 MW(t); Installed electric capacity: 165 MW(e); Load following capability: 100-40-100%; Availability: ≥ 95%; Core configuration: Vertical with fixed centre graphite reflector; Fuel: TRISO ceramic coated U-235 in graphite spheres; Primary coolant: Helium; Primary coolant pressure: 9 MPa; Moderator: Graphite; Core outlet temperature: 900 C.; Core inlet temperature: 500 C.; Cycle type: Direct; Number of circuits: 1; Cycle

Fluid effects on the core seismic behavior of a liquid metal reactor

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Lee, Jae Han

2004-01-01

In this paper, a numerical application algorithm for applying the CFAM (Consistent Fluid Added Mass) matrix for a core seismic analysis is developed and applied to the 7-ducts core system to investigate the fluid effects on the dynamic characteristics and the seismic time history responses. To this end, three cases such as the in-air condition, the in-water condition without the fluid coupling terms, and the in-water condition with the fluid coupling terms are considered in this paper. From modal analysis, the core duct assemblies revealed strongly coupled out-of-phase vibration modes unlike the other cases with the fluid coupling terms considered. From the results of the seismic time history analysis, it was also verified that the fluid coupling terms in the CFAM matrix can significantly affect the impact responses and the seismic displacement responses of the ducts

Innovative research reactor core designed. Estimation and analysis of gamma heating distribution

International Nuclear Information System (INIS)

Setiyanto

2014-01-01

The Gamma heating value is an important factor needed for safety analysis of each experiments that will be realized on research reactor core. Gamma heat is internal heat source occurs in each irradiation facilities or any material irradiated in reactor core. This value should be determined correctly because of the safety related problems. The gamma heating value is in general depend on. reactor core characteristics, different one and other, and then each new reactor design should be completed by gamma heating data. The Innovative Research Reactor is one of the new reactor design that should be completed with any safety data, including the gamma heating value. For this reasons, calculation and analysis of gamma heating in the hole of reactor core and irradiation facilities in reflector had been done by using of modified and validated Gamset computer code. The result shown that gamma heating value of 11.75 W/g is the highest value at the center of reactor core, higher than gamma heating value of RSG-GAS. However, placement of all irradiation facilities in reflector show that safety characteristics for irradiation facilities of innovative research reactor more better than RSG-GAS reactor. Regarding the results obtained, and based on placement of irradiation facilities in reflector, can be concluded that innovative research reactor more safe for any irradiation used. (author)

Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

Energy Technology Data Exchange (ETDEWEB)

Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

2017-06-01

Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

Core/corona modeling of diode-imploded annular loads

Science.gov (United States)

Terry, R. E.; Guillory, J. U.

1980-11-01

The effects of a tenuous exterior plasma corona with anomalous resistivity on the compression and heating of a hollow, collisional aluminum z-pinch plasma are predicted by a one-dimensional code. As the interior ("core") plasma is imploded by its axial current, the energy exchange between core and corona determines the current partition. Under the conditions of rapid core heating and compression, the increase in coronal current provides a trade-off between radial acceleration and compression, which reduces the implosion forces and softens the pitch. Combined with a heuristic account of energy and momentum transport in the strongly coupled core plasma and an approximate radiative loss calculation including Al line, recombination and Bremsstrahlung emission, the current model can provide a reasonably accurate description of imploding annular plasma loads that remain azimuthally symmetric. The implications for optimization of generator load coupling are examined.

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)

Coupling of AST-500 heating reactors with desalination facilities

International Nuclear Information System (INIS)

Kourachenkov, A.V.

1998-01-01

The general issues regarding NHR and desalination facility joint operation for potable water production are briefly considered. AST-500 reactor plant and DOU GTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. Similarity of NHR operation for a heating grid and a desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author)

Heterogeneous cores for fast breeder reactor

International Nuclear Information System (INIS)

Schroeder, R.; Spenke, H.

1980-01-01

Firstly, the motivation for heterogeneous cores is discussed. This is followed by an outline of two reactor designs, both of which are variants of the combined ring and island core. These designs are presented by means of figures and detailed tables. Subsequently, a description of two international projects at fast critical zero energy facilities is given. Both of them support the nuclear design of heterogeneous cores. In addition to a survey of these projects, a typical experiment is discussed: the measurement of rate distributions. (orig.) [de

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

International Nuclear Information System (INIS)

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

1977-01-01

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

17 CFR 37.6 - Compliance with core principles.

Science.gov (United States)

2010-04-01

... 17 Commodity and Securities Exchanges 1 2010-04-01 2010-04-01 false Compliance with core principles. 37.6 Section 37.6 Commodity and Securities Exchanges COMMODITY FUTURES TRADING COMMISSION DERIVATIVES TRANSACTION EXECUTION FACILITIES § 37.6 Compliance with core principles. (a) In general. To...

Initial Coupling of the RELAP-7 and PRONGHORN Applications

Energy Technology Data Exchange (ETDEWEB)

J. Ortensi; D. Andrs; A.A. Bingham; R.C. Martineau; J.W. Peterson

2012-10-01

Modern nuclear reactor safety codes require the ability to solve detailed coupled neutronic- thermal fluids problems. For larger cores, this implies fully coupled higher dimensionality spatial dynamics with appropriate feedback models that can provide enough resolution to accurately compute core heat generation and removal during steady and unsteady conditions. The reactor analysis code PRONGHORN is being coupled to RELAP-7 as a first step to extend RELAP’s current capabilities. This report details the mathematical models, the type of coupling, and the testing results from the integrated system. RELAP-7 is a MOOSE-based application that solves the continuity, momentum, and energy equations in 1-D for a compressible fluid. The pipe and joint capabilities enable it to model parts of the power conversion unit. The PRONGHORN application, also developed on the MOOSE infrastructure, solves the coupled equations that define the neutron diffusion, fluid flow, and heat transfer in a full core model. The two systems are loosely coupled to simplify the transition towards a more complex infrastructure. The integration is tested on a simplified version of the OECD/NEA MHTGR-350 Coupled Neutronics-Thermal Fluids benchmark model.

Development of the flow control irradiation facility for JOYO

International Nuclear Information System (INIS)

Soroi, Masatoshi; Miyakawa, Shun-ichi

1998-05-01

This report describes the present situation and problems with the development of the flow control irradiation facility (FLORA). The purpose of FLORA is to run the cladding breach (RTCB) irradiation test under loss of flow conditions in the experimental fast reactor 'JOYO'. FLORA is a facility like FPTF (Fuel Performance Test Facility) plus BFTF (Breached Fuel Test Facility) in EBR-II, USA. The technical feature of FLORA is its annular linear induction pump (A-LIP), which was developed in response to a need identified through the experiences in the mechanical flow control of FPTF. We have already designed the basic system facility of FLORA for the JOYO MK-II core. However, to put FLORA to practical use in the future, we have to confirm the stability of the JOYO MK-III core condition, solve problems and improve the design. We are going to freeze and review the FLORA project, taking into consideration the fuel development situation and the research project of JOYO MK-III core. (J.P.N.)

NICHD Microscopy and Imaging Core (MIC)

Data.gov (United States)

Federal Laboratory Consortium — The NICHD Microscopy and Imaging Core (MIC) is designed as a multi-user research facility providing training and instrumentation for high resolution microscopy and...

Evaporation Of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Effluent Management Facility Core Simulant

Energy Technology Data Exchange (ETDEWEB)

Adamson, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Mcclane, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-01

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation, and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator, in the Effluent Management Facility (EMF), and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator, so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would reduce the need for closely integrated operation of the LAW melter and the Pretreatment Facilities. Long-term implementation of this option after WTP start-up would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other operational complexities such a recycle stream presents. In order to accurately plan for the disposition path, it is key to experimentally determine the fate of contaminants. To do this, testing is needed to accurately account for the buffering chemistry of the components, determine the achievable evaporation end point, identify insoluble solids that form, and determine the distribution of key regulatory-impacting constituents. The LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures, have limited solubility in the glass waste form, and represent a materials corrosion concern, such as halides and sulfate. Because this stream will recycle within WTP, these components will accumulate in the Melter Condensate

Blowdown hydraulic influence on core thermal response in LOFT nuclear experiment L2-3

International Nuclear Information System (INIS)

Reeder, D.L.

1979-01-01

Experimental research into pressurized water reactor (PWR) loss-of-coolant phenomena conducted in the Loss-of-Fluid Test (LOFT) facility has given results indicating that for very large pipe breaks the core thermal response is tightly coupled to the fluid hydraulic phenomena during the blowdown phase of the loss-of-coolant transient. This summary presents and discusses data supporting this conclusion. LOFT Loss-of-Coolant Experiment (LOCE) L2-3 simulated a complete double-ended offset shear break of a primary coolant reactor vessel inlet pipe in a commercial PWR. The LOFT system conditions at experiment initiation were: fuel rod maximum linear heat generation rate (MLHGR) of 39.4 +- 3 kW/m, hot leg temperature of 593 +- 3 K, core ΔT of 32.2 +- 4 K, system pressure of 15.06 +- 0.03 MPa, and flow rate/system volume of 25.6 +- 0.8 kg/m 3 . These conditions are typical of those in commercial PWR systems at normal operating conditions

Rearrangement of a polar core provides a conserved mechanism for constitutive activation of class B G protein-coupled receptors.

Science.gov (United States)

Yin, Yanting; de Waal, Parker W; He, Yuanzheng; Zhao, Li-Hua; Yang, Dehua; Cai, Xiaoqing; Jiang, Yi; Melcher, Karsten; Wang, Ming-Wei; Xu, H Eric

2017-06-16

The glucagon receptor (GCGR) belongs to the secretin-like (class B) family of G protein-coupled receptors (GPCRs) and is activated by the peptide hormone glucagon. The structures of an activated class B GPCR have remained unsolved, preventing a mechanistic understanding of how these receptors are activated. Using a combination of structural modeling and mutagenesis studies, we present here two modes of ligand-independent activation of GCGR. First, we identified a GCGR-specific hydrophobic lock comprising Met-338 and Phe-345 within the IC3 loop and transmembrane helix 6 (TM6) and found that this lock stabilizes the TM6 helix in the inactive conformation. Disruption of this hydrophobic lock led to constitutive G protein and arrestin signaling. Second, we discovered a polar core comprising conserved residues in TM2, TM3, TM6, and TM7, and mutations that disrupt this polar core led to constitutive GCGR activity. On the basis of these results, we propose a mechanistic model of GCGR activation in which TM6 is held in an inactive conformation by the conserved polar core and the hydrophobic lock. Mutations that disrupt these inhibitory elements allow TM6 to swing outward to adopt an active TM6 conformation similar to that of the canonical β 2 -adrenergic receptor complexed with G protein and to that of rhodopsin complexed with arrestin. Importantly, mutations in the corresponding polar core of several other members of class B GPCRs, including PTH1R, PAC1R, VIP1R, and CRFR1, also induce constitutive G protein signaling, suggesting that the rearrangement of the polar core is a conserved mechanism for class B GPCR activation. © 2017 by The American Society for Biochemistry and Molecular Biology, Inc.

Analysis of core physics and thermal-hydraulics results of control rod withdrawal experiments in the LOFT facility

International Nuclear Information System (INIS)

Varacalle, D.J. Jr.; Chen, T.H.; Harvego, E.A.; Ollikkala, H.

1983-01-01

Two anticipated transient experiments simulating an uncontrolled control rod withdrawal event in a pressurized water reactor (PWR) were conducted in the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The scaled LOFT 50-MW(t) PWR includes most of the principal features of larger commercial PWRs. The experiments tested the ability of reactor analysis codes to accurately calculate core reactor physics and thermal-hydraulic phenomena in an integral reactor system. The initial conditions and scaled operating parameters for the experiments were representative of those expected in a commercial PWR. In both experiments, all four LOFT control rod assemblies were withdrawn at a reactor power of 37.5 MW and a system pressure of 14.8 MPa

Coupling of AST-500 heating reactors with desalination facilities

International Nuclear Information System (INIS)

Gureyeva, L.V.; Egorov, V.V.; Podberezniy, V.L.

1997-01-01

The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab

Coupling of AST-500 heating reactors with desalination facilities

Energy Technology Data Exchange (ETDEWEB)

Gureyeva, L V; Egorov, V V [OKBM, Nizhny Novgorod (Russian Federation); Podberezniy, V L [Scientific Research Inst. of Machine Building, Ekaterinburg (Russian Federation)

1997-09-01

The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab.

Solution of the 6th dynamic AER benchmark using the coupled core DYN3D/ATHLET

International Nuclear Information System (INIS)

Seidel, A.; Kliem, S.

2001-01-01

The 6 th dynamic benchmark is a logical continuation of the work to validate systematically coupled neutron kinetics/thermohydraulics code systems for the estimation of the transient behaviour of WWER type nuclear power plant which was started in the 5 th dynamic benchmark. This benchmark concerns a double ended break of the main steam line (asymmetrical MSLB) in a WWER plant. The core is at the end of first cycle in full power conditions. The asymmetric leak causes a different depressurization of all steam generators. New features in comparison to the 5 th dynamic benchmark were included: asymmetric operation of the feed water system, consideration of incomplete coolant mixing in the reactor vessel, and the definition of a fixed isothermal recriticality temperature for normalising the nuclear data (Authors)

Core heatup prediction during SB LOCA with RELAP5/MOD3.2.2 Gamma

International Nuclear Information System (INIS)

Parzer, I.; Mavko, B.; Petelin, S.

2001-01-01

The paper focuses on the phenomena leading to core uncovering and heatup during the SB LOCA and the ability of RELAP5/MOD3.2.2 Gamma to predict core overheating. The code prediction has been compared to the three experiments, one conducted on the separate effect test facility NEPTUN in Switzerland and the other two conducted on two integral test facilities, PMK-2 in Hungary and PACTEL facility in Finland. In the case of a series of boiloff experiments performed on the NEPTUN test facility the influence of the two correlations available in MOD3.2.2 Gamma for determining interphase drag has been studied. In the case of IAEA-SPE-4 experiment simulation on PMK-2 facility the main goal of the analysis was to study the adequate modeling of the hexagonal core channel with 19-rod bundle and the phenomena during the core uncovering. The third analyzed experiment, OECD-ISP-33, was performed on PACTEL facility to study different natural circulation modes during SB LOCA. The analysis also focused on the final stage of this SB LOCA experiment, when core dryout and heatup was observed due to gradual emptying of the primary system. Following the experience the appropriate modeling options have been used to achieve better representation of the important phenomena during the SB LOCA.(author)

Prismatic Core Coupled Transient Benchmark

International Nuclear Information System (INIS)

Ortensi, J.; Pope, M.A.; Strydom, G.; Sen, R.S.; DeHart, M.D.; Gougar, H.D.; Ellis, C.; Baxter, A.; Seker, V.; Downar, T.J.; Vierow, K.; Ivanov, K.

2011-01-01

The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts that have existed for some time. Several prismatic units have operated in the world (DRAGON, Fort St. Vrain, Peach Bottom) and one unit is still in operation (HTTR). The deterministic neutronics and thermal-fluids transient analysis tools and methods currently available for the design and analysis of PMRs have lagged behind the state of the art compared to LWR reactor technologies. This has motivated the development of more accurate and efficient tools for the design and safety evaluations of the PMR. In addition to the work invested in new methods, it is essential to develop appropriate benchmarks to verify and validate the new methods in computer codes. The purpose of this benchmark is to establish a well-defined problem, based on a common given set of data, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events. The benchmark-working group is currently seeking OECD/NEA sponsorship. This benchmark is being pursued and is heavily based on the success of the PBMR-400 exercise.

Quantitative conformational analysis of the core region of N-glycans using residual dipolar couplings, aqueous molecular dynamics, and steric alignment

International Nuclear Information System (INIS)

Almond, Andrew; Duus, Jens O.

2001-01-01

A method is described for quantitatively investigating the dynamic conformation of small oligosaccharides containing an α(1 → 6) linkage. It was applied to the oligosaccharide Man-α(1 → 3) {Man-α (1 → 6)}Man-α-O-Me, which is a core region frequently observed in N-linked glycans. The approach tests an aqueous molecular dynamics simulation, capable of predicting microscopic dynamics, against experimental residual dipolar couplings, by assuming that alignment is caused purely by steric hindrance. The experimental constraints were heteronuclear and homonuclear residual dipolar couplings, and in particular those within the α(1 → 6) linkage itself. Powerful spin-state-selective pulse sequences and editing schemes were used to obtain the most relevant couplings for testing the model. Molecular dynamics simulations in water over a period of 50 ns were not able to predict the correct rotamer population at the α(1 → 6) linkage to agree with the experimental data. However, this sampling problem could be corrected using a simple maximum likelihood optimisation, indicating that the simulation was modelling local dynamics correctly. The maximum likelihood prediction of the residual dipolar couplings was found to be an almost equal population of the gg and gt rotamer conformations at the α(1 → 6) linkage, and the tg conformation was predicted to be unstable and unpopulated in aqueous solution. In this case all twelve measured residual dipolar couplings could be satisfied. This conformer population could also be used to make predictions of scalar couplings with the use of a previously derived empirical equation, and is qualitatively in agreement with previous predictions based on NMR, X-ray crystallography and optical data

Jahn-Teller effect in Rydberg series: A multi-state vibronic coupling problem

International Nuclear Information System (INIS)

Staib, A.; Domcke, W.; Sobolewski, A.L.

1990-01-01

Two simple limiting cases of Jahn-Teller (JT) coupling in Rydberg states of polyatomic molecules are considered, namely (i) JT coupling in Rydberg orbitals as well as in the ionization continuum (nondegenerate ion core, degenerate Rydberg series) and (ii) JT coupling in the ion core (degenerate ion core, nondegenerate Rydberg series). For both models simple and efficient algorithms for the computation of spectra (dynamical JT effect) are developed. The orbital JT effect is shown to represent a novel type of multi-state vibronic coupling, giving rise to interesting spectroscopic phenomena, among them resonant inter-Rydberg perturbations and JT induced autoionization. Particular attention is paid to the demonstration of the characteristic spectroscopic signatures of the two types of JT coupling in Rydberg states. (orig.)

Direct chemoselective synthesis of glyconanoparticles from unprotected reducing glycans and glycopeptide aldehydes

DEFF Research Database (Denmark)

Thygesen, Mikkel Boas; Sørensen, Kasper Kildegaard; Cló, Emiliano

2009-01-01

Chemoselective oxime coupling was used for facile conjugation of unprotected, reducing glycans and glycopeptide aldehydes with core-shell gold nanoparticles carrying reactive aminooxy groups on the organic shell.......Chemoselective oxime coupling was used for facile conjugation of unprotected, reducing glycans and glycopeptide aldehydes with core-shell gold nanoparticles carrying reactive aminooxy groups on the organic shell....

Photonic crystal fiber design for broadband directional coupling

DEFF Research Database (Denmark)

Lægsgaard, Jesper; Bang, Ole; Bjarklev, Anders Overgaard

2004-01-01

A novel design for a broadband directional coupler based on a photonic crystal fiber is investigated numerically. It is shown that suitable index-depressing doping of the core regions in an index-guiding twin-core photonic crystal fiber can stabilize the coupling coefficient between the cores over...

Current status of irradiation facilities in JRR-3 and JRR-4

International Nuclear Information System (INIS)

Hori, Naohiko; Wada, Shigeru; Sasajima, Fumio; Kusunoki, Tsuyoshi

2006-01-01

The Department of Research Reactor has operated two research reactors, JRR-3 and JRR-4. These reactors were constructed in the Tokai Research Establishment. Many researchers and engineers use these joint-use facilities. JRR-3 is a light water moderated and cooled, pool type research reactor using low-enriched silicide fuel. JRR-3's maximum thermal power is 20MW. JRR-3 has nine vertical irradiation holes for RI production, nuclear fuels and materials irradiation at reactor core area. JRR-3 has many kinds of irradiation holes in a heavy water tank around the reactor core. These are two hydraulic rabbit irradiation facilities, two pneumatic rabbit irradiation facilities, one activation analysis irradiation facilities, one uniform irradiation facility, one rotating irradiation facility and one capsule irradiation facility. JRR-3 has nine horizontal experimental holes, that are used by many kinds of neutron beam experimental facilities using these holes. JRR-4 is a light water moderated and cooled, swimming pool type research reactor using low-enriched silicide fuel. JRR-4's maximum thermal power is 3.5MW. JRR-4 has five vertical irradiation tubes at reactor core area, three capsule irradiation facilities, one hydraulic rabbit irradiation facility, and one pneumatic rabbit irradiation facility. JRR-4 has a neutron beam hole, and it has used neutron beam experiments, irradiations for activation analysis and medical neutron irradiations. (author)

An investigation of core liquid level depression in small break loss-of-coolant accidents

International Nuclear Information System (INIS)

Schultz, R.R.; Watkins, J.C.; Motley, F.E.; Stumpf, H.; Chen, Y.S.

1991-08-01

Core liquid level depression can result in partial core dryout and heatup early in a small break loss-of-coolant accident (SBLOCA) transient. Such behavior occurs when steam, trapped in the upper regions of the reactor primary system (between the loop seal and the core inventory), moves coolant out of the core region and uncovers the rod upper elevations. The net result is core liquid level depression. Core liquid level depression and subsequent core heatups are investigated using subscale data from the ROSA-IV Program's 1/48-scale Large Scale Test Facility (LSTF) and the 1/1705-scale Semiscale facility. Both facilities are Westinghouse-type, four-loop, pressurized water reactor simulators. The depression phenomena and factors which influence the minimum core level are described and illustrated using examples from the data. Analyses of the subject experiments, conducted using the TRAC-PF1/MOD1 (Version 12.7) thermal-hydraulic code, are also described and summarized. Finally, the response of a typical Westinghouse four-loop plant (RESAR-3S) was calculated to qualitatively study coal liquid level depression in a full-scale system. 31 refs., 37 figs., 6 tabs

Consequences of the conversion of research reactor cores on experimental facilities at the example of a cold neutron source

International Nuclear Information System (INIS)

Roegler, H.-J.; Goebs, H.; Stroemich, A.

1985-01-01

The consequences for and specifically the potential reduction of the performance of research reactors have been in discussions very often within the last five years as one of the draw-backs which has to be paid for the reduction of the proliferation risk at research reactor plants. Up to now and up to our knowledge the available results are restricted to unperturbated fluxes. Thus, this contribution makes the attempt to demonstrate the consequence of core conversion on an example of a real experimental facility and - at the same time - on one that is going to be used in the next decade a lot, i.e. a cold neutron source (CNS). (author)

Improvement of Cycle Dependent Core Model for NPP Simulator

International Nuclear Information System (INIS)

Song, J. S.; Koo, B. S.; Kim, H. Y. and others

2003-11-01

The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations

Improvement of Cycle Dependent Core Model for NPP Simulator

Energy Technology Data Exchange (ETDEWEB)

Song, J. S.; Koo, B. S.; Kim, H. Y. and others

2003-11-15

The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations.

BWR power oscillation evaluation methodologies in core design

International Nuclear Information System (INIS)

Hotta, Akitoshi

1995-01-01

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

Edge and core dynamics in harness

International Nuclear Information System (INIS)

Ball, R.

2007-01-01

Resistive kink oscillations in tokamak plasmas are usually treated as core localized events, yet there there are several mechanisms by which they may interact with the edge dynamics. This suggests that we may regulate edge oscillatory behaviour, or ELMs, by harnessing the natural or contrived sawtooth period and amplitude. In this work I investigate core-edge oscillatory entrainment through direct propagation of heat pulses, inductive coupling, and global higher order resonance effects. In the core of auxiliary heated tokamak plasmas the ineluctable rhythm of slow buildup and rapid conversion of potential energy governs electron and heat radial transport. The growth phase of the sawtooth is accompanied by significant reconnection, then during the collapse the temperature and density in the core fall dramatically. There is evidence from experiments in reversed field pinch devices that ensuing energy fluxes can affect flow shear and confinement at the edge. The basis for this study is the dynamical (BDS) model for edge plasma behavior that was derived from electrostatic resistive MHD equations. The BDS model reflects the major qualitative features of edge dynamics that have been observed, such as L-H transitions and associated ELMs, hysteresis, and spontaneous reversal of poloidal shear flow. Under poorly dissipative conditions the transient behavior of the model can exhibit period-doubling, blue-sky, homoclinic, and other exotic bifurcations. Thus we might ask questions such as: Is it possible to mode-lock the edge dynamics to the core sawteeth? Can we induce, or prevent, a change in direction of shear flow? What about MHD effects? Is core-edge communication one way or is there some feedback? In the simplest prototype for coupled core-edge dynamics I model the sawtooth crash as a periodic power input to the edge potential energy reservoir. This is effected by coupling the BDS model to the dynamical system u = u(1 - u 2 - x 2 ) - ω s x, x = x(1-u 2 -x 2 ) + ω s u

Description of low-lying states in odd-odd deformed nuclei taking account of the coupling with core rotations and vibrations. 1

International Nuclear Information System (INIS)

Kvasil, J.; Hrivnacova, I.; Nesterenko, V.O.

1990-01-01

The microscopic approach for description of low-lyinig states in deformed odd-odd nuclei is formulated as a generalization of the quasiparticle-phonon model (QPM) with including the rotational degrees of freedom and n-p interaction between external nucleons into the QPM. In comparison with other models, the approach proposed includes all three the most important effects coupling with rotational and vibrational degrees of freedom of doubly-even core and p-n interaction mentioned above even treates them on the microscopic base. 36 refs

Core polarization and Coulomb displacement energies

International Nuclear Information System (INIS)

Shlomo, S.; Love, W.G.

1982-01-01

The contributions of core polarization terms (other than the Auerbach-Kahana-Weneser (AKW) effect) to Coulomb displacement energies of mirror nuclei near A = 16 and A = 40 are examined within the particle-vibration coupling model. The parameters of the model are determined using updated data on the locations and strengths of multipole core excitations. In the absence of relevant data an energy-weighted sum rule (EWSR) is exploited. Taking into account multipole excitations up to L = 5 and subtracting the contributions which are due to short-range correlations, significant contributions (1-3%) to ΔEsub(c) are found. These corrections arise from particle coupling to low-lying collective states (long-range correlations). The implications of these results on the Coulomb energy problem are discussed. (Auth.)

Safety and core design of large liquid-metal cooled fast breeder reactors

Science.gov (United States)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

Improved core-edge tokamak transport simulations with the CORSICA 2 code

International Nuclear Information System (INIS)

Tarditi, A.; Cohen, R.H.; Crotinger, J.A.

1996-01-01

The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented

Density functional theory studies on the structures and electronic communication of meso-ferrocenylporphyrins: long range orbital coupling via porphyrin core.

Science.gov (United States)

Zhang, Lijuan; Qi, Dongdong; Zhang, Yuexing; Bian, Yongzhong; Jiang, Jianzhuang

2011-02-01

The molecular and electronic structures together with the electronic absorption spectra of a series of metal free meso-ferrocenylporphyrins, namely 5-ferrocenylporphyrin (1), 5,10-diferrocenylporphyrin (2), 5,15-diferrocenylporphyrin (3), 5,10,15-triferrocenylporphyrin (4), and 5,10,15,20-tetraferrocenylporphyrin (5) have been studied with the density functional theory (DFT) and time-dependent density functional theory (TD-DFT) methods. For the purpose of comparative studies, metal free porphyrin without any ferrocenyl group (0) and isolated ferrocene (6) were also calculated. The effects of the number and position of meso-attached ferrocenyl substituents on their molecular and electronic structures, atomic charges, molecular orbitals, and electronic absorption spectra of 1-5 were systematically investigated. The orbital coupling is investigated in detail, explaining well the long range coupling of ferrocenyl substituents connected via porphyrin core and the systematic change in the electronic absorption spectra of porphyrin compounds. Copyright © 2010 Elsevier Inc. All rights reserved.

The in-core experimental program at the MIT Research Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

Analysis on working pressure selection of ACME integral test facility

International Nuclear Information System (INIS)

Chen Lian; Chang Huajian; Li Yuquan; Ye Zishen; Qin Benke

2011-01-01

An integral effects test facility, advanced core cooling mechanism experiment facility (ACME) was designed to verify the performance of the passive safety system and validate its safety analysis codes of a pressurized water reactor power plant. Three test facilities for AP1000 design were introduced and review was given. The problems resulted from the different working pressures of its test facilities were analyzed. Then a detailed description was presented on the working pressure selection of ACME facility as well as its characteristics. And the approach of establishing desired testing initial condition was discussed. The selected 9.3 MPa working pressure covered almost all important passive safety system enables the ACME to simulate the LOCAs with the same pressure and property similitude as the prototype. It's expected that the ACME design would be an advanced core cooling integral test facility design. (authors)

Facile one-step hydrothermal synthesis toward strongly coupled TiO{sub 2}/graphene quantum dots photocatalysts for efficient hydrogen evolution

Energy Technology Data Exchange (ETDEWEB)

Min, Shixiong, E-mail: sxmin@nun.edu.cn [School of Chemistry and Chemical Engineering, Beifang University of Nationalities, Yinchuan, 750021, Ningxia Province (China); Hou, Jianhua; Lei, Yonggang; Ma, Xiaohua [School of Chemistry and Chemical Engineering, Beifang University of Nationalities, Yinchuan, 750021, Ningxia Province (China); Lu, Gongxuan [State Key Laboratory for Oxo Synthesis and Selective Oxidation, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou, 730000 (China)

2017-02-28

Highlights: • TiO{sub 2}/GQDs composites were prepared by a facile one-step hydrothermal method. • GQDs were strongly coupled onto the surface of TiO{sub 2} nanoparticles by this method. • The TiO{sub 2}/GQDs showed enhanced light absorption and charge separation efficiency. • The TiO{sub 2}/GQDs exhibited higher photocatalytic H{sub 2} evolution activity than pure TiO{sub 2}. • GQDs play synergistic roles by acting as both photosensitizer and electron acceptor. - Abstract: The coupling of semiconductor photocatalysts with graphene quantum dots (GQDs) has been proven to be an effective strategy to enhance the photocatalytic and photoelectrical conversion performances of the resulted composites; however, the preparation of semiconductor/GQDs composites usually involves several time-inefficient and tedious post-treatment steps. Herein, we present a facile one-step hydrothermal route for the preparation of GQDs coupled TiO{sub 2} (TiO{sub 2}/GQDs) photocatalysts using 1,3,6-trinitropyrene (TNP) as the sole precursor of GQDs. During the hydrothermal process, TNP molecules undergo an intramolecular fusion to form GQDs, which simultaneously decorate on the surface of TiO{sub 2} nanoparticles, leading to a strong surface interaction between the two components. The effective coupling of GQDs on TiO{sub 2} can effectively extend the light absorption of the TiO{sub 2} to visible region and enhance the charge separation efficiency of TiO{sub 2}/GQDs composites as a result of GQDs acting as a photosensitizer and an excellent electron acceptor. These key advances make the TiO{sub 2}/GQDs photocatalyst highly active towards the H{sub 2} evolution reaction, resulting in 7 and 3 times higher H{sub 2} evolution rate and photocurrent response at optimal GQDs content than TiO{sub 2} alone, respectively. This study provides a new methodology for the development of high-performance GQDs modified semiconductor photocatalysts for energy conversion applications.

The GUINEVERE-project: the first zero-power fast lead reactor coupled to a 14 MeV neutron generator (GENEPI)

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

The GUINEVERE project is an European project in the framework of FP6 IP-EUROTRANS. The IP-EUROTRANS project aims at addressing the main issues for ADS development in the framework of partitioning and transmutation for nuclear waste volume and radio toxicity reduction. The GUINEVERE-project is carried out in the context of domain 2 of IP-EUROTRANS, ECATS, devoted to specific experiments for the coupling of an accelerator, a target and a subcritical core. These experiments should provide an answer to the questions of on-line reactivity monitoring, sub-criticality determination and operational procedures (loading, start-up, shut-down) in an ADS by 2009-2010. During the definition of the experimental programme ECATS, it was judged that there was a strong need for a European managed experiment in the line of the FP5 MUSE-project. Reanalyzing the outcome of MUSE, two points were left open for significant improvement. To validate the methodology for reactivity monitoring, a continuous beam is needed, which was not present in the MUSE-project. In the definition of the MUSE-project, from the beginning a strong request was made for a lead core in order to have representative conditions of a lead-cooled ADS which was only partially answered by the MUSE-programme. Therefore, there is a need for a lead fast critical facility connected to a continuous beam accelerator. Since such a programme/installation is not present at the European nor at the international level, SCK-CEN has proposed to use a modified VENUS critical facility located at its Mol-site and to couple it to a modified GENEPI deuteron accelerator (used in MUSE) working in current mode delivering 14 MeV neutrons by bombardment of deuterons on a tritium-target: the GUINEVERE-project (Generator of Uninterrupted Intense NEutrons at the lead VEnus REactor). This proposal was formally accepted by the Governing Council of IP-Eurotrans in December 2006. This project represents a close collaboration between SCK-CEN, CEA and

Automated Core Design

International Nuclear Information System (INIS)

Kobayashi, Yoko; Aiyoshi, Eitaro

2005-01-01

Multistate searching methods are a subfield of distributed artificial intelligence that aims to provide both principles for construction of complex systems involving multiple states and mechanisms for coordination of independent agents' actions. This paper proposes a multistate searching algorithm with reinforcement learning for the automatic core design of a boiling water reactor. The characteristics of this algorithm are that the coupling structure and the coupling operation suitable for the assigned problem are assumed and an optimal solution is obtained by mutual interference in multistate transitions using multiagents. Calculations in an actual plant confirmed that the proposed algorithm increased the convergence ability of the optimization process

VVER-1000 coolant transient benchmark. Phase 1 (V1000CT-1). Vol. 3: summary results of exercise 2 on coupled 3-D kinetics/core thermal-hydraulics

International Nuclear Information System (INIS)

2007-01-01

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as current applications. (authors) Recently developed best-estimate computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present volume is a follow-up to the first two volumes. While the first described the specification of the benchmark, the second presented the results of the first exercise that identified the key parameters and important issues concerning the thermal-hydraulic system modelling of the simulated transient caused by the switching on of a main coolant pump when the other three were in operation. Volume 3 summarises the results for Exercise 2 of the benchmark that identifies the key parameters and important issues concerning the 3-D neutron kinetics modelling of the simulated transient. These studies are based on an experiment that was conducted by Bulgarian and Russian engineers during the plant-commissioning phase at the VVER-1000 Kozloduy Unit 6. The final volume will soon be published, completing Phase 1 of this study. (authors)

Irradiation facilities in JRR-3M

International Nuclear Information System (INIS)

Ohtomo, Akitoshi; Sigemoto, Masamitsu; Takahashi, Hidetake

1992-01-01

Irradiation facilities have been installed in the upgraded JRR-3 (JRR-3M) in Japan Atomic Energy Research Institute (JAERI). There are hydraulic rabbit facilities (HR), pneumatic rabbit facilities (PN), neutron activation analysis facility (PN3), uniform irradiation facility (SI), rotating irradiation facility and capsule irradiation facilities to carry out the neutron irradiation in the JRR-3M. These facilities are operated using a process control computer system to centerize the process information. Some of the characteristics for the facilities were satisfactorily measured at the same time of reactor performance test in 1990. During reactor operation, some of the tests are continued to confirm the basic characteristics on facilities, for example, PN3 was confirmed to have enough performance for activation analysis. Measurement of neutron flux at all irradiation positions has been carried out for the equilibrium core. (author)

Ministry of ordinance determining the technical standard concerning atomic energy facilities for power generation

International Nuclear Information System (INIS)

1985-01-01

The ministerial ordinance provides for the technical standards for the power generation of nuclear facilities; i.e., electric power facilities generating electricity with nuclear energy for motive power, according to the Electricity Enterprises Act. The contents are as follows: protection against fires, aseismatic design, radiation protective barriers, structural protection for sitings, reactor installation, safety measures, materials and structures, safety valves, pressure resistance tests, reactor core, radiation shields, reactor cooling, emergency core cooling system, facility equipment, alarm system, reactor control system, reactor control room, fuel storage facility, fuel handling facility, ventilation equipment, radioactive contamination prevention, radioactive waste management facility, reactor containment facility, and so on. (Kubozono, M.)

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)

Full Core Multiphysics Simulation with Offline Mesh Deformation

Energy Technology Data Exchange (ETDEWEB)

Merzari, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Obabko, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Jain, Rajeev [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay [Argonne National Lab. (ANL), Argonne, IL (United States); Solberg, Jerome [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferencz, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-12-21

In this report, building on previous reports issued in FY13 we describe our continued efforts to integrate thermal/hydraulics, neutronics, and structural mechanics modeling codes to perform coupled analysis of a representative fast sodium-cooled reactor core. The focus of the present report is a full core simulation with off-line mesh deformation.

PWR core safety analysis with 3-dimensional methods

International Nuclear Information System (INIS)

Gensler, A.; Kühnel, K.; Kuch, S.

2015-01-01

Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

Large scale reflood test with cylindrical core test facility (CCTF). Core I. FY 1979 tests

International Nuclear Information System (INIS)

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

1982-03-01

This report presents the results of analysis of the data obtained in the CCTF Core I test series (19 tests) in FY. 1979 as an interim report. The Analysis of the test results showed that: (1) The present safety evaluation model on the reflood phenomena during LOCA conservatively represents the phenomena observed in the tests except for the downcomer thermohydrodynamic behavior. (2) The downcomer liquid level rose slowly and it took long time for the water to reach a terminal level or the spill-over level. It was presume that such a results was due to an overly conservative selection of the ECC flow rate. This presumption will be checked against a future test result for an increased flow rate. The loop-seal-water filling test was unsuccessful due to a premature power shutdown by the core protection circuit. The test will be conducted again. The tests to be performed in the future are summerized. Tests for investigation of the refill phenomena were also proposed. (author)

Improved core protection calculator system algorithm

International Nuclear Information System (INIS)

Yoon, Tae Young; Park, Young Ho; In, Wang Kee; Bae, Jong Sik; Baeg, Seung Yeob

2009-01-01

Core Protection Calculator System (CPCS) is a digitized core protection system which provides core protection functions based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels which adapted a two out of four trip logic. CPCS algorithm improvement for the newly designed core protection calculator system, RCOPS (Reactor COre Protection System), is described in this paper. New features include the improvement of DNBR algorithm for thermal margin, the addition of pre trip alarm generation for auxiliary trip function, VOPT (Variable Over Power Trip) prevention during RPCS (Reactor Power Cutback System) actuation and the improvement of CEA (Control Element Assembly) signal checking algorithm. To verify the improved CPCS algorithm, CPCS algorithm verification tests, 'Module Test' and 'Unit Test', would be performed on RCOPS single channel facility. It is expected that the improved CPCS algorithm will increase DNBR margin and enhance the plant availability by reducing unnecessary reactor trips

Critical experiments of JMTRC MEU cores

International Nuclear Information System (INIS)

Nagaoka, Y.; Takeda, K.; Shimakawa, S.; Koike, S.; Oyamada, R.

1984-01-01

The JMTRC, the critical facility of the Japan Materials Testing Reactor (JMTR), went critical on August 29, 1983, with 14 medium enriched uranium (MEU, 45%) fuel elements. Experiments are now being carried out to measure the change in various reactor characteristics between the previous HEU core and the new MEU fueled core. This paper describes the results obtained thus far on critical mass, excess reactivity, control rod worths and flux distribution, including preliminary neutronics calculations for the experiments using the SRAC code. (author)

Micron-scale variations in coupled δ13C-N abundance core-rim traverses in octahedral diamonds: insights into the processes and sources of episodic diamond formation beneath the Siberian craton

NARCIS (Netherlands)

Wiggers de Vries, D.F.; Bulanova, G.; de Corte, K.; Pearson, D.G.; Davies, G.R.

2013-01-01

The internal structure and growth history of six macro-diamonds from kimberlite pipes in Yakutia (Russia) were investigated with cathodoluminescence imaging and coupled carbon isotope and nitrogen abundance analyses along detailed core to rim traverses. The diamonds are characterised by octahedral

Novel Crosstalk Measurement Method for Multi-Core Fiber Fan-In/Fan-Out Devices

DEFF Research Database (Denmark)

Ye, Feihong; Ono, Hirotaka; Abe, Yoshiteru

2016-01-01

We propose a new crosstalk measurement method for multi-core fiber fan-in/fan-out devices utilizing the Fresnel reflection. Compared with the traditional method using core-to-core coupling between a multi-core fiber and a single-mode fiber, the proposed method has the advantages of high reliability...

CCTF CORE I test results

International Nuclear Information System (INIS)

Murao, Yoshio; Sudoh, Takashi; Akimoto, Hajime; Iguchi, Tadashi; Sugimoto, Jun; Fujiki, Kazuo; Hirano, Kenmei

1982-07-01

This report presents the results of the following CCTF CORE I tests conducted in FY. 1980. (1) Multi-dimensional effect test, (2) Evaluation model test, (3) FLECHT coupling test. On the first test, one-dimensional treatment of the core thermohydrodynamics was discussed. On the second and third tests, the test results were compared with the results calculated by the evaluation model codes and the results of the corresponding FLECHT-SET test (Run 2714B), respectively. The work was performed under contracts with the Atomic Energy Bureau of Science and Technology Agency of Japan. (author)

Zero-bias tunneling anomaly at a vortex core

International Nuclear Information System (INIS)

Overhauser, A.W.; Daemen, L.L.

1989-01-01

The sharp peak in the tunneling conductance at a vortex core, reported by Hess et al. in NbSe 2 , is attributed to self-energy corrections of the normal electrons (in the core) caused by their coupling to excitations of the superconducting region (outside the core). The shape of the zero-bias anomaly is reproduced without benefit from adjustable parameters, though the predicted size is a little too large. If the critical currents in the superconducting region (outside the core) are recognized by letting the excitation density (at zero energy) be finite, then a perfect fit can be obtained

[caCORE: core architecture of bioinformation on cancer research in America].

Science.gov (United States)

Gao, Qin; Zhang, Yan-lei; Xie, Zhi-yun; Zhang, Qi-peng; Hu, Zhang-zhi

2006-04-18

A critical factor in the advancement of biomedical research is the ease with which data can be integrated, redistributed and analyzed both within and across domains. This paper summarizes the Biomedical Information Core Infrastructure built by National Cancer Institute Center for Bioinformatics in America (NCICB). The main product from the Core Infrastructure is caCORE--cancer Common Ontologic Reference Environment, which is the infrastructure backbone supporting data management and application development at NCICB. The paper explains the structure and function of caCORE: (1) Enterprise Vocabulary Services (EVS). They provide controlled vocabulary, dictionary and thesaurus services, and EVS produces the NCI Thesaurus and the NCI Metathesaurus; (2) The Cancer Data Standards Repository (caDSR). It provides a metadata registry for common data elements. (3) Cancer Bioinformatics Infrastructure Objects (caBIO). They provide Java, Simple Object Access Protocol and HTTP-XML application programming interfaces. The vision for caCORE is to provide a common data management framework that will support the consistency, clarity, and comparability of biomedical research data and information. In addition to providing facilities for data management and redistribution, caCORE helps solve problems of data integration. All NCICB-developed caCORE components are distributed under open-source licenses that support unrestricted usage by both non-profit and commercial entities, and caCORE has laid the foundation for a number of scientific and clinical applications. Based on it, the paper expounds caCORE-base applications simply in several NCI projects, of which one is CMAP (Cancer Molecular Analysis Project), and the other is caBIG (Cancer Biomedical Informatics Grid). In the end, the paper also gives good prospects of caCORE, and while caCORE was born out of the needs of the cancer research community, it is intended to serve as a general resource. Cancer research has historically

33-GVA interrupter test facility

International Nuclear Information System (INIS)

Parsons, W.M.; Honig, E.M.; Warren, R.W.

1979-01-01

The use of commercial ac circuit breakers for dc switching operations requires that they be evaluated to determine their dc limitations. Two 2.4-GVA facilities have been constructed and used for this purpose at LASL during the last several years. In response to the increased demand on switching technology, a 33-GVA facility has been constructed. Novel features incorporated into this facility include (1) separate capacitive and cryogenic inductive energy storage systems, (2) fiber-optic controls and optically-coupled data links, and (3) digital data acquisition systems. Facility details and planned tests on an experimental rod-array vacuum interrupter are presented

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)

IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility

International Nuclear Information System (INIS)

Dietze, Klaus; Klippel, Henk Th.; Koning, Arjan; Jacqmin, Robert

2003-01-01

1 - Description: The STEK-experiments have been performed to check neutron data of the most important reactor materials, especially of fission product nuclides, fuel isotopes and structural materials. The measured central reactivity worths (CRW) of small samples were compared with calculated values. These C/E-ratios have been used then for data corrections or in adjustment procedures. The reactors STEK (ECN Petten/ Netherlands) was a fast-thermal coupled facility of zero power. The annular thermal drivers were filled by fuel assemblies and moderated by water. The inner insertion lattices were loaded with pellets of fuel and other materials producing the fast neutron flux. The characteristics of the neutron and adjoint spectra were obtained by special arrangements of these pellets in unit cells. In this way, a hard or soft neutron spectrum or a special energy behavior of the adjoint function could be reached. The samples were moved by means of tubes to the central position (pile-oscillation technique). The original information about the facility and measurements is compiled in RCN-209, ECN-10 The 5 STEK configurations cover a broad energy range due to their increasing softness. The experiments are very valuable because of the extensive program of sample reactivity measurements with many fission product nuclides important in reactor burn-up calculations. At first, analyses of the experiments have been performed in Petten. Newer analyses were done later in Cadarache / CEA France using the European scheme for reactor calculation JEF-2.2 / ECCO / ERANOS (see Note Techniques and JEF/DOC-746). Furthermore, re-analyses were performed in O-arai / JNC Japan with the JNC standard route JENDL-3.2 / SLAROM / CITATION / PERKY. Results obtained with both code systems and different data evaluations (JEF-2.2 and JENDL-3.2) are compared in JEF/DOC-861. It contains the following documents: 31 Reports, 2 publications, 5 JEF documents, 4 conferences. 2 - Related or auxiliary programs

Lead coolant test facility systems design, thermal hydraulic analysis and cost estimate

Energy Technology Data Exchange (ETDEWEB)

Khericha, Soli, E-mail: slk2@inel.gov [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Harvego, Edwin; Svoboda, John; Evans, Robert [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Dalling, Ryan [ExxonMobil Gas and Power Marketing, Houston, TX 77069 (United States)

2012-01-15

The Idaho National Laboratory prepared a preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed below: Bullet Develop and demonstrate feasibility of submerged heat exchanger. Bullet Develop and demonstrate open-lattice flow in electrically heated core. Bullet Develop and demonstrate chemistry control. Bullet Demonstrate safe operation. Bullet Provision for future testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimated. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 Degree-Sign C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

CMT scaling analysis and distortion evaluation in passive integral test facility

International Nuclear Information System (INIS)

Deng Chengcheng; Qin Benke; Wang Han; Chang Huajian

2013-01-01

Core makeup tank (CMT) is the crucial device of AP1000 passive core cooling system, and reasonable scaling analysis of CMT plays a key role in the design of passive integral test facilities. H2TS method was used to perform scaling analysis for both circulating mode and draining mode of CMT. And then, the similarity criteria for CMT important processes were applied in the CMT scaling design of the ACME (advanced core-cooling mechanism experiment) facility now being built in China. Furthermore, the scaling distortion results of CMT characteristic Ⅱ groups of ACME were calculated. At last, the reason of scaling distortion was analyzed and the distortion evaluation was conducted for ACME facility. The dominant processes of CMT circulating mode can be adequately simulated in the ACME facility, but the steam condensation process during CMT draining is not well preserved because the excessive CMT mass leads to more energy to be absorbed by cold metal. However, comprehensive analysis indicates that the ACME facility with high-pressure simulation scheme is able to properly represent CMT's important phenomena and processes of prototype nuclear plant. (authors)

America's Next Great Ship: Space Launch System Core Stage Transitioning from Design to Manufacturing

Science.gov (United States)

Birkenstock, Benjamin; Kauer, Roy

2014-01-01

The Space Launch System (SLS) Program is essential to achieving the Nation's and NASA's goal of human exploration and scientific investigation of the solar system. As a multi-element program with emphasis on safety, affordability, and sustainability, SLS is becoming America's next great ship of exploration. The SLS Core Stage includes avionics, main propulsion system, pressure vessels, thrust vector control, and structures. Boeing manufactures and assembles the SLS core stage at the Michoud Assembly Facility (MAF) in New Orleans, LA, a historical production center for Saturn V and Space Shuttle programs. As the transition from design to manufacturing progresses, the importance of a well-executed manufacturing, assembly, and operation (MA&O) plan is crucial to meeting performance objectives. Boeing employs classic techniques such as critical path analysis and facility requirements definition as well as innovative approaches such as Constraint Based Scheduling (CBS) and Cirtical Chain Project Management (CCPM) theory to provide a comprehensive suite of project management tools to manage the health of the baseline plan on both a macro (overall project) and micro level (factory areas). These tools coordinate data from multiple business systems and provide a robust network to support Material & Capacity Requirements Planning (MRP/CRP) and priorities. Coupled with these tools and a highly skilled workforce, Boeing is orchestrating the parallel buildup of five major sub assemblies throughout the factory. Boeing and NASA are transforming MAF to host state of the art processes, equipment and tooling, the most prominent of which is the Vertical Assembly Center (VAC), the largest weld tool in the world. In concert, a global supply chain is delivering a range of structural elements and component parts necessary to enable an on-time delivery of the integrated Core Stage. SLS is on plan to launch humanity into the next phase of space exploration.

Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

1998-03-01

The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation

Directory of Open Access Journals (Sweden)

Avramović Ivana

2007-01-01

Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.

Critical experiment study on uranyl nitrate solution experiment facility

International Nuclear Information System (INIS)

Zhu Qingfu; Shi Yongqian; Wang Jinrong

2005-01-01

The Uranyl Nitrate Solution Experiment Facility was constructed for the research on nuclear criticality safety. In this paper, the configuration of the facility is introduced; a series of critical experiments on uranyl nitrate solution is described later, which were performed for various uranium concentrations under different conditions, i.e. with or without neutron absorbers in the core and with or without water-reflector outside the core. Critical volume and the minimum 235U critical mass for different uranium concentrations are presented. Finally, theoretical analysis is made on the experimental results. (authors)

Nonlocal elasticity tensors in dislocation and disclination cores

International Nuclear Information System (INIS)

Taupin, V.; Gbemou, K.; Fressengeas, C.; Capolungo, L.

2017-01-01

We introduced nonlocal elastic constitutive laws for crystals containing defects such as dislocations and disclinations. Additionally, the pointwise elastic moduli tensors adequately reflect the elastic response of defect-free regions by relating stresses to strains and couple-stresses to curvatures, elastic cross-moduli tensors relating strains to couple-stresses and curvatures to stresses within convolution integrals are derived from a nonlocal analysis of strains and curvatures in the defects cores. Sufficient conditions are derived for positive-definiteness of the resulting free energy, and stability of elastic solutions is ensured. The elastic stress/couple stress fields associated with prescribed dislocation/disclination density distributions and solving the momentum and moment of momentum balance equations in periodic media are determined by using a Fast Fourier Transform spectral method. Here, the convoluted cross-moduli bring the following results: (i) Nonlocal stresses and couple stresses oppose their local counterparts in the defects core regions, playing the role of restoring forces and possibly ensuring spatio-temporal stability of the simulated defects, (ii) The couple stress fields are strongly affected by nonlocality. Such effects favor the stability of the simulated grain boundaries and allow investigating their elastic interactions with extrinsic defects, (iii) Driving forces inducing grain growth or refinement derive from the self-stress and couple stress fields of grain boundaries in nanocrystalline configurations.

The SHIP facility at CERN

Science.gov (United States)

De Lellis, Giovanni

2016-04-01

Searches for new physics with accelerators are being performed at the LHC, looking for high massive particles coupled to matter with ordinary strength. A new experimental facility meant to search for very weakly coupled particles in the few GeV mass domain has been recently proposed. The existence of such particles, foreseen in different theoretical models beyond the Standard Model, is largely unexplored from the experimental point of view. A beam dump facility, built at CERN in the north area, using 400 GeV protons is a copious factory of charmed hadrons and could be used to probe the existence of such particles. The beam dump is also an ideal source of tau neutrinos, the less known particle in the Standard Model. In particular, tau anti-neutrinos have not been directly observed so far. We report the physics potential of such an experiment and outline the performances of a detector operating at the same facility for the search for the τ → μμμ decay.

Constraints on The Coupled Thermal Evolution of the Earth's Core and Mantle, The Age of The Inner Core, And The Origin of the 186Os/188Os Core(?) Signal in Plume-Derived Lavas

Science.gov (United States)

Lassiter, J. C.

2005-12-01

Thermal and chemical interaction between the core and mantle has played a critical role in the thermal and chemical evolution of the Earth's interior. Outer core convection is driven by core cooling and inner core crystallization. Core/mantle heat transfer also buffers mantle potential temperature, resulting in slower rates of mantle cooling (~50-100 K/Ga) than would be predicted from the discrepancy between current rates of surface heat loss (~44 TW) and internal radioactive heat production (~20 TW). Core/mantle heat transfer may also generate thermal mantle plumes responsible for ocean island volcanic chains such as the Hawaiian Islands. Several studies suggest that mantle plumes, in addition to transporting heat from the core/mantle boundary, also carry a chemical signature of core/mantle interaction. Elevated 186Os/188Os ratios in lavas from Hawaii, Gorgona, and in the 2.8 Ga Kostomuksha komatiites have been interpreted as reflecting incorporation of an outer core component with high time-integrated Pt/Os and Re/Os ( Brandon et al., 1999, 2003; Puchtel et al., 2005). Preferential partitioning of Os relative to Re and Pt into the inner core during inner core growth may generate elevated Re/Os and Pt/Os ratios in the residual outer core. Because of the long half-life of 190Pt (the parent of 186Os, t1/2 = 489 Ga), an elevated 186Os/188Os outer core signature in plume lavas requires that inner core crystallization began early in Earth history, most likely prior to 3.5 Ga. This in turn requires low time-averaged core/mantle heat flow (<~2.5 TW) or large quantities of heat-producing elements in the core. Core/mantle heat flow may be estimated using boundary-layer theory, by measuring the heat transported in mantle plumes, by estimating the heat transported along the outer core adiabat, or by comparing the rates of heat production, surface heat loss, and secular cooling of the mantle. All of these independent methods suggest time-averaged core/mantle heat flow of ~5

Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

International Nuclear Information System (INIS)

Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

2005-01-01

Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data

International Nuclear Information System (INIS)

Akimoto, Hajime; Iguchi, Tadashi; Murao, Yoshio

1985-01-01

In the reactor safety assessment during reflood phase of a PWR-LOCA, it is assumed implicitly that the core thermal hydraulic behavior is evaluated by the one-dimensional model with an average power rod. In order to assess the applicability of the one-dimensional treatment, integral tests were performed with various core radial power profiles using the Cylindrical Core Test Facility (CCTF) whose core includes about 2,000 heater rods. The CCTF results confirm that the core radial power profile has weak effect on the thermal hydraulic behavior in the primary system except core. It is also confirmed that the core differential pressure in the axial direction is predicted by the one-dimensional core model with an average power rod even in the case with a steep radial power profile in the core. Even though the core heat transfer coefficient is dependent on the core radial power profile, it is found that the error of the peak clad surface temperature calculation is less than 15 K using the one-dimensional model in the CCTF tests. The CCTF results support the one-dimensional treatment assumed in the reactor safety assessment. (author)

Near-Edge X-ray Absorption Fine Structure within Multilevel Coupled Cluster Theory.

Science.gov (United States)

Myhre, Rolf H; Coriani, Sonia; Koch, Henrik

2016-06-14

Core excited states are challenging to calculate, mainly because they are embedded in a manifold of high-energy valence-excited states. However, their locality makes their determination ideal for local correlation methods. In this paper, we demonstrate the performance of multilevel coupled cluster theory in computing core spectra both within the core-valence separated and the asymmetric Lanczos implementations of coupled cluster linear response theory. We also propose a visualization tool to analyze the excitations using the difference between the ground-state and excited-state electron densities.

The Swedish facility for boron neutron capture therapy

Energy Technology Data Exchange (ETDEWEB)

Skoeld, K.; Capala, J. [Studsvik Medical AB (Sweden); Kierkegaard, J.; Haakansson, R. [Studsvik Nuclear AB (Sweden); Gudowska, I. [Karolinska Institute (Sweden)

2000-10-01

A BNCT (Boron Neutron Capture Therapy) facility has been constructed at the R2-0 reactor at Studsvik, Sweden. R2-0 is a 1 MW, open core, pool reactor. The reactor core is suspended on a movable tower and can be positioned anywhere in the pool. The BNCT facility includes two adjacent, parallel filter/moderator configurations and the reactor core is positioned in front of any of them as appropriate. One of the resulting neutron beams has been optimized for clinical irradiations with a filter/moderator system that allows easy variation of the neutron spectrum from the thermal to the epithermal energy range and with an extended collimator for convenient patient positioning. The other beam has been designed for radiobiological research and is equipped with a heavy water moderator and a large irradiation cavity with a uniform field of thermal neutrons. (author)

The Swedish facility for boron neutron capture therapy

International Nuclear Information System (INIS)

Skoeld, K.; Capala, J.; Kierkegaard, J.; Haakansson, R.; Gudowska, I.

2000-01-01

A BNCT (Boron Neutron Capture Therapy) facility has been constructed at the R2-0 reactor at Studsvik, Sweden. R2-0 is a 1 MW, open core, pool reactor. The reactor core is suspended on a movable tower and can be positioned anywhere in the pool. The BNCT facility includes two adjacent, parallel filter/moderator configurations and the reactor core is positioned in front of any of them as appropriate. One of the resulting neutron beams has been optimized for clinical irradiations with a filter/moderator system that allows easy variation of the neutron spectrum from the thermal to the epithermal energy range and with an extended collimator for convenient patient positioning. The other beam has been designed for radiobiological research and is equipped with a heavy water moderator and a large irradiation cavity with a uniform field of thermal neutrons. (author)

Use of Ground Penetrating Radar at the FAA's National Airport Pavement Test Facility

Science.gov (United States)

Injun, Song

2015-04-01

The Federal Aviation Administration (FAA) in the United States has used a ground-coupled Ground Penetrating Radar (GPR) at the National Airport Pavement Test Facility (NAPTF) since 2005. One of the primary objectives of the testing at the facility is to provide full-scale pavement response and failure information for use in airplane landing gear design and configuration studies. During the traffic testing at the facility, a GSSI GPR system was used to develop new procedures for monitoring Hot Mix Asphalt (HMA) pavement density changes that is directly related to pavement failure. After reviewing current setups for data acquisition software and procedures for identifying different pavement layers, dielectric constant and pavement thickness were selected as dominant parameters controlling HMA properties provided by GPR. A new methodology showing HMA density changes in terms of dielectric constant variations, called dielectric sweep test, was developed and applied in full-scale pavement test. The dielectric constant changes were successfully monitored with increasing airplane traffic numbers. The changes were compared to pavement performance data (permanent deformation). The measured dielectric constants based on the known HMA thicknesses were also compared with computed dielectric constants using an equation from ASTM D4748-98 Standard Test Method for Determining the Thickness of Bound Pavement Layers Using Short-Pulse Radar. Six inches diameter cylindrical cores were taken after construction and traffic testing for the HMA layer bulk specific gravity. The measured bulk specific gravity was also compared to monitor HMA density changes caused by aircraft traffic conditions. Additionally this presentation will review the applications of the FAA's ground-coupled GPR on embedded rebar identification in concrete pavement, sewer pipes in soil, and gage identifications in 3D plots.

Qualification of the coupled RELAP5/PANTHER/COBRA code package for licensing applications

International Nuclear Information System (INIS)

Schneidesch, C.R.; Zhang Jinzhao

2004-01-01

A coupled thermal hydraulics-neutronics code package has been developed at Tractebel Engineering (TE), in which the best-estimate thermal-hydraulic system code, RELAP5/mod2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via the dynamic data exchange interface, TALINK. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by the sub-channel thermal-hydraulic analysis code COBRA-3C. The package provides the capability to accurately simulate the key physical phenomena in nuclear power plant accidents with strong asymmetric behaviours and system-core interactions. This paper presents the TE coupled code package and focuses on the methodology followed for qualifying it for licensing applications. The qualification of the coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric PWR accidents with strong core-system interactions

Fourth intercomparison of personal dosemeters used in US Department of Energy accelerator facilities

CERN Document Server

Stewart, R D; Otto, T; Loesch, R M

2000-01-01

Personal neutron dosemeters from seven US Department of Energy (DOE) laboratories were mailed to the European Laboratory for Particle Physics (CERN) and irradiated using the well-characterised CERN reference radiation facility (CERF). Neutron dose equivalents determined using the DOE personal dosemeters have been compared to the reference dose equivalent as determined using a tissue-equivalent proportional counter (TEPC). In the 0.5 to 5 mSv dose equivalent range, the comparison of results suggests that the neutron personal dosemeters in use at DOE facilities are capable of estimating dose equivalents for high energy neutrons to within a factor of at least 2 or 3. If a field-specific calibration factor is used to correct the dose equivalent responses, the agreement with the reference dose equivalent for these dosemeters can be improved to better than about 25 to 65at is decoupled from the core in /sup 183,185/Au, becomes the 3/2[532] state (h9/2 parentage) strongly coupled in the doubly-odd /sup 184/Au nucleu...

Role of Absorbing Nanocrystal Cores in Soft Photonic Crystals: A Spectroscopy and SANS Study.

Science.gov (United States)

Rauh, Astrid; Carl, Nico; Schweins, Ralf; Karg, Matthias

2018-01-23

Periodic superstructures of plasmonic nanoparticles have attracted significant interest because they can support coupled plasmonic modes, making them interesting for plasmonic lasing, metamaterials, and as light-management structures in thin-film optoelectronic devices. We have recently shown that noble metal hydrogel core-shell colloids allow for the fabrication of highly ordered 2-dimensional plasmonic lattices that show surface lattice resonances as the result of plasmonic/diffractive coupling (Volk, K.; Fitzgerald, J. P. S.; Ruckdeschel, P.; Retsch, M.; König, T. A. F.; Karg, M. Reversible Tuning of Visible Wavelength Surface Lattice Resonances in Self-Assembled Hybrid Monolayers. Adv. Optical Mater. 2017, 5, 1600971, DOI: 10.1002/adom.201600971). In the present work, we study the photonic properties and structure of 3-dimensional crystalline superstructures of gold hydrogel core-shell colloids and their pitted counterparts without gold cores. We use far-field extinction spectroscopy to investigate the optical response of these superstructures. Narrow Bragg peaks are measured, independently of the presence or absence of the gold cores. All crystals show a significant reduction in low-wavelength scattering. This leads to a significant enhancement of the plasmonic properties of the samples prepared from gold-nanoparticle-containing core-shell colloids. Plasmonic/diffractive coupling is not evident, which we mostly attribute to the relatively small size of the gold cores limiting the effective coupling strength. Small-angle neutron scattering is applied to study the crystal structure. Bragg peaks of several orders clearly assignable to an fcc arrangement of the particles are observed for all crystalline samples in a broad range of volume fractions. Our results indicate that the nanocrystal cores do not influence the overall crystallization behavior or the crystal structure. These are important prerequisites for future studies on photonic materials built from core

Field and laboratory investigations of coring-induced damage in core recovered from Marker Bed 139 at the waste isolation pilot plant underground facility

International Nuclear Information System (INIS)

Holcomb, D.J.; Zeuch, D.H.; Morin, K.; Hardy, R.; Tormey, T.V.

1995-09-01

A combined laboratory and field investigation was carried out to determine the extent of coring-induced damage done to samples cored from Marker Bed 139 at the WIPP site. Coring-induced damage, if present, has the potential to significantly change the properties of the material used for laboratory testing relative to the in situ material properties, resulting in misleading conclusions. In particular, connected, crack-like damage could make the permeability of cored samples orders of magnitude greater than the in situ permeabilities. Our approach compared in situ velocity and resistivity measurements with laboratory measurements of the same properties. Differences between in situ and laboratory results could be attributed to differences in the porosity due to cracks. The question of the origin of the changes could not be answered directly from the results of the measurements. Pre-existing cracks, held closed by the in situ stress, could open when the core was cut free, or new cracks could be generated by coring-induced damage. We used core from closely spaced boreholes at three orientations (0 degree, ±45 degrees relative to vertical) to address the origin of cracks. The absolute orientation of pre-existing cracks would be constant, independent of the borehole orientation. In contrast, cracks induced by coring were expected to show an orientation dependent on that of the source borehole

Field and laboratory investigations of coring-induced damage in core recovered from Marker Bed 139 at the waste isolation pilot plant underground facility

Energy Technology Data Exchange (ETDEWEB)

Holcomb, D.J.; Zeuch, D.H.; Morin, K.; Hardy, R.; Tormey, T.V.

1995-09-01

A combined laboratory and field investigation was carried out to determine the extent of coring-induced damage done to samples cored from Marker Bed 139 at the WIPP site. Coring-induced damage, if present, has the potential to significantly change the properties of the material used for laboratory testing relative to the in situ material properties, resulting in misleading conclusions. In particular, connected, crack-like damage could make the permeability of cored samples orders of magnitude greater than the in situ permeabilities. Our approach compared in situ velocity and resistivity measurements with laboratory measurements of the same properties. Differences between in situ and laboratory results could be attributed to differences in the porosity due to cracks. The question of the origin of the changes could not be answered directly from the results of the measurements. Pre-existing cracks, held closed by the in situ stress, could open when the core was cut free, or new cracks could be generated by coring-induced damage. We used core from closely spaced boreholes at three orientations (0{degree}, {plus_minus}45{degrees} relative to vertical) to address the origin of cracks. The absolute orientation of pre-existing cracks would be constant, independent of the borehole orientation. In contrast, cracks induced by coring were expected to show an orientation dependent on that of the source borehole.

Imaging quality evaluation method of pixel coupled electro-optical imaging system

Science.gov (United States)

He, Xu; Yuan, Li; Jin, Chunqi; Zhang, Xiaohui

2017-09-01

With advancements in high-resolution imaging optical fiber bundle fabrication technology, traditional photoelectric imaging system have become ;flexible; with greatly reduced volume and weight. However, traditional image quality evaluation models are limited by the coupling discrete sampling effect of fiber-optic image bundles and charge-coupled device (CCD) pixels. This limitation substantially complicates the design, optimization, assembly, and evaluation image quality of the coupled discrete sampling imaging system. Based on the transfer process of grayscale cosine distribution optical signal in the fiber-optic image bundle and CCD, a mathematical model of coupled modulation transfer function (coupled-MTF) is established. This model can be used as a basis for following studies on the convergence and periodically oscillating characteristics of the function. We also propose the concept of the average coupled-MTF, which is consistent with the definition of traditional MTF. Based on this concept, the relationships among core distance, core layer radius, and average coupled-MTF are investigated.

Network coupling via a current-limiting throttle with a high-Tc superconductor core

International Nuclear Information System (INIS)

Bochenek, E.; Fischer, R.; Lampen, U.; Voigt, H.

1989-01-01

A current-limiting concept is tested by means of a choke with a current-responsive inductivity for linking three-phase current supplies. The choke has a core of a material with a high transition point T c . In the case of nominal current, the core is superconductive and keeps the resulting inductance of the choke low by shield currents. In the case of overload, the core passes into the normal conductive state due to the increased magnetic field of the winding. The resulting inductance of the choke rises and, in doing so, effects current limitation. (orig.) [de

Performance testing of a mixed TRIGA core

Energy Technology Data Exchange (ETDEWEB)

Schumacher, R F; Godsey, T A; Feltz, D E; Randall, J D [Texas A and M University (United States)

1974-07-01

The major operational problem experienced by the Nuclear Science Center Reactor at Texas A and M University is full burnup. With two shift operation caused by the high utilization of the facility, the reactor is operated more than 100 megawatt days per year. The solution chosen for this problem was conversion to FLIP fuel. Since funds were not available to load an entire FLIP core, a mixed core comprised of approximately one third FLIP fuel located in the central region was designed. The design core was loaded and went critical on July 1, 1973. The results of the following measurements on the mixed core are presented: Determination of Rod worths; Measurement of Reactivity Effects; Determination of Flux values; Measurement of Fuel temperatures; Preliminary Fuel Burnup Rate; Pulsing Calibration. (author)

Mineralogy in the Waste Isolation Pilot Plant (WIPP) facility stratigraphic horizon

International Nuclear Information System (INIS)

Stein, C.L.

1985-09-01

Forty-six samples were selected for this study from two cores, one extending 50 ft up through the roof of the WIPP facility and the other penetrating 50 ft below the facility floor. These samples, selected from approximately every other foot of core length, represent the major lithologies present in the immediate vicinity of the WIPP facility horizon: ''clean'' halite, polyhalitic halite, argillaceous halite, and mixed polyhalitic-argillaceous halite. Samples were analyzed for non-NaCl mineralogy by determining weight percents of water- and EDTA-insoluble residues, which were then identified by x-ray diffraction. In general, WIPP halite contains at most 5 wt % non-NaCl residue. The major mineral constituents are quartz, magnesite, anhydrite, gypsum, polyhalite, and clays. Results of this study confirm that, in previous descriptions of WIPP core, trace mineral quantities have been visually overestimated by approximately an order of magnitude. 9 refs., 5 figs., 5 tabs

Effective particle magnetic moment of multi-core particles

Energy Technology Data Exchange (ETDEWEB)

Ahrentorp, Fredrik; Astalan, Andrea; Blomgren, Jakob; Jonasson, Christian [Acreo Swedish ICT AB, Arvid Hedvalls backe 4, SE-411 33 Göteborg (Sweden); Wetterskog, Erik; Svedlindh, Peter [Department of Engineering Sciences, Uppsala University, Box 534, SE-751 21 Uppsala (Sweden); Lak, Aidin; Ludwig, Frank [Institute of Electrical Measurement and Fundamental Electrical Engineering, TU Braunschweig, D‐38106 Braunschweig Germany (Germany); IJzendoorn, Leo J. van [Department of Applied Physics, Eindhoven University of Technology, 5600 MB Eindhoven (Netherlands); Westphal, Fritz; Grüttner, Cordula [Micromod Partikeltechnologie GmbH, D ‐18119 Rostock (Germany); Gehrke, Nicole [nanoPET Pharma GmbH, D ‐10115 Berlin Germany (Germany); Gustafsson, Stefan; Olsson, Eva [Department of Applied Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Johansson, Christer, E-mail: christer.johansson@acreo.se [Acreo Swedish ICT AB, Arvid Hedvalls backe 4, SE-411 33 Göteborg (Sweden)

2015-04-15

In this study we investigate the magnetic behavior of magnetic multi-core particles and the differences in the magnetic properties of multi-core and single-core nanoparticles and correlate the results with the nanostructure of the different particles as determined from transmission electron microscopy (TEM). We also investigate how the effective particle magnetic moment is coupled to the individual moments of the single-domain nanocrystals by using different measurement techniques: DC magnetometry, AC susceptometry, dynamic light scattering and TEM. We have studied two magnetic multi-core particle systems – BNF Starch from Micromod with a median particle diameter of 100 nm and FeraSpin R from nanoPET with a median particle diameter of 70 nm – and one single-core particle system – SHP25 from Ocean NanoTech with a median particle core diameter of 25 nm.

Effective particle magnetic moment of multi-core particles

International Nuclear Information System (INIS)

Ahrentorp, Fredrik; Astalan, Andrea; Blomgren, Jakob; Jonasson, Christian; Wetterskog, Erik; Svedlindh, Peter; Lak, Aidin; Ludwig, Frank; IJzendoorn, Leo J. van; Westphal, Fritz; Grüttner, Cordula; Gehrke, Nicole; Gustafsson, Stefan; Olsson, Eva; Johansson, Christer

2015-01-01

In this study we investigate the magnetic behavior of magnetic multi-core particles and the differences in the magnetic properties of multi-core and single-core nanoparticles and correlate the results with the nanostructure of the different particles as determined from transmission electron microscopy (TEM). We also investigate how the effective particle magnetic moment is coupled to the individual moments of the single-domain nanocrystals by using different measurement techniques: DC magnetometry, AC susceptometry, dynamic light scattering and TEM. We have studied two magnetic multi-core particle systems – BNF Starch from Micromod with a median particle diameter of 100 nm and FeraSpin R from nanoPET with a median particle diameter of 70 nm – and one single-core particle system – SHP25 from Ocean NanoTech with a median particle core diameter of 25 nm

Effective particle magnetic moment of multi-core particles

Science.gov (United States)

Ahrentorp, Fredrik; Astalan, Andrea; Blomgren, Jakob; Jonasson, Christian; Wetterskog, Erik; Svedlindh, Peter; Lak, Aidin; Ludwig, Frank; van IJzendoorn, Leo J.; Westphal, Fritz; Grüttner, Cordula; Gehrke, Nicole; Gustafsson, Stefan; Olsson, Eva; Johansson, Christer

2015-04-01

In this study we investigate the magnetic behavior of magnetic multi-core particles and the differences in the magnetic properties of multi-core and single-core nanoparticles and correlate the results with the nanostructure of the different particles as determined from transmission electron microscopy (TEM). We also investigate how the effective particle magnetic moment is coupled to the individual moments of the single-domain nanocrystals by using different measurement techniques: DC magnetometry, AC susceptometry, dynamic light scattering and TEM. We have studied two magnetic multi-core particle systems - BNF Starch from Micromod with a median particle diameter of 100 nm and FeraSpin R from nanoPET with a median particle diameter of 70 nm - and one single-core particle system - SHP25 from Ocean NanoTech with a median particle core diameter of 25 nm.

Ultra-short pulse delivery at high average power with low-loss hollow core fibers coupled to TRUMPF's TruMicro laser platforms for industrial applications

Science.gov (United States)

Baumbach, S.; Pricking, S.; Overbuschmann, J.; Nutsch, S.; Kleinbauer, J.; Gebs, R.; Tan, C.; Scelle, R.; Kahmann, M.; Budnicki, A.; Sutter, D. H.; Killi, A.

2017-02-01

Multi-megawatt ultrafast laser systems at micrometer wavelength are commonly used for material processing applications, including ablation, cutting and drilling of various materials or cleaving of display glass with excellent quality. There is a need for flexible and efficient beam guidance, avoiding free space propagation of light between the laser head and the processing unit. Solid core step index fibers are only feasible for delivering laser pulses with peak powers in the kW-regime due to the optical damage threshold in bulk silica. In contrast, hollow core fibers are capable of guiding ultra-short laser pulses with orders of magnitude higher peak powers. This is possible since a micro-structured cladding confines the light within the hollow core and therefore minimizes the spatial overlap between silica and the electro-magnetic field. We report on recent results of single-mode ultra-short pulse delivery over several meters in a lowloss hollow core fiber packaged with industrial connectors. TRUMPF's ultrafast TruMicro laser platforms equipped with advanced temperature control and precisely engineered opto-mechanical components provide excellent position and pointing stability. They are thus perfectly suited for passive coupling of ultra-short laser pulses into hollow core fibers. Neither active beam launching components nor beam trackers are necessary for a reliable beam delivery in a space and cost saving packaging. Long term tests with weeks of stable operation, excellent beam quality and an overall transmission efficiency of above 85 percent even at high average power confirm the reliability for industrial applications.

Imaging of High-Z doped, Imploded Capsule Cores

Science.gov (United States)

Prisbrey, Shon T.; Edwards, M. John; Suter, Larry J.

2006-10-01

The ability to correctly ascertain the shape of imploded fusion capsules is critical to be able to achieve the spherical symmetry needed to maximize the energy yield of proposed fusion experiments for the National Ignition Facility. Implosion of the capsule creates a hot, dense core. The introduction of a high-Z dopant into the gas-filled core of the capsule increases the amount of bremsstrahlung radiation produced in the core and should make the imaging of the imploded core easier. Images of the imploded core can then be analyzed to ascertain the symmetry of the implosion. We calculate that the addition of Ne gas into a deuterium gas core will increase the amount of radiation emission while preserving the surrogacy of the radiation and hydrodynamics in the indirect drive NIF hohlraum in the proposed cryogenic hohlraums. The increased emission will more easily enable measurement of asymmetries and tuning of the implosion.

On-line generation of core monitoring power distribution in the SCOMS couppled with core design code

International Nuclear Information System (INIS)

Lee, K. B.; Kim, K. K.; In, W. K.; Ji, S. K.; Jang, M. H.

2002-01-01

The paper provides the description of the methodology and main program module of power distribution calculation of SCOMS(SMART COre Monitoring System). The simulation results of the SMART core using the developed SCOMS are included. The planar radial peaking factor(Fxy) is relatively high in SMART core because control banks are inserted to the core at normal operation. If the conventional core monitoring method is adapted to SMART, highly skewed planar radial peaking factor Fxy yields an excessive conservatism and reduces the operation margin. In addition to this, the error of the core monitoring would be enlarged and thus operating margin would be degraded, because it is impossible to precalculate the core monitoring constants for all the control banks configurations taking into account the operation history in the design stage. To get rid of these drawbacks in the conventional power distribution calculation methodology, new methodology to calculate the three dimensional power distribution is developed. Core monitoring constants are calculated with the core design code (MASTER) which is on-line coupled with SCOMS. Three dimensional (3D) power distribution and the several peaking factors are calculated using the in-core detector signals and core monitoring constant provided at real time. Developed methodology is applied to the SMART core and the various core states are simulated. Based on the simulation results, it is founded that the three dimensional peaking factor to calculate the Linear Power Density and the pseudo hot-pin axial power distribution to calculate the Departure Nucleate Boiling Ratio show the more conservative values than those of the best-estimated core design code, and SCOMS adapted developed methodology can secures the more operation margin than the conventional methodology

Titanium dioxide@polypyrrole core-shell nanowires for all solid-state flexible supercapacitors

Science.gov (United States)

Yu, Minghao; Zeng, Yinxiang; Zhang, Chong; Lu, Xihong; Zeng, Chenghui; Yao, Chenzhong; Yang, Yangyi; Tong, Yexiang

2013-10-01

Herein, we developed a facile two-step process to synthesize TiO2@PPy core-shell nanowires (NWs) on carbon cloth and reported their improved electrochemical performance for flexible supercapacitors (SCs). The fabricated solid-state SC device based on TiO2@PPy core-shell NWs not only has excellent flexibility, but also exhibits remarkable electrochemical performance.Herein, we developed a facile two-step process to synthesize TiO2@PPy core-shell nanowires (NWs) on carbon cloth and reported their improved electrochemical performance for flexible supercapacitors (SCs). The fabricated solid-state SC device based on TiO2@PPy core-shell NWs not only has excellent flexibility, but also exhibits remarkable electrochemical performance. Electronic supplementary information (ESI) available: Experimental details, XRD pattern, FT-IR absorption spectrum and CV curves of TiO2@PPy NWs, and SEM images of the PPy. See DOI: 10.1039/c3nr03578f

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II forced feed reflood tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1987-01-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase of a PWR-LOCA. It was revealed in the previous Slab Core Test Facility (SCTF) Core-II test results that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. In order to separately evaluate the effect of the radial power (Q) distribution itself and the effect of the radial temperature (T) distribution, four tests were performed with steep Q and T, flat Q and T, steep Q and flat T, and flat Q and steep T. Based on the test results, it was concluded that the radial temperature distribution which accompanied the radial power distribution was the dominant factor of the two-dimensional thermal-hydraulic behavior in the core during the initial period. Selected data from these four tests are also presented in this report. Some data from Test S2-12 (steep Q, T) were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

Neutron spectrometric methods for core inventory verification in research reactors

International Nuclear Information System (INIS)

Ellinger, A.; Filges, U.; Hansen, W.; Knorr, J.; Schneider, R.

2002-01-01

In consequence of the Non-Proliferation Treaty safeguards, inspections are periodically made in nuclear facilities by the IAEA and the EURATOM Safeguards Directorate. The inspection methods are permanently improved. Therefore, the Core Inventory Verification method is being developed as an indirect method for the verification of the core inventory and to check the declared operation of research reactors

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

International Nuclear Information System (INIS)

Tombakoglu, M.; Cecen, Y.

2001-01-01

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

Design of Test Facility to Evaluate Boric Acid Precipitation Following a LOCA

Energy Technology Data Exchange (ETDEWEB)

Suh, Jeong-Kwan; Song, Yong-Jae [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The U.S.NRC has identified a concern that debris associated with generic safety issue (GSI) - 191 may affect the potential precipitation of boric acid due to one or more of the following phenomena: - Reducing mass transport (i.e. mixing) between the core and the lower plenum (should debris accumulate at the core inlet) - Reduced lower plenum volume (should debris settle in the lower plenum), and, - Increased potential for boric acid precipitation (BAP) in the core (should debris accumulate in suspension in the core) To address these BAP issues, KHNP is planning to conduct validation tests by constructing a BAP test facility. This paper describes the design of test facility to evaluate BAP following a LOCA. The design of BAP test facility has been developed by KHNP. To design the test facility, test requirements and success criteria were established, and scaling analysis of power-to-volume method, Ishii-Kataoka method, and hierarchical two-tiered method were investigated. The test section is composed of two fuel assemblies with half of full of prototypic FA height. All the fuel rods are heated by the electric power supplier. The BAP tests in the presence of debris, buffering agents, and boron will be performed following the test matrix.

Phase 1 sampling and analysis plan for the 304 Concretion Facility closure activities

International Nuclear Information System (INIS)

Adler, J.G.

1994-01-01

This document provides guidance for the initial (Phase 1) sampling and analysis activities associated with the proposed Resource Conservation and Recovery Act of 1976 (RCRA) clean closure of the 304 Concretion Facility. Over its service life, the 304 Concretion Facility housed the pilot plants associated with cladding uranium cores, was used to store engineering equipment and product chemicals, was used to treat low-level radioactive mixed waste, recyclable scrap uranium generated during nuclear fuel fabrication, and uranium-titanium alloy chips, and was used for the repackaging of spent halogenated solvents from the nuclear fuels manufacturing process. The strategy for clean closure of the 304 Concretion Facility is to decontaminate, sample (Phase 1 sampling), and evaluate results. If the evaluation indicates that a limited area requires additional decontamination for clean closure, the limited area will be decontaminated, resampled (Phase 2 sampling), and the result evaluated. If the evaluation indicates that the constituents of concern are below action levels, the facility will be clean closed. Or, if the evaluation indicates that the constituents of concern are present above action levels, the condition of the facility will be evaluated and appropriate action taken. There are a total of 37 sampling locations comprising 12 concrete core, 1 concrete chip, 9 soil, 11 wipe, and 4 asphalt core sampling locations. Analysis for inorganics and volatile organics will be performed on the concrete core and soil samples. Separate concrete core samples will be required for the inorganic and volatile organic analysis (VOA). Analysis for inorganics only will be performed on the concrete chip, wipe, and asphalt samples

Vessel coolant mass depletion during a 5% SBLOCA in the Semiscale Mod-2C facility

International Nuclear Information System (INIS)

Shaw, R.A.; Loomis, G.G.

1985-01-01

Experimental results are presented from two 5% small-break loss-of-coolant accident (SBLOCA) simulations in the Semiscale Mod-2C facility. In performing the simulated 5% SBLOCAs, boundary conditions scaled from a pressurized water reactor (PWR) were used. The experiment was run with initial conditions typical of a PWR (15.6 MPa pressure and 35 K core differential temperature). The Mod-2C facility represents the state-of-the-art in small facilities scaled from PWRs. Phenomena which occurred during the transient included: primary fluid saturation (change from subcooled to saturated blowdown), break uncovery (a centerline break was simulated), condensation-induced liquid hold-up in the steam generator primary tubes, pump suction liquid seal formation and core level depression with resulting core rod temperature excursion, pump suction liquid seal clearance, loop fluid mass redistribution, and gradual core rewet. The influence of core bypass flow is also discussed. 11 refs., 13 figs

Status report on SHARP coupling framework.

Energy Technology Data Exchange (ETDEWEB)

Caceres, A.; Tautges, T. J.; Lottes, J.; Fischer, P.; Rabiti, C.; Smith, M. A.; Siegel, A.; Yang, W. S.; Palmiotti, G.

2008-05-30

This report presents the software engineering effort under way at ANL towards a comprehensive integrated computational framework (SHARP) for high fidelity simulations of sodium cooled fast reactors. The primary objective of this framework is to provide accurate and flexible analysis tools to nuclear reactor designers by simulating multiphysics phenomena happening in complex reactor geometries. Ideally, the coupling among different physics modules (such as neutronics, thermal-hydraulics, and structural mechanics) needs to be tight to preserve the accuracy achieved in each module. However, fast reactor cores in steady state mode represent a special case where weak coupling between neutronics and thermal-hydraulics is usually adequate. Our framework design allows for both options. Another requirement for SHARP framework has been to implement various coupling algorithms that are parallel and scalable to large scale since nuclear reactor core simulations are among the most memory and computationally intensive, requiring the use of leadership-class petascale platforms. This report details our progress toward achieving these goals. Specifically, we demonstrate coupling independently developed parallel codes in a manner that does not compromise performance or portability, while minimizing the impact on individual developers. This year, our focus has been on developing a lightweight and loosely coupled framework targeted at UNIC (our neutronics code) and Nek (our thermal hydraulics code). However, the framework design is not limited to just using these two codes.

Space Launch System, Core Stage, Structural Test Design and Implementation

Science.gov (United States)

Shaughnessy, Ray

2017-01-01

As part of the National Aeronautics and Space Administration's (NASA) Space Launch System (SLS) Program, engineers at NASA's Marshall Space Flight Center (MSFC) in Huntsville, Alabama are working to design, develop and implement the SLS Core Stage structural testing. The SLS will have the capability to return humans to the Moon and beyond and its first launch is scheduled for December of 2017. The SLS Core Stage consist of five major elements; Forward Skirt, Liquid Oxygen (LOX) tank, Intertank (IT), Liquid Hydrogen (LH2) tank and the Engine Section (ES). Structural Test Articles (STA) for each of these elements are being designed and produced by Boeing at Michoud Assembly Facility located in New Orleans, La. The structural test for the Core Stage STAs (LH2, LOX, IT and ES) are to be conducted by the MSFC Test Laboratory. Additionally, the MSFC Test Laboratory manages the Structural Test Equipment (STE) design and development to support the STAs. It was decided early (April 2012) in the project life that the LH2 and LOX tank STAs would require new test stands and the Engine Section and Intertank would be tested in existing facilities. This decision impacted schedules immediately because the new facilities would require Construction of Facilities (C of F) funds that require congressional approval and long lead times. The Engine Section and Intertank structural test are to be conducted in existing facilities which will limit lead times required to support the first launch of SLS. With a SLS launch date of December, 2017 Boeing had a need date for testing to be complete by September of 2017 to support flight certification requirements. The test facilities were required to be ready by October of 2016 to support test article delivery. The race was on to get the stands ready before Test Article delivery and meet the test complete date of September 2017. This paper documents the past and current design and development phases and the supporting processes, tools, and

Considerations in setting up and planning a graft processing facility.

Science.gov (United States)

Koh, Mickey B C

2017-12-01

The graft processing facility forms one of the core components of a clinical haematopoietic stem cell transplant program. The quality of a graft is instrumental in leading to consistent and reproducible outcomes of engraftment and other parameters. As such, meticulous planning and consideration is required and will include core elements including physical design and clinical correlates. The successful running of such a facility depends on an overarching quality program and adherence to local and international regulatory guidelines. Copyright © 2017 King Faisal Specialist Hospital & Research Centre. Published by Elsevier B.V. All rights reserved.

Validation of reactor core protection system

International Nuclear Information System (INIS)

Lee, Sang-Hoon; Bae, Jong-Sik; Baeg, Seung-Yeob; Cho, Chang-Ho; Kim, Chang-Ho; Kim, Sung-Ho; Kim, Hang-Bae; In, Wang-Kee; Park, Young-Ho

2008-01-01

Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

Coupled fast-thermal system at the RB, masters thesis; Spregnuti brzo-termicki sistem na reaktoru RB, magistarski rad

Energy Technology Data Exchange (ETDEWEB)

Pesic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1984-05-15

Coupled fast-thermal system at the RB reactor was formed owing to availability of highly enriched fuel. This paper deals with reactor parameters calculations and measurements of coupled core taking into account safety constraints. Validity of applied calculation methods was confirmed. The following parameters were analyzed: critical height of the core; reactivity dependent on heavy water level in the core; fast neutron spectrum in the fast region channel; spatial distribution of thermal. epithermal and fat neutrons in the fast region channel; reactivity of safety rods; neutron and gamma absorption doses in the center of the coupled core.

Automated reactivity anomaly surveillance in the Fast Flux Test Facility

International Nuclear Information System (INIS)

Knutson, B.J.; Harris, R.A.; Honeyman, D.J.; Shook, A.T.; Krohn, C.N.

1985-01-01

The automated technique for monitoring core reactivity during power operation used at the Fast Flux Test Facility (FFTF) is described. This technique relies on comparing predicted to measured rod positions to detect any anomalous (or unpredicted) core reactivity changes. It is implemented on the Plant Data System (PDS) computer and, thus, provides rapid indication of any abnormal core conditions. The prediction algorithms use thermal-hydraulic, control rod position and neutron flux sensor information to predict the core reactivity state

Mechanical strength calculation of the disk type windings with elastic couplings by the finite element method

International Nuclear Information System (INIS)

Sivkova, G.N.; Spirchenko, Yu.V.; Chvartatskij, P.V.

1981-01-01

Stressed-deformed state of toroidal field coils of the disc type with elastic couplings of the tokamaks has been investigated with provision for the effect of the central core pliability by means of the two-dimensional version of the finite element method. Numerical solution of the finite element method is performed by means of the ES 1040 computer according to the computer code permitting taking account of boundary conditions of elastic support. The calculation has been performed using as the example the project of T-20 facility coil of the disc type. Consideration of pliability of the central core of the facility inductor is accomplished by the introduction of additional rigidities to the complete matrix of rigidity. Scheme of the structure distretization includes 141 units, 211 elements. The accuracy of solution depends on the reduction accuracy of the volume load to unit forces and on the number of finite elements. Analysis of the solution convergence is performed by the comparison of solutions obtained for three different schemes of the disk discretization without regard for the inductor pliability. The comparative analysis of the results shows that transfer epures for all the three discretization versions practically coincide and stresses differ not more than by 10%. On the whole the above investigation has demonstrated good convergence of the problem solution [ru

The SHiP facility at CERN

CERN Document Server

AUTHOR|(SzGeCERN)452451

2016-01-01

Searches for new physics with accelerators are being performed at the LHC, looking for high massive particles coupled to matter with ordinary strength. A new experimental facility meant to search for very weakly coupled particles in the few GeV mass domain has been recently proposed. The existence of such particles, foreseen in dierent theoretical models beyond the Standard Model, is largely unexplored from the experimental point of view. A beam dump facility, built at CERN in the north area, using 400 GeV protons is a copious factory of charmed hadrons and could be used to probe the existence of such particles. The beam dump is also an ideal source of tau neutrinos, the less known particle in the Standard Model. In particular, tau anti-neutrinos have not been directly observed so far. We report the physics potential of such an experiment and outline the performances of a detector operating at the same facility for the search for the tau --> mu mu mu decay

Coupling of THALES and FROST using MPI Method

International Nuclear Information System (INIS)

Park, Jin Woo; Ryu, Seok Hee; Jung, Chan Do; Jung, Jee Hoon; Um, Kil Sup; Lee, Jae Il

2013-01-01

This paper presents the coupling method between THALES and FROST and the simulation results with the coupled code system. In this study, subchannel analysis code THALES and transient fuel performance code FROST were coupled using MPI method as the first stage of the development of the multi-dimensional safety analysis methodology. As a part of the validation, the CEA ejection accident was simulated using the coupled THALES-FROST code and the results were compared with the ShinKori 3 and 4 FSAR. Comparison results revealed that CHASER using MPI method predicts fuel temperatures and heat flux quantitatively well. Thus it was confirmed that the THALES and FROST are properly coupled. In near future, ASTRA, multi-dimensional core neutron kinetics code, will be linked to THALESFROST code for the detailed three-dimensional CEA ejection analysis. The current safety analysis methodology for a CEA ejection accident based on numerous conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KNF is developing the multi-dimensional safety analysis methodology to enhance the consequences of the CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, subchannel analysis code THALES, and transient fuel performance analysis code FROST are being coupled using message passing interface(MPI). For the first step, THALES and FROST are coupled and tested

Constraints on the coupled thermal evolution of the Earth's core and mantle, the age of the inner core, and the origin of the 186Os/188Os “core signal” in plume-derived lavas

Science.gov (United States)

Lassiter, J. C.

2006-10-01

The possibility that some mantle plumes may carry a geochemical signature of core/mantle interaction has rightly generated considerable interest and attention in recent years. Correlated 186Os- 187Os enrichments in some plume-derived lavas (Hawaii, Gorgona, Kostomuksha) have been interpreted as deriving from an outer core with elevated Pt/Os and Re/Os ratios due to the solidification of the Earth's inner core (c.f., [A.D. Brandon, R.J. Walker, The debate over core-mantle interaction, Earth Planet. Sci. Lett. 232 (2005) 211-225.] and references therein). Conclusive identification of a "core signal" in plume-derived lavas would profoundly influence our understanding of mantle convection and evolution. This paper reevaluates the Os-isotope evidence for core/mantle interaction by examining other geochemical constraints on core/mantle interaction, geophysical constraints on the thermal evolution of the outer core, and geochemical and cosmochemical constraints on the abundance of heat-producing elements in the core. Additional study of metal/silicate and sulfide/silicate partitioning of K, Pb, and other trace elements is needed to more tightly constrain the likely starting composition of the Earth's core. However, available data suggest that the observed 186Os enrichments in Hawaiian and other plume-derived lavas are unlikely to derive from core/mantle interaction. 1) Core/mantle interaction sufficient to produce the observed 186Os enrichments would likely have significant effects on other tracers such as Pb- and W-isotopes that are not observed. 2) Significant partitioning of K or other heat-producing elements into the core would produce a "core depletion" pattern in the Silicate Earth very different from that observed. 3) In the absence of heat-producing elements in the core, core/mantle heat flow of ˜ 6-15 TW estimated from several independent geophysical constraints suggests an inner core age (< ˜ 2.5 Ga) too young for the outer core to have developed a significant

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

Experimental possibilities and fast neutron dose map of the fast neutron fields at the RB reactor facility

International Nuclear Information System (INIS)

Sokcic-Kostic, M.; Pesic, M.; Antic, D.; Ninkovic, M.

1993-01-01

The RB is an unshielded, zero power nuclear facility with natural and enriched uranium fuel (2% and 80%) and D 2 O as moderator. It is possible to create different configurations of non-reflected and partially reflected critical systems and to make experiments in the fields of thermal neutrons. The fields of fast neutrons with 'softened' fission spectrum are made by modifying the system: modified experimental fuel channel EFC, coupled fast-thermal system in two configurations CFTS-1 and CFTS-2, coupled fast-thermal core HERBE. The intermediate and fast neutron absorbed doses in fast neutron fields are given. In first configuration of RB reactor it was almost impossible to perform dosimetric and other experiments. By creating these fields, with in our circumstances available fuel elements, the possibilities for different experiments are greatly improved. Now we can irradiate food samples, soil samples, electronic devices, study material properties, perform various dosimetry experiments, etc. (1 tab.)

Optimizing a three-element core design for the Advanced Neutron Source Reactor

International Nuclear Information System (INIS)

West, C.D.

1995-01-01

Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ''no new inventions'' rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high flux reactors, containing highly enriched U silicide particles. DOE commissioned a study of the use of medium- or low-enriched U; a three-element core design was studied as a means to provide extra volume to accommodate the additional U compound required when the fissionable 235 U has to be diluted with 238 U to reduce the enrichment. This paper describes the design and optimization of that three-element core

Efficient tungsten oxide/bismuth oxyiodide core/shell photoanode for photoelectrochemical water splitting

Science.gov (United States)

Ma, Haipeng; Zhang, Jing; Liu, Zhifeng

2017-11-01

The novel WO3 nanorods (NRs)/BiOI core/shell structure composite is used as an efficient photoanode applied in photoelectrochemical (PEC) water splitting for the first time. It is synthesized via facile hydrothermal method and, successive ionic layer adsorption and reaction (SILAR) process. This facile synthesis route can achieve uniform WO3/BiOI core/shell composite nanostructures and obtain varied BiOI morphologies simultaneously. The WO3 NRs/BiOI-20 composite exhibits enhanced PEC activity compared to pristine WO3 with a photocurrent density of 0.79 mA cm-2 measured at 0.8 V vs. RHE under AM 1.5G. This excellent performance benefits from the broader absorption spectrum and suppressed electron-hole recombination. This novel core/shell composite may provide insight in developing more efficient solar driven photoelectrodes.

A core management system for JRR-3

International Nuclear Information System (INIS)

Soyama, Kazuhiko; Tsuruta, Harumichi; Ichikawa, Hiroki; Nemoto, Hiroyuki.

1991-05-01

Japan Research Reactor No.3 (JRR-3) was upgraded to the thermal output with 20 MW by replacing the core, cooling system and utilization facilities. It is a water moderated and cooled, pool type reactor using 20% enriched U · Alx fuel. A core management system for JRR-3 has been made. This code system can manage of reactivity, power distribution and burn up in consideration of the position of control rod, fuel arrangement and operation pattern. This report is the user's manual of this code system. (author)

HYDRATE CORE DRILLING TESTS

Energy Technology Data Exchange (ETDEWEB)

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

2002-11-01

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

Neural networks within multi-core optic fibers.

Science.gov (United States)

Cohen, Eyal; Malka, Dror; Shemer, Amir; Shahmoon, Asaf; Zalevsky, Zeev; London, Michael

2016-07-07

Hardware implementation of artificial neural networks facilitates real-time parallel processing of massive data sets. Optical neural networks offer low-volume 3D connectivity together with large bandwidth and minimal heat production in contrast to electronic implementation. Here, we present a conceptual design for in-fiber optical neural networks. Neurons and synapses are realized as individual silica cores in a multi-core fiber. Optical signals are transferred transversely between cores by means of optical coupling. Pump driven amplification in erbium-doped cores mimics synaptic interactions. We simulated three-layered feed-forward neural networks and explored their capabilities. Simulations suggest that networks can differentiate between given inputs depending on specific configurations of amplification; this implies classification and learning capabilities. Finally, we tested experimentally our basic neuronal elements using fibers, couplers, and amplifiers, and demonstrated that this configuration implements a neuron-like function. Therefore, devices similar to our proposed multi-core fiber could potentially serve as building blocks for future large-scale small-volume optical artificial neural networks.

Core clamping device for a nuclear reactor

International Nuclear Information System (INIS)

Guenther, R.W.

1974-01-01

The core clamping device for a fast neutron reactor includes clamps to support the fuel zone against the pressure vessel. The clamps are arranged around the circumference of the core. They consist of torsion bars arranged parallel at some distance around the core with lever arms attached to the ends whose force is directed in the opposite direction, pressing against the wall of the pressure vessel. The lever arms and pressure plates also actuated by the ends of the torsion bars transfer the stress, the pressure plates acting upon the fuel elements or fuel assemblies. Coupling between the ends of the torsion bars and the pressure plates is achieved by end carrier plates directly attached to the torsion bars and radially movable. This clamping device follows the thermal expansions of the core, allows specific elements to be disengaged in sections and saves space between the core and the neutron reflectors. (DG) [de

Inner core tilt and polar motion

Science.gov (United States)

Dumberry, Mathieu; Bloxham, Jeremy

2002-11-01

A tilted inner core permits exchange of angular momentum between the core and the mantle through gravitational and pressure torques and, as a result, changes in the direction of Earth's axis of rotation with respect to the mantle. We have developed a model to calculate the amplitude of the polar motion that results from an equatorial torque at the inner core boundary which tilts the inner core out of alignment with the mantle. We specifically address the issue of the role of the inner core tilt in the decade polar motion known as the Markowitz wobble. We show that a decade polar motion of the same amplitude as the observed Markowitz wobble requires a torque of 1020 N m which tilts the inner core by 0.07 degrees. This result critically depends on the viscosity of the inner core; for a viscosity less than 5 × 1017 Pa s, larger torques are required. We investigate the possibility that a torque of 1020 N m with decadal periodicity can be produced by electromagnetic coupling between the inner core and torsional oscillations of the flow in the outer core. We demonstrate that a radial magnetic field at the inner core boundary of 3 to 4 mT is required to obtain a torque of such amplitude. The resulting polar motion is eccentric and polarized, in agreement with the observations. Our model suggests that equatorial torques at the inner core boundary might also excite the Chandler wobble, provided there exists a physical mechanism that can generate a large torque at a 14 month period.

Core homogenization method for pebble bed reactors

International Nuclear Information System (INIS)

Kulik, V.; Sanchez, R.

2005-01-01

This work presents a core homogenization scheme for treating a stochastic pebble bed loading in pebble bed reactors. The reactor core is decomposed into macro-domains that contain several pebble types characterized by different degrees of burnup. A stochastic description is introduced to account for pebble-to-pebble and pebble-to-helium interactions within a macro-domain as well as for interactions between macro-domains. Performance of the proposed method is tested for the PROTEUS and ASTRA critical reactor facilities. Numerical simulations accomplished with the APOLLO2 transport lattice code show good agreement with the experimental data for the PROTEUS reactor facility and with the TRIPOLI4 Monte Carlo simulations for the ASTRA reactor configuration. The difference between the proposed method and the traditional volume-averaged homogenization technique is negligible while only one type of fuel pebbles present in the system, but it grows rapidly with the level of pebble heterogeneity. (authors)

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.

Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 1. ZPPR analysis results

International Nuclear Information System (INIS)

Mantourov, Guennadi

2001-05-01

This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess of Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. The data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS and ZPPR fast reactor cores applying JNC core calculation code CITATION. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the first flight collision probability method and subgroup approach. Especially a converting program was written to transmit the prepared effective cross sections to JNC standard PDS files. Then the CITATION code was applied for 3-D XYZ neutronics calculations of BFS and ZPPR JUPITER experiments series cores. The effects of nuclear data library have been studied by comparing the former results based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for ZPPR-9, ZPPR-13A and ZPPR-17A cores are presented. The calculated correction factor in all cases was less than 1.0%. So the uncertainty in C value caused by possible errors in calculation of these corrections is expected to be less than 0.3% in case of ZPPR-13A and ZPPR-17A cores, and rather less for ZPPR-9 core. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC calculation route revealed a large enough discrepancy in k-eff for ZPPR-9 (about 0.6%) and ZPPR-17A (about 0.5%) cores. For BFS-62-1 and BFS-62-2 cores such analysis is in progress. Stretch cell models for both BFS cores were formed and cell calculations using FFCP code have started. Some results of cell calculations are presented. (author)

Thermal hydraulic modelling of the Mo and Iridium irradiation facilities of the RA10 reactor

International Nuclear Information System (INIS)

Gramajo, M.; García, J.; Marcel, C.P.

2013-01-01

The RA-10 reactor is a multipurpose, open pool research reactor. The core consists of a rectangular array of MTR type fuel. The produced thermal power is 30 MW which is extracted by the refrigeration system via an ascendant flow through the core. The core reflector is D 2 O contained in a watertight tank. The design of the reactor includes a number of out-core facilities which are meant to be used for industrial, medical and research purposes. Among all the facilities, the most important ones are the Molybdenum and Iridium ones which we modeled in this work. During the normal operation of the reactor, the manipulation and the on-line extraction of the irradiation facilities is foreseen. Therefore the study of the head loss during the normal operation as well as during the extraction maneuvers plays a relevant role in the design and safety analysis. In this work a CFD commercial code is use dto perform the calculations needed to guarantee the design requirements.In addition, a full detailed geometric model for both, the Molybdenum and Iridium facilities,is used to perform the required simulations. The obtained results allow to evaluating the thermal-hydraulic performance of the proposed facilities designs. (author)

Use of EBR-II as a principal fast breeder reactor irradiation test facility in the U.S

International Nuclear Information System (INIS)

Staker, R.G.; Seim, O.S.; Beck, W.N.; Golden, G.H.; Walters, L.C.

1975-01-01

The EBR-II as originally designed and operated by the Argonne National Laboratory was successful in demonstrating the operation of a sodium-cooled fast breeder power plant with a closed fuel reprocessing cycle. Subsequent operation has been as an experimental facility where thousands of irradiation tests have been performed. Conversion to this application entailed the design and fabrication of special irradiation subassemblies for in-core irradiations, additions to existing facilities for out-of-core irradiations, and additions to existing facilities for out-of-core experiments. Experimental subassemblies now constitute about one third of the core, and changes in the core configuration occur about monthly, requiring neutronic and thermal-hydraulics analyses and monitoring of the reactor dynamic behavior. The surveillance programs provided a wealth of information on irradiation induced swelling and creep, in-reactor fracture behavior, and the compatibility of materials with liquid sodium. (U.S.)

Enhancing the Relationship Adjustment of South Asian Canadian Couples Using a Systemic-Constructivist Approach to Couple Therapy.

Science.gov (United States)

Ahmad, Saunia; Reid, David W

2016-10-01

The effectiveness of systemic-constructivist couple therapy (SCCT) in improving the relationship adjustment of South Asian Canadian couples in ways that attend to their culture was evaluated. The SCCT interventions engage partners in reflexive processing of both their own and their partner's ways of construing, and the reciprocity between these two. A core change mechanism of SCCT, couple identity ("we-ness"), that connotes the ability for thinking and experiencing relationally, was coded from verbatim transcripts of partners' within-session dialogue. As predicted, South Asian partners' relationship adjustment improved significantly from the first to final session of SCCT, and concurrent increases in each partner's couple identity mediated such improvements. The implications for considering culture and couple identity in couple therapy are discussed. Video Abstract is found in the online version of the article. © 2016 American Association for Marriage and Family Therapy.

Development of a core management tool for MYRRHA

International Nuclear Information System (INIS)

Jalůvka, David; Van den Eynde, Gert; Vandewalle, Stefan

2013-01-01

Highlights: • An in-core fuel management tool is being developed for the flexible irradiation machine MYRRHA. • Specific issues of the MYRRHA in-core fuel management are briefly discussed. • The tool addresses the loading pattern optimization problem. • Illustrative in-core fuel management optimization problems are solved using the tool. - Abstract: MYRRHA is an advanced multi-purpose irradiation facility under development at SCK• CEN in Mol, Belgium. In order to ensure an economical and safe operation of the reactor, an in-core fuel management tool is being developed within the project to address the loading pattern optimization problem. In the paper, the current version of the tool – its architecture and design, unique features, and the field of its application, are presented. In the second part of the paper, the tool’s capabilities are demonstrated on simple MYRRHA in-core fuel management optimization problems

Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 5. TMI-1 Benchmark Performed by Different Coupled Three-Dimensional Neutronics Thermal- Hydraulic Codes

International Nuclear Information System (INIS)

D'Auria, F.; Galassi, G.M.; Spadoni, A.; Gago, J.L.; Grgic, D.

2001-01-01

A comprehensive analysis of a double-ended main-steam-line-break (MSLB) accident assumed to have occurred in the Babcock and Wilcox Three Mile Island (TMI) Unit 1 nuclear power plant (NPP) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy. The research has been carried out in cooperation with the University of Zagreb, Croatia, and with partial financial support from the European Union through a grant to one of the authors. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development Committee on the Safety of Nuclear Installations-Nuclear Science Committee PWR MSLB Benchmark. Different code versions have been adopted in the analysis. Results from the following codes (or code versions) are described in this paper: 1. RELAP5/mod 3.2.2, gamma version, coupled with the three-dimensional (3-D) neutron kinetics PARCS code; 2. RELAP5/mod 3.2.2, gamma version, coupled with the 3-D neutron kinetics QUABBOX code; 3. RELAP5/3D code coupled with the 3-D neutron kinetics NESTLE code. Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by The Pennsylvania State University in cooperation with GPU Nuclear (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission (NRC). The main challenge for the calculation was the prediction of the return to power (RTP) following the inlet of cold water into the core and one 'stuck-withdrawn' control rod. Non-realistic assumptions were proposed to augment the core power peak following scram. Zero-dimensional neutronics codes were capable of detecting the RTP after scram. However, the application of 3-D neutronics codes to the same scenario allowed the calculation of a similar value for overall core power peak but showed power increase occurrence in about one-tenth of the core volume. The results achieved in phase 1 of

NM-Net Gigabit-based Implementation on Core Network Facilities and Network Design Hierarchy

International Nuclear Information System (INIS)

Raja Murzaferi Raja Moktar; Mohd Fauzi Haris; Siti Nurbahyah Hamdan

2011-01-01

Nuclear Malaysia computing network or NM the main backbone of internet working on operational staffs. Main network operating center or NOC is situated in Block 15 and linkup via fiber cabling to adjacent main network blocks (18, 29, 11 connections. Pre 2009 infrastructure; together to form the core networking switch. of the core network infrastructure were limited by the up link between core switches that is the Pair (UTP) Category 6 Cable. Furthermore, majority of the networking infrastructure throughout the agency were mainly built with Fast Ethernet Based specifications to date. With current research and operational tasks highly dependent on IT infrastructure that is being enabled through NM-Net, the performance NM-Net implementing gigabit-based networking system achieve optimal performance of internet networking services in the agency thus catalyze initiative. (author)

The coupling of the Star-Cd software to a whole-core neutron transport code Decart for PWR applications

International Nuclear Information System (INIS)

Thomas, J.W.; Lee, H.C.; Downar, T.J.; Sofu, T.; Weber, D.P.; Joo, H.G.; Cho, J.Y.

2003-01-01

As part of a U.S.- Korea collaborative U.S. Department of Energy INERI project, a comprehensive high-fidelity reactor-core modeling capability is being developed for detailed analysis of existing and advanced PWR reactor designs. An essential element of the project has been the development of an interface between the computational fluid dynamics (CFD) module, STAR-CD, and the neutronics module, DeCART. Since the computational mesh for CFD and neutronics calculations are generally different, the capability to average and decompose data on these different meshes has been an important part of code coupling activities. An averaging process has been developed to extract neutronics zone temperatures in the fuel and coolant and to generate appropriate multi group cross sections and densities. Similar procedures have also been established to map the power distribution from the neutronics zones to the mesh structure used in the CFD module. Since MPI is used as the parallel model in STAR-CD and conflicts arise during initiation of a second level of MPI, the interface developed here is based on using TCP/IP protocol sockets to establish communication between the CFD and neutronics modules. Preliminary coupled calculations have been performed for PWR fuel assembly size problems and converged solutions have been achieved for a series of steady-state problems ranging from a single pin to a 1/8 model of a 17 x 17 PWR fuel assembly. (authors)

Capabilities of the Power Burst Facility

International Nuclear Information System (INIS)

Spencer, W.A.; Jensen, A.M.; McCardell, R.K.

1982-01-01

The unique and diverse test capabilities of the Power Burst Facility (PBF) are described in this paper. The PBF test reactor, located at the Idaho National Engineering Laboratory, simulates normal, off-normal, and accident operating conditions of light water reactor fuel rods. An overview description is given, with specific detail on design and operating characteristics of the driver core, experiment test loop, fission product detection system, test train assembly facility, and support equipment which make the testing capability of the PBF so versatile

Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

International Nuclear Information System (INIS)

Guelol Oezdere, Oya

2003-01-01

Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

Controlled self-assembly of multiferroic core-shell nanoparticles exhibiting strong magneto-electric effects

Energy Technology Data Exchange (ETDEWEB)

Sreenivasulu, Gollapudi; Hamilton, Sean L.; Lehto, Piper R.; Srinivasan, Gopalan, E-mail: srinivas@oakland.edu [Physics Department, Oakland University, Rochester, Michigan 48309-4401 (United States); Popov, Maksym [Physics Department, Oakland University, Rochester, Michigan 48309-4401 (United States); Radiophysics Department, Taras Shevchenko National University of Kyiv, Kyiv 01601 (Ukraine); Chavez, Ferman A. [Chemistry Department, Oakland University, Rochester, Michigan 48309-4401 (United States)

2014-02-03

Ferromagnetic-ferroelectric composites show strain mediated coupling between the magnetic and electric sub-systems due to magnetostriction and piezoelectric effects associated with the ferroic phases. We have synthesized core-shell multiferroic nano-composites by functionalizing 10–100 nm barium titanate and nickel ferrite nanoparticles with complementary coupling groups and allowing them to self-assemble in the presence of a catalyst. The core-shell structure was confirmed by electron microscopy and magnetic force microscopy. Evidence for strong strain mediated magneto-electric coupling was obtained by static magnetic field induced variations in the permittivity over 16–18 GHz and polarization and by electric field induced by low-frequency ac magnetic fields.

Construction of STACY (Static Experiment Critical Facility)

International Nuclear Information System (INIS)

Murakami, Kiyonobu; Onodera, Seiji; Hirose, Hideyuki

1998-08-01

Two critical assemblies, STACY (Static Experiment Critical Facility) and TRACY (Transient Experiment Critical Facility), were constructed in NUCEF (Nuclear Fuel Cycle Safety Engineering Research Facility) to promote researches on the criticality safety at a reprocessing facility. STACY aims at providing critical data of uranium nitrate solution, plutonium nitrate solution and their mixture while varying concentration of solution fuel, core tank shape and size and neutron reflecting condition. STACY achieved first criticality in February 1995, and passed the licensing inspection by STA (Science and Technology Agency of Japan) in May. After that a series of critical experiments commenced with 10 w/o enriched uranium solution. This report describes the outline of STACY at the end of FY 1996. (author)

Modelling guidelines for core exit temperature simulations with system codes

Energy Technology Data Exchange (ETDEWEB)

Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)

2015-05-15

Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.

Reactor core performance estimating device

International Nuclear Information System (INIS)

Tanabe, Akira; Yamamoto, Toru; Shinpuku, Kimihiro; Chuzen, Takuji; Nishide, Fusayo.

1995-01-01

The present invention can autonomously simplify a neural net model thereby enabling to conveniently estimate various amounts which represents reactor core performances by a simple calculation in a short period of time. Namely, a reactor core performance estimation device comprises a nerve circuit net which divides the reactor core into a large number of spacial regions, and receives various physical amounts for each region as input signals for input nerve cells and outputs estimation values of each amount representing the reactor core performances as output signals of output nerve cells. In this case, the nerve circuit net (1) has a structure of extended multi-layered model having direct coupling from an upper stream layer to each of downstream layers, (2) has a forgetting constant q in a corrected equation for a joined load value ω using an inverse error propagation method, (3) learns various amounts representing reactor core performances determined using the physical models as teacher signals, (4) determines the joined load value ω decreased as '0' when it is to less than a predetermined value upon learning described above, and (5) eliminates elements of the nerve circuit net having all of the joined load value decreased to 0. As a result, the neural net model comprises an autonomously simplifying means. (I.S.)

Evaluation of Coupled Precipitator Two

International Nuclear Information System (INIS)

Stone, M.E.

1999-01-01

The offline testing of the Coupled Precipitator Two (CP-2) has been completed. The tests were conducted and are documented. The tests were conducted at an offline test rack near the Drain Tube Test Stand facility in 672-T

Birefringent hollow core fibers

DEFF Research Database (Denmark)

Roberts, John

2007-01-01

Hollow core photonic crystal fiber (HC-PCF), fabricated according to a nominally non-birefringent design, shows a degree of un-controlled birefringence or polarization mode dispersion far in excess of conventional non polarization maintaining fibers. This can degrade the output pulse in many...... applications, and places emphasis on the development of polarization maintaining (PM) HC-PCF. The polarization cross-coupling characteristics of PM HC-PCF are very different from those of conventional PM fibers. The former fibers have the advantage of suffering far less from stress-field fluctuations...... and an increased overlap between the polarization modes at the glass interfaces. The interplay between these effects leads to a wavelength for optimum polarization maintenance, lambda(PM), which is detuned from the wavelength of highest birefringence. By a suitable fiber design involving antiresonance of the core...

Computed Tomography Scanning and Geophysical Measurements of Core from the Coldstream 1MH Well

Energy Technology Data Exchange (ETDEWEB)

Crandall, Dustin M.; Brown, Sarah; Moore, Johnathan E.; Mackey, Paige E.; Paronish, Thomas J.

2018-03-05

The computed tomography (CT) facilities and the Multi-Sensor Core Logger (MSCL) at the National Energy Technology Laboratory (NETL) Morgantown, West Virginia site were used to characterize core of the Marcellus Shale from a vertical well, the Coldstream 1MH Well in Clearfield County, PA. The core is comprised primarily of the Marcellus Shale from a depth of 7,002 to 7,176 ft.

The primary impetus of this work is a collaboration between West Virginia University (WVU) and NETL to characterize core from multiple wells to better understand the structure and variation of the Marcellus and Utica shale formations. As part of this effort, bulk scans of core were obtained from the Coldstream 1MH well, provided by the Energy Corporation of America (now Greylock Energy). This report, and the associated scans, provide detailed datasets not typically available from unconventional shales for analysis. The resultant datasets are presented in this report, and can be accessed from NETL's Energy Data eXchange (EDX) online system using the following link: https://edx.netl.doe.gov/dataset/coldstream-1mh-well.

All equipment and techniques used were non-destructive, enabling future examinations to be performed on these cores. None of the equipment used was suitable for direct visualization of the shale pore space, although fractures and discontinuities were detectable with the methods tested. Low resolution CT imagery with the NETL medical CT scanner was performed on the entire core. Qualitative analysis of the medical CT images, coupled with x-ray fluorescence (XRF), P-wave, and magnetic susceptibility measurements from the MSCL were useful in identifying zones of interest for more detailed analysis as well as fractured zones. En echelon fractures were observed at 7,100 ft and were CT scanned using NETL’s industrial CT scanner at higher resolution. The ability to quickly identify key areas for more detailed study with higher resolution will save time and

Coupled simulation of steam line break accident

International Nuclear Information System (INIS)

Royer, E.; Raimond, E.; Caruge, D.

2000-01-01

The steam line break is a PWR type reactor design accident, which concerns coupled physical phenomena. To control these problems simulation are needed to define and validate the operating procedures. The benchmark OECD PWR MSLB (Main Steam Line Break) has been proposed by the OECD to validate the feasibility and the contribution of the multi-dimensional tools in the simulation of the core transients. First the benchmark OECD PWR MSLB is presented. Then the analysis of the three exercises (system with pinpoint kinetic, three-dimensional core and whole system with three-dimensional core) are discussed. (A.L.B.)

DOE Radiological Control Manual Core Training Program

International Nuclear Information System (INIS)

Scott, H.L.; Maisler, J.

1993-01-01

Over the past year, the Department of Energy (DOE) Office of Health (EH-40) has taken a leading role in the development of new standardized radiological control training programs for use throughout the DOE complex. The Department promulgated its Radiological Control (RadCon) Manual in June 1992. To ensure consistent application of the criteria presented in the RadCon Manual, standardized radiological control core training courses and training materials have been developed for implementation at all DOE facilities. In producing local training programs, standardized core courses are to be supplemented with site-specific lesson plans, viewgraphs, student handbooks, qualification standards, question banks, and wallet-sized training certificates. Training programs for General Employee Radiological Training, Radiological Worker I and II Training, and Radiological Control Technician Training have been disseminated. Also, training committees under the direction of the Office of Health (EH-40) have been established for the development of additional core training courses, development of examination banks, and the update of the existing core training courses. This paper discusses the current activities and future direction of the DOE radiological control core training program

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)

Effect of the design change of the LSSBP on core flow distribution of APR+ Reactor

International Nuclear Information System (INIS)

Kim, Kihwan; Euh, Dong-Jin; Choi, Hae-Seob; Kwon, Tae-Soon

2014-01-01

The uniform core inlet flow distribution of an Advanced Power Reactor Plus (APR+) is required to prevent the failure rate of the HIPER fuel assembly and improve the core thermal margin. KEPCO-E and C and KAERI proposed a design change of the Lower Support Structure Bottom Plate (LSSBP), since the core flow rates were intense near the outer region of the intact LSSBP in a previous study. In this study, an experiment was carried out to evaluate the effect of the design change of the LSSBP on the core flow distribution using the APR+ Core Flow and Pressure (ACOP) test facility. The results showed great improvement on the core flow distribution under a 4-pump balanced flow condition. Under the 4-pump balanced flow condition, fifteen tests were repeated using the ACOP test facility to verify the effect of the 50% blocked flow area at the outer region of the LSSBP on the core inlet flow distribution. The profiles of the core inlet mass flow rates were analyzed using ensemble averaged values, and compared with that of the intact LSSBP. The results showed great improvement for the overall core region. The change in design of the LSSBP is expected to improve the hydraulic performance of an APR+ reactor

Demonstration of the importance of a dedicated neutron beam monitoring system for BNCT facility

International Nuclear Information System (INIS)

Chao, Der-Sheng; Liu, Yuan-Hao; Jiang, Shiang-Huei

2016-01-01

The neutron beam monitoring system is indispensable to BNCT facility in order to achieve an accurate patient dose delivery. The neutron beam monitoring of a reactor-based BNCT (RB-BNCT) facility can be implemented through the instrumentation and control system of a reactor provided that the reactor power level remains constant during reactor operation. However, since the neutron flux in reactor core is highly correlative to complicated reactor kinetics resulting from such as fuel depletion, poison production, and control blade movement, some extent of variation may occur in the spatial distribution of neutron flux in reactor core. Therefore, a dedicated neutron beam monitoring system is needed to be installed in the vicinity of the beam path close to the beam exit of the RB-BNCT facility, where it can measure the BNCT beam intensity as closely as possible and be free from the influence of the objects present around the beam exit. In this study, in order to demonstrate the importance of a dedicated BNCT neutron beam monitoring system, the signals originating from the two in-core neutron detectors installed at THOR were extracted and compared with the three dedicated neutron beam monitors of the THOR BNCT facility. The correlation of the readings between the in-core neutron detectors and the BNCT neutron beam monitors was established to evaluate the improvable quality of the beam intensity measurement inferred by the in-core neutron detectors. In 29 sampled intervals within 16 days of measurement, the fluctuations in the mean value of the normalized ratios between readings of the three BNCT neutron beam monitors lay within 0.2%. However, the normalized ratios of readings of the two in-core neutron detectors to one of the BNCT neutron beam monitors show great fluctuations of 5.9% and 17.5%, respectively. - Highlights: • Two in-core neutron detectors and three BNCT neutron beam monitors were compared. • BNCT neutron beam monitors improve the stability in neutron

Investigation research of core-basic information associated with the coupling analysis. Outline report

International Nuclear Information System (INIS)

Kataoka, Shinichi; Matsunaga, Kenichi; Ishihara, Yoshinao; Kawahara, Kenichi; Neyama, Atsushi; Nakagawa, Koichi; Iwata, Hiroshi; Mori, Koji

2001-03-01

The newest literature information in the foreign countries was researched, and this research showed the basic concept of the coupling analysis code to realize coupling analysis in near field of the geological disposal system. The outline of this research is shown in the following. (1) The combination of M (Mechanical) and C (Chemistry) is placed on the weak relations, because coupling analysis of the United States Yucca Mountain limits a site and the specifications of engineered barrier. (2) One of the purposes of this research is information collecting about coupling analysis code NUFT-C adopted in the United States Yucca Mountain. Therefore, we carried out an information exchange with the United States Lawrence Livermore National Laboratory. We could collect the development purpose of analysis code, key function, and information such as a test case analysis. (3) The investigation of the analysis code concerned with the newest information of coupling analysis which contains the geochemistry process and 2 phase system was done based on the public information for the purpose of building some concept of the coupling analysis code, the extraction of the development issues. It could be understood about the future development strategy and the precaution in addition to a phenomenon to deal with, the current status of the coupling analysis technique as a result of the investigation. (4) It was cleared about the mission of the coupling analysis code and the requirement items (function, quality) by this research. Then, some development options were presented. (5) It was studied about the procedure of developing it to satisfy the above requirement toward the conditions that a site isn't selected, the short development. The tool (Diffpack) which could cope with the speed-up of the calculation time and visualization flexibly was effective, and it was summarized about the test case by using this tool, the key function of this tool as that result. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

AREVA main steam line break fully coupled methodology based on CATHARE-ARTEMIS - 15496

International Nuclear Information System (INIS)

Denis, L.; Jasserand, L.; Tomatis, D.; Segond, M.; Royere, C.; Sauvage, J.Y.

2015-01-01

The CATHARE code developed since 1979 by AREVA, CEA, EDF and IRSN is one of the major thermal-hydraulic system codes worldwide. In order to have at disposal realistic methodologies based on CATHARE for the whole transient and accident analysis in Chapter 15 of Safety Reports, a coupling with the code ARTEMIS was developed. ARTEMIS is the core code in AREVA's new reactor simulator system ARCADIA, using COBRA-FLX to model the thermal-hydraulics in the core. The Fully Coupled Methodology was adapted to the CATHARE-ARTEMIS coupling to perform Main Steam Line Break studies. This methodology, originally applied to the MANTA-SMART-FLICA coupling, is dedicated to Main Steam Line Break transients at zero power. The aim of this paper is to present the coupling between CATHARE and ARTEMIS and the application of the Fully Coupled Methodology in a different code environment. (authors)

Relationship of core exit-temperature noise to thermal-hydraulic conditions in PWRs

International Nuclear Information System (INIS)

Sweeney, F.J.; Upadhyaya, B.R.

1983-01-01

Core exit thermocouple temperature noise and neutron detector noise measurements were performed at the Loss of Fluid Test Facility (LOFT) reactor and a Westinghouse, 1148 MW(e) PWR to relate temperature noise to core thermal-hydraulic conditions. The noise analysis results show that the RMS of the temperature noise increases linearly with increasing core δT at LOFT and the commercial PWR. Out-of-core test loop temperature noise has shown similar behavior. The phase angle between core exit temperature noise and in-core or ex-core neutron noise is directly related to the core coolant flow velocity. However, if the thermocouple response time is slow, compared to the coolant transit time between the sensors, velocities inferred from the phase angle are lower than measured coolant flow velocities

SLIMarray: Lightweight software for microarray facility management

Directory of Open Access Journals (Sweden)

Marzolf Bruz

2006-10-01

Full Text Available Abstract Background Microarray core facilities are commonplace in biological research organizations, and need systems for accurately tracking various logistical aspects of their operation. Although these different needs could be handled separately, an integrated management system provides benefits in organization, automation and reduction in errors. Results We present SLIMarray (System for Lab Information Management of Microarrays, an open source, modular database web application capable of managing microarray inventories, sample processing and usage charges. The software allows modular configuration and is well suited for further development, providing users the flexibility to adapt it to their needs. SLIMarray Lite, a version of the software that is especially easy to install and run, is also available. Conclusion SLIMarray addresses the previously unmet need for free and open source software for managing the logistics of a microarray core facility.

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II cold leg injection tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1985-07-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase in a PWR-LOCA. In order to investigate the effects of radial power profile, three cold leg injection tests with different radial power profiles under the same total heating power and core stored energy were performed by using the Slab Core Test Facility (SCTF) Core-II. It was revealed by comparing these three tests that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. The turnaround temperature in the high power bundles were evaluated to be reduced by about 40 to 120 K. On the other hand, a two-dimensional flow in the core was also induced by the non-uniform water accumulation in the upper plenum and the quench was delayed resultantly in the bundles corresponding to the peripheral bundles of a PWR. However, the effect of the non-uniform upper plenum water accumulation on the turnaround temperature was small because the effect dominated after the turnaround of the cladding temperature. Selected data from Tests S2-SH1, S2-SH2 and S2-O6 are also presented in this report. Some data from Tests S2-SH1 and S2-SH2 were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

The Current Status of the Space Station Biological Research Project: a Core Facility Enabling Multi-Generational Studies under Slectable Gravity Levels

Science.gov (United States)

Santos, O.

2002-01-01

The Space Station Biological Research Project (SSBRP) has developed a new plan which greatly reduces the development costs required to complete the facility. This new plan retains core capabilities while allowing for future growth. The most important piece of equipment required for quality biological research, the 2.5 meter diameter centrifuge capable of accommodating research specimen habitats at simulated gravity levels ranging from microgravity to 2.0 g, is being developed by NASDA, the Japanese space agency, for the SSBRP. This is scheduled for flight to the ISS in 2007. The project is also developing a multi-purpose incubator, an automated cell culture unit, and two microgravity habitat holding racks, currently scheduled for launch in 2005. In addition the Canadian Space Agency is developing for the project an insect habitat, which houses Drosophila melanogaster, and provides an internal centrifuge for 1 g controls. NASDA is also developing for the project a glovebox for the contained manipulation and analysis of biological specimens, scheduled for launch in 2006. This core facility will allow for experimentation on small plants (Arabidopsis species), nematode worms (C. elegans), fruit flies (Drosophila melanogaster), and a variety of microorganisms, bacteria, yeast, and mammalian cells. We propose a plan for early utilization which focuses on surveys of changes in gene expression and protein structure due to the space flight environment. In the future, the project is looking to continue development of a rodent habitat and a plant habitat that can be accommodated on the 2.5 meter centrifuge. By utilizing the early phases of the ISS to broadly answer what changes occur at the genetic and protein level of cells and organisms exposed to the ISS low earth orbit environment, we can generate interest for future experiments when the ISS capabilities allow for direct manipulation and intervention of experiments. The ISS continues to hold promise for high quality, long

XGC developments for a more efficient XGC-GENE code coupling

Science.gov (United States)

Dominski, Julien; Hager, Robert; Ku, Seung-Hoe; Chang, Cs

2017-10-01

In the Exascale Computing Program, the High-Fidelity Whole Device Modeling project initially aims at delivering a tightly-coupled simulation of plasma neoclassical and turbulence dynamics from the core to the edge of the tokamak. To permit such simulations, the gyrokinetic codes GENE and XGC will be coupled together. Numerical efforts are made to improve the numerical schemes agreement in the coupling region. One of the difficulties of coupling those codes together is the incompatibility of their grids. GENE is a continuum grid-based code and XGC is a Particle-In-Cell code using unstructured triangular mesh. A field-aligned filter is thus implemented in XGC. Even if XGC originally had an approximately field-following mesh, this field-aligned filter permits to have a perturbation discretization closer to the one solved in the field-aligned code GENE. Additionally, new XGC gyro-averaging matrices are implemented on a velocity grid adapted to the plasma properties, thus ensuring same accuracy from the core to the edge regions.

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

Diamond Ordinance Radiation Facility (DORF) reactor operating experiences

International Nuclear Information System (INIS)

Gieseler, Walter

1970-01-01

The Diamond Ordnance Radiation Facility Mark F Reactor is described and some of the problems encountered with its operation are discussed. In a period from reactor startup in September 1961 to June 1964, when the aluminum-clad core was changed to a stainless-steel clad core, a total of 30 fuel elements were removed from reactor service because of excessive growth. One leaking fuel element was detected during the lifetime of the aluminum- clad core. In June 1964, the core was changed to the stainless-steel-clad high hydride fuel elements. Since the installation of the stainless-steel-clad fuel element core, there has been a gradual decline of excess reactivity. Various theories were discussed as the cause but the investigations have resulted in no definitive conclusion that could account for the total reactivity loss

Lead Coolant Test Facility Technical and Functional Requirements, Conceptual Design, Cost and Construction Schedule

International Nuclear Information System (INIS)

Soli T. Khericha

2006-01-01

This report presents preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements of basis are identified: Develop and Demonstrate Prototype Lead/Lead-Bismuth Liquid Metal Flow Loop Develop and Demonstrate Feasibility of Submerged Heat Exchanger Develop and Demonstrate Open-lattice Flow in Electrically Heated Core Develop and Demonstrate Chemistry Control Demonstrate Safe Operation and Provision for Future Testing. These five broad areas are divided into twenty-one (21) specific requirements ranging from coolant temperature to design lifetime. An overview of project engineering requirements, design requirements, QA and environmental requirements are also presented. The purpose of this T and FRs is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M. It is also estimated that the facility will require two years to be constructed and ready for operation

Inclusion of orbital relaxation and correlation through the unitary group adapted open shell coupled cluster theory using non-relativistic and scalar relativistic Hamiltonians to study the core ionization potential of molecules containing light to medium-heavy elements

Science.gov (United States)

Sen, Sangita; Shee, Avijit; Mukherjee, Debashis

2018-02-01

The orbital relaxation attendant on ionization is particularly important for the core electron ionization potential (core IP) of molecules. The Unitary Group Adapted State Universal Coupled Cluster (UGA-SUMRCC) theory, recently formulated and implemented by Sen et al. [J. Chem. Phys. 137, 074104 (2012)], is very effective in capturing orbital relaxation accompanying ionization or excitation of both the core and the valence electrons [S. Sen et al., Mol. Phys. 111, 2625 (2013); A. Shee et al., J. Chem. Theory Comput. 9, 2573 (2013)] while preserving the spin-symmetry of the target states and using the neutral closed-shell spatial orbitals of the ground state. Our Ansatz invokes a normal-ordered exponential representation of spin-free cluster-operators. The orbital relaxation induced by a specific set of cluster operators in our Ansatz is good enough to eliminate the need for different sets of orbitals for the ground and the core-ionized states. We call the single configuration state function (CSF) limit of this theory the Unitary Group Adapted Open-Shell Coupled Cluster (UGA-OSCC) theory. The aim of this paper is to comprehensively explore the efficacy of our Ansatz to describe orbital relaxation, using both theoretical analysis and numerical performance. Whenever warranted, we also make appropriate comparisons with other coupled-cluster theories. A physically motivated truncation of the chains of spin-free T-operators is also made possible by the normal-ordering, and the operational resemblance to single reference coupled-cluster theory allows easy implementation. Our test case is the prediction of the 1s core IP of molecules containing a single light- to medium-heavy nucleus and thus, in addition to demonstrating the orbital relaxation, we have addressed the scalar relativistic effects on the accuracy of the IPs by using a hierarchy of spin-free Hamiltonians in conjunction with our theory. Additionally, the contribution of the spin-free component of the two

Inclusion of orbital relaxation and correlation through the unitary group adapted open shell coupled cluster theory using non-relativistic and scalar relativistic Hamiltonians to study the core ionization potential of molecules containing light to medium-heavy elements.

Science.gov (United States)

Sen, Sangita; Shee, Avijit; Mukherjee, Debashis

2018-02-07

The orbital relaxation attendant on ionization is particularly important for the core electron ionization potential (core IP) of molecules. The Unitary Group Adapted State Universal Coupled Cluster (UGA-SUMRCC) theory, recently formulated and implemented by Sen et al. [J. Chem. Phys. 137, 074104 (2012)], is very effective in capturing orbital relaxation accompanying ionization or excitation of both the core and the valence electrons [S. Sen et al., Mol. Phys. 111, 2625 (2013); A. Shee et al., J. Chem. Theory Comput. 9, 2573 (2013)] while preserving the spin-symmetry of the target states and using the neutral closed-shell spatial orbitals of the ground state. Our Ansatz invokes a normal-ordered exponential representation of spin-free cluster-operators. The orbital relaxation induced by a specific set of cluster operators in our Ansatz is good enough to eliminate the need for different sets of orbitals for the ground and the core-ionized states. We call the single configuration state function (CSF) limit of this theory the Unitary Group Adapted Open-Shell Coupled Cluster (UGA-OSCC) theory. The aim of this paper is to comprehensively explore the efficacy of our Ansatz to describe orbital relaxation, using both theoretical analysis and numerical performance. Whenever warranted, we also make appropriate comparisons with other coupled-cluster theories. A physically motivated truncation of the chains of spin-free T-operators is also made possible by the normal-ordering, and the operational resemblance to single reference coupled-cluster theory allows easy implementation. Our test case is the prediction of the 1s core IP of molecules containing a single light- to medium-heavy nucleus and thus, in addition to demonstrating the orbital relaxation, we have addressed the scalar relativistic effects on the accuracy of the IPs by using a hierarchy of spin-free Hamiltonians in conjunction with our theory. Additionally, the contribution of the spin-free component of the two

Coupled 3D neutronics/thermal hydraulics modeling of the SAFARI-1 MTR

International Nuclear Information System (INIS)

Rosenkrantz, Adam; Avramova, Maria; Ivanov, Kostadin; Prinsloo, Rian; Botes, Danniëll; Elsakhawy, Khalid

2014-01-01

Highlights: • Development of 3D coupled neutronics/thermal–hydraulic model of SAFARI-1. • Verification of 3D steady-state NEM based neutronics model for SAFARI-1. • Verification of 3D COBRA-TF based thermal–hydraulic model of SAFARI-1. • Quantification of the effect of correct modeling of thermal–hydraulic feedback. - Abstract: The purpose of this study was to develop a coupled accurate multi-physics model of the SAFARI-1 Material Testing Reactor (MTR), a facility that is used for both research and the production of medical isotopes. The model was developed as part of the SAFARI-1 benchmarking project as a cooperative effort between the Pennsylvania State University (PSU) and the South African Nuclear Energy Corporation (Necsa). It was created using a multi-physics coupling of state of the art nuclear reactor simulation tools, consisting of a neutronics code and a thermal hydraulics code. The neutronics tool used was the PSU code NEM, and the results from this component were verified using the Necsa neutronics code OSCAR-4, which is utilized for SAFARI-1 core design and fuel management. On average, the multiplication factors of the neutronics models agreed to within 5 pcm and the radial assembly-averaged powers agreed to within 0.2%. The thermal hydraulics tool used was the PSU version of COBRA-TF (CTF) sub-channel code, and the results of this component were verified against another thermal hydraulics code, the RELAP5-3D system code, used at Necsa for thermal–hydraulics analysis of SAFARI-1. Although only assembly-averaged results from RELAP5-3D were available, they fell within the range of values for the corresponding assemblies in the comprehensive CTF solution. This comparison allows for the first time to perform a quantification of steady-state errors for a low-powered MTR with an advanced thermal–hydraulic code such as CTF on a per-channel basis as compared to simpler and coarser-mesh RELAP5-3D modeling. Additionally, a new cross section

Developing confidence in a coupled TH model based on the results of experiment by using engineering scale test facility, 'COUPLE'

International Nuclear Information System (INIS)

Fujisaki, Kiyoshi; Suzuki, Hideaki; Fujita, Tomoo

2008-03-01

It is necessary to understand quantitative changes of near-field conditions and processes over time and space for modeling the near-field evolution after emplacement of engineered barriers. However, the coupled phenomena in near-field are complicated because thermo-, hydro-, mechanical, chemical processes will interact each other. The question is, therefore, whether the applied model will represent the coupled behavior adequately or not. In order to develop confidence in the modeling, it is necessary to compare with results of coupled behavior experiments in laboratory or in site. In this report, we evaluated the applicability of a coupled T-H model under the conditions of simulated near-field for the results of coupled T-H experiment in laboratory. As a result, it has been shown that the fitting by the modeling with the measured data is reasonable under this condition. (author)

High Performance Computing Facility Operational Assessment 2015: Oak Ridge Leadership Computing Facility

Energy Technology Data Exchange (ETDEWEB)

Barker, Ashley D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Bernholdt, David E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Bland, Arthur S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Gary, Jeff D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Hack, James J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; McNally, Stephen T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Rogers, James H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Smith, Brian E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Straatsma, T. P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Sukumar, Sreenivas Rangan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Thach, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Tichenor, Suzy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Vazhkudai, Sudharshan S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility; Wells, Jack C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Oak Ridge Leadership Computing Facility

2016-03-01

Oak Ridge National Laboratory’s (ORNL’s) Leadership Computing Facility (OLCF) continues to surpass its operational target goals: supporting users; delivering fast, reliable systems; creating innovative solutions for high-performance computing (HPC) needs; and managing risks, safety, and security aspects associated with operating one of the most powerful computers in the world. The results can be seen in the cutting-edge science delivered by users and the praise from the research community. Calendar year (CY) 2015 was filled with outstanding operational results and accomplishments: a very high rating from users on overall satisfaction that ties the highest-ever mark set in CY 2014; the greatest number of core-hours delivered to research projects; the largest percentage of capability usage since the OLCF began tracking the metric in 2009; and success in delivering on the allocation of 60, 30, and 10% of core hours offered for the INCITE (Innovative and Novel Computational Impact on Theory and Experiment), ALCC (Advanced Scientific Computing Research Leadership Computing Challenge), and Director’s Discretionary programs, respectively. These accomplishments, coupled with the extremely high utilization rate, represent the fulfillment of the promise of Titan: maximum use by maximum-size simulations. The impact of all of these successes and more is reflected in the accomplishments of OLCF users, with publications this year in notable journals Nature, Nature Materials, Nature Chemistry, Nature Physics, Nature Climate Change, ACS Nano, Journal of the American Chemical Society, and Physical Review Letters, as well as many others. The achievements included in the 2015 OLCF Operational Assessment Report reflect first-ever or largest simulations in their communities; for example Titan enabled engineers in Los Angeles and the surrounding region to design and begin building improved critical infrastructure by enabling the highest-resolution Cybershake map for Southern

Design of Multi-core Fiber Patch Panel for Space Division Multiplexing Implementations

DEFF Research Database (Denmark)

Gonzalez, Luz E.; Morales, Alvaro; Rommel, Simon

2018-01-01

A multi-core fiber (MCF) patch panel was designed, allowing easy coupling of individual signals to and from a 7-core MCF. The device was characterized, measuring insertion loss and cross talk, finding highest insertion loss and lowest crosstalk at 1300 nm with values of 9.7 dB and -36.5 d...

Simulation an Accelerator driven Subcritical Reactor core with thorium fuel

International Nuclear Information System (INIS)

Shirmohammadi, L.; Pazirandeh, A.

2011-01-01

The main purpose of this work is simulation An Accelerator driven Subcritical core with Thorium as a new generation nuclear fuel. In this design core , A subcritical core coupled to an accelerator with proton beam (E p =1 GeV) is simulated by MCNPX code .Although the main purpose of ADS systems are transmutation and use MA (Minor Actinides) as a nuclear fuel but another use of these systems are use thorium fuel. This simulated core has two fuel assembly type : (Th-U) and (U-Pu) . Consequence , Neutronic parameters related to ADS core are calculated. It has shown that Thorium fuel is use able in this core and less nuclear waste ,Although Iran has not Thorium reserves but study on Thorium fuel cycle can open a new horizontal in use nuclear energy as a clean energy and without nuclear waste

Evaluation report on CCTF Core-II reflood tests C2-AC1 (run 51) and C2-4 (run 62)

International Nuclear Information System (INIS)

Sugimoto, Jun; Iguchi, Tadashi; Murao, Yoshio

1984-02-01

A reflood test program has been conducted at Japan Atomic Energy Research Institute (JAERI) using large scale test facilities named Cylindrical Core Test Facility (CCTF) and Slab Core Test Facility (SCTF). The present report describes the effect of the initial clad temperature i.e., the initial stored energy on reflood phenomena observed in CCTF Core-II tests C2-ACl and C2-4. The peak clad temperatures of tests C2-ACl and C2-4 were 863 K and 1069 K, respectively at reflood initiation. With higher initial clad temperature, obtained were lower water accumulation in the core and upper plenum, and higher loop mass flow rate in an early reflood transient due to larger heat release of the stored energy in the core. Core inlet flow conditions were only affected shortly after the reflood initiation, causing the suppressed flooding rate and the larger U-tube flow oscillation between the core and the downcomer. In the core, with higher initial clad temperature, slower quench front propagation and higher turnaround temperature were observed. Responses to a higher initial clad temperature were similar to those observed in CCTF Core-I and FLECHT tests. Thus, the lower temperature rise with higher initial clad temperature was experimentally confirmed. The importance of higher flooding rate at initial period was analytically shown for further decreasing the temperature rise. (author)

The WR-1 corrosion test facility

International Nuclear Information System (INIS)

Murphy, E.V.; Simmons, G.R.

1978-07-01

This report describes a new Corrosion Test Facility which has recently been installed in the WR-1 organic-cooled research reactor. The irradiation facility is a single insert, installed in a reactor site, which can deliver a fast neutron flux density of 2.65 x 10 17 neutrons/(m 2 .s) to specimens under irradiation. A self-contained controlled-chemistry cooling water circuit removes the gamma- and neutron-heat generated in the insert and specimens. Specimen temperatures typically vary from 245 deg C to 280 deg C across the insert core region. (author)