Thermo-hydraulic analysis of the cool-down of the EDIPO test facility

Science.gov (United States)

Lewandowska, Monika; Bagnasco, Maurizio

2011-09-01

The first cool-down of the EDIPO (European DIPOle) test facility is foreseen to take place in 2011 by means of the existing 1.2 kW cryoplant at EPFL-CRPP Villigen. In this work, the thermo-hydraulic analysis of the EDIPO cool-down is performed in order both to assess the its duration and to optimize the procedure. The cool-down is driven by the helium flowing in both the outer cooling channel and in the windings connected hydraulically in parallel. We take into account limitations due to the pressure drop in the cooling circuit and the refrigerator capacity as well as heat conduction in the iron yoke. Two schemes of the hydraulic cooling circuit in the EDIPO windings are studied (coils connected in series and coils connected in parallel). The analysis is performed by means of an analytical model complemented by and numerical model. The results indicate that the cool-down to 5 K can be achieved in about 12 days.

Preliminary study of the thermo-hydraulic behaviour of the binary breeder reactor

International Nuclear Information System (INIS)

Silveira Luz, M. da; Ferreira, W.J.

1984-06-01

Continuing the development of the Binary Breeder Reactor, its physical configuration and the advantages of differents types of spacers are analysed. In order to simulate the thermo-hydraulic behaviour and obtain data for a preliminary evaluation of the core geometry, the COBRA III C code was used to study the effects of the lenght and diameter of the fuel element, the coolant inlet temperature, the system pressure, helicoidal pitch and the pitch to diameter ratio. (Author) [pt

Numerical study of the thermo-hydraulic behavior for the Candu type fuel channel

International Nuclear Information System (INIS)

Lazaro, Pavel Gabriel; Balas Ghizdeanu, Elena Nineta

2008-01-01

Candu type reactors use fuel channel in a horizontal lattice. The fuel bundles are positioned in two Zircaloy tubes: the pressure tube surrounded by calandria tube. Inside the pressure tube the coolant heavy water flows. The coolant reaches high temperatures and pressures. Due to irregular neutron spatial distribution, the fuel channel stress differs from one channel to other. In one improbable event of severe accident, the fuel channel behaves differently according to its normal function history. Over the years, there have been many research projects trying to analyze thermal hydraulic performance of the design and to add some operational improvements in order to achieve an efficient thermal hydraulic distribution. This paper discusses the thermo hydraulic behavior (influence of the temperature and velocity distribution) of the most solicited channel, simulated with Fluent 6.X. Code. Moreover it will be commented the results obtained using different models and mesh applied. (authors)

Fundamental study on thermo-hydraulic behaviors during power transient, 2

International Nuclear Information System (INIS)

Shinano, M.; Inoue, A.

1988-01-01

Thermo-hydraulic behaviors during power transient of nuclear reactors are studied. Boiling around test rod heated transiently forces to flow out liquid in the test section and generates high pressure pulse. In this study, it is investigated experimentally and analytically that magnitude of pressure pulse and energy conversion efficiency to the mechanical works in cases of fragmentation and non-fragmentation. In analysis, effects of increasing of heat transfer and of interaction area due to fragmentation is considered. Consequently, 1) magnitude of pressure pulse on fragmentation is about 10 times greater than that on non-fragmentation. 2) analytical model can show characteristics of fragmentation processes qualitatively. (author)

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

International Nuclear Information System (INIS)

Liu, X.J.; Cheng, X.

2009-01-01

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

Study of the thermo-mechanical performances of the IFMIF-EVEDA Lithium Test Loop target assembly

Energy Technology Data Exchange (ETDEWEB)

Di Maio, P.A., E-mail: dimaio@din.unipa.it [Dipartimento dell' Energia, Universita di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Arena, P.; Bongiovi, G. [Dipartimento dell' Energia, Universita di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, R.; Micciche, G.; Tincani, A. [ENEA C. R. Brasimone, 40032 Camugnano, Bologna (Italy)

2012-08-15

Highlights: Black-Right-Pointing-Pointer IFMIF-EVEDA target assembly thermo-mechanical behavior has been investigated. Black-Right-Pointing-Pointer Finite element method has been followed and a commercial code has been used. Black-Right-Pointing-Pointer Nominal, design and pressure test steady state scenarios and start-up transient conditions have been investigated. Black-Right-Pointing-Pointer Steady state results have shown that back-plate yielding may occur only under the design scenario. Black-Right-Pointing-Pointer Transient analysis has indicated that TA start-up lasts for {approx}60 h. - Abstract: Within the framework of the IFMIF R and D program and in close cooperation with ENEA-Brasimone, at the Department of Energy of the University of Palermo a research campaign has been launched to investigate the thermo-mechanical behavior of the target assembly under both steady state and start-up transient conditions. A theoretical approach based on the finite element method (FEM) has been followed and a well-known commercial code has been adopted. A realistic 3D FEM model of the target assembly has been set-up and optimized by running a mesh independency analysis. A proper set of loads and boundary conditions, mainly concerned with radiation heat transfer between the target assembly external walls and the inner walls of its containment vessel, have been considered and the target assembly thermo-mechanical behavior under nominal, design and pressure test steady state scenarios and start-up transient conditions has been investigated. Results are herewith reported and discussed.

Hydrogen and methane generation from large hydraulic plant: Thermo-economic multi-level time-dependent optimization

International Nuclear Information System (INIS)

Rivarolo, M.; Magistri, L.; Massardo, A.F.

2014-01-01

Highlights: • We investigate H 2 and CH 4 production from very large hydraulic plant (14 GW). • We employ only “spilled energy”, not used by hydraulic plant, for H 2 production. • We consider the integration with energy taken from the grid at different prices. • We consider hydrogen conversion in chemical reactors to produce methane. • We find plants optimal size using a time-dependent thermo-economic approach. - Abstract: This paper investigates hydrogen and methane generation from large hydraulic plant, using an original multilevel thermo-economic optimization approach developed by the authors. Hydrogen is produced by water electrolysis employing time-dependent hydraulic energy related to the water which is not normally used by the plant, known as “spilled water electricity”. Both the demand for spilled energy and the electrical grid load vary widely by time of year, therefore a time-dependent hour-by-hour one complete year analysis has been carried out, in order to define the optimal plant size. This time period analysis is necessary to take into account spilled energy and electrical load profiles variability during the year. The hydrogen generation plant is based on 1 MWe water electrolysers fuelled with the “spilled water electricity”, when available; in the remaining periods, in order to assure a regular H 2 production, the energy is taken from the electrical grid, at higher cost. To perform the production plant size optimization, two hierarchical levels have been considered over a one year time period, in order to minimize capital and variable costs. After the optimization of the hydrogen production plant size, a further analysis is carried out, with a view to converting the produced H 2 into methane in a chemical reactor, starting from H 2 and CO 2 which is obtained with CCS plants and/or carried by ships. For this plant, the optimal electrolysers and chemical reactors system size is defined. For both of the two solutions, thermo

Hydraulic shock damper for fuel assemblies of nuclear reactors

International Nuclear Information System (INIS)

Jabson, F.S.

1978-01-01

A typical embodiment of this invention provides a hydraulic mechanism for alleviating the effect of seismic forces and other stresses that are applied to a fuel assembly in a nuclear reactor. Illustratively, hollow guide posts potrude into a fuel assembly end fitting grid from biased spring pads. Plungers that move with the spring pads plug one end of each of the respective guide posts. Plates on the end fitting grid that have individual holes for fluid discharge partially plug the other ends of the respective guide posts, thereby providing a hydraulic means for absorbing the longitudinal component of seismic shocks and other anticipated forces. (Auth.)

Basic researches on thermo-hydraulic non-equilibrium phenomena related to nuclear reactor safety

International Nuclear Information System (INIS)

Sakurai, Akira; Kataoka, Isao; Aritomi, Masanori.

1989-01-01

A review was made of recent developments of fundamental researches on thermo-hydraulic non-equilibrium phenomena related to light water reactor safety, in relation to problems to be solved for the improvement of safety analysis codes. As for the problems related to flow con ditions, fundamental researches on basic conservation equations and constitutive equations for transient two-phase flow were reviewed. Regarding to the problems related to thermal non-equilibrium phenomena, fundamental researches on film boiling in pool and forced convection, transient boiling heat transfer and flow behavior caused by pressure transients were reviewed. (author)

Related research with thermo hydraulics safety by means of Trace code

International Nuclear Information System (INIS)

Chaparro V, F. J.; Del Valle G, E.; Rodriguez H, A.; Gomez T, A. M.; Sanchez E, V. H.; Jager, W.

2014-10-01

In this article the results of the design of a pressure vessel of a BWR/5 similar to the type of Laguna Verde NPP are presented, using the Trace code. A thermo hydraulics Vessel component capable of simulating the behavior of fluids and heat transfer that occurs within the reactor vessel was created. The Vessel component consists of a three-dimensional cylinder divided into 19 axial sections, 4 azimuthal sections and two concentric radial rings. The inner ring is used to contain the core and the central part of the reactor, while the outer ring is used as a down comer. Axial an azimuthal divisions were made with the intention that the dimensions of the internal components, heights and orientation of the external connections match the reference values of a reactor BWR/5 type. In the model internal components as, fuel assemblies, steam separators, jet pumps, guide tubes, etc. are included and main external connections as, steam lines, feed-water or penetrations of the recirculation system. The model presents significant simplifications because the object is to keep symmetry between each azimuthal section of the vessel. In most internal components lack a detailed description of the geometry and initial values of temperature, pressure, fluid velocity, etc. given that it only considered the most representative data, however with these simulations are obtained acceptable results in important parameters such as the total flow through the core, the pressure in the vessel, percentage of vacuums fraction, pressure drop in the core and the steam separators. (Author)

Thermo-hydraulic Quench Propagation at the LHC Superconducting Magnet String

CERN Document Server

Rodríguez-Mateos, F; Serio, L

1998-01-01

The superconducting magnets of the LHC are protected by heaters and cold by-pass diodes. If a magnet quenches, the heaters on this magnet are fired and the magnet chain is de-excited in about two minu tes by opening dump switches in parallel to a resistor. During the time required for the discharge, adjacent magnets might quench due to thermo-hydraulic propagation in the helium bath and/or heat con duction via the bus bar. The number of quenching magnets depends on the mechanisms for the propagation. In this paper we report on quench propagation experiments from a dipole magnet to an adjacent ma gnet. The mechanism for the propagation is hot helium gas expelled from the first quenching magnet. The propagation changes with the pressure opening settings of the quench relief valves.

Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC

International Nuclear Information System (INIS)

Gomez Garcia-Torano, I.; Jimenez, G.

2013-01-01

The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

Directory of Open Access Journals (Sweden)

Jakubec Jakub

2017-04-01

Full Text Available The paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.

TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

International Nuclear Information System (INIS)

Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

1983-01-01

TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

International Nuclear Information System (INIS)

Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

2002-12-01

In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

Development of LILAC-meltpool for the thermo-hydraulic analysis of core melt relocated in a reactor vessel

International Nuclear Information System (INIS)

Kim, Jong Tae; Kim, Sang Baik; Kim, Hee Dong

2002-03-01

LILAC-meltpool has been developed to study thermo-hydraulic behavior of molten pool and thermal behavior of vessel wall during severe accident. To validate LILAC-meltpool code several two and three dimensional thermo-hydraulic problems were selected and solved. The benchmark problems have experimental results or verified numerical results. Through the validation it was found that LILAC-meltpool reproduces very accurate numerical results. Two-layered semicircular pool was solved to study thermal and hydraulic characteristics of pool stratification. The LAVA experiment using alumina/ferrite molten pool was calculated and compared with computed results. Cooling of alumina/ferrite two-layered pool was affected by stratification. In the numerical results temperature of vessel inner was highest at a location below the interface. Crust was developed from upper surface and lower outer surface, but in the area near the interface corium simulant existed as molten state for long time. LAVA-4 experiment was studied using gap-cooling model in LILAC-meltpool code. Temperature increase of LAVA vessel after alumina melt relocation was strongly dependent on gap formation mechanism. Calculated cooling rates of the vessel were very similar to experimental results. For LAVA experiments which do not have heat generation coolant penetrates easily into a gap and it is found that gap-cooling is very effective for cooling of vessel, but it is thought that coolant penetration could be limited near upper part of gap because of decay heat and high temperature of corium crust

On the application of reynolds theory to thermo-piezo-viscous lubrication in oil hydraulics

DEFF Research Database (Denmark)

Johansen, Per; Roemer, Daniel Beck; Andersen, Torben O.

2015-01-01

The efficiency of fluid power motors and pumps is a subject to research, which has generated numerous publications during the last three decades. The main incentives for this research are optimization of reliability and efficiency through the study of loss and wear mechanisms, which are very....... In this paper the derivation of Reynolds equation from the continuum assumption is reviewed and it is shown that the validity of Reynolds theory based pressure field solutions in oil hydraulic thermo-piezo-viscous lubrication models are subject to maximum bounds on the pressure and temperature field gradients...

Thermo hydraulic and quench propagation characteristics of SST-1 TF coil

Energy Technology Data Exchange (ETDEWEB)

Sharma, A.N., E-mail: ansharma@ipr.res.in [Institute for Plasma Research, Gandhinagar (India); Pradhan, S. [Institute for Plasma Research, Gandhinagar (India); Duchateau, J.L. [CEA Cadarache, 13108 St Paul lez Durance Cedex (France); Khristi, Y.; Prasad, U.; Doshi, K.; Varmora, P.; Patel, D.; Tanna, V.L. [Institute for Plasma Research, Gandhinagar (India)

2014-02-15

Highlights: • Details of SST-1 TF coils, CICC. • Details of SST-1 TF coil cold test. • Quench analysis of TF magnet. • Flow changes following quench. • Predictive analysis of assembled magnet system. - Abstract: SST-1 toroidal field (TF) magnet system is comprising of sixteen superconducting modified ‘D’ shaped TF coils. During single coil test campaigns spanning from June 10, 2010 till January 24, 2011; the electromagnetic, thermal hydraulic and mechanical performances of each TF magnet have been qualified at its respective nominal operating current of 10,000 A in either two-phase or supercritical helium cooling conditions. During the current charging experiments, few quenches have initiated either as a consequence of irrecoverable normal zones or being induced in some of the TF magnets. Quench evolution in the TF coils have been analyzed in detail in order to understand the thermal hydraulic and quench propagation characteristics of the SST-1 TF magnets. The same were also simulated using 1D code Gandalf. This paper elaborates the details of the analyses and the quench simulation results. A predictive quench propagation analysis of 16 assembled TF magnets system has also been reported in this paper.

Development of gas-cooled fast reactor and its thermo-hydraulics

International Nuclear Information System (INIS)

Kawamura, Hiroshi

1977-10-01

Development, thermo-hydraulics and safety of GCFR are reviewed. The Development of Gas-Cooled Fast Reactor (GCFR) utilizes helium technology of HTGR and fuel technology of LMFBR. The breeding ratio of GCFR will be larger than that of LMFBR by about 0.2. Features of GCFR are a fuel with roughened surface to raise the heat transfer and vent system for the pressure equalization in the fuel rod. Helium as coolant of GCFR is chemically stable and stays in the single phase. So, there is no fuel-coolant interaction unlike the case of LMFBR. Since the helium must be pressurized, possibility of a depressurization accident is not negligible. In the United States, a 300MWe demonstration plant program is about to start; the collaboration with European countries is now quite active in this field. Though the development of GCFR started behind that of LMFBR, GCFR is equally promising as a fast breeder reactor. When realized, it will present possibility of a choice between these two. (auth.)

Thermo-hydraulic design of earth-air heat exchangers

Energy Technology Data Exchange (ETDEWEB)

Paepe, M. de [Ghent University (Belgium). Department of Flow, Heat and Combustion Mechanics; Janssens, A. [Ghent University (Belgium). Department of Architecture and Urbanism

2003-05-01

Earth-air heat exchangers, also called ground tube heat exchangers, are an interesting technique to reduce energy consumption in a building. They can cool or heat the ventilation air, using cold or heat accumulated in the soil. Several papers have been published in which a design method is described. Most of them are based on a discretisation of the one-dimensional heat transfer problem in the tube. Three-dimensional complex models, solving conduction and moisture transport in the soil are also found. These methods are of high complexity and often not ready for use by designers. In this paper, a one-dimensional analytical method is used to analyse the influence of the design parameters of the heat exchanger on the thermo-hydraulic performance. A relation is derived for the specific pressure drop, linking thermal effectiveness with pressure drop of the air inside the tube. The relation is used to formulate a design method which can be used to determine the characteristic dimensions of the earth-air heat exchanger in such a way that optimal thermal effectiveness is reached with acceptable pressure loss. The choice of the characteristic dimensions, becomes thus independent of the soil and climatological conditions. This allows designers to choose the earth-air heat exchanger configuration with the best performance. (author)

Thermo-hydraulic design of earth-air heat exchangers

Energy Technology Data Exchange (ETDEWEB)

De Paepe, M. [Department of Flow, Heat and Combustion Mechanics, Ghent University, Ghent (Belgium); Janssens, A. [Department of Architecture and Urbanism, Ghent University, Ghent (Belgium)

2003-07-01

Earth-air heat exchangers, also called ground tube heat exchangers, are an interesting technique to reduce energy consumption in a building. They can cool or heat the ventilation air, using cold or heat accumulated in the soil. Several papers have been published in which a design method is described. Most of them are based on a discretisation of the one-dimensional heat transfer problem in the tube. Three-dimensional complex models, solving conduction and moisture transport in the soil are also found. These methods are of high complexity and often not ready for use by designers. In this paper, a one-dimensional analytical method is used to analyse the influence of the design parameters of the heat exchanger on the thermo-hydraulic performance. A relation is derived for the specific pressure drop, linking thermal effectiveness with pressure drop of the air inside the tube. The relation is used to formulate a design method which can be used to determine the characteristic dimensions of the earth-air heat exchanger in such a way that optimal thermal effectiveness is reached with acceptable pressure loss. The choice of the characteristic dimensions, becomes thus independent of the soil and climatological conditions. This allows designers to choose the earth-air heat exchanger configuration with the best performance. (author)

A novel thermo-hydraulic coupling model to investigate the crater formation in electrical discharge machining

Science.gov (United States)

Tang, Jiajing; Yang, Xiaodong

2017-09-01

A novel thermo-hydraulic coupling model was proposed in this study to investigate the crater formation in electrical discharge machining (EDM). The temperature distribution of workpiece materials was included, and the crater formation process was explained from the perspective of hydrodynamic characteristics of the molten region. To better track the morphology of the crater and the movement of debris, the level-set method was introduced in this study. Simulation results showed that the crater appears shortly after the ignition of the discharge, and the molten material is removed by vaporizing in the initial stage, then by splashing at the following time. The driving force for the detachment of debris in the splashing removal stage comes from the extremely large pressure difference in the upper part of the molten region, and the morphology of the crater is also influenced by the shearing flow of molten material. It was found that the removal ratio of molten material is only about 7.63% under the studied conditions, leaving most to form the re-solidification layer on the surface of the crater. The size of the crater reaches the maximum at the end of discharge duration then experiences a slight reduction because of the reflux of molten material after the discharge. The results of single pulse discharge experiments showed that the morphologies and sizes between the simulation crater and actual crater are good at agreement, verifying the feasibility of the proposed thermo-hydraulic coupling model in explaining the mechanisms of crater formation in EDM.

Thermo-Fluid Verification of Fuel Column with Crossflow Gap

International Nuclear Information System (INIS)

Lee, Sung Nam; Tak, Nam Il; Kim, Min Hwan; Noh, Jae Man

2013-01-01

Korea Atomic Energy Research Institute (KAERI) has been developing thermal-hydraulic code to design a safe and effective VHTR. Core reliable Optimization and Network thermo-fluid Analysis (CORONA) is a code that solves the fluid region as 1-D and the solid domain as 3-D. The postulated event is modeled to secure safety during design process. The reactor core of VHTR is piled with multi-fuel block layers. The helium gas goes through coolant channel holes after distributed from upper plenum. The fuel blocks are irradiated during operation and there might be cross gaps between blocks. These cross gaps change the passage of coolant channels and could affect the temperature of fuel compact. Therefore, two types of single fuel assembly (i. e., standard and Reserved Shutdown Control (RSC) hole fuel assemblies) were investigated in this study. The CORONA, thermo-fluid analysis code, has been developing to compute the reactor core of VHTR. Crossflow model was applied to predict temperature and flow distribution between fuel blocks in this study. The calculated results are compared with the data of commercial software, CFX. The temperature variations along the axial direction well agree for both standard / RSC fuel assemblies. The flow redistribution due to crossflow matches well. The hot spot temperature and locations might differ depending on the cross gap size. This research will be done in detail for further study

Hydraulic Design of the CARA Fuel Assembly for Atucha-I

International Nuclear Information System (INIS)

Juanico, Luis; Brasnarof, Daniel

2000-01-01

In this paper a hydraulic model of the CARA fuel assembly within the Atucha I fuel channel is developed. Besides, a experimental test running in the CBP low pressure loop have been designed.This model is used for design purpose of the assembly system such as the whole channel pressure drop remains the same that it is at the present.It is observed that choosing the right thickness and hole surface of the assembly system, it is possible tune up the CARA pressure drop, releases the azimuth alignment condition on the fuel element neighbors

Thermo-hydraulic consequence of pressure suppression containment vessel during blowdown, 2

International Nuclear Information System (INIS)

Aya, Izuo; Nariai, Hideki; Kobayashi, Michiyuki

1980-01-01

As a part of the safety research works for the integral-type marine reactor, an analytical code SUPPAC-2V was developed to simulate the thermo-hydraulic consequence of a pressure suppression containment system during blowdown and the code was applied to the Model Experimental Facility of the Safety of Integral Type Marine Reactors (explained already in Part 1). SUPPAC-2V is much different from existing codes in the following points. A nonhomogeneous model for the gaseous region in the drywell, a new correlation for condensing heat transfer coefficient at drywell wall based on existing data and approximation of air bubbles in wetwell water by one dimensional bubble rising model are adopted in this code. In comparing calculational results with experimental results, values of predominant input parameters were evaluated and discussed. Moreover, the new code was applied also to the NSR-7 marine reactor, conceptually designed at the Shipbuilding Research Association in Japan, of which suppression system had been already analysed by CONTEMPT-PS. (author)

Thermal-hydraulic analysis of PWR small assembly for irradiation test of CARR

International Nuclear Information System (INIS)

Yin Hao; Zou Yao; Liu Xingmin

2015-01-01

The thermal-hydraulic behaviors of the PWR 4 × 4 small assembly tested in the high temperature and high pressure loop of China Advanced Research Reactor were analyzed. The CFD method was used to carry out 3D simulation of the model, thus detailed thermal-hydraulic parameters were obtained. Firstly, the simplified model was simulated to give the 3D temperature and velocity distributions and analyze the heat transfer process. Then the whole scale small assembly model was simulated and the simulation results were compared with those of simplified rod bundle model. Its flow behavior was studied and flow mixing characteristics of the grids were analyzed, and the mixing factor of the grid was calculated and can be used for further thermal-hydraulic study. It is shown that the highest temperature of the fuel rod meets the design limit and the mixing effect of the grid is obvious. (authors)

Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

Energy Technology Data Exchange (ETDEWEB)

Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-05-15

{sub th} power and electricity generation with 100 kW{sub th} idle power. Consequently, KANUTER has the characteristics of a compact and lightweight system, excellent propellant efficiency, bimodal capability, and mission versatility as indicated in the reference design parameters. This thermo-hydraulic design analysis was carried out to estimate the optimum FWT of the unique SLHC fuel design in the core and thereby the maximum rocket performance. The FWT affects the mechanical strength of the SLHC fuel assembly as well as the thermo-hydraulic capability mainly depending on the heat transfer area of fuel. The thicker fuel wafer is mechanically strong with low pressure drop, while the thinner fuel wafer is thermally robust with less mechanical strength and higher shear stress in the core.

Computer code HYDRO-ACE for analyzing thermo-hydraulic phenomena in the BWR core

International Nuclear Information System (INIS)

Abe, Kiyoharu; Naito, Yoshitaka

1979-10-01

A computer code HYDRO-ACE has been developed for analyzing thermo-hydraulic phenomena in the BWR core under forced or natural circulation of cooling water. The code is composed of two main calculation routines for single channels such as riser, separator, and downcommer and multiple channels such as the reactor core with a heated zone. Functionally the code is divided into many subroutines to be connected straightforwardly, and so that the user can choose a given course freely by simply arranging the subroutines. In the program, void fraction is calculated by Maurer's method, two-phase frictional pressure drop by Maltinelli-Nelson's, and critical heat flux ratio by Hench-Levy's. The coolant flow distributions in the JPDR-II core calculated by the code are in good agreement with those measured. (author)

Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

International Nuclear Information System (INIS)

Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

2016-01-01

To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

Thermal hydraulic evaluation of advanced wire-wrapped assemblies

International Nuclear Information System (INIS)

Wei, J.P.

1975-01-01

The thermal-hydraulic analyses presented in this report are based on application of the subchannel concept in association with the use of bulk parameters for coolant velocity and coolant temperature within a subchannel. The interactions between subchannels are due to turbulent interchange, pressure-induced diversion crossflow, directed sweeping crossflow induced by the helical wire wrap, and transverse thermal conduction. The FULMIX-II computer program was successfully developed to perform the steady-state temperature predictions for LMFBR fuel assemblies with the reference straight-start design and the advanced wire-wrap designs. Predicted steady-state temperature profiles are presented for a typical CRBRP 217-rod wire-wrapped assembly with the selected wire-wrap designs

Coupled neutronics/thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly

International Nuclear Information System (INIS)

Waata, C.L.

2006-07-01

The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence

Fabrication of Thermo-Responsive Molecular Layers from Self-Assembling Elastin-Like Oligopeptides Containing Cell-Binding Domain for Tissue Engineering

Directory of Open Access Journals (Sweden)

Tomoyuki Koga

2015-01-01

Full Text Available Novel thermo-responsive elastin-like oligopeptides containing cell-binding epitope (Arg-Gly-Asp-Ser sequence; arginine-glycine-aspartic acid-serine (RGDS-elastin-like peptides (ELP and RGDS-deg-ELP; were newly prepared as building blocks of self-assembled molecular layer for artificial extra cellular matrix. A detailed analysis of the conformation of the oligo(ELPs in water and their self-assembling behavior onto hydrophobic surfaces were performed by using circular dichroism, Fourier transform infrared spectroscopy (FTIR, atomic force microscopy and water contact angle measurements. The experimental results revealed that both oligo(ELPs self-assembled onto hydrophobic surfaces and formed molecular layers based on their thermo-responsive conformational change from hydrous random coil to dehydrated β-turn structure. Effective cell adhesion and spreading behaviors were observed on these self-assembled oligo(ELP layers. In addition, attached cells were found to be recovered successfully as a cell-sheet by temperature-induced disassembly of oligo(ELP layer. This achievement provides an important insight to construct novel oligopeptide-based nano-surfaces for the design of smart artificial extra-cellular matrix.

The delay function in finite difference models for nuclear channels thermo-hydraulic transients

International Nuclear Information System (INIS)

Agazzi, A.

1977-01-01

The study of the thermo-hydraulic transients in a nuclear reactor core often requires a bi- or tri-dimensional mathematical simulation of a reactor channel. The equations involved are generally solved by means of finite-difference methods. The determination of the spatial mesh-width and the time interval is strongly conditioned by the necessity of a good accuracy in the description of the delay function which defines the transfer of thermal perturbations along the cooling channel. In this paper the effects of both space and time discretization on the delay function are considered and for the classical cases of inlet temperature step and ramp universal functions and diagrams are given in order to make possible the determination of optimal spatial mesh-width and time interval, once the requested accuracy of the model is fixed in advance

Analysis Thermo-hydraulic of trajectories related to procedures for operation of Emergency (POE). Application to the loss of a train of the DTH; Analisis termohidraulico de trayectorias vinculadas a Procedimientos de Operacion de emergencia (POE). Aplicacion a la perdida de un tren de RHR

Energy Technology Data Exchange (ETDEWEB)

Sanchez-Saez, F.; Martorell Alsina, S.; Carlos Alberola, A.; Villanueva Lopez, J. F.; Martorell Aygues, P.

2012-07-01

This work explores different possible sequences at the loss of a train of the DTH when the plant is lowering power. The study of the different possible trajectories has been done through the collapse tool and study thermo-hydraulic each of these paths is done by the code TRACE Thermo-hydraulic.

Thermo-responsive human α-elastin self-assembled nanoparticles for protein delivery.

Science.gov (United States)

Kim, Jae Dong; Jung, Youn Jae; Woo, Chang Hee; Choi, Young Chan; Choi, Ji Suk; Cho, Yong Woo

2017-01-01

Self-assembled nanoparticles based on PEGylated human α-elastin were prepared as a potential vehicle for sustained protein delivery. The α-elastin was extracted from human adipose tissue and modified with methoxypolyethyleneglycol (mPEG) to control particle size and enhance the colloidal stability. The PEGylated human α-elastin showed sol-to-particle transition with a lower critical solution temperature (LCST) of 25°C-40°C in aqueous media. The PEGylated human α-elastin nanoparticles (PhENPs) showed a narrow size distribution with an average diameter of 330±33nm and were able to encapsulate significant amounts of insulin and bovine serum albumin (BSA) upon simple mixing at low temperature in water and subsequent heating to physiological temperature. The release profiles of insulin and BSA showed sustained release for 72h. Overall, the thermo-responsive self-assembled PhENPs provide a useful tool for a range of protein delivery and tissue engineering applications. Copyright © 2016 Elsevier B.V. All rights reserved.

Assessment of TRAC-PD2 reflood core thermo-hydraulic model by CCTF Test C1-16

International Nuclear Information System (INIS)

Sugimoto, Jun

1982-11-01

The TRAC-PD2 reflood core thermo-hydraulic model was assessed by CCTF Test C1-16. The measured data were utilized as core boundary conditions in the TRAC calculations. The results indicate that the core inlet liquid temperature and the core heater rod temperatures are in reasonable agreement with data, but the pressure distribution in the core and water pool formation in the upper plenum are not in good agreement. The parametric effects of the droplet critical Weber number, the material properties of the heater rod, the noding of the upper plenum, and the minimum stable film boiling temperature are also discussed. (author)

The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

Energy Technology Data Exchange (ETDEWEB)

Kim, K. S.; Ju, H. G.; Jeon, T. H. and others

2005-03-15

A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well.

The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

International Nuclear Information System (INIS)

Kim, K. S.; Ju, H. G.; Jeon, T. H. and others

2005-03-01

A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well

Coupled hydro-thermo-mechanical modeling of hydraulic fracturing in quasi-brittle rocks using BPM-DEM

Directory of Open Access Journals (Sweden)

Ingrid Tomac

2017-02-01

Full Text Available This paper presents an improved understanding of coupled hydro-thermo-mechanical (HTM hydraulic fracturing of quasi-brittle rock using the bonded particle model (BPM within the discrete element method (DEM. BPM has been recently extended by the authors to account for coupled convective–conductive heat flow and transport, and to enable full hydro-thermal fluid–solid coupled modeling. The application of the work is on enhanced geothermal systems (EGSs, and hydraulic fracturing of hot dry rock (HDR is studied in terms of the impact of temperature difference between rock and a flowing fracturing fluid. Micro-mechanical investigation of temperature and fracturing fluid effects on hydraulic fracturing damage in rocks is presented. It was found that fracture is shorter with pronounced secondary microcracking along the main fracture for the case when the convective–conductive thermal heat exchange is considered. First, the convection heat exchange during low-viscosity fluid infiltration in permeable rock around the wellbore causes significant rock cooling, where a finger-like fluid infiltration was observed. Second, fluid infiltration inhibits pressure rise during pumping and delays fracture initiation and propagation. Additionally, thermal damage occurs in the whole area around the wellbore due to rock cooling and cold fluid infiltration. The size of a damaged area around the wellbore increases with decreasing fluid dynamic viscosity. Fluid and rock compressibility ratio was found to have significant effect on the fracture propagation velocity.

Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

Energy Technology Data Exchange (ETDEWEB)

Han, Kyu Hyun [Korea Institute of Nuclear Safety, 19, Guseong-dong, Yuseong-gu, Daejeon, 305-338 (Korea, Republic of)

2008-07-01

Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

2015-12-31

A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

Sensitiveness Analysis of Neutronic Parameters Due to Uncertainty in Thermo-hydraulic parameters on CAREM-25 Reactor

International Nuclear Information System (INIS)

Serra, Oscar

2000-01-01

Some studies were done about the effect of the uncertainty in the values of several thermo-hydraulic parameters on the core behaviour of the CAREM-25 reactor.By using the chain codes CITVAP-THERMIT and the perturbation the reference states, it was found that concerning to the total power, the effects were not very important, but were much bigger for the pressure.Furthermore were hardly significant in the presence of any perturbation on the void fraction calculation and the fuel temperature.The reactivity and the power peaking factor had highly important changes in the case of the coolant flow.We conclude that the use of this procedure is adequate and useful to our purpose

Thermo-hydro-mechanical simulation of a 3D fractured porous rock: preliminary study of coupled matrix-fracture hydraulics

International Nuclear Information System (INIS)

Canamon, I.; Javier Elorza, F.; Ababou, R.

2007-01-01

We present a problem involving the modeling of coupled flow and elastic strain in a 3D fractured porous rock, which requires prior homogenization (up-scaling) of the fractured medium into an equivalent Darcian anisotropic continuum. The governing equations form a system of PDE's (Partial Differential Equations) and, depending on the case being considered, this system may involve two different types of 'couplings' (in a real system, both couplings (1) and (2) generally take place): 1) Hydraulic coupling in a single (no exchange) or in a dual matrix-fracture continuum (exchange); 2) Thermo-Hydro-Mechanical interactions between fluid flow, pressure, elastic stress, strain, and temperature. We present here a preliminary model and simulation results with FEMLAB R , for the hydraulic problem with anisotropic heterogeneous coefficients. The model is based on data collected at an instrumented granitic site (FEBEX project) for studying a hypothetical nuclear waste repository at the Grimsel Test Site in the Swiss Alps. (authors)

Thermo-hydro-mechanical simulation of a 3D fractured porous rock: preliminary study of coupled matrix-fracture hydraulics

Energy Technology Data Exchange (ETDEWEB)

Canamon, I.; Javier Elorza, F. [Universidad Politecnica de Madrid, Dept. de Matematica Aplicada y Metodos Informaticas, ETSI Minas (UPM) (Spain); Ababou, R. [Institut de Mecanique des Fluides de Toulouse (IMFT), 31 (France)

2007-07-01

We present a problem involving the modeling of coupled flow and elastic strain in a 3D fractured porous rock, which requires prior homogenization (up-scaling) of the fractured medium into an equivalent Darcian anisotropic continuum. The governing equations form a system of PDE's (Partial Differential Equations) and, depending on the case being considered, this system may involve two different types of 'couplings' (in a real system, both couplings (1) and (2) generally take place): 1) Hydraulic coupling in a single (no exchange) or in a dual matrix-fracture continuum (exchange); 2) Thermo-Hydro-Mechanical interactions between fluid flow, pressure, elastic stress, strain, and temperature. We present here a preliminary model and simulation results with FEMLAB{sup R}, for the hydraulic problem with anisotropic heterogeneous coefficients. The model is based on data collected at an instrumented granitic site (FEBEX project) for studying a hypothetical nuclear waste repository at the Grimsel Test Site in the Swiss Alps. (authors)

Enhancement of thermo-stability and product tolerance of Pseudomonas putida nitrile hydratase by fusing with self-assembling peptide.

Science.gov (United States)

Liu, Yi; Cui, Wenjing; Liu, Zhongmei; Cui, Youtian; Xia, Yuanyuan; Kobayashi, Michihiko; Zhou, Zhemin

2014-09-01

Self-assembling amphipathic peptides (SAPs) are the peptides that can spontaneously assemble into ordered nanostructures. It has been reported that the attachment of SAPs to the N- or C-terminus of an enzyme can benefit the thermo-stability of the enzyme. Here, we discovered that the thermo-stability and product tolerance of nitrile hydratase (NHase) were enhanced by fusing with two of the SAPs (EAK16 and ELK16). When the ELK16 was fused to the N-terminus of β-subunit, the resultant NHase (SAP-NHase-2) became an active inclusion body; EAK16 fused NHase in the N-terminus of β-subunit (SAP-NHase-1) and ELK16 fused NHase in the C-terminus of β-subunit (SAP-NHase-10) did not affect NHase solubility. Compared with the deactivation of the wild-type NHase after 30 min incubation at 50°C, SAP-NHase-1, SAP-NHase-2 and SAP-NHase-10 retained 45%, 30% and 50% activity; after treatment in the buffer containing 10% acrylamide, the wild-type retained 30% activity, while SAP-NHase-1, SAP-NHase-2 and SAP-NHase-10 retained 52%, 42% and 55% activity. These SAP-NHases with enhanced thermo-stability and product tolerance would be helpful for further industrial applications of the NHase. Copyright © 2014 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

Thermal-hydraulic and thermo-mechanical design of plasma facing components for SST-1 tokamak

International Nuclear Information System (INIS)

Chaudhuri, Paritosh; Santra, P.; Chenna Reddy, D.; Parashar, S.K.S.

2014-01-01

The Plasma Facing Components (PFCs) are one of the major sub-systems of ssT-1 tokamak. PFC of ssT-1 consisting of divertors, passive stabilizers, baffles and limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC cooling is the steady state heat removal of up to 1 MW/m 2 . The PFC has been designed to withstand the peak heat fluxes and also without significant erosion such that frequent replacement of the armor is not necessary. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to carry out the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. Thermal analysis of the PFC is carried out with the purpose of evaluating the thermal mechanical behavior of PFCs. The detailed thermal-hydraulic and thermo-mechanical designs of PFCs of ssT-1 are discussed in this paper. (authors)

Thermo-mechanical design of the SINGAP accelerator grids for ITER NB Injectors

International Nuclear Information System (INIS)

Agostinetti, P.; Dal Bello, S.; Palma, M.D.; Zaccaria, P.

2006-01-01

The SINGle Aperture - SINgle GAP (SINGAP) accelerator for ITER neutral beam injector foresees four grids for the extraction and acceleration of negative ions, instead of the seven grids of the Multi Aperture Multi Grid (MAMuG) reference configuration. Optimized geometry of the SINGAP grids (plasma, extraction, pre-acceleration, and grounded grid) was identified by CEA Association considering specific requirements for ions extraction and beam generation referring to experimental data and code simulations. This paper focuses on the thermo-hydraulic and thermo-mechanical design of the grids carried out by Consorzio RFX for the design of the first ITER NB Injector and the ITER NB Test Facility. The cooling circuit design (position and shape of the channels) and the cooling parameters (water coolant temperatures, pressure and velocity) were optimized with thermo-hydraulic and thermo-mechanical sensitivity analyses in order to satisfy the grid functional requirements (temperatures, in plane and out of plane deformations). A complete and detailed thermo-structural design assessment of the SINGAP grids was accomplished applying the structural design rules for ITER in-vessel components and considering both the reference load conditions and the maximum load provided by the power supplies. The design required a complete modelling of the grids and their support frames by means of 3D FE and CAD models. The grids were finally integrated with the support and cooling systems inside the beam source vessel. The main results of the thermo-hydraulic and thermo-mechanical analyses are presented. The open issues are then reported, mainly regarding the material properties characterization (static and fatigue tests) and the qualification of technologies for OFHC copper electro-deposition, brazing, and welding of heterogeneous materials. (author)

Preliminary CFD analysis methodology for flow in a LFR fuel assembly

International Nuclear Information System (INIS)

Catana, A.; Ioan, M.; Serbanel, M.

2013-01-01

In this paper a preliminary Computational Fluid Dynamics (CFD) analysis was performed in order to setup a methodology to be used for more complex coolant flow analysis inside ALFRED nuclear reactor fuel assembly. The core contains 171 separated fuel assembly, each consisting in a hexagonal array of 127 fuel rods. Three honey comb spacer grids are proposed along fuel rods with the aim to keep flow geometry intact during reactor operation. The main goal of this paper is to compute some hydraulic parameters: pressure, velocity, wall shear stress and turbulence parameters with and without spacer grids. In this analysis we consider an adiabatic case, so far no heat transfer is considered but we pave the road toward more complex thermo hydraulic analysis for ALFRED (LFR in general). The CAELINUX CFD distribution was used with its main components: Salome-Meca (for geometry and mesh) and Code-Saturne as mono-phase CFD solver. Paraview and Visist Postprocessors were used for data extraction and graphical displays. (authors)

Hydraulic Actuators with Autonomous Hydraulic Supply for the Mainline Aircrafts

Directory of Open Access Journals (Sweden)

I. S. Shumilov

2014-01-01

pipelines, as well as their increasing reliability. It is also possible, in addition, in addition to increase reliability of the remained pipelines, having applied the last developments, e.g. introduction of one-piece connections (thermo-mechanical ones, high-strength steels for pipelines with σв˃85 кг/мм 2 σ to increase control of residual assembly tension, and so on;- to eliminate essentially all the shortcomings of hydraulic actuators, which constrain their introduction in aircraft industry;- to simplify essentially steering drive structures and designs, which allow to apply the tried and tested components and principles;- to simplify essentially a solution for cooling of working liquid;- to simplify essentially a solution for the steering drive configuration in a zone of control vanes;- to simplify essentially a solution for meeting requirements for dynamic rigidity and dynamic sensitivity of hydraulic actuators;- to simplify essentially a solution for the aircraft fire safety, etc.

Thermo-mechanical behavior of power electronic packaging assemblies: From characterization to predictive simulation of lifetimes

Science.gov (United States)

Dalverny, O.; Alexis, J.

2018-02-01

This article deals with thermo-mechanical behavior of power electronic modules used in several transportation applications as railway, aeronautic or automotive systems. Due to a multi-layered structures, involving different materials with a large variation of coefficient of thermal expansion, temperature variations originated from active or passive cycling (respectively from die dissipation or environmental constraint) induces strain and stresses field variations, giving fatigue phenomenon of the system. The analysis of the behavior of these systems and their dimensioning require the implementation of complex modeling strategies by both the multi-physical and the multi-scale character of the power modules. In this paper we present some solutions for studying the thermomechanical behavior of brazed assemblies as well as taking into account the interfaces represented by the numerous metallizations involved in the process assembly.

Development of whole core thermal-hydraulic analysis program ACT. 4. Simplified fuel assembly model and parallelization by MPI

International Nuclear Information System (INIS)

Ohshima, Hiroyuki

2001-10-01

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 the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)

A Thermo-Hydraulic Tool for Automatic Virtual Hazop Evaluation

Directory of Open Access Journals (Sweden)

Pugi L.

2014-12-01

Full Text Available Development of complex lubrication systems in the Oil&Gas industry has reached high levels of competitiveness in terms of requested performances and reliability. In particular, the use of HazOp (acronym of Hazard and Operability analysis represents a decisive factor to evaluate safety and reliability of plants. The HazOp analysis is a structured and systematic examination of a planned or existing operation in order to identify and evaluate problems that may represent risks to personnel or equipment. In particular, P&ID schemes (acronym of Piping and Instrument Diagram according to regulation in force ISO 14617 are used to evaluate the design of the plant in order to increase its safety and reliability in different operating conditions. The use of a simulation tool can drastically increase speed, efficiency and reliability of the design process. In this work, a tool, called TTH lib (acronym of Transient Thermal Hydraulic Library for the 1-D simulation of thermal hydraulic plants is presented. The proposed tool is applied to the analysis of safety relevant components of compressor and pumping units, such as lubrication circuits. Opposed to the known commercial products, TTH lib has been customized in order to ease simulation of complex interactions with digital logic components and plant controllers including their sensors and measurement systems. In particular, the proposed tool is optimized for fixed step execution and fast prototyping of Real Time code both for testing and production purposes. TTH lib can be used as a standard SimScape-Simulink library of components optimized and specifically designed in accordance with the P&ID definitions. Finally, an automatic code generation procedure has been developed, so TTH simulation models can be directly assembled from the P&ID schemes and technical documentation including detailed informations of sensor and measurement system.

Inkjet-Printed Biofunctional Thermo-Plasmonic Interfaces for Patterned Neuromodulation.

Science.gov (United States)

Kang, Hongki; Lee, Gu-Haeng; Jung, Hyunjun; Lee, Jee Woong; Nam, Yoonkey

2018-02-27

Localized heat generation by the thermo-plasmonic effect of metal nanoparticles has great potential in biomedical engineering research. Precise patterning of the nanoparticles using inkjet printing can enable the application of the thermo-plasmonic effect in a well-controlled way (shape and intensity). However, a universally applicable inkjet printing process that allows good control in patterning and assembly of nanoparticles with good biocompatibility is missing. Here we developed inkjet-printing-based biofunctional thermo-plasmonic interfaces that can modulate biological activities. We found that inkjet printing of plasmonic nanoparticles on a polyelectrolyte layer-by-layer substrate coating enables high-quality, biocompatible thermo-plasmonic interfaces across various substrates (rigid/flexible, hydrophobic/hydrophilic) by induced contact line pinning and electrostatically assisted nanoparticle assembly. We experimentally confirmed that the generated heat from the inkjet-printed thermo-plasmonic patterns can be applied in micrometer resolution over a large area. Lastly, we demonstrated that the patterned thermo-plasmonic effect from the inkjet-printed gold nanorods can selectively modulate neuronal network activities. This inkjet printing process therefore can be a universal method for biofunctional thermo-plasmonic interfaces in various bioengineering applications.

Application of hydraulically assembled shaft coupling hubs to large agitators

International Nuclear Information System (INIS)

Murray, W.E.; Anderson, T.D.; Bethmann, H.K.

1991-01-01

This paper describes the basis for and implementation of hydraulically assembled shaft coupling hubs for large tank-mounted agitators. This modification to the original design was intended to minimize maintenance personnel exposure to ionizing radiation and also provide for disassembly capability without damage to shafts or hubs. In addition to realizing these objectives, test confirmed that the modified couplings reduced agitator shaft end runouts approximately 65%, thereby reducing bearing loads and increasing service life, a significant enhancement for a nuclear facility. 5 refs

Thermo-controlled rheology of electro-assembled polyanionic polysaccharide (alginate) and polycationic thermo-sensitive polymers.

Science.gov (United States)

Niang, Pape Momar; Huang, Zhiwei; Dulong, Virginie; Souguir, Zied; Le Cerf, Didier; Picton, Luc

2016-03-30

Several thermo-sensitive polyelectrolyte complexes were prepared by ionic self-association between an anionic polysaccharide (alginate) and a monocationic copolymer (polyether amine, Jeffamine®-M2005) with a 'Low Critical Solubility Temperature' (LCST). We show that electro-association must be established below the aggregation temperature of the free Jeffamine®, after which the organization of the system is controlled by the thermo-association of Jeffamine® that was previously electro-associated with the alginate. Evidence for this comes primarily from the rheology in the semi-dilute region. Electro- and thermo-associative behaviours are optimal at a pH corresponding to maximum ionization of both compounds (around pH 7). High ionic strength could prevent the electro-association. The reversibility of the transition is possible only at temperatures lower than the LCST of Jeffamine®. Similar behaviour has been obtained with carboxymethyl cellulose (CMC), which suggests that this behaviour can be observed using a range of anionic polyelectrolytes. In contrast, no specific properties have been found for pullulan, which is a neutral polysaccharide. Copyright © 2015 Elsevier Ltd. All rights reserved.

Parametric study of the stability properties of a thermo hydraulic channel coupled to punctual kinetics

International Nuclear Information System (INIS)

Cecenas F, M.; Campos G, R.M.

2005-01-01

The reason of decay is the indicator of stability usually used in the literature to evaluate stability of boiling water reactors, however, in the operation of this type of reactors is considered the length of boiling like an auxiliary parameter for the evaluation of stability. In this work its are studied the variation of these two indicators when modifying a given an operation parameter in a model of a thermo hydraulic channel coupled to punctual kinetics, maintaining all the other input constant variables. The parameters selected for study are the axial profile of power, the subcooling, the flow of coolant and the thermal power. The study is supplemented by means of real data of plant using the one Benchmark of Ringhals, and the results for the case of the ratio of decay its are compared with the decay reasons obtained by means of autoregression models of the local instrumentation of neutron flux. (Author)

CRAB-II: a computer program to predict hydraulics and scram dynamics of LMFBR control assemblies and its validation

International Nuclear Information System (INIS)

Carelli, M.D.; Baker, L.A.; Willis, J.M.; Engel, F.C.; Nee, D.Y.

1982-01-01

This paper presents an analytical method, the computer code CRAB-II, which calculates the hydraulics and scram dynamics of LMFBR control assemblies of the rod bundle type and its validation against prototypic data obtained for the Clinch River Breeder Reactor (CRBR) primary control assemblies. The physical-mathematical model of the code is presented, followed by a description of the testing of prototypic CRBR control assemblies in water and sodium to characterize, respectively, their hydraulic and scram dynamics behavior. Comparison of code predictions against the experimental data are presened in detail; excellent agreement was found. Also reported are experimental data and empirical correlations for the friction factor of the absorber bundle in the entire flow range (laminar to turbulent) which represent an extension of the state-of-the-art, since only fuel and blanket assemblies friction factor correlations were previously reported in the open literature

ASCOT-1: a computer program for analyzing the thermo-hydraulic behavior in a PWR core during a LOCA

International Nuclear Information System (INIS)

Kobayashi, Kensuke; Sato, Kazuo

1978-09-01

A digital computer code ASCOT-1 has been developed to analyze the thermo-hydraulic behavior in a PWR core during a loss-of-coolant accident. The core is assumed to be axi-symmetric two-dimensional and the conservation laws are solved by the method of characteristics. For the temperature response of representative fuels of the concentric annular subregions into which the core is divided, the heat conduction equations are solved by the explicit method with the averaged flow conditions decided above. The boundary conditions at the upper and lower plenum are given as inputs. The program is of an adjustable dimension so there are no restrictions to the numbers of meshes. ASCOT-1 is written in FORTRAN-IV for FACOM230-75. (author)

Modifications in Compacted MX-80 Bentonite Due to Thermo-Hydraulic Treatment; Modificaciones en la Bentonita MX-80 Compactada Sometida a Tratamiento Termo-Hidraulico

Energy Technology Data Exchange (ETDEWEB)

Gomez-Espina, R.; Villar, M. V.

2013-09-01

The thermo-hydraulic tests reproduce the thermal and hydraulic conditions to which bentonite is subjected in the engineered barrier of a deep geological repository of radioactive waste. The results of thermo-hydraulic test TBT1500, which was running for approximately 1500 days, are presented. This is a continuation to the Technical Report Ciemat 1199, which presented results of test TBT500, performed under similar conditions but with duration of 500 days. In both tests the MX-80 bentonite was used with initial density and water content similar to those of the large-scale test TBT. The bentonite column was heated at the bottom at 140 degree centigrade and hydrated on top with deionized water. At the end of the test a sharp water content gradient was observed along the column, as well as an inverse dry density gradient. Hydration modified also the bentonite microstructure. Besides, an overall decrease of the smectite content with respect to the initial value took place, especially in the most hydrated areas where the percentage of interest ratified illite increased and in the longer test. On the other hand, the content of cristobalite, feldspars and calcite increased. Smectite dissolution processes (probably colloidal) occurred, particularly in the more hydrated areas and in the longer test. Due to the dissolution of low-solubility species and to the loss of exchangeable positions in the smectite, the content of soluble salts in the pore water increased with respect to the original one, especially in the longer test. The solubilized ions were transported; sodium, calcium, magnesium and sulphate having a similar mobility, which was in turn lower than that of potassium and chloride. The cationic exchange complex was also modified. (Author)

Thermal-hydraulics associated with nuclear education and research

International Nuclear Information System (INIS)

Yokobori, Seiichi

2011-01-01

This article was the rerecording of the author's lecture at the fourth 'Future Energy Forum' (aiming at improving nuclear safety and economics) held in December 2010. The lecture focused on (1) importance of thermal hydraulics associated with nuclear education and research (critical heat flux, two-phase flow and multiphase flow), (2) emerging trend of maintenance engineering (fluid induced vibration, flow accelerated corrosion and stress corrosion cracks), (3) fostering sensible nuclear engineer with common engineering sense, (4) balanced curriculum of basics and advanced research, (5) computerized simulation and fluid mechanics, (6) crucial point of thermo hydraulics education (viscosity, flux, steam and power generation), (7) safety education and human resources development (indispensable technologies such as defence in depth) and (8) topics of thermo hydraulics research (vortices of curbed pipes and visualization of two-phase flow). (T. Tanaka)

Thermo-hydraulic test of the moderator cell of liquid hydrogen cold neutron source for the Budapest research reactor

International Nuclear Information System (INIS)

Grosz, Tamas; Rosta, Laszlo; Hargitai, Tibor; Mityukhlyaev, V.A.; Serebrov, A.P.; Zaharov, A.A.

1999-01-01

Thermo-hydraulic experiment was carried out in order to test performance of the direct cooled liquid hydrogen moderator cell to be installed at the research reactor of the Budapest Neutron Center. Two electric hearers up to 300 W each imitated the nuclear heat release in the liquid hydrogen as well as in construction material. The test moderator cell was also equipped with temperature gauges to measure the hydrogen temperature at different positions as well as the inlet and outlet temperature of cooling he gas. The hydrogen pressure in the connected buffer volume was also controlled. At 140 w expected total heat load the moderator cell was filled with liquid hydrogen within 4 hours. The heat load and hydrogen pressure characteristics of the moderator cell are also presented. (author)

Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

Energy Technology Data Exchange (ETDEWEB)

Arndt, S.A.

1997-07-01

The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for code use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities.

Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

International Nuclear Information System (INIS)

Arndt, S.A.

1997-01-01

The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for code use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities

Related research with thermo hydraulics safety by means of Trace code; Investigaciones relacionadas con seguridad termohidraulica con el codigo TRACE

Energy Technology Data Exchange (ETDEWEB)

Chaparro V, F. J.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico); Rodriguez H, A.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Sanchez E, V. H.; Jager, W., E-mail: evalle@esfm.ipn.mx [Karlsruhe Institute of Technology, Hermann-von-Helmholtz Platz I, D-76344 Eggenstein - Leopoldshafen (Germany)

2014-10-15

In this article the results of the design of a pressure vessel of a BWR/5 similar to the type of Laguna Verde NPP are presented, using the Trace code. A thermo hydraulics Vessel component capable of simulating the behavior of fluids and heat transfer that occurs within the reactor vessel was created. The Vessel component consists of a three-dimensional cylinder divided into 19 axial sections, 4 azimuthal sections and two concentric radial rings. The inner ring is used to contain the core and the central part of the reactor, while the outer ring is used as a down comer. Axial an azimuthal divisions were made with the intention that the dimensions of the internal components, heights and orientation of the external connections match the reference values of a reactor BWR/5 type. In the model internal components as, fuel assemblies, steam separators, jet pumps, guide tubes, etc. are included and main external connections as, steam lines, feed-water or penetrations of the recirculation system. The model presents significant simplifications because the object is to keep symmetry between each azimuthal section of the vessel. In most internal components lack a detailed description of the geometry and initial values of temperature, pressure, fluid velocity, etc. given that it only considered the most representative data, however with these simulations are obtained acceptable results in important parameters such as the total flow through the core, the pressure in the vessel, percentage of vacuums fraction, pressure drop in the core and the steam separators. (Author)

Integrated Modeling and Experiments to Characterize Coupled Thermo-hydro-geomechanical-chemical processes in Hydraulic Fracturing

Science.gov (United States)

Viswanathan, H. S.; Carey, J. W.; Karra, S.; Porter, M. L.; Rougier, E.; Kang, Q.; Makedonska, N.; Hyman, J.; Jimenez Martinez, J.; Frash, L.; Chen, L.

2015-12-01

Hydraulic fracturing phenomena involve fluid-solid interactions embedded within coupled thermo-hydro-mechanical-chemical (THMC) processes over scales from microns to tens of meters. Feedbacks between processes result in complex dynamics that must be unraveled if one is to predict and, in the case of unconventional resources, facilitate fracture propagation, fluid flow, and interfacial transport processes. The proposed work is part of a broader class of complex systems involving coupled fluid flow and fractures that are critical to subsurface energy issues, such as shale oil, geothermal, carbon sequestration, and nuclear waste disposal. We use unique LANL microfluidic and triaxial core flood experiments integrated with state-of-the-art numerical simulation to reveal the fundamental dynamics of fracture-fluid interactions to characterize the key coupled processes that impact hydrocarbon production. We are also comparing CO2-based fracturing and aqueous fluids to enhance production, greatly reduce waste water, while simultaneously sequestering CO2. We will show pore, core and reservoir scale simulations/experiments that investigate the contolling mechanisms that control hydrocarbon production.

Cradle modification for hydraulic ram

International Nuclear Information System (INIS)

Koons, B.M.

1995-01-01

The analysis of the cradle hydraulic system considers stress, weld strength, and hydraulic forces required to lift and support the cradle/pump assembly. The stress and weld strength of the cradle modifications is evaluated to ensure that they meet the requirements of the American Institute for Steel Construction (AISC 1989). The hydraulic forces are evaluated to ensure that the hydraulic system is capable of rotating the cradle and pump assembly to the vertical position (between 70 degrees and 90 degrees)

Thermo-mechanical analysis of a user filter assembly for undulator/wiggler operations at the Advanced Photon Source

International Nuclear Information System (INIS)

Nian, H.L.T.; Kuzay, T.M.; Collins, J.; Shu, D.; Benson, C.; Dejus, R.

1996-01-01

This paper reports a thermo-mechanical study of a beamline filter (user filter) for undulator/wiggler operations. It is deployed in conjunction with the current commissioning window assembly on the APS insertion device (ID) front ends. The beamline filter at the Advanced Photon Source (APS) will eventually be used in windowless operations also. Hence survival and reasonable life expectancy of the filters under intense insertion device (ID) heat flu are crucial to the beamline operations. To accommodate various user requirements, the filter is configured to be a multi-choice type and smart to allow only those filter combinations that will be safe to operate with a given ring current and beamline insertion device gap. However, this paper addresses only the thermo-mechanical analysis of individual filter integrity and safety in all combinations possible. The current filter design is configured to have four filter frames in a cascade with each frame holding five filters. This allows a potential 625 total filter combinations. Thermal analysis for all of these combinations becomes a mammoth task considering the desired choices for filter materials (pyrolitic graphite and metallic filters), filter thicknesses, undulator gaps, and the beam currents. The paper addresses how this difficult task has been reduced to a reasonable effort and computational level. Results from thermo-mechanical analyses of the filter combinations are presented both in tabular and graphical format

Thermo-hydraulic-mechanical analysis of the SS-050 sodium loop during a thermal shock of 2000C/s

International Nuclear Information System (INIS)

Jesus Miranda, C.A. de; Gebrin, A.N.

1988-01-01

An analytical thermo-hydraulic model was developed to obtain the temperature of the sodium flowing between the mixing tank TM of constant volume and the drain tank of the SS-050 sodium test facility. The piping connecting these two tanks is considered in the analysis. The sodium enters in the TM through a tube with lateral holes immersed in the TM's sodium. The model and relative computer program were tested and a typical situation was studied: a thermal shock with -200 0 C/s of thermal gradient in the test section. The sodium temperature time-histories along the piping length are presented. For the thermal shock situation, the temperature field in the TM bottom and outlet nozzle was calculated and the stresses were evaluated. The final thermal stresses will allow a detailed verification of the circuit design. (author) [pt

Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

Energy Technology Data Exchange (ETDEWEB)

Mays, Brian [AREVA Federal Services, Lynchburg, VA (United States); Jackson, R. Brian [TerraPower, Bellevue, WA (United States)

2017-03-08

The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services. The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.

Thermo-hydraulic characterization of a self-pumping corrugated wall heat exchanger

International Nuclear Information System (INIS)

Schmidmayer, Kevin; Kumar, Prashant; Lavieille, Pascal; Miscevic, Marc; Topin, Frédéric

2017-01-01

Compactness, efficiency and thermal control of the heat exchanger are of critical significance for many electronic industry applications. In this view, a new concept of heat exchanger at millimeter scale is proposed and numerically studied. It consists in dynamically deforming at least one of its walls by a progressive wave in order to create an active corrugated channel. Systematic studies were performed in single-phase flow on the different deformation parameters that allow obtaining the thermo-hydraulic characteristics of the system. It has been observed the dynamic wall deformation induces a significant pumping effect. Intensification of heat transfer remains very important even for highly degraded waveforms although the pumping efficiency is reduced in this case. The mechanical power applied on the upper wall to deform it dynamically is linked to the wave shape, amplitude, frequency and outlet-inlet pressure difference. The overall performance of the proposed system has been evaluated and compared to existing static channels. The performance of the proposed heat exchanger evolved in two steps for a given wall deformation. It declines slightly up to a critical value of mechanical power applied on the wall. When this critical value is exceeded, it deteriorates significantly, reaching the performance of existing conventional systems. - Highlights: • A new concept of heat exchanger within channel at millimeter scale is proposed. • Upper wall is deformed dynamically by applying external mechanical power. • Pumping effect is observed and is linked to the wave shape, amplitude and frequency. • Efficient proposed system in low Reynolds number range. • Overall performance is significantly high compared to static corrugated and straight channels.

Influence of the Lubricant Thermo-Piezo-Viscous Property on Hydrostatic Bearings in Oil Hydraulics

DEFF Research Database (Denmark)

Johansen, Per; Roemer, Daniel Beck; Andersen, Torben O.

2016-01-01

adds to the discrepancy of such simple design approach. In this paper the hydrostatic pressure force calculation is reviewed in terms of thermohydrodynamic (THD) lubrication theory, and simple analytical approximations of the hydrostatic pressure force, incorporating the piezo-viscous and thermo...... of these analytical approximations are explored in order to clarify the limits of application. In conclusion, it is found that the spatial gradient of the thermal field on the bearing surface is the significant factor in the thermo-viscous effect on the hydrostatic pressure profile, which leads to the conclusion...... that design engineers need to understand the thermodynamics of hydrostatic bearings, when using the conventional simple analytical approach, neglecting thermo-piezo-viscosity, in hydrostatic pressure force calculations....

Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC; Desarrollo de un modelo del NSSS de un reactor PWR con el codigo termo-hidraulico GOTHIC

Energy Technology Data Exchange (ETDEWEB)

Gomez Garcia-Torano, I.; Jimenez, G.

2013-07-01

The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

Modeling and Parameter Identification of the Vibration Characteristics of Armature Assembly in a Torque Motor of Hydraulic Servo Valves under Electromagnetic Excitations

Directory of Open Access Journals (Sweden)

Jinghui Peng

2014-07-01

Full Text Available The resonance of the armature assembly is the main problem leading to the fatigue of the spring pipe in a torque motor of hydraulic servo valves, which can cause the failure of servo valves. To predict the vibration characteristics of the armature assembly, this paper focuses on the mathematical modeling of the vibration characteristics of armature assembly in a hydraulic servo valve and the identification of parameters in the models. To build models more accurately, the effect of the magnetic spring is taken into account. Vibration modal analysis is performed to obtain the mode shapes and natural frequencies, which are necessary to implement the identification of damping ratios in the mathematical models. Based on the mathematical models for the vibration characteristics, the harmonic responses of the armature assembly are analyzed using the finite element method and measured under electromagnetic excitations. The simulation results agree well with the experimental studies.

Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

Directory of Open Access Journals (Sweden)

Skrzypek Maciej

2015-09-01

Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

Hydro-methane and methanol combined production from hydroelectricity and biomass: Thermo-economic analysis in Paraguay

International Nuclear Information System (INIS)

Rivarolo, M.; Bellotti, D.; Mendieta, A.; Massardo, A.F.

2014-01-01

Highlights: • We investigate H 2 /O 2 production from large hydraulic plant by water electrolysis. • We produce methanol and hydro-methane from H 2 /O 2 obtained. • We investigate two different configurations of the plant. • We perform a thermo-economic analysis for three scenarios in Paraguay. • We find plants optimal size using a time-dependent thermo-economic approach. - Abstract: A thermo-economic analysis regarding large scale hydro-methane and methanol production from renewable sources (biomass and renewable electricity) is performed. The study is carried out investigating hydrogen and oxygen generation by water electrolysis, mainly employing the hydraulic energy produced from the 14 GW Itaipu Binacional Plant, owned by Paraguay and Brazil. Oxygen is employed in biomass gasification to synthesize methanol; the significant amount of CO 2 separated in the process is mixed with hydrogen produced by electrolysis in chemical reactors to produce hydro-methane. Hydro-methane is employed to supply natural gas vehicles in Paraguay, methanol is sold to Brazil, that is the largest consumer in South America. The analysis is performed employing time-dependent hydraulic energy related to the water that would normally not be used by the plant, named “spilled energy”, when available; in the remaining periods, electricity is acquired at higher cost by the national grid. For the different plant lay-outs, a thermo-economic analysis has been performed employing two different software, one for the design point and one for the time-dependent one entire year optimization, since spilled energy is strongly variable throughout the year. Optimal sizes for the generation plants have been determined, investigating the influence of electricity cost, size and plant configuration

Thermo-hydraulic modelling of the South East Gas Pipeline System - an integrated model; Modelagem termo-hidraulica do Sistema de Gasodutos do Sudeste : um modelo integrado

Energy Technology Data Exchange (ETDEWEB)

Vianna Neto, Armando M.; Santos, Arnaldo M.; Mercon, Eduardo G. [TRANSPETRO - PETROBRAS Transportes, Rio de Janeiro, RJ (Brazil)

2003-07-01

This paper presents the development of an integrated simulation model, for the numerical calculation of thermal-hydraulic behaviors in the Brazilian southeast onshore gas pipeline flow system, remotely operated by TRANSPETRO's Gas Pipeline Control Centre (CCG). In its final application, this model is supposed to provide simulated results at the closer range to reality, in order to improve gas pipeline simulation studies and evaluations for the system in question. Considering the fact that numerical thermo-hydraulic simulation becomes the CCG's most important tool to analyze the boundary conditions to adjust the mentioned gas flow system, this paper seeks and takes aim to the optimization of the following prime attributions of a gas pipeline control centre: verification of system behaviors, face to some unit maintenance stop or procedure, programmed or not, or to some new gas outlet or inlet connection to the system; daily operational compatibility analysis between programmed and realized gas volumes; gas technical expedition and delivery analysis. Finally, all this work was idealized and carried out within the one-phase flow domain (dry gas) (author)

Layer-by-Layer Assembly of Biopolyelectrolytes onto Thermo/pH-Responsive Micro/Nano-Gels

Directory of Open Access Journals (Sweden)

Ana M. Díez-Pascual

2014-11-01

Full Text Available This review deals with the layer-by-layer (LbL assembly of polyelectrolyte multilayers of biopolymers, polypeptides (i.e., poly-l-lysine/poly-l-glutamic acid and polysaccharides (i.e., chitosan/dextran sulphate/sodium alginate, onto thermo- and/or pH-responsive micro- and nano-gels such as those based on synthetic poly(N-isopropylacrylamide (PNIPAM and poly(acrylic acid (PAA or biodegradable hyaluronic acid (HA and dextran-hydroxyethyl methacrylate (DEX-HEMA. The synthesis of the ensembles and their characterization by way of various techniques is described. The morphology, hydrodynamic size, surface charge density, bilayer thickness, stability over time and mechanical properties of the systems are discussed. Further, the mechanisms of interaction between biopolymers and gels are analysed. Results demonstrate that the structure and properties of biocompatible multilayer films can be finely tuned by confinement onto stimuli-responsive gels, which thus provides new perspectives for biomedical applications, particularly in the controlled release of biomolecules, bio-sensors, gene delivery, tissue engineering and storage.

Thermo-Hydraulic Analysis of Heat Storage Filled with the Ceramic Bricks Dedicated to the Solar Air Heating System.

Science.gov (United States)

Nemś, Magdalena; Nemś, Artur; Kasperski, Jacek; Pomorski, Michał

2017-08-12

This article presents the results of a study into a packed bed filled with ceramic bricks. The designed storage installation is supposed to become part of a heating system installed in a single-family house and eventually to be integrated with a concentrated solar collector adapted to climate conditions in Poland. The system's working medium is air. The investigated temperature ranges and air volume flow rates in the ceramic bed were dictated by the planned integration with a solar air heater. Designing a packed bed of sufficient parameters first required a mathematical model to be constructed and heat exchange to be analyzed, since heat accumulation is a complex process influenced by a number of material properties. The cases discussed in the literature are based on differing assumptions and different formulas are used in calculations. This article offers a comparison of various mathematical models and of system operating parameters obtained from these models. The primary focus is on the Nusselt number. Furthermore, in the article, the thermo-hydraulic efficiency of the investigated packed bed is presented. This part is based on a relationship used in solar air collectors with internal storage.

Thermo-Hydraulic Analysis of Heat Storage Filled with the Ceramic Bricks Dedicated to the Solar Air Heating System

Science.gov (United States)

Nemś, Magdalena; Nemś, Artur; Kasperski, Jacek; Pomorski, Michał

2017-01-01

This article presents the results of a study into a packed bed filled with ceramic bricks. The designed storage installation is supposed to become part of a heating system installed in a single-family house and eventually to be integrated with a concentrated solar collector adapted to climate conditions in Poland. The system’s working medium is air. The investigated temperature ranges and air volume flow rates in the ceramic bed were dictated by the planned integration with a solar air heater. Designing a packed bed of sufficient parameters first required a mathematical model to be constructed and heat exchange to be analyzed, since heat accumulation is a complex process influenced by a number of material properties. The cases discussed in the literature are based on differing assumptions and different formulas are used in calculations. This article offers a comparison of various mathematical models and of system operating parameters obtained from these models. The primary focus is on the Nusselt number. Furthermore, in the article, the thermo-hydraulic efficiency of the investigated packed bed is presented. This part is based on a relationship used in solar air collectors with internal storage. PMID:28805703

Thermo-hydraulic simulations of the experimental fast reactor core

International Nuclear Information System (INIS)

Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

1985-01-01

A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author) [pt

Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

International Nuclear Information System (INIS)

Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

1983-01-01

The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

Thermal-hydraulic calculation and analysis on helium cooled ceramic breeder pebble bed assembly for in-pile irradiation and in-situ tritium extraction

International Nuclear Information System (INIS)

Guo Chunqiu; Xie Jiachun; Liu Xingmin

2013-01-01

In-pile irradiation and in-situ tritium extraction experiment is one of associated domestic research projects in ITER special program. According to the technical requirements of in-pile irradiation experiment of helium cooled ceramic breeder (ceramic) pebble bed assembly in a research reactor, the feasibility of the design for the in-pile irradiation and in-situ tritium extraction experiment of ceramic pebble bed assembly was evaluated. By conducting thermal-hydraulic design calculation with different in-pile irradiation channels, locations and structure parameters for ceramic pebble bed assembly, a reasonable design scheme of ceramic pebble bed assembly satisfying the design requirements for in-pile irradiation was obtained. (authors)

Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Saxena, Aakanksha

2014-01-01

The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR). First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry. The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (∼20 C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (∼1 C-2 C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10 Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures. (author) [fr

Thermo hydraulic analysis of narrow channel effect in supercritical-pressure light water reactor

International Nuclear Information System (INIS)

Zhou Tao; Chen Juan; Cheng Wanxu

2012-01-01

Highlights: ► Detailed thermal analysis with different narrow gaps between fuel rods is given. ► Special characteristics of narrow channels effect on heat transfer in supercritical pressure are shown. ► Reasonable size selection of gaps between fuel rods is proposed for SCWR. - Abstract: The size of the gap between fuel rods has important effects on flow and heat transfer in a supercritical-pressure light water reactor. Based on thermal analysis at different coolant flow rates, the reasonable value range of gap size between fuel rods is obtained, for which the maximum cladding temperature safety limits and installation technology are comprehensively considered. Firstly, for a given design flow rate of coolant, thermal hydraulic analysis of supercritical pressure light water reactor with different gap sizes is provided by changing the fuel rod pitch only. The results show that, by means of reducing the gap size between fuel rods, the heat transfer coefficients between coolant and fuel rod, as well as the heat transfer coefficient between coolant and water rod, would both increase noticeably. Furthermore, the maximum cladding temperature will significantly decrease when the moderator temperature is decreased but coolant temperature remains essentially constant. Meanwhile, the reduction in the maximum cladding temperature in the inner assemblies is much larger than that in the outer assemblies. In addition, the maximum cladding temperature could be further reduced by means of increasing coolant flow rate for each gap size. Finally, the characteristics of narrow channels effect are proposed, and the maximum allowable gap between fuel rods is obtained by making full use of the enhancing narrow channels effect on heat transfer, and concurrently considering installation. This could provide a theoretical reference for supercritical-pressure light water reactor design optimization, in which the effects of gap size and flow rate on heat transfer are both considered.

Thermo-mechanical analysis of RMP coil system for EAST tokamak

International Nuclear Information System (INIS)

Wang, Songke; Ji, Xiang; Song, Yuntao; Zhang, Shanwen; Wang, Zhongwei; Sun, Youwen; Qi, Minzhong; Liu, Xufeng; Wang, Shengming; Yao, Damao

2014-01-01

Highlights: • Thermal design requirements for EAST RMP coils are summarized. • Cooling parameters based on both theoretical and numerical solutions are determined. • Compromise between thermal design and structural design is made on number of turns. • Thermo-mechanical calculations are made to validate its structural performance. - Abstract: Resonant magnetic perturbation (RMP) has been proved to be an efficient approach on edge localized modes (ELMs) control, resistive wall mode (RWM) control, and error field correction (EFC), RMP coil system under design in EAST tokamak will realize the above-mentioned multi-functions. This paper focuses on the thermo-mechanical analysis of EAST RMP coil system on the basis of sensitivity analysis, both normal and off-normal working conditions are considered. The most characteristic set of coil system is chosen with a complete modelling by means of three-dimensional (3D) finite element method, thermo-hydraulic and thermal-structural performances are investigated adequately, both locally and globally. The compromise is made between thermal performance and structural design requirements, and the results indicate that the optimized design is feasible and reasonable

Support of Construction and Verification of Out-of-Pile Fuel Assembly Test Facilities

International Nuclear Information System (INIS)

Park, Nam Gyu; Kim, K. T.; Park, J. K.

2006-12-01

Fuel assembly and components should be verified by the out-of-pile test facilities in order to load the developed fuel in reactor. Even though most of the component-wise tests have been performed using the facilities in land, the assembly-wise tests has been depended on the oversees' facility due to the lack of the facilities. KAERI started to construct the assembly-wise mechanical/hydraulic test facilities and KNF, as an end user, is supporting the mechanical/hydraulic test facility construction by using the technologies studied through the fuel development programs. The works performed are as follows: - Test assembly shipping container design and manufacturing support - Fuel handling tool design : Gripper, Upper and lower core simulators for assembly mechanical test facility, Internals for assembly hydraulic test facility - Manufacture of test specimens : skeleton and assembly for preliminary functional verification of assembly mechanical/hydraulic test facilities, two assemblies for the verification of assembly mechanical/hydraulic test facilities, Instrumented rod design and integrity evaluation - Verification of assembly mechanical/hydraulic test facilities : test data evaluation

Support of Construction and Verification of Out-of-Pile Fuel Assembly Test Facilities

Energy Technology Data Exchange (ETDEWEB)

Park, Nam Gyu; Kim, K. T.; Park, J. K. [KNF, Daejeon (Korea, Republic of)] (and others)

2006-12-15

Fuel assembly and components should be verified by the out-of-pile test facilities in order to load the developed fuel in reactor. Even though most of the component-wise tests have been performed using the facilities in land, the assembly-wise tests has been depended on the oversees' facility due to the lack of the facilities. KAERI started to construct the assembly-wise mechanical/hydraulic test facilities and KNF, as an end user, is supporting the mechanical/hydraulic test facility construction by using the technologies studied through the fuel development programs. The works performed are as follows: - Test assembly shipping container design and manufacturing support - Fuel handling tool design : Gripper, Upper and lower core simulators for assembly mechanical test facility, Internals for assembly hydraulic test facility - Manufacture of test specimens : skeleton and assembly for preliminary functional verification of assembly mechanical/hydraulic test facilities, two assemblies for the verification of assembly mechanical/hydraulic test facilities, Instrumented rod design and integrity evaluation - Verification of assembly mechanical/hydraulic test facilities : test data evaluation.

Nupec thermal hydraulic test to evaluate post-DNB characteristics for PWR fuel assemblies (1. general test plan and results)

International Nuclear Information System (INIS)

Norio, Kono; Kenji, Murai; Kaichiro, Misima; Takayuki, Suemura; Yoshiei, Akiyama; Keiichi, Hori

2001-01-01

In the present thermal hydraulic design of Pressurized Water Reactor (PWR), a departure from nucleate boiling (DNB) under anticipated transient conditions is not allowed. However, it is recognized that the DNB dose not cause a fuel rod failure immediately, and a suitable reactor trip can prevent the core from severe damages. If the fuel rod temperature under the post-DNB conditions can be accurately evaluated, the potentially existing margin in the present design method will be quantitatively assessed. To establish the heat transfer evaluation method on post-DNB event for PWR thermal hydraulic design, Nuclear Power Engineering Corporation (NUPEC) started a program, NUPEC Thermal Hydraulic Test to Evaluate Post-DNB Characteristics for PWR Fuel Assemblies (NUPEC-TH-P), in 1995 (hereinafter the year means fiscal year) under the sponsorship of Ministry of Economy, Trade and industry (METI). This program is now under going until 2001. This paper is to show the overall plan and the status of NUPEC-TH-P. (authors)

Acceptance Test Report for 241-SY Pump Cradle Hydraulic System

International Nuclear Information System (INIS)

Koons, B.M.

1995-01-01

The purpose of this ATP is to verify that hydraulic system/cylinder procured to replace the cable/winch system on the 101-SY Mitigation Pump cradle assembly fulfills its functional requirements for raising and lowering the cradle assembly between 70 and 90 degrees, both with and without pump. A system design review was performed on the 101-SY Cradle Hydraulic System by the vendor before shipping (See WHC-SD-WM-DRR-045, 241-SY-101 Cradle Hydraulic System Design Review). The scope of this plan focuses on verification of the systems ability to rotate the cradle assembly and any load through the required range of motion

Thermo-mechanical design of the SINGAP accelerator grids for ITER NB injectors

Energy Technology Data Exchange (ETDEWEB)

Agostinetti, P. [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, I35127 Padova (Italy)], E-mail: piero.agostinetti@igi.cnr.it; Dal Bello, S.; Dalla Palma, M.; Zaccaria, P. [Consorzio RFX, Euratom-ENEA Association, Corso Stati Uniti 4, I35127 Padova (Italy)

2007-10-15

The SINGle Aperture-SINgle GAP (SINGAP) accelerator for ITER neutral beam injector foresees four grids for the extraction and acceleration of negative ions, instead of the seven grids of the Multi-Aperture Multi-Grid (MAMuG) reference configuration. The grids have to fulfil specific requirements coming from ion extraction, beam optics and thermo-mechanical issues. This paper focuses on the thermo-hydraulic and thermo-mechanical design of the grids carried out by Consorzio RFX for the design of the first ITER NB injector and the ITER NB Test Facility. The cooling circuit design (position and shape of the channels) and the cooling parameters (water coolant temperatures, pressure and velocity) were optimized with sensitivity analyses in order to satisfy the grid functional requirements (temperatures, stresses, in plane and out of plane deformations). The design required a complete modelling of the grids and their support frames by means of 3D FE and CAD models.

A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

Energy Technology Data Exchange (ETDEWEB)

Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)

1997-12-31

Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.

A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

International Nuclear Information System (INIS)

Marguet, S.D.

1997-01-01

Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF's neutronic code COCCINELLE uses the Rowland's formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission's products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on 'low' configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.)

Study of thermal - hydraulic sensors signal fluctuations in PWR

International Nuclear Information System (INIS)

Hennion, F.

1987-10-01

This thesis deals with signal fluctuations of thermal-hydraulic sensors in the main coolant primary of a pressurized water reactor. The aim of this work is to give a first response about the potentiality of use of these noise signals for the functionning monitoring. Two aspects have been studied: - the modelisation of temperature fluctuations of core thermocouples, by a Monte-Carlo method, gives the main characteristics of these signals and their domain of application. - the determination of eigenfrequency in the primary by an acoustic representation could permit the monitoring of local and global thermo-hydraulic conditions [fr

Thermo-hydraulic Analysis of a Water-cooled Printed Circuit Heat Exchanger in a Small-scale Nitrogen Loop

International Nuclear Information System (INIS)

Kim, Chan Soo; Hong, Sung Deok; Kim, Min Hwan; Shim, Jaesool; Lee, Gyung Dong

2013-01-01

The development of high-temperature heat exchangers is very important because of its higher operation temperature and pressure than those of common light water reactors and industrial process plants. In particular, the intermediate heat exchanger is a key-challenged high temperature component in a Very High Temperature gas-cooled Reactor (VHTR). A printed circuit heat exchanger is one of the candidates for an intermediate heat exchanger in a VHTR. The printed circuit heat exchanger (PCHE) was developed and commercialized by HEATRIC. The compactness is better than any other heat exchanger types, because its core matrices are fabricated by diffusion bonding with photo-chemically etched micro-channels. Various tests and analysis have been performed to verify the performance of PCHE. The thermal stress analysis of the high temperature PCHE is necessary to endure the extremely operation condition of IHX. In this study, the thermo-hydraulic analysis for the laboratory-scale PCHE is performed to provide the input data for the boundary conditions of a structural analysis. The results from the first-principal calculation are compared with those from computational fluid dynamics code analysis. COMSOL 4.3a analysis is successfully performed at the uniform pressure drop condition in a set of flow channel stacks. The heat-exchanged region concentrated to the nitrogen inlet cause the uniform mass velocity distribution in the channels, therefore there is little difference between two analytical results

Parallelization of TWOPORFLOW, a Cartesian Grid based Two-phase Porous Media Code for Transient Thermo-hydraulic Simulations

Science.gov (United States)

Trost, Nico; Jiménez, Javier; Imke, Uwe; Sanchez, Victor

2014-06-01

TWOPORFLOW is a thermo-hydraulic code based on a porous media approach to simulate single- and two-phase flow including boiling. It is under development at the Institute for Neutron Physics and Reactor Technology (INR) at KIT. The code features a 3D transient solution of the mass, momentum and energy conservation equations for two inter-penetrating fluids with a semi-implicit continuous Eulerian type solver. The application domain of TWOPORFLOW includes the flow in standard porous media and in structured porous media such as micro-channels and cores of nuclear power plants. In the latter case, the fluid domain is coupled to a fuel rod model, describing the heat flow inside the solid structure. In this work, detailed profiling tools have been utilized to determine the optimization potential of TWOPORFLOW. As a result, bottle-necks were identified and reduced in the most feasible way, leading for instance to an optimization of the water-steam property computation. Furthermore, an OpenMP implementation addressing the routines in charge of inter-phase momentum-, energy- and mass-coupling delivered good performance together with a high scalability on shared memory architectures. In contrast to that, the approach for distributed memory systems was to solve sub-problems resulting by the decomposition of the initial Cartesian geometry. Thread communication for the sub-problem boundary updates was accomplished by the Message Passing Interface (MPI) standard.

Thermo-Mechanical Modeling of Laser-Mig Hybrid Welding (lmhw)

Science.gov (United States)

Kounde, Ludovic; Engel, Thierry; Bergheau, Jean-Michel; Boisselier, Didier

2011-01-01

Hybrid welding is a combination of two different technologies such as laser (Nd: YAG, CO2…) and electric arc welding (MIG, MAG / TIG …) developed to assemble thick metal sheets (over 3 mm) in order to reduce the required laser power. As a matter of fact, hybrid welding is a lso used in the welding of thin materials to benefit from process, deep penetration and gap limit. But the thermo-mechanical behaviour of thin parts assembled by LMHW technology for railway cars production is far from being controlled the modeling and simulation contribute to the assessment of the causes and effects of the thermo mechanical behaviour in the assembled parts. In order to reproduce the morphology of melted and heat-affected zones, two analytic functions were combined to model the heat source of LMHW. On one hand, we applied a so-called "diaboloïd" (DB) which is a modified hyperboloid, based on experimental parameters and the analysis of the macrographs of the welds. On the other hand, we used a so-called "double ellipsoïd" (DE) which takes the MIG only contribution including the bead into account. The comparison between experimental result and numerical result shows a good agreement.

Thermo-pneumatic canning; Le gainage thermopneumatique

Energy Technology Data Exchange (ETDEWEB)

Gauthron, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

In the thermo-pneumatic canning, the fuel is enclosed in its can with a clearance that must be reduced by external heated gas pressure. The principal applications are: a) binding magnesium cans on to uranium in fuel elements of reactors cooled by CO{sub 2} under pressure, b) application of a can to a hollow bar of uranium too thin to resist the pressure of cold hydraulic canning, c) application of an aluminium can to a bar, with an initial diametrical clearance between uranium and can too great to sustain cold hydraulic canning without buckling, d) detection of major leakage in the slugs. (author) [French] Ce procede consiste a appliquer une gaine sur une barre d'uranium par pression hydrostatique d'un gaz chaud. Les principales applications sont: a) le frettage des gaines de magnesium des elements combustibles des piles refroidies au CO{sub 2} sous pression, b) le gainage d'un barreau creux qui serait ecrase a froid, c) le gainage avec un jeu initial trop fort pour etre effectue a froid sans plisser, d) la detection des fuites de cartouches. (auteur)

Engine including hydraulically actuated valvetrain and method of valve overlap control

Science.gov (United States)

Cowgill, Joel [White Lake, MI

2012-05-08

An exhaust valve control method may include displacing an exhaust valve in communication with the combustion chamber of an engine to an open position using a hydraulic exhaust valve actuation system and returning the exhaust valve to a closed position using the hydraulic exhaust valve actuation assembly. During closing, the exhaust valve may be displaced for a first duration from the open position to an intermediate closing position at a first velocity by operating the hydraulic exhaust valve actuation assembly in a first mode. The exhaust valve may be displaced for a second duration greater than the first duration from the intermediate closing position to a fully closed position at a second velocity at least eighty percent less than the first velocity by operating the hydraulic exhaust valve actuation assembly in a second mode.

Thermal-hydraulic behaviour of the ITER TF system during a quench development

International Nuclear Information System (INIS)

Nicollet, S.; Lacroix, B.; Bessette, D.; Copetti, R.; Duchateau, J.L.; Coatanea-Gouachet, M.; Rodriguez-Mateos, F.

2011-01-01

In order to ensure the safety of the ITER TF magnets, a primary quench detection system has been foreseen, based on voltage detection. In addition, a secondary quench detection could rely on signals of thermo-hydraulic nature. As a matter of fact, the development of a quench in a CICC leads to significant variations of pressure and mass flow at the quenched pancake extremities. Analyses of the quench development have thus been performed using the coupled GANDALF and FLOWER codes. This tool allows to simulate the thermo-hydraulic behaviour of one CICC with a model of the external cryogenic circuit. The study has focused on the first seconds of the quench development, supposing that the quench has not been detected earlier by the primary detector. It is shown that signals regarding pressure, mass flow and temperature reach significant high values especially in the connecting feeder associated with the helium inlet. More detailed studies will be needed to select a secondary detector in this region.

IAEA coordinated research programme on heat transfer behavior and thermo-hydraulics code testing for super critical water cooled reactors

International Nuclear Information System (INIS)

Bilbao y Leon, Sama; Aksan, Nusret

2009-01-01

One of the key roles of the IAEA is to foster the collaboration among Member States on the development of advances in technology for advanced nuclear power plants. There is high international interest, both in developing and industrialized countries, in innovative supercritical water-cooled reactors (SCWRs), primarily because such concepts will achieve high thermal efficiencies (44-45%) and promise improved economic competitiveness utilizing and building upon the recent developments for highly efficient fossil power plants. The SCWR has been selected as one of the promising concepts for development by the Generation-IV International Forum. Following the advice of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and TWG-HWR), with the feedback from the Gen-IV SCWR Steering Committee, and in coordination with the OECD-NEA, IAEA has recently started a Coordinated Research Programme (CRP) in the areas of heat transfer behaviour and testing of thermo-hydraulic computer methods for Supercritical Water-Cooled Reactors. The first Research Coordination Meeting (RCM) of the CRP was held at the IAEA Headquarters, in Vienna, Austria in July 2008. This paper summarizes the current status of the CRP, including the Integrated Research Plan and the general schedule for the CRP. (author)

Hydraulic nuts (HydraNuts) for reactor vessel tensioning

International Nuclear Information System (INIS)

Greenwell, Steve

2008-01-01

The paper will present how the introduction of hydraulic nuts - HydraNuts, has reduced critical path times, dose exposure for workers and improved working safety conditions around the reactor vessel during tensioning or de-tensioning operations. It will focus upon detailing the advantages realized by utilities that have introduced the technology and providing examples of the improvements made to the process as well as discussing the engineering design change packages required to make the conversion to the new system. HydraNuts replace the traditional mechanical nut/stud tensioning equipment, combining the two functions into a single system, designed for easy installation and operation by one individual. The primary components of the HydraNut can be assembled without the need for external crane or hoist support and are designed so that each sub assembly can be fitted separately. Once all HydraNuts are fitted to the Rx vessel studs and are sitting on the main Rx vessel head flange, then a system of flexible hydraulic hoses is connected to them, forming a closed loop hydraulic harness, which will allow for simultaneous pressurization of all HydraNuts. Hydraulic pressure is obtained by the use of a hydraulic pumping unit and the resultant load generated in each HydraNut is transferred to the stud and main flange closure is obtained. While maintaining hydraulic pressure, a locking ring is rotated into place on the HydraNut assembly that will support the tensioned load mechanically when the hydraulic pressure is released from the hose harness assembly. The hose harness is removed and the HydraNut is now functioning as a mechanical nut retaining the tensioned load. The HydraNut system for Rx vessel applications was first introduced into a plant in the U.S. in October 2006 and based upon the benefits realized subsequent projects are under way within the Asian and U.S. operating fleet. (author)

Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

Energy Technology Data Exchange (ETDEWEB)

Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

2016-08-01

To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

Investigation research on the evaluation of a coupled thermo-hydro-mechanical-chemical phenomena. 2. Result report

International Nuclear Information System (INIS)

Ishihara, Yoshinao; Ito Takaya; Chijimatsu, Masakazu; Amemiya, Kiyoshi; Shiozaki, Isao; Neyama, Atsushi; Tanaka, Yumiko

2003-02-01

In order to realize a coupling analysis in the near field of the geological disposal system, the coupling analysis code on the thermo-hydro-mechanical-chemical phenomena by THAMES, Dtransu and phreeqe60, which are existing analysis code, is developed in this study. And we carried out the case analysis on the thermo-hydro-mechanical-chemical phenomena by this code. (1) We have developed coupling analysis system to manage coupling analysis and to control coupling process automatically for THAMES (thermo-hydro-mechanical analysis code), Dtransu (mass transport analysis code) and phreeqe60 (geochemical analysis code). (2) Some supporting module, which includes transfer of dissolution concentration and total concentration (dissolution + precipitation concentration), was prepared as a functional expansion. And in order to treat multi-chemical elements, we have codified mass transport analysis code. (3) We have prepared hydraulic conductivity module of buffer material depending on change of dry density due to chemical equilibrium (dissolution and precipitation of minerals), and change of concentration of NaCl solutions. After THAMES, Dtransu, phreeqe60 and hydraulic conductivity module were installed in the COUPLYS, sensitivity analysis was carried out to check basic operation. (4) In order to confirm the applicability of the developed THMC analysis code, we have carried out case analysis on 1-dimensional and 3-dimensional model which including vitrified waste, over-pack, buffer material and rock in the HLW near-field. (author)

Thermo-mechanical design of the extraction grids for RF negative ion source at HUST

Energy Technology Data Exchange (ETDEWEB)

Zuo, Chen; Liu, Kaifeng, E-mail: kfliuhust@hust.edu.cn; Li, Dong; Mei, Zhiyuan; Zhang, Zhe; Chen, Dezhi

2017-01-15

Highlights: • An extraction system with cooling channels has been designed for HUST negative ion source. • Corresponding heat loads onto three grids has been used in thermo-mechanical analysis. • The analysis results could be very useful for driving the engineering design. - Abstract: Huazhong University of Science and Technology (HUST) is developing a small radio frequency negative ion source experimental setup to promote research on neutral beam injection ion sources. The extraction system for the negative ion source is the key component to obtain the negative ions. The extraction system is composed of three grids: the plasma grid, the extraction grid and the grounded grid. Each grid is impacted by different heat loads. As the grids have to fulfil specific requirements regarding ion extraction, beam optics, and thermo-mechanical issues, grid cooling systems have been included for ensuring reliable operation. This paper focuses on the thermo-hydraulic and thermo-mechanical design of the grids. Finite element calculations have been carried out to analyse the temperature and deformation of the grids under heat loads using the fluid dynamics code CFX. Based on these results, the cooling circuit design and cooling parameters are optimised to satisfy the grid requirements.

Thermo-osmosis in Membrane Systems: A Review

Science.gov (United States)

Barragán, V. María; Kjelstrup, Signe

2017-06-01

We give a first review of experimental results for a phenomenon little explored in the literature, namely thermal osmosis or thermo-osmosis. Such systems are now getting increased attention because of their ability to use waste heat for separation purposes. We show that this volume transport of a solution or a pure liquid caused by a temperature difference across a membrane can be understood as a property of the membrane system, i. e. the membrane with its adjacent solutions. We present experimental values found in the literature of thermo-osmotic coefficients of neutral and hydrophobic as well as charged and hydrophilic membranes, with water and other permeant fluids as well as electrolyte solutions. We propose that the coefficient can be qualitatively explained by a formula that contains the entropy of adsorption of permeant into the membrane, the hydraulic permeability, and a factor that depends on the interface resistance to heat transfer. A variation in the entropy of adsorption with hydrophobic/hydrophilic membranes and structure breaking/structure making cations could then explain the sign of the permeant flux. Systematic experiments in the field are lacking and we propose an experimental program to mend this situation.

Thermo Wigner operator in thermo field dynamics: its introduction and application

International Nuclear Information System (INIS)

Fan Hongyi; Jiang Nianquan

2008-01-01

Because in thermo-field dynamics (TFD) the thermo-operator has a neat expression in the thermo-entangled state representation, we need to introduce the thermo-Wigner operator (THWO) in the same representation. We derive the THWO in a direct way, which brings much conveniece to calculating the Wigner functions of thermo states in TFD. We also discuss the condition for existence of a wavefunction corresponding to a given Wigner function in the context of TFD by using the explicit form of the THWO.

Study of thermal hydraulic behavior of supercritical water flowing through fuel rod bundles

International Nuclear Information System (INIS)

Thakre, Sachin; Lakshmanan, S.P.; Kulkarni, Vinayak; Pandey, Manmohan

2009-01-01

Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases. (author)

Thermo-hydro-mechanical behavior of fractured rock mass

International Nuclear Information System (INIS)

Coste, F.

1997-12-01

The purpose of this research is to model Thermo-Hydro-Mechanical behavior of fractured rock mass regarding a nuclear waste re-depository. For this, a methodology of modeling was proposed and was applied to a real underground site (EDF site at Nouvelle Romanche). This methodology consists, in a first step, to determine hydraulic and mechanical REV. Beyond the greatest of these REV, development of a finite element code allows to model all the fractures in an explicit manner. The homogenized mechanical properties are determined in drained and undrained boundary conditions by simulating triaxial tests that represent rock mass subject to loading. These simulations allow to study the evolution of hydraulic and mechanical properties as a function of stress state. Drained and undrained boundary conditions enable to discuss the validity of assimilation of a fractured rock mass to a porous medium. The simulations lead to a better understanding of the behavior of the fractured rock masses and allow to show the dominant role of the shear behavior of the fractures on the hydraulic and mechanical homogenized properties. From a thermal point of view, as long as conduction is dominant, thermal properties of the rock mass are almost the same as those the intact rock. (author)

Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

Energy Technology Data Exchange (ETDEWEB)

Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

2012-02-01

The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

Whole Core Thermal-Hydraulic Design of a Sodium Cooled Fast Reactor Considering the Gamma Energy Transport

International Nuclear Information System (INIS)

Choi, Sun Rock; Back, Min Ho; Park, Won Seok; Kim, Sang Ji

2012-01-01

Since a fuel cladding failure is the most important parameter in a core thermal-hydraulic design, the conceptual design stage only involves fuel assemblies. However, although non-fuel assemblies such as control rod, reflector, and B4C generate a relatively smaller thermal power compared to fuel assemblies, they also require independent flow allocation to properly cool down each assembly. The thermal power in non-fuel assemblies is produced from both neutron and gamma energy, and thus the core thermal-hydraulic design including non-fuel assemblies should consider an energy redistribution by the gamma energy transport. To design non-fuel assemblies, the design-limiting parameters should be determined considering the thermal failure modes. While fuel assemblies set a limiting factor with cladding creep temperature to prevent a fission product ejection from the fuel rods, non-fuel assemblies restrict their outlet temperature to minimize thermally induced stress on the upper internal structure (UIS). This work employs a heat generation distribution reflecting both neutron and gamma transport. The whole core thermal-hydraulic design including fuel and non-fuel assemblies is then conducted using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code. The other procedures follow from the previous conceptual design

Integration of Flex Nozzle System and Electro Hydraulic Actuators to Solid Rocket Motors

Science.gov (United States)

Nayani, Kishore Nath; Bajaj, Dinesh Kumar

2017-10-01

A rocket motor assembly comprised of solid rocket motor and flex nozzle system. Integration of flex nozzle system and hydraulic actuators to the solid rocket motors are done after transportation to the required place where integration occurred. The flex nozzle system is integrated to the rocket motor in horizontal condition and the electro hydraulic actuators are assembled to the flex nozzle systems. The electro hydraulic actuators are connected to the hydraulic power pack to operate the actuators. The nozzle-motor critical interface are insulation diametrical compression, inhibition resin-28, insulation facial compression, shaft seal `O' ring compression and face seal `O' ring compression.

Feasibility study of application of ductless fuel assembly to FBR

International Nuclear Information System (INIS)

Itoh, K.; Shibahara, I.

1996-01-01

Feasibility studies on an application of the ductless fuel concept to an FBR core have been carried out in order to evaluate the basic features of the ductless core, especially in the fields of the thermal-hydraulic aspects and the mechanical behaviors. Regarding thermal-hydraulic aspects, analyses were performed by using a whole core thermal-hydraulic analysis code by making some modification for this study on the PWR code THINC. A small scaled ductless core model was prepared and a hydraulic experiment was carried out to study basic hydraulic characteristics of a ductless core. Core mechanical behaviors were analyzed focusing on the core irradiation bowing aspects and the seismic behaviors. Following features are revealed on the core structural behaviors: (1) the bowing stiffness of the ductless assembly is around 1/5 to 1/10 of that of the duct type assembly; (2) the contact loads between assemblies by the bowing effects are small through core cycles; (3) the damping of the ductless assemblies are so large that the seismic responses are small and the loads between assemblies are small due to occurring many contact points. Through this study it is expected that the concept of the ductless fuel can be applicable to FBR cores from the design view points of thermal-hydraulic and core mechanical behaviors

Analysis of unsaturated clayey materials hydration incorporating the effect of thermo-osmotic flow

International Nuclear Information System (INIS)

Sanchez, M.; Arson, C.

2012-01-01

Document available in extended abstract form only. The hydraulic gradient is the main physical phenomenon influencing the movement of water in permeable porous media. It is, however, not the only one. Figure 1 presents the main kinds of flow that can occur in a porous media alongside with the corresponding gradient responsible for the movements. The word 'law' is generally used for the diagonal terms associated with the direct flow phenomena, and the name 'effect' is reserved to the non-diagonal ones, called also 'coupled processes'. Lippmann (1907) discovered and named the phenomenon of thermo-osmosis. He discovered it experimentally by separating a volume of water into two parts by means of a membrane. Different temperatures were held in the two regions of the system. The thermal gradient caused a flow of water through the membrane from the cold to the hot side. In permeable reservoirs, the non-diagonal coefficients are relatively small and negligible compared to the diagonal terms. That is the reason why the coupled processes are generally ignored when analyzing problems in aquifers. However, in non-isothermal problems involving low permeability media and/or low hydraulic gradients thermo-osmosis may play a more influential role. Srivastava and Avasthi (1975) and Horseman and McEwen (1996) showed that water flux due to thermo-osmosis can easily exceed Darcy flux in low permeability clays. The 'phenomenological coefficient' that links each flow with the corresponding driving gradient must be measured experimentally. Accounting for thermo-osmosis is assuming that the transport of heat may modify the transport of fluids. The counterpart phenomenon of thermo-osmosis is thermo-filtration, which reflects the influence of a pressure gradient on heat flow. Thermo-osmosis and thermo-filtration are generally formulated as reciprocal relations, so that the coupled conductivity terms related to each phenomenon are set equal. Thermo-osmotic effects have been studied in the

Core design options for high conversion BWRs operating in Th–233U fuel cycle

International Nuclear Information System (INIS)

Shaposhnik, Y.; Shwageraus, E.; Elias, E.

2013-01-01

Highlights: • BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. • Seed blanket optimization that includes assembly size array and axial dimensions. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal-hydraulic analysis includes MCPR observation. -- Abstract: Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone “sandwiched” between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233 U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating. Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design

Thermo hydraulics of a steam boiler forced circulation

International Nuclear Information System (INIS)

Tucakovic, Dragan; Zivanovic, Titoslav; Stevanovic, Vladimir

2006-01-01

In order to minimize the dryout at the steam boiler furnace in the Thermal Power Plant Kolubara B, designed are inner rifled wall tubes. This type of tubes, with many spiral grooves cut into the bore, prevents film boiling and enables the nucleate boiling be still maintained under the condition of vapour quality being app. 1. To verify the choice of the rifled tubes instead of the cheaper, smooth tubes type being justified, analyzed is the change of the actual and critical vapour quality with the furnace height, under uniform and non-uniform heat flu through evaporator walls. Furthermore, made are hydraulic calculations for various steam boiler loads, in case of both rifled and smooth tubes types, with the purpose to check the rifles influence to pressure drop increase in comparison with the smooth tubes. Also, checked is the selection of the circulation pump. Key words: evaporator, forced circulation, rifled tubes, critical vapour quality, pressure drop

3. Workshop for IAEA ICSP on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents. Presentations

International Nuclear Information System (INIS)

2012-04-01

Most advanced nuclear power plant designs adopted several kinds of passive systems. Natural circulation is used as a key driving force for many passive systems and even for core heat removal during normal operation such as NuScale, CAREM, ESBWR and Indian AHWR designs. Simulation of natural circulation phenomena is very challenging since the driving force of it is weak compared to forced circulation and involves a coupling between primary system and containment for integral type reactor. The IAEA ICSP (International Collaborative Standard Problem) on 'Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents' was proposed within the CRP on 'Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems that utilize Natural Circulation'. Oregon State University (OSU) of USA offered to host this ICSP. This ICSP plans to conduct the following experiments and blind/open simulations with system codes: 1. Quasi-steady state operation with different core power levels: Conduct quasi-steady state operation with step-wise increase of core power level in order to observe single phase natural circulation flow according to power level. The experimental facility and operating conditions for an integral PWR will be used. 2. Thermo-hydraulic Coupling between Primary system and Containment: Conduct a loss of feedwater transient with subsequent ADS blowdown and long term cooling to determine the progression of a loss of feedwater transient by natural circulation through primary and containment systems. These tests would examine the blowdown phase as well as the long term cooling using sump natural circulation by coupling the primary to containment systems. This data could be used for the evaluation of system codes to determine if they model specific phenomena in an accurate manner. OSU completed planned two ICSP tests in July 2011 and real initial and boundary conditions measured from the

Thermo-Hydraulic behaviour of dual-channel superconducting Cable-In-Conduit Conductors for ITER

International Nuclear Information System (INIS)

Renard, B.

2006-09-01

In an effort to optimise the cryogenics of large superconducting coils for fusion applications (ITER), dual channel Cable-In-Conduit Conductors (CICC) are designed with a central channel spiral to provide low hydraulic resistance and faster helium circulation. The qualitative and economic rationale of the conductor central channel is here justified to limit the superconductor temperature increase, but brings more complexity to the conductor cooling characteristics. The pressure drop of spirals is experimentally evaluated in nitrogen and water and an explicit hydraulic friction model is proposed. Temperatures in the cable must be quantified to guarantee superconductor margin during coil operation under heat disturbance and set adequate inlet temperature. Analytical one-dimensional thermal models, in steady state and in transient, allow to better understand the thermal coupling of CICC central and annular channels. The measurement of a heat transfer characteristic space and time constants provides cross-checking experimental estimations of the internal thermal homogenization. A simple explicit model of global inter-channel heat exchange coefficient is proposed. The risk of thermosyphon between the two channels is considered since vertical portions of fusion coils are subject to gravity. The new hydraulic model, heat exchange model and gravitational risk ratio allow the thermohydraulic improvement of CICC central spirals. (author)

PWR fuel thermomechanics

International Nuclear Information System (INIS)

Traccucci, R.; Leclercq, J.

1986-01-01

Fuel thermo-mechanics means the studies of mechanical and thermal effects, and more generally, the studies of the behavior of the fuel assembly under stresses including thermal and mechanical loads, hydraulic effects and phenomena induced by materials irradiation. This paper describes the studies dealing with the fuel assembly behavior, first in normal operating conditions, and then in accidental conditions. 43 refs [fr

ASSEMBLY DESIGN OPTIMIZATION FOR GEAR PUMP HYDRAULIC UNITS

Directory of Open Access Journals (Sweden)

ŞCHEAUA Fanel

2012-09-01

Full Text Available This paper presents a model for gear pump assembly design realized in Solid Edge V20. The aim is to highlight modelling aspects for solid part components and how to achieve an assembly from several component parts. Can be noted that computer aided design (CAD software can provide multiple options of representing various designed components, assemblies containing up to hundreds of items and part component motion simulation.

HCPB TBM thermo mechanical design: Assessment with respect codes and standards and DEMO relevancy

International Nuclear Information System (INIS)

Cismondi, F.; Kecskes, S.; Aiello, G.

2011-01-01

In the frame of the activities of the European TBM Consortium of Associates the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) is developed in Karlsruhe Institute of Technology (KIT). After performing detailed thermal and fluid dynamic analyses of the preliminary HCPB TBM design, the thermo mechanical behaviour of the TBM under typical ITER loads has to be assessed. A synthesis of the different design options proposed has been realized building two different assemblies of the HCPB-TBM: these two assemblies and the analyses performed on them are presented in this paper. Finite Element thermo-mechanical analyses of two detailed 1/4 scaled models of the HCPB-TBM assemblies proposed have been performed, with the aim of verifying the accordance of the mechanical behaviour with the criteria of the design codes and standards. The structural design limits specified in the codes and standard are discussed in relation with the EUROFER available data and possible damage modes. Solutions to improve the weak structural points of the present design are identified and the DEMO relevancy of the present thermal and structural design parameters is discussed.

Thermo-mechanical design of the Plasma Driver Plate for the MITICA ion source

Energy Technology Data Exchange (ETDEWEB)

Pavei, Mauro, E-mail: mauro.pavei@igi.cnr.it [Consorzio RFX, EURATOM-ENEA Association, Corso Stati Uniti 4, I-35127 Padova (Italy); Palma, Mauro Dalla; Marcuzzi, Diego [Consorzio RFX, EURATOM-ENEA Association, Corso Stati Uniti 4, I-35127 Padova (Italy)

2010-12-15

In the framework of the activities for the development of the Neutral Beam Injector (NBI) for ITER, the detailed design of the Radio-Frequency (RF) negative ion source has been carried out. One of the most heated components of the RF source is the rear vertical plate, named Plasma Driver Plate (PDP), where the Back-Streaming positive Ions (BSI+) generated from stripping losses in the accelerator and back scattered on the plasma source impinge on. The heat loads that result are huge and concentrated, with first estimate of the power densities up to 60 MW/m{sup 2}. The breakdowns that occur into the accelerator cause such heat loads to act cyclically, so that the PDP is thermo-mechanically fatigue loaded. Moreover, the surface of the PDP facing the plasma is functionally required to be temperature controlled and to be molybdenum or tungsten coated. The thermo-hydraulic design of the plate has been carried out considering active cooling with ultra-pure water. Different heat sink materials, hydraulic circuit layout and manufacturing processes have been considered. The heat exhaust has been optimized by changing the channels geometry, the path of the heat flux in the heat sink, the thickness of the plate and maximizing the Heat Transfer Coefficient. Such optimization has been carried out by utilizing 3D Finite Element (FE) models. Afterwards all the suitable mechanical (aging, structural monotonic and cyclic) verifications have been carried out post-processing the results of the thermo-mechanical 3D FE analyses in accordance to specific procedures for nuclear components exposed to high temperature. The effect of sputtering phenomenon due to the high energy BSI+ impinging on the plate has been considered and combined with fatigue damage for the mechanical verification of the PDP. Alternative solutions having molybdenum (or tungsten coatings) facing the plasma, aiming to reduce the sputtering rate and the consequent plasma pollution, have been evaluated and related 3D FE

Hydromechanical transmission with three simple planetary assemblies, one sun gear being mounted on the output shaft and the other two on a common shaft connected to an input-driven hydraulic module

Science.gov (United States)

Orshansky, Jr., deceased, Elias; Weseloh, William E.

1978-01-01

A power transmission having three simple planetary assemblies, each having its own carrier and its own planet, sun, and ring gears. A speed-varying module is connected in driving relation to the input shaft and in driving relationship to the sun gears of the first two planetary assemblies, these two sun gears being connected together on a common shaft. The speed-varying means may comprise a pair of hydraulic units hydraulically interconnected so that one serves as a pump while the other serves as a motor and vice versa, one of the units having a variable stroke and being connected in driving relation to the input shaft, the other unit, which may have a fixed stroke, being connected in driving relation to the sun gears. The input shaft is also connected to drive the second ring gear and, furthermore is clutchable to the carrier of the third planetary assembly. A brake grounds the first carrier in the first range and in reverse and causes drive to be delivered to the output through the first ring gear in a hydrostatic mode. The carrier of the second planetary assembly drives the ring gear of the third planetary assembly, which is clutchable to the output shaft, and the sun gear of the third planetary assembly is mounted rigidly to the output shaft.

Sensitiveness Analysis of Neutronic Parameters Due to Uncertainty in Thermo-hydraulic parameters on CAREM-25 Reactor; Analisis de Sensibilidad de los Parametros Neutronicos ante Incertezas en los Parametros Termohidraulicos en el Reactor CAREM-25

Energy Technology Data Exchange (ETDEWEB)

Serra, Oscar [Comision Nacional de Energia Atomica, Centro Atomico Bariloche (Argentina)

2000-07-01

Some studies were done about the effect of the uncertainty in the values of several thermo-hydraulic parameters on the core behaviour of the CAREM-25 reactor.By using the chain codes CITVAP-THERMIT and the perturbation the reference states, it was found that concerning to the total power, the effects were not very important, but were much bigger for the pressure.Furthermore were hardly significant in the presence of any perturbation on the void fraction calculation and the fuel temperature.The reactivity and the power peaking factor had highly important changes in the case of the coolant flow.We conclude that the use of this procedure is adequate and useful to our purpose.

Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

International Nuclear Information System (INIS)

Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

2017-01-01

A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

Energy Technology Data Exchange (ETDEWEB)

Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

Fluid flow test for KMRR fuel assemblies

International Nuclear Information System (INIS)

Chung, Moon Ki; Yang, Sun Kyu; Chung, Chang Hwan; Chun, See Young; Song, Chul Hha; Jun, Hyung Gil; Chung, Heung Joon; Won, Soon Yeun; Cho, Young Rho; Kim, Bok Deuk

1991-01-01

Hydraulic and velocity measurment tests were carried out for the KMRR fuel assembly. Two types of the KMRR fuel assembly are consist of longitudinally finned rods. Experimental data of the pressure drops and friction factors for the KMRR fuel assemlby were produced. The measurement technique for the turbulent flow structure in subchannels using the LDV was obtained. The measurement of the experimental constant of the thermal hydraulic analysis code was investigated. The results in this study are used as the basic data for the development of an analysis code. The measurement technique acquired in this study can be applied to the KMRR thermal hydraulic commissioning test and development of the domestic KMRR fuel fabrication. (Author)

Shock absorbing structure for nuclear fuel assemblies

International Nuclear Information System (INIS)

Jabsen, F.S.

1981-01-01

A hydraulic apparatus is described that absorbs shocks that may be applied to fuel assemblies. Spring pads mounted on the upper end fittings of the fuel assemblies have plungers that move within hollow guide posts attached to the upper grids of the fuel assemblies. (L.L.)

ERP-IV-A program for transient thermal-hydraulic analysis of PWR plant

International Nuclear Information System (INIS)

Dai Anguo; Tang Jiahuan; Qian Huifu; Gao Zhikang

1987-12-01

The author deal with the descriptions of physical model of transient process in PWR plant and the function of ERP-IV (ERR-IV Transient Thermo-Hydraulic Analysis Code). The code has been developed for safety analysis and design transient. The code is characterized by the multi-loop long-term, short term, wide-range plant simulation with the capability to analyze natural circulation condition. The description of ERP-IV includes following parts: reactor, primary coolant loops, pressurizer, steam generators, main steam system, turbine, feedwater system, steam dump, relive valves, and safety valves in secondary side, etc.. The code can use for accident analysis, such as loss of all A.C. power to power plant auxiliaries (a station blackout), loss of normal feedwater, loss of load, loss of condenser vacuum and other events causing a turbine trip, complete loss of forced reactor coolant flow, uncontrolled rod cluster control assembly bank withdrawal. It can also be used for accident analysis of the emergency and limiting conditions, such as feedwater line break and main steam line rupture. It can also be utilized as a tool for system design studies, component design, setpoint studies and design transition studies, etc

Hydraulic performance of Compacted Clay Liners (CCLs) under combined temperature and leachate exposures.

Science.gov (United States)

Aldaeef, A A; Rayhani, M T

2014-12-01

Experimental investigations were carried out to investigate the effect of thermo-chemical exposures on the hydraulic performance of Compacted Clay Liners (CCLs) in landfills. Hydraulic conductivity of most CCL specimens was increased by two to three times their initial values when exposed to 55 °C for 75 days. CCL specimens also experienced increases in their hydraulic conductivities when exposed to leachate at room temperature. This behaviour could be due to the decrease in viscosity when the permeant was changed from tap water to leachate. However, as the leachate exposure time exceeded the first 15 days, hydraulic conductivity readings decreased to as much as one order of magnitude after 75 days of leachate permeation at room temperature. The gradual decrease in the CCLs hydraulic conductivities was most likely due to chemical precipitation and clogging of pore voids within the soils which seemed to reduce the effective pore volume. The rate of hydraulic conductivity reduction due to leachate permeation was slower at higher temperatures, which was attributed to the lower permeant viscosity and lower clogging occurrence. The observed hydraulic behaviours were correlated to the physical, mineral, and chemical properties of the CCLs and described below. Copyright © 2014 Elsevier Ltd. All rights reserved.

Fracture mechanics in new designed power module under thermo-mechanical loads

Directory of Open Access Journals (Sweden)

Durand Camille

2014-06-01

Full Text Available Thermo-mechanically induced failure is a major reliability issue in the microelectronic industry. On this account, a new type of Assembly Interconnected Technology used to connect MOSFETs in power modules has been developed. The reliability is increased by using a copper clip soldered on the top side of the chip, avoiding the use of aluminium wire bonds, often responsible for the failure of the device. Thus the new designed MOSFET package does not follow the same failure mechanisms as standard modules. Thermal and power cycling tests were performed on these new packages and resulting failures were analyzed. Thermo-mechanical simulations including cracks in the aluminium metallization and intermetallics (IMC were performed using Finite Element Analysis in order to better understand crack propagation and module behaviour.

Automatically closing swing gate closure assembly

Science.gov (United States)

Chang, Shih-Chih; Schuck, William J.; Gilmore, Richard F.

1988-01-01

A swing gate closure assembly for nuclear reactor tipoff assembly wherein the swing gate is cammed open by a fuel element or spacer but is reliably closed at a desired closing rate primarily by hydraulic forces in the absence of a fuel charge.

Implementation of the optimization for the methodology of the neutronic calculation and thermo-hydraulic in IEA-R1 reactor

International Nuclear Information System (INIS)

Stefani, Giovanni Laranjo de; Conti, Thadeu das Neves; Fedorenko, Giuliana G.; Castro, Vinicius A.; Maio, Mireia F.; Santos, Thiago Augusto dos

2011-01-01

This work objective was to create a manager program that would automate the programs and computer codes in use for neutronic calculation and thermo-hydraulic in IEA-R1 reactor thus making the process for calculation of safety parameters and for configuration change up to 98% faster than that used in the reactor today. This process was tested in combination with the reactor operators and is being implemented by the quality department. The main codes and programs involved in the calculations of configuration change are Leopard, Hammier-Technion, Twodb, Citation and Cobra. Calculations of delayed neutron and criticality coefficients given in the process of safety parameters calculation are given by the Hammer-Technion and Citation in a process that involves about eleven repetitions so that it meets all the necessary conditions (such different temperatures of the moderator and fuel). The results are entirely consistent with the expected and absolutely the same as those given by manual process. Thus the work shows its reliability as well the advantage of saving time, once a process that could take up to four hours was turned in one that takes around five minutes when done in a home computer. Much of this advantage is due to the fact that were created subprograms to treat the output of each program used and transform them into the input of the other programs, removing from it the intermediate essential data for this to occur, thus avoiding also a possible human error by handling the various data supplied. (author)

An immersed body method for coupled neutron transport and thermal hydraulic simulations of PWR assemblies

International Nuclear Information System (INIS)

Jewer, S.; Buchan, A.G.; Pain, C.C.; Cacuci, D.G.

2014-01-01

Highlights: • A new method of coupled radiation transport, heat and momentum exchanges on fluids, and heat transfer simulations. • Simulation of the thermal hydraulics and radiative properties within whole PWR assemblies. • An immersed body method for modelling complex solid domains on practical computational meshes. - Abstract: A recently developed immersed body method is adapted and used to model a typical pressurised water reactor (PWR) fuel assembly. The approach is implemented with the numerical framework of the finite element, transient criticality code, FETCH which is composed of the neutron transport code, EVENT, and the CFD code, FLUIDITY. Within this framework the neutron transport equation, Navier–Stokes equations and a fluid energy conservation equation are solved in a coupled manner on a coincident structured or unstructured mesh. The immersed body method has been used to model the solid fuel pins. The key feature of this method is that the fluid/neutronic domain and the solid domain are represented by overlapping and non-conforming meshes. The main difficulty of this approach, for which a solution is proposed in this work, is the conservative mapping of the energy and momentum exchange between the fluid/neutronic mesh and the solid fuel pin mesh. Three numerical examples are presented which include a validation of the fuel pin submodel against an analytical solution; an uncoupled (no neutron transport solution) PWR fuel assembly model with a specified power distribution which was validated against the COBRA-EN subchannel analysis code; and finally a coupled model of a PWR fuel assembly with reflective neutron boundary conditions. Coupling between the fluid and neutron transport solutions is through the nuclear cross sections dependence on Doppler fuel temperature, coolant density and temperature, which was taken into account by using pre-calculated cross-section lookup tables generated using WIMS9a. The method was found to show good agreement

Influence of osmotic processes on the excess-hydraulic head measured in the Toarcian/Domerian argillaceous formation of Tournemire

International Nuclear Information System (INIS)

Tremosa, J.

2010-01-01

In the framework of the studies dealing on ability to store radioactive wastes in argillaceous formations, signification of interstitial pressures is an important point to understand water and solutes transport. In very low permeability argillaceous formations, like those studied in the Callovo-Oxfordian of the Paris basin by ANDRA, pore pressure is frequently higher than the theoretical hydrostatic pressure or than the pressure in the surrounding aquifers. Such an overpressure is also measured in the Toarcian/Domerian argillaceous formation (k = 10 -21 m 2 ), studied by the IRSN in the underground research laboratory of Tournemire (Aveyron, France). The hydraulic head profile has been specified in this manuscript and found to present a 30 ±10 m excess head. This excess-head can be due to compaction disequilibrium of the argillaceous formation, diagenetic evolution of the rock, tectonic compression, changes in hydrodynamic boundary conditions or osmotic processes. Amongst these potential causes, chemical osmosis and thermo-osmosis, a fluid flow under a chemical concentration and a temperature gradient, respectively, are expected to develop owing to the small pore size and the electrostatic interactions related to the charged surface of clay minerals. The goal of the work presented here was to study and quantify the contribution of each cause to the measured excess-head. Chemo-osmotic and thermo-osmotic permeabilities were obtained by experiments and using theoretical models. Theoretical models are based on the reproduction of the interactions occurring between the charged surface of clay minerals and pore solution and their up-scaling at the representative elementary volume macroscopic scale. Chemical osmosis phenomenon is related to anionic exclusion and the determination of the chemo-osmotic efficiency requires the resolution of an electrical interactions model. A triple-layer-model which considers diffuse layers overlapping was improved during this thesis to be

Integral nuclear fuel element assembly

International Nuclear Information System (INIS)

Schluderberg, D. C.

1985-01-01

An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

Effects of temperature and thermally-induced microstructure change on hydraulic conductivity of Boom Clay

Directory of Open Access Journals (Sweden)

W.Z. Chen

2017-06-01

Full Text Available Boom Clay is one of the potential host rocks for deep geological disposal of high-level radioactive nuclear waste in Belgium. In order to investigate the mechanism of hydraulic conductivity variation under complex thermo-mechanical coupling conditions and to better understand the thermo-hydro-mechanical (THM coupling behaviour of Boom Clay, a series of permeability tests using temperature-controlled triaxial cell has been carried out on the Boom Clay samples taken from Belgian underground research laboratory (URL HADES. Due to its sedimentary nature, Boom Clay presents across-anisotropy with respect to its sub-horizontal bedding plane. Direct measurements of the vertical (Kv and horizontal (Kh hydraulic conductivities show that the hydraulic conductivity at 80 °C is about 2.4 times larger than that at room temperature (23 °C, and the hydraulic conductivity variation with temperature is basically reversible during heating–cooling cycle. The anisotropic property of Boom Clay is studied by scanning electron microscope (SEM tests, which highlight the transversely isotropic characteristics of intact Boom Clay. It is shown that the sub-horizontal bedding feature accounts for the horizontal permeability higher than the vertical one. The measured increment in hydraulic conductivity with temperature is lower than the calculated one when merely considering the changes in water kinematic viscosity and density with temperature. The nuclear magnetic resonance (NMR tests have also been carried out to investigate the impact of microstructure variation on the THM properties of clay. The results show that heating under unconstrained boundary condition will produce larger size of pores and weaken the microstructure. The discrepancy between the hydraulic conductivity experimentally measured and predicted (considering water viscosity and density changes with temperature can be attributed to the microstructural weakening effect on the thermal volume change

THERMIT, 3-D Thermo-Hydraulics of BWR and PWR

International Nuclear Information System (INIS)

Kazimi, M.S.; Kao, S.P.; Kelly, J.E.

1984-01-01

1 - Description of program or function: THERMIT2, the most recent release of THERMIT, is intended for thermal-hydraulic analysis of both boiling and pressurized water reactor cores. It solves the three-dimensional, two-fluid equations describing the two-phase flow and heat transfer dynamics in rectangular coordinates. The two-fluid model uses separate partial differential equations expressing conservation of mass, momentum, and energy for each fluid. THERMIT2 offers the choice of either pressure or velocity boundary conditions at the top and bottom of the core. THERMIT2 includes a two-phase turbulent mixing model which provides subchannel analysis capability. THERMIT2 also solves the radial heat conduction equations for fuel pin temperatures, and calculates the heat flux from fuel pin to coolant with appropriate heat transfer models described by a boiling curve. 2 - Method of solution: By expressing the exchange of mass, momentum, and energy between the fluids with physically-based mathematical models, the relative motion and thermal non-equilibrium between the fluids can exist

In-Pile Section(IPS) Inner Assembly Manufacturing Report

International Nuclear Information System (INIS)

Lee, Jong Min; Shim, Bong Sik; Lee, Chung Yong

2009-12-01

The objective of this report is to present the manufacturing, assembling and testing process of IPS Inner Assembly used in Fuel Test Loop(FTL) pre-operation test. The majority of the manufactured components are test fuels, inner assembly structures and subsidiary tools that is needed during the assembly process. In addition, Mock-up test for the welding and brazing is included at this stage. Lower structure, such as test fuels, fuel carrier legs are assembled and following structures, such as fuel carrier stem in the middle structure, top flange in the top structure are assembled together each other. To Verify the Reactor Coolant Pressure Boundary(RCPB) function in IPS Inner Assembly helium leak test and hydraulic test is performed with its acceptance criteria. According to the ASME III code Authorized Nuclear Inspector(ANI) is required during the hydraulic test. As-built measurement and insulation resistance test are performed to the structures and instrumentations after the test process. All requirements are satisfied and the IPS Inner Assembly was loaded in HANARO IR-1 hole in September 25, 2009

Thermo-hydraulic characteristics of serpentine tubing in the boilers of gas cooled reactors under condition of rapid and slow depressurization

International Nuclear Information System (INIS)

Abouhadra, D.S.; Byrne, J.E.

2003-01-01

In nuclear reactors of the magnox or advanced gas cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accidents using two phase flow codes requires knowledge of the heat transfer behaviour of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear electric . The tests were carried out on the thermal hydraulics experimental research assembly (THERA) loop at manchester university. Depressurization from an initial pressure of 60 bar, with fluid subcooling of 5 k, 50 k, and 100 k was controlled by discharging the test section contents through suitably chosen orifices to produce blowdown to 10% of the initial pressure over a time scale of 30 s to 3600 s. pressures and temperatures in the serpentine were measured at average time intervals of approximately 1 s

Hydraulic Profiling of a Parallel Channel Type Reactor Core

International Nuclear Information System (INIS)

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

2006-01-01

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

GCFR thermal-hydraulic experiments

International Nuclear Information System (INIS)

Schlueter, G.; Baxi, C.B.; Dalle Donne, M.; Gat, U.; Fenech, H.; Hanson, D.; Hudina, M.

1980-01-01

The thermal-hydraulic experimental studies performed and planned for the Gas-Cooled Fast Reactor (GCFR) core assemblies are described. The experiments consist of basic studies performed to obtain correlations, and bundle experiments which provide input for code validation and design verification. These studies have been performed and are planned at European laboratories, US national laboratories, Universities in the US, and at General Atomic Company

Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.

1981-01-01

The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru

A Thermoelastic Hydraulic Fracture Design Tool for Geothermal Reservoir Development

Energy Technology Data Exchange (ETDEWEB)

Ahmad Ghassemi

2003-06-30

Geothermal energy is recovered by circulating water through heat exchange areas within a hot rock mass. Geothermal reservoir rock masses generally consist of igneous and metamorphic rocks that have low matrix permeability. Therefore, cracks and fractures play a significant role in extraction of geothermal energy by providing the major pathways for fluid flow and heat exchange. Thus, knowledge of conditions leading to formation of fractures and fracture networks is of paramount importance. Furthermore, in the absence of natural fractures or adequate connectivity, artificial fracture are created in the reservoir using hydraulic fracturing. At times, the practice aims to create a number of parallel fractures connecting a pair of wells. Multiple fractures are preferred because of the large size necessary when using only a single fracture. Although the basic idea is rather simple, hydraulic fracturing is a complex process involving interactions of high pressure fluid injections with a stressed hot rock mass, mechanical interaction of induced fractures with existing natural fractures, and the spatial and temporal variations of in-situ stress. As a result it is necessary to develop tools that can be used to study these interactions as an integral part of a comprehensive approach to geothermal reservoir development, particularly enhanced geothermal systems. In response to this need we have set out to develop advanced thermo-mechanical models for design of artificial fractures and rock fracture research in geothermal reservoirs. These models consider the significant hydraulic and thermo-mechanical processes and their interaction with the in-situ stress state. Wellbore failure and fracture initiation is studied using a model that fully couples poro-mechanical and thermo-mechanical effects. The fracture propagation model is based on a complex variable and regular displacement discontinuity formulations. In the complex variable approach the displacement discontinuities are

Experimental studies of thermo-hydraulic processes during passive safety systems operation in new WWER NPP projects

International Nuclear Information System (INIS)

Morozov, A.V.; Remizov, O.V.; Kalyakin, D.S.

2014-01-01

The results of experimental study of thermal-hydraulic processes during operation of the passive safety systems of WWER reactors of new generation are given. The interaction processes of counter flows of saturated steam and cold water in vertical steam-line of the auxiliary passive core reflood system from secondary hydraulic accumulator are studied. The peculiarities of undeveloped boiling on single horizontal tube heating by steam and steam-gas mixture, which is character for WWER steam generator condensing mode, are investigated [ru

Free-standing thermo-responsive nanoporous membranes from high molecular weight PS-PNIPAM block copolymers synthesized via RAFT polymerization

NARCIS (Netherlands)

Cetintas, Merve; de Grooth, Joris; Hofman, Anton H.; van der Kooij, Hanne M.; Loos, Katja; de Vos, Wiebe Matthijs; Kamperman, Marleen

2017-01-01

The incorporation of stimuli-responsive pores in nanoporous membranes is a promising approach to facilitate the cleaning process of the membranes. Here we present fully reversible thermo-responsive nanoporous membranes fabricated by self-assembly and non-solvent induced phase separation (SNIPS) of

Parametric study of the stability properties of a thermo hydraulic channel coupled to punctual kinetics; Estudio parametrico de las propiedades de estabilidad de un canal termohidraulico acoplado a cinetica puntual

Energy Technology Data Exchange (ETDEWEB)

Cecenas F, M.; Campos G, R.M. [Instituto de Investigaciones Electricas, Reforma 113, Col. Palmira, Temixco, Morelos (Mexico)]. e-mail: mcf@iie.org.mx

2005-07-01

The reason of decay is the indicator of stability usually used in the literature to evaluate stability of boiling water reactors, however, in the operation of this type of reactors is considered the length of boiling like an auxiliary parameter for the evaluation of stability. In this work its are studied the variation of these two indicators when modifying a given an operation parameter in a model of a thermo hydraulic channel coupled to punctual kinetics, maintaining all the other input constant variables. The parameters selected for study are the axial profile of power, the subcooling, the flow of coolant and the thermal power. The study is supplemented by means of real data of plant using the one Benchmark of Ringhals, and the results for the case of the ratio of decay its are compared with the decay reasons obtained by means of autoregression models of the local instrumentation of neutron flux. (Author)

Mid-sized omnidirectional robot with hydraulic drive and steering

Science.gov (United States)

Wood, Carl G.; Perry, Trent; Cook, Douglas; Maxfield, Russell; Davidson, Morgan E.

2003-09-01

Through funding from the US Army-Tank-Automotive and Armaments Command's (TACOM) Intelligent Mobility Program, Utah State University's (USU) Center for Self-Organizing and Intelligent Systems (CSOIS) has developed the T-series of omni-directional robots based on the USU omni-directional vehicle (ODV) technology. The ODV provides independent computer control of steering and drive in a single wheel assembly. By putting multiple omni-directional (OD) wheels on a chassis, a vehicle is capable of uncoupled translational and rotational motion. Previous robots in the series, the T1, T2, T3, ODIS, ODIS-T, and ODIS-S have all used OD wheels based on electric motors. The T4 weighs approximately 1400 lbs and features a 4-wheel drive wheel configuration. Each wheel assembly consists of a hydraulic drive motor and a hydraulic steering motor. A gasoline engine is used to power both the hydraulic and electrical systems. The paper presents an overview of the mechanical design of the vehicle as well as potential uses of this technology in fielded systems.

Suppression of Squeal Noise Excited by the Pressure Pulsation from the Flapper-Nozzle Valve inside a Hydraulic Energy System

Directory of Open Access Journals (Sweden)

Meng Chen

2018-04-01

Full Text Available Squeal noise often occurs in a two-stage electrohydraulic servo-valve, which is an unfavorable issue of modern hydraulic energy systems. The root causes of such noise from the servo-valve are still unclear. The objective of this paper is to explore the noise mechanism in a servo-valve excited by the pressure pulsations from the hydraulic energy system perspective. The suppressing capability of squeal noise energy is investigated by changing the pressure pulsation frequency and natural frequency of the flapper-armature assembly. The frequencies of the pressure pulsations are adjusted by setting different speeds of the hydraulic pump varying from 10,400–14,400 rpm, and two flapper-armature assemblies with different armature lengths are used in the tested hydraulic energy system. The first eight vibration mode shapes and natural frequencies of the flapper-armature assembly are obtained by numerical modal analysis using two different armature lengths. The characteristics of pressure pulsations at the pump outlet and in the chamber of the flapper-nozzle valve, armature vibration and noise are tested and compared with the natural frequencies of the flapper-armature assembly. The results reveal that the flapper-armature assembly vibrates and makes the noise with the same frequencies as the pressure pulsations inside the hydraulic energy system. Resonance appears when the frequency of the pressure pulsations coincides with the natural frequency of the flapper-armature assembly. Therefore, it can be concluded that the pressure pulsation energy from the power supply may excite the vibration of the flapper-armature assembly, which may consequently cause the squeal noise inside the servo-valve. It is verified by the numerical simulations and experiments that setting the pressure pulsation frequencies different from the natural frequencies of the flapper-armature assembly can suppress the resonance and squeal noise.

Thermo-hydro-mechanical modelling of buffer, synthesis report

International Nuclear Information System (INIS)

Toprak, E.; Mokni, N.; Olivella, S.; Pintado, X.

2013-08-01

This study addresses analyses of coupled thermo-hydro-mechanical (THM) processes in a scheme considered for the spent nuclear fuel repository in Olkiluoto (Finland). The finite element code CODE B RIGHT is used to perform modelling calculations. The objective of the THM modelling was to study some fundamental design parameters. The time required to reach full saturation, the maximum temperature reached in the canister, the deformations in the buffer-backfill interface, the stress-deformation balance between the buffer and the backfill, the swelling pressure developed and the homogenization process development are critical variables. Because of the complexity of the THM processes developed, only a single deposition hole has been modelled with realistic boundary conditions which take into account the entire repository. A thermal calculation has been performed to adopt appropriate boundary conditions for a reduced domain. The modelling has been done under axisymmetric conditions. As a material model for the buffer bentonite and backfill soil, the Barcelona Basic Model (BBM) has been used. Simulation of laboratory tests conducted at B and Tech under supervision of Posiva has been carried out in order to determine the fundamental mechanical parameters for modelling the behaviour of MX-80 bentonite using the BBM model. The modelling process of the buffer-backfill interface is an essential part of tunnel backfill design. The calculations will aim to determine deformations in this intersection, the behaviour of which is important for the buffer swelling. The homogenization process is a key issue as well. Porosity evolution during the saturation process is evaluated in order to check if the final saturated density accomplishes the homogenization requirements. This report also describes the effect of the existence of an air-filled gap located between the canister and the bentonite block rings in thermo-hydro-mechanical behaviour of the future spent nuclear fuel repository in

Thermo-hydro-mechanical modelling of buffer, synthesis report

Energy Technology Data Exchange (ETDEWEB)

Toprak, E.; Mokni, N.; Olivella, S. [Universitat Politecnica de Catalunya, Barcelona (Spain); Pintado, X. [B and Tech Oy, Helsinki (Finland)

2013-08-15

This study addresses analyses of coupled thermo-hydro-mechanical (THM) processes in a scheme considered for the spent nuclear fuel repository in Olkiluoto (Finland). The finite element code CODE{sub B}RIGHT is used to perform modelling calculations. The objective of the THM modelling was to study some fundamental design parameters. The time required to reach full saturation, the maximum temperature reached in the canister, the deformations in the buffer-backfill interface, the stress-deformation balance between the buffer and the backfill, the swelling pressure developed and the homogenization process development are critical variables. Because of the complexity of the THM processes developed, only a single deposition hole has been modelled with realistic boundary conditions which take into account the entire repository. A thermal calculation has been performed to adopt appropriate boundary conditions for a reduced domain. The modelling has been done under axisymmetric conditions. As a material model for the buffer bentonite and backfill soil, the Barcelona Basic Model (BBM) has been used. Simulation of laboratory tests conducted at B and Tech under supervision of Posiva has been carried out in order to determine the fundamental mechanical parameters for modelling the behaviour of MX-80 bentonite using the BBM model. The modelling process of the buffer-backfill interface is an essential part of tunnel backfill design. The calculations will aim to determine deformations in this intersection, the behaviour of which is important for the buffer swelling. The homogenization process is a key issue as well. Porosity evolution during the saturation process is evaluated in order to check if the final saturated density accomplishes the homogenization requirements. This report also describes the effect of the existence of an air-filled gap located between the canister and the bentonite block rings in thermo-hydro-mechanical behaviour of the future spent nuclear fuel

Thermo-hydraulic instability of natural circulation BWRs at low pressure star-up. Experimental estimation of instability region with test facility considering scaling law

International Nuclear Information System (INIS)

Inada, F.; Furuya, M.; Yasuo, A.; Tabata, H.; Yoshioka, Y.; Kim, H.T.

1995-01-01

In natural circulation BWRs developed for advanced light water reactors with simplified passive safety systems, thermo-hydraulic stability should be confirmed especially at low pressure start-up. In this paper, nondimensional parameters to estimate the hydrodynamic stability to reactors at low pressure start-up were obtained by transformation of the basic equations of drift-flux model in the two-phase region into nondimensional form. A test facility based on these parameters was then constructed. The height of the test facility is 70% of SBWR and many nondimensional test facility parameters are almost the same as those of the reactor. Reactor stability was estimated experimentally. Stability maps below 0.5MPa were obtained on the heat flux - channel inlet subcooling place. It was found that there were two stability boundaries, between which the flow became unstable. Flow was stable in the high and low channel inlet subcooling regions. Typical conditions of SBWR at low pressure start-up were noted in the high channel inlet subcooling stable region. The heat flux at typical SBWR start-up was about one fifth that of the stability boundary. Though some nondimensional parameters of the test facility did not exactly agree with those of SBWR, it was suggested that the flow in SBWR was stable below 0.5MPa because of the large margin. (author)

Clinical application of transcatheter arterial thermo-chemotherapy and thermo-lipiodol embolization in treatment of hepatocellular carcinoma

International Nuclear Information System (INIS)

Wang Xuan; Chen Xiaofei; Dong Weihua

2007-01-01

Objective: To evaluate the clinical efficacy of thermo-chemotherapy and thermo-lipiodol embolization in treatment of primary hepatocellular carcinoma(PHC). Methods: One hundred and sixteen cases of PHC were divided into three groups. Group A (38 cases)was treated with normal temperature chemotherapy and normal temperature lipiodol, Group B(40 cases)with thermo-chemotherapy and normal temperature lipiodol and group C (38 cases)with thermo-chemotherapy and thermo-lipiodol. Group B and group C were called the thermotherapy group. Results: In the thermotherapy groups, the rates of tumor size reduction were significantly greater than those in the normal group. There were no significant different in the hepatic function tests among the three groups. The 6-, 12-, 18-, and 24- month survival rates of the normal group and thermotherapy groups were 97%, 58%, 39% and 18%, versus 99%, 79%, 57% and 36%, respectively. No significant differences were found in the rates of reduction of tumor size and survival rates between group B and group C. Conclusion: Thermo-chemotherapy and thermo-embolization possess significant effect on PHC but without conspicuous damage to liver function. (authors)

Hydraulic modelling of the CARA Fuel element

International Nuclear Information System (INIS)

Brasnarof, Daniel O.; Juanico, Luis; Giorgi, M.; Ghiselli, Alberto M.; Zampach, Ruben; Fiori, Jose M.; Yedros, Pablo A.

2004-01-01

The CARA fuel element is been developing by the National Atomic Energy Commission for both Argentinean PHWRs. In order to keep the hydraulic restriction in their fuel channels, one of CARA's goals is to keep its similarity with both present fuel elements. In this paper is presented pressure drop test performed at a low-pressure facility (Reynolds numbers between 5x10 4 and 1,5x10 5 ) and rational base models for their spacer grid and rod assembly. Using these models, we could estimate the CARA hydraulic performance in reactor conditions that have shown to be satisfactory. (author) [es

The applicability of CFD to simulate and study the mixing process and the thermo-hydraulic consequences of a main steam line break in PWR model

Directory of Open Access Journals (Sweden)

Farkas Istvan

2017-01-01

Full Text Available This paper focuses on the validation and applicability of CFD to simulate and analyze the thermo-hydraulic consequences of a main steam line break. Extensive validation data come from experiments performed using the Rossendorf coolant mixing model facility. For the calculation, the range of 9 to 12 million hexahe¬dral cells was constructed to capture all details in the interrogation domain in the system. The analysis was performed by running a time-dependent calculation, Detailed analyses were made at different cross-sections in the system to evaluate not only the value of the maximum and minimum temperature, but also the loca¬tion and the time at which it occurs during the transient which is considered to be indicator for the quality of mixing in the system. CFD and experimental results were qualitatively compared; mixing in the cold legs with emergency core cooling systems was overestimated. This could be explained by the sensitivity to the bound¬ary conditions. In the downcomer, the experiments displayed higher mixing: by our assumption this related to the dense measurement grid (they were not modelled. The temperature distribution in the core inlet plane agreed with the measurement results. Minor deviations were seen in the quantitative comparisons: the maximum temperature difference was 2ºC.

System Design and Performance Test of Hydraulic Intensifier

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Eui; Lee, Gi Chun [Korea Institute of Machinery and Materials, Daejeon (Korea, Republic of); Kim, Jae Hoon [Chungnam National University, Daejeon (Korea, Republic of)

2010-07-15

Components such as pressure vessel, hydraulic hose assembly, accumulator, hydraulic cylinder, hydraulic valve, pipe, etc., are tested under the impulse-pressure conditions prescribed in ISO and SAE standards. The impulse pressure test machine needs to have a high pressure, a precise control system and a long life. It should satisfy the requirements for fabrication of the impulse tester to generate ultra high pressure in the hydraulic system. In the impulse tester, a servo-valve control system is adopted; although the control application is convenient, it is expensive owing to the cost of developing the system. The type of the control system determines the pressure wave, which affects the components that are tested. In this study, the manufacturing process and the intensifier system design related to the flow, pressure, and the increasing rate of pressure are investigated. The results indicate the ultra high pressure waves in the system.

Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations

Energy Technology Data Exchange (ETDEWEB)

Pacio, J., E-mail: Julio.pacio@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Wetzel, T. [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Doolaard, H.; Roelofs, F. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Van Tichelen, K. [Belgian Nuclear Reseach Center (SCK-CEN), Boeretang 200, Mol (Belgium)

2017-02-15

Heavy liquid metals (HLMs), such as lead-bismuth eutectic (LBE) and pure lead are prominent candidate coolants for many advanced systems based on fast neutrons. In particular, LBE is used in the first-of-its-kind MYRRHA fast reactor, to be built in Mol (Belgium), which can be operated either in critical mode or as a sub-critical accelerator-driven system. With a strong focus on safety, key thermal-hydraulic aspects of these systems, such as the proper cooling of fuel assemblies, must be assessed. Considering the complex geometry and low Prandtl number of LBE (Pr ∼ 0.025), this flow scenario is challenging for the models used in Computational Fluid Dynamics (CFD), e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data for the relevant scenario are needed for validation. In this general context, this topic is studied both experimentally and numerically in the framework of the European FP7 project SEARCH (2011–2015). An experimental campaign, including a 19-rod bundle with wire spacers, cooled by LBE is undertaken at KIT. With prototypical geometry and operating conditions, it is intended to evaluate the validity of current empirical correlations for the MYRRHA conditions and, at the same time, to provide validation data for the CFD simulations performed at NRG. The results of one benchmarking case are presented in this work. Moreover, this validated approach is then used for simulating a complete MYRRHA fuel assembly (127 rods).

Thermo-elastic optical coherence tomography.

Science.gov (United States)

Wang, Tianshi; Pfeiffer, Tom; Wu, Min; Wieser, Wolfgang; Amenta, Gaetano; Draxinger, Wolfgang; van der Steen, Antonius F W; Huber, Robert; Soest, Gijs van

2017-09-01

The absorption of nanosecond laser pulses induces rapid thermo-elastic deformation in tissue. A sub-micrometer scale displacement occurs within a few microseconds after the pulse arrival. In this Letter, we investigate the laser-induced thermo-elastic deformation using a 1.5 MHz phase-sensitive optical coherence tomography (OCT) system. A displacement image can be reconstructed, which enables a new modality of phase-sensitive OCT, called thermo-elastic OCT. An analysis of the results shows that the optical absorption is a dominating factor for the displacement. Thermo-elastic OCT is capable of visualizing inclusions that do not appear on the structural OCT image, providing additional tissue type information.

Study of the Thermo-Mechanical Behavior of the CLIC Two-Beam Modules

CERN Document Server

Rossi, F; Riddone, G; Österberg, K; Kossyvakis, I; Gudkov, D; Samochkine, A

2013-01-01

The final luminosity target of the Compact LInear Collider (CLIC) imposes a micron-level stability requirement on the two-meter repetitive two-beam modules constituting the main linacs. Two-beam prototype modules are being assembled to extensively study their thermo-mechanical behaviour under different operation modes. The power dissipation occurring in the modules will be reproduced and the efficiency of the corresponding cooling systems validated. At the same time, the real environmental conditions present in the CLIC tunnel will be studied. Air conditioning and ventilation systems have been installed in the dedicated laboratory. The air temperature will be changed from 20 to 40°C, while the air flow rate will be varied up to 0.8 m/s. During all experimental tests, the alignment of the RF structures will be monitored to investigate the influence of power dissipation and air temperature on the overall thermo-mechanical behaviour. \

A thermo-mechanical benchmark calculation of an hexagonal can in the BTI accident with ABAQUS code

International Nuclear Information System (INIS)

Zucchini, A.

1988-07-01

The thermo-mechanical behaviour of an hexagonal can in a benchmark problem (simulating the conditions of a BTI accident in a fuel assembly) is examined by means of the ABAQUS code: the effects of the geometric nonlinearity are shown and the results are compared with those of a previous analysis performed with the INCA code. (author)

Maximum thermal loading test of BWR fuel assembly

International Nuclear Information System (INIS)

Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

1987-01-01

Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

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

Aqueous corrosion in static capsule tests representing multi-metal assemblies in the hydraulic circuit of Tore Supra

Energy Technology Data Exchange (ETDEWEB)

Lipa, M. [Association Euratom-CEA, CEA/DSM/DRFC, Centre de Cadarache, 13108 Saint-Paul-Lez-Durance (France)], E-mail: manfred.lipa@cea.fr; Blanchet, J.; Feron, D. [CEA/DEN/SCCME, Centre de Saclay, 91191 Gif sur Yvette (France); Cellier, F. [AREVA ANP, Centre Technique, 71380 Saint Marcel (France)

2008-12-15

Tore supra (TS) in vessel components represent a unique combination of metals in the hydraulic circuit. Different materials, e.g. stainless steel, copper alloys, nickel, etc., were joined together by fusion welding, brazing and friction. Since the operation and baking temperature of all in vessel components has been defined to be set at 230 deg. C/40 bars a special water chemistry of the cooling water loop was suggested in order to prevent eventual water leaks due to corrosion at relative high temperatures and pressures in tubes, bellows, coils and coolant plant ancillary equipments. Following experiences with water chemistry in Pressurised Water Reactors, an all volatile chemical treatment (AVT) has been defined for the cooling water quality of TS. Since then, a simplified static (no fluid circulation) corrosion test program at relatively high temperature and pressure has been performed using capsule-type samples made of above mentioned multi-metal assemblies.

Thermo-plasmonics of Irradiated Metallic Nanostructures

DEFF Research Database (Denmark)

Ma, Haiyan

Thermo-plasmonics is an emerging field in photonics which aims at harnessing the kinetic energy of light to generate nanoscopic sources of heat. Localized surface plasmons (LSP) supported by metallic nanostructures greatly enhance the interactions of light with the structure. By engineering...... delivery, nano-surgeries and thermo-transportations. Apart from generating well-controlled temperature increase in functional thermo-plasmonic devices, thermo-plasmonics can also be used in understanding complex phenomena in thermodynamics by creating drastic temperature gradients which are not accessible...... using conventional techniques. In this thesis, we present novel experimental and numerical tools to characterize thermo-plasmonic devices in a biologically relevant environment, and explore the thermodiffusion properties and measure thermophoretic forces for particles in temperature gradients ranging...

Fabrication details for wire wrapped fuel assembly components

International Nuclear Information System (INIS)

Bosy, B.J.

1978-09-01

Extensive hydraulic testing of simulated LMFBR blanket and fuel assemblies is being carried out under this MIT program. The fabrication of these test assemblies has involved development of manufacturing procedures involving the wire wrapped pins and the flow housing. The procedures are described in detail in the report

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)

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)

Development and validation of the ASTEC-Na thermal-hydraulic models

Energy Technology Data Exchange (ETDEWEB)

Herranz, L. E.; Perez, S.; Bandini, G.; Jacq, F.; Parisi, C.; Berna, C.

2014-07-01

Last years the interest in sodium-cooled fast reactors (SFR) has been fostered worldwide by the search for higher nuclear energy sustainability. This has been reflected in the various international initiatives like GEN-IV International Forum, INPRO or ESNII platforms. At the same time, innovative nuclear reactor designs, particularly SFR, are aiming at even higher safety standards than current LWRs. A proof of it is the consideration of severe accidents since the earliest stages of reactor design. commonalities of LWR and SFR severe accident scenarios suggest that some of the knowledge achieved in the LWR arena might be applicable to some extent to SFRs. This is the spirit underneath of the EU-JASMIN project, which generic goal is developing the ASTEC-Na code from the LWR ASTEC platform. This will entail to t extend and adapt some existing models as well as to implement new ones in all the areas covered, from neutronics and pin thermo-mechanics and pin thermo-mechanics to the in-containment source term behavior by these, going through the indispensable Na thermal-hydraulics. (Author)

High fidelity thermal-hydraulic analysis using CFD and massively parallel computers

International Nuclear Information System (INIS)

Weber, D.P.; Wei, T.Y.C.; Brewster, R.A.; Rock, Daniel T.; Rizwan-uddin

2000-01-01

Thermal-hydraulic analyses play an important role in design and reload analysis of nuclear power plants. These analyses have historically relied on early generation computational fluid dynamics capabilities, originally developed in the 1960s and 1970s. Over the last twenty years, however, dramatic improvements in both computational fluid dynamics codes in the commercial sector and in computing power have taken place. These developments offer the possibility of performing large scale, high fidelity, core thermal hydraulics analysis. Such analyses will allow a determination of the conservatism employed in traditional design approaches and possibly justify the operation of nuclear power systems at higher powers without compromising safety margins. The objective of this work is to demonstrate such a large scale analysis approach using a state of the art CFD code, STAR-CD, and the computing power of massively parallel computers, provided by IBM. A high fidelity representation of a current generation PWR was analyzed with the STAR-CD CFD code and the results were compared to traditional analyses based on the VIPRE code. Current design methodology typically involves a simplified representation of the assemblies, where a single average pin is used in each assembly to determine the hot assembly from a whole core analysis. After determining this assembly, increased refinement is used in the hot assembly, and possibly some of its neighbors, to refine the analysis for purposes of calculating DNBR. This latter calculation is performed with sub-channel codes such as VIPRE. The modeling simplifications that are used involve the approximate treatment of surrounding assemblies and coarse representation of the hot assembly, where the subchannel is the lowest level of discretization. In the high fidelity analysis performed in this study, both restrictions have been removed. Within the hot assembly, several hundred thousand to several million computational zones have been used, to

Regional-scale geomechanical impact assessment of underground coal gasification by coupled 3D thermo-mechanical modeling

Science.gov (United States)

Otto, Christopher; Kempka, Thomas; Kapusta, Krzysztof; Stańczyk, Krzysztof

2016-04-01

Underground coal gasification (UCG) has the potential to increase the world-wide coal reserves by utilization of coal deposits not mineable by conventional methods. The UCG process involves combusting coal in situ to produce a high-calorific synthesis gas, which can be applied for electricity generation or chemical feedstock production. Apart from its high economic potentials, UCG may induce site-specific environmental impacts such as fault reactivation, induced seismicity and ground subsidence, potentially inducing groundwater pollution. Changes overburden hydraulic conductivity resulting from thermo-mechanical effects may introduce migration pathways for UCG contaminants. Due to the financial efforts associated with UCG field trials, numerical modeling has been an important methodology to study coupled processes considering UCG performance. Almost all previous UCG studies applied 1D or 2D models for that purpose, that do not allow to predict the performance of a commercial-scale UCG operation. Considering our previous findings, demonstrating that far-field models can be run at a higher computational efficiency by using temperature-independent thermo-mechanical parameters, representative coupled simulations based on complex 3D regional-scale models were employed in the present study. For that purpose, a coupled thermo-mechanical 3D model has been developed to investigate the environmental impacts of UCG based on a regional-scale of the Polish Wieczorek mine located in the Upper Silesian Coal Basin. The model size is 10 km × 10 km × 5 km with ten dipping lithological layers, a double fault and 25 UCG reactors. Six different numerical simulation scenarios were investigated, considering the transpressive stress regime present in that part of the Upper Silesian Coal Basin. Our simulation results demonstrate that the minimum distance between the UCG reactors is about the six-fold of the coal seam thickness to avoid hydraulic communication between the single UCG

Thermo-hydraulic behaviour of Boom Clay using a heating cell. experimental and numerical study

International Nuclear Information System (INIS)

Lima, A.; Romero, E.; Vaunat, J.; Gens, A.; Li, X.L.

2010-01-01

Document available in extended abstract form only. Boom clay is the subject of extensive research in Belgium dealing with all phenomena that may possibly affect the performance of this geological formation as potential radioactive waste repository. Specifically, thermal loads may play an important role on this low-permeability clay. There are a number of laboratory results concerning the saturated hydro-mechanical behaviour of Boom clay under different temperatures and recent studies on this area are described in Le (2008). Nevertheless, information on clay hydro-mechanical response on heating and cooling paths under small-scale laboratory conditions is less known. To this aim, non-isothermal tests on intact borehole samples were carried out using an axisymmetric heating cell described in Munoz et al. (2009). Heating and cooling paths under nearly constant volume and different target temperatures (maximum 85 deg. C) were performed under controlled hydraulic boundary conditions. The paper presents results of an exhaustive experimental programme performed on a fully-instrumented cell (sample 75 mm in diameter and 100 mm high) with a controlled-power heater installed along the axis of the sample. Different transducers were monitoring the sample response: two miniature pore water pressure transducers located at different heights of the lateral wall of the cell, three thermocouples (two at equivalent locations in relation to the pressure transducers and one near the heater), and top and lateral strain gauges attached to reduced thickness sections of the walls. The cell has top and bottom valves to apply the hydraulic conditions. The protocol of the tests presented three important phases: hydration, heating and cooling. Throughout the heating and cooling phases, the bottom drainage was maintained open at a constant water pressure of 1 MPa using an automatic pressure/volume controller, while the upper valve was kept closed. This back-pressure was important since it

Analysis of the thermo-chemo-mechanical behavior of massive concrete structures oat early-age

International Nuclear Information System (INIS)

Honorio, T.; Bary, B.; Benboudjema, F.

2014-01-01

The prediction of the thermo-chemo-mechanical behavior of concrete structures at early ages is important in the context of the feasibility of massive structures. Different phenomena affecting the thermal response of the structure are studied, namely the influence of the change on convection conditions due to wind, the influence of solar radiation, the influence of ambient temperature and the influence of assembly date. A mechanical analysis accounting for autogenous shrinkage and creep strains, besides thermal strains, is performed for the latter case. The results point out the importance of considering the solar radiation and wind conditions on the thermal response of the structure. The ambient temperature impacts directly the maximum temperature reached within the structure. Finally, although the temperature profiles seem just to shift according to the assembly date, the mechanical response is less favorable to early assembly dates. (authors)

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

International Nuclear Information System (INIS)

Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

2015-01-01

Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

Energy Technology Data Exchange (ETDEWEB)

Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

2015-02-15

Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

Hydraulic release oil tool

International Nuclear Information System (INIS)

Mims, M.G.; Mueller, M.D.; Ehlinger, J.C.

1992-01-01

This patent describes a hydraulic release tool. It comprises a setting assembly; a coupling member for coupling to drill string or petroleum production components, the coupling member being a plurality of sockets for receiving the dogs in the extended position and attaching the coupling member the setting assembly; whereby the setting assembly couples to the coupling member by engagement of the dogs in the sockets of releases from and disengages the coupling member in movement of the piston from its setting to its reposition in response to a pressure in the body in exceeding the predetermined pressure; and a relief port from outside the body into its bore and means to prevent communication between the relief port and the bore of the body axially of the piston when the piston is in the setting position and to establish such communication upon movement of the piston from the setting position to the release position and reduce the pressure in the body bore axially of the piston, whereby the reduction of the pressure signals that the tool has released the coupling member

On numerical simulation of fuel assembly bow in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Horváth, Ákos, E-mail: akoshorvath@t-online.hu [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Budapest University of Technology and Economics, Department of Aircraft and Ships, Stoczek Street 6, Building J, H-1111 Budapest (Hungary); Dressel, Bernd [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany)

2013-12-15

Highlights: • Simulation of fuel assembly bow by coupled CFD and finite element method. • Comparison of calculated and experimentally measured bow shapes. • Investigation of boundary condition effect on bow pattern of a fuel assembly row. • Highlighting importance of consideration of fluid–structure interaction. • Assessment of flow redistribution within the fuel assembly row model. - Abstract: Fuel assembly bow in pressurized water reactor cores is largely triggered by lateral hydraulic forces together with creep processes generated by neutron flux. A detailed understanding of the flow induced bow behaviour is, therefore, an important issue. The experimental feedbacks and laboratory tests on fuel assembly bow show that it is characterized to a high degree by fluid–structure interaction (FSI) effects, therefore, consideration of FSI is essential and indispensable in full comprehension of the bow mechanism. In the present study, coupled computational fluid dynamics (CFD) and finite element simulations are introduced, calculating fuel assembly deformation under different conditions as a quasi-stationary phenomenon. The aim has been, on the one hand, to develop such a simplified fuel assembly CFD model, which allows set up of fuel assembly rows without loosing its main hydraulic characteristic; on the other hand, to investigate the bow pattern of a given fuel assembly row under different boundary conditions. The former one has been achieved by comparing bow shapes obtained with different fuel assembly (spacer grid) modelling approaches and mesh resolutions with experimental data. In the second part of the paper a row model containing 7.5 fuel assemblies is introduced, investigating the effect of flow distribution at inlet and outlet boundary regions on fuel assembly bow behaviour. The post processing has been focused on the bow pattern, lateral hydraulic forces, and horizontal flow distribution. The results have revealed importance of consideration of

Heat transfer and friction characteristics of rotor-assembled strand heat exchanger studied by uniform design experiment

Directory of Open Access Journals (Sweden)

Yan Wei

2015-10-01

Full Text Available The uniform distribution and experimental design is employed to study the thermo-hydraulic characteristics of a heat exchanger, which consists of the rotor-assembled strands mounted in circular smooth tubes. The uniform distribution and experimental design parameters include multiple rotor parameters such as rotor diameters, rotor lead, and height of blade, with the aim of studying their influence on the PEC, that is, ( ( Nu z / Nu g / ( f g / f z 1 / 3 , which stands for the heat transfer and friction characteristics. The best matching schemes of rotor-assembled strand, which significantly improves PEC to 2.01, are given by the regression analysis of uniform distribution and experimental design table. The single-factor experiments are performed to compare a tube installed with different kinds of rotor-assembled strands with a smooth tube without any strands when the Reynolds number changes between 20,000 and 60,000. The experimental result is in good agreement with the result obtained by the regression analysis of uniform distribution and experimental design. It is shown that the rotor diameters play important role in the heat transfer, and the optimal PEC value is obtained under the case that the rotor diameter is 21 mm. The rotor lead also contributes to the improvement of heat transfer and its optimal value is 700 mm in this study. The Nusselt number, friction factor and PEC increase with the increase in blade height. It shows that the uniform distribution and experimental design is an efficient method to find out the optimal parameters.

Spacer grid with mixing blades for nuclear fuel assembly

International Nuclear Information System (INIS)

Noailly, J.

1986-01-01

The spacer grid for nuclear fuel assembly has two sets of intersecting metal plates provided with blades and defining cells. The plates are fitted only with half-blades associated with a single grid opening. The half-blades of adjacent cells are arranged at 90deg C to each other and each plate has at most one half-blade at each corner of a cell. The invention concerns fuel assemblies of pressurized water reactors. The blades arranged on a single side of the plate provide a good hydraulic uniformity. The invention provides a uniform distribution of blades (and thus of absorbing material in each hydraulic cell) [fr

Liquid metal thermal-hydraulics

International Nuclear Information System (INIS)

Kottowski-Duemenil, H.M.

1994-01-01

This textbook is a report of the 26 years activity of the Liquid Metal Boiling Working Group (LMBWG). It summarizes the state of the art of liquid metal thermo-hydraulics achieved through the collaboration of scientists concerned with the development of the Fast Breeder Reactor. The first chapter entitled ''Liquid Metal Boiling Behaviour'', presents the background and boiling mechanisms. This section gives the reader a brief but thorough survey on the superheat phenomena in liquid metals. The second chapter of the text, ''A Review of Single and Two-Phase Flow Pressure Drop Studies and Application to Flow Stability Analysis of Boiling Liquid Metal Systems'' summarizes the difficulty of pressure drop simulation of boiling sodium in core bundles. The third chapter ''Liquid Metal Dry-Out Data for Flow in Tubes and Bundles'' describes the conditions of critical heat flux which limits the coolability of the reactor core. The fourth chapter dealing with the LMFBR specific topic of ''Natural Convection Cooling of Liquid Metal Systems''. This chapter gives a review of both plant experiments and out-of-pile experiments and shows the advances in the development of computing power over the past decade of mathematical modelling ''Subassembly Blockages Suties'' are discussed in chapter five. Chapter six is entitled ''A Review of the Methods and Codes Available for the Calculation on Thermal-Hydraulics in Rod-Cluster and other Geometries, Steady state and Transient Boiling Flow Regimes, and the Validation achieves''. Codes available for the calculation of thermal-hydraulics in rod-clusters and other geometries are reviewed. Chapter seven, ''Comparative Studies of Thermohydraulic Computer Code Simulations of Sodium Boiling under Loss of Flow Conditions'', represents one of the key activities of the LMBWG. Several benchmark exercises were performed with the aim of transient sodium boiling simulation in single channels and bundle blockages under steady state conditions and loss of

Thermo-sensitive intelligent track membrane

International Nuclear Information System (INIS)

Pang Deling; Ren Lihua; Qian Zhilin; Huang Gang; Zhang Jinhua

1999-01-01

Using N-isopropylacryl-amide (NIP AAm) thermo-sensitive function material as monomer and nuclear track microporous membrane (NTMM) as baseline material, a thermo-sensitive intelligent track membrane (TsITM) has been prepared by the over-oxidization and pre-irradiation grafting techniques. The TsITM can be used to make a micro-switch controlled by temperature and to adjust particle screening and osmosis. To obtain sub-micron responsive grafted track pores only a very thin thermo-sensitive layer is needed. The TsITM pores are capable of swelling and shrinking rapidly and respond more sensitively to temperature

The AFA 3G fuel assembly: a proven design for high burnups

International Nuclear Information System (INIS)

Forat, C.; Florentin, F.

1999-01-01

The AFA 3G fuel assembly design is based on the wide experience gained with the AFA 2G fuel assembly. More than 9500 AFA 2G fuel assemblies have been loaded in different reactors, worldwide, reaching discharged burnups in the range of 45 - 55 GWd/tU. This experience confirmed the features of the AFA 2G, such as the grids and the vanes arrangement for thermal hydraulic performance, the concept of the fuel rod support within the grid which avoids any rod fretting or vibration phenomenon, the efficiency of the anti-debris device. The AFA 3G also relies on and benefits from the results of the world's largest R and D program, in-pile and out-of(pile testing by Framatome with EDF and CEA, with a special focus on corrosion-resistant fuel rod cladding. The AFA 3G exhibits the following enhancements: a reinforced structure, which improves resistance to assembly bow as well as its consequences in terms of RCCA insertion fuel handling and core physics obtained from: MONOBLOC TM guide thimbles, characterized by a thickened and enlarged tube and reinforced dash-pot; a hold down spring system which has been optimized to accommodate fuel assembly hydraulic lift-off forces and to meet the fuel assembly bow resistance requirement; widened recrystallized Zircaloy-4 spacer grids; a high resistance to corrosion due to the M5 TM Zirconium-Niobium-Oxygen alloy for the fuel rod cladding, which contributes also to the bow resistance of the fuel assembly; an enhanced thermal-hydraulic behavior promoted by well proven mixing vane array of AFA 2G spacer grids, combined with three additional Mid Span Mixing Grids; a very effective debris protection with the use of the TRAPPER TM bottom nozzle. With these improvements, the AFA 3G fuel assembly is able to reach discharge burnup of 60 GWd/tU with margins on important characteristics like corrosion behavior, assembly bow and thermal-hydraulic performance. The AFA 3G design is so convincing that major utilities have decided to shift their fuel

Thermo-elastic optical coherence tomography

NARCIS (Netherlands)

Wang, Tianshi; Pfeiffer, Tom; Wu, Min; Wieser, Wolfgang; Amenta, Gaetano; Draxinger, Wolfgang; van der Steen, A.F.W.; Huber, Robert; Van Soest, Gijs

2017-01-01

The absorption of nanosecond laser pulses induces rapid thermo-elastic deformation in tissue. A sub-micrometer scale displacement occurs within a few microseconds after the pulse arrival. In this Letter, we investigate the laser-induced thermo-elastic deformation using a 1.5 MHz phase-sensitive

Preliminary study on the feasibility of ductless fuel assembly for fast reactors

International Nuclear Information System (INIS)

Shibahara, Itaru; Enokido, Yuji

1988-01-01

Preliminary study on the feasibility of ductless fuel assembly for fast reactors has been conducted. The primary concern is with forecasting the thermal hydraulic characteristics and the heat removal efficiency from the core. The thermal hydraulic analysis revealed the coolant mixing in the core at steady state operating condition was not intensive and the coolant temperature increase was almost proportional to the power of each assembly. The hot spot analysis of the ductless core indicated that the hottest temperature in the core could be comparable with the temperature of the conventional ducted core, even in case the radial power flattening was not actively pursued but with adopting ducted radial blanket assemblies. Under off-normal conditions, the ductless core had improved heat removal capability which was caused by inter-assembly coolant flow. The study has indicated the feasibility of the ductless fuel assembly for fast reactors. The experiments to demonstrate the feasibility will be the next key process for the development. (author)

Thermal-hydraulics of actinide burner reactors

International Nuclear Information System (INIS)

Takizuka, Takakazu; Mukaiyama, Takehiko; Takano, Hideki; Ogawa, Toru; Osakabe, Masahiro.

1989-07-01

As a part of conceptual study of actinide burner reactors, core thermal-hydraulic analyses were conducted for two types of reactor concepts, namely (1) sodium-cooled actinide alloy fuel reactor, and (2) helium-cooled particle-bed reactor, to examine the feasibility of high power-density cores for efficient transmutation of actinides within the maximum allowable temperature limits of fuel and cladding. In addition, calculations were made on cooling of actinide fuel assembly. (author)

Physics of thermo-acoustic sound generation

Science.gov (United States)

Daschewski, M.; Boehm, R.; Prager, J.; Kreutzbruck, M.; Harrer, A.

2013-09-01

We present a generalized analytical model of thermo-acoustic sound generation based on the analysis of thermally induced energy density fluctuations and their propagation into the adjacent matter. The model provides exact analytical prediction of the sound pressure generated in fluids and solids; consequently, it can be applied to arbitrary thermal power sources such as thermophones, plasma firings, laser beams, and chemical reactions. Unlike existing approaches, our description also includes acoustic near-field effects and sound-field attenuation. Analytical results are compared with measurements of sound pressures generated by thermo-acoustic transducers in air for frequencies up to 1 MHz. The tested transducers consist of titanium and indium tin oxide coatings on quartz glass and polycarbonate substrates. The model reveals that thermo-acoustic efficiency increases linearly with the supplied thermal power and quadratically with thermal excitation frequency. Comparison of the efficiency of our thermo-acoustic transducers with those of piezoelectric-based airborne ultrasound transducers using impulse excitation showed comparable sound pressure values. The present results show that thermo-acoustic transducers can be applied as broadband, non-resonant, high-performance ultrasound sources.

Transcatheter hepatic arterial thermo-chemotherapy and thermo-lipiodol embolization for the treatment of hepatic metastases from colorectal carcinoma

International Nuclear Information System (INIS)

Wang Xuan; Chen Xiaofei

2009-01-01

Objective: To evaluate the clinical efficacy of transcatheter hepatic arterial thermo-chemotherapy and thermo-lipiodol embolization in the treatment of hepatic metastases from colorectal carcinoma. Methods: Sixty-eight cases with hepatic metastases from colorectal carcinoma were equally and randomly divided into two groups. The patients in study group were treated with transcatheter hepatic arterial thermo-chemotherapy and thermo-lipiodol embolization, while the patients in control group were treated with conventional (normal temperature) transcatheter hepatic arterial chemotherapy lipiodol embolization. Results: The effective rate of study group and control group was 65%(22/34) and 32%(11/34) respectively, the difference between two groups was statistically significant (P<0.05). No significant difference in the postoperative changes of hepatic function tests was found between the two groups. The survival rate at 6,12,18 and 24 months after the treatment was 100%, 82%, 44% and 18% respectively in study group, while it was 91%, 47%, 15% and 6% respectively in control group. Conclusion: Transcatheter hepatic arterial thermo-chemotherapy and thermo-lipiodol embolization is an effective and safe treatment for the hepatic metastases from colorectal carcinoma and has no obvious damage to the hepatic function. (authors)

The Sentinel-4 UVN focal plane assemblies

Science.gov (United States)

Hinger, Jürgen; Hohn, Rüdiger; Gebhardt, Eyk; Reichardt, Jörg

2017-09-01

The Sentinel-4 UVN Instrument is a dispersive imaging spectrometer covering the UV-VIS and the NIR wavelength. It is developed and built under an ESA contract by an industrial consortium led by Airbus Defence and Space. It will be accommodated on board of the MTG-S (Meteosat Third Generation - Sounder) satellite that will be placed in a geostationary orbit over Europe sampling data for generating two-dimensional maps of a number of atmospheric trace gases. The incoming light is dispersed by reflective gratings and detected by the two (UVVIS and NIR) CCDs mounted inside the focal plane assemblies. Both CCD detectors acquire spectral channels and spatial sampling in two orthogonal directions and will be operated at about 215 K mainly to minimize random telegraph signal effects and to reduce dark current. Stringent detector temperature as well as alignment stability requirements of less than +/-0.1 K per day respectively of less than 2 micrometers/2 arcseconds from ground to orbit are driving the FPA thermo-mechanical design. A specific FPA design feature is the redundant LED-calibration system for bad pixel detection as well as pixel gain and linearity monitoring. This paper reports on the design and qualification of the Focal Plane Assemblies with emphasis on thermo-mechanical as well as alignment stability verification.

Full vessel CFD analysis on thermal-hydraulic characteristics of CPR1000 PWR

International Nuclear Information System (INIS)

Chao Yanmeng; Yang Lixin; Zhang Mingqian

2014-01-01

To obtain flow distributions and thermal-hydraulic properties in a full vessel PWR under limited computation ability and time, a full vessel simulation model of CPR1000 was built based on two simplification methods. One simplified the inner geometry of the control rod guide tubes using equivalent flow area. Another substituted the core by a porous domain to maintain the pressure drop and temperature rise. After the computation, global and localized flow distributions, hydraulic loads of some main assemblies were obtained, as well as other thermal-hydraulic properties. The results indicate the flow distribution in the full vessel is asymmetrical. Therefore it is essential to use the full vessel model to simulate. The calculated thermal-hydraulic characteristics agree well with the operation statistics, providing the reference data for the reactor safety operation. (authors)

Equations of macrotransport in reactor fuel assemblies

International Nuclear Information System (INIS)

Sorokin, A.P.; Zhukov, A.V.; Kornienko, Yu.N.; Ushakov, P.A.

1986-01-01

The rigorous statement of equations of macrotransport is obtained. These equations are bases for channel-by-channel methods of thermohydraulic calculations of reactor fuel assemblies within the scope of the model of discontinuous multiphase coolant flow (including chemical reactions); they also describe a wide range of problems on thermo-physical reactor fuel assembly justification. It has been carried out by smoothing equations of mass, momentum and enthalpy transfer in cross section of each phase of the elementary fuel assembly subchannel. The equation for cross section flows is obtaind by smoothing the equation of momentum transfer on the interphase. Interaction of phases on the channel boundary is described using the Stanton number. The conclusion is performed using the generalized equation of substance transfer. The statement of channel-by-channel method without the scope of homogeneous flow model is given

Sufficient conditions for Hadamard well-posedness of a coupled thermo-chemo-poroelastic system

Directory of Open Access Journals (Sweden)

Tetyana Malysheva

2016-01-01

Full Text Available This article addresses the well-posedness of a coupled parabolic-elliptic system modeling fully coupled thermal, chemical, hydraulic, and mechanical processes in porous formations that impact drilling and borehole stability. The underlying thermo-chemo-poroelastic model is a system of time-dependent parabolic equations describing thermal, solute, and fluid diffusions coupled with Navier-type elliptic equations that attempt to capture the elastic behavior of rock around a borehole. An existence and uniqueness theory for a corresponding initial-boundary value problem is an open problem in the field. We give sufficient conditions for the well-posedness in the sense of Hadamard of a weak solution to a fully coupled parabolic-elliptic initial-boundary value problem describing homogeneous and isotropic media.

Development of NTD Hydraulic Rotation System for Kijang Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Kang, Hanok; Park, Kijung; Park, Yongsoo; Kim, Seong Hoon; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The KJRR will be mainly utilized for isotope production, NTD (Neutron Transmutation Doping) production, and related research activities. During irradiation for the NTD process, the irradiation rigs containing the silicon ingot rotate at a constant speed to ensure precisely defined homogeneity of the irradiation. The NTDHRS requires only hydraulic piping conveniently routed to the rotating devices inside the reactor pool. The resulting layout leaves the pool area clear of obstructions which might obscure vision and hinder target handling for operators. Pump banks and control valves are located remotely in a dedicated plant room allowing easy access and online maintenance. The necessities and major characteristic of NTD hydraulic rotation system are described in this study. A new NTD hydraulic rotation system are being developed to rotate the irradiation rigs at a constant speed and supply cooling flow for the irradiation rigs and reflector assembly. The configuration of the NTD hydraulic rotation device is discussed and practical methods to improve the rotational performance are suggested.

Development of NTD Hydraulic Rotation System for Kijang Research Reactor

International Nuclear Information System (INIS)

Kang, Hanok; Park, Kijung; Park, Yongsoo; Kim, Seong Hoon; Park, Cheol

2014-01-01

The KJRR will be mainly utilized for isotope production, NTD (Neutron Transmutation Doping) production, and related research activities. During irradiation for the NTD process, the irradiation rigs containing the silicon ingot rotate at a constant speed to ensure precisely defined homogeneity of the irradiation. The NTDHRS requires only hydraulic piping conveniently routed to the rotating devices inside the reactor pool. The resulting layout leaves the pool area clear of obstructions which might obscure vision and hinder target handling for operators. Pump banks and control valves are located remotely in a dedicated plant room allowing easy access and online maintenance. The necessities and major characteristic of NTD hydraulic rotation system are described in this study. A new NTD hydraulic rotation system are being developed to rotate the irradiation rigs at a constant speed and supply cooling flow for the irradiation rigs and reflector assembly. The configuration of the NTD hydraulic rotation device is discussed and practical methods to improve the rotational performance are suggested

Thermal-hydraulic process for cooling, heating and power production with low-grade heat sources in residential sector

International Nuclear Information System (INIS)

Borgogno, R.; Mauran, S.; Stitou, D.; Marck, G.

2017-01-01

Highlights: • Assessment of solar thermal-hydraulic process for tri-generation application. • Choice of the most suitable working fluid pair (R1234yf/R1233zd). • Evaluation of the global annual performance in Mediterranean climate. • Global annual COP and heat amplification achieving 0.24 and 1.2 respectively. • Global annual performance achieving an electric efficiency of 3.7%. - Abstract: A new process based on thermal-hydraulic conversion actuated by low-grade thermal energy is investigated. Input thermal energy can be provided by the means of solar collectors, as well as other low temperature energy sources. In the following article, “thermo-hydraulic” term refers to a process involving an incompressible fluid used as an intermediate medium to transfer work hydraulically between different thermal operated components or sub-systems. The system aims at providing trigeneration energy features for the residential sector, that is providing heating, cooling and electrical power for meeting the energy needs of domestic houses. This innovative system is made of two dithermal processes (working at two different levels of temperatures) and featuring two different working fluids. The first process is able to directly supply either electrical energy generated by an hydraulic turbine or drives the second process thanks to the incompressible fluid, which is similar to a heat pump effect for heating or cooling purposes. The innovative aspect of this process relies on the use of an hydraulic transfer fluid to transfer the work between each sub-system and therefore simplifying the conversion chain. A model, assuming steady-state operation, is developed to assess the energy performances of different variants of this thermo-hydraulic process with various heat source temperatures (80–110 °C) or heat sinks (0–30 °C), as well as various pairs of working fluids. For instance, in the frame of a single-family home, located in the Mediterranean region, the working

APA: U free Pu pin in a heterogeneous assembly to improve Pu loading in a PWR - neutronic, thermo-hydraulic and manufacturing studies

International Nuclear Information System (INIS)

Porta, J.; Puill, A.; Bauer, M.; Matheron, P.

1999-01-01

After having presented the specific context of France with respect to the fuel cycle and reprocessing, the problem of plutonium fuel utilization is posed. If one of the solutions, a pressurized water reactor (PWR) with an increased moderation ratio seems possible, it entails making excessive changes to the reactor, the control systems, and the general architecture of the steam supply system. Another solution consists in modifying the fuel itself so as to eliminate conversion on 238 U by using plutonium (Pu) in a neutronically inert matrix. However, the disadvantage of this type of fuel is that it has very low Doppler and draining coefficients and a very small delayed neutron fraction. To enable using these fuels, a heterogeneous assembly has to be defined, in which standard UO 2 rods provide the physical properties required to ensure acceptable safety coefficients. (author)

Biomass thermo-conversion. Research trends

International Nuclear Information System (INIS)

Rodriguez Machin, Lizet; Perez Bermudez, Raul; Quintana Perez, Candido Enrique; Ocanna Guevara, Victor Samuel; Duffus Scott, Alejandro

2011-01-01

In this paper is studied the state of the art in order to identify the main trends of the processes of thermo conversion of biomass into fuels and other chemicals. In Cuba, from total supply of biomass, wood is the 19% and sugar cane bagasse and straw the 80%, is why research in the country, should be directed primarily toward these. The methods for energy production from biomass can be group into two classes: thermo-chemical and biological conversion routes. The technology of thermo-chemical conversion includes three subclasses: pyrolysis, gasification, and direct liquefaction. Although pyrolysis is still under development, in the current energy scenario, has received special attention, because can convert directly biomass into solid, liquid and gaseous by thermal decomposition in absence of oxygen. The gasification of biomass is a thermal treatment, where great quantities of gaseous products and small quantities of char and ash are produced. In Cuba, studies of biomass thermo-conversion studies are limited to slow pyrolysis and gasification; but gas fuels, by biomass, are mainly obtained by digestion (biogas). (author)

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

International Nuclear Information System (INIS)

Nash, C.A.; Blake, J.E.; Rush, G.C.

1990-01-01

A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m 2 ) (1.1E+6 BTU/(ft 2 hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient

ECOSIM - Applied to a study on the thermo-hydraulic behaviour of feedwater heaters

International Nuclear Information System (INIS)

Huelamo Martinez, E.; Casado Flores, E.; Bosch Aparicio, F.

1998-01-01

In order to carry out a behaviour study on the secondary circuit of a nuclear power plant operating at a load level higher than originally planned, it is essential to know if the cycle heaters are valid from the thermo-dynamic point of view. This paper describes the models which were used for the study of certain heaters; these models were validated by checking that they faithfully reproduced the behaviour of the equipment (TTD and DCA) in areas where data from the manufacturer was available. The behaviour of said equipment was later obtained in the foreseen operating range. The calculations necessary for these studies were carried out by building ECOSIM models, taking into account that the behaviour of the feedwater heaters depends both on the entry conditions of the extraction steam and also on the remaining mass and energy inputs. For this reason the actual plant layout was taken into consideration, as it was different from the original design. This paper describes the starting hypothesis, the correlations used, the results obtained, an analysis of said results, and a comparison with the manufacturer's data where available. (Author)

Results of laboratory and in-situ measurements for the description of coupled thermo-hydro-mechanical processes in clays

Energy Technology Data Exchange (ETDEWEB)

Goebel, Ingeborg; Alheid, Hans-Joachim [BGR Hannover, Stilleweg 2, D-30655 Hannover (Germany); Jockwer, Norbert [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Theodor-Heuss-Str. 4, 38122 Braunschweig (Germany); Mayor, Juan Carlos [ENRESA, Emilio Vargas 7, E-Madrid (Spain); Garcia-Sineriz, Jose Luis [AITEMIN, c/ Alenza, 1 - 28003 Madrid (Spain); Alonso, Eduardo [International Center for Numerical Methods in Engineering, CIMNE, Edificio C-1, Campus Norte UPC, C/Gran Capitan, s/n, 08034 Barcelona (Spain); Weber, Hans Peter [NAGRA, Hardstrasse 73, CH-5430 Wettingen (Switzerland); Ploetze, Michael [ETHZ, Eidgenoessische Technische Hochschule Zuerich, ETH Zentrum, HG Raemistrasse 101, CH-8092 Zuerich (Switzerland); Klubertanz, Georg [COLENCO Power Engineering Ltd, CPE, Taefern Str. 26, 5405 Baden-Daettwil (Switzerland); Ammon, Christian [Rothpletz, Lienhard, Cie AG, Schifflaendestrasse 35, 5001 Aarau (Switzerland)

2004-07-01

The Heater Experiment at the Mont Terri Underground Laboratory aims at producing a validated model of thermo-hydro-mechanically (THM) coupled processes. The experiment consists of an engineered barrier system where in a vertical borehole, a heater is embedded in bentonite blocks, surrounded by the host rock, Opalinus Clay. The experimental programme comprises permanent monitoring before, during, and after the heating phase, complemented by geotechnical, hydraulic, and seismic in-situ measurements as well as laboratory analyses of mineralogical and rock mechanics properties. After the heating, the experiment was dismantled for further investigations. Major results of the experimental findings are outlined. (authors)

Thermo-hydrodynamic lubrication in hydrodynamic bearings

CERN Document Server

Bonneau, Dominique; Souchet, Dominique

2014-01-01

This Series provides the necessary elements to the development and validation of numerical prediction models for hydrodynamic bearings. This book describes the thermo-hydrodynamic and the thermo-elasto-hydrodynamic lubrication. The algorithms are methodically detailed and each section is thoroughly illustrated.

Thermo-stimulated current and dielectric loss in composite materials

International Nuclear Information System (INIS)

Nishijima, S.; Hagihara, T.; Okada, T.

1986-01-01

Thermo-stimulated current and dielectric loss measurements have been performed on five kinds of commercially available composite materials in order to study the electric properties of composite materials at low temperatures. Thermo-stimulated current measurements have been made on the composite materials in which the matrix quality was changed intentionally. The changes in the matrices were introduced by gamma irradiation or different curing conditions. Thermo-stimulated current and dielectric loss measurements revealed the number and the molecular weight of dipolar molecules. The different features of thermo-stimulated current and dielectric losses were determined for different composite materials. The gamma irradiation and the curing conditions especially affect the thermo-stimulated current features. The changes in macroscopic mechanical properties reflect those of thermo-stimulated current. It was found that the change in quality and/or degradation of the composite materials could be detected by means of thermo-stimulated current and/or dielectric loss measurements

THE LIQUID NITROGEN SYSTEM FOR CHAMBER A; A CHANGE FROM ORIGINAL FORCED FLOW DESIGN TO A NATURAL FLOW (THERMO SIPHON) SYSTEM

International Nuclear Information System (INIS)

Homan, J.; Montz, M.; Ganni, V.; Sidi-Yekhlef, A.; Knudsen, P.; Creel, J.; Arenius, D.; Garcia, S.

2010-01-01

NASA at the Johnson Space Center (JSC) in Houston is presently working toward modifying the original forced flow liquid nitrogen cooling system for the thermal shield in the space simulation chamber-A in Building 32 to work as a natural flow (thermo siphon) system. Chamber A is 19.8 m (65 ft) in diameter and 35.66 m (117 ft) high. The LN 2 shroud environment within the chamber is approximately 17.4 m (57 ft) in diameter and 28 m (92 ft) high. The new thermo siphon system will improve the reliability, stability of the system. Also it will reduce the operating temperature and the liquid nitrogen use to operate the system. This paper will present the requirements for the various operating modes. System level thermodynamic comparisons of the existing system to the various options studied and the final option selected will be outlined. A thermal and hydraulic analysis to validate the selected option for the conversion of the current forced flow to natural flow design will be discussed. The proposed modifications to existing system to convert to natural circulation (thermo siphon) system and the design features to help improve the operations, and maintenance of the system will be presented.

Coupled thermo-hydro-mechanical calculations of the water saturation phase of a KBS-3 deposition hole. Influence of hydraulic rock properties on the water saturation phase

International Nuclear Information System (INIS)

Boergesson, Lennart; Hernelind, J.

1999-12-01

The wetting process in deposition holes designed according to the KBS-3-concept has been simulated with finite element calculations of the thermo-hydro-mechanical processes in the buffer, backfill and surrounding rock. The buffer material has been modelled according to the preliminary material models developed for swelling clay. The properties of the rock have been varied in order to investigate the influence of the rock properties and the hydraulic conditions on the wetting processes. In the modelling of the test holes the permeability of the rock matrix, the water supply from the backfill, the water pressure in the surrounding rock, the permeability of the disturbed zone around the deposition hole, the water retention properties of the rock, and the transmissivity of two fractures intersecting the deposition hole have been varied. The calculations indicate that the wetting takes about 5 years if the water pressure in the rock is high and if the permeability of the rock is so high that the properties of the bentonite determine the wetting rate. However, it may take considerably more than 30 years if the rock is very tight and the water pressure in the rock is low. The calculations also show that the influence of the rock structure is rather large except for the influence of the transmissivity T of the fractures, which turned out to be insignificant for the values used in the calculations

Thermal-hydraulic effects of transition to improved System 80TM fuel

International Nuclear Information System (INIS)

Rodack, T.; Joffre, P.F.; Kapoor, R.K.

2004-01-01

ABB CE's improved System 80 TM PWR fuel design includes GUARDIAN debris-resistant features and laser-welded Zircaloy grids. The GUARDIAN features include an Inconel grid with debris-filtering features located just above the Lower End Fitting, and a solid fuel rod bottom end cap that extends above the filtering features. Tests and analyses were done to establish the impact of these design improvements on fuel assembly hydraulic performance. Further analysis was done to determine the mixed core thermal-hydraulic performance as the transition is made over two fuel cycles to a full core of the improved System 80 TM fuel. Results confirm that the Thermal-Hydraulic (T-H) effects of the reduction in hydraulic resistance between the improved and resident fuel due to the laser-welded Zircaloy grids offsets the effects of the increased resistance GUARDIAN grid. Therefore, the mechanically improved System 80 TM fuel can be implemented with no net impact on Departure from Nucleate Boiling (DNB) margin in transition cores. (author)

Thermo-electric pump

International Nuclear Information System (INIS)

Georges, J.-L.; Veyret, J.-F.

1973-01-01

Description is given of a thermo-pump for electrically conductive liquid fluids, e.g. for a liquid metal such as sodium. This pump is characterized in that the piping for the circulation of the conductive liquid is constituted by a plurality of conduits defined by two co-axial cylinders and two walls parallel to their axis. Each conduit limited outside by a magnet, inside by a mild-iron tube, and laterally by two materials forming a thermocouple. The electric current generated by that thermo-couple and the magnetic flux generated by the magnets both loop the loop through an outer cylindrical nickel shell. This can be applied to sodium circulation loops for testing nuclear fuel elements [fr

Neutronics and thermal hydraulics coupling scheme for design improvement of liquid metal fast systems

International Nuclear Information System (INIS)

Sanchez-Espinoza, V.H.; Jaeger, W.; Travleev, A.; Monti, L.; Doern, R.

2009-01-01

Many advanced reactor concepts are nowadays under investigations within the Generation IV international initiative as well as in European research programs including subcritical and critical fast reactor systems cooled by liquid metal, gas and supercritical water. The Institute of Neutron Physics and Reactor Technology (INR) at the Forschungszentrum Karlsruhe GmbH is involved in different European projects like IP EUROTRANS, ELSY, ESFR. The main goal of these projects is, among others, to assess the technical feasibility of proposed concepts regarding safety, economics and transmutation requirements. In view of increased computer capabilities, improved computational schemes, where the neutronic and the thermal hydraulic solution is iteratively coupled, become practicable. The codes ERANOS2.1 and TRACE are being coupled to analyze fuel assembly or core designs of lead-cooled fast reactors (LFR). The neutronic solution obtained with the coupled system for a LFR fuel assembly was compared with the MCNP5 solution. It was shown that the coupled system is predicting physically sound results. The iterative coupling scheme was realized using Perlscripts and auxiliary Fortran programs to ensure that the mapping between the neutronic and the thermal hydraulic part is consistent. The coupled scheme is very flexible and appropriate for the neutron physical and thermal hydraulic investigation of fuel assemblies and of cores of lead cooled fast reactors. The developed methods and the obtained results will be presented and discussed. (author)

Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core

International Nuclear Information System (INIS)

Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park

2000-01-01

This paper summarizes the methodology for thermal hydraulic conceptual design and performance analysis which is used for KALIMER core, especially the preliminary methodology for flow grouping and peak pin temperature calculation in detail. And the major technical results of the conceptual design for the KALIMER 98.03 core was shown and compared with those of KALIMER 97.07 design core. The KALIMER 98.03 design core is proved to be more optimized compared to the 97.07 design core. The number of flow groups are reduced from 16 to 11, and the equalized peak cladding midwall temperature from 654 deg. C to 628 deg. C. It was achieved from the nuclear and thermal hydraulic design optimization study, i.e. core power flattening and increase of radial blanket power fraction. Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in core thermal hydraulic analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each bundle, thus pin cladding damage accumulation and pin reliability. The flow grouping and the peak pin temperature calculation for the preliminary conceptual design is performed with the modules ORFCE-F60 and ORFCE-T60 respectively. The basic subchannel analysis will be performed with the SLTHEN code, and the detailed subchannel analysis will be done with the MATRA-LMR code which is under development for the K-Core system. This methodology was proved practical to KALIMER core thermal hydraulic design from the related benchmark calculation studies, and it is used to KALIMER core thermal hydraulic conceptual design. (author)

Near-field NanoThermoMechanical memory

International Nuclear Information System (INIS)

Elzouka, Mahmoud; Ndao, Sidy

2014-01-01

In this letter, we introduce the concept of NanoThermoMechanical Memory. Unlike electronic memory, a NanoThermoMechanical memory device uses heat instead of electricity to record, store, and recover data. Memory function is achieved through the coupling of near-field thermal radiation and thermal expansion resulting in negative differential thermal resistance and thermal latching. Here, we demonstrate theoretically via numerical modeling the concept of near-field thermal radiation enabled negative differential thermal resistance that achieves bistable states. Design and implementation of a practical silicon based NanoThermoMechanical memory device are proposed along with a study of its dynamic response under write/read cycles. With more than 50% of the world's energy losses being in the form of heat along with the ever increasing need to develop computer technologies which can operate in harsh environments (e.g., very high temperatures), NanoThermoMechanical memory and logic devices may hold the answer

Thermo-msf-parser: an open source Java library to parse and visualize Thermo Proteome Discoverer msf files.

Science.gov (United States)

Colaert, Niklaas; Barsnes, Harald; Vaudel, Marc; Helsens, Kenny; Timmerman, Evy; Sickmann, Albert; Gevaert, Kris; Martens, Lennart

2011-08-05

The Thermo Proteome Discoverer program integrates both peptide identification and quantification into a single workflow for peptide-centric proteomics. Furthermore, its close integration with Thermo mass spectrometers has made it increasingly popular in the field. Here, we present a Java library to parse the msf files that constitute the output of Proteome Discoverer. The parser is also implemented as a graphical user interface allowing convenient access to the information found in the msf files, and in Rover, a program to analyze and validate quantitative proteomics information. All code, binaries, and documentation is freely available at http://thermo-msf-parser.googlecode.com.

Development of hydraulic analysis code for optimizing thermo-chemical is process reactors

International Nuclear Information System (INIS)

Terada, Atsuhiko; Hino, Ryutaro; Hirayama, Toshio; Nakajima, Norihiro; Sugiyama, Hitoshi

2007-01-01

The Japan Atomic Energy Agency has been conducting study on thermochemical IS process for water splitting hydrogen production. Based on the test results and know-how obtained through the bench-scale test, a pilot test plant, which has a hydrogen production performance of 30 Nm 3 /h, is being designed conceptually as the next step of the IS process development. In design of the IS pilot plant, it is important to make chemical reactors compact with high performance from the viewpoint of plant cost reduction. A new hydraulic analytical code has been developed for optimizing mixing performance of multi-phase flow involving chemical reactions especially in the Bunsen reactor. Complex flow pattern with gas-liquid chemical interaction involving flow instability will be characterized in the Bunsen reactor. Preliminary analytical results obtained with above mentioned code, especially flow patterns induced by swirling flow agreed well with that measured by water experiments, which showed vortex breakdown pattern in a simplified Bunsen reactor. (author)

Hydraulic Hybrid Parcel Delivery Truck Deployment, Testing & Demonstration

Energy Technology Data Exchange (ETDEWEB)

Gallo, Jean-Baptiste [Calstart Incorporated, Pasadena, CA (United States)

2014-03-07

Although hydraulic hybrid systems have shown promise over the last few years, commercial deployment of these systems has primarily been limited to Class 8 refuse trucks. In 2005, the Hybrid Truck Users Forum initiated the Parcel Delivery Working Group including the largest parcel delivery fleets in North America. The goal of the working group was to evaluate and accelerate commercialization of hydraulic hybrid technology for parcel delivery vehicles. FedEx Ground, Purolator and United Parcel Service (UPS) took delivery of the world’s first commercially available hydraulic hybrid parcel delivery trucks in early 2012. The vehicle chassis includes a Parker Hannifin hydraulic hybrid drive system, integrated and assembled by Freightliner Custom Chassis Corp., with a body installed by Morgan Olson. With funding from the U.S. Department of Energy, CALSTART and its project partners assessed the performance, reliability, maintainability and fleet acceptance of three pre-production Class 6 hydraulic hybrid parcel delivery vehicles using information and data from in-use data collection and on-road testing. This document reports on the deployment of these vehicles operated by FedEx Ground, Purolator and UPS. The results presented provide a comprehensive overview of the performance of commercial hydraulic hybrid vehicles in parcel delivery applications. This project also informs fleets and manufacturers on the overall performance of hydraulic hybrid vehicles, provides insights on how the technology can be both improved and more effectively used. The key findings and recommendations of this project fall into four major categories: -Performance, -Fleet deployment, -Maintenance, -Business case. Hydraulic hybrid technology is relatively new to the market, as commercial vehicles have been introduced only in the past few years in refuse and parcel delivery applications. Successful demonstration could pave the way for additional purchases of hydraulic hybrid vehicles throughout the

RAP-3A Computer code for thermal and hydraulic calculations in steady state conditions for fuel element clusters

International Nuclear Information System (INIS)

Popescu, C.; Biro, L.; Iftode, I.; Turcu, I.

1975-10-01

The RAP-3A computer code is designed for calculating the main steady state thermo-hydraulic parameters of multirod fuel clusters with liquid metal cooling. The programme provides a double accuracy computation of temperatures and axial enthalpy distributions of pressure losses and axial heat flux distributions in fuel clusters before boiling conditions occur. Physical and mathematical models as well as a sample problem are presented. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer

Standardization of the methodology used for fuel pressure drop evaluation to improve hydraulic calculation of heterogeneous cores

International Nuclear Information System (INIS)

Le Borgne, E.; Mattei, A.

1994-01-01

Continuous searching for safer and more efficient fuel, and diversification of fuel supply have as a consequence a possible change in the characteristics of the fuel assemblies used in nuclear reactors. By partially refueling cores with new assemblies, nuclear power plant operators are confronted with the problem of heterogeneous cores. The complexity of the problem increases as products diversify in isotopic concentration, types of alloy, size and shape of structure components. This document will focus strictly on the differences in hydraulic resistance related to the modifications in grid structures having no effect on DNB correlations. Although this is an extremely simplified approach to the problem, establishing data to evaluate the hydraulic compatibility between two different assemblies can be difficult, and if not controlled closely, can lead to false conclusions that may affect the operation and safety of the reactor. (authors). 2 figs

Methodology for local verification of flow regimes in fuel assemblies charts

International Nuclear Information System (INIS)

Igor, Sharaevsky; Elena, Sharaevskaya; Domashev, E.D.; Alexander, Arkhypov; Vladimir, Kolochko

2003-01-01

The best estimate thermal hydraulic codes describe adequately two-phase flows in nuclear energy facilities if there is proper system of closed relations. It could be obtained from the reliable information on structure forms of two-phase flows, its boundaries and reliable regime charts. In the paper the methodology of automatic recognition of the boundaries of the main types of two phase flows for rod fuel assemblies is presented. The methodology is based on definition of thermal hydraulic parameters distribution in experimental fuel assembly. The measurements were carried out using ASD signals of acoustic noise. In the paper data on two-phase flow regimes boundaries recognition especially low boundaries of bubble flow are summarized for experimental fuel assembly. The methodology of flow regimes charts applied to recognition of upper boundaries of boiling crisis regime was verificated. The satisfactory coincidence with experimental results have been shown. (author)

stepping motor - hydraulic motor servo drives for an nc milling machine

African Journals Online (AJOL)

Dr Obe

stepping motor Drive Assembly especially Designed for CNC systems". 13th Machine Tool Design and. Research. (MTDR) conference,. University of Birmingham, 1972. 2 Ertongur, N.A. "Investigation into the instability in an electro hydraulic control system for machine tools" Ph.D. Thesis, University of. Birmingham, UK. 1966 ...

An integrated approach to develop, validate and operate thermo-physiological human simulator for the development of protective clothing.

Science.gov (United States)

Psikuta, Agnes; Koelblen, Barbara; Mert, Emel; Fontana, Piero; Annaheim, Simon

2017-12-07

Following the growing interest in the further development of manikins to simulate human thermal behaviour more adequately, thermo-physiological human simulators have been developed by coupling a thermal sweating manikin with a thermo-physiology model. Despite their availability and obvious advantages, the number of studies involving these devices is only marginal, which plausibly results from the high complexity of the development and evaluation process and need of multi-disciplinary expertise. The aim of this paper is to present an integrated approach to develop, validate and operate such devices including technical challenges and limitations of thermo-physiological human simulators, their application and measurement protocol, strategy for setting test scenarios, and the comparison to standard methods and human studies including details which have not been published so far. A physical manikin controlled by a human thermoregulation model overcame the limitations of mathematical clothing models and provided a complementary method to investigate thermal interactions between the human body, protective clothing, and its environment. The opportunities of these devices include not only realistic assessment of protective clothing assemblies and equipment but also potential application in many research fields ranging from biometeorology, automotive industry, environmental engineering, and urban climate to clinical and safety applications.

Pressure drop performance evaluation for test assemblies with the newly developed top and bottom nozzles

International Nuclear Information System (INIS)

Lee, S. K.; Park, N. K.; Su, J. M.; Kim, H. K.; Lee, J. N.; Kim, K. T.

2003-01-01

To perform the hydraulic test for the newly developed top and bottom nozzles, two kinds of test assemblies were manufactured i. e. one is the test assembly which has the newly developed top and bottom nozzles and the other is Guardian test assembly which is commercially in mass production now. The test results show that the test assembly with one top nozzle and two bottom nozzles has a greater pressure loss coefficient than Guardian test assembly by 60.9% and 90.4% at the bottom nozzle location. This cause is due to the debris filtering plate for bottom nozzle to improve a filtering efficiency aginst foreign material. In the region of mid grid and top nozzle, there is no difference in pressure loss coefficient between the test assemblies since the componet features in these regions are very similar or same each other. The loss coefficients are 14.2% and 21.9% for model A and B respectively in the scale of test assembly, and the value would be within the 10% in the scale of real fuel assembly. As a result of hydraulic performance evaluation, model A is superior to model B

The TRPM2 channel: A thermo-sensitive metabolic sensor.

Science.gov (United States)

Kashio, Makiko; Tominaga, Makoto

2017-09-03

Living organisms continually experience changes in ambient temperature. To detect such temperature changes for adaptive behavioral responses, we evolved the ability to sense temperature. Thermosensitive transient receptor potential (TRP) channels, so-called thermo-TRPs, are involved in many physiologic functions in diverse organisms and constitute important temperature sensors. One of the important roles of thermo-TRPs is detecting ambient temperature in sensory neurons. Importantly, the functional expression of thermo-TRPs is observed not only in sensory neurons but also in tissues and cells that are not exposed to drastic temperature changes, indicating that thermo-TRPs are involved in many physiologic functions within the body's normal temperature range. Among such thermo-TRPs, this review focuses on one thermo-sensitive metabolic sensor in particular, TRPM2, and summarizes recent progress to clarify the regulatory mechanisms and physiologic functions of TRPM2 at body temperature under various metabolic states.

Thermo-hydrodynamical modelling of a flooded deep mine reservoir - Case of the Lorraine Coal Basin

International Nuclear Information System (INIS)

Reichart, Guillaume

2015-01-01

Since 2006, cessation of dewatering in Lorraine Coal Basin (France) led to the flooding of abandoned mines, resulting in a new hydrodynamic balance in the area. Recent researches concerning geothermal exploitation of flooded reservoirs raised new questions, which we propose to answer. Our work aimed to understand the thermos-hydrodynamic behaviour of mine water in a flooding or flooded system. Firstly, we synthesized the geographical, geological and hydrogeological contexts of the Lorraine Coal Basin, and we chose a specific area for our studies. Secondly, temperature and electric conductivity log profiles were measured in old pits of the Lorraine Coal Basin, giving a better understanding of the water behaviour at a deep mine shaft scale. We were able to build a thermos-hydrodynamic model and simulate water behaviour at this scale. Flow regime stability is also studied. Thirdly, a hydrodynamic spatialized meshed model was realized to study the hydrodynamic behaviour of a mine reservoir as a whole. Observed water-table rise was correctly reproduced: moreover, the model can be used in a predictive way after the flooding. Several tools were tested, improved or developed to ease the study of flooded reservoirs, as three-dimensional up-scaling of hydraulic conductivities and a coupled spatialized meshed model with a pipe network. (author) [fr

Coupled thermo-hydro-mechanical experiment at Kamaishi mine. Technical note 15-99-02. Experimental results

Energy Technology Data Exchange (ETDEWEB)

Chijimatsu, Masakazu; Sugita, Yutaka; Fujita, Tomoo [Tokai Works, Waste Management and Fuel Cycle Research Center, Waste Isolation Research Division, Barrier Performance Group, Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan); Amemiya, Kiyoshi [Hazama Corp., Tokyo (Japan)

1999-07-01

It is an important part of the near field performance assessment of nuclear waste disposal to evaluate coupled thermo-hydro-mechanical (T-H-M) phenomena, e.g., thermal effects on groundwater flow through rock matrix and water seepage into the buffer material, the generation of swelling pressure of the buffer material, and thermal stresses potentially affecting porosity and fracture apertures of the rock. An in-situ T-H-M experiment named Engineered Barrier Experiment' has been conducted at the Kamaishi Mine, of which host rock is granodiorite, in order to establish conceptual models of the coupled T-H-M processes and to build confidence in mathematical and computer codes. In 1995, fourteen boreholes were excavated in order to install the various sensors. After the hydraulic tests, mechanical tests were carried out to obtain the rock properties. After that, a test pit, 1.7 m in diameter and 5.0 m in depth, was excavated. During the excavation, the change of pore pressure, displacement and temperature of rock mass were measured. In 1996, the buffer material and heater were set up in the test pit, and then coupled thermo-hydro-mechanical test was started. The duration of heating phase was 250 days and that of cooling phase was 180 days. The heater surface was controlled to be 100degC during heating phase. Measurement was carried out by a number of sensors installed in both buffer and rock mass during the test. The field experiment leads to a better understanding of the behavior of the coupled thermo-hydro-mechanical phenomena in the near field. (author)

The time-dependent 3D discrete ordinates code TORT-TD with thermal-hydraulic feedback by ATHLET models

International Nuclear Information System (INIS)

Seubert, A.; Velkov, K.; Langenbuch, S.

2008-01-01

This paper describes the time-dependent 3D discrete ordinates transport code TORT-TD. Thermal-hydraulic feedback is considered by coupling TORT-TD with the thermal-hydraulics system code ATHLET. The coupled code TORT-TD/ATHLET 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. The nuclear cross sections are interpolated between pre-calculated table values of fuel temperature, moderator density and boron concentration. For verification of the implementation, selected test cases have been calculated by TORT-TD/ATHLET. They include a control rod ejection transient in a small PWR fuel assembly arrangement and a local boron concentration change in a single PWR fuel assembly. In the latter, special attention has been paid to study the influence of the thermal-hydraulic feedback modelling in ATHLET. The results obtained for a control rod ejection accident in a PWR quarter core demonstrate the applicability of TORT-TD/ATHLET. (authors)

Thermo-mechanical ratcheting in jointed rock masses

KAUST Repository

Pasten, C.

2015-09-01

Thermo-mechanical coupling takes place in jointed rock masses subjected to large thermal oscillations. Examples range from exposed surfaces under daily and seasonal thermal fluctuations to subsurface rock masses affected by engineered systems such as geothermal operations. Experimental, numerical and analytical results show that thermo-mechanical coupling can lead to wedging and ratcheting mechanisms that result in deformation accumulation when the rock mass is subjected to a biased static-force condition. Analytical and numerical models help in identifying the parameter domain where thermo-mechanical ratcheting can take place.

Thermo-mechanical ratcheting in jointed rock masses

KAUST Repository

Pasten, C.; Garcí a, M.; Santamarina, Carlos

2015-01-01

Thermo-mechanical coupling takes place in jointed rock masses subjected to large thermal oscillations. Examples range from exposed surfaces under daily and seasonal thermal fluctuations to subsurface rock masses affected by engineered systems such as geothermal operations. Experimental, numerical and analytical results show that thermo-mechanical coupling can lead to wedging and ratcheting mechanisms that result in deformation accumulation when the rock mass is subjected to a biased static-force condition. Analytical and numerical models help in identifying the parameter domain where thermo-mechanical ratcheting can take place.

Novel thermo-sensitive core-shell nanoparticles for targeted paclitaxel delivery

International Nuclear Information System (INIS)

Li Yuanpei; Pan Shirong; Zhang Wei; Du Zhuo

2009-01-01

Novel thermo-sensitive nanoparticles self-assembled from poly(N,N-diethylacrylamide- co-acrylamide)-block-poly(γ-benzyl L-glutamate) were designed for targeted drug delivery in localized hyperthermia. The lower critical solution temperature (LCST) of nanoparticles was adjusted to a level between physiological body temperature (37 deg. C) and that used in local hyperthermia (about 43 deg. C). The temperature-dependent performances of the core-shell nanoparticles were systemically studied by nuclear magnetic resonance (NMR), circular dichroism (CD), fluorescence spectroscopy, dynamic light scattering (DLS), and atom force microscopy (AFM). The mean diameter of the nanoparticles increased slightly from 110 to 129 nm when paclitaxel (PTX), a poorly water-soluble anti-tumor drug, was encapsulated. A stability study in bovine serum albumin (BSA) solution indicated that the PTX loaded nanoparticles may have a long circulation time under physiological environments as the LCST was above physiological body temperature and the shell remained hydrophilic at 37 deg.C. The PTX release profiles showed thermo-sensitive controlled behavior. The proliferation inhibiting activity of PTX loaded nanoparticles was evaluated against Hela cells in vitro, compared with Taxol (a formulation of paclitaxel dissolved in Cremophor EL and ethanol). The cytotoxicity of PTX loaded nanoparticles increased obviously when hyperthermia was performed. The nanoparticles synthesized here could be an ideal candidate for thermal triggered anti-tumor PTX delivery system.

On nonlinear thermo-electro-elasticity.

Science.gov (United States)

Mehnert, Markus; Hossain, Mokarram; Steinmann, Paul

2016-06-01

Electro-active polymers (EAPs) for large actuations are nowadays well-known and promising candidates for producing sensors, actuators and generators. In general, polymeric materials are sensitive to differential temperature histories. During experimental characterizations of EAPs under electro-mechanically coupled loads, it is difficult to maintain constant temperature not only because of an external differential temperature history but also because of the changes in internal temperature caused by the application of high electric loads. In this contribution, a thermo-electro-mechanically coupled constitutive framework is proposed based on the total energy approach. Departing from relevant laws of thermodynamics, thermodynamically consistent constitutive equations are formulated. To demonstrate the performance of the proposed thermo-electro-mechanically coupled framework, a frequently used non-homogeneous boundary-value problem, i.e. the extension and inflation of a cylindrical tube, is solved analytically. The results illustrate the influence of various thermo-electro-mechanical couplings.

Soil hydraulic properties of topsoil along two elevation transects affected by soil erosion

Czech Academy of Sciences Publication Activity Database

Nikodem, A.; Kodešová, R.; Jakšík, O.; Jirků, V.; Fér, M.; Klement, A.; Žigová, Anna

2013-01-01

Roč. 15, - (2013) ISSN 1607-7962. [EGU General Assembly /10./. 07.04.2013-12.04.2013, Vienna] Institutional support: RVO:67985831 Keywords : topsoil * hydraulic properties * erosion processes Subject RIV: DF - Soil Science http://meetingorganizer.copernicus.org/EGU2013/EGU2013-7924.pdf

Education for hydraulics and pneumatics in Nihon University; Nihon Daigaku ni okeru yukuatsu kyoiku

Energy Technology Data Exchange (ETDEWEB)

Ouchi, M. [Nihon Univ., Chiba (Japan). Coll. of Industrial Technology

2000-03-15

Described herein is education of hydraulics and pneumatics in Nihon University. Department of Mechanical Engineering of Faculty of Production Engineering has been holding up the educational aims of bringing up engineers and researchers who have ability and intelligence to cope with internationalization and contribute to society, and of bringing about creativity, among others. Control equipment is an optional subject for the sophomore class in the second semester, and is centered by mechatronics, including hydraulic and pneumatic control systems and equipment. The related subjects include fluid dynamics, control engineering, system controlling, hydraulic machines, robotics and automobile engineering. The drill course includes disassembling and assembling gear pumps, drills on pneumatic devices, system behavior and mechatronics, experiments on fan and hydraulic control circuits and on servo mechanisms, and machinery designs and drawings. Seminars are led by full-time or part-time lecturers for the themes related to hydraulic power. Many students are interested in hydraulic and pneumatic themes for their graduation theses, because of their relations with control, environments, energy saving and so on. We are now in the age of composite technologies, and hydraulic power basics are prerequisite for engineers, and important for education of students. (NEDO)

Impact of land management on soil structure and soil hydraulic properties

Czech Academy of Sciences Publication Activity Database

Kodešová, R.; Jirků, V.; Nikodem, A.; Mühlhanselová, M.; Žigová, Anna

2010-01-01

Roč. 12, - (2010) ISSN 1029-7006. [European Geosciences Union General Assembly 2010. 02.05.2010-07.05.2010, Wienna] R&D Projects: GA ČR GA526/08/0434 Institutional research plan: CEZ:AV0Z30130516 Keywords : land management * soil structure * soil hydraulic properties * micromorphology Subject RIV: DF - Soil Science

Spent fuel pool thermal-hydraulic analysis using RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Ramos, M. C.; Fernandes, G.H.N.; Costa, A.L.; Pereira, F.; Pereira, C., E-mail: marc5663@gmail.com, E-mail: ghnfernandes@pq.cnpq.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

In order to analyze the thermo-hydraulic behavior of spent fuel pools, and taking as reference a hypothetic PWR nuclear plant, a model of RELAP-3D for a spent fuel pool has been built. This model has been used to simulate a loss of coolant in SPF. This study focuses on the loss of coolant flow accident in spent fuel storage pool which is modelled by using RELAP5-3D code to observe the coolant level reduction and fuel uncovery because of decay heat generation of the spent fuel in the pool. The results have been compared with the available data. The developed model demonstrated that the RELAP5-3D is capable of reproduce the thermal behavior of SPF in a transient scenario. (author)

Hydraulically centered control rod

International Nuclear Information System (INIS)

Horlacher, W.R.; Sampson, W.T.; Schukei, G.E.

1981-01-01

A control rod suspended to reciprocate in a guide tube of a nuclear fuel assembly has a hydraulic bearing formed at its lower tip. The bearing includes a plurality of discrete pockets on its outer surface into which a flow of liquid is continuously provided. In one embodiment the flow is induced by the pressure head in a downward facing chamber at the end of the bearing. In another embodiment the flow originates outside the guide tube. In both embodiments the flow into the pockets produces pressure differences across the bearing which counteract forces tending to drive the rod against the guide tube wall. Thus contact of the rod against the guide tube is avoided

Steady state thermal-hydraulic analyses of the MITICA cooling circuits

Energy Technology Data Exchange (ETDEWEB)

Zaupa, M., E-mail: matteo.zaupa@igi.cnr.it [Università degli Studi di Padova, Via 8 Febbraio 2, Padova 35122 (Italy); Consorzio RFX, Corso Stati Uniti 4, Padova 35127 (Italy); Sartori, E.; Dalla Palma, M.; Fellin, F.; Marcuzzi, D.; Pavei, M.; Rizzolo, A. [Consorzio RFX, Corso Stati Uniti 4, Padova 35127 (Italy)

2016-02-15

Megavolt ITER Injector Concept Advancement is the full scale prototype of the heating and current drive neutral beam injectors for ITER, to be built at Consorzio RFX (Padova). The engineering design of its components is challenging: the total heat loads they will be subjected to (expected between 2 and 19 MW), the high heat fluxes (up to 20 MW/m{sup 2}), and the beam pulse duration up to 1 h, set demanding requirements for reliable active cooling circuits. In support of the design, the thermo-hydraulic behavior of each cooling circuit under steady state condition has been investigated by using one-dimensional models. The final results, obtained considering a number of optimizations for the cooling circuits, show that all the requirements in terms of flow rate, temperature, and pressure drop are properly fulfilled.

Enhanced pathway efficiency of Saccharomyces cerevisiae by introducing thermo-tolerant devices.

Science.gov (United States)

Liu, Yueqin; Zhang, Genli; Sun, Huan; Sun, Xiangying; Jiang, Nisi; Rasool, Aamir; Lin, Zhanglin; Li, Chun

2014-10-01

In this study, thermo-tolerant devices consisting of heat shock genes from thermophiles were designed and introduced into Saccharomyces cerevisiae for improving its thermo-tolerance. Among ten engineered thermo-tolerant yeasts, T.te-TTE2469, T.te-GroS2 and T.te-IbpA displayed over 25% increased cell density and 1.5-4-fold cell viability compared with the control. Physiological characteristics of thermo-tolerant strains revealed that better cell wall integrity, higher trehalose content and enhanced metabolic energy were preserved by thermo-tolerant devices. Engineered thermo-tolerant strain was used to investigate the impact of thermo-tolerant device on pathway efficiency by introducing β-amyrin synthesis pathway, showed 28.1% increased β-amyrin titer, 28-35°C broadened growth temperature range and 72h shortened fermentation period. The results indicated that implanting heat shock proteins from thermophiles to S. cerevisiae would be an efficient approach to improve its thermo-tolerance. Copyright © 2014 Elsevier Ltd. All rights reserved.

Use of sensitivity-information for the adaptive simulation of thermo-hydraulic system codes

International Nuclear Information System (INIS)

Kerner, Alexander M.

2011-01-01

Within the scope of this thesis the development of methods for online-adaptation of dynamical plant simulations of a thermal-hydraulic system code to measurement data is depicted. The described approaches are mainly based on the use of sensitivity-information in different areas: statistical sensitivity measures are used for the identification of the parameters to be adapted and online-sensitivities for the parameter adjustment itself. For the parameter adjustment the method of a ''system-adapted heuristic adaptation with partial separation'' (SAHAT) was developed, which combines certain variants of parameter estimation and control with supporting procedures to solve the basic problems. The applicability of the methods is shown by adaptive simulations of a PKL-III experiment and by selected transients in a nuclear power plant. Finally the main perspectives for the application of a tracking simulator on a system code are identified.

Long term thermo-hydro-mechanical interaction behavior study of the saturated, discontinuous granitic rock mass around the radwaste repository using a steady state flow algorithm

Energy Technology Data Exchange (ETDEWEB)

Kim, Jhin Wung; Bae, Dae Suk; Kang, Chul Hyung; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-02-01

The objective of the present study is to understand the long term (500 years) thermo-hydro-mechanical interaction behavior of the 500 m depth underground radwaste repository in the saturated, discontinuous granitic rock mass using a steady state flow algorithm. The numerical model includes a saturated granitic rock mass with joints around the repository and a 45 .deg. C fault passing through the tunnel roof-wall intersection, and a canister with PWR spent fuels surrounded by the compacted bentonite and mixed-bentonite. Barton-Bandis joint constitutive model from the UDEC code is used for the joints. For the hydraulic analysis, a steady state flow algorithm is used for the groundwater flow through the rock joints. For the thermal analysis, heat transfer is modeled as isotropic conduction and heat decays exponentially with time. The results show that the variations of the hydraulic aperture, hydraulic conductivity, normal stress, normal displacements, and shear displacements of the joints are high in the vicinity of the repository and stay fairly constant on the region away from the repository. 14 refs., 15 figs., 11 tabs. (Author)

Theory and modeling of cylindrical thermo-acoustic transduction

Energy Technology Data Exchange (ETDEWEB)

Tong, Lihong, E-mail: lhtong@ecjtu.edu.cn [School of Civil Engineering and Architecture, East China Jiaotong University, Nanchang, Jiangxi (China); Lim, C.W. [Department of Architecture and Civil Engineering, City University of Hong Kong, Kowloon, Hong Kong SAR (China); Zhao, Xiushao; Geng, Daxing [School of Civil Engineering and Architecture, East China Jiaotong University, Nanchang, Jiangxi (China)

2016-06-03

Models both for solid and thinfilm-solid cylindrical thermo-acoustic transductions are proposed and the corresponding acoustic pressure solutions are obtained. The acoustic pressure for an individual carbon nanotube (CNT) as a function of input power is investigated analytically and it is verified by comparing with the published experimental data. Further numerical analysis on the acoustic pressure response and characteristics for varying input frequency and distance are also examined both for solid and thinfilm-solid cylindrical thermo-acoustic transductions. Through detailed theoretical and numerical studies on the acoustic pressure solution for thinfilm-solid cylindrical transduction, it is concluded that a solid with smaller thermal conductivity favors to improve the acoustic performance. In general, the proposed models are applicable to a variety of cylindrical thermo-acoustic devices performing in different gaseous media. - Highlights: • Theory and modeling both for solid and thinfilm-solid cylindrical thermo-acoustic transductions are proposed. • The modeling is verified by comparing with the published experimental data. • Acoustic response characteristics of cylindrical thermo-acoustic transductions are predicted by the proposed model.

The InterFrost benchmark of Thermo-Hydraulic codes for cold regions hydrology - first inter-comparison results

Science.gov (United States)

Grenier, Christophe; Roux, Nicolas; Anbergen, Hauke; Collier, Nathaniel; Costard, Francois; Ferrry, Michel; Frampton, Andrew; Frederick, Jennifer; Holmen, Johan; Jost, Anne; Kokh, Samuel; Kurylyk, Barret; McKenzie, Jeffrey; Molson, John; Orgogozo, Laurent; Rivière, Agnès; Rühaak, Wolfram; Selroos, Jan-Olof; Therrien, René; Vidstrand, Patrik

2015-04-01

The impacts of climate change in boreal regions has received considerable attention recently due to the warming trends that have been experienced in recent decades and are expected to intensify in the future. Large portions of these regions, corresponding to permafrost areas, are covered by water bodies (lakes, rivers) that interact with the surrounding permafrost. For example, the thermal state of the surrounding soil influences the energy and water budget of the surface water bodies. Also, these water bodies generate taliks (unfrozen zones below) that disturb the thermal regimes of permafrost and may play a key role in the context of climate change. Recent field studies and modeling exercises indicate that a fully coupled 2D or 3D Thermo-Hydraulic (TH) approach is required to understand and model the past and future evolution of landscapes, rivers, lakes and associated groundwater systems in a changing climate. However, there is presently a paucity of 3D numerical studies of permafrost thaw and associated hydrological changes, and the lack of study can be partly attributed to the difficulty in verifying multi-dimensional results produced by numerical models. Numerical approaches can only be validated against analytical solutions for a purely thermic 1D equation with phase change (e.g. Neumann, Lunardini). When it comes to the coupled TH system (coupling two highly non-linear equations), the only possible approach is to compare the results from different codes to provided test cases and/or to have controlled experiments for validation. Such inter-code comparisons can propel discussions to try to improve code performances. A benchmark exercise was initialized in 2014 with a kick-off meeting in Paris in November. Participants from USA, Canada, Germany, Sweden and France convened, representing altogether 13 simulation codes. The benchmark exercises consist of several test cases inspired by existing literature (e.g. McKenzie et al., 2007) as well as new ones. They

Mechanical seal assembly

Science.gov (United States)

Kotlyar, Oleg M.

2001-01-01

An improved mechanical seal assembly is provided for sealing rotating shafts with respect to their shaft housings, wherein the rotating shafts are subject to substantial axial vibrations. The mechanical seal assembly generally includes a rotating sealing ring fixed to the shaft, a non-rotating sealing ring adjacent to and in close contact with the rotating sealing ring for forming an annular seal about the shaft, and a mechanical diode element that applies a biasing force to the non-rotating sealing ring by means of hemispherical joint. The alignment of the mechanical diode with respect to the sealing rings is maintained by a series of linear bearings positioned axially along a desired length of the mechanical diode. Alternative embodiments include mechanical or hydraulic amplification components for amplifying axial displacement of the non-rotating sealing ring and transferring it to the mechanical diode.

Mechanical Seal Assembly

Energy Technology Data Exchange (ETDEWEB)

Kotlyar, Oleg M.

1999-06-18

An improved mechanical seal assembly is provided for sealing rotating shafts with respect to their shaft housings, wherein the rotating shafts are subject to substantial axial vibrations. The mechanical seal assembly generally includes a rotating sealing ring fixed to the shaft, a non-rotating sealing ring adjacent to and in close contact with the rotating sealing ring for forming an annular seal about the shaft, and a mechanical diode element that applies a biasing force to the non-rotating sealing ring by means of hemispherical joint. The alignment of the mechanical diode with respect to the sealing rings is maintained by a series of linear bearings positioned axially along a desired length of the mechanical diode. Alternative embodiments include mechanical or hydraulic amplification components for amplifying axial displacement of the non-rotating sealing ring and transferring it to the mechanical diode.

Qualification test for ITER HCCR-TBS mockups with high heat flux test facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Suk-Kwon, E-mail: skkim93@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Seong Dae; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2016-11-01

Highlights: • The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated. • A thermo-hydraulic analysis was performed using a high heat flux test facility by using electron beam. • The plan for qualification tests was developed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization. - Abstract: The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated, and an integrity and thermo-hydraulic performance test should be completed under the same or similar operation conditions of ITER. The test plan for a thermo-hydraulic analysis was developed by using a high heat flux test facility, called the Korean heat load test facility by using electron beam (KoHLT-EB). This facility is utilized for a qualification test of the plasma facing component (PFC) for the ITER first wall and DEMO divertor, and for the thermo-hydraulic experiments. In this work, KoHLT-EB will be used for the plan of the performance qualification test of the ITER HCCR-TBS mockups. This qualification tests should be performed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization (IO), which describe the specifications and qualifications of the heat flux test facility and test procedure for ITER PFC.

A Mathematical Model of the Thermo-Anemometric Flowmeter.

Science.gov (United States)

Korobiichuk, Igor; Bezvesilna, Olena; Ilchenko, Andriі; Shadura, Valentina; Nowicki, Michał; Szewczyk, Roman

2015-09-11

A thermo-anemometric flowmeter design and the principles of its work are presented in the article. A mathematical model of the temperature field in a stream of biofuel is proposed. This model allows one to determine the fuel consumption with high accuracy. Numerical modeling of the heater heat balance in the fuel flow of a thermo-anemometric flowmeter is conducted and the results are analyzed. Methods for increasing the measurement speed and accuracy of a thermo-anemometric flowmeter are proposed.

The Thermos process heat reactor

International Nuclear Information System (INIS)

Lerouge, Bernard

1979-01-01

The THERMOS process heat reactor was born from the following idea: the hot water energy vector is widely used for heating purposes in cities, so why not save on traditional fossil fuels by simply substituting a nuclear boiler of comparable power for the classical boiler installed in the same place. The French Atomic Energy Commission has techniques for heating in the big French cities which provide better guarantees for national independence and for the environment. This THERMOS technique would result in a saving of 40,000 to 80,000 tons of oil per year [fr

Investigation of the possibility to use a fine-mesh solver for resolving coupled neutronics and thermal-hydraulics

International Nuclear Information System (INIS)

Jareteg, K.; Vinai, P.; Demaziere, C.

2013-01-01

The development of a fine-mesh coupled neutronic/thermal-hydraulic solver is touched upon in this paper. The reported work investigates the feasibility of using finite volume techniques to discretize a set of conservation equations modeling neutron transport, fluid dynamics, and heat transfer within a single numerical tool. With the long-term objective of developing fine-mesh computing capabilities for a few selected fuel assemblies in a nuclear core, this preliminary study considers an infinite array of a single fuel assembly having a finite height. Thermal-hydraulic conditions close to the ones existing in PWRs are taken as a first test case. The neutronic modeling relies on the diffusion approximation in a multi-energy group formalism, with cross-sections pre-calculated and tabulated at the sub-pin level using a Monte Carlo technique. The thermal-hydraulics is based on the Navier-Stokes equations, complemented by an energy conservation equation. The non-linear coupling terms between the different conservation equations are fully resolved using classical iteration techniques. Early tests demonstrate that the numerical tool provides an unprecedented level of details of the coupled solution estimated within the same numerical tool and thus avoiding any external data transfer, using fully consistent models between the neutronics and the thermal-hydraulics. (authors)

Thermo-economic optimization of an endoreversible four-heat-reservoir absorption-refrigerator

International Nuclear Information System (INIS)

Qin Xiaoyong; Chen Lingen; Sun Fengrui; Wu Chih

2005-01-01

Based on an endoreversible four-heat-reservoir absorption-refrigeration-cycle model, the optimal thermo-economic performance of an absorption-refrigerator is analyzed and optimized assuming a linear (Newtonian) heat-transfer law applies. The optimal relation between the thermo-economic criterion and the coefficient of performance (COP), the maximum thermo-economic criterion, and the COP and specific cooling load for the maximum thermo-economic criterion of the cycle are derived using finite-time thermodynamics. Moreover, the effects of the cycle parameters on the thermo-economic performance of the cycle are studied by numerical examples

A coupled hydraulic and structure-dynamic model for prediction of RCCA drop time under hypothetical FA deformation

International Nuclear Information System (INIS)

Ren, Mingmin; Dressel, Bernd

2009-01-01

The ability of the RCCA (Rod Control Cluster Assemblies) in a pressurized water reactor (PWR) to be fully inserted into the core and to reach the dashpot within a required time limit is one of the important safety requirements for quick shutdown. This kind of quick shutdown in a PWR is initiated by allowing the control rod with the drive rod together to fall into the core by gravity. During normal operation, the RCCA drop time is mainly influenced by the weight of control assembly, hydraulic resistance in the CRDM (Control Rod Drive Mechanism), control rod guide assembly and guide thimbles and by the mechanical friction forces between the RCCA and its surroundings. In the case of an accident, e.g. earthquake, an additional influence of horizontal vibrations of the RCCA and its surroundings has to be considered [1]. A coupled hydraulic and structure-dynamic model is presented in this paper for prediction of RCCA drop time down to dashpot under hypothetical fuel assembly (FA) deformations. This coupled model was verified by RCCA static and dynamic drop tests with a deformed FA and by RCCA drop tests under operational conditions. (orig.)

Preparation of nano-aluminum and studies on thermo-reaction properties

International Nuclear Information System (INIS)

Wei Sheng; Wang Chaoyang; Huang Yong; Wu Weidong; Tang Yongjian; Wei Jianjun

2002-01-01

The author presents the fabrication of nano-aluminum powders by evaporation-condensation method. The thermo gravimetric-differential scanning calorimetry technique is used to characterize the thermo-reaction properties between nano-aluminum powders and N 2 or Ar. The experiment results confirm the different thermo-reaction properties between block- and nano-aluminum

The CEA and nuclear energy applications

International Nuclear Information System (INIS)

1980-01-01

With PWR reactors, CEA has applied a large part of its activities on steam generators, whilst other technical studies have involved components, maintenance, thermo-hydraulics, safety, materials, instrumentation apparatus and controls. For small light-water reactors, studies carried out have led to development of the Thermos Project: demonstrating the validity of urban heating derived from a pool-type reactor. Other studies have involved fast reactors (manufacture of fissile fuel assemblies, contributions toward the development of the Superphenix project and longer-term studies involving the overall breeder line). Finally, studies on the retreatment of irradiated fuels: aside from the retreatment of irradiated fuel programmes, CEA is pursuing its work on the TOR Project (large-scale pilot for retreatment of fast-neutron fuels) [fr

Coupled Model of channels in parallel and neutron kinetics in two dimensions

International Nuclear Information System (INIS)

Cecenas F, M.; Campos G, R.M.; Valle G, E. del

2004-01-01

In this work an arrangement of thermohydraulic channels is presented that represent those four quadrants of a nucleus of reactor type BWR. The channels are coupled to a model of neutronic in two dimensions that allow to generate the radial profile of power of the reactor. Nevertheless that the neutronic pattern is of two dimensions, it is supplemented with axial additional information when considering the axial profiles of power for each thermo hydraulic channel. The stationary state is obtained the one it imposes as frontier condition the same pressure drop for all the channels. This condition is satisfied to iterating on the flow of coolant in each channel to equal the pressure drop in all the channels. This stationary state is perturbed later on when modifying the values for the effective sections corresponding to an it assembles. The calculation in parallel of the neutronic and the thermo hydraulic is carried out with Vpm (Virtual parallel machine) by means of an outline teacher-slave in a local net of computers. (Author)

Coupled Model of channels in parallel and neutron kinetics in two dimensions; Modelo acoplado de canales en paralelo y cinetica neutronica en dos dimensiones

Energy Technology Data Exchange (ETDEWEB)

Cecenas F, M.; Campos G, R.M. [Instituto de Investigaciones Electricas, Av. Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)]. E-mail: mcf@iie.org.mx; Valle G, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

2004-07-01

In this work an arrangement of thermohydraulic channels is presented that represent those four quadrants of a nucleus of reactor type BWR. The channels are coupled to a model of neutronic in two dimensions that allow to generate the radial profile of power of the reactor. Nevertheless that the neutronic pattern is of two dimensions, it is supplemented with axial additional information when considering the axial profiles of power for each thermo hydraulic channel. The stationary state is obtained the one it imposes as frontier condition the same pressure drop for all the channels. This condition is satisfied to iterating on the flow of coolant in each channel to equal the pressure drop in all the channels. This stationary state is perturbed later on when modifying the values for the effective sections corresponding to an it assembles. The calculation in parallel of the neutronic and the thermo hydraulic is carried out with Vpm (Virtual parallel machine) by means of an outline teacher-slave in a local net of computers. (Author)

Application of a new thermo-mechanical model for the study of the nuclear waste disposal in clay rocks

International Nuclear Information System (INIS)

Dizier, A.; Li, X.L.; Francois, B.; Collin, F.; Charlier, R.

2012-01-01

Document available in extended abstract form only. One of the cornerstones of the nuclear waste disposal researches concerns the evolution of the damaged zone which can offer a preferential path for migration of radionuclide through modifications of its mechanical and hydraulic properties. Even if the thermo-mechanical behaviour of clays is well documented in the literature, the development of the damaged zone induced by an excavation with temperature is not well known. To investigate this problem, a new thermo-mechanical constitutive law has been implemented in the non-linear finite element code LAGAMINE developed at ULg (Universite de Liege) and has been used to model the PRACLAY experiment (Preliminary demonstration test for clay disposal of vitrified high level radioactive waste) at Mol URL (Underground Research Laboratory). Though several models are being to reproduce the different phenomena met when a thermal loading is applied to a clay specimen, the applications of such thermo-mechanical models to simulate large scale in-situ experiment are rare. Based on the work of Sultan a new thermo-mechanical constitutive law has been implemented in combination with a Cap model in the code LAGAMINE. The Cap model is a combination of a frictional criterion, a Cam-Clay model and a traction criterion. The influence of the temperature is considered through the thermo-mechanical law developed by Cui et al. (2000). This law permits to reproduce common features of the thermo-mechanical behaviour of clay, such as the decrease of the pre-consolidation pressure with temperature, the volume change, the thermal hardening, the transition between thermal dilation and thermal contraction for over-consolidated clays. These aspects are modelled with two curves in the (p',T) plane. The first one is related to the generation of the thermal volumetric plastic strains (TY curve (Thermal Yield)). The second one reproduces the decrease of the pre-consolidation pressure with the temperature

Thermo Techno Modern Analytical Equipment for Research and Industrial Laboratories

Directory of Open Access Journals (Sweden)

Khokhlov, S.V.

2014-03-01

Full Text Available A brief overview of some models of Thermo Techno analytical equipment and possible areas of their application is given. Thermo Techno Company was created in 2000 as a part of representative office of international corporation Thermo Fisher Scientific — world leader in manufacturing analytical equipments. Thermo Techno is a unique company in its integrated approach in solving the problems of the user, which includes a series of steps: setting the analytical task, selection of effective analysis methods, sample delivery and preparation as well as data transmitting and archiving.

Fuel assembly

International Nuclear Information System (INIS)

Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

1995-01-01

The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

Improving intrinsic corrosion reliability of printed circuit board assembly

DEFF Research Database (Denmark)

Ambat, Rajan; Conseil, Helene

2016-01-01

conditions, therefore the protection of electronic devices is becoming a critical factor in system design. Humidity and local condensation inside electronic enclosures can significantly alter the performance of electronic devices. The presence of moisture in a PCB alters its quality, functionality, thermal...... performance, and thermo-mechanical properties, while condensation on the surface of printed circuit board assemblies (PCBAs) can lead to electrical failures, like electrochemical migration. The result is reduced life span for electronic products and heavy economic loss due to failures....

Library of neutron cross sections of the Thermos code; Biblioteca de secciones eficaces de neutrones del codigo Thermos

Energy Technology Data Exchange (ETDEWEB)

Alonso V, G; Hernandez L, H [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

1991-10-15

The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

Stability study of MMC tubes and advanced assemblies for telescope structures

Science.gov (United States)

Nivet-Lutz, Martine; Pommatau, Gilles

2017-11-01

This paper presents new advances concerning the development of an Aluminum Matrix Composite for dimensionally stable satellite structures. Feasibility of thermally stable thin-walled tubes have been acquired through microstructure observation and Coefficient of Thermal Expansion measurement. In order to fit the thermo-mechanical stability domain of tubes on specifications, 3 thermal cycles have been tested, regarding to relaxation of internal stress and changes in macroscopic thermo-mechanical properties. Experimental expansion curves and microstructure observation show that thermal treatments permitts such a good fitting. For a better understanding of physical internal phenomena, internal stress has been measured by neutron diffraction on tube samples after each thermal treatment. Results show a significant decrease of stress due to cycling in cold temperature. In order to decrease the absolute value of CTE of assemblies, a new concept of thermo-mechanical stable linkage has been developped, which consists in a common alumlinum infiltration of superposed carbon preforms. Structural bonding, which usually affects stability properties and impose surface treatments and polymerization, can so be avoided. The study has been achieved through CNES (Centre National d'Etudes Spatiales) and French Ministry of Defense (DGA) supports.

On the general theory of thermo-elastic friction

NARCIS (Netherlands)

Alblas, J.B.

1961-01-01

A theory of the thermo-elastic dissipation in vibrating bodies is developed, starting from the three-dimensional thermo-elastic equations. After a discussion of the basic thermodynamical foundations, some general considerations on the problem of the conversion of mechanical energy into heat are

Laser thermal effect on silicon nitride ceramic based on thermo-chemical reaction with temperature-dependent thermo-physical parameters

International Nuclear Information System (INIS)

Pan, A.F.; Wang, W.J.; Mei, X.S.; Wang, K.D.; Zhao, W.Q.; Li, T.Q.

2016-01-01

Highlights: • A two-dimensional thermo-chemical reaction model is creatively built. • Thermal conductivity and heat capacity of β-Si_3N_4 are computed accurately. • The appropriate thermo-chemical reaction rate is fitted and reaction element length is set to assure the constringency. • The deepest ablated position was not the center of the ablated area due to plasma absorption. • The simulation results demonstrate the thermo-chemical process cant be simplified to be physical phase transition. - Abstract: In this study, a two-dimensional thermo-chemical reaction model with temperature-dependent thermo-physical parameters on Si_3N_4 with 10 ns laser was developed to investigate the ablated size, volume and surface morphology after single pulse. For model parameters, thermal conductivity and heat capacity of β-Si_3N_4 were obtained from first-principles calculations. Thermal-chemical reaction rate was fitted by collision theory, and then, reaction element length was deduced using the relationship between reaction rate and temperature distribution. Furthermore, plasma absorption related to energy loss was approximated as a function of electron concentration in Si_3N_4. It turned out that theoretical ablated volume and radius increased and then remained constant with increasing laser energy, and the maximum ablated depth was not in the center of the ablated zone. Moreover, the surface maximum temperature of Si_3N_4 was verified to be above 3000 K within pulse duration, and it was much higher than its thermal decomposition temperature of 1800 K, which indicated that Si_3N_4 was not ablated directly above the thermal decomposition temperature. Meanwhile, the single pulse ablation of Si_3N_4 was performed at different powers using a TEM_0_0 10 ns pulse Nd:YAG laser to validate the model. The model showed a satisfactory consistence between the experimental data and numerical predictions, presenting a new modeling technology that may significantly increase the

Development of CFD software for the simulation of thermal hydraulics in advanced nuclear reactors. Final report

International Nuclear Information System (INIS)

Bachar, Abdelaziz; Haslinger, Wolfgang; Scheuerer, Georg; Theodoridis, Georgios

2015-01-01

The objectives of the project were: Improvement of the simulation accuracy for nuclear reactor thermo-hydraulics by coupling system codes with three-dimensional CFD software; Extension of CFD software to predict thermo-hydraulics in advanced reactor concepts; Validation of the CFD software by simulation different UPTF TRAM-C test cases and development of best practice guidelines. The CFD module was based on the ANSYS CFD software and the system code ATHLET of GRS. All three objectives were met: The coupled ATHLET-ANSYS CFD software is in use at GRS and TU Muenchen. Besides the test cases described in the report, it has been used for other applications, for instance the TALL-3D experiment of KTH Stockholm. The CFD software was extended with material properties for liquid metals, and validated using existing data. Several new concepts were tested when applying the CFD software to the UPTF test cases: Simulations with Conjugate Heat Transfer (CHT) were performed for the first time. This led to better agreement between predictions and data and reduced uncertainties when applying temperature boundary conditions. The meshes for the CHT simulation were also used for a coupled fluid-structure-thermal analysis which was another novelty. The results of the multi-physics analysis showed plausible results for the mechanical and thermal stresses. The workflow developed as part of the current project can be directly used for industrial nuclear reactor simulations. Finally, simulations for two-phase flows with and without interfacial mass transfer were performed. These showed good agreement with data. However, a persisting problem for the simulation of multi-phase flows are the long simulation times which make use for industrial applications difficult.

Measurements of subchannel velocity and pressure drop for HANARO fuel assembly

International Nuclear Information System (INIS)

Yang, Sun Kyu; Jeong, Heung Jun; Cho, Suk; Min, Kyung Ho; Jeong, Moon Ki

1996-07-01

This report presents the hydraulic test results for HANARO fuel assemblies, which are performed to obtain the axial velocity and pressure drop data to be used to validate the code calculation model. For both 18 and 36-element fuel assemblies axial velocities of the entrance and exit regions are obtained, and developing axial velocity profiles along the flow direction for the fuel region of 18-element fuel assembly are also obtained. Varying the pressure tap locations, pressure drop data for each component of fuel assembly are obtained for various flow conditions. From the pressure drop test results it is noted that the pressure drops across the fuel assembly are 214 kPa and 205 kPa for the 18-element and 36-element fuel assembly respectively. 39 tabs., 12 figs., 5 refs. (Author)

Wigner Function of Thermo-Invariant Coherent State

International Nuclear Information System (INIS)

Xue-Fen, Xu; Shi-Qun, Zhu

2008-01-01

By using the thermal Winger operator of thermo-field dynamics in the coherent thermal state |ξ) representation and the technique of integration within an ordered product of operators, the Wigner function of the thermo-invariant coherent state |z,ℵ> is derived. The nonclassical properties of state |z,ℵ> is discussed based on the negativity of the Wigner function. (general)

Thermo-driven microcrawlers fabricated via a microfluidic approach

International Nuclear Information System (INIS)

Wang Wei; Yao Chen; Zhang Maojie; Ju Xiaojie; Xie Rui; Chu Liangyin

2013-01-01

A novel thermo-driven microcrawler that can transform thermal stimuli into directional mechanical motion is developed by a simple microfluidic approach together with emulsion-template synthesis. The microcrawler is designed with a thermo-responsive poly(N-isopropylacrylamide) (PNIPAM) hydrogel body and a bell-like structure with an eccentric cavity. The asymmetric shrinking–swelling circulation of the microcrawlers enables a thermo-driven locomotion responding to repeated temperature changes, which provides a novel model with symmetry breaking principle for designing biomimetic soft microrobots. The microfluidic approach offers a novel and promising platform for design and fabrication of biomimetic soft microrobots. (paper)

Thermo-cleavable polymers: Materials with enhanced photochemical stability

DEFF Research Database (Denmark)

Manceau, Matthieu; Petersen, Martin Helgesen; Krebs, Frederik C

2010-01-01

Photochemical stability of three thermo-cleavable polymers was investigated as thin films under atmospheric conditions. A significant increase in lifetime was observed once the side-chain was cleaved emphasizing the detrimental effect of solubilizing groups on the photochemical stability of conju......Photochemical stability of three thermo-cleavable polymers was investigated as thin films under atmospheric conditions. A significant increase in lifetime was observed once the side-chain was cleaved emphasizing the detrimental effect of solubilizing groups on the photochemical stability...... of conjugated polymers. In addition to their ease of processing, thermo-cleavable polymers thus also offer a greater intrinsic stability under illumination....

European liquid metal thermal-hydraulics R and D: present and future

International Nuclear Information System (INIS)

Roelofs, F.; Batta, A.; Bandini, G.; Van Tichelen, K.; Gerschenfeld, A.; Cheng, X.

2014-01-01

A large role is attributed in the future within the European Sustainable Nuclear Energy Technology Platform (SNE-TP) and especially the underlying European Sustainable Nuclear Industry Initiative (ESNII) to the application of fast reactors for sustainable nuclear energy production. Specifically, fast reactors are considered attractive because of their possibility to use natural resources efficiently and to reduce the volume and lifetime of nuclear waste. Currently four demonstration projects have a promising outlook in Europe, i.e. the ASTRID project in France, the MYRRHA project in Belgium, the ALFRED project developed in Europe and to be built in Romania, and the ELECTRA project in Sweden. Sodium and lead(-alloys) are envisaged as coolants for these reactors. Obviously, in the development of these reactors, thermal-hydraulics is recognized as a key challenge with emphasis on safety issues. This paper will discuss the present development status of liquid metal cooled reactor thermal-hydraulics as an outcome of the European 7. framework programme THINS (Thermal-Hydraulics for Innovative Nuclear Systems) project. The main project results with respect to liquid metal cooled reactors will be summarized, i.e. turbulence heat transfer model development, fuel assembly analysis, pool thermal-hydraulics, system behaviour, multi-phase physics, and multiscale thermal-hydraulics simulation. In conclusion, the main challenges for future developments will be indicated. Emphasis will be put on the important experimental and numerical challenges. (authors)

Thermal hydraulic behavior of sub-assembly local blockage in China experiment fast reactor

International Nuclear Information System (INIS)

Yang Zhimin

2000-01-01

The geometrical parameter ratio of pitch to diameter of China Experiment Fast Reactor (CEFR) subassembly is 1,167. To address the thermal hydraulic behavior of subassembly local blockage which may be caused by deformation of cladding due to severe swelling and thermal stresses and by space swirl loosening etc., the porous numerical model and SIMPLE-P code used to solve Navier-Stokes and energy equations in porous medium was developed, and the bundle experiment with 19 pins with 24 subchannels blocked in the sodium coolant was carried on in China Institute of Atomic Energy (CIAE). The comparison of code predictions against experiments (including non-blockage and ten blockage conditions) seems well. The thermal hydraulic behavior of fuel subassembly with 61 fuel pins blockage of CEFR is calculated with SIMPLE-P code. The results indicate that the maximum temperature is 815 deg. C when the blockage area is about 37% (54 central subchannels are blocked). In this case the cladding won't be damaged and no sodium coolant boiling takes place. (author)

Hydraulic engine valve actuation system including independent feedback control

Science.gov (United States)

Marriott, Craig D

2013-06-04

A hydraulic valve actuation assembly may include a housing, a piston, a supply control valve, a closing control valve, and an opening control valve. The housing may define a first fluid chamber, a second fluid chamber, and a third fluid chamber. The piston may be axially secured to an engine valve and located within the first, second and third fluid chambers. The supply control valve may control a hydraulic fluid supply to the piston. The closing control valve may be located between the supply control valve and the second fluid chamber and may control fluid flow from the second fluid chamber to the supply control valve. The opening control valve may be located between the supply control valve and the second fluid chamber and may control fluid flow from the supply control valve to the second fluid chamber.

Design of a hydraulic loop for characterization of nuclear fuels for the Rech-1

International Nuclear Information System (INIS)

Munoz Reveco, David Hernan

2016-01-01

The Chilean Nuclear Energy Commission (CCHEN), in particular the Fuels Elements Plant (PEC), wants to increase its capacity of design of nuclear fuel elements (ECN). The International Atomic Energy Agency (IAEA) stipulates that in order to develop new ECN designs must be met with neutron, structural, thermal and hydraulics. The CCHEN in the first instance, wishes to implement the hydraulic tests with a test bench which is called 'Hydraulic Loop (LH)'. The general objective of the project is to design a LH at the level of detailed engineering, with the purpose to characterize ECN hydraulically of the RECH-1 reactor. The specific objectives are: (i) Carry out the conceptual design and control philosophy applicable to the LH, ii) Develop the design LH baseline including layout, equipment, instrumentation and assembly; and iii) Carry out the detailed design for LH construction and assembly including technical specifications for the acquisition, construction and assembly of LH, and estimated investments to materialize the project. To carry out the conceptual design of the LH proceed to review piping projects provided by specialists, fluid pressure drop analysis is carried out in pipelines and papers are reviewed on existing LHs (Petten and IPEN). The basic design is developed by setting the conditions system, taking as a design criterion the operating the RECH-2 reactor. For the design of the test zone, the ASME code VIII division 1 section 13-7. The pond develops according to the capacity of fluid contained in the system. The pump is selected by system losses, overestimating the load of the pump. In addition, the characteristic curves of the pump are compared with the curve feature of the system. Detail planes are generated with Autodesk Inventor software Professional 2014-student version. The instrumentation selection is carried out with the advice of a company dedicated to the measurement and control of industrial activities (VETO). The main results of the project

Hydraulic structures

CERN Document Server

Chen, Sheng-Hong

2015-01-01

This book discusses in detail the planning, design, construction and management of hydraulic structures, covering dams, spillways, tunnels, cut slopes, sluices, water intake and measuring works, ship locks and lifts, as well as fish ways. Particular attention is paid to considerations concerning the environment, hydrology, geology and materials etc. in the planning and design of hydraulic projects. It also considers the type selection, profile configuration, stress/stability calibration and engineering countermeasures, flood releasing arrangements and scouring protection, operation and maintenance etc. for a variety of specific hydraulic structures. The book is primarily intended for engineers, undergraduate and graduate students in the field of civil and hydraulic engineering who are faced with the challenges of extending our understanding of hydraulic structures ranging from traditional to groundbreaking, as well as designing, constructing and managing safe, durable hydraulic structures that are economical ...

Development of a detailed BWR core thermal-hydraulic analysis method based on the Japanese post-BT standard using a best-estimate code

International Nuclear Information System (INIS)

Ono, H.; Mototani, A.; Kawamura, S.; Abe, N.; Takeuchi, Y.

2004-01-01

The post-BT standard is a new fuel integrity standard or the Atomic Energy Society of Japan that allows temporary boiling transition condition in the evaluation for BWR anticipated operational occurrences. For application of the post-BT standard to BWR anticipated operational occurrences evaluation, it is important to identify which fuel assemblies and which axial, radial positions of fuel rods have temporarily experienced the post-BT condition and to evaluates how high the fuel cladding temperature rise was and how long the dryout duration continued. Therefore, whole bundle simulation, in which each fuel assembly is simulated independently by one thermal-hydraulic component, is considered to be an effective analytical method. In the present study, a best-estimate thermal-hydraulic code, TRACG02, has been modified to extend it predictive capability by implementing the post-BT evaluation model such as the post-BT heat transfer correlation and rewetting correlation and enlarging the number of components used for BWR plant simulation. Based on new evaluation methods, BWR core thermal-hydraulic behavior has been analyzed for typical anticipated operational occurrence conditions. The location where boiling transition occurs and the severity of fuel assembly in the case of boiling transition conditions such as fuel cladding temperature, which are important factors in determining whether the reuse of the fuel assembly can be permitted, were well predicted by the proposed evaluation method. In summary, a new evaluation method for a detailed BWR core thermal-hydraulic analysis based on the post-BT standard of the Atomic Energy Society of Japan has been developed and applied to the evaluation of the post-BT standard during the actual BWR plant anticipated operational occurrences. (author)

Linear hydraulic drive system for a Stirling engine

Science.gov (United States)

Walsh, Michael M.

1984-02-21

A hydraulic drive system operating from the periodic pressure wave produced by a Stirling engine along a first axis thereof and effecting transfer of power from the Stirling engine to a load apparatus therefor and wherein the movable, or working member of the load apparatus is reciprocatingly driven along an axis substantially at right angles to the first axis to achieve an arrangement of a Stirling engine and load apparatus assembly which is much shorter and the components of the load apparatus more readily accessible.

Numerical Simulation of a Single-Phase Closed-Loop Thermo-Siphon in LORELEI Test Device

International Nuclear Information System (INIS)

Gitelman, D.; Shenha, H.; Gonnier, Ch.; Tarabelli, D.; Sasson, A.; Weiss, Y.; Katz, M.

2014-01-01

The LORELEI experimental setup in the Jules Horowitz Reactor (JHR) is dedicated for the study of fuel during a Loss of Coolant Accident (LOCA). The main objective of the LORELEI(2) (Light-Water One-Rod Equipment for LOCA Experimental Investigation) is to study the thermal-mechanical behavior of fuel during such an accident and to produce a short half-life fission products source term. In order to study those phenomena, the fuel sample will experience a transient neutron flux field, which in turn will generate a Linear Heat Generation Rate (LHGR) and determine the temperature of the fuel and its cladding, simulating the behavior of the fuel and the cladding during a LOCA accident. In order to reproduce a LOCA-type transient sequence, the experimental test device will be located on a displacement device. The displacement device moves the test device in the flux field in order to generate a representing LHGR in the fuel or temperature of its cladding. The LOCA-type transient sequence has four major features: „h An adiabatic heating of the fuel up to the ballooning and burst occurrence. „h High temperature plateau which will promote clad oxidation. „h Passive precooling by thermal inertia. „h Water re-flooding and quenching. The challenge in the thermo-hydraulic design of the LORELEI test section is in defining a one closed water capsule design that can operate as a thermo-siphon at re-irradiation phase and also can reproduce all LOCA-type transient sequence phases. This design should be validated and verified to fill all safety and regulation requirements. This work aims to investigate fluid flow behavior of a single-phase thermo-siphon in the LORELEI test device, as part of the conceptual design and optimization study. The complexity of the flow field in the LORELEI test device, as a closed-loop thermo-siphon, is due to the opposing forces in the device - buoyancy forces and natural convection flow generated (mainly) by the fuel power in the hot channel

Design analysis of a new SCWR fuel assembly using a coupled method

International Nuclear Information System (INIS)

Liu Xiaojing; Yang Ting; Cheng Xu

2011-01-01

Among the six GEN-Ⅳ reactor concepts recommended by the Gen-Ⅳ International Forum (GIF), supercritical water-cooled reactor (SCWR) is the only reactor type with water as coolant. Compared to the existing reactors, it has economic advantage and technology continuity. Based on the newly developed coupling code, analysis on the square SCWR assembly is carried out in this paper. A new design concept of SCWR fuel assembly is proposed. The results achieved so far indicate favorable thermal-hydraulic performance and neutron-physical behavior of the new fuel assembly compared to the previous ones. (authors)

Hydraulic System Design of Hydraulic Actuators for Large Butterfly Valves

Directory of Open Access Journals (Sweden)

Ye HUANG

2014-09-01

Full Text Available Hydraulic control systems of butterfly valves are presently valve-controlled and pump-controlled. Valve-controlled hydraulic systems have serious power loss and generate much heat during throttling. Pump-controlled hydraulic systems have no overflow or throttling losses but are limited in the speed adjustment of the variable-displacement pump, generate much noise, pollute the environment, and have motor power that does not match load requirements, resulting in low efficiency under light loads and wearing of the variable-displacement pump. To overcome these shortcomings, this article designs a closed hydraulic control system in which an AC servo motor drives a quantitative pump that controls a spiral swinging hydraulic cylinder, and analyzes and calculates the structure and parameters of a spiral swinging hydraulic cylinder. The hydraulic system adjusts the servo motor’s speed according to the requirements of the control system, and the motor power matches the power provided to components, thus eliminating the throttling loss of hydraulic circuits. The system is compact, produces a large output force, provides stable transmission, has a quick response, and is suitable as a hydraulic control system of a large butterfly valve.

COUPLED SIMULATION OF GAS COOLED FAST REACTOR FUEL ASSEMBLY WITH NESTLE CODE SYSTEM

Directory of Open Access Journals (Sweden)

Filip Osusky

2018-05-01

Full Text Available The paper is focused on coupled calculation of the Gas Cooled Fast Reactor. The proper modelling of coupled neutronics and thermal-hydraulics is the corner stone for future safety assessment of the control and emergency systems. Nowadays, the system and channel thermal-hydraulic codes are accepted by the national regulatory authorities in European Union for license purposes, therefore the code NESTLE was used for the simulation. The NESTLE code is a coupled multigroup neutron diffusion code with thermal-hydraulic sub-channel code. In the paper, the validation of NESTLE code 5.2.1 installation is presented. The processing of fuel assembly homogeneous parametric cross-section library for NESTLE code simulation is made by the sequence TRITON of SCALE code package system. The simulated case in the NESTLE code is one fuel assembly of GFR2400 concept with reflective boundary condition in radial direction and zero flux boundary condition in axial direction. The results of coupled calculation are presented and are consistent with the GFR2400 study of the GoFastR project.

Activated carbon from thermo-compressed wood and other lignocellulosic precursors

Directory of Open Access Journals (Sweden)

Capart, R.

2007-05-01

Full Text Available The effects of thermo-compression on the physical properties such as bulk density, mass yield, surface area, and also adsorption capacity of activated carbon were studied. The activated carbon samples were prepared from thermo-compressed and virgin fir-wood by two methods, a physical activation with CO2 and a chemical activation with KOH. A preliminary thermo-compression method seems an easy way to confer to a tender wood a bulk density almost three times larger than its initial density. Thermo-compression increased yield regardless of the mode of activation. The physical activation caused structural alteration, which enhanced the enlargement of micropores and even their degradation, leading to the formation of mesopores. Chemical activation conferred to activated carbon a heterogeneous and exclusively microporous nature. Moreover, when coupled to chemical activation, thermo-compression resulted in a satisfactory yield (23%, a high surface area (>1700 m2.g-1, and a good adsorption capacity for two model pollutants in aqueous solution: methylene blue and phenol. Activated carbon prepared from thermo-compressed wood exhibited a higher adsorption capacity for both the pollutants than did a commercial activated carbon.

Thermo-emf of cermet films based on rare earth borides

International Nuclear Information System (INIS)

Islamgaliev, R.K.; Zyrin, A.V.; Shulishova, O.I.; Shcherbak, I.A

1987-01-01

Thermo-emf and electric conductivity of granulated films which contain a solid solution of europium and praseodymium borides Eu 0.5 Pr 0.5 B 6 as a conducting phase, and glass-crystal binder on the base of alummomagnesial fluosilicates as a dielectric phase are studied within the temperature range of 100-1100 K. Thermo-emf of films has a negative sign within the temperature range of 100-500 K and does not exceed 5 μkV/K according to the absolute value which is close to the value of the conducting phase thermo-emf. A negative sign and a small value of thermo-emf are indicative of the charge transfer in granulated films by electrons. Contribution of each of the components into the general thermo-emf is different at high temperatures in different temperature ranges and depends on the individual physico-chemical properties of the used materials

Test-element assembly and loading parameters for the in-pile test of HCPB ceramic pebble beds

Energy Technology Data Exchange (ETDEWEB)

Laan, J.G. van der E-mail: vanderlaan@nrg-nl.com; Boccaccini, L.V.; Conrad, R.; Fokkens, J.H.; Jong, M.; Magielsen, A.J.; Pijlgroms, B.J.; Reimann, J.; Stijkel, M.P.; Malang, S

2002-11-01

In the framework of developing the helium cooled pebble-bed (HCPB) blanket an irradiation test of pebble-bed assemblies is prepared at the HFR Petten. The test objective is to concentrate on the effect of neutron irradiation on the thermal-mechanical behaviour of the HCPB breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. The paper reports on the project status, and presents the results of pre-tests, material characteristics, the manufacturing of the pebble-bed assemblies, and the nuclear and thermo-mechanical loading parameters.

Thermo-hydraulic free piston engine as a primary propulsion unit in mobile hydraulic drives; Die thermohydraulische Freikolbenmaschine - ein neues Antriebskonzept fuer hydraulische angetriebene Fahrzeuge

Energy Technology Data Exchange (ETDEWEB)

Brunner, H. [Technische Univ. Dresden (Germany)

2004-07-01

The principle function of a free piston engine was tested on a test stand. The engine can drive hydraulic loads as a primary aggregate in a storage-based constant pressure network. Its power is independent of the loads. The engine is operated in intermittent operation and at the optimal operating point. There are no idle or part-load fractions. Measurements so far have shown that the performance of the new system equals that of a current combination of internal combustion engine and axial piston pump in their optimal operating point. In cyclic operation, the performance is even better. (orig.)

Characteristics of Core Thermal-Hydraulic Design of SMART-P

International Nuclear Information System (INIS)

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

2006-01-01

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

Cell for receipting and dismantling nuclear fuel assembly

International Nuclear Information System (INIS)

Beneck, J.A.; Quayre, C.

1989-01-01

The cell has a vertical structure with a right section corresponding at that of the assembly to receive, a mechanism for keeping fuel pins at their nominal separation in the form of at least two combs and mechanisms of holding grids and bottom nozzle. The comb arrangements are moved into position by hydraulic actuators so that they cross each other to form a lattice round the fuel pins. The mechanism for holding grid assemblies consist of joints that articulate from a free position to a position where the joints press of the grid on all sides [fr

Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor

International Nuclear Information System (INIS)

Belem, J.A.T.

1993-09-01

The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs

Thermal hydraulic performance of naturally aspirated control rod housing assemblies

International Nuclear Information System (INIS)

Geiger, G.T.; Randolph, H.W.; Paik, I.K.; Foti, D.J.

1992-01-01

Savannah River Site reactors are comprised of heat generating fuel/target assemblies, control rods which regulate reactor power, and heavy water which acts as the coolant and as a moderator. The fuel/target assemblies are cooled by the downflow of heavy water while the control rods are cooled via upflow. Five control rods are grouped with two safety rods in seven-channel assemblies called septifoils. Under normal operating conditions, the reactor power level, radial shape flux and axial power flux are regulated by the positioning of the control rods. The control rods are solid rods of a lithium-aluminum alloy with an thin aluminum outer sheath. Lithium is a good absorber of neutrons and, thus control rod temperatures rise with reactor power. At conditions of sufficiently high reactor power and degraded coolant flow, the control rods could heat sufficiently to cause a metallurigical failure of the sheath leading to molten material coming in contact with water and the possibility of a steam explosion. An accident has been postulated as part of the analysis involving the safety upgrade of Savannah River Site reactors in which the housing is not seated on the pin. Coolant from the upflow pin would not be directed into the housing but, into the moderator space surrounding the housing. Only naturally aspirated cooling due to buoyancy effects would be available to cool the control rods and the coolant mass flow rate would drop significantly from its nominal value. In this study, the mechanisms and limits of cooling heated rods housed in an unseated septifoil are addressed. Experiments were conducted on a shortened, prototypic housing with electrically heated rods to gain an understanding of the phenomena governing the cooling in such a case and develop data which can be used to evaluate predictive models. These experiments are described, their results discussed, and the predictions of current models is presented

Model with Peach Bottom Turbine trip and thermal-Hydraulic code TRACE V5P3

International Nuclear Information System (INIS)

Mesado, C.; Miro, R.; Barrachina, T.; Verdu, G.

2014-01-01

This work is the continuation of the work presented previously in the thirty-ninth meeting annual of the Spanish Nuclear society. The semi-automatic translation of the Thermo-hydraulic model TRAC-BF1 Peach Bottom Turbine Trip to TRACE was presented in such work. This article is intended to validate the model obtained in TRACE, why compare the model results result from the translation with the Benchmark results: NEA/OECD BWR Peach Bottom Turbine Trip (PBTT), in particular is of the extreme scenario 2 of exercise 3, in which there is SCRAM in the reactor. Among other data present in the (transitional) Benchmark , are: total power, axial profile of power, pressure Dome, total reactivity and its components. (Author)

Hydraulic Arm Modeling via Matlab SimHydraulics

Czech Academy of Sciences Publication Activity Database

Věchet, Stanislav; Krejsa, Jiří

2009-01-01

Roč. 16, č. 4 (2009), s. 287-296 ISSN 1802-1484 Institutional research plan: CEZ:AV0Z20760514 Keywords : simulatin modeling * hydraulics * SimHydraulics Subject RIV: JD - Computer Applications, Robotics

Thermo-mechanical tests on W7-X current lead flanges

International Nuclear Information System (INIS)

Dhard, Chandra Prakash; Rummel, Thomas; Zacharias, Daniel; Bykov, Victor; Moennich, Thomas; Buscher, Klaus-Peter

2013-01-01

Highlights: • There are significant mechanical loads on the cryostat and radial flanges for W7-X current leads. • These are due to evacuation of W7-X cryostat, cool-down of cold mass, electro-magnetic forces and self weight of leads. • The actual mechanical loads were reduced to simplify the experimental set-up. • The tests were carried out on mock-up flanges test assembly at ambient temperature and at 77 K. • The thermo-mechanical tests on W7-X current lead flanges validate the design and joints of these flanges to the leads. -- Abstract: Fourteen pieces of high temperature superconducting current leads (CL) arranged in seven pairs, will be installed on the outer vessel of Wendelstein 7-X (W7-X) stellarator. In order to support the CL, it is provided with two glass fiber reinforce plastic (GFRP) flanges, namely, the lower cryostat flange (CF) remaining at room temperature and upper radial flange (RF) at about 5 K. Both the flanges i.e. CF and RF experience high mechanical loads with respect to the CL, due to the evacuation of W7-X cryostat, cool-down of cold mass including the CL, electro-magnetic forces due to current and plasma operations and self weight of CL. In order to check the integrity of these flanges for such mechanical loads, thermo-mechanical tests were carried out on these flanges at room temperatures and at liquid nitrogen (LN2) temperatures. The details of test set-up, results and modeling are described in the paper

Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

International Nuclear Information System (INIS)

Cui, Shijie; Zhang, Dalin; Cheng, Jie; Tian, Wenxi; Su, G.H.

2017-01-01

As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

Energy Technology Data Exchange (ETDEWEB)

Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

2017-01-15

As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

Thermo-Mechanical tests for the CLIC two-beam module study

CERN Document Server

Xydou, A; Riddone, G; Daskalaki, E

2014-01-01

The luminosity goal of CLIC requires micron level precision with respect to the alignment of the components on its two-meter long modules, composing the two main linacs. The power dissipated inside the module components introduces mechanical deformations affecting their alignment and therefore the resulting machine performance. Several two-beam prototype modules must be assembled to extensively measure their thermo-mechanical behavior under different operation modes. In parallel, the real environmental conditions present in the CLIC tunnel should be studied. The air conditioning and ventilation system providing specified air temperature and flow has been installed in the dedicated laboratory. The power dissipation occurring in the modules is being reproduced by the electrical heaters inserted inside the RF structure mock-ups and the quadrupoles. The efficiency of the cooling systems is being verified and the alignment of module components is monitored. The measurement results will be compared to finite elemen...

Thermal-hydraulic modeling of porous bed reactors

International Nuclear Information System (INIS)

Araj, K.J.; Nourbakhsh, H.P.

1987-01-01

Optimum design of nuclear reactor cores requires an iterative approach between the thermal-hydraulic, neutronic, and operational analysis. This paper will concentrate on the thermal-hydraulic behavior of a hydrogen-cooled small particle bed reactor (PBR). The PBR core modeled here consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flows, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit to a common plenum. A fast running one-dimensional lumped-parameter steady-state code (FTHP) was developed to evaluate the effects of design changes in fuel assembly and power distribution. Another objective for the code was to investigate various methods of coolant control to minimize hot channel effects and maximize outlet temperatures

Did Life Emerge in Thermo-Acidic Conditions?

Science.gov (United States)

Holmes, D. S.

2017-12-01

There is widespread, but not unanimous, agreement that life emerged in hot conditions by exploiting redox and pH disequilibria found on early earth. Although there are several hypotheses to explain the postulated pH disequilibria, few of these consider that life evolved at very low pH (biological evolution. This presentation will evaluate the pros and cons of the hypothesis that the early evolution of life occurred in thermo-acidic conditions. Such environments are thought to have been abundant on early earth and were probably rich in hydrogen and soluble metals including iron and sulfur that could have served as sources and sinks of electrons. Extant thermo-acidophiles thrive in such conditions. Low pH environments are rich in protons that are the major drivers of energy conservation by coupling to phosphorylation in virtually all organisms on earth; this may be a "biochemical fossil" reflecting the use of protons (low pH) in primitive energy conservation. It has also been proposed that acidic conditions favored the evolution of an RNA world with expanded catalytic activities. On the other hand, the idea that life emerged in thermo-acidic conditions can be challenged because of the proposed difficulties of folding and stabilizing proteins simultaneously exposed to high temperature and low pH. In addition, although thermo-acidophiles root to the base of the phylogenetic tree of life, consistent with the proposition that they evolved early, yet there are problems of interpretation of their subsequent evolution that cloud this simplistic phylogenetic view. We propose solutions to these problems and hypothesize that life evolved in thermo-acidic conditions.

Analogy between dynamics of thermo-rheological and piezo-rheological pendulums

International Nuclear Information System (INIS)

Hedrih, K

2008-01-01

The constitutive stress-strain relations of the standard thermo-rheological and piezo-rheological hereditary element in differential form as well as in two different integro-differential forms are defined. The considered problem of a thermo-rheological hereditary discrete system nonlinear dynamics in the form of thermo-rheological double pendulum system with coupled pendulums gets the significance of two constrained bodies in plane motion problem, as a problem important for studying a sensor dynamics or actuator dynamics in active structure dynamics. System of the averaged equations in the first approximation for amplitudes and phases are derived and qualitatively analyzed. Analogy between nonlinear dynamics of the double pendulum systems with thermo-rheological and piezo-rheological properties between pendulums is pointed out

Thermal and hydraulic properties of the concrete used in the ILW repository of El Cabril (Spain). Preliminary laboratory tests

International Nuclear Information System (INIS)

Villar, Maria Victoria

2012-01-01

This work is a contribution to the understanding of the behaviour of concrete barriers in surface (low and intermediate-level) waste disposal facilities, in particular in the Spanish disposal facility of El Cabril, where the waste containers are placed inside concrete cells. The durability of concrete and its mechanical properties are intrinsically bound to moisture transport effects, especially when it is subjected to repeated wetting and drying regimes, and that is why a detailed thermo-hydraulic characterisation is necessary to model its behaviour. The concrete used in this experimental work has a characteristic strength of 350 kp/cm 2 and uses ordinary Portland cement, resistant to sulphates and seawater, with a water/cement ratio of 0.43. Its average pore size is 0.03 micrometer. In addition to the determination of the thermal conductivity of concrete as a function of water content, a hydraulic characterisation - including determination of saturated permeability, permeability to gas for different degrees of saturation and water retention curves - has been performed

Optimised Iteration in Coupled Monte Carlo - Thermal-Hydraulics Calculations

Science.gov (United States)

Hoogenboom, J. Eduard; Dufek, Jan

2014-06-01

This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration method are also tested and it is concluded that the presented iteration method is near optimal.

Optimized iteration in coupled Monte-Carlo - Thermal-hydraulics calculations

International Nuclear Information System (INIS)

Hoogenboom, J.E.; Dufek, J.

2013-01-01

This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration methods are also tested and it is concluded that the presented iteration method is near optimal. (authors)

Critical analysis of soil hydraulic conductivity determination using monoenergetic gamma radiation attenuation

International Nuclear Information System (INIS)

Portezan Filho, Otavio

1997-01-01

Three soil samples of different textures: LVA (red yellow latosol), LVE (dark red latosol) and LRd (dystrophic dark red latosol) were utilized for unsaturated hydraulic conductivity K(θ) measurements. Soil bulk densities and water contents during internal water drainage were measured by monoenergetic gamma radiation attenuation, using homogeneous soil columns assembled in the laboratory. The measurements were made with a collimated gamma beam of 0.003 m in diameter using a Nal(Tl) (3'' x 3 '') detector and a 137 Cs gamma source of 74 X 10 8 Bq and 661.6 KeV. Soil columns were scanned with the gamma beam from 0.01 to 0.20 m depth, in 0.01m steps, for several soil water redistribution times. The results show a great variability of the unsaturated hydraulic conductivity relation K(θ), even though homogeneous soils were used. The variability among methods is significantly smaller in relation to variability in space. The assumption of unit hydraulic gradient during redistribution of soil water utilized in the methods of Hillel, Libardi and Sisson leads to hydraulic conductivity values that increase in depth. The exponential character of the K(θ) relationship, is responsible for the difficulty of estimating soil hydraulic conductivity, which is a consequence of small variations in the porous arrangement, even in samples supposed to be homogeneous. (author)

Preparation of thermo-responsive graft copolymer by using a novel macro-RAFT agent and its application for drug delivery.

Science.gov (United States)

Song, Cunfeng; Yu, Shirong; Liu, Cheng; Deng, Yuanming; Xu, Yiting; Chen, Xiaoling; Dai, Lizong

2016-05-01

A methodology to prepare thermo-responsive graft copolymer by using a novel macro-RAFT agent was proposed. The macro-RAFT agent with pendant dithioester (ZC(S)SR) was facilely prepared via the combination of RAFT polymerization and esterification reaction. By means of ZC(S)SR-initiated RAFT polymerization, the thermo-responsive graft copolymer consisting of poly(methyl methacrylate-co-hydroxylethyl methacrylate) (P(MMA-co-HEMA)) backbone and hydrophilic poly(N-isopropylacrylamide) (PNIPAAm) side chains was constructed through the "grafting from" approach. The chemical compositions and molecular weight distributions of the synthesized polymers were respectively characterized by (1)H nuclear magnetic resonance ((1)H NMR) and gel permeation chromatography (GPC). Self-assembly behavior of the amphiphilic graft copolymers (P(MMA-co-HEMA)-g-PNIPAAm) was studied by transmission electron microscopy (TEM), dynamic light scattering (DLS) and spectrofluorimeter. The critical micelle concentration (CMC) value was 0.052 mg mL(-1). These micelles have thermo-responsibility and a low critical solution temperature (LCST) of 33.5°C. Further investigation indicated that the guest molecule release property of these micelles, which can be well described by a first-order kinetic model, was significantly affected by temperature. Besides, the micelles exhibited excellent biocompatibility and cellular uptake property. Hence, these micelles are considered to have potential application in controlled drug delivery. Copyright © 2016 Elsevier B.V. All rights reserved.

Extension of BEPU methods to Sub-channel Thermal-Hydraulics and to Coupled Three-Dimensional Neutronics/Thermal-Hydraulics Codes

International Nuclear Information System (INIS)

Avramova, M.; Ivanov, K.; Arenas, C.

2013-01-01

The principles that support the risk-informed regulation are to be considered in an integrated decision-making process. Thus, any evaluation of licensing issues supported by a safety analysis would take into account both deterministic and probabilistic aspects of the problem. The deterministic aspects will be addressed using Best Estimate code calculations and considering the associated uncertainties i.e. Plus Uncertainty (BEPU) calculations. In recent years there has been an increasing demand from nuclear research, industry, safety and regulation for best estimate predictions to be provided with their confidence bounds. This applies also to the sub-channel thermal-hydraulic codes, which are used to evaluate local safety parameters. The paper discusses the extension of BEPU methods to the sub-channel thermal-hydraulic codes on the example of the Pennsylvania State University (PSU) version of COBRA-TF (CTF). The use of coupled codes supplemented with uncertainty analysis allows to avoid unnecessary penalties due to incoherent approximations in the traditional decoupled calculations, and to obtain more accurate evaluation of margins regarding licensing limit. This becomes important for licensing power upgrades, improved fuel assembly and control rod designs, higher burn-up and others issues related to operating LWRs as well as to the new Generation 3+ designs being licensed now (ESBWR, AP-1000, EPR-1600 and etc.). The paper presents the application of Generalized Perturbation Theory (GPT) to generate uncertainties associated with the few-group assembly homogenized neutron cross-section data used as input in coupled reactor core calculations. This is followed by a discussion of uncertainty propagation methodologies, being implemented by PSU in cooperation of Technical University of Catalonia (UPC) for reactor core calculations and for comprehensive multi-physics simulations. (authors)

Inert gas narcosis has no influence on thermo-tactile sensation.

Science.gov (United States)

Jakovljević, Miroljub; Vidmar, Gaj; Mekjavic, Igor B

2012-05-01

Contribution of skin thermal sensors under inert gas narcosis to the raising hypothermia is not known. Such information is vital for understanding the impact of narcosis on behavioural thermoregulation, diver safety and judgment of thermal (dis)comfort in the hyperbaric environment. So this study aimed at establishing the effects of normoxic concentration of 30% nitrous oxide (N(2)O) on thermo-tactile threshold sensation by studying 16 subjects [eight females and eight males; eight sensitive (S) and eight non-sensitive (NS) to N(2)O]. Their mean (SD) age was 22.1 (1.8) years, weight 72.8 (15.3) kg, height 1.75 (0.10) m and body mass index 23.8 (3.8) kg m(-2). Quantitative thermo-tactile sensory testing was performed on forearm, upper arm and thigh under two experimental conditions: breathing air (air trial) and breathing normoxic mixture of 30% N(2)O (N(2)O trial) in the mixed sequence. Difference in thermo-tactile sensitivity thresholds between two groups of subjects in two experimental conditions was analysed by 3-way mixed-model analysis of covariance. There were no statistically significant differences in thermo-tactile thresholds either between the Air and N(2)O trials, or between S and NS groups, or between females and males, or with respect to body mass index. Some clinically insignificant lowering of thermo-tactile thresholds occurred only for warm thermo-tactile thresholds on upper arm and thigh. The results indicated that normoxic mixture of 30% N(2)O had no influence on thermo-tactile sensation in normothermia.

A two-step approach for the preliminary evaluation of the thermal-hydraulics and safety of the ELSY open square core design

International Nuclear Information System (INIS)

Meloni, Paride; Bandini, Giacomino; Polidori, Massimiliano; Cervone, Antonio; Manservisi, Sandro

2009-01-01

Several innovative solutions for a liquid metal fast reactor design have been investigated in the EURATOM Sixth Framework Programme and an open-assembly core design for the ELSY (European Lead-cooled System) reactor has been proposed by ENEA. The development of this new reactor, based on innovative neutronic and safety considerations, requires a new approach to the thermal-hydraulic (T/H) core design. In this paper a new two-step approach of the T/H analysis for this open-assembly core is presented and, in particular is used for the evaluation of the preliminary core design of a 1500 MW lead fast reactor with open square lattice and three fuel radial zones with different levels of enrichment. In the first step a preliminary thermal-hydraulic and safety evaluation of the core neutronic design is investigated by using a one-dimensional RELAP5 model for independent channel analysis. Then two and three-dimensional effects are taken into account by using a dedicated tool for the evaluation of assembly mixing effects. The RELAP5 model, based on pressure loss and heat transfer correlations available for heavy liquid metal flows in rod bundle, consists of completely independent assemblies and therefore it can be used for a conservative evaluation of the thermal-hydraulics of the core reactor. Due to the open-lattice configuration, the two and three-dimensional effects are important and they are taken into account by using a simplified three-dimensional numerical model of an open square lattice reactor core, developed with the purpose of analyzing the whole core behavior. The numerical simulation is performed at assembly length level taking into account the local fluctuations of turbulent viscosity and energy exchange coefficients at sub-channel level through transfer operators based on parametric coefficients. A preliminary evaluation of the mixing effects between assembly flows on the temperature field has been performed by using an average assembly turbulent viscosity

Test Specifications and the Design of the Wire Wrapped 37-Pin Fuel Assembly for Hydrodynamic Experiments

International Nuclear Information System (INIS)

Chang, S. K.; Euh, D. J.; Bae, H.; Lee, H. Y.; Choi, S. R.

2013-01-01

Most influencing parameters on uncertainties and sensitivities of the CFD analyses are the friction coefficient and the mixing coefficient. The friction coefficient is related to the flow distribution in reactor sub-channels. The mixing coefficient is defined with the cross flow between neighboring sub-channels. The eventual purpose of the thermal hydraulic design considering these parameters is to guarantee the fuel cladding integrity as the design limit parameter. At the moment, the experimental program is being undertaken to quantify these friction and mixing parameters which characterize the flow distribution in sub-channels, and the wire wrapped 37-pin rod assembly and its hexagonal test rig have been designed and fabricated. The quantified thermal hydraulic experimental data from this program are utilized primarily to estimate the accuracy of the safety analysis codes and their thermal hydraulic model. A wire wrapped 37 pin fuel assembly has been designed for the measurements of the flow distribution, where the measurements are utilized to quantify the friction coefficient and the mixing coefficient. The test rig of the wire wrapped 37 pin fuel assembly has been fabricated considering the geometric and flow dynamic similarities. It comprises four components i. e., the upper plenum, the fuel housing, the lower plenum, and the wire wrapped 37 pin fuel assembly. At further works, the quantified friction and mixing coefficients through the experiments are going to be utilized for insuring the reliability of the CFD analysis results

The development of flow test technology for PWR fuel assembly

International Nuclear Information System (INIS)

Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

1988-05-01

KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

Novel thermo-responsive double-hydrophilic and hydrophobic MPEO-b-PEtOx-b-PCL triblock terpolymers: synthesis, characterization and self-assembly studies

Czech Academy of Sciences Publication Activity Database

Petrova, Svetlana; Venturini, Cristina Garcia; Jäger, Alessandro; Jäger, Eliezer; Černoch, Peter; Kereiche, S.; Kováčik, L.; Raška, I.; Štěpánek, Petr

2015-01-01

Roč. 59, 24 February (2015), s. 215-225 ISSN 0032-3861 R&D Projects: GA ČR GAP208/10/1600; GA MŠk(CZ) 7F14009 Institutional support: RVO:61389013 Keywords : MPEO-b-PEtOx-b-PCL triblock terpolymers * light-scattering * thermo-responsive nanoparticles Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 3.586, year: 2015

The upgrade of integrity analysis module and the mechanical behavior evaluation for the assembly duct

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Lee, Dong Uk; Kim, Young Il [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-05-01

The high neutron fluxes and operating temperatures associated with KALIMER are inducing the important radiation damage phenomena, which can cause significant dimensional changes in the core components of the reactor.The thermo-mechanical analysis of the assembly ducts for KALIMER are mainly performed to evaluate the following items.1) change of reactivity. 2) force at pads on core assemblies. 3) withdrawal force at refueling. 4) loading and refueling deviation of assembly ducts. 5) bowing modes for control assembly. In this report, the model for the evaluation of reactivity change as well as the refueling model and the withdrawl force model are upgraded. And the reactivity change is considered as the most important parameter among the above items. Therefore, the sensitivity analyses mainly associated with reactivity change are carried out. As the results, the pad gap between the assembly ducts preliminary driven for keeping the (-) reactivity change. 9 refs., 24 figs., 2 tabs. (Author)

Surface enhanced thermo lithography

KAUST Repository

Coluccio, Maria Laura

2017-01-13

We used electroless deposition to fabricate clusters of silver nanoparticles (NPs) on a silicon substrate. These clusters are plasmonics devices that induce giant electromagnetic (EM) field increments. When those EM field are absorbed by the metal NPs clusters generate, in turn, severe temperature increases. Here, we used the laser radiation of a conventional Raman set-up to transfer geometrical patterns from a template of metal NPs clusters into a layer of thermo sensitive Polyphthalaldehyde (PPA) polymer. Temperature profile on the devices depends on specific arrangements of silver nanoparticles. In plane temperature variations may be controlled with (i) high nano-meter spatial precision and (ii) single Kelvin temperature resolution on varying the shape, size and spacing of metal nanostructures. This scheme can be used to generate strongly localized heat amplifications for applications in nanotechnology, surface enhanced thermo-lithography (SETL), biology and medicine (for space resolved cell ablation and treatment), nano-chemistry.

Surface enhanced thermo lithography

KAUST Repository

Coluccio, Maria Laura; Alabastri, Alessandro; Bonanni, Simon; Majewska, Roksana; Dattoli, Elisabetta; Barberio, Marianna; Candeloro, Patrizio; Perozziello, Gerardo; Mollace, Vincenzo; Di Fabrizio, Enzo M.; Gentile, Francesco

2017-01-01

We used electroless deposition to fabricate clusters of silver nanoparticles (NPs) on a silicon substrate. These clusters are plasmonics devices that induce giant electromagnetic (EM) field increments. When those EM field are absorbed by the metal NPs clusters generate, in turn, severe temperature increases. Here, we used the laser radiation of a conventional Raman set-up to transfer geometrical patterns from a template of metal NPs clusters into a layer of thermo sensitive Polyphthalaldehyde (PPA) polymer. Temperature profile on the devices depends on specific arrangements of silver nanoparticles. In plane temperature variations may be controlled with (i) high nano-meter spatial precision and (ii) single Kelvin temperature resolution on varying the shape, size and spacing of metal nanostructures. This scheme can be used to generate strongly localized heat amplifications for applications in nanotechnology, surface enhanced thermo-lithography (SETL), biology and medicine (for space resolved cell ablation and treatment), nano-chemistry.

Thermo-responsive cell culture carrier: Effects on macrophage functionality and detachment efficiency.

Science.gov (United States)

Rennert, Knut; Nitschke, Mirko; Wallert, Maria; Keune, Natalie; Raasch, Martin; Lorkowski, Stefan; Mosig, Alexander S

2017-01-01

Harvesting cultivated macrophages for tissue engineering purposes by enzymatic digestion of cell adhesion molecules can potentially result in unintended activation, altered function, or behavior of these cells. Thermo-responsive polymer is a promising tool that allows for gentle macrophage detachment without artificial activation prior to subculture within engineered tissue constructs. We therefore characterized different species of thermo-responsive polymers for their suitability as cell substrate and to mediate gentle macrophage detachment by temperature shift. Primary human monocyte- and THP-1-derived macrophages were cultured on thermo-responsive polymers and characterized for phagocytosis and cytokine secretion in response to lipopolysaccharide stimulation. We found that both cell types differentially respond in dependence of culture and stimulation on thermo-responsive polymers. In contrast to THP-1 macrophages, primary monocyte-derived macrophages showed no signs of impaired viability, artificial activation, or altered functionality due to culture on thermo-responsive polymers compared to conventional cell culture. Our study demonstrates that along with commercially available UpCell carriers, two other thermo-responsive polymers based on poly(vinyl methyl ether) blends are attractive candidates for differentiation and gentle detachment of primary monocyte-derived macrophages. In summary, we observed similar functionality and viability of primary monocyte-derived macrophages cultured on thermo-responsive polymers compared to standard cell culture surfaces. While this first generation of custom-made thermo-responsive polymers does not yet outperform standard culture approaches, our results are very promising and provide the basis for exploiting the unique advantages offered by custom-made thermo-responsive polymers to further improve macrophage culture and recovery in the future, including the covalent binding of signaling molecules and the reduction of

Hydraulic turbines

International Nuclear Information System (INIS)

Meluk O, G.

1998-01-01

The hydraulic turbines are defined according to the specific speed, in impulse turbines and in reaction turbines. Currently, the Pelton turbines (of impulse) and the Francis and Kaplan turbines (of reaction), they are the most important machines in the hydroelectric generation. The hydraulic turbines are capable of generating in short times, large powers, from its loads zero until the total load and reject the load instantly without producing damages in the operation. When the hydraulic resources are important, the hydraulic turbines are converted in the axle of the electric system. Its combination with thermoelectric generation systems, it allow the continuing supply of the variations in demand of energy system. The available hydraulic resource in Colombia is of 93085 MW, of which solely 9% is exploited, become 79% of all the electrical country generation, 21% remaining is provided by means of the thermoelectric generation

Structural behaviour of fuel assemblies for water cooled reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2005-07-01

At the invitation of the Government of France and in response to a proposal of the IAEA Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT), the IAEA convened a Technical Meeting on Fuel Assembly Structural Behaviour in Cadarache, France, from 22 to 26 November 2004. The meeting was hosted by the CEA Cadarache Centre, AREVA Framatome-ANP and Electricite de France. The meeting aimed to provide in depth technical exchanges on PWR and WWER operational experience in the field of fuel assembly mechanical behaviour and the potential impact of future high burnup fuel management on fuel reliability. It addressed in-service experience and remedial solutions, loop testing experience, qualification and damage assessment methods (analytic or experimental ones), mechanical behaviour of the fuel assembly including dynamic and fluid structure interaction aspects, modelling and numerical analysis methods, and impact of the in-service evolution of the structural materials. Sixty-seven participants from 17 countries presented 30 papers in the course of four sessions. The topics covered included the impact of hydraulic loadings on fuel assembly (FA)performance, FA bow and control rod (CR) drop kinetics, vibrations and rod-to-grid wear and fretting, and, finally, evaluation and modelling of accident conditions, mainly from seismic causes. FA bow, CR drop kinetics and hydraulics are of great importance under conditions of higher fuel duties including burnup increase, thermal uprates and longer fuel cycles. Vibrations and rod-to-grid wear and fretting have been identified as a key cause of fuel failure at PWRs during the past several years. The meeting demonstrated that full-scale hydraulic tests and modelling provide sufficient information to develop remedies to increase FA skeleton resistance to hydraulic loads, including seismic ones, vibrations and wear. These proceedings are presented as a book with an attached CD-ROM. The first part of the CD

THERMOS, district central heating nuclear reactors

International Nuclear Information System (INIS)

Patarin, L.

1981-02-01

In order to expand the penetration of uranium in the national energy balance sheet, the C.E.A. has been studying nuclear reactors for several years now, that are capable of providing heat at favourable economic conditions. In this paper the THERMOS model is introduced. After showing the attraction of direct town heating by nuclear energy, the author describes the THERMOS project, defines the potential market, notably in France, and applies the lay-out study to the Grenoble Nuclear Study Centre site with district communal heating in mind. The economic aspects of the scheme are briefly mentioned [fr

Thermo-hydrodynamic and inductive modelling of a glass melt elaborated in cold inductive crucible

International Nuclear Information System (INIS)

Sauvage, E.

2009-11-01

Within the context of a search for a new vitrification process for nuclear wastes with a replacement of the presently used metallic pot by an inductive cold crucible, this research thesis deals with the numerical modelling of this technology. After having recalled the interest of nuclear waste vitrification, this report presents the new process based on the use of a cold crucible, describing principles and objectives of this method, and the characteristic physical phenomena associated with the flow and the thermodynamics of the glassy melt in such a crucible. It also recalls and comments the existing works on modelling. The main objective of this research is then to demonstrate the feasibility of 3D thermo-hydraulic and inductive simulations. He describes and analyses the glass physical properties (electrical properties, viscosity, thermal properties), the electromagnetic, hydrodynamic and thermal phenomena. He presents in detail the bubbling mixing modelling, reports 3D induction and fluid mechanical coupling calculations, and specific thermal investigations (radiating transfers, thermal limit conditions)

Preliminary verification of structure design for CN HCCB TBM with 1 × 4 configuration

Energy Technology Data Exchange (ETDEWEB)

Zhao, Zhou, E-mail: zhaozhou@swip.ac.cn; Zhou, Bing; Wang, Qijie; Cao, Qixiang; Feng, Kaiming; Wang, Xiaoyu; Zhang, Guoshu

2016-02-15

Highlights: • A new and simplification structural design scheme with 1 × 4 configuration is proposed for CN HCCB TBM. • The detail conceptual structural design for 1 × 4 TBM is completed. • The preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis for 1 × 4 TBM had been carried out. - Abstract: Based on the conceptual design of CN HCCB TBM with 1 × 4 configuration, the preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis had been carried out for it. Hydraulic and thermo-hydraulic analyses show that the coolant manifold system could meet the fluid design requirement preliminarily and the temperature of RAFMs structural parts, Be and Li{sub 4}SiO{sub 4} pebble beds are within the allowable range, and no zone shows a stress higher than the allowable limit in the preliminary structural analysis. These results indicate the design for CN HCCB TBM with 1 × 4 configuration is preliminary reasonable.

Preliminary verification of structure design for CN HCCB TBM with 1 × 4 configuration

International Nuclear Information System (INIS)

Zhao, Zhou; Zhou, Bing; Wang, Qijie; Cao, Qixiang; Feng, Kaiming; Wang, Xiaoyu; Zhang, Guoshu

2016-01-01

Highlights: • A new and simplification structural design scheme with 1 × 4 configuration is proposed for CN HCCB TBM. • The detail conceptual structural design for 1 × 4 TBM is completed. • The preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis for 1 × 4 TBM had been carried out. - Abstract: Based on the conceptual design of CN HCCB TBM with 1 × 4 configuration, the preliminary hydraulic analysis, thermo-hydraulic analysis and structural analysis had been carried out for it. Hydraulic and thermo-hydraulic analyses show that the coolant manifold system could meet the fluid design requirement preliminarily and the temperature of RAFMs structural parts, Be and Li_4SiO_4 pebble beds are within the allowable range, and no zone shows a stress higher than the allowable limit in the preliminary structural analysis. These results indicate the design for CN HCCB TBM with 1 × 4 configuration is preliminary reasonable.

Study on heat transfer and hydraulic model of spiral-fin fuel rods based on equivalent annulus method

International Nuclear Information System (INIS)

Zhang Dan; Liu Changwen; Lu Jianchao

2011-01-01

Tight lattice fuel assembly usually adopts spiral-fin fuel elements. Compared with the traditional PWR fuel rods, the closely packed and spiral fin spacers make the heat transfer and hydraulic phenomena in sub-channels very complicated, and: there was no suitable model and correlation to study it. This paper studied the effect of spiral spacers on the channel geometry in the equivalent annulus and physical performance based on the Rehme equivalent annulus methods, and the heat transfer of the spiral fin fuel rods and hydraulic model were obtained. The new model was verified with the traditional one, and the verification showed that two new models agreed well, which could provide certain theoretical explanation to the effect of the spiral spacer on the thermal hydraulics. (authors)

Spacer grid for a nuclear reactor fuel assembly

International Nuclear Information System (INIS)

Jabsen, F.S.

1978-01-01

The spacer grid consists of pairs of plates forming rectangular cells and enclosing the cylindrical fuel assemblies. They have got rigid as well as elastic projections extending into the cells and holding the fuel assemblies. Additional pairs of plates are arranged in about the center of the grid of plates. They have got only elastic projections extending on both sides of the plates into one cell each. This spacer grid may be used for reactor cores with and without fuel channels. By the combination of spring-elastic and rigid projections there is obtained a reinforced outer tie. Hydraulic pressure losses, parasitic neutron capture, and hot spots are essentially reduced. (DG) [de

Investigation research on the evaluation of a coupled thermo-hydro-mechanical-chemical phenomena. 4

International Nuclear Information System (INIS)

Ishihara, Yoshinao; Sagawa, Hiroshi; Matsuoka, Fushiki; Chijimatsu, Masakazu; Amemiya, Kiyoshi

2005-02-01

In order to realize a coupling analysis in the near field of the geological disposal system, the coupling analysis code 'COUPLYS (Coupling analysis system)' on the Thermo-Hydro-Mechanical-Chemical (THMC) phenomena by THAMES, Dtransu-3D·EL and PHREEQC, those are existing analysis code, is developed in this study. (1) We have introduced 8 nodes element for THAMES code in order to solve the coupled thermal, hydraulic and mechanical phenomena. Furthermore, in order to obtain the reliable resolution, each phenomenon is solved separately instead of full coupling. (2) In order to upgrade Dtransu-3D·EL model, we have introduced gas diffusion independent on aqueous element. (3) We have adopted surface site density for the bentonite depend on water content and CSH solid phase based on the ratio of C/S for cementitious material in the geochemistry module, and studied on the methodology of time mesh for kinetic model and separate method for pore water chemistry in the bentonite. (4) In order to develop THMC code, we have modified Multi p hreeqc to keep efficiency distributed processing for geochemical calculation and modified COUPLYS to calculate continuous treatment, and studied on the coupling module. After THAMES, Dtransu, PHREEQC and the hydraulic conductivity module were installed in COUPLYS, verification study was carried out to check basic function. (5) In order to ensure efficiency of analysis processor, we have developed supporting tool for graphic processor for THMC code and supporting tool of interpretation for geochemistry results. (author)

Modeling and stability of electro-hydraulic servo of hydraulic excavator

Science.gov (United States)

Jia, Wenhua; Yin, Chenbo; Li, Guo; Sun, Menghui

2017-11-01

The condition of the hydraulic excavator is complicated and the working environment is bad. The safety and stability of the control system is influenced by the external factors. This paper selects hydraulic excavator electro-hydraulic servo system as the research object. A mathematical model and simulation model using AMESIM of servo system is established. Then the pressure and flow characteristics are analyzed. The design and optimization of electro-hydraulic servo system and its application in engineering machinery is provided.

A Hydraulic Stress Measurement System for Deep Borehole Investigations

Science.gov (United States)

Ask, Maria; Ask, Daniel; Cornet, Francois; Nilsson, Tommy

2017-04-01

Luleå University of Technology (LTU) is developing and building a wire-line system for hydraulic rock stress measurements, with funding from the Swedish Research Council and Luleå University of Technology. In this project, LTU is collaborating with University of Strasbourg and Geosigma AB. The stress state influences drilling and drillability, as well as rock mass stability and permeability. Therefore, knowledge about the state of in-situ stress (stress magnitudes, and orientations) and its spatial variation with depth is essential for many underground rock engineering projects, for example for underground storage of hazardous material (e.g. nuclear waste, carbon dioxide), deep geothermal exploration, and underground infrastructure (e.g. tunneling, hydropower dams). The system is designed to conduct hydraulic stress testing in slim boreholes. There are three types of test methods: (1) hydraulic fracturing, (2) sleeve fracturing and (3) hydraulic testing of pre-existing fractures. These are robust methods for determining in situ stresses from boreholes. Integration of the three methods allows determination of the three-dimensional stress tensor and its spatial variation with depth in a scientific unambiguously way. The stress system is composed of a downhole and a surface unit. The downhole unit consists of hydraulic fracturing equipment (straddle packers and downhole imaging tool) and their associated data acquisition systems. The testing system is state of the art in several aspects including: (1) Large depth range (3 km), (2) Ability to test three borehole dimensions (N=76 mm, H=96 mm, and P=122 mm), (3) Resistivity imager maps the orientation of tested fracture; (4) Highly stiff and resistive to corrosion downhole testing equipment; and (5) Very detailed control on the injection flow rate and cumulative volume is obtained by a hydraulic injection pump with variable piston rate, and a highly sensitive flow-meter. At EGU General Assembly 2017, we would like to

The problems of calculation of heat transfer crisis in fuel assemblies of PW reactors based on modern versions of thermohydraulic codes

International Nuclear Information System (INIS)

Fialko, N.M.; Sharaevskij, G.I.; Sharaevskaya, E.I.; Babak, E.I.

2014-01-01

The article gives an analysis of the adequacy of computer software systems FASCICLE BM-DF and COBRA, which are designed to calculate the main parameters of the safety of water-cooled nuclear reactors. This calculation is based on determining the local thermal-hydraulic parameters of the flow of coolant in the fuel rod assembled elements. In this article introduced the results of the comparison of experiments performed to determine the distribution of the main thermal-hydraulic flow parameters characteristic of subchannels of fuel rod assembled elements with the data for calculating these parameters on the basis of declared computer codes. Particular attention is paid to the analysis of experimental and calculation data, by definition, burnout in rod fuel assembled elements

DYNREL - the reference calculation (coupled code utilization on analysis of RIA-transient)

International Nuclear Information System (INIS)

Strmensky, C.; Darilek, P.

2003-01-01

DYNREL is coupled code, comprising DYN3D and RELAP5 programs. The coupled code has been developed during four years. Now DYNREL is tested on selected RIA and thermo-hydraulic transient calculations. This material describes some results from selected RIA transient calculation (initiated by control rod movement). DYNREL modelled the whole nuclear reactors. The core is modeled as 313 or 349 independent thermo-hydraulic channels with 10 or 20 axial layers. Thermo-hydraulic part contains about 700 components that covered the six loops' model of nuclear power plant in detail. The calculated results are compared with DYN3D/M3, DYN3D/H1.1 results (Authors)

Thermo-economic evaluation and optimization of the thermo-chemical conversion of biomass into methanol

International Nuclear Information System (INIS)

Peduzzi, Emanuela; Tock, Laurence; Boissonnet, Guillaume; Maréchal, François

2013-01-01

In a carbon and resources constrained world, thermo-chemical conversion of lignocellulosic biomass into fuels and chemicals is regarded as a promising alternative to fossil resources derived products. Methanol is one potential product which can be used for the synthesis of various chemicals or as a fuel in fuel cells and internal combustion engines. This study focuses on the evaluation and optimization of the thermodynamic and economic performance of methanol production from biomass by applying process integration and optimization techniques. Results reveal the importance of the energy integration and in particular of the cogeneration of electricity for the efficient use of biomass. - Highlights: • A thermo-economic model for biomass conversion into methanol is developed. • Process integration and multi-objective optimization techniques are applied. • Results reveal the importance of energy integration for electricity co-generation

Preparation of thermo-responsive graft copolymer by using a novel macro-RAFT agent and its application for drug delivery

International Nuclear Information System (INIS)

Song, Cunfeng; Yu, Shirong; Liu, Cheng; Deng, Yuanming; Xu, Yiting; Chen, Xiaoling; Dai, Lizong

2016-01-01

A methodology to prepare thermo-responsive graft copolymer by using a novel macro-RAFT agent was proposed. The macro-RAFT agent with pendant dithioester (ZC(S)SR) was facilely prepared via the combination of RAFT polymerization and esterification reaction. By means of ZC(S)SR-initiated RAFT polymerization, the thermo-responsive graft copolymer consisting of poly(methyl methacrylate-co-hydroxylethyl methacrylate) (P(MMA-co-HEMA)) backbone and hydrophilic poly(N-isopropylacrylamide) (PNIPAAm) side chains was constructed through the “grafting from” approach. The chemical compositions and molecular weight distributions of the synthesized polymers were respectively characterized by "1H nuclear magnetic resonance ("1H NMR) and gel permeation chromatography (GPC). Self-assembly behavior of the amphiphilic graft copolymers (P(MMA-co-HEMA)-g-PNIPAAm) was studied by transmission electron microscopy (TEM), dynamic light scattering (DLS) and spectrofluorimeter. The critical micelle concentration (CMC) value was 0.052 mg mL"−"1. These micelles have thermo-responsibility and a low critical solution temperature (LCST) of 33.5 °C. Further investigation indicated that the guest molecule release property of these micelles, which can be well described by a first-order kinetic model, was significantly affected by temperature. Besides, the micelles exhibited excellent biocompatibility and cellular uptake property. Hence, these micelles are considered to have potential application in controlled drug delivery. - Highlights: • A novel macro-RAFT agent with ZC(S)SR was used for preparing graft copolymer. • P(MMA-co-HEMA)-g-PNIPAAm was successful prepared via the “grafting from” approach. • Thermo-responsibility of the P(MMA-co-HEMA)-g-PNIPAAm micelles was investigated. • The drug release behavior of the P(MMA-co-HEMA)-g-PNIPAAm micelles was studied. • These micelles exhibited excellent biocompatibility and cellular uptake property.

Preparation of thermo-responsive graft copolymer by using a novel macro-RAFT agent and its application for drug delivery

Energy Technology Data Exchange (ETDEWEB)

Song, Cunfeng; Yu, Shirong [Department of Materials Science and Engineering, College of Materials, Xiamen University, Xiamen 361005 (China); Liu, Cheng; Deng, Yuanming; Xu, Yiting [Department of Materials Science and Engineering, College of Materials, Xiamen University, Xiamen 361005 (China); Fujian Provincial Key Laboratory of Fire Retardant Materials, Xiamen University, Xiamen 361005 (China); Chen, Xiaoling, E-mail: tinachen0628@163.com [Department of Endodontics, Xiamen Stomatology Hospital, Teaching Hospital of Fujian Medical University, Xiamen 361003 (China); Dai, Lizong, E-mail: lzdai@xmu.edu.cn [Department of Materials Science and Engineering, College of Materials, Xiamen University, Xiamen 361005 (China); Fujian Provincial Key Laboratory of Fire Retardant Materials, Xiamen University, Xiamen 361005 (China)

2016-05-01

A methodology to prepare thermo-responsive graft copolymer by using a novel macro-RAFT agent was proposed. The macro-RAFT agent with pendant dithioester (ZC(S)SR) was facilely prepared via the combination of RAFT polymerization and esterification reaction. By means of ZC(S)SR-initiated RAFT polymerization, the thermo-responsive graft copolymer consisting of poly(methyl methacrylate-co-hydroxylethyl methacrylate) (P(MMA-co-HEMA)) backbone and hydrophilic poly(N-isopropylacrylamide) (PNIPAAm) side chains was constructed through the “grafting from” approach. The chemical compositions and molecular weight distributions of the synthesized polymers were respectively characterized by {sup 1}H nuclear magnetic resonance ({sup 1}H NMR) and gel permeation chromatography (GPC). Self-assembly behavior of the amphiphilic graft copolymers (P(MMA-co-HEMA)-g-PNIPAAm) was studied by transmission electron microscopy (TEM), dynamic light scattering (DLS) and spectrofluorimeter. The critical micelle concentration (CMC) value was 0.052 mg mL{sup −1}. These micelles have thermo-responsibility and a low critical solution temperature (LCST) of 33.5 °C. Further investigation indicated that the guest molecule release property of these micelles, which can be well described by a first-order kinetic model, was significantly affected by temperature. Besides, the micelles exhibited excellent biocompatibility and cellular uptake property. Hence, these micelles are considered to have potential application in controlled drug delivery. - Highlights: • A novel macro-RAFT agent with ZC(S)SR was used for preparing graft copolymer. • P(MMA-co-HEMA)-g-PNIPAAm was successful prepared via the “grafting from” approach. • Thermo-responsibility of the P(MMA-co-HEMA)-g-PNIPAAm micelles was investigated. • The drug release behavior of the P(MMA-co-HEMA)-g-PNIPAAm micelles was studied. • These micelles exhibited excellent biocompatibility and cellular uptake property.

Thermo-Plasmonics for Localized Graphitization and Welding of Polymeric Nanofibers

Directory of Open Access Journals (Sweden)

Ahnaf Usman Zillohu

2014-01-01

Full Text Available There is a growing interest in modulating the temperature under the illumination of light. As a heat source, metal nanoparticles (NPs have played an important role to pave the way for a new branch of plasmonics, i.e., thermo-plasmonics. While thermo-plasmonics have been well established in photo-thermal therapy, it has received comparatively less attention in materials science and chemistry. Here, we demonstrate the first proof of concept experiment of local chemistry and graphitization of metalized polymeric nanofibers through thermo-plasmonic effect. In particular, by tuning the plasmonic absorption of the nanohybrid through a change in the thickness of the deposited silver film on the fibers, the thermo-plasmonic effect can be adjusted in such a way that high enough temperature is generated enabling local welding and graphitization of the polymeric nanofibers.

Thermo-chemical pretreatment and enzymatic hydrolysis for enhancing saccharification of catalpa sawdust.

Science.gov (United States)

Jin, Shuguang; Zhang, Guangming; Zhang, Panyue; Li, Fan; Fan, Shiyang; Li, Juan

2016-04-01

To improve the reducing sugar production from catalpa sawdust, thermo-chemical pretreatments were examined and the chemicals used including NaOH, Ca(OH)2, H2SO4, and HCl. The hemicellulose solubilization and cellulose crystallinity index (CrI) were significantly increased after thermo-alkaline pretreatments, and the thermo-Ca(OH)2 pretreatment showed the best improvement for reducing sugar production comparing to other three pretreatments. The conditions of thermo-Ca(OH)2 pretreatment and enzymatic hydrolysis were systematically optimized. Under the optimal conditions, the reducing sugar yield increased by 1185.7% comparing to the control. This study indicates that the thermo-Ca(OH)2 pretreatment is ideal for the saccharification of catalpa sawdust and that catalpa sawdust is a promising raw material for biofuel. Copyright © 2016 Elsevier Ltd. All rights reserved.

Hydraulic design of Three Gorges right bank powerhouse turbine for improvement of hydraulic stability

International Nuclear Information System (INIS)

Shi, Q

2010-01-01

This paper presents the hydraulic design of Three Gorges Right Bank Powerhouse turbine for improvement of hydraulic stability. The technical challenges faced in the hydraulic design of the turbine are given. The method of hydraulic design for improving the hydraulic stability and particularly for eliminating the upper part load pressure pulsations is clarified. The final hydraulic design results of Three Gorges Right Bank Powerhouse turbine based on modern hydraulic design techniques are presented.

Hydraulic design of Three Gorges right bank powerhouse turbine for improvement of hydraulic stability

Energy Technology Data Exchange (ETDEWEB)

Shi, Q, E-mail: qhshi@dfem.com.c [Dong Fang Electrical Machinery Co., Ltd., DEC 188, Huanghe West Road, Deyang, 618000 (China)

2010-08-15

This paper presents the hydraulic design of Three Gorges Right Bank Powerhouse turbine for improvement of hydraulic stability. The technical challenges faced in the hydraulic design of the turbine are given. The method of hydraulic design for improving the hydraulic stability and particularly for eliminating the upper part load pressure pulsations is clarified. The final hydraulic design results of Three Gorges Right Bank Powerhouse turbine based on modern hydraulic design techniques are presented.

Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

International Nuclear Information System (INIS)

Vaiana, F.

2009-11-01

This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

Calculation and analysis of thermal–hydraulics fluctuations in pressurized water reactors

International Nuclear Information System (INIS)

Malmir, Hessam; Vosoughi, Naser

2015-01-01

Highlights: • Single-phase thermal–hydraulics noise equations are originally derived in the frequency domain. • The fluctuations of all the coolant parameters are calculated, without any simplifying assumptions. • The radial distribution of the temperature fluctuations in the fuel, gap and cladding are taken into account. • The closed-loop calculations are performed by means of the point kinetics noise theory. • Both the space- and frequency-dependence of the thermal–hydraulics fluctuations are analyzed. - Abstract: Analysis of thermal–hydraulics fluctuations in pressurized water reactors (e.g., local and global temperature or density fluctuations, as well as primary and charging pumps fluctuations) has various applications in calculation or measurement of the core dynamical parameters (temperature or density reactivity coefficients) in addition to thermal–hydraulics surveillance and diagnostics. In this paper, the thermal–hydraulics fluctuations in PWRs are investigated. At first, the single-phase thermal–hydraulics noise equations (in the frequency domain) are originally derived, without any simplifying assumptions. The fluctuations of all the coolant parameters, as well as the radial distribution of the temperature fluctuations in the fuel, gap and cladding are taken into account. Then, the derived governing equations are discretized using the finite volume method (FVM). Based on the discretized equations and the proposed algorithm of solving, a single heated channel noise calculation code (SHC-Noise) is developed, by which the steady-state and fluctuating parameters of PWR fuel assemblies can be calculated. The noise sources include the inlet coolant temperature and velocity fluctuations, in addition to the power density noises. The developed SHC-Noise code is benchmarked in different cases and scenarios. Furthermore, to show the effects of the power feedbacks, the closed-loop calculations are performed by means of the point kinetics noise

Thermo-Fluid Dynamics of Two-Phase Flow

CERN Document Server

Ishii, Mamrou

2011-01-01

"Thermo-fluid Dynamics of Two-Phase Flow, Second Edition" is focused on the fundamental physics of two-phase flow. The authors present the detailed theoretical foundation of multi-phase flow thermo-fluid dynamics as they apply to: Nuclear reactor transient and accident analysis; Energy systems; Power generation systems; Chemical reactors and process systems; Space propulsion; Transport processes. This edition features updates on two-phase flow formulation and constitutive equations and CFD simulation codes such as FLUENT and CFX, new coverage of the lift force model, which is of part

What Controls Thermo-osmosis? Molecular Simulations Show the Critical Role of Interfacial Hydrodynamics

Science.gov (United States)

Fu, Li; Merabia, Samy; Joly, Laurent

2017-11-01

Thermo-osmotic and related thermophoretic phenomena can be found in many situations from biology to colloid science, but the underlying molecular mechanisms remain largely unexplored. Using molecular dynamics simulations, we measure the thermo-osmosis coefficient by both mechanocaloric and thermo-osmotic routes, for different solid-liquid interfacial energies. The simulations reveal, in particular, the crucial role of nanoscale interfacial hydrodynamics. For nonwetting surfaces, thermo-osmotic transport is largely amplified by hydrodynamic slip at the interface. For wetting surfaces, the position of the hydrodynamic shear plane plays a key role in determining the amplitude and sign of the thermo-osmosis coefficient. Finally, we measure a giant thermo-osmotic response of the water-graphene interface, which we relate to the very low interfacial friction displayed by this system. These results open new perspectives for the design of efficient functional interfaces for, e.g., waste-heat harvesting.

What Controls Thermo-osmosis? Molecular Simulations Show the Critical Role of Interfacial Hydrodynamics.

Science.gov (United States)

Fu, Li; Merabia, Samy; Joly, Laurent

2017-11-24

Thermo-osmotic and related thermophoretic phenomena can be found in many situations from biology to colloid science, but the underlying molecular mechanisms remain largely unexplored. Using molecular dynamics simulations, we measure the thermo-osmosis coefficient by both mechanocaloric and thermo-osmotic routes, for different solid-liquid interfacial energies. The simulations reveal, in particular, the crucial role of nanoscale interfacial hydrodynamics. For nonwetting surfaces, thermo-osmotic transport is largely amplified by hydrodynamic slip at the interface. For wetting surfaces, the position of the hydrodynamic shear plane plays a key role in determining the amplitude and sign of the thermo-osmosis coefficient. Finally, we measure a giant thermo-osmotic response of the water-graphene interface, which we relate to the very low interfacial friction displayed by this system. These results open new perspectives for the design of efficient functional interfaces for, e.g., waste-heat harvesting.

Nuclear materials thermo-physical property database and property analysis using the database

International Nuclear Information System (INIS)

Jeong, Yeong Seok

2002-02-01

It is necessary that thermo-physical properties and understand of nuclear materials for evaluation and analysis to steady and accident states of commercial and research reactor. In this study, development of nuclear materials thermo-properties database and home page. In application of this database, it is analyzed of thermal conductivity, heat capacity, enthalpy, and linear thermal expansion of fuel and cladding material and compared thermo-properties model in nuclear fuel performance evaluation codes with experimental data in database. Results of compare thermo-property model of UO 2 fuel and cladding major performance evaluation code, both are similar

Experimental Study of Hydraulic Control Rod Drive Mechanism for Passive IN-core Cooling System of Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol [KAERI, Daejeon (Korea, Republic of)

2016-05-15

CAREM 25 (27 MWe safety systems using hydraulic control rod drives (CRD) studied critical issues that were rod drops with interrupted flow [3]. Hydraulic control rod drive suggested fast shutdown condition using a large gap between piston and cylinder in order to fast drop of neutron absorbing rods. A Passive IN-core Cooling system (PINCs) was suggested for safety enhancement of pressurized water reactors (PWR), small modular reactor (SMR), sodium fast reactor (SFR) in UNIST. PINCs consist of hydraulic control rod drive mechanism (Hydraulic CRDM) and hybrid control rod assembly with heat pipe combined with control rod. The schematic diagram of the hydraulic CRDM for PINCs is shown in Fig. 1. The experimental results show the steady state and transient behavior of the upper cylinder at a low pressure and low temperature. The influence of the working fluid temperature and cylinder mass are investigated. Finally, the heat removal between evaporator section and condenser section is compared with or without the hybrid control rod. Heat removal test of the hybrid heat pipe with hydraulic CRDM system showed the heat transfer coefficient of the bundle hybrid control rod and its effect on evaporator pool. The preliminary test both hydraulic CRDM and heat removal system was conducted, which showed the possibility of the in-core hydraulic drive system for application of PINCs.

Investigation of Surface Pre-Treatment Methods for Wafer-Level Cu-Cu Thermo-Compression Bonding

Directory of Open Access Journals (Sweden)

Koki Tanaka

2016-12-01

Full Text Available To increase the yield of the wafer-level Cu-Cu thermo-compression bonding method, certain surface pre-treatment methods for Cu are studied which can be exposed to the atmosphere before bonding. To inhibit re-oxidation under atmospheric conditions, the reduced pure Cu surface is treated by H2/Ar plasma, NH3 plasma and thiol solution, respectively, and is covered by Cu hydride, Cu nitride and a self-assembled monolayer (SAM accordingly. A pair of the treated wafers is then bonded by the thermo-compression bonding method, and evaluated by the tensile test. Results show that the bond strengths of the wafers treated by NH3 plasma and SAM are not sufficient due to the remaining surface protection layers such as Cu nitride and SAMs resulting from the pre-treatment. In contrast, the H2/Ar plasma–treated wafer showed the same strength as the one with formic acid vapor treatment, even when exposed to the atmosphere for 30 min. In the thermal desorption spectroscopy (TDS measurement of the H2/Ar plasma–treated Cu sample, the total number of the detected H2 was 3.1 times more than the citric acid–treated one. Results of the TDS measurement indicate that the modified Cu surface is terminated by chemisorbed hydrogen atoms, which leads to high bonding strength.

Using statistical sensitivities for adaptation of a best-estimate thermo-hydraulic simulation model

International Nuclear Information System (INIS)

Liu, X.J.; Kerner, A.; Schaefer, A.

2010-01-01

On-line adaptation of best-estimate simulations of NPP behaviour to time-dependent measurement data can be used to insure that simulations performed in parallel to plant operation develop synchronously with the real plant behaviour even over extended periods of time. This opens a range of applications including operator support in non-standard-situations, improving diagnostics and validation of measurements in real plants or experimental facilities. A number of adaptation methods have been proposed and successfully applied to control problems. However, these methods are difficult to be applied to best-estimate thermal-hydraulic codes, such as TRACE and ATHLET, with their large nonlinear differential equation systems and sophisticated time integration techniques. This paper presents techniques to use statistical sensitivity measures to overcome those problems by reducing the number of parameters subject to adaptation. It describes how to identify the most significant parameters for adaptation and how this information can be used by combining: -decomposition techniques splitting the system into a small set of component parts with clearly defined interfaces where boundary conditions can be derived from the measurement data, -filtering techniques to insure that the time frame for adaptation is meaningful, -numerical sensitivities to find minimal error conditions. The suitability of combining those techniques is shown by application to an adaptive simulation of the PKL experiment.

Thermohydraulics of a horizontal diphasic flow of superfluid helium; Thermo-hydraulique d'un ecoulement horizontal d'helium superfluide diphasique

Energy Technology Data Exchange (ETDEWEB)

Perraud, S

2007-12-15

This study aims at characterizing helium two phase flows, and to identify the dependence of their characteristics on various thermo-hydraulic parameters: vapour velocity, liquid height, vapour density, specificities of superfluidity. Both the engineer and the physicist's points of view are taken into consideration: the first one in terms of optimization of a particular cooling scheme based on a two-phase flow, and these second one in terms of more fundamental atomization-related questions. It has been shown that for velocities around 3 to 4 m/s, the liquid phase that was initially stratified undergoes an atomization through the presence of a drop haze carried by the vapor phase.This happens for superfluid helium as well as for normal helium without main differences on atomization.

Thermal-hydraulics verification of a coarse-mesh OpenFOAM-based solver for a Sodium Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Bonet López, M.

2015-07-01

Recently, in the Institute Swiss Paul Scherrer Institut, is has developed a platform Multiphysics, based in OpenFOAM, that is capable of performing an analysis multidimensional of a reactor nuclear. One of the main objectives of this project is to verify the part of the code responsible for the Thermo-hydraulic analysis of the reactor. To carry out simulations this part of the code uses the approximation of thick mesh based on the equations of a porous medium. Therefore, the other objective is demonstrate that this method is applicable to the analysis of a reactor nuclear fast of sodium, focusing is in his capacity of predict the transfer of heat between a subset and the space vacuum between subsets of the core of the reactor. (Author)

Improving Thermo-Oxidative Stability of Nitrile Rubber Composites by Functional Graphene Oxide.

Science.gov (United States)

Zhong, Rui; Zhang, Zhao; Zhao, Hongguo; He, Xianru; Wang, Xin; Zhang, Rui

2018-05-30

Graphene oxide (GO), modified with anti-aging agent p -phenylenediamine (PPD), was added into nitrile rubber (NBR) in order to improve the thermo-oxidative stability of NBR. The modification of GO and the transformation of functional groups were characterized by Fourier transform infrared spectroscopy (FTIR), Raman, and X-ray diffraction (XRD). Mechanical performances of NBR composites before and after the thermo-oxidative aging were recorded. The results of dynamic mechanical analysis (DMA) show an increased storage modulus (G') and a decreased value of area of tan δ peak after introducing modified GO into NBR. It indicates that filler particles show positive interaction with molecular chains. The thermo-oxidative stability of composites was investigated by thermogravimetric analysis (TG) and differential scanning calorimetry (DSC). Then, the thermo-oxidative aging kinetic parameters were obtained by the Flynn⁻Wall⁻Ozawa (FWO) equation. The results of aging tests show that the thermo-oxidative stability of rubber matrix increases obviously after introducing GO⁻PPD. In addition, mechanical properties (tensile strength and elongation at break) of both before and after aged NBR/GO⁻PPD composites were superior to that of NBR. This work provides meaningful guidance for achieving multifunction thermo-oxidative aging resistance rubber composites.

Thermally Actuated Hydraulic Pumps

Science.gov (United States)

Jones, Jack; Ross, Ronald; Chao, Yi

2008-01-01

Thermally actuated hydraulic pumps have been proposed for diverse applications in which direct electrical or mechanical actuation is undesirable and the relative slowness of thermal actuation can be tolerated. The proposed pumps would not contain any sliding (wearing) parts in their compressors and, hence, could have long operational lifetimes. The basic principle of a pump according to the proposal is to utilize the thermal expansion and contraction of a wax or other phase-change material in contact with a hydraulic fluid in a rigid chamber. Heating the chamber and its contents from below to above the melting temperature of the phase-change material would cause the material to expand significantly, thus causing a substantial increase in hydraulic pressure and/or a substantial displacement of hydraulic fluid out of the chamber. Similarly, cooling the chamber and its contents from above to below the melting temperature of the phase-change material would cause the material to contract significantly, thus causing a substantial decrease in hydraulic pressure and/or a substantial displacement of hydraulic fluid into the chamber. The displacement of the hydraulic fluid could be used to drive a piston. The figure illustrates a simple example of a hydraulic jack driven by a thermally actuated hydraulic pump. The pump chamber would be a cylinder containing encapsulated wax pellets and containing radial fins to facilitate transfer of heat to and from the wax. The plastic encapsulation would serve as an oil/wax barrier and the remaining interior space could be filled with hydraulic oil. A filter would retain the encapsulated wax particles in the pump chamber while allowing the hydraulic oil to flow into and out of the chamber. In one important class of potential applications, thermally actuated hydraulic pumps, exploiting vertical ocean temperature gradients for heating and cooling as needed, would be used to vary hydraulic pressures to control buoyancy in undersea research

Contribution at the turbulence study in tridimensional flow: application at nuclear reactors

International Nuclear Information System (INIS)

Rodet, J.C.

1985-01-01

Inside hexagonal assemblies of 3 or 19 pins of nuclear reactor type Super Phenix parietal pressure ranges, mean kinematical ranges and turbulent ranges has been measured. For these latest a measurement methodology taking into account the orientation constraints binded at the geometry of assembly is developed for a simple probe having two crossed hot wires. The mean kinematic range obtained is tridimensional; it puts in clearness periodicities, a peripheral flow along the case and render an account of kink values. Local analysis of Reynold tensors permits to return localy and in selected references at tangential bidimensional situations. Evaluations of mixing lenghts are executed in subchannels, then integrated in a numerical code of thermo hydraulic prediction with object to improve it [fr

Thermo-mechanical analysis of PWR bolts susceptible to IASCC

International Nuclear Information System (INIS)

Matteoli, C.; Hannink, M.H.C.; Blom, F.J.; Marck, S.C. van der; Charpin-Jacobs, F.

2015-01-01

Irradiation Assisted Stress Corrosion Cracking (IASCC) is considered a primary ageing issue for the Reactor Pressure Vessel (RPV) internals of Pressurized Water Reactors (PWR). In particular, this complex phenomenon which develops in an environment featuring thermal and mechanical stresses, interaction with corrosive compounds and irradiation, is affecting the bolts connecting the baffles and the formers in the Nuclear Power Plants' RPVs. The baffle-former assembly is the structure that borders the fuel assemblies region, contributing to keep them in position and separating in the radial direction, the core region from the downcomer region. An evaluation of the stresses and temperatures reached in the baffle-former bolts during normal operation was performed by means of a coupled thermo-mechanical study which uses reactor physics calculations to obtain the fluence in the reactor core and as a consequence the heat deposition in the RPV internals. The heat deposition data are coupled with a finite element model of the bolts and the RPV internals in order to perform a complete analysis taking in account thermal, mechanical and radiation loadings. The study is first carried out focusing on a section of the RPV internals, showing a single row of baffle-former bolts. Then the work is extended to the full core height. The model set up in this work, includes an in-depth study of the behavior of the core internals, in particular baffle-former bolts. The model has the capability of understanding the mechanical and thermal behavior of essential internal components in a PWR. (authors)

Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

International Nuclear Information System (INIS)

Onozuka, M.; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H.

2001-01-01

Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region

Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Onozuka, M. E-mail: onozukm@itereu.de; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H

2001-11-01

Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region.

Magnetic Field and Torque Output of Packaged Hydraulic Torque Motor

Directory of Open Access Journals (Sweden)

Liang Yan

2018-01-01

Full Text Available Hydraulic torque motors are one key component in electro-hydraulic servo valves that convert the electrical signal into mechanical motions. The systematic characteristics analysis of the hydraulic torque motor has not been found in the previous research, including the distribution of the electromagnetic field and torque output, and particularly the relationship between them. In addition, conventional studies of hydraulic torque motors generally assume an evenly distributed magnetic flux field and ignore the influence of special mechanical geometry in the air gaps, which may compromise the accuracy of analyzing the result and the high-precision motion control performance. Therefore, the objective of this study is to conduct a detailed analysis of the distribution of the magnetic field and torque output; the influence of limiting holes in the air gaps is considered to improve the accuracy of both numerical computation and analytical modeling. The structure and working principle of the torque motor are presented first. The magnetic field distribution in the air gaps and the magnetic saturation in the iron blocks are analyzed by using a numerical approach. Subsequently, the torque generation with respect to the current input and assembly errors is analyzed in detail. This shows that the influence of limiting holes on the magnetic field is consistent with that on torque generation. Following this, a novel modified equivalent magnetic circuit is proposed to formulate the torque output of the hydraulic torque motor analytically. The comparison among the modified equivalent magnetic circuit, the conventional modeling approach and the numerical computation is conducted, and it is found that the proposed method helps to improve the modeling accuracy by taking into account the effect of special geometry inside the air gaps.

Thermo-responsive magnetic liposomes for hyperthermia-triggered local drug delivery.

Science.gov (United States)

Dai, Min; Wu, Cong; Fang, Hong-Ming; Li, Li; Yan, Jia-Bao; Zeng, Dan-Lin; Zou, Tao

2017-06-01

We prepared and characterised thermo-responsive magnetic liposomes, which were designed to combine features of magnetic targeting and thermo-responsive control release for hyperthermia-triggered local drug delivery. The particle size and zeta-potential of the thermo-responsive magnetic ammonium bicarbonate (MagABC) liposomes were about 210 nm and -14 mV, respectively. The MagABC liposomes showed encapsulation efficiencies of about 15% and 82% for magnetic nanoparticles (mean crystallite size 12 nm) and doxorubicin (DOX), respectively. The morphology of the MagABC liposomes was visualised using transmission electron microscope (TEM). The MagABC liposomes showed desired thermo-responsive release. The MagABC liposomes, when physically targeted to tumour cells in culture by a permanent magnetic field yielded a substantial increase in intracellular accumulation of DOX as compared to non-magnetic ammonium bicarbonate (ABC) liposomes. This resulted in a parallel increase in cytotoxicity for DOX loaded MagABC liposomes over DOX loaded ABC liposomes in tumour cells.

Construction of database server system for fuel thermo-physical properties

International Nuclear Information System (INIS)

Park, Chang Je; Kang, Kwon Ho; Song, Kee Chan

2003-12-01

To perform the evaluation of various fuels in the nuclear reactors, not only the mechanical properties but also thermo-physical properties are required as one of most important inputs for fuel performance code system. The main objective of this study is to make a database system for fuel thermo-physical properties and a PC-based hardware system has been constructed for ease use for the public with visualization such as web-based server system. This report deals with the hardware and software which are used in the database server system for nuclear fuel thermo-physical properties. It is expected to be highly useful to obtain nuclear fuel data without such a difficulty through opening the database of fuel properties to the public and is also helpful to research of development of various fuel of nuclear industry. Furthermore, the proposed models of nuclear fuel thermo-physical properties will be enough utilized to the fuel performance code system

Analytical Expressions for Thermo-Osmotic Permeability of Clays

Science.gov (United States)

Gonçalvès, J.; Ji Yu, C.; Matray, J.-M.; Tremosa, J.

2018-01-01

In this study, a new formulation for the thermo-osmotic permeability of natural pore solutions containing monovalent and divalent cations is proposed. The mathematical formulation proposed here is based on the theoretical framework supporting thermo-osmosis which relies on water structure alteration in the pore space of surface-charged materials caused by solid-fluid electrochemical interactions. The ionic content balancing the surface charge of clay minerals causes a disruption in the hydrogen bond network when more structured water is present at the clay surface. Analytical expressions based on our heuristic model are proposed and compared to the available data for NaCl solutions. It is shown that the introduction of divalent cations reduces the thermo-osmotic permeability by one third compared to the monovalent case. The analytical expressions provided here can be used to advantage for safety calculations in deep underground nuclear waste repositories.

Fiber Optic Thermo-Hygrometers for Soil Moisture Monitoring.

Science.gov (United States)

Leone, Marco; Principe, Sofia; Consales, Marco; Parente, Roberto; Laudati, Armando; Caliro, Stefano; Cutolo, Antonello; Cusano, Andrea

2017-06-20

This work deals with the fabrication, prototyping, and experimental validation of a fiber optic thermo-hygrometer-based soil moisture sensor, useful for rainfall-induced landslide prevention applications. In particular, we recently proposed a new generation of fiber Bragg grating (FBGs)-based soil moisture sensors for irrigation purposes. This device was realized by integrating, inside a customized aluminum protection package, a FBG thermo-hygrometer with a polymer micro-porous membrane. Here, we first verify the limitations, in terms of the volumetric water content (VWC) measuring range, of this first version of the soil moisture sensor for its exploitation in landslide prevention applications. Successively, we present the development, prototyping, and experimental validation of a novel, optimized version of a soil VWC sensor, still based on a FBG thermo-hygrometer, but able to reliably monitor, continuously and in real-time, VWC values up to 37% when buried in the soil.

Multi-scale modeling of the thermo-hydro- mechanical behaviour of heterogeneous materials. Application to cement-based materials under severe loads

International Nuclear Information System (INIS)

Grondin, Frederic Alain

2005-01-01

The work of modeling presented here relates to the study of the thermo-hydro- mechanical behaviour of porous materials based on hydraulic binder such as concrete, High Performance Concrete or more generally cement-based materials. This work is based on the exploitation of the Digital Concrete model, of the finite element code Symphonie developed in the Scientific and Technical Centre for Building (CSTB), in coupling with the homogenization methods to obtain macroscopic behaviour laws drawn from the Micro-Macro relations. Scales of investigation, macroscopic and microscopic, has been exploited by simulation in order to allow the comprehension fine of the behaviour of cement-based materials according to thermal, hydrous and mechanical loads. It appears necessary to take into account various scales of modeling. In order to study the behaviour of the structure, we are brought to reduce the scale of investigation to study the material more particularly. The research tasks presented suggest a new approach for the identification of the multi-physic behaviour of materials by simulation. In complement of the purely experimental approach, based on observations on the sample with measurements of the apparent parameters on the macroscopic scale, this new approach allows to obtain the fine analysis of elementary mechanisms in acting within the material. These elementary mechanisms are at the origin of the evolution of the macroscopic parameters measured in experimental tests. In this work, coefficients of the thermo-hydro-mechanical behaviour law of porous materials and the equivalent hydraulic conductivity were obtained by a multi-scales approach. Applications has been carried out on the study of the damaged behaviour of cement-based materials, in the objective to determine the elasticity tensor and the permeability tensor of a High Performance Concrete at high temperatures under a mechanical load. Also, the study of the strain evolution of cement-based materials at low

Mine drivage in hydraulic mines

Energy Technology Data Exchange (ETDEWEB)

Ehkber, B Ya

1983-09-01

From 20 to 25% of labor cost in hydraulic coal mines falls on mine drivage. Range of mine drivage is high due to the large number of shortwalls mined by hydraulic monitors. Reducing mining cost in hydraulic mines depends on lowering drivage cost by use of new drivage systems or by increasing efficiency of drivage systems used at present. The following drivage methods used in hydraulic mines are compared: heading machines with hydraulic haulage of cut rocks and coal, hydraulic monitors with hydraulic haulage, drilling and blasting with hydraulic haulage of blasted rocks. Mining and geologic conditions which influence selection of the optimum mine drivage system are analyzed. Standardized cross sections of mine roadways driven by the 3 methods are shown in schemes. Support systems used in mine roadways are compared: timber supports, roof bolts, roof bolts with steel elements, and roadways driven in rocks without a support system. Heading machines (K-56MG, GPKG, 4PU, PK-3M) and hydraulic monitors (GMDTs-3M, 12GD-2) used for mine drivage are described. Data on mine drivage in hydraulic coal mines in the Kuzbass are discussed. From 40 to 46% of roadways are driven by heading machines with hydraulic haulage and from 12 to 15% by hydraulic monitors with hydraulic haulage.

Basic hydraulics

CERN Document Server

Smith, P D

1982-01-01

BASIC Hydraulics aims to help students both to become proficient in the BASIC programming language by actually using the language in an important field of engineering and to use computing as a means of mastering the subject of hydraulics. The book begins with a summary of the technique of computing in BASIC together with comments and listing of the main commands and statements. Subsequent chapters introduce the fundamental concepts and appropriate governing equations. Topics covered include principles of fluid mechanics; flow in pipes, pipe networks and open channels; hydraulic machinery;

Complex investigation of thermo-technical parameters of Ruskov andesite

Directory of Open Access Journals (Sweden)

František Krepelka

2006-12-01

Full Text Available The research of thermo-technical parameters of Ruskov andesite was made as a part of the complex research of its properties as well as of rock disintegration by the action of chemical flame on the rock surface, i.e. thermal spalling in particular. Thermal spalling is a process in which thermal stresses are induced in the surface layer of rock whose surface is thereby disintegrated into small parts, the so called spalls, by the brittle manner. The evaluation of thermo-technical properties of the studied rocks is necessary for the qualification and quantification of the thermal spalling process. The measured and evaluated parameters were the coefficient of linear thermal expansion, the coefficient of thermal conductivity, the specific heat capacity and the coefficient of thermal diffusivity. Andesite from the Ruskov locality was chosen as a basic experimental material for the investigation of thermal spalling upon preliminary experiments. The estimated thermo-technical parameters were analyzed regarding the application of thermal spalling for the disintegration of the Ruskov andesite. The outcome as that the values of determine thermo-technical parameters established an expectation for its successful application.

Comprehensive thermal-hydraulic and thermal-mechanical analysis of core and fuel rods for the safety validation of real refueling at the Kozloduy WWER-440

Energy Technology Data Exchange (ETDEWEB)

Stefanova, S; Panajotov, D; Ilieva, B; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1996-12-31

Safety analysis aimed at determination of thermal-hydraulic and thermal-mechanical margins of core and fuel rods has been carried out using computer codes COBSOFM and PIN-micro. Thermal-hydraulic calculations for the part of the core with maximum heat flux during steady-state regime show that the coolant, cladding and fuel temperatures are within the design limits. A severe accident with reactor blackout has been simulated. It is found that at 95% probability level there is no boiling crisis anywhere in the core. The thermal-mechanical parameters of working assembly fuel rod with maximum load have been calculated. The assembly linear power reached a maximum of 25 kW/m during the second fuel cycle, the fuel temperature remaining well below 1000{sup o} C. As the fuel assembly with typical power history has enough safety margins, it was proposed to use it for one more cycle. 4 refs., 12 figs.

Multi-sector thermo-physiological head simulator for headgear research

Science.gov (United States)

Martinez, Natividad; Psikuta, Agnes; Corberán, José Miguel; Rossi, René M.; Annaheim, Simon

2017-02-01

A novel thermo-physiological human head simulator for headgear testing was developed by coupling a thermal head manikin with a thermo-physiological model. As the heat flux at head-site is directly measured by the head manikin, this method provides a realistic quantification of the heat transfer phenomena occurring in the headgear, such as moisture absorption-desorption cycles, condensation, or moisture migration across clothing layers. Before coupling, the opportunities of the head manikin for representing the human physiology were evaluated separately. The evaluation revealed reduced precision in forehead and face temperature predictions under extreme heterogeneous temperature distributions and no initial limitation for simulating temperature changes observed in the human physiology. The thermo-physiological model predicted higher sweat rates when applied for coupled than for pure virtual simulations. After coupling, the thermo-physiological human head simulator was validated using eight human experiments. It precisely predicted core, mean skin, and forehead temperatures with average rmsd values within the average experimental standard deviation (rmsd of 0.20 ± 0.15, 0.83 ± 0.34, and 1.04 ± 0.54 °C, respectively). However, in case of forehead, precision was lower for the exposures including activity than for the sedentary exposures. The representation of the human sweat evaporation could be affected by a reduced evaporation efficiency and the manikin sweat dynamics. The industry will benefit from this thermo-physiological human head simulator leading to the development of helmet designs with enhanced thermal comfort and, therefore, with higher acceptance by users.

Multi-sector thermo-physiological head simulator for headgear research.

Science.gov (United States)

Martinez, Natividad; Psikuta, Agnes; Corberán, José Miguel; Rossi, René M; Annaheim, Simon

2017-02-01

A novel thermo-physiological human head simulator for headgear testing was developed by coupling a thermal head manikin with a thermo-physiological model. As the heat flux at head-site is directly measured by the head manikin, this method provides a realistic quantification of the heat transfer phenomena occurring in the headgear, such as moisture absorption-desorption cycles, condensation, or moisture migration across clothing layers. Before coupling, the opportunities of the head manikin for representing the human physiology were evaluated separately. The evaluation revealed reduced precision in forehead and face temperature predictions under extreme heterogeneous temperature distributions and no initial limitation for simulating temperature changes observed in the human physiology. The thermo-physiological model predicted higher sweat rates when applied for coupled than for pure virtual simulations. After coupling, the thermo-physiological human head simulator was validated using eight human experiments. It precisely predicted core, mean skin, and forehead temperatures with average rmsd values within the average experimental standard deviation (rmsd of 0.20 ± 0.15, 0.83 ± 0.34, and 1.04 ± 0.54 °C, respectively). However, in case of forehead, precision was lower for the exposures including activity than for the sedentary exposures. The representation of the human sweat evaporation could be affected by a reduced evaporation efficiency and the manikin sweat dynamics. The industry will benefit from this thermo-physiological human head simulator leading to the development of helmet designs with enhanced thermal comfort and, therefore, with higher acceptance by users.

The axial power distribution validation of the SCWR fuel assembly with coupled neutronics-thermal hydraulics method

Energy Technology Data Exchange (ETDEWEB)

Xi, Xi [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Xiao, Zejun, E-mail: fabulous_2012@sina.com [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Yan, Xiao; Li, Yongliang; Huang, Yanping [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China)

2013-05-15

Highlights: ► CFX and MCNP codes are suitable to calculate the axial power profile of the FA. ► The partition method in the calculation will affect the final result. ► The density feedback has little effect on the axial power profile of CSR1000 FA. -- Abstract: SCWR (super critical water reactor) is one of the IV generation nuclear reactors in the world. In a typical SCWR the water enters the reactor from the cold leg with a temperature of 280 °C and then leaves the core with a temperature of 500 °C. Due to the sharp change in temperature, there is a huge density change of the water along the axial direction of the fuel assembly (FA), which will affect the moderating power of the water. So the axial power distribution of the SCWR FA could be different from the traditional PWR FA.In this paper, it is the first time that the thermal hydraulics code CFX and neutronics code MCNP are used to analyze the axial power distribution of the SCWR FA. First, the factors in the coupled method which could affect the result are analyzed such as the initialization value or the partition method especially in the MCNP code. Then the axial power distribution of the Europe HPLWR FA is obtained by the coupled method with the two codes and the result is compared with that obtained by Waata and Reiss. There is a good agreement among the three kinds of results. At last, this method is used to calculate the axial power distribution of the Chinese SCWR (CSR1000) FA. It is found the axial power profile of the CSR1000 FA is not so sensitive to the change of the moderator density.

Library of neutron cross sections of the Thermos code

International Nuclear Information System (INIS)

Alonso V, G.; Hernandez L, H.

1991-10-01

The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

Thermo-hydric characterization of partially saturated porous media; Caracterisation thermo-hydrique de milieux poreux partiellement satures d'eau

Energy Technology Data Exchange (ETDEWEB)

Simon Salager; Frederic Jamin; Moulay Said El Youssoufi; Christian Saix [Laboratoire de Mecanique et Genie Civil, Universite Montpellier II, cc 048, Place Eugene Bataillon, 34095 Montpellier (France)

2005-07-01

We present a contribution to the thermo-hydric characterization of partially saturated porous media by water, through the characteristic curve. This curve defines the relation between suction and degree of saturation. Using this curve for a given temperature, a model is used to predict it for other temperatures. An experimental device called pressure cell was made in a thermo-regulated environment. The model was validated by several tests on a ceramic and silty clayey sand, at 20 and 60 C. The results obtained lead to a characteristic surface which can be considered as a generalization of the classical characteristic curve. (authors)

Holddown device for nuclear fuel assembly

International Nuclear Information System (INIS)

Anthony, A.J.

1978-01-01

An apparatus for preventing ''floating'' of nuclear-reactor fuel assemblies due to hydraulic forces is disclosed. The apparatus uses a holddown column made of the same material as the core barrel. The column is positioned in a center guide-tube location in the fuel assembly in such a manner as to enable it either to slide within the center guide tube or, if the center guide tube is replaced by the column, to slide through openings in the spacer grids. The lower end of the holddown column engages the lower end fitting of the fuel assembly, and the upper end of the column engages a flow plate to which holddown force is applied. As a consequence of this arrangement, holddown force is transmitted from the flow plate through the holddown column to the lower end fitting. Movement of the fuel assembly is thereby prevented without a compression load being applied to the fuel-assemb1ly structure. In addition, variations due to thermal expansion in the distance between the lower core plate and the upper core plate are largely made up for by corresponding variations in the holddown column because the holddown column and the core barrel can be made of the same material

Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

Energy Technology Data Exchange (ETDEWEB)

Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

2017-07-01

To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

International Nuclear Information System (INIS)

Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

2017-01-01

To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

Thermo-responsive methylcellulose hydrogels as temporary substrate for cell sheet biofabrication.

Science.gov (United States)

Altomare, Lina; Cochis, Andrea; Carletta, Andrea; Rimondini, Lia; Farè, Silvia

2016-05-01

Methylcellulose (MC), a water-soluble polymer derived from cellulose, was investigated as a possible temporary substrate having thermo-responsive properties favorable for cell culturing. MC-based hydrogels were prepared by a dispersion technique, mixing MC powder (2, 4, 6, 8, 10, 12 % w/v) with selected salts (sodium sulphate, Na2SO4), sodium phosphate, calcium chloride, or phosphate buffered saline, to evaluate the influence of different compositions on the thermo-responsive behavior. The inversion test was used to determine the gelation temperatures of the different hydrogel compositions; thermo-mechanical properties and thermo-reversibility of the MC hydrogels were investigated by rheological analysis. Gelation temperatures and rheological behavior depended on the MC concentration and type and concentration of salt used in hydrogel preparation. In vitro cytotoxicity tests, performed using L929 mouse fibroblasts, showed no toxic release from all the tested hydrogels. Among the investigated compositions, the hydrogel composed of 8 % w/v MC with 0.05 M Na2SO4 had a thermo-reversibility temperature at 37 °C. For that reason, this formulation was thus considered to verify the possibility of inducing in vitro spontaneous detachment of cells previously seeded on the hydrogel surface. A continuous cell layer (cell sheet) was allowed to grow and then detached from the hydrogel surface without the use of enzymes, thanks to the thermo-responsive behavior of the MC hydrogel. Immunofluorescence observation confirmed that the detached cell sheet was composed of closely interacting cells.

Thermal hydraulics of accelerator breeder systems for regeneration of reactor fuel assemblies

International Nuclear Information System (INIS)

Yu, W.S.; Powell, J.R.

1979-01-01

The following conclusions are obtained with regard to the thermal-hydraulic behavior of the Linear Accelerator Fuel Regenerator for PWR and CANDU fuel: (1) two-phase flow is a feasible coolant option for fuel element heat fluxes up to 1 x PWR (or CANDU) average value, which is the maximum design value for a LAFR; (2) two-phase flow pressure drops are low (typically 10 to 30 psi) and film temperature drops very low (typically approx. 10 0 F) for PWR fuel, with inlet velocity range (50 to 75 ft/sec). A somewhat higher inlet velocity range (75 to 100 ft/sec) and pressure drop (50 to 100 psi) is necessary for CANDU fuel, however, to prevent dry out

Thermo-mechanical design and testing of a microbalance for space applications

Science.gov (United States)

Scaccabarozzi, Diego; Saggin, Bortolino; Tarabini, Marco; Palomba, Ernesto; Longobardo, Andrea; Zampetti, Emiliano

2014-12-01

This work focuses on the thermo-mechanical design of the microbalance used for the VISTA (Volatile In Situ Thermogravimetry Analyzer) sensor. VISTA has been designed to operate in situ in different space environments (asteroids, Mars, icy satellites). In this paper we focus on its application on Mars, where the expected environmental conditions are the most challenging for the thermo-mechanical design. The microbalance holding system has been designed to ensure piezoelectric crystal integrity against the high vibration levels during launch and landing and to cope with the unavoidable thermo-elastic differential displacements due to CTE and temperature differences between the microbalance elements. The crystal holding system, based on three symmetrical titanium supports, provides also the electrical connections needed for crystal actuation, microbalance heating and temperature measurement on the electrode area. On the microbalance crystal surfaces the electrodes, a micro film heater (optimized to perform thermo-gravimetric analysis up to 400 °C) and a resistive thermometer are deposited through a vacuum sputtering process. A mockup of the system has been manufactured and tested at the expected vibration levels and the thermal control effectiveness has been verified in thermo-vacuum environment.

ROSA-IV Large Scale Test Facility (LSTF) system description for second simulated fuel assembly

International Nuclear Information System (INIS)

1990-10-01

The ROSA-IV Program's Large Scale Test Facility (LSTF) is a test facility for integral simulation of thermal-hydraulic response of a pressurized water reactor (PWR) during small break loss-of-coolant accidents (LOCAs) and transients. In this facility, the PWR core nuclear fuel rods are simulated using electric heater rods. The simulated fuel assembly which was installed during the facility construction was replaced with a new one in 1988. The first test with this second simulated fuel assembly was conducted in December 1988. This report describes the facility configuration and characteristics as of this date (December 1988) including the new simulated fuel assembly design and the facility changes which were made during the testing with the first assembly as well as during the renewal of the simulated fuel assembly. (author)

General thermo-elastic solution of radially heterogeneous, spherically isotropic rotating sphere

Energy Technology Data Exchange (ETDEWEB)

Bayat, Yahya; EkhteraeiToussi, THamid [Ferdowsi University of Mashhad, Mashhad (Iran, Islamic Republic of)

2015-06-15

A thick walled rotating spherical object made of transversely isotropic functionally graded materials (FGMs) with general types of thermo-mechanical boundary conditions is studied. The thermo-mechanical governing equations consisting of decoupled thermal and mechanical equations are represented. The centrifugal body forces of the rotation are considered in the modeling phase. The unsymmetrical thermo-mechanical boundary conditions and rotational body forces are expressed in terms of the Legendre series. The series method is also implemented in the solution of the resulting equations. The solutions are checked with the known literature and FEM based solutions of ABAQUS software. The effects of anisotropy and heterogeneity are studied through the case studies and the results are represented in different figures. The newly developed series form solution is applicable to the rotating FGM spherical transversely isotropic vessels having nonsymmetrical thermo-mechanical boundary condition.

Sensitivity analyses of biodiesel thermo-physical properties under diesel engine conditions

DEFF Research Database (Denmark)

Cheng, Xinwei; Ng, Hoon Kiat; Gan, Suyin

2016-01-01

This reported work investigates the sensitivities of spray and soot developments to the change of thermo-physical properties for coconut and soybean methyl esters, using two-dimensional computational fluid dynamics fuel spray modelling. The choice of test fuels made was due to their contrasting...... saturation-unsaturation compositions. The sensitivity analyses for non-reacting and reacting sprays were carried out against a total of 12 thermo-physical properties, at an ambient temperature of 900 K and density of 22.8 kg/m3. For the sensitivity analyses, all the thermo-physical properties were set...... as the baseline case and each property was individually replaced by that of diesel. The significance of individual thermo-physical property was determined based on the deviations found in predictions such as liquid penetration, ignition delay period and peak soot concentration when compared to those of baseline...

Improving Thermo-Oxidative Stability of Nitrile Rubber Composites by Functional Graphene Oxide

Directory of Open Access Journals (Sweden)

Rui Zhong

2018-05-01

Full Text Available Graphene oxide (GO, modified with anti-aging agent p-phenylenediamine (PPD, was added into nitrile rubber (NBR in order to improve the thermo-oxidative stability of NBR. The modification of GO and the transformation of functional groups were characterized by Fourier transform infrared spectroscopy (FTIR, Raman, and X-ray diffraction (XRD. Mechanical performances of NBR composites before and after the thermo-oxidative aging were recorded. The results of dynamic mechanical analysis (DMA show an increased storage modulus (G’ and a decreased value of area of tan δ peak after introducing modified GO into NBR. It indicates that filler particles show positive interaction with molecular chains. The thermo-oxidative stability of composites was investigated by thermogravimetric analysis (TG and differential scanning calorimetry (DSC. Then, the thermo-oxidative aging kinetic parameters were obtained by the Flynn–Wall–Ozawa (FWO equation. The results of aging tests show that the thermo-oxidative stability of rubber matrix increases obviously after introducing GO–PPD. In addition, mechanical properties (tensile strength and elongation at break of both before and after aged NBR/GO–PPD composites were superior to that of NBR. This work provides meaningful guidance for achieving multifunction thermo-oxidative aging resistance rubber composites.

Athermalization of resonant optical devices via thermo-mechanical feedback

Science.gov (United States)

Rakich, Peter; Nielson, Gregory N.; Lentine, Anthony L.

2016-01-19

A passively athermal photonic system including a photonic circuit having a substrate and an optical cavity defined on the substrate, and passive temperature-responsive provisions for inducing strain in the optical cavity of the photonic circuit to compensate for a thermo-optic effect resulting from a temperature change in the optical cavity of the photonic circuit. Also disclosed is a method of passively compensating for a temperature dependent thermo-optic effect resulting on an optical cavity of a photonic circuit including the step of passively inducing strain in the optical cavity as a function of a temperature change of the optical cavity thereby producing an elasto-optic effect in the optical cavity to compensate for the thermo-optic effect resulting on an optical cavity due to the temperature change.

The ways for increasing the technical capabilities of antiseismic hydraulic shock absorbers

International Nuclear Information System (INIS)

Kaznovskij, S.P.; Lenskij, V.S.; Plyaskov, A.S.

1986-01-01

Basic achievements in a sphere of production of a new type of atomic power equipment-hydraulic shock absorbers-intended for NPP antiseismic fixation of pipelines are considered in short. In designing the new shock absorbers emphasis is placed on maximum unification of the most labor-consuming units, introduction of promising technological processes for mechanical treatment, optimization of the type and dimensional series, impovement of operational characteristics and simplification of assemblying and preparation for putting into operation

The off-line computation system for supervising performance of JOYO: JOYPAC system, 2

International Nuclear Information System (INIS)

Suzuki, Tomoo; Hasegawa, Akira; Akimoto, Masayuki; Miyamoto, Yoshiaki; Katsuragi, Satoru

1976-10-01

HONEYCOMB is a code for detailed calculations in analyzing nuclear characteristics of the reactor. It performs criticality calculation in diffusion model and burn up calculation, for 3-dimensional hexagonal-z geometry. It can predict the critical insertion depth of control rods and calculate the 3-dimensional power distribution required by thermo-hydraulic calculation. Power distribution and burn up are also obtained for fuel pins, if necessary, as well as for assemblies. FDCAL-2 predicts coolant flow distribution in every coolant channel between inlet and outlet plenums in the reactor vessel. In calculating the flow distribution in the assemblies, the subchannel model is used, and the thermal mixing effect is expressed in terms of an apparent heat transfer coefficient. FATEC-3 calculates temperature distribution within some assemblies, optionally specified in the given core matrix. At the same time, it estimates the hot-spot temperature, one of the informations for confirming the safe operation. FACAL-2 and FATEC-3 have been combined so as to remove their unnecessary overlapping parts, and have consequently formed a detailed calculation code for analyzing thermo-hydraulic characteristics of the reactor, FDCAL-3. FDCAL-3 has been linked to HONEYCOMB as a segment of overlay structure, and this combination of HONEYCOMB and FDCAL-3 forms the detailed calculation subsystem in the JOYPAC system. The detailed calculation subsystem produces the data file of the detailed fundamental informations such as distributions of neutron flux, power etc. about the reactor under stationary performance. This file is required by the quick and simple calculation subsystem SMART and the recording subsystem MASTOR described in Part I. Thus, times of resorting to the time-consuming detailed calculation are reduced as far as possible, and supervision of reactor performance is realized in both features of practically sufficient accuracy and reasonable computer cost. (JPN)

Thermal-Hydraulic Simulations of Single Pin and Assembly Sector for IVG- 1M Reactor

Energy Technology Data Exchange (ETDEWEB)

Kraus, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Garner, P. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanan, N. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-01-15

Thermal-hydraulic simulations have been performed using computational fluid dynamics (CFD) for the highly-enriched uranium (HEU) design of the IVG.1M reactor at the Institute of Atomic Energy (IAE) at the National Nuclear Center (NNC) in the Republic of Kazakhstan. Steady-state simulations were performed for both types of fuel assembly (FA), i.e. the FA in rows 1 & 2 and the FA in row 3, as well as for single pins in those FA (600 mm and 800 mm pins). Both single pin calculations and bundle sectors have been simulated for the most conservative operating conditions corresponding to the 10 MW output power, which corresponds to a pin unit cell Reynolds number of only about 7500. Simulations were performed using the commercial code STAR-CCM+ for the actual twisted pin geometry as well as a straight-pin approximation. Various Reynolds-Averaged Navier-Stokes (RANS) turbulence models gave different results, and so some validation runs with a higher-fidelity Large Eddy Simulation (LES) code were performed given the lack of experimental data. These singled out the Realizable Two-Layer k-ε as the most accurate turbulence model for estimating surface temperature. Single-pin results for the twisted case, based on the average flow rate per pin and peak pin power, were conservative for peak clad surface temperature compared to the bundle results. Also the straight-pin calculations were conservative as compared to the twisted pin simulations, as expected, but the single-pin straight case was not always conservative with regard to the straight-pin bundle. This was due to the straight-pin temperature distribution being strongly influenced by the pin orientation, particularly near the outer boundary. The straight-pin case also predicted the peak temperature to be in a different location than the twisted-pin case. This is a limitation of the straight-pin approach. The peak temperature pin was in a different location from the peak power pin in every case simulated, and occurred at an

Hydraulic manipulator research at ORNL

International Nuclear Information System (INIS)

Kress, R.L.; Jansen, J.F.; Love, L.J.

1997-01-01

Recently, task requirements have dictated that manipulator payload capacity increase to accommodate greater payloads, greater manipulator length, and larger environmental interaction forces. General tasks such as waste storage tank cleanup and facility dismantlement and decommissioning require manipulator life capacities in the range of hundreds of pounds rather than tens of pounds. To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned once again to hydraulics as a means of actuation. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem), sophisticated modeling, analysis, and control experiments are usually needed. Oak Ridge National Laboratory (ORNL) has a history of projects that incorporate hydraulics technology, including mobile robots, teleoperated manipulators, and full-scale construction equipment. In addition, to support the development and deployment of new hydraulic manipulators, ORNL has outfitted a significant experimental laboratory and has developed the software capability for research into hydraulic manipulators, hydraulic actuators, hydraulic systems, modeling of hydraulic systems, and hydraulic controls. The purpose of this article is to describe the past hydraulic manipulator developments and current hydraulic manipulator research capabilities at ORNL. Included are example experimental results from ORNL's flexible/prismatic test stand

Hydraulic manipulator research at ORNL

Energy Technology Data Exchange (ETDEWEB)

Kress, R.L.; Jansen, J.F. [Oak Ridge National Lab., TN (United States); Love, L.J. [Oak Ridge Inst. for Science and Education, TN (United States)

1997-03-01

Recently, task requirements have dictated that manipulator payload capacity increase to accommodate greater payloads, greater manipulator length, and larger environmental interaction forces. General tasks such as waste storage tank cleanup and facility dismantlement and decommissioning require manipulator life capacities in the range of hundreds of pounds rather than tens of pounds. To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned once again to hydraulics as a means of actuation. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem), sophisticated modeling, analysis, and control experiments are usually needed. Oak Ridge National Laboratory (ORNL) has a history of projects that incorporate hydraulics technology, including mobile robots, teleoperated manipulators, and full-scale construction equipment. In addition, to support the development and deployment of new hydraulic manipulators, ORNL has outfitted a significant experimental laboratory and has developed the software capability for research into hydraulic manipulators, hydraulic actuators, hydraulic systems, modeling of hydraulic systems, and hydraulic controls. The purpose of this article is to describe the past hydraulic manipulator developments and current hydraulic manipulator research capabilities at ORNL. Included are example experimental results from ORNL`s flexible/prismatic test stand.

Thermo-hydro-mechanical behaviour of Boom clay; Comportement thermo-hydro-mecanique de l'argile de Boom

Energy Technology Data Exchange (ETDEWEB)

Le, T.T

2008-01-15

This thesis studied the thermo-hydro-mechanical properties of Boom clay, which was chosen to be the host material for the radioactive waste disposal in Mol, Belgium. Firstly, the research was concentrated on the soil water retention properties and the hydro-mechanical coupling by carrying out axial compression tests with suction monitoring. The results obtained permitted elaborating a rational experimental procedure for triaxial tests. Secondly, the systems for high pressure triaxial test at controlled temperature were developed to carry out compression, heating, and shearing tests at different temperatures. The obtained results showed clear visco-elasto-plastic behaviour of the soil. This behaviour was modelled by extending the thermo-elasto-plastic model of Cui et al. (2000) to creep effect. (author)

Analysis of Thermo-Acoustic Emission from Damage in Composite Laminates under Thermal Cyclic Loading

International Nuclear Information System (INIS)

Kim, Young Bok; Min, Dae Hong; Lee, Deok Bo; Choi, Nak Sam

2001-01-01

An investigation on nondestructive evaluation of thermal stress-reduced damage in the composite laminates (3mm in thickness and [+45 6 /-45 6 ] S lay-up angles) has been performed using the thermo-acoustic emission technique. Reduction of thermo-AE events due to repetitive thermal load cycles showed a Kaiser effect. An analysis of the thermo-AE behavior determined the stress free temperature of composite laminates. Fiber fracture and matrix cracks were observed using the optical microscopy, scanning electron microscopy and ultrasonic C-sean. Short-Time Fourier Transform of thermo-AE signals offered the time-frequency characteristics which might classify the thermo-AE as three different types to estimate the damage processes of the composites

Development of whole core thermal-hydraulic analysis program ACT. 4. Incorporation of three-dimensional upper plenum model

International Nuclear Information System (INIS)

Ohshima, Hiroyuki

2003-03-01

The thermal-hydraulic analysis computer program ACT is under development for the evaluation of detailed flow and temperature fields in a core region of fast breeder reactors under various operation conditions. The purpose of this program development is to contribute not only to clarifying thermal hydraulic characteristics that cannot be revealed by experiments due to measurement difficulty but also to performing rational safety design and assessment. This report describes the incorporation of a three-dimensional upper plenum model to ACT and its verification study as part of the program development. To treat the influence of three-dimensional thermal-hydraulic behavior in a upper plenum on the in-core temperature field, the multi-dimensional general purpose thermal-hydraulic analysis program AQUA, which was developed and validated at JNC, was applied as the base of the upper plenum analysis module of ACT. AQUA enables to model the upper plenum configuration including immersed heat exchangers of the direct reactor auxiliary cooling system (DRACS). In coupling core analysis module that consists of the fuel-assembly and the inter-wrapper gap calculation parts with the upper plenum module, different types of computation mesh systems were jointed using the staggered quarter assembly mesh scheme. A coupling algorithm among core, upper plenum and heat transport system modules, which can keep mass, momentum and energy conservation, was developed and optimized in consideration of parallel computing. ACT was applied to analyzing a sodium experiment (PLANDTL-DHX) performed at JNC, which simulated the natural circulation decay heat removal under DRACS operation conditions for the program verification. From the calculation result, the validity of the improved program was confirmed. (author)

Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

Energy Technology Data Exchange (ETDEWEB)

Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

2014-12-15

Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

Thermo-ecological optimization of a solar collector

International Nuclear Information System (INIS)

Szargut, J.; Stanek, W.

2007-01-01

The depletion of non-renewable natural exergy resources (the thermo-ecological cost) has been accepted as the objective function for thermo-ecological optimization. Its general formulation has been cited. A detailed form of the objective function has been formulated for a solar collector producing hot water for household needs. The following design parameters have been accepted as the decision variables: the collector area per unit of the heat demand, the diameter of collector pipes, the distance of the pipe axes in the collector plate. The design parameters of the internal installation (the pipes, the hot water receiver) have not been taken into account, because they are very individual. The accumulation ability of hot water comprising one day has been assumed. The objective function contains the following components: the thermo-ecological cost of copper plate, copper pipes, glass plate, steel box, thermal insulation, heat transfer liquid, electricity for driving the pump of liquid, fuel for the peak boiler. The duration curves of the flux of solar radiation and absorbed heat have been elaborated according to meteorological data and used in the calculations. The objective function for economic optimization may have a similar form, only the cost values would be different

Ground source thermo-pumps for individual residential houses; Les thermopompes a capteur enterres dans les residences individuelles

Energy Technology Data Exchange (ETDEWEB)

Ossant, G. [Societe Syrec (France)

1997-12-31

The main principles, performances and constraints of the various types of ground source thermo-pumps for individual houses, i.e. ground/ground thermo-pumps, glycol water/water thermo-pumps and ground/water thermo-pumps are reviewed, and their energy consumptions are discussed. The design and operating conditions of a reverse ground source thermo-pump (Syrec) for space heating and air conditioning through a hot and cold floor system and a Syrec ground source thermo-pump for water heating, are presented

Insights into the Mechanism and Kinetics of Thermo-Oxidative Degradation of HFPE High Performance Polymer.

Science.gov (United States)

Kunnikuruvan, Sooraj; Parandekar, Priya V; Prakash, Om; Tsotsis, Thomas K; Nair, Nisanth N

2016-06-02

The growing requisite for materials having high thermo-oxidative stability makes the design and development of high performance materials an active area of research. Fluorination of the polymer backbone is a widely applied strategy to improve various properties of the polymer, most importantly the thermo-oxidative stability. Many of these fluorinated polymers are known to have thermo-oxidative stability up to 700 K. However, for space and aerospace applications, it is important to improve its thermo-oxidative stability beyond 700 K. Molecular-level details of the thermo-oxidative degradation of such polymers can provide vital information to improve the polymer. In this spirit, we have applied quantum mechanical and microkinetic analysis to scrutinize the mechanism and kinetics of the thermo-oxidative degradation of a fluorinated polymer with phenylethenyl end-cap, HFPE. This study gives an insight into the thermo-oxidative degradation of HFPE and explains most of the experimental observations on the thermo-oxidative degradation of this polymer. Thermolysis of C-CF3 bond in the dianhydride component (6FDA) of HFPE is found to be the rate-determining step of the degradation. Reaction pathways that are responsible for the experimentally observed weight loss of the polymer is also scrutinized. On the basis of these results, we propose a modification of HFPE polymer to improve its thermo-oxidative stability.

Forced synchronization and asynchronous quenching in a thermo-acoustic system

Science.gov (United States)

Mondal, Sirshendu; Pawar, Samadhan A.; Sujith, Raman

2017-11-01

Forced synchronization, which has been extensively studied in theory and experiments, occurs through two different mechanisms known as phase locking and asynchronous quenching. The latter indicates the suppression of oscillation amplitude. In most practical combustion systems such as gas turbine engines, the main concern is high amplitude pressure oscillations, known as thermo-acoustic instability. Thermo-acoustic instability is undesirable and needs to be suppressed because of its damaging consequences to an engine. In the present study, a systematic experimental investigation of forced synchronization is performed in a prototypical thermo-acoustic system, a Rijke tube, in its limit cycle operation. Further, we show a qualitatively similar behavior using a reduced order model. In the phase locking region, the simultaneous occurrence of synchronization and resonant amplification leads to high amplitude pressure oscillations. However, a reduction in the amplitude of natural oscillations by about 78% of the unforced amplitude is observed when the forcing frequency is far lower than the natural frequency. This shows the possibility of suppression of the oscillation amplitude through asynchronous quenching in thermo-acoustic systems.

Handbook of hydraulic fluid technology

CERN Document Server

Totten, George E

2011-01-01

""The Handbook of Hydraulic Fluid Technology"" serves as the foremost resource for designing hydraulic systems and for selecting hydraulic fluids used in engineering applications. Featuring new illustrations, data tables, as well as practical examples, this second edition is updated with essential information on the latest hydraulic fluids and testing methods. The detailed text facilitates unparalleled understanding of the total hydraulic system, including important hardware, fluid properties, and hydraulic lubricants. Written by worldwide experts, the book also offers a rigorous overview of h

A thermo-degradable hydrogel with light-tunable degradation and drug release.

Science.gov (United States)

Hu, Jingjing; Chen, Yihua; Li, Yunqi; Zhou, Zhengjie; Cheng, Yiyun

2017-01-01

The development of thermo-degradable hydrogels is of great importance in drug delivery. However, it still remains a huge challenge to prepare thermo-degradable hydrogels with inherent degradation, reproducible, repeated and tunable dosing. Here, we reported a thermo-degradable hydrogel that is rapidly degraded above 44 °C by a facile chemistry. Besides thermo-degradability, the hydrogel also undergoes rapid photolysis with ultraviolet light. By embedding photothermal nanoparticles or upconversion nanoparticles into the gel, it can release the entrapped cargoes such as dyes, enzymes and anticancer drugs in an on-demand and dose-tunable fashion upon near-infrared light exposure. The smart hydrogel works well both in vitro and in vivo without involving sophisticated syntheses, and is well suited for clinical cancer therapy due to the high transparency and non-invasiveness features of near-infrared light. Copyright © 2016 Elsevier Ltd. All rights reserved.

Thermo-siphon Mock-up Test for the HANARO-CNS

Energy Technology Data Exchange (ETDEWEB)

Choi, Jungwoon; Lee, Kye Hong; Kim, Hark Rho; Kim, Youngki; Kim, Myong Seop; Wu, Sang Ik; Kim, Bong Su

2006-04-15

In order to moderate thermal neutrons into cold neutrons, the liquid hydrogen is selected as a moderator for the HANARO CNS. By the non-nuclear heat load and nuclear heat load induced from collision of gamma-ray, beta-ray, and thermal neutrons, the liquid hydrogen in the moderator cell evaporates and flows into the heat exchanger. This evaporated hydrogen gas is liquefied by the cryogenic helium supplied from the helium refrigeration system,, then flows back to the moderator cell. This is so-called two-phase thermo-siphon. The most important point in the stable thermo-siphon is to have the good balance between the cooling capacity of the HRS and the heat load on the moderator cell so as to maintain the stable two-phase liquid level in the moderator cell. Accordingly, for not only the experience of the cryogenic two-phase thermo-siphon but also setup of the operation procedure, the full-scaled mock-up test has been performed using the liquid hydrogen. Through the test, the stable thermo-siphon establishment is confirmed at the cold normal operation; furthermore, the detail design parameter is validated. On top of the normal operation procedure setup, the abnormal operation procedure is settled based on the understanding the abnormal pressure and temperature transient dynamics in the hydrogen system.

Beneath the surface of water. Hydraulic structures and human skeletal remains in Ancient Italy

Directory of Open Access Journals (Sweden)

Vera Zanoni

2013-12-01

Full Text Available Recent findings from the area of Modena, in Northern Italy, have revitalized the debate on the association between human skeletal remains and artificial hydraulic structures. In this paper, our intention is to assemble the relevant archaeological and anthropological data on the matter in order to establish whether these findings are exceptional and isolated or indicate instead a structured and specific cultural behaviour which persists through time.

Thermo-mechanical properties of SOFC components investigated by a combined method

DEFF Research Database (Denmark)

Teocoli, Francesca; Esposito, Vincenzo; Ramousse, Severine

, and differential thermo-mechanical behavior at each layer. The combination of such factors can have a critical effect on the final shape and microstructure, and on the mechanical integrity. Thermo-mechanical properties and sintering mechanisms of important SOFC materials (CGO, YSZ, ScYSZ) were systematically...

The Metabolic Basis of Pollen Thermo-Tolerance: Perspectives for Breeding

Directory of Open Access Journals (Sweden)

Marine J. Paupière

2014-09-01

Full Text Available Crop production is highly sensitive to elevated temperatures. A rise of a few degrees above the optimum growing temperature can lead to a dramatic yield loss. A predicted increase of 1–3 degrees in the twenty first century urges breeders to develop thermo-tolerant crops which are tolerant to high temperatures. Breeding for thermo-tolerance is a challenge due to the low heritability of this trait. A better understanding of heat stress tolerance and the development of reliable methods to phenotype thermo-tolerance are key factors for a successful breeding approach. Plant reproduction is the most temperature-sensitive process in the plant life cycle. More precisely, pollen quality is strongly affected by heat stress conditions. High temperature leads to a decrease of pollen viability which is directly correlated with a loss of fruit production. The reduction in pollen viability is associated with changes in the level and composition of several (groups of metabolites, which play an important role in pollen development, for example by contributing to pollen nutrition or by providing protection to environmental stresses. This review aims to underline the importance of maintaining metabolite homeostasis during pollen development, in order to produce mature and fertile pollen under high temperature. The review will give an overview of the current state of the art on the role of various pollen metabolites in pollen homeostasis and thermo-tolerance. Their possible use as metabolic markers to assist breeding programs for plant thermo-tolerance will be discussed.

The metabolic basis of pollen thermo-tolerance: perspectives for breeding.

Science.gov (United States)

Paupière, Marine J; van Heusden, Adriaan W; Bovy, Arnaud G

2014-09-30

Crop production is highly sensitive to elevated temperatures. A rise of a few degrees above the optimum growing temperature can lead to a dramatic yield loss. A predicted increase of 1-3 degrees in the twenty first century urges breeders to develop thermo-tolerant crops which are tolerant to high temperatures. Breeding for thermo-tolerance is a challenge due to the low heritability of this trait. A better understanding of heat stress tolerance and the development of reliable methods to phenotype thermo-tolerance are key factors for a successful breeding approach. Plant reproduction is the most temperature-sensitive process in the plant life cycle. More precisely, pollen quality is strongly affected by heat stress conditions. High temperature leads to a decrease of pollen viability which is directly correlated with a loss of fruit production. The reduction in pollen viability is associated with changes in the level and composition of several (groups of) metabolites, which play an important role in pollen development, for example by contributing to pollen nutrition or by providing protection to environmental stresses. This review aims to underline the importance of maintaining metabolite homeostasis during pollen development, in order to produce mature and fertile pollen under high temperature. The review will give an overview of the current state of the art on the role of various pollen metabolites in pollen homeostasis and thermo-tolerance. Their possible use as metabolic markers to assist breeding programs for plant thermo-tolerance will be discussed.

Hydraulics and heat transfer in the IFMIF liquid lithium target: CFD calculations

OpenAIRE

Peña, A.; Esteban, G.A.; Sancho, J.; Kolesnik, V.; Abánades Velasco, Alberto

2009-01-01

CFD (Computational fluid dynamics) calculation turns out to be a good approximation to the real behavior of the lithium (Li) flow of the target of the international fusion materials irradiation facility (IFMIF). A three-dimensional (3D) modelling of the IFMIF design Li target assembly, made with the CFD commercial code ANSYS-FLUENT has been carried out. The simulation by a structural mesh is focused on the thermal-hydraulic analysis inside the Li jet flow. For, this purpose, the two deuteron ...

Thermal-hydraulic experiment and analysis for interim dry storage of spent nuclear fuel

International Nuclear Information System (INIS)

Yoo, Seung Hun

2011-02-01

The experimental and numerical studies of interim storages for nuclear spent fuels have been performed to investigate thermal-hydraulic characteristics of the dry storage systems and to propose new methodologies for the analysis and the design. Three separate researches have been performed in the present study: (a) Development of a scaling methodology and thermal-hydraulic experiment of a single spent fuel assembly simulating a dry storage cask: (b) Full-scope simulation of a dry storage cask by the use of Computational Fluid Dynamics (CFD) code: (c) Thermal-hydraulic design of a tunnel-type interim storage facility. In the first study, a scaling methodology has been developed to design a scaled-down canister. The scaling was performed in two steps. For the first step, the height of a spent fuel assembly was reduced from full height to half height. In order to consider the effect of height reduction on the natural convection, the scaling law of Ishii and Kataoka (1984) was employed. For the second step, the quantity of spent fuel assemblies was reduced from multiple assemblies to a single assembly. The scaling methodology was validated through the comparison with the experiment of the TN24P cask. The Peak Cladding Temperature (PCT), temperature gradients, and the axial and radial temperature distribution in the nondimensional forms were in good agreement with the experimental data. Based on the developed methodology, we have performed a single assembly experiment which was designed to simulate the full scale of the TN24P cask. The experimental data was compared with the CFD calculations. It turns out that their PCTs were less than the maximum allowable temperature for the fuel cladding and that the differences of their PCTs were agreed within 3 .deg. C, which was less than measurement uncertainty. In the second study, the full-scope simulations of the TN24P cask were performed by FLUENT. In order to investigate the sensitivity of the numerical and physical

Comparison of thermo-hydraulic analysis with measurements for HELIOS. The scaled integral test loop for PEACER

International Nuclear Information System (INIS)

Cho, Jae Hyun; Lim, Jun; Kim, Ji Hak; Hwang, Il Soon

2009-01-01

A scaled-down Lead-Bismuth Eutectic circulating integral test loop named as HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER) has been employed to characterize steady-state isothermal forced circulation behavior and non-isothermal natural circulation capability of the lead and lead-alloy cooled advanced nuclear energy systems (LACANES). In this time, thermal-hydraulic experiments have been carried out using HELIOS following rigorous calibration campaigns on sensors for temperature and pressure, especially isothermal steady-state forced convection using by the pump. The isothermal steady-state forced convection test was performed to obtain the pressure loss information including friction loss coefficients and form loss coefficients. Then its data were compared with multi-approaching analysis including hand calculation results and computer simulation code results. (MARS-LBE, CFX). We report the results of comparisons between the analysis and measurements together. (author)

Thermo-Optic Characterization of Silicon Nitride Resonators for Cryogenic Photonic Circuits

NARCIS (Netherlands)

Elshaari, A.W.A.; Esmaeil Zadeh, I.; Jöns, K.D.; Zwiller, Val

2016-01-01

In this paper, we characterize the Thermo-optic properties of silicon nitride ring resonators between 18 and 300 K. The Thermo-optic coefficients of the silicon nitride core and the oxide cladding are measured by studying the temperature dependence of the resonance wavelengths. The resonant modes

Analysis code for medium and small rupture accidents in ATR. LOTRAC/HEATUP

International Nuclear Information System (INIS)

1997-08-01

In the evaluation of thermo-hydraulic and fuel temperature transient changes in the events which are classified in medium and small rupture accidents of reactor coolant loss that is the safety evaluation event of the ATR, the analysis code for synthetic thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC and the detailed analysis code for fuel temperature HEATUP are used, respectively. By using the LOTAC, the thermo-hydraulic behavior of reactor cooling facility and the temperature behavior of fuel at the time of blow-down are analyzed, and also the characteristics of changing reactor thermal output is analyzed, considering the functioning characteristics of emergency core cooling system. Based on the data of thermo-hydraulic behavior obtained by the LOTRAC, the time of beginning the turn-around of fuel cladding tube temperature obtained by the data of ECCS pouring characteristics, the heat transfer rate after the turn-around and so on, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The LOTRAC code, the HEATUP code, various analysis models, and rupture simulation experiment are reported. (K.I.)

Thermo-, photo-, and mechano-responsive liquid crystal networks enable tunable photonic crystals.

Science.gov (United States)

Akamatsu, N; Hisano, K; Tatsumi, R; Aizawa, M; Barrett, C J; Shishido, A

2017-10-25

Tunable photonic crystals exhibiting optical properties that respond reversibly to external stimuli have been developed using liquid crystal networks (LCNs) and liquid crystal elastomers (LCEs). These tunable photonic crystals possess an inverse opal structure and are photo-responsive, but circumvent the usual requirement to contain dye molecules in the structure that often limit their applicability and cause optical degradation. Herein, we report tunable photonic crystal films that reversibly tune the reflection peak wavelength under thermo-, photo- and mechano-stimuli, through bilayering a stimuli-responsive LCN including azobenzene units with a colourless inverse opal film composed of non-responsive, flexible durable polymers. By mechanically deforming the azobenzene containing LCN via various stimuli, the reflection peak wavelength from the bilayered film assembly could be shifted on demand. We confirm that the reflection peak shift occurs due to the deformation of the stimuli-responsive layer propagating towards and into the inverse opal layer to change its shape in response, and this shift behaviour is repeatable without optical degradation.

Thermo-mechanical modelling and experimental validation of CLIC prototype module type 0

CERN Document Server

Kortelainen, Lauri; Koivurova, Hannu; Riddone, Germana; Österberg, Kenneth

Micron level stability of the two-meter repetitive modules constituting the two main linacs is one of the most important requirements to achieve the luminosity goal for the Compact Linear Collider. Structural deformations due to thermal loads and related to the RF power dissipated inside the modules affect the alignment of the linacs and therefore the resulting luminosity performance. A CLIC prototype module has been assembled in a dedicated laboratory and a thermal test program has been started in order to study its thermo-mechanical behaviour. This thesis focuses on the finite elements modelling of the first CLIC prototype module 0. The aim of the modelling is to examine the temperature distributions and the resulting deformations of the module in different operating conditions defined in the thermal test program. The theoretical results have been compared to the experimental ones; the comparison shows that the results are in good agreement both for the thermal behaviour of the module and for the resulting ...

Intelligent Materials Used in Hydraulic, Fuel, and Rudder Control Systems of Aircrafts

Directory of Open Access Journals (Sweden)

D. B. Chernov

2017-01-01

Full Text Available The device is really intelligent, only if it is capable to respond to changing external conditions. The devices, which "feel" the external environment and can change their characteristics, have many advantages compared to the conventional devices: they are more efficient, wear out more slowly, and have lower operating costs.The scope of smart products is truly infinite. Alloys with memory effect also apply to intellectual content. Natural piezoelectric crystals such as silicon dioxide (intellectual material have been known for over a hundred years. They have greater stiffness and can be used at high operating frequencies. Due to the direct piezoelectric effect, they have been successfully used as a strain gage. Later came artificial ceramic piezoelectric materials; they are used as mechanical transducers. Thus, an inverse piezoelectric effect is usually used. It consists in the change of dimensions when an electric field is applied. Control of intellectual structure can be provided by heat fluxes, electromagnetic, hydraulic or piezoelectric forces and through application of electro-rheological, and magneto-rheological fluids. The article examines the intellectual materials and technologies that are already in place or will find its application in aviation hydraulic and fuel systems and control systems of rudders (CSR of aircrafts in the near future.The paper considers in detail the shape memory effect alloys (SMEA as "intelligent" materials. Actuators made from SMEA have a number of advantages: high working power; large recoverable deformation; different types of strain (tensile, compressive, bending and torsional; most specific value of the work per unit mass. All the SMEA advantages may be well used for the so-called thermo-mechanical connections (TMС of pipelines where SMEA drawbacks in this application, practically, do not affect the quality of TMC. In aircraft engineering the TMC were first used in hydraulic systems of the aircraft TU204

Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

International Nuclear Information System (INIS)

Niceno, Bojan; Pouchon, Manuel

2014-01-01

The accident in Fukushima has drastically shown the drawbacks of Zircaloy claddings despite their beneficial properties in normal use. The effect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an alternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on thermal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident management guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these issues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from

Thermal energy storage using thermo-chemical heat pump

International Nuclear Information System (INIS)

Hamdan, M.A.; Rossides, S.D.; Haj Khalil, R.

2013-01-01

Highlights: ► Understanding of the performance of thermo chemical heat pump. ► Tool for storing thermal energy. ► Parameters that affect the amount of thermal stored energy. ► Lithium chloride has better effect on storing thermal energy. - Abstract: A theoretical study was performed to investigate the potential of storing thermal energy using a heat pump which is a thermo-chemical storage system consisting of water as sorbet, and sodium chloride as the sorbent. The effect of different parameters namely; the amount of vaporized water from the evaporator, the system initial temperature and the type of salt on the increase in temperature of the salt was investigated and hence on the performance of the thermo chemical heat pump. It was found that the performance of the heat pump improves with the initial system temperature, with the amount of water vaporized and with the water remaining in the system. Finally it was also found that lithium chloride salt has higher effect on the performance of the heat pump that of sodium chloride.

Recent advances in modeling and validation of nuclear thermal-hydraulics applications with NEPTUNE CFD - 15471

International Nuclear Information System (INIS)

Guingo, M.; Baudry, C.; Hassanaly, M.; Lavieville, J.; Mechitouna, N.; Merigoux, N.; Mimouni, S.; Bestion, D.; Coste, P.; Morel, C.

2015-01-01

NEPTUNE CFD is a Computational Multi-(Fluid) Dynamics code dedicated to the simulation of multiphase flows, primarily targeting nuclear thermo-hydraulics applications, such as the departure from nuclear boiling (DNB) or the two-phase Pressurized Thermal Shock (PTS). It is co-developed within the joint research/development project NEPTUNE (AREVA, CEA, EDF, IRSN) since 2001. Over the years, to address the aforementioned applications, dedicated physical models and numerical methods have been developed and implemented in the code, including specific sets of models for turbulent boiling flows and two-phase non-adiabatic stratified flows. This paper aims at summarizing the current main modeling capabilities of the code, and gives an overview of the associated validation database. A brief summary of emerging applications of the code, such as containment simulation during a potential severe accident or in-vessel retention, is also provided. (authors)

PLC based control system for RAM assembly test facility

International Nuclear Information System (INIS)

Kulkarni, S.S.; Kumar, Vinaya; Chandra, Umesh

1994-01-01

The flexibility, expandability, ease of programming and diagnostic features makes the programmable logic controller (PLC) suitable for a variety of control applications in engineering system test facilities. A PLC based control system for RAM assembly test facility (RATF) and for testing the related hydraulic components is being developed and installed at BARC. This paper describes the approach taken for meeting the control requirements and illustrates the PLC software that has been developed. (author). 1 fig

Control rod drive hydraulic device

International Nuclear Information System (INIS)

Takekawa, Toru.

1994-01-01

The device of the present invention can reliably prevent a possible erroneous withdrawal of control rod driving mechanism when the pressure of a coolant line is increased by isolation operation of hydraulic control units upon periodical inspection for a BWR type reactor. That is, a coolant line is connected to the downstream of a hydraulic supply device. The coolant line is connected to a hydraulic control unit. A coolant hydraulic detection device and a pressure setting device are disposed to the coolant line. A closing signal line and a returning signal line are disposed, which connect the hydraulic supply device and a flow rate control valve for the hydraulic setting device. In the device of the present invention, even if pressure of supplied coolants is elevated due to isolation of hydraulic control units, the elevation of the hydraulic pressure can be prevented. Accordingly, reliability upon periodical reactor inspection can be improved. Further, the facility is simplified and the installation to an existent facility is easy. (I.S.)

ε-Polylysine-based thermo-responsive adsorbents for immunoglobulin adsorption-desorption under mild conditions.

Science.gov (United States)

Maruyama, Masashi; Shibuya, Keisuke

2017-08-22

Thermo-responsive adsorbents for immunoglobulin G (IgG) employing ε-polylysine (EPL) as a polymer backbone were developed. The introduction of mercaptoethylpyridine (MEP) as an IgG-binding ligand and hydrophobization of side chains afforded thermo-responsive IgG adsorbents, whose thermo-responsive IgG desorption ratio was up to 88% (EPL/MEP derivative 3m). The changes in surface densities of active MEP groups, which are caused by thermal conformational changes of the adsorbents, play key roles for IgG desorption. Although a trade-off of IgG adsorption capacity and IgG desorption ratio was observed, the present study offers a novel molecular design for thermo-responsive adsorbents with high synthetic accessibility and potentially low toxicity.

Study on Characteristics of Hydraulic Servo System for Force Control of Hydraulic Robots

International Nuclear Information System (INIS)

Kim, Hyo-gon; Han, Changsoo; Lee, Jong-won; Park, Sangdeok

2015-01-01

Because a hydraulic actuator has high power and force densities, this allows the weight of the robot's limbs to be reduced. This allows for good dynamic characteristics and high energy efficiency. Thus, hydraulic actuators are used in some exoskeleton robots and quadrupedal robots that require high torque. Force control is useful for robot compliance with a user or environment. However, force control of a hydraulic robot is difficult because a hydraulic servo system is highly nonlinear from a control perspective. In this study, a nonlinear model was used to develop a simulation program for a hydraulic servo system consisting of a servo valve, transmission lines, and a cylinder. The problems and considerations with regard to the force control performance for a hydraulic servo system were investigated. A force control method using the nonlinear model was proposed, and its effect was evaluated with the simulation program

Study on Characteristics of Hydraulic Servo System for Force Control of Hydraulic Robots

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyo-gon; Han, Changsoo [Hanyang University, Seoul (Korea, Republic of); Lee, Jong-won [Korea University of Science and Technology, Seoul (Korea, Republic of); Park, Sangdeok [Korea Institute of Industrial Technology, Seoul (Korea, Republic of)

2015-02-15

Because a hydraulic actuator has high power and force densities, this allows the weight of the robot's limbs to be reduced. This allows for good dynamic characteristics and high energy efficiency. Thus, hydraulic actuators are used in some exoskeleton robots and quadrupedal robots that require high torque. Force control is useful for robot compliance with a user or environment. However, force control of a hydraulic robot is difficult because a hydraulic servo system is highly nonlinear from a control perspective. In this study, a nonlinear model was used to develop a simulation program for a hydraulic servo system consisting of a servo valve, transmission lines, and a cylinder. The problems and considerations with regard to the force control performance for a hydraulic servo system were investigated. A force control method using the nonlinear model was proposed, and its effect was evaluated with the simulation program.

Fundamental topics for thermo-elastic stress analyses

International Nuclear Information System (INIS)

Biermann, M.

1989-01-01

This paper delivers a consistent collection of theoretical fundamentals needed to perform rather sound experimental stress analyses on thermo-elastic materials. An exposition of important concepts of symmetry and so-called peer groups, yielding the very base for a rational description of materials, goes ahead and is followed by an introduction to the constitutive theory of simple materials. Neat distinction is made between stress contributions determined by deformational and thermal impressions, on the one part, and stress constraints not accessible to strain gauging, on the other part. The mathematical formalism required for establishing constitutive equations is coherently developed from scratch and aided, albeit not subrogated, by intuition. The main intention goes to turning some of the recent advances in the nonlinear field theories of thermomechanics to practical account. A full success therein, obviously, results under the restriction to thermo-elasticity. In adverting to more particular subjects, the elementary static effects of nonlinear isotropic elasticity are pointed out. Due allowance is made for thermal effects likely to occur in heat conducting materials also beyond the isothermal or isentropic limit cases. Linearization of the constitutive equations for anisotropic thermo-elastic materials is then shown to entail the formulas of the classical theory. (orig./MM) [de

Hydraulic Hybrid Vehicles

Science.gov (United States)

EPA and the United Parcel Service (UPS) have developed a hydraulic hybrid delivery vehicle to explore and demonstrate the environmental benefits of the hydraulic hybrid for urban pick-up and delivery fleets.

Thermo-power in Brazil: diagnosis of control and monitoring of gas emissions

International Nuclear Information System (INIS)

Xavier, E.E.; Magrini, Alessandra; Rosa, L.P.; Santos, M.A. dos

2004-01-01

In parallel to Brazil's recent supply crisis, the privatization process of its power sector has drastically reshaped the nation's energy matrix. From a profile based mainly on hydro-power generation, this sector is being reshaped through a thermo-power plant construction program whose environmental repercussions will certainly be felt over the next few years. This paper offers a description of the thermo-power segment currently in operation, under construction and on the drawing board in Brazil, followed by the results of a diagnosis of the control and monitoring of the gas emissions by this segment. The methodology used for the exploratory analysis and to prepare the diagnosis consists of surveys through questionnaires completed by companies owning the thermo-power plants. After consolidating, processing and analyzing the findings reached through the replies sent in by the companies, it is concluded that thermo-power plants currently in operation lack control systems that would help reduce atmospheric pollution, and are not equipped with monitoring systems for these emissions. The thermo-power plants currently under construction and on the drawing board indicate a trend towards including these systems in their project designs, due to more stringent licensing processes

Enhanced Thermo-Optical Switching of Paraffin-Wax Composite Spots under Laser Heating.

Science.gov (United States)

Said, Asmaa; Salah, Abeer; Fattah, Gamal Abdel

2017-05-12

Thermo-optical switches are of particular significance in communications networks where increasingly high switching speeds are required. Phase change materials (PCMs), in particular those based on paraffin wax, provide wealth of exciting applications with unusual thermally-induced switching properties, only limited by paraffin's rather low thermal conductivity. In this paper, the use of different carbon fillers as thermal conductivity enhancers for paraffin has been investigated, and a novel structure based on spot of paraffin wax as a thermo-optic switch is presented. Thermo-optical switching parameters are enhanced with the addition of graphite and graphene, due to the extreme thermal conductivity of the carbon fillers. Differential Scanning Calorimetry (DSC) and Scanning electron microscope (SEM) are performed on paraffin wax composites, and specific heat capacities are calculated based on DSC measurements. Thermo-optical switching based on transmission is measured as a function of the host concentration under conventional electric heating and laser heating of paraffin-carbon fillers composites. Further enhancements in thermo-optical switching parameters are studied under Nd:YAG laser heating. This novel structure can be used in future networks with huge bandwidth requirements and electric noise free remote aerial laser switching applications.

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

International Nuclear Information System (INIS)

Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

2000-03-01

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

The hydraulic wheel

International Nuclear Information System (INIS)

Alvarez Cardona, A.

1985-01-01

The present article this dedicated to recover a technology that key in disuse for the appearance of other techniques. It is the hydraulic wheel with their multiple possibilities to use their energy mechanical rotational in direct form or to generate electricity directly in the fields in the place and to avoid the high cost of transport and transformation. The basic theory is described that consists in: the power of the currents of water and the hydraulic receivers. The power of the currents is determined knowing the flow and east knowing the section of the flow and its speed; they are given you formulate to know these and direct mensuration methods by means of floodgates, drains and jumps of water. The hydraulic receivers or properly this hydraulic wheels that are the machines in those that the water acts like main force and they are designed to transmit the biggest proportion possible of absolute work of the water, the hydraulic wheels of horizontal axis are the common and they are divided in: you rotate with water for under, you rotate with side water and wheels with water for above. It is analyzed each one of them, their components are described; the conditions that should complete to produce a certain power and formulate them to calculate it. There are 25 descriptive figures of the different hydraulic wheels

Thermo-mechanical response and fatigue behavior of shape memory alloy

Energy Technology Data Exchange (ETDEWEB)

Kusagawa, Masaki; Asada, Yasuhide; Nakamura, Toshiya [Tokyo Univ. (Japan). Dept. of Mechanical Engineering

1998-11-01

Mechanical, thermo-mechanical and fatigue behaviors of Ni-Ti-Nb shape memory alloy (SMA) have been studied to prepare material data for a design purpose. Presented are testing devices, testing procedure and test results of monotonic tensile, recovery of inelastic deformation due to post heating (thermo-mechanical recovery) and fatigue for future use of the SMA as a structural material of nuclear incore structures. (orig.)

Thermo-mechanical response and fatigue behavior of shape memory alloy

International Nuclear Information System (INIS)

Kusagawa, Masaki; Asada, Yasuhide; Nakamura, Toshiya

1998-01-01

Mechanical, thermo-mechanical and fatigue behaviors of Ni-Ti-Nb shape memory alloy (SMA) have been studied to prepare material data for a design purpose. Presented are testing devices, testing procedure and test results of monotonic tensile, recovery of inelastic deformation due to post heating (thermo-mechanical recovery) and fatigue for future use of the SMA as a structural material of nuclear incore structures. (orig.)

A thermal-hydraulic test rig for advanced fast reactor fuel assemblies

International Nuclear Information System (INIS)

Rapier, A.C.

1989-03-01

A new design of fast reactor fuel assemblies has been proposed in which the pins are supported in grids attached to the wrapper by flexible skirts. Coolant mixing is enhanced by the skirts diverting flow into the cluster of pins at each grid. There are insufficient empirical data available for the detailed design of the skirt or for the input to computer calculations of flow and heat transfer. A test rig to provide these data has been designed and built. (author)

Boostream: a dynamic fluid flow process to assemble nanoparticles at liquid interface

Science.gov (United States)

Delléa, Olivier; Lebaigue, Olivier

2017-12-01

CEA-LITEN develops an original process called Boostream® to manipulate, assemble and connect micro- or nanoparticles of various materials, sizes, shapes and functions to obtain monolayer colloidal crystals (MCCs). This process uses the upper surface of a liquid film flowing down a ramp to assemble particles in a manner that is close to the horizontal situation of a Langmuir-Blodgett film construction. In presence of particles at the liquid interface, the film down-flow configuration exhibits an unusual hydraulic jump which results from the fluid flow accommodation to the particle monolayer. In order to master our process, the fluid flow has been modeled and experimentally characterized by optical means, such as with the moiré technique that consists in observing the reflection of a succession of periodic black-and-red fringes on the liquid surface mirror. The fringe images are deformed when reflected by the curved liquid surface associated with the hydraulic jump, the fringe deformation being proportional to the local slope of the surface. This original experimental setup allowed us to get the surface profile in the jump region and to measure it along with the main process parameters (liquid flow rate, slope angle, temperature sensitive fluid properties such as dynamic viscosity or surface tension, particle sizes). This work presents the experimental setup and its simple model, the different experimental characterization techniques used and will focus on the way the hydraulic jump relies on the process parameters.

Experimental study of thermo-hydro-mechanical behaviour of Callovo-Oxfordian Clay-stone

International Nuclear Information System (INIS)

Mohajerani, M.

2011-01-01

During the different phases of the exothermic radioactive waste deep disposal (excavation, operation) and after permanent closure, the host rock is submitted to various coupled mechanical, hydraulic and thermal phenomena. Hence, a thorough investigation of the thermo-hydro-mechanical behaviour of the rock is necessary to complete existing data and to better understand and model the short and long term behaviour of the Callovo-Oxfordian (COx) clay formation in Bure (Meuse/Haute-Marne - M/HM), considered by ANDRA as a potential host rock in France.In this work, the compression - swelling behaviour of the COx Clay-stone was first investigated by carrying out a series of high-pressure oedometric tests. The results, interpreted in terms of coupling between damage and swelling, showed that the magnitude of swelling was linked to the density of the fissures created during compression. In a second step, the hydro-mechanical and thermo-hydro-mechanical behaviour of the saturated Clay-stone under a mean stress close to the in situ one were investigated by using two devices with short drainage path (10 mm), namely a isotropic cell and a newly designed hollow cylinder triaxial cell with local displacement measurements. These devices helped to solve two majors problems related to testing very low permeability materials: i) a satisfactory previous sample saturation (indicated by good Skempton values) and ii) satisfactory drainage conditions. Some typical constitutive parameters (Skempton and Biot's coefficients, drained and undrained compressibility coefficients) have been determined at ambient temperature through isotropic compression tests that also confirmed the transverse isotropy of the Clay-stone. The consistency of the obtained parameters has been checked in a saturated poro-elastic framework. Two aspects of the thermo-hydro-mechanical behaviour of the COx Clay-stone have then been investigated through different heating tests and through drained and undrained isotropic

Validation of a CFD code for Unsteady Flows with cyclic boundary Conditions

International Nuclear Information System (INIS)

Kim, Jong-Tae; Kim, Sang-Baik; Lee, Won-Jae

2006-01-01

Currently Lilac code is under development to analyze thermo-hydraulics of a high-temperature gas-cooled reactor (GCR). Interesting thermo-hydraulic phenomena in a nuclear reactor are usually unsteady and turbulent. The analysis of the unsteady flows by using a three dimension CFD code is time-consuming if the flow domain is very large. Hopefully, flow domains commonly encountered in the nuclear thermo-hydraulics is periodic. So it is better to use the geometrical characteristics in order to reduce the computational resources. To get the benefits from reducing the computation domains especially for the calculations of unsteady flows, the cyclic boundary conditions are implemented in the parallelized CFD code LILAC. In this study, the parallelized cyclic boundary conditions are validated by solving unsteady laminar and turbulent flows past a circular cylinder

Mechanical design and thermo-hydraulic simulation of the infrared thermography diagnostic of the WEST tokamak

Energy Technology Data Exchange (ETDEWEB)

Micolon, Frédéric, E-mail: frederic.micolon@cea.fr; Courtois, Xavier; Aumeunier, Marie-Hélène; Chenevois, Jean-Pierre; Larroque, Sébastien

2015-10-15

The WEST (Tungsten (W) Environment in Steady state Tokamak) project is a partial rebuild of the Tore Supra tokamak to make it an X-point metallic environment machine aimed at testing ITER technologies in relevant plasma environment. For the safe operation of the WEST tokamak, infra-red (IR) thermography is a crucial diagnostic as it is a sound and reliable way to detect hotspots or abnormal heating patterns on the plasma facing components (PFCs). Thus WEST will be fitted with middle/short-IR (1.5–2 μm or 3–5 μm) cameras in the upper port plugs to get a full view of the critical PFCs (in particular the new lower divertor) and radio-frequency (RF) heating antennas and one camera at the equatorial level to monitor the new upper divertor and the first wall. This paper describes the design of the up-to-date optical system along with the hydraulic analysis and the thermal and mechanical finite element analysis conducted to ensure adequate heat extraction capabilities. Boundary conditions and simulation results will be presented and discussed as well as technological solutions retained.

Water Hydraulic Systems

DEFF Research Database (Denmark)

Conrad, Finn

2005-01-01

The paper presents research results using IT-Tools for CAD and dynamic modelling, simulation, analysis, and design of water hydraulic actuators for motion control of machines, lifts, cranes and robots. Matlab/Simulink and CATIA are used as IT-Tools. The contributions include results from on......-going research projects on fluid power and mechatronics based on tap water hydraulic servovalves and linear servo actuators and rotary vane actuators for motion control and power transmission. Development and design a novel water hydraulic rotary vane actuator for robot manipulators. Proposed mathematical...... modelling, control and simulation of a water hydraulic rotary vane actuator applied to power and control a two-links manipulator and evaluate performance. The results include engineering design and test of the proposed simulation models compared with IHA Tampere University’s presentation of research...

Analysis of the equalizing holes resistance in fuel assembly spike for lead-based reactor

International Nuclear Information System (INIS)

Zhang, Guangyu; Jin, Ming; Wang, Jianye; Song, Yong

2017-01-01

Highlights: • A RELAP5 model for a 10 MWth lead-based reactor was built to study the hydrodynamic characteristics between the equalizing holes in the fuel assembly spike. • Different fuel assembly total blockage scenarios and different resistances for different fuel assemblies were examined. • The inherent safety characteristics of the lead-based reactor was improved by optimizing the configuration of equalizing holes in the fuel assembly spike. - Abstract: To avoid the damage of the fuel rod cladding when a fuel assembly (FA) is totally blocked, a special configuration of the fuel assembly spike was designed with some equalizing holes in the center region which can let the coolant to flow during the totally blockage scenarios of FA. To study the hydrodynamic characteristics between the equalizing holes and an appropriate resistance, a RELAP5 model was developed for a 10 MWth lead-based reactor which used lead-bismuth as coolant. Several FA total blockage and partial core blockage scenarios were selected. The simulation results indicated that when all the FA spike equalizing holes had the same hydraulic resistance, only a narrow range of suitable equalizing holes resistances could be chosen when a FA was blocked. However, in the two or more FA blockage scenarios, there were no appropriate resistances to meet the requirement. In addition, with different FA spike equalizing holes with different resistances, a large range of suitable equalizing hole resistances could be chosen. Especially a series of suitable resistances were selected when the small power FA resistance was 1/2, 1/4, 1/8 of the large one. Under these circumstances, one, two or three FA blockages would not damage the core. These demonstrated that selecting a series of suitable hydraulic resistances for the equalizing holes could improve the safety characteristics of the reactor effectively.

Characteristic Length Scales in Fracture Networks: Hydraulic Connectivity through Periodic Hydraulic Tests

Science.gov (United States)

Becker, M.; Bour, O.; Le Borgne, T.; Longuevergne, L.; Lavenant, N.; Cole, M. C.; Guiheneuf, N.

2017-12-01

Determining hydraulic and transport connectivity in fractured bedrock has long been an important objective in contaminant hydrogeology, petroleum engineering, and geothermal operations. A persistent obstacle to making this determination is that the characteristic length scale is nearly impossible to determine in sparsely fractured networks. Both flow and transport occur through an unknown structure of interconnected fracture and/or fracture zones making the actual length that water or solutes travel undetermined. This poses difficulties for flow and transport models. For, example, hydraulic equations require a separation distance between pumping and observation well to determine hydraulic parameters. When wells pairs are close, the structure of the network can influence the interpretation of well separation and the flow dimension of the tested system. This issue is explored using hydraulic tests conducted in a shallow fractured crystalline rock. Periodic (oscillatory) slug tests were performed at the Ploemeur fractured rock test site located in Brittany, France. Hydraulic connectivity was examined between three zones in one well and four zones in another, located 6 m apart in map view. The wells are sufficiently close, however, that the tangential distance between the tested zones ranges between 6 and 30 m. Using standard periodic formulations of radial flow, estimates of storativity scale inversely with the square of the separation distance and hydraulic diffusivity directly with the square of the separation distance. Uncertainty in the connection paths between the two wells leads to an order of magnitude uncertainty in estimates of storativity and hydraulic diffusivity, although estimates of transmissivity are unaffected. The assumed flow dimension results in alternative estimates of hydraulic parameters. In general, one is faced with the prospect of assuming the hydraulic parameter and inverting the separation distance, or vice versa. Similar uncertainties exist

Investigation research on the evaluation of a coupled thermo-hydro-mechanical-chemical phenomena. 3. Result Report

International Nuclear Information System (INIS)

Ishihara, Yoshinao; Ito, Takaya; Chijimatsu, Masakazu; Amemiya, Kiyoshi; Shiozaki, Isao

2004-02-01

In order to realize a coupling analysis in the near field of the geological disposal system, the coupling analysis code 'COUPLYS (Coupling analysis system)' on the Thermo-Hydro-Mechanical-Chemical (THMC) phenomena by THAMES, Dtransu and phreeqc, which are existing analysis code, is developed in this study. And some case analyses on THMC phenomena are carried out by this code. (1) Some supporting modules, which include the transfer of dissolution concentration and total concentration (dissolution + precipitation concentration), were prepared as a functional expansion. And in order to add on the function of treat de-gases and gases diffusion, accumulation and dilution phenomena, the mass transport analysis code was modified. (2) We have modified reactive transport module to treat ionic exchange, surface reaction and kinetic reaction in the each barrier. (3) We have prepared hydraulic conductivity module of buffer material depending on change of dry density due to chemical equilibrium (dissolution and precipitation of minerals), degradation of buffer material such as Ca-type bentonite and change of concentration of NaCl solutions. After THAMES, Dtransu, phreeqc and the hydraulic conductivity module were installed in COUPLYS (Coupling Analysis), verification study was carried out to check basic function. And we have modified COUPLYS to control coupling process. (4) In order to confirm the applicability of the developed THMC analysis code (existing analysis code and COUPLYS), we have carried out case analyses on 1-dimensional and 3-dimensional model which are including vitrified waste, over-pack, buffer material and rock in the HLW near-field. (author)

Hydrophilic magnetic nanoclusters with thermo-responsive properties and their drug controlled release

International Nuclear Information System (INIS)

Meerod, Siraprapa; Rutnakornpituk, Boonjira; Wichai, Uthai; Rutnakornpituk, Metha

2015-01-01

Synthesis and drug controlled release properties of thermo-responsive magnetic nanoclusters grafted with poly(N-isopropylacrylamide) (poly(NIPAAm)) and poly(NIPAAm-co-poly(ethylene glycol) methyl ether methacrylate) (PEGMA) copolymers were described. These magnetic nanoclusters were synthesized via an in situ radical polymerization in the presence of acrylamide-grafted magnetic nanoparticles (MNPs). Poly(NIPAAm) provided thermo-responsive properties, while PEGMA played a role in good water dispersibility to the nanoclusters. The ratios of PEGMA to NIPAAm in the (co)polymerization in the presence of the MNPs were fine-tuned such that the nanoclusters with good water dispersibility, good magnetic sensitivity and thermo responsiveness were obtained. The size of the nanoclusters was in the range of 50–100 nm in diameter with about 100–200 particles/cluster. The nanoclusters were well dispersible in water at room temperature and can be suddenly agglomerated when temperature was increased beyond the lower critical solution temperature (LCST) (32 °C). The release behavior of an indomethacin model drug from the nanoclusters was also investigated. These novel magnetic nanoclusters with good dispersibility in water and reversible thermo-responsive properties might be good candidates for the targeting drug controlled release applications. - Highlights: • Nanoclusters with good water dispersibility and magnetic response were prepared. • They were grafted with thermo-responsive poly(NIPAAm) and/or poly(PEGMA). • Poly(NIPAAm) provided thermo-responsive properties to the nanoclusters. • Poly(PEGMA) provided good water dispersibilityto the nanoclusters. • Accelerated and controllable releases of a drug from the nanoclusters were shown

A low cost apparatus for measuring the xylem hydraulic conductance in plants

Directory of Open Access Journals (Sweden)

Luciano Pereira

2012-01-01

Full Text Available Plant yield and resistance to drought are directly related to the efficiency of the xylem hydraulic conductance and the ability of this system to avoid interrupting the flow of water. In this paper we described in detail the assembling of an apparatus proposed by TYREE et al. (2002, and its calibration, as well as low cost adaptations that make the equipment accessible for everyone working in this research area. The apparatus allows measuring the conductance in parts of roots or shoots (root ramifications or branches, or in the whole system, in the case of small plants or seedlings. The apparatus can also be used to measure the reduction of conductance by embolism of the xylem vessels. Data on the hydraulic conductance of eucalyptus seedlings obtained here and other reports in the literature confirm the applicability of the apparatus in physiological studies on the relationship between productivity and water stress.

Experimental Study of Thermo-hydraulic Characteristics of Surfaces with In-line Dimple Arrangement

Directory of Open Access Journals (Sweden)

S. A. Burtsev

2015-01-01

Full Text Available The paper presents a conducted experimental study of the heat exchange intensification on the surfaces covered with a regular vortex-generating relief that is an in-line array of the shallow hemispherical dimples. Using 12 configuration options with the Reynolds numbers in the range of (0.2-7.0 106 as an example, it analyses how a longitudinal and cross step of the in-line dimple array (density dimples effects on the processes of heat exchange intensification and resistance.The monocomponent strain-gauge balance allows us to define a value of the resistance coefficient by direct weighing of models (located in parallel in a flow of "relief" and smooth "reference" ones being under study. Distribution fields of heat – transfer factor are determined by recording a cooling process of the surface of studied models having high spatial and temporary resolution. All researches were conducted with one-shot data record of these thermal and hydraulic measurements for the smooth (reference surfaces and the studied surfaces covered with a regular vortex-generating relief (dimples. The error of determined parameters was no more than ±5%.The oil-sooty method allows us to visualize flow around a regular relief and obtain a flow pattern for 12 options of dimples configuration. The analysis has been carried out and a compliance of the flow patterns with the field of heat-transfer factors has been obtained.It has been found that for the in-line configuration a Reynolds analogy factor for most models is nonlinearly dependent on the Reynolds number. The friction intensification, at first, falls (to some Reynolds number and, further, starts increasing, tending to the friction intensification value with self-similarity flow around. Thus with increasing Reynolds number, the heattransfer factor intensification falls (more slowly than resistance intensification.

Quality control of FWC during assembly/commissioning on SST-1

International Nuclear Information System (INIS)

Patel, Hiteshkumar; Santra, Prosenjit; Jaiswal, Snehal

2015-01-01

First Wall components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring and port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 under going a rigorous quality control and checks at every stage of the assembly process. This paper will present the quality control and checks of FWC from commencement of assembly procedure, namely material test reports, leak testing of high temperature baked components, assembled dimensional tolerances, leak testing of all welded joints, graphite tile tightening torques, electrical continuity of passive stabilizers, and electrical isolation of passive stabilizers from vacuum vessel, baking and cooling hydraulic connections inside vacuum vessel. (author)

Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics

International Nuclear Information System (INIS)

Won-Seok Kim; Young-Gyun Kim

2000-01-01

In the sodium cooled liquid metal reactors, the design limit are imposed on the maximum temperatures of claddings and fuel pins. Thus an accurate prediction of core coolant/fuel temperature distribution is essential to the LMR core thermal-hydraulic design. The detailed subchannel thermal-hydraulic analysis code MATRA-LMR (Multichannel Analyzer for Steady States and Transients in Rod Arrays for Liquid Metal Reactors) is being developed for KALIMER core design and analysis, based on COBRA-IV-i and MATRA. The major modifications and improvements implemented into MATRA-LMR are as follows: a) nonuniform axial noding capability, b) sodium properties calculation subprogram, c) sodium coolant heat transfer correlations, and d) most recent pressure drop correlations, such as Novendstern, Chiu-Rohsenow-Todreas and Cheng-Todreas. To assess the development status of this code, the benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR for ORNL 19-pin assembly tests and EBR-II 91-pin experiments were compared to the measurements, and to SABRE4 and SLTHEN code calculation results, respectively. In this comparison, the differences are found among the three codes because of the pressure drop and the thermal mixing modellings. Finally, the major technical results of the conceptual design for the KALIMER 98.03 core have been compared with the calculations of MATRA-LMR, SABRE4 and SLTHEN codes. (author)

One-dimensional two-phase thermal hydraulics (ENSTA course); Thermo-hydraulique diphasique monodimensionnelle. Cours ENSTA

Energy Technology Data Exchange (ETDEWEB)

Olive, J

1995-11-01

This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends.

Fuel assemblies for nuclear reactors

International Nuclear Information System (INIS)

Leclercg, J.

1985-01-01

Improvements to guide tubes for the fuel assemblies of light water nuclear reactors, said assemblies being immersed in operation in the cooling water of the core of such a reactor, the guide tubes being of the type made from zircaloy and fixed at their two ends respectively to an upper end part and a lower end part made from stainless steel or Irconel and which incorporate devices for braking the fall of the control rods which they house during the rapid shutdown of the reactor, wherein the said braking devices are constituted by means for restricting the diameter of the guide tubes comprising for each guide tube a zircaloy inner sleeve spot welded to the said guide tube and whose internal diameter permits the passage, with a calibrated clearance, of the corresponding control rod, the sleeve being distributed over the lower portion of each guide tube and associated with orifices made in the actual guide tubes to produce the progressive hydraulic absorption of the end of the fall of the control rods

Finite Element Modeling of Thermo Creep Processes Using Runge-Kutta Method

Directory of Open Access Journals (Sweden)

Yu. I. Dimitrienko

2015-01-01

Full Text Available Thermo creep deformations for most heat-resistant alloys, as a rule, nonlinearly depend on stresses and are practically non- reversible. Therefore, to calculate the properties of these materials the theory of plastic flow is most widely used. Finite-element computations of a stress-strain state of structures with account of thermo creep deformations up to now are performed using main commercial software, including ANSYS package. However, in most cases to solve nonlinear creep equations, one should apply explicit or implicit methods based on the Euler method of approximation of time-derivatives. The Euler method is sufficiently efficient in terms of random access memory in computations, however this method is cumbersome in computation time and does not always provide a required accuracy for creep deformation computations.The paper offers a finite-element algorithm to solve a three-dimensional problem of thermo creep based on the Runge-Kutta finite-difference schemes of different orders with respect to time. It shows a numerical test example to solve the problem on the thermo creep of a beam under tensile loading. The computed results demonstrate that using the Runge-Kutta method with increasing accuracy order allows us to obtain a more accurate solution (with increasing accuracy order by 1 a relative error decreases, approximately, by an order too. The developed algorithm proves to be efficient enough and can be recommended for solving the more complicated problems of thermo creep of structures.

Dosimetric characteristics of Thermo-Shield material for orthovoltage photon beams

International Nuclear Information System (INIS)

Bahmaid, Mohammad; Kim, Siyong; Liu, Chihray R.; Palta, Jatinder R.

2003-01-01

Conventionally, lead has been used for field shaping in orthovoltage radiation therapy. Recently, a compensator material named Thermo-Shield was presented for field shaping in electron beams. Thermo-Shield is composed of nontoxic, high atomic weight metal particles dispersed in a thermoplastic matrix. It is manually moldable and conforms to human anatomy or any shape at temperatures of 108-132 degree sign F. It is reusable and can be continuously reshaped to better fit the treatment field. Dosimetric characteristics of thermoplastic material were studied for Philips RT250 orthovoltage photon beams ranging from 75 to 250 kVp. It was found that Thermo-Shield should be four to five times thicker than lead to achieve the same transmission (less than 5%). However, it did not cause significant degradation in penumbra. Clinical procedures for use are discussed

Thermal hydraulic considerations and mock-up tests for developing two-phase thermo-siphon loop of CARR-CNS

International Nuclear Information System (INIS)

Shejiao, Du; Qincheng, Bi; Tingkuan, Chen; Quanke, Feng

2005-01-01

The main component of the China Advanced Research Reactor Cold Neutron Source (CARR-CNS), which is under design, is a two-phase thermo-siphon loop of hydrogen. It consists of a condenser, a single tube with counter current flow avoiding flooding and a cylindrical-annulus moderator cell. The mockup tests were carried out using a full-scale loop with Freon-113, to validate the self-regulating characteristics of the loop, void fraction less than 20% in the liquid of the moderator cell and the requirements for establishing the condition under which the inner shell of the moderator cell has only vapor and the outer shell liquid. In the case of these mockup tests the density ratio of liquid to vapor and the volumetric vapor evaporation rate due to heat load are kept the same as those in normal operation of the CARR-CNS. The results show that the loop has the self-regulating characteristics and the inner shell of the moderator cell contains only vapor, the outer shell liquid. The average void fraction of the moderator cell was verified less than 20% under the volumetric vapor generation of 0.65 l/s corresponding to the nuclear heating of 800 W in the case of the liquid hydrogen. The local void fraction in the liquid hydrogen increases with the increase of the loop pressure under the condition of a constant volumetric evaporation

Simulation of thermo-mechanical effect in bulk-silicon FinFETs

OpenAIRE

Burenkov, Alex; Lorenz, Jürgen

2016-01-01

The thermo-mechanical effect in bulk-silicon FinFETs of the 14 nm CMOS technology node is studied by means of numerical simulation. The electrical performance of such devices is significantly enhanced by the intentional introduction of mechanical stress during the device processing. The thermo-mechanical effect modifies the mechanical stress distribution in active regions of the transistors when they are heated. This can lead to a modification of the electrical performance. Numerical simulati...

Power Management in Mobile Hydraulic Applications - An Approach for Designing Hydraulic Power Supply Systems

DEFF Research Database (Denmark)

Pedersen, Henrik Clemmensen

2004-01-01

Throughout the last three decades energy consumption has become one of the primary design aspects in hydraulic systems, especially for mobile hydraulic systems, as power and cooling capacity here is at limited disposal. Considering the energy usage, this is dependent on component efficiency, but ...... the hydraulic power supply in the most energy efficient way, when considering a number of load situations. Finally an example of the approach is shown to prove its validity.}......Throughout the last three decades energy consumption has become one of the primary design aspects in hydraulic systems, especially for mobile hydraulic systems, as power and cooling capacity here is at limited disposal. Considering the energy usage, this is dependent on component efficiency...

Application of Carbon Nanotube Assemblies for Sound Generation and Heat Dissipation

Science.gov (United States)

Kozlov, Mikhail; Haines, Carter; Oh, Jiyoung; Lima, Marcio; Fang, Shaoli

2011-03-01

Nanotech approaches were explored for the efficient transformation of an electrical signal into sound, heat, cooling action, and mechanical strain. The studies are based on the aligned arrays of multi-walled carbon nanotubes (MWNT forests) that can be grown on various substrates using a conventional CVD technique. They form a three-dimensional conductive network that possesses uncommon electrical, thermal, acoustic and mechanical properties. When heated with an alternating current or a near-IR laser modulated in 0.01--20 kHz range, the nanotube forests produce loud, audible sound. High generated sound pressure and broad frequency response (beyond 20 kHz) show that the forests act as efficient thermo-acoustic (TA) transducers. They can generate intense third and fourth TA harmonics that reveal peculiar interference-like patterns from ac-dc voltage scans. A strong dependence of the patterns on forest height can be used for characterization of carbon nanotube assemblies and for evaluation of properties of thermal interfaces. Because of good coupling with surrounding air, the forests provide excellent dissipation of heat produced by IC chips. Thermoacoustic converters based on forests can be used for thermo- and photo-acoustic sound generation, amplification and noise cancellation.

Hydraulic Yaw System

DEFF Research Database (Denmark)

Stubkier, Søren; Pedersen, Henrik C.; Mørkholt, M.

a hydraulic soft yaw system, which is able to reduce the loads on the wind turbine significantly. A full scale hydraulic yaw test rig is available for experiments and tests. The test rig is presented as well as the system schematics of the hydraulic yaw system....... the HAWC2 aeroelastic code and an extended model of the NREL 5MW turbine combined with a simplified linear model of the turbine, the parameters of the soft yaw system are optimized to reduce loading in critical components. Results shows that a significant reduction in fatigue and extreme loads to the yaw...... system and rotor shaft when utilizing the soft yaw drive concept compared to the original stiff yaw system. The physical demands of the hydraulic yaw system are furthermore examined for a life time of 20 years. Based on the extrapolated loads, the duty cycles show that it is possible to construct...

Hydraulic lifter of a drilling unit

Energy Technology Data Exchange (ETDEWEB)

Velikovskiy, L S; Demin, A V; Shadchinov, L M

1979-01-08

The invention refers to drilling equipment, in particular, devices for lowering and lifting operations during drilling. A hydraulic lifter of the drilling unit is suggested which contains a hydraulic cylinder, pressure line and hollow plunger whose cavities are hydraulically connected. In order to improve the reliability of the hydraulic lifter by balancing the forces of compression in the plunger of the hydraulic cylinder, a closed vessel is installed inside the plunger and rigidly connected to its ends. Its cavity is hydraulically connected to the pressure line.

Synthesis of new thermo/pH sensitive drug delivery systems based on tragacanth gum polysaccharide.

Science.gov (United States)

Hemmati, Khadijeh; Ghaemy, Mousa

2016-06-01

In this study, new pH/temperature responsive graft copolymers were synthesized based on natural Tragacanth Gum (TG) carbohydrate and their controlled drug release was investigated. Amphiphilic alkyne terminated terpolymers (mPEG-PCL-PDMAEMA-CCH)s consist of methylated poly(ethyleneglycol) (mPEG), polycaprolactone (PCL), and poly(dimethylaminoethylmethacrylate) (PDMAEMA) were synthesized by using ring opening polymerization (ROP) and atom transfer radical polymerization (ATRP), and then were grafted onto azide-functionalized TG molecules by click chemistry. Different techniques such as FT-IR, (1)H NMR, gel permeation chromatography (GPC), thermo-gravimetrical analysis (TGA) and scanning electron microscopy (SEM) were used to verify the successful synthesis of graft copolymers (TG-g-PDMAEMA-PCL-mPEG)s. The graft copolymers self-assembled to single micelles in aqueous solution and upon pH changes further assembled into micellar aggregates. These micelles were used to prepare quercetin loaded nanocarriers by probe sonication method. Size and morphology of the nanocarriers were studied by dynamic light scattering (DLS) and SEM. The in vitro release behavior of quercetin from these micelles showed pH-dependence. The results showed that release profile of quercetin best followed the first order model. Copyright © 2016 Elsevier B.V. All rights reserved.

Vibration of hydraulic machinery

CERN Document Server

Wu, Yulin; Liu, Shuhong; Dou, Hua-Shu; Qian, Zhongdong

2013-01-01

Vibration of Hydraulic Machinery deals with the vibration problem which has significant influence on the safety and reliable operation of hydraulic machinery. It provides new achievements and the latest developments in these areas, even in the basic areas of this subject. The present book covers the fundamentals of mechanical vibration and rotordynamics as well as their main numerical models and analysis methods for the vibration prediction. The mechanical and hydraulic excitations to the vibration are analyzed, and the pressure fluctuations induced by the unsteady turbulent flow is predicted in order to obtain the unsteady loads. This book also discusses the loads, constraint conditions and the elastic and damping characters of the mechanical system, the structure dynamic analysis, the rotor dynamic analysis and the system instability of hydraulic machines, including the illustration of monitoring system for the instability and the vibration in hydraulic units. All the problems are necessary for vibration pr...

Control of the wavelength dependent thermo-optic coefficients in structured fibres

DEFF Research Database (Denmark)

Sørensen, Henrik Rokkjær; Canning, J.; Lægsgaard, Jesper

2006-01-01

By controlling the fibre geometry, the fraction of optical field within the holes and the inserted material of a photonic crystal fibre, we demonstrate that it is possible to engineer any arbitrary wavelength-dependent thermo-optic coefficient. The possibility of making a fibre with a zero temper...... temperature dependent thermo-optic coefficient, ideal for packaging of structured fibre gratings, is proposed and explored....

Increasing the production efficiency and reducing the environmental impacts of hydraulic fracturing

Science.gov (United States)

Viswanathan, H. S.

2016-12-01

Shale gas is an unconventional fossil energy resource profoundly impacting US energy independence and is projected to last for at least 100 years. Production of methane and other hydrocarbons from low permeability shale involves hydraulic fracturing of rock, establishing fracture connectivity, and multiphase fluid-flow and reaction processes all of which are poorly understood. The result is inefficient extraction with many environmental concerns. A science-based capability is required to quantify the governing mesoscale fluid-solid interactions, including microstructural control of fracture patterns and the interaction of engineered fluids with hydrocarbon flow. These interactions depend on coupled thermo-hydro-mechanical-chemical (THMC) processes over scales from microns to tens of meters. Determining the key mechanisms in subsurface THMC systems has been impeded due to the lack of sophisticated experimental methods to measure fracture aperture and connectivity, multiphase permeability, and chemical exchange capacities at the high temperature, pressure, and stresses present in the subsurface. In this study, we developed and prototyped the microfluidic and triaxial core flood experiments required to reveal the fundamental dynamics of fracture-fluid interactions. The goal is transformation of hydraulic fracturing from present ad hoc approaches to science-based strategies while safely enhancing production. Specifically, we have demonstrated an integrated experimental/modeling approach that allows for a comprehensive characterization of fluid-solid interactions and develop models that can be used to determine the reservoir operating conditions necessary to gain a degree of control over fracture generation, fluid flow, and interfacial processes over a range of subsurface conditions.

Optimization in Friction Stir Welding - With Emphasis on Thermo-mechanical Aspects

DEFF Research Database (Denmark)

Tutum, Cem Celal

combined with classical single-objective and evolutionary multi-objective optimization algorithms (i.e. SQP and NSGA-II), to find the optimum process parameters (heat input, rotational and traverse welding speeds) that would result in favorable thermo-mechanical conditions for the process.......This book deals with the challenging multidisciplinary task of combining variant thermal and thermo-mechanical simulations for the manufacturing process of friction stir welding (FSW) with numerical optimization techniques in the search for optimal process parameters. The FSW process...... is characterized by multiphysics involving solid material flow, heat transfer, thermal softening, recrystallization and the formation of residual stresses. Initially, the thermal models were addressed since they in essence constitute the basis of all other models of FSW. Following this, several integrated thermo-mechanical...

Thermal stress analysis and thermo-mechanical fatigue for gas turbine blade

International Nuclear Information System (INIS)

Hyun, J. S.; Kim, B. S.; Kang, M. S.; Ha, J. S.; Lee, Y. S.

2002-01-01

The numerical analysis for gas turbine blades were carried out under several conditions by compounding temperature field, velocity field, thermal conduction of blade, and cooling heat transfer. The three types of 1,100 deg. C class 1st-stage gas turbine blades were analyzed. The analysis results are applied to the study on evaluating the remaining life for thermo-mechanical fatigue life. The thermo-mechanical fatigue experiments under out-of-phase and in-phase have been performed. The physical-based life prediction models which considered the contribution of different damage mechanisms have been applied. These models were applied to the temperature and strain rate dependences of isothermal cycling fatigue lives, and the strain-temperature history effect on the thermo-mechanical fatigue lives

Thermo-elektrische materialen : Peltier energy harvesting

NARCIS (Netherlands)

Beurden, K.M.M. (Karin); Goselink, E.A. (Erik)

2013-01-01

Thermo-elektrische materialen zijn al sinds de 19e eeuw bekend. In 1834 ontdekte de Franse natuurkundige Jean Peltier dat er warmte wordt getransporteerd van de overgang tussen twee metalen wanneer er een elektrische stroom vloeit door het grensvlak. Het grote voordeel van Peltier elementen is dat

Method of determination of thermo-acoustic coolant instability boundaries in reactor core at NPPs with WWER

International Nuclear Information System (INIS)

Skalozubov, Volodymyr; Kolykhanov, Viktor; Kovryzhkin, Yuriy

2007-01-01

The regulatory body of Ukraine, the National Atomic Energy Company and the Scientific and Production Centre have led team-works concerned with previously unstudied factors or phenomena affecting reactor safety. As a result it is determined that the thermo-acoustic coolant instability conditions can appear in the core at definite operating WWER regimes. Considerable cyclic dynamic loads affect fuel claddings over thermo-acoustic pressure oscillations. These loads can result in inadmissible cassette design damage and containment damage. Taking into account calculation and experimental research authors submit a method of on-line assessment of WWER core state concerning thermo-acoustic coolant instability. According to this method, the thermo-acoustic coolant instability appearance conditions can be estimated using normal registered parameters (pressure, temperature, heat demand etc.). At operative modes, a WWER-1000 core is stable to tracheotomies oscillations, but reduction of coolant discharge through the core for some times can result in thermo-acoustic coolant instability. Thermo-acoustic instability appears at separate transitional modes concerned with reactor scram and unloading/loading at all power units. When thermo-acoustic instability begins in transitional modes, core elements are under influence of high-frequency coolant pressure pulsations for a long time (tens of hours)

Objective and subjective measures of exercise intensity during thermo-neutral and hot yoga.

Science.gov (United States)

Boyd, Corinne N; Lannan, Stephanie M; Zuhl, Micah N; Mora-Rodriguez, Ricardo; Nelson, Rachael K

2018-04-01

While hot yoga has gained enormous popularity in recent years, owing in part to increased environmental challenge associated with exercise in the heat, it is not clear whether hot yoga is more vigorous than thermo-neutral yoga. Therefore, the aim of this study was to determine objective and subjective measures of exercise intensity during constant intensity yoga in a hot and thermo-neutral environment. Using a randomized, crossover design, 14 participants completed 2 identical ∼20-min yoga sessions in a hot (35.3 ± 0.8 °C; humidity: 20.5% ± 1.4%) and thermo-neutral (22.1 ± 0.2 °C; humidity: 27.8% ± 1.6%) environment. Oxygen consumption and heart rate (HR) were recorded as objective measures (percentage of maximal oxygen consumption and percentage of maximal HR (%HRmax)) and rating of perceived exertion (RPE) was recorded as a subjective measure of exercise intensity. There was no difference in exercise intensity based on percentage of maximal oxygen consumption during hot versus thermo-neutral yoga (30.9% ± 2.3% vs. 30.5% ± 1.8%, p = 0.68). However, exercise intensity was significantly higher during hot versus thermo-neutral yoga based on %HRmax (67.0% ± 2.3% vs. 60.8% ± 1.9%, p = 0.01) and RPE (12 ± 1 vs. 11 ± 1, p = 0.04). According to established exercise intensities, hot yoga was classified as light-intensity exercise based on percentage of maximal oxygen consumption but moderate-intensity exercise based on %HRmax and RPE while thermo-neutral yoga was classified as light-intensity exercise based on percentage of maximal oxygen uptake, %HRmax, and RPE. Despite the added hemodynamic stress and perception that yoga is more strenuous in a hot environment, we observed similar oxygen consumption during hot versus thermo-neutral yoga, classifying both exercise modalities as light-intensity exercise.

LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

International Nuclear Information System (INIS)

Khan, E.U.

1980-04-01

A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist

Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

International Nuclear Information System (INIS)

Ohnuki, A.; Kureta, M.; Liu, W.; Tamai, H.; Akimoto, H.

2004-01-01

Research and development project for investigating thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured light-water reactor technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important issues for the RMWR because of the tight-lattice configuration. The project has mainly consisted of a large-scale thermal-hydraulic test and development of analytical methods named modeling engineering. In the large-scale test, 37-rod bundle experiments can be performed. Steady-state critical power experiments have been achieved in the test facility and the experimental data reveal the feasibility of RMWR

Hydraulic Structures : Caissons

NARCIS (Netherlands)

Voorendt, M.Z.; Molenaar, W.F.; Bezuyen, K.G.

These lecture notes on caissons are part of the study material belonging to the course 'Hydraulic Structures 1' (code CTB3355), part of the Bachelor of Science education and the Hydraulic Engineering track of the Master of Science education for civil engineering students at Delft University of

Determining the Conditions for the Hydraulic Impacts Emergence at Hydraulic Systems

Directory of Open Access Journals (Sweden)

Mazurenko A.S.

2017-08-01

Full Text Available This research aim is to develop a method for modeling the conditions for the critical hydrau-lic impacts emergence on thermal and nuclear power plants’ pipeline systems pressure pumps depart-ing from the general provisions of the heat and hydrodynamic instability theory. On the developed method basis, the conditions giving rise to the reliability-critical hydraulic impacts emergence on pumps for the thermal and nuclear power plants’ typical pipeline system have been determined. With the flow characteristic minimum allowable (critical sensitivity, the flow velocity fluctuations ampli-tude reaches critical values at which the pumps working elements’ failure occurs. The critical hydrau-lic impacts emergence corresponds to the transition of the vibrational heat-hydrodynamic instability into an aperiodic one. As research revealed, a highly promising approach as to the preventing the criti-cal hydraulic impacts related to the foreground use of pumps having the most sensitive consumption (at supply network performance (while other technical characteristics corresponding to that parame-ter. The research novelty refers to the suggested method elaborated by the authors’ team, which, in contrast to traditional approaches, is efficient in determining the pump hydraulic impact occurrence conditions when the vibrational heat-hydrodynamic instability transition to the aperiodic instability.

Role of thermo-analytical techniques in compositional characterization of nuclear materials

International Nuclear Information System (INIS)

Raje, Naina

2015-01-01

The study of heat effects on different materials has a long history. Extraction of metals from the ores, pottery production, glasses making etc. are the examples, where the performance of products obtained from raw materials depends on the processing temperatures. Concrete, pottery, bricks etc., are severely damaged due to uncontrolled high temperatures. Therefore, the heating of raw materials in controlled manner is of pivotal importance to get products of the desired quality. Thermo-analytical techniques provide the information on the effect of heat under controlled heating conditions. In thermo-analytical techniques, physical properties of materials are measured as a function of temperature. Simultaneous thermo-analytical techniques are beneficial in comparison to any single thermo-analytical technique. Simultaneous techniques refer to the measurement of two or more signals on the same sample at the same time in the same instrument. Nowadays, simultaneous thermo-analytical technique are extensively in use for the analysis of materials. Ammonium diuranate (ADU) and magnesium diuranate (MDU), also known as yellowcake, are intermediate precursors in fuel fabrication process, with stringent specifications along with the need to understand its thermal behavior. In the processing of lowgrade ores, higher levels of impurities are being encountered in the leach solution that affects the properties of ADU/MDU. In order to meet the fuel specifications, quality assurance of these nuclear materials is essential. Current studies describe the application of simultaneous Thermogravimetry (TG) - differential thermal analysis (DTA) - evolved gas analysis (EGA) techniques for the compositional characterization of ADU/MDU with respect to the impurities present in the matrices

Water hydraulic applications in hazardous environments

International Nuclear Information System (INIS)

Siuko, M.; Koskinen, K.T.; Vilenius, M.J.

1996-01-01

Water hydraulic technology provides several advantages for devices operating in critical environment. Though water hydraulics has traditionally been used in very rough applications, gives recent strong development of components possibility to build more sophisticated applications and devices with similar capacity and control properties than those of oil hydraulics without the disadvantages of oil hydraulic systems. In this paper, the basic principles, possibilities and advantages of water hydraulics are highlighted, some of the most important design considerations are presented and recent developments of water hydraulic technology are presented. Also one interesting application area, ITER fusion reactor remote handling devices, are discussed. (Author)

Uncertainty Evaluation of the SFR Subchannel Thermal-Hydraulic Modeling Using a Hot Channel Factors Analysis

International Nuclear Information System (INIS)

Choi, Sun Rock; Cho, Chung Ho; Kim, Sang Ji

2011-01-01

In an SFR core analysis, a hot channel factors (HCF) method is most commonly used to evaluate uncertainty. It was employed to the early design such as the CRBRP and IFR. In other ways, the improved thermal design procedure (ITDP) is able to calculate the overall uncertainty based on the Root Sum Square technique and sensitivity analyses of each design parameters. The Monte Carlo method (MCM) is also employed to estimate the uncertainties. In this method, all the input uncertainties are randomly sampled according to their probability density functions and the resulting distribution for the output quantity is analyzed. Since an uncertainty analysis is basically calculated from the temperature distribution in a subassembly, the core thermal-hydraulic modeling greatly affects the resulting uncertainty. At KAERI, the SLTHEN and MATRA-LMR codes have been utilized to analyze the SFR core thermal-hydraulics. The SLTHEN (steady-state LMR core thermal hydraulics analysis code based on the ENERGY model) code is a modified version of the SUPERENERGY2 code, which conducts a multi-assembly, steady state calculation based on a simplified ENERGY model. The detailed subchannel analysis code MATRA-LMR (Multichannel Analyzer for Steady-State and Transients in Rod Arrays for Liquid Metal Reactors), an LMR version of MATRA, was also developed specifically for the SFR core thermal-hydraulic analysis. This paper describes comparative studies for core thermal-hydraulic models. The subchannel analysis and a hot channel factors based uncertainty evaluation system is established to estimate the core thermofluidic uncertainties using the MATRA-LMR code and the results are compared to those of the SLTHEN code

First results of the post-irradiation examination of the Ceramic Breeder materials from the Pebble Bed Assemblies Irradiation for the HCPB Blanket concept

International Nuclear Information System (INIS)

Hegeman, J.; Magielsen, A.J.; Peeters, M.; Stijkel, M.P.; Fokkens, J.H.; Laan, J.G. van der

2006-01-01

In the framework of developing the European Helium Cooled Pebble-Bed (HCPB) blanket an irradiation test of pebble-bed assemblies is performed in the HFR Petten. The experiment is focused on the thermo-mechanical behavior of the HCPB type breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. To achieve representative conditions a section of the HCPB is simulated by EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. Floating Eurofer-97 steel plates separate the pebble-beds. The structural integrity of the ceramic breeder materials is an issue for the design of the Helium Cooled Pebble Bed concept. Therefore the objective of the post irradiation examination is to study deformation of pebbles and the pebble beds and to investigate the microstructure of the ceramic pebbles from the Pebble Bed Assemblies. This paper concentrates on the Post Irradiation Examination (PIE) of the four ceramic pebble beds that have been irradiated in the Pebble Bed Assembly experiment for the HCPB blanket concept. Two assemblies with Li 4 SiO 4 pebble-beds are operated at different maximum temperatures of approximately 600 o C and 800 o C. Post irradiation computational analysis has shown that both have different creep deformation. Two other assemblies have been loaded with a ceramic breeder bed of two types of Li 2 TiO 3 beds having different sintering temperatures and consequently different creep behavior. The irradiation maximum temperature of the Li 2 TiO 3 was 800 o C. To support the first PIE result, the post irradiation thermal analysis will be discussed because thermal gradients have influence on the pebble-bed thermo-mechanical behavior and as a result it may have impact on the structural integrity of the ceramic breeder materials. (author)

Specific storage and hydraulic conductivity tomography through the joint inversion of hydraulic heads and self-potential data

Science.gov (United States)

Ahmed, A. Soueid; Jardani, A.; Revil, A.; Dupont, J. P.

2016-03-01

Transient hydraulic tomography is used to image the heterogeneous hydraulic conductivity and specific storage fields of shallow aquifers using time series of hydraulic head data. Such ill-posed and non-unique inverse problem can be regularized using some spatial geostatistical characteristic of the two fields. In addition to hydraulic heads changes, the flow of water, during pumping tests, generates an electrical field of electrokinetic nature. These electrical field fluctuations can be passively recorded at the ground surface using a network of non-polarizing electrodes connected to a high impedance (> 10 MOhm) and sensitive (0.1 mV) voltmeter, a method known in geophysics as the self-potential method. We perform a joint inversion of the self-potential and hydraulic head data to image the hydraulic conductivity and specific storage fields. We work on a 3D synthetic confined aquifer and we use the adjoint state method to compute the sensitivities of the hydraulic parameters to the hydraulic head and self-potential data in both steady-state and transient conditions. The inverse problem is solved using the geostatistical quasi-linear algorithm framework of Kitanidis. When the number of piezometers is small, the record of the transient self-potential signals provides useful information to characterize the hydraulic conductivity and specific storage fields. These results show that the self-potential method reveals the heterogeneities of some areas of the aquifer, which could not been captured by the tomography based on the hydraulic heads alone. In our analysis, the improvement on the hydraulic conductivity and specific storage estimations were based on perfect knowledge of electrical resistivity field. This implies that electrical resistivity will need to be jointly inverted with the hydraulic parameters in future studies and the impact of its uncertainty assessed with respect to the final tomograms of the hydraulic parameters.

Evaluation of thermo-mechanical properties data of carbon-based plasma facing materials

International Nuclear Information System (INIS)

Ulrickson, M.; Barabash, V.R.; Matera, R.; Roedig, M.; Smith, J.J.; Janev, R.K.

1991-03-01

This Report contains the proceedings, results and conclusions of the work done and the analysis performed during the IAEA Consultants' Meeting on ''Evaluation of thermo-mechanical properties data of carbon-based plasma facing materials'', convened on December 17-21, 1990, at the IAEA Headquarters in Vienna. Although the prime objective of the meeting was to critically assess the available thermo-mechanical properties data for certain types of carbon-based fusion relevant materials, the work of the meeting went well beyond this task. The meeting participants discussed in depth the scope and structure of the IAEA material properties database, the format of data presentation, the most appropriate computerized system for data storage, retrieval, exchange and management. The existing IAEA ALADDIN system was adopted as a convenient tool for this purpose and specific ALADDIN labelling schemes and dictionaries were established for the material properties data. An ALADDIN formatted test-file for the thermo-physical and thermo-mechanical properties of pyrolytic graphite is appended to this Report for illustrative purposes. (author)

Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

Energy Technology Data Exchange (ETDEWEB)

In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

2016-12-15

The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

Comparative study of methods to estimate hydraulic parameters in the hydraulically undisturbed Opalinus Clay (Switzerland)

Energy Technology Data Exchange (ETDEWEB)

Yu, C.; Matray, J.-M. [Institut de Radioprotection et de Sûreté Nucléaire, Fontenay-aux-Roses, (France); Yu, C.; Gonçalvès, J. [Aix Marseille Université UMR 6635 CEREGE Technopôle Environnement Arbois-Méditerranée Aix-en-Provence, Cedex 4 (France); and others

2017-04-15

The deep borehole (DB) experiment gave the opportunity to acquire hydraulic parameters in a hydraulically undisturbed zone of the Opalinus Clay at the Mont Terri rock laboratory (Switzerland). Three methods were used to estimate hydraulic conductivity and specific storage values of the Opalinus Clay formation and its bounding formations through the 248 m deep borehole BDB-1: application of a Poiseuille-type law involving petrophysical measurements, spectral analysis of pressure time series and in situ hydraulic tests. The hydraulic conductivity range in the Opalinus Clay given by the first method is 2 × 10{sup -14}-6 × 10{sup -13} m s{sup -1} for a cementation factor ranging between 2 and 3. These results show low vertical variability whereas in situ hydraulic tests suggest higher values up to 7 × 10{sup -12} m s{sup -1}. Core analysis provides economical estimates of the homogeneous matrix hydraulic properties but do not account for heterogeneities at larger scale such as potential tectonic conductive features. Specific storage values obtained by spectral analysis are consistent and in the order of 10{sup -6} m{sup -1}, while formulations using phase shift and gain between pore pressure signals were found to be inappropriate to evaluate hydraulic conductivity in the Opalinus Clay. The values obtained are globally in good agreement with the ones obtained previously at the rock laboratory. (authors)

Feasibility of fully ceramic microencapsulated (FCM) replacement fuel assembly for OPR-1000 core fully loaded with FCM fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Lee, W.J.; Lee, K.H.; Kwon, H.; Chun, J.H.; Kim, Y.M. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of); Venneri, F. [Ultra Safe Nuclear Corp., Los Alamos, NM (United States)

2014-07-01

The feasibility of replacing conventional UO{sub 2} fuel assemblies (FAs) of light water reactors with accident-tolerant fully ceramic microencapsulated (FCM) FAs has been explored referencing OPR-1000, 1000MW{sub e} PWR. An optimum FCM FA design, 16x16 FCM FA with Silicon Carbide-coated Zircaloy cladding, was selected based on core-level scoping analysis for five FCM FA design candidates screened from FA-level study. For the selected FCM FA design, detailed core following analysis from initial to equilibrium cores, initially fully loaded with the FCM FAs, was carried out to quantify core physics parameters. Using these parameters, the core thermal-hydraulics and coated fuel particle performance of the FCM core was assessed, and the safety margin and accident-tolerance of the FCM core was evaluated for limiting design- and beyond design-basis-accidents. From the study, it has been demonstrated that the FCM fuel is a viable option in replacing the OPR-1000 core with enhanced safety and accident tolerance while maintaining the core neutronics, thermal-hydraulics and mechanical compatibility. (author)

Characterize the hydraulic behaviour of grate inlet in urban drainage to prevent the urban's flooding

Science.gov (United States)

Tellez Alvarez, Jackson David; Gomez, Manuel; Russo, Beniamino; Redondo, Jose M.

2016-04-01

image velocimetry to measure surface flow velocities has been developed and validated with the experiments assays with the grate inlets [3 - 4]. Indeed, the Methodology carried out can become a useful tools to understand the hydraulics behavior of the flow approaching the inlet where the traditional measuring equipment have serious problems and limitations [5 - 6]. References [1] Gómez, M., Macchione, F. and Russo, B. (2006). Inlet systems and risk criteria associated to street runoff application to urban drainage catchments. 27 Corso di aggiornamiento in techniche per la difesa dall'inquinamento. [2] Russo, B., Gómez, M., & Tellez, J. (2013). Methodology to Estimate the Hydraulic Efficiency of Nontested Continuous Transverse Grates. Journal of Irrigation and Drainage Engineering, 139(10), 864-871. doi:10.1061/(ASCE)IR.1943-4774.0000625 [3] DigiFlow. User Guide. (2012), (June). [4] Vila, T., Tellez, J., Sanchez, J.M., Sotillos, L., Diez, M., and Redondo, J.M. (2014). Diffusion in fractal wakes and convective thermoelectric flows. Geophysical Research Abstracts - EGU General Assembly 2014. [5] Tellez, J., Gómez, M., Russo, B. and Redondo, J.M. (2014). A simple technique to measuring surface flow velocity to analyze the behavior of fields velocities in hydraulics engineer applications. Geophysical Research Abstracts - EGU General Assembly 2015. [6] Tellez, J., Gómez, M. and Russo, B. (2015). Técnica para la obtención del campo de velocidad del flujo superficial en proximidad de rejas de alcantarillado. IV Jornadas de Ingeniería del Agua. La precipitación y los procesos erosivos.

3D Hydraulic tomography from joint inversion of the hydraulic heads and self-potential data. (Invited)

Science.gov (United States)

Jardani, A.; Soueid Ahmed, A.; Revil, A.; Dupont, J.

2013-12-01

Pumping tests are usually employed to predict the hydraulic conductivity filed from the inversion of the head measurements. Nevertheless, the inverse problem is strongly underdetermined and a reliable imaging requires a considerable number of wells. We propose to add more information to the inversion of the heads by adding (non-intrusive) streaming potentials (SP) data. The SP corresponds to perturbations in the local electrical field caused directly by the fow of the ground water. These SP are obtained with a set of the non-polarising electrodes installed at the ground surface. We developed a geostatistical method for the estimation of the hydraulic conductivity field from measurements of hydraulic heads and SP during pumping and injection experiments. We use the adjoint state method and a recent petrophysical formulation of the streaming potential problem in which the streaming coupling coefficient is derived from the hydraulic conductivity allowed reducing of the unknown parameters. The geostatistical inverse framework is applied to three synthetic case studies with different number of the wells and electrodes used to measure the hydraulic heads and the streaming potentials. To evaluate the benefits of the incorporating of the streaming potential to the hydraulic data, we compared the cases in which the data are coupled or not to map the hydraulic conductivity. The results of the inversion revealed that a dense distribution of electrodes can be used to infer the heterogeneities in the hydraulic conductivity field. Incorporating the streaming potential information to the hydraulic head data improves the estimate of hydraulic conductivity field especially when the number of piezometers is limited.

Parametric thermo-hydraulic analysis of the TF system of JT-60SA during fast discharge

International Nuclear Information System (INIS)

Polli, Gian Mario; Lacroix, Benoit; Zani, Louis; Besi Vetrella, Ugo; Cucchiaro, Antonio

2013-01-01

Highlights: • We modeled the central clock-wise pancake of JT-60SA TF magnet at the EOB. • We simulated a quench followed by a fast discharge. • We evaluated the temperature and pressure rises in the nominal configuration. • We evaluated the effect of several parameter changes on the thermal-hydraulic response of the system. -- Abstract: The evolution of the conductor temperature and of the helium pressure of the central pancake of the TF superconducting magnet of the JT-60SA tokamak in a quench scenario are here discussed. The quench is triggered by a heat disturbance applied at the end of burning and followed by a fast safety discharge. A parametric study aimed at assessing the robustness of the calculation is also addressed with special regard to the voltage threshold, used to define the occurrence of the quench, and to the time delay, that cover all the possible delays in the fast discharge after quench detection. Finally, due to sensitivity analyses the influences of different parameters were assessed: the material properties of the strands (RRR, copper fraction), the magnitude and the spatial length of the triggering disturbance and the magnetic field distribution. The numerical evaluations were performed in the framework of the Broader Approach Agreement in collaboration with CEA, ENEA and the JT-60SA European Home Team using the 1D code Gandalf [1

Comparison of square and hexagonal fuel lattices for high conversion PWRs

International Nuclear Information System (INIS)

Kotlyar, D.; Shwageraus, E.

2011-01-01

This paper reports on an investigation into fuel design choices of a PWR operating in a self sustainable Th- 233 U fuel cycle. Achieving such self-sustainable with respect to fissile material fuel cycle would practically eliminate concerns over nuclear fuel supply hundreds of years into the future. Moreover, utilization of light water reactor technology and its associated vast experience would allow faster deployment of such fuel cycle without immediate need for development of fast reactor technology, which tends to be more complex and costly. In order to evaluate feasibility of this concept, two types of fuel assembly lattices were considered: square and hexagonal. The hexagonal lattice may offer some advantages over the square one. For example, the fertile blanket fuel can be packed more tightly reducing the blanket volume fraction in the core and potentially allowing to achieve higher core average power density. Furthermore, hexagonal lattice may allow more uniform leakage of neutrons from fissile to fertile regions and therefore more uniform neutron captures in thorium blanket. The calculations were carried out with Monte-Carlo based BGCore system, which includes neutronic, fuel depletion and thermo-hydraulic modules. The results were compared to those obtained from Serpent Monte-Carlo code and deterministic fuel assembly transport code BOXER. One of the major design challenges associated with the square seed-blanket concept is high power peaking due to the high concentration of fissile material in the seed region. In order to explore feasibility of the studied designs, the calculations were extended to include 3D fuel assembly analysis with thermal-hydraulic feedback. The coupled neutronic - thermal-hydraulic calculations were performed with BGCore code system. The analysis showed that both hexagonal and square seed-blanket fuel assembly designs have a potential of achieving net breeding. While no major neutronic advantages were observed for either fuel

Hydraulic Shearing and Hydraulic Jacking Observed during Hydraulic Stimulations in Fractured Geothermal Reservoir in Pohang, Korea

Science.gov (United States)

Min, K. B.; Park, S.; Xie, L.; Kim, K. I.; Yoo, H.; Kim, K. Y.; Choi, J.; Yoon, K. S.; Yoon, W. S.; Lee, T. J.; Song, Y.

2017-12-01

Enhanced Geothermal System (EGS) relies on sufficient and irreversible enhancement of reservoir permeability through hydraulic stimulation and possibility of such desirable change of permeability is an open question that can undermine the universality of EGS concept. We report results of first hydraulic stimulation campaign conducted in two deep boreholes in fractured granodiorite geothermal reservoir in Pohang, Korea. Borehole PX-1, located at 4.22 km, was subjected to the injection of 3,907 m3 with flow rate of up to 18 kg/s followed by bleeding off of 1,207 m3. The borehole PX-2, located at 4.35 km, was subjected to the injection of 1,970 m3 with flow rate of up to 46 kg/sIn PX-1, a sharp distinct decline of wellhead pressure was observed at around 16 MPa of wellhead pressure which was similar to the predicted injection pressure to induce hydraulic shearing. Injectivity interpretation before and after the hydraulic shearing indicates that permanent increase of permeability was achieved by a factor of a few. In PX-2, however, injectivity was very small and hydraulic shearing was not observed due possibly to the near wellbore damage made by the remedying process of lost circulation such as using lost circulation material during drilling. Flow rate of larger than 40 kg/s was achieved at very high well head pressure of nearly 90 MPa. Hydraulic jacking, that is reversible opening and closure of fracture with change of injection pressure, was clearly observed. Although sharp increase of permeability due to fracture opening was achieved with elevated injection pressure, the increased permeability was reversed with decreased injection pressure.Two contrasting response observed in the same reservoir at two different boreholes which is apart only 600 m apart provide important implication that can be used for the stimulation strategy for EGS.This work was supported by the New and Renewable Energy Technology Development Program of the Korea Institute of Energy Technology

Preparation, optical properties and 1 Multiplication-Sign 2 polymeric thermo-optic switch of polyurethane-urea

Energy Technology Data Exchange (ETDEWEB)

Qiu, Fengxian, E-mail: fxqiuchem@163.com [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 212013 (China); Cao, Zhijuan [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 212013 (China); Cao, Guorong; Guan, Yijun; Shen, Qiang [Department of Physics, Jiangsu University, Zhenjiang 212013 (China); Wang, Qing; Yang, Dongya [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 212013 (China)

2012-08-15

A polyurethane-urea (PUU) containing azo chromophore, polyether polyol (NJ-220) and isophorone diisocyanate (IPDI) was prepared. The structure, thermal property and mechanical properties of obtained PUU were characterized and measured by the UV-visible spectroscopy, Fourier transform infrared, Differential scanning calorimetry (DSC) and thermogravimetric analysis (TGA). The refractive index (n) of PUU was determined at different temperature and wavelength (532 nm, 650 nm and 850 nm) using attenuated total reflection (ATR) technique, and the thermo-optic coefficients (dn/dT) were -5.3643 Multiplication-Sign 10{sup -4} Degree-Sign C{sup -1}, -5.2500 Multiplication-Sign 10{sup -4} Degree-Sign C{sup -1} and -4.6071 Multiplication-Sign 10{sup -4} Degree-Sign C{sup -1}, respectively. Using the Charge Coupled Device (CCD) digital imaging devices, transmission loss of PUU was measured and the value was 0.659 dB cm{sup -1}. A 1 Multiplication-Sign 2 polymeric thermo-optic switch based on the thermo-optic effect of PUU film was proposed. With branching angle of 0.143 Degree-Sign and the finite difference beam propagation method (FD-BPM), the polymeric thermo-optic switch was simulated. The result showed that the power consumption of the thermo-optic switch could be only 0.72 mW, and the response time of the switch was about 3.0 ms. The obtained PUU has a significant improvement in reducing the power consumption and response time compared with those of the normal polymeric thermo-optic switches. -- Highlights: Black-Right-Pointing-Pointer Preparation and structural characterization of a novel azo polyurethane-urea (PUU). Black-Right-Pointing-Pointer The mechanical performance and thermal property of PUU film. Black-Right-Pointing-Pointer The thermo-optic property, transmission loss and dispersion property of PUU. Black-Right-Pointing-Pointer Proposed a new 1 Multiplication-Sign 2 polymeric thermo-optic switch.

Hydraulic Stability of Accropode Armour

DEFF Research Database (Denmark)

Jensen, T.; Burcharth, H. F.; Frigaard, Peter

The present report describes the hydraulic model tests of Accropode armour layers carried out at the Hydraulics Laboratory at Aalborg University from November 1995 through March 1996. The objective of the model tests was to investigate the hydraulic stability of Accropode armour layers...... with permeable core (crushed granite with a gradation of 5-8 mm). The outcome of this study is described in "Hydraulic Stability of Single-Layer Dolos and Accropode Armour Layers" by Christensen & Burcharth (1995). In January/February 1996, Research Assistant Thomas Jensen carried out a similar study...

Assimilation of temperature and hydraulic gradients for quantifying the spatial variability of streambed hydraulics

Science.gov (United States)

Huang, Xiang; Andrews, Charles B.; Liu, Jie; Yao, Yingying; Liu, Chuankun; Tyler, Scott W.; Selker, John S.; Zheng, Chunmiao

2016-08-01

Understanding the spatial and temporal characteristics of water flux into or out of shallow aquifers is imperative for water resources management and eco-environmental conservation. In this study, the spatial variability in the vertical specific fluxes and hydraulic conductivities in a streambed were evaluated by integrating distributed temperature sensing (DTS) data and vertical hydraulic gradients into an ensemble Kalman filter (EnKF) and smoother (EnKS) and an empirical thermal-mixing model. The formulation of the EnKF/EnKS assimilation scheme is based on a discretized 1D advection-conduction equation of heat transfer in the streambed. We first systematically tested a synthetic case and performed quantitative and statistical analyses to evaluate the performance of the assimilation schemes. Then a real-world case was evaluated to calculate assimilated specific flux. An initial estimate of the spatial distributions of the vertical hydraulic gradients was obtained from an empirical thermal-mixing model under steady-state conditions using a constant vertical hydraulic conductivity. Then, this initial estimate was updated by repeatedly dividing the assimilated specific flux by estimates of the vertical hydraulic gradients to obtain a refined spatial distribution of vertical hydraulic gradients and vertical hydraulic conductivities. Our results indicate that optimal parameters can be derived with fewer iterations but greater simulation effort using the EnKS compared with the EnKF. For the field application in a stream segment of the Heihe River Basin in northwest China, the average vertical hydraulic conductivities in the streambed varied over three orders of magnitude (5 × 10-1 to 5 × 102 m/d). The specific fluxes ranged from near zero (qz < ±0.05 m/d) to ±1.0 m/d, while the vertical hydraulic gradients were within the range of -0.2 to 0.15 m/m. The highest and most variable fluxes occurred adjacent to a debris-dam and bridge pier. This phenomenon is very likely

NRC wants plant-specific responses on Thermo-Lag

International Nuclear Information System (INIS)

Anon.

1994-01-01

Dissatisfied with recent industry-backed efforts to assure fire safety at nuclear power plants, the Nuclear Regulatory Commission announced on November 24 that it would direct all nuclear plant owners to specify the actions they would take to assure that the use of the Thermo-Lag 330 fire barrier material would not lead to insufficient protection of electrical cables connected to safe-shutdown systems. Previously, the NRC had been content to let the matter wait until tests sponsored by the Nuclear Management and Resources Council (Numarc) could show whether Thermo-Lag, used and installed in certain ways, would provide sufficient protection, but the NRC and Numarc have disagreed over the test methodology, and the Numarc tests are now considered to be several months behind schedule

Black Holes Versus Firewalls and Thermo-Field Dynamics

Science.gov (United States)

Chowdhury, Borun D.

2013-09-01

In this paper, we examine the implications of the ongoing black holes versus firewalls debate for the thermo-field dynamics of black holes by analyzing a conformal field theory (CFT) in a thermal state in the context of anti-de Sitter/CFT. We argue that the thermo-field doubled copy of the thermal CFT should be thought of not as a fictitious system, but as the image of the CFT in the heat bath. In case of strong coupling between the CFT and the heat bath, this image allows for free infall through the horizon and the system is described by a black hole. Conversely, firewalls are the appropriate dual description in case of weak interaction of the CFT with its heat bath.

Thermo-electrical systems for the generation of electricity

International Nuclear Information System (INIS)

Bitschi, A.; Froehlich, K.