Transient thermal camouflage and heat signature control

Science.gov (United States)

Yang, Tian-Zhi; Su, Yishu; Xu, Weikai; Yang, Xiao-Dong

2016-09-01

Thermal metamaterials have been proposed to manipulate heat flux as a new way to cloak or camouflage objects in the infrared world. To date, however, thermal metamaterials only operate in the steady-state and exhibit detectable, transient heat signatures. In this letter, the theoretical basis for a thermal camouflaging technique with controlled transient diffusion is presented. This technique renders an object invisible in real time. More importantly, the thermal camouflaging device instantaneously generates a pre-designed heat signature and behaves as a perfect thermal illusion device. A metamaterial coating with homogeneous and isotropic thermal conductivity, density, and volumetric heat capacity was fabricated and very good camouflaging performance was achieved.

Transient thermal characteristics of a core channel in a molten salt reactor

International Nuclear Information System (INIS)

Sakashita, H.; Ishiguro, R.; Sugiyama, K.

1987-01-01

The present paper deals with the thermal characteristics of Molten Salt Reactor (MSR). Analyses of the fundamental behavior of internal heat generating fluid and graphite contiguous to the fluid are performed. As a result, it is known that the transient thermal characteristics of MSR differ fundamentally from those of a solid-fuel reactor, and the simplified method of thermal analysis which is commonly used for solid-fuel reactors gives optimistic predictions than the actual phenomena. (author)

Enhancing the ABAQUS Thermomechanics Code to Simulate Steady and Transient Fuel Rod Behavior

International Nuclear Information System (INIS)

Williamson, R.L.; Knoll, D.A.

2009-01-01

A powerful multidimensional fuels performance capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. The various modeling capabilities are demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multi-pellet fuel rod, during both steady and transient operation. Computational results demonstrate the importance of a multidimensional fully-coupled thermomechanics treatment. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermo-mechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

International Nuclear Information System (INIS)

El-Morshedy, Salah El-Din; Hassanein, Ahmed

2009-01-01

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m 2 plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

Energy Technology Data Exchange (ETDEWEB)

El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu

2009-12-15

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

Thermal fatigue behavior of valves

International Nuclear Information System (INIS)

Moinereau, D.; Scliffet, L.; Capion, J.C.; Genette, P.

1991-01-01

This paper reports that valves of pressurized water reactors are exposed to thermal shocks during transient operations. The numerous thermal shock tests performed on valves on the EDF test facilities have shown the sensibility of fillets and geometrical discontinuities to thermal fatigue: cracks can appear in those areas and grow through the valve body. Valves systems designated as level 1 must be designed to withstand fatigue up to the second isolation valve: the relevant rule is specified in the paragraph B 3500 of the French RCCM code. It is a simplified method which doesn't require finite element calculations. Many valve systems have been designed according to this rule and have been operated without accident. However, in one case, important cracks were found in the fillet of a check-valve after numerous thermal shocks. Calculation of the valve's behavior according to the RCCM code to estimate the fatigue damage resulting from thermal shocks led to a low damage factor, which doesn't agree with the experimental results. This was confirmed by new testings and showed the inadequacy of B 3500 rule for thermal transients. On this base a new rule is proposed to estimate fatigue damage resulting from thermal shocks. An experimental program has been realized to validate this rule. Axisymetrical analytical mock-ups with different geometries and one check-valve in austenitic stainless steel 316 L have been submitted to hot thermal shocks of 210 degrees C magnitude

Apparatus and method for transient thermal infrared emission spectrometry

Science.gov (United States)

McClelland, John F.; Jones, Roger W.

1991-12-24

A method and apparatus for enabling analysis of a solid material (16, 42) by applying energy from an energy source (20, 70) top a surface region of the solid material sufficient to cause transient heating in a thin surface layer portion of the solid material (16, 42) so as to enable transient thermal emission of infrared radiation from the thin surface layer portion, and by detecting with a spectrometer/detector (28, 58) substantially only the transient thermal emission of infrared radiation from the thin surface layer portion of the solid material. The detected transient thermal emission of infrared radiation is sufficiently free of self-absorption by the solid material of emitted infrared radiation, so as to be indicative of characteristics relating to molecular composition of the solid material.

Mitigation of thermal transients by tube bundle inlet plenum design

International Nuclear Information System (INIS)

Oras, J.J.; Kasza, K.E.

1984-06-01

A multiphase program aimed at investigating the importance of thermal buoyancy to LMFBR steam-generator and heat-exchanger thermal hydraulics under low-flow transient conditions is being conducted in the Argonne Mixing Components Test Facility (MCTF) on a 60 0 sector shell-side flow model of the Westinghouse straight-tube steam generator being developed under the US/DOE large-component development program. A series of shell-side constant-flow thermal-downramp transient tests have been conducted focusing on the phenomenon of thermal-buoyancy-induced-flow channeling. In addition, it was discovered that a shell-inlet flow-distribution plenum can play a significant role in mitigating the severity of a thermal transient entering a steam generator or heat exchanger

Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

Science.gov (United States)

Domalapally, Phani; Di Caro, Marco

2018-05-01

Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

Analysis of piping response to thermal and operational transients

International Nuclear Information System (INIS)

Wang, C.Y.

1987-01-01

The reactor piping system is an extremely complex three-dimensional structure. Maintaining its structural integrity is essential to the safe operation of the reactor and the steam-supply system. In the safety analysis, various transient loads can be imposed on the piping which may cause plastic deformation and possible damage to the system, including those generated from hydrodynamic wave propagations, thermal and operational transients, as well as the seismic events. At Argonne National Laboratory (ANL), a three-dimensional (3-D) piping code, SHAPS, aimed for short-duration transients due to wave propagation, has been developed. Since 1984, the development work has been shifted to the long-duration accidents originating from the thermal and operational transient. As a result, a new version of the code, SHAPS-2, is being established. This paper describes many features related to this later development. To analyze piping response generated from thermal and operational transients, a 3-D implicit finite element algorithm has been developed for calculating the hoop, flexural, axial, and torsional deformations induced by the thermomechanical loads. The analysis appropriately accounts for stresses arising from the temperature dependence of the elastic material properties, the thermal expansion of the materials, and the changes in the temperature-dependent yield surface. Thermal softening, failure, strain rate, creep, and stress ratching can also be considered

Mitigation method of thermal transient stress by thermalhydraulic-structure total analysis

International Nuclear Information System (INIS)

Kasahara, Naoto; Jinbo, Masakazu; Hosogai, Hiromi

2003-01-01

This study proposes a rational evaluation and mitigation method of thermal transient loads in fast reactor components by utilizing relationships among plant system parameters and stresses induced by thermal transients of plants. A thermalhydraulic-structure total analysis procedure helps us to grasp relationship among system parameters and thermal stresses. Furthermore, it enables mitigation of thermal transient loads by adjusting system parameters. In order to overcome huge computations, a thermalhydraulic-structure total analysis code and the Design of Experiments methodology are utilized. The efficiency of the proposed mitigation method is validated through thermal stress evaluation of an intermediate heat exchanger in Japanese demonstration fast reactor. (author)

Transient stress control of aeroengine disks based on active thermal management

International Nuclear Information System (INIS)

Ding, Shuiting; Wang, Ziyao; Li, Guo; Liu, Chuankai; Yang, Liu

2016-01-01

Highlights: • The essence of cooling in turbine system is a process of thermal management. • Active thermal management is proposed to control transient stress of disks. • The correlation between thermal load and transient stress of disks is built. • Stress level can be declined by actively adjusting the thermal load distribution. • Artificial temperature gradient can be used to counteract stress from rotating. - Abstract: The physical essence of cooling in the turbine system is a process of thermal management. In order to overcome the limits of passive thermal management based on thermal protection, the concept of active thermal management based on thermal load redistribution has been proposed. On this basis, this paper focuses on a near real aeroengine disk during a transient process and studies the stress control mechanism of active thermal management in transient conditions by a semi-analytical method. Active thermal management is conducted by imposing extra heating energy on the disk hub, which is represented by the coefficient of extra heat flow η. The results show that the transient stress level can be effectively controlled by actively adjusting the thermal load distribution. The decline ratio of the peak equivalent stress of the disk hub can be 9.0% for active thermal management load condition (η = 0.2) compared with passive condition (η = 0), even at a rotation speed of 10,000 r/min. The reason may be that the temperature distribution of the disk turns into an artificial V-shape because of the extra heating energy on the hub, and the resulting thermal stresses induced by the negative temperature gradients counteract parts of the stress from rotating.

Transient thermal analysis of Vega launcher structures

Energy Technology Data Exchange (ETDEWEB)

Gori, F. [University of Rome ' Tor Vergata' , Rome (Italy); De Stefanis, M. [Thales Alenia Space Italia, Rome (Italy); Worek, W.M. [University of Illinois at Chicago, Chicago (United States)], E-mail: wworek@uic.edu; Minkowycz, W.J. [University of Illinois at Chicago, Chicago (United States)

2008-12-15

A transient thermal analysis is carried out to verify the base cover thermal protection system of Vega 2nd stage Solid Rocket Motor (SRM) and the flange coupling of the inter-stage 2/3. The analysis is performed with a finite element code. The work has developed suitable numerical Fortran subroutines to assign radiation and convection boundary conditions. The thermal behaviour of the structures is presented.

Simulation of the effects of grain boundary fission gas during thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Emerson, J.E.; Beiersdorf, B.A.

1984-11-01

This report presents the results of an initial set of out-of-cell transient heating experiments performed on unirradiated UO 2 pellets fabricated to simulate the effect of grain boundary fission gas on fuel swelling and cladding failure. The fabrication involved trapping high-pressure argon on internal pores by sintering annular UO 2 pellets in a hot isostatic press (HIP). The pellet stack was subjected to two separate transients (DGF83-03A and -03B). Figures show photomicrographs of HIPped and non-HIPped UO 2 , respectively, and the adjacent cladding after DGF83-03B. Fuel melting occurred at the center of both the HIPped and non-HIPped pellets; however, a dark ring is present near the center in the HIPped fuel but not in the non-HIPped fuel. This dark band is a high-porosity region due to increased grain boundary/edge swelling in that pellet. In contrast, grain boundary/edge swelling did not occur in the non-HIPped pellets. Thus, the presence of the high-pressure argon trapped on internal pores during sintering in the HIP altered the microstructural behavior. Results of these preliminary tests indicate that the microstructural behavior of HIPped fuel during thermal transients is different from the behavior of conventionally fabricated fuel

Transient thermal performance analysis of micro heat pipes

International Nuclear Information System (INIS)

Liu, Xiangdong; Chen, Yongping

2013-01-01

A theoretical analysis of transient fluid flow and heat transfer in a triangular micro heat pipes (MHP) has been conducted to study the thermal response characteristics. By introducing the system identification theory, the quantitative evaluation of the MHP's transient thermal performance is realized. The results indicate that the evaporation and condensation processes are both extended into the adiabatic section. During the start-up process, the capillary radius along axial direction of MHP decreases drastically while the liquid velocity increases quickly at the early transient stage and an approximately linear decrease in wall temperature arises along the axial direction. The MHP behaves as a first-order LTI control system with the constant input power as the 'step input' and the evaporator wall temperature as the 'output'. Two corresponding evaluation criteria derived from the control theory, time constant and temperature constant, are able to quantitatively evaluate the thermal response speed and temperature level of MHP under start-up, which show that a larger triangular groove's hydraulic diameter within 0.18–0.42 mm is able to accelerate the start-up and decrease the start-up temperature level of MHP. Additionally, the MHP starts up fastest using the fluid of ethanol and most slowly using the working fluid of methanol, and the start-up temperature reaches maximum level for acetone and minimum level for the methanol. -- Highlights: • Transient thermal response of micro heat pipe is simulated by an improved model. • Control theory is introduced to quantify the thermal response of micro heat pipe. • Evaluation criteria are proposed to represent thermal response of micro heat pipe. • Effects of groove dimensions and working fluids on start-up of micro heat pipe are evaluated

Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

Energy Technology Data Exchange (ETDEWEB)

Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

2017-11-01

A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

The simulation of transients in thermal plant. Part II: Applications

International Nuclear Information System (INIS)

Morini, G.L.; Piva, S.

2008-01-01

This paper deals with the simulation of the transients of thermal plant with control systems. In the companion paper forming part I of this article [G.L. Morini, S. Piva, The simulation of transients in thermal plant. Part I: Mathematical model, Applied Thermal Engineering 27 (2007) 2138-2144] it has been described how a 'thermal-library' of customised blocks can be built and used, in an intuitive way, to study the transients of any kind of thermal plant. Each component of plant such as valves, boilers, and pumps, is represented by a single block. In this paper, the 'thermal-library' approach is demonstrated by the analysis of the dynamic behaviour of a central heating plant of a typical apartment house during a sinusoidal variation of the external temperature. A comparison of the behaviour of such a plant with three way valve working either in flow rate or in temperature control, is presented and discussed. Finally, the results show the delaying effect of the thermal capacity of the building on the performance of the control system

Application of Thermal Network Model to Transient Thermal Analysis of Power Electronic Package Substrate

Directory of Open Access Journals (Sweden)

Masaru Ishizuka

2011-01-01

Full Text Available In recent years, there is a growing demand to have smaller and lighter electronic circuits which have greater complexity, multifunctionality, and reliability. High-density multichip packaging technology has been used in order to meet these requirements. The higher the density scale is, the larger the power dissipation per unit area becomes. Therefore, in the designing process, it has become very important to carry out the thermal analysis. However, the heat transport model in multichip modules is very complex, and its treatment is tedious and time consuming. This paper describes an application of the thermal network method to the transient thermal analysis of multichip modules and proposes a simple model for the thermal analysis of multichip modules as a preliminary thermal design tool. On the basis of the result of transient thermal analysis, the validity of the thermal network method and the simple thermal analysis model is confirmed.

On transient irradiation behavior of HTGR fuel particles

International Nuclear Information System (INIS)

Mortenson, S.C.; Okrent, D.

1977-01-01

An examination of HTGR TRISO coated fuel particles was made in which the particles' stress-strain histories were determined during both steady-state and transient operating conditions. The basis for the examination was a modified version of a computer code written by Kaae which assumed spherical symmetry, isotropic thermal expansion, isotropic elastic constants, time-temperature-irradiation invariant materials properties, and steady state operation during particle exposure. Additionally, the Kaae code modelled potential separation of layers at the SiC-inner PyC interface and considered that several entrapped fission products could exist in either the gaseous or solid state, dependent upon particle operating conditions. Using the modified code which modelled transient behavior in a quasi-static fashion, a series of both steady-state and transient operating condition computer simulations was made. For the former set of runs, a candidate set of particle dimensions and a nominal set of materials' properties was assumed. Layer thicknesses were assumed to be normally distributed about the nominal thickenesses and a probability distribution of SiC tensile stresses was generated; sensitivity of the stress distribution to assumed standard deviation of the layer thicknesses was acute. Further, this series of steady-state runs demonstrated that for certain combinations of the assumed PyC-SiC bond interface strength and irradiation-induced creep constant, anomalous predicted stresses may be obtained in the PyC layers. The steady-state runs also suggest that transient behavior would most likely not be significant at fast neutron exposures below about 10 21 NVT due to both low fission gas pressure and likely beneficial interface separation

Transient thermal stresses of work roll by coupled thermoelasticity

Science.gov (United States)

Lai, W. B.; Chen, T. C.; Weng, C. I.

1991-01-01

A numerical method, based on a two-dimensional plane strain model, is developed to predict the transient responses (that include distributions of temperature, thermal deformation, and thermal stress) of work roll during strip rolling by coupled thermoelasticity. The method consists of discretizing the space domain of the problem by finite element method first, and then treating the time domain by implicit time integration techniques. In order to avoid the difficulty in analysis due to relative movement between work roll and its thermal boundary, the energy equation is formulated with respect to a fixed Eulerian reference frame. The effect of thermoelastic coupling term, that is generally disregarded in strip rolling, can be considered and assessed. The influences of some important process parameters, such as rotational speed of the roll and intensity of heat flux, on transient solutions are also included and discussed. Furthermore, since the stress history at any point of the roll in both transient and steady state could be accurately evaluated, it is available to perform the analysis of thermal fatigue for the roll by means of previous data.

Thermal hydraulic behavior of SCWR sliding pressure startup

International Nuclear Information System (INIS)

Fu Shengwei; Zhou Chong; Xu Zhihong; Yang Yanhua

2011-01-01

The modification to ATHLET-SC code is introduced in this paper, which realizes the simulation of trans-critical transients using two-phase model. With the modified code, the thermal-hydraulic dynamic behavior of the mixed SCWR core during the startup process is simulated. The startup process is similar to the design of SCLWR-H sliding pressure startup. The results show that maximum temperature of cladding-surface does not exceed 650℃ in the whole startup process, and the sudden change of water properties in the trans-critical transients will not cause harmful influence to the heat transfer of the fuel cladding. (authors)

Lumped thermal capacitance analysis of transient heat conduction ...

African Journals Online (AJOL)

Lumped thermal capacitance analysis has been undertaken to investigate the transient temperature variations, associated induced thermal stress distributions, and the structural integrity of Ghana Research Reactor-1 (GHAR R-1) vessel after 15 years of operation. The beltline configuration of the cylindrical vessel of the ...

47 CFR 90.214 - Transient frequency behavior.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Transient frequency behavior. 90.214 Section 90... PRIVATE LAND MOBILE RADIO SERVICES General Technical Standards § 90.214 Transient frequency behavior... Behavior for Equipment Designed to Operate on 25 kHz Channels t1 4 ±25.0 kHz 5.0 ms 10.0 ms t2 ±12.5 kHz 20...

Transient behavior of flare-associated solar wind. II - Gas dynamics in a nonradial open field region

Science.gov (United States)

Nagai, F.

1984-01-01

Transient behavior of flare-associated solar wind in the nonradial open field region is numerically investigated, taking into account the thermal and dynamical coupling between the chromosphere and the corona. A realistic steady solar wind is constructed which passes through the inner X-type critical point in the rapidly diverging region. The wind speed shows a local maximum at the middle, O-type, critical point. The wind's density and pressure distributions decrease abruptly in the rapidly diverging region of the flow tube. The transient behavior of the wind following flare energy deposition includes ascending and descending conduction fronts. Thermal instability occurs in the lower corona, and ascending material flows out through the throat after the flare energy input ceases. A local density distribution peak is generated at the shock front due to the pressure deficit just behind the shock front.

A method of lines solution of the transient behavior of the helium cooled power leads for the SSC

International Nuclear Information System (INIS)

Demko, J.A.; Schiesser, W.E.; Carcagno, R.; McAshan, M.

1995-01-01

In this study, a detailed numerical thermal mode of a 6.5 kA power lead for the Superconducting Super Collider has been developed, which was adapted from the dynamic model developed by Schiesser. The transient behavior of the power leads was modeled using, a method of lines (MOL) approach. The model was developed to pmvide a tool for analyzing coolant control strategies as well as an understanding of the behavior of the leads under presumed system transients. Results for a current ramp up to 4970 amps are favorably compared with measurements. Also, a loss of cooling situation is predicted to determine the transient temperature distribution under an off-design condition

TRANSPA: a code for transient thermal analysis of a single fuel pin

International Nuclear Information System (INIS)

Prenger, F.C.

1985-02-01

An analytical model (TRANSPA) for the transient thermal analysis of a single uranium carbide fuel pin was developed. This model uses thermal boundary conditions obtained from COBRA-WC output and calculates the transient thermal response of a single fuel pin to changes in internal power generation, coolant flowrate, or fuel pin physical configuration. The model uses the MITAS finite difference thermal analyzer. MITAS provides the means to input separate conductance models through the use of a user subroutine input capability. The model is a lumped-mass representation of the fuel pin using 26 nodes and 42 conductors. Run time for each transient analysis is approximately one minute of central processor time on the NOS operating system

Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

International Nuclear Information System (INIS)

Williamson, R.L.

2011-01-01

Highlights: → The ABAQUS thermomechanics code is enhanced to enable simulation of nuclear fuel behavior. → Comparisons are made between discrete and smeared fuel pellet analysis. → Multidimensional and multipellet analysis is important for accurate prediction of PCMI. → Fully coupled thermomechanics results in very smooth prediction of fuel-clad gap closure. → A smeared-pellet approximation results in significant underprediction of clad radial displacements and plastic strain. - Abstract: A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO 2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

Investigation of transient thermal dissipation in thinned LSI for advanced packaging

Science.gov (United States)

Araga, Yuuki; Shimamoto, Haruo; Melamed, Samson; Kikuchi, Katsuya; Aoyagi, Masahiro

2018-04-01

Thinning of LSI is necessary for superior form factor and performance in dense cutting-edge packaging technologies. At the same time, degradation of thermal characteristics caused by the steep thermal gradient on LSIs with thinned base silicon is a concern. To manage a thermal environment in advanced packages, thermal characteristics of the thinned LSIs must be clarified. In this study, static and dynamic thermal dissipations were analyzed before and after thinning silicon to determine variations of thermal characteristics in thinned LSI. Measurement results revealed that silicon thinning affects dynamic thermal characteristics as well as static one. The transient variations of thermal characteristics of thinned LSI are precisely verified by analysis using an equivalent model based on the thermal network method. The results of analysis suggest that transient thermal characteristics can be easily estimated by employing the equivalent model.

Shock-induced thermal wave propagation and response analysis of a viscoelastic thin plate under transient heating loads

Science.gov (United States)

Li, Chenlin; Guo, Huili; Tian, Xiaogeng

2018-04-01

This paper is devoted to the thermal shock analysis for viscoelastic materials under transient heating loads. The governing coupled equations with time-delay parameter and nonlocal scale parameter are derived based on the generalized thermo-viscoelasticity theory. The problem of a thin plate composed of viscoelastic material, subjected to a sudden temperature rise at the boundary plane, is solved by employing Laplace transformation techniques. The transient responses, i.e. temperature, displacement, stresses, heat flux as well as strain, are obtained and discussed. The effects of time-delay and nonlocal scale parameter on the transient responses are analyzed and discussed. It can be observed that: the propagation of thermal wave is dynamically smoothed and changed with the variation of time-delay; while the displacement, strain, and stress can be rapidly reduced by nonlocal scale parameter, which can be viewed as an important indicator for predicting the stiffness softening behavior for viscoelastic materials.

BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

International Nuclear Information System (INIS)

Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

2012-01-01

The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

Transient thermal model of passenger car's cabin and implementation to saturation cycle with alternative working fluids

International Nuclear Information System (INIS)

Lee, Hoseong; Hwang, Yunho; Song, Ilguk; Jang, Kilsang

2015-01-01

A transient thermal model of a passenger car's cabin is developed to investigate the dynamic behavior of cabin thermal conditions. The model is developed based on a lumped-parameter model and solved using integral methods. Solar radiation, engine heat through the firewall, and engine heat to the air ducts are all considered. Using the thermal model, transient temperature profiles of the interior mass and cabin air are obtained. This model is used to investigate the transient behavior of the cabin under various operating conditions: the recirculation mode in the idling state, the fresh air mode in the idling state, the recirculation mode in the driving state, and fresh air mode in the driving state. The developed model is validated by comparing with experimental data and is within 5% of deviation. The validated model is then applied for evaluating the mobile air conditioning system's design. The study found that a saturation cycle concept (four-stage cycle with two-phase refrigerant injection) could improve the system efficiency by 23.9% and reduce the power consumption by 19.3%. Lastly, several alternative refrigerants are applied and their performance is discussed. When the saturation cycle concept is applied, R1234yf MAC (mobile air conditioning) shows the largest COP (coefficient of performance) improvement and power consumption reduction. - Highlights: • The transient thermal model of the passenger car cabin is developed. • The developed model is validated with experimental data and showed 5% deviation. • Saturation cycle concept is applied to the developed cabin model. • There is 24% COP improvement by applying the saturation cycle concept. • R1234yf showed the highest potential when it is applied to the saturation cycle.

Transient thermal driven bubble's surface and its potential ultrasound-induced damage

Science.gov (United States)

Movahed, Pooya; Freund, Jonathan B.

2017-11-01

Ultrasound-induced bubble activity in soft tissues is well-known to be a potential injury mechanism in therapeutic ultrasound treatments. We consider damage by transient thermal effects, including a hypothetical mechanism based on transient thermal phenomena, including viscous dissipation. A spherically symmetric compressible Navier-Stokes discretization is developed to solve the full governing equations, both inside and outside of the bubble, without the usual simplifications in the Rayleigh-Plesset bubble dynamics approach. Equations are solved in the Lagrangian framework, which provides a sharp and accurate representation of the interface as well as the viscous dissipation and thermal transport effects, which preclude reduction to the usual Rayleigh-Plesset ordinary differential equation. This method is used to study transient thermal effects at different frequencies and pressure amplitudes relevant to therapeutic ultrasound treatments. High temperatures achieved in the surrounding medium during the violent bubble collapse phase due to the viscous dissipation in the surrounding medium and thermal conduction from the bubble are expected to cause damage. This work was supported by NIH NIDDK Grant P01-DK043881.

TPX vacuum vessel transient thermal and stress conditions

International Nuclear Information System (INIS)

Feldshteyn, Y.; Dinkevich, S.; Feng, T.; Majumder, D.

1995-01-01

The TPX vacuum vessel provides the vacuum boundary for the plasma and the mechanical support for the internal components. Another function of the vacuum vessel is to contain neutron shielding water in the double wall space during normal operation. This double wall space serves as a heat reservoir for the entire vacuum vessel during bakeout. The vacuum vessel and the internal components are subjected to thermal stresses induced by a nonuniform temperature distribution within the structure during bakeout. A successful Conceptual Design Review in March 1993 has established superheated steam as the heating source of the vacuum vessel. A transient bakeout mode of the vacuum vessel and in-vessel components has been analyzed to evaluate transient period duration, proper temperature level, actual thermal stresses and performance of the steam equipment. Thermally, the vacuum vessel structure may be considered as an adiabatic system because it is perfectly insulated by the strong surrounding vacuum and multiple layers of superinsulation. Important aspects of the analysis are described herein

Investigation of effective factors of transient thermal stress of the MONJU-System components

Energy Technology Data Exchange (ETDEWEB)

Inoue, Masaaki; Hirayama, Hiroshi; Kimura, Kimitaka; Jinbo, M. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

1999-03-01

Transient thermal stress of each system Component in the fast breeder reactor is an uncertain factor on it's structural design. The temperature distribution in a system component changes over a wide range in time and in space. An unified evaluation technique of thermal, hydraulic, and structural analysis, in which includes thermal striping, temperature stratification, transient thermal stress and the integrity of the system components, is required for the optimum design of tho fast reactor plant. Thermal boundary conditions should be set up by both the transient thermal stress analysis and the structural integrity evaluation of each system component. The reasonable thermal boundary conditions for the design of the MONJU and a demonstration fast reactor, are investigated. The temperature distribution analysis models and the thermal boundary conditions on the Y-piece structural parts of each system component, such as reactor vessel, intermediate heat exchanger, primary main circulation pump, steam generator, superheater and upper structure of reactor core, are illustrated in the report. (M. Suetake)

Transient Thermal State of an Active Braille Matrix with Incorporated Thermal Actuators by Means of Finite Element Method

Science.gov (United States)

Alutei, Alexandra-Maria; Szelitzky, Emoke; Mandru, Dan

2013-01-01

In this article the authors present the transient thermal analysis for a developed thermal linear actuator based on wax paraffin used to drive the cells of a Braille device. A numerical investigation of transient heat transfer phenomenon during paraffin melting and solidification in an encapsulated recipient has been carried out using the ANSYS…

Estimation of inelastic behavior for a tapered nozzle in vessel due to thermal transient load using stress redistribution locus method

International Nuclear Information System (INIS)

Kobayashi, Ken-ichi; Yamada, Jun-ichi

2010-01-01

Simplified inelastic design procedures for elevated temperature components have been required to reduce simulation cost and to shorten a period of time for developing new projects. Stress redistribution locus (SRL) method has been proposed to provide a reasonable estimate employing both the elastic FEM analysis and a unique hyperbolic curve: ε tilde={1/σ tilde + (κ - 1)σ tilde}/κ, where ε tilde and σ tilde show dimensionless strain and stress normalized by corresponding elastic ones, respectively. This method is based on a fact that stress distribution in well deformed or high temperature components would change with deformation or time, and that the relation between the dimensionless stress and strain traces a kind of the elastic follow-up locus in spite of the constitutive equation of material and loading modes. In this paper, FEM analyses incorporating plasticity and creep in were performed for a tapered nozzle in reactor vessel under some thermal transient loads through the nozzle thickness. The normalized stress and strain was compared with the proposed SRL curve. Calculation results revealed that a critical point in the tapered nozzle due to the thermal transient load depended on a descending rate of temperature from the higher temperature in the operation cycle. Since the inelastic behavior in the nozzle resulted in a restricted area, the relationship between the normalized stress and strain was depicted inside the proposed SRL curve: Coefficient κ of the SRL in analyses is greater than the proposed one, and the present criterion guarantees robust structures for complicated components involving inelastic deformation. (author)

GAPCON-THERMAL-3

International Nuclear Information System (INIS)

Mohr, C.L.; Lanning, D.D.; Panisko, F.E.

1979-01-01

The fuel performance code GAPCON-THERMAL-3 has been expanded to include recent transient material deformation constitutive relations and the FLECHT heat transfer correlation. The modifications make it possible to compute the thermal and mechanical response of nuclear fuel to postulated Loss of Coolant Accidents (LOCA). The numerical formulation has the capability of predicting both steady state and transient behavior of a fuel rod using a single analytical procedure. GAPCON-THERMAL-3 (G-T-3) uses a specialized finite element procedure for mechanics predictions and the method of weighted residuals and finite difference techniques to compute temperature and thermal behavior. Fuel behavior, gas release models, gas conductance models, and stored energy calculations are applicable to both steady state and transient conditions. The code has been used to perform scoping analysis for in-reactor LOCA simulation testing. (orig.)

Thermal analysis of LOFT modular DTT for LOCE transient

International Nuclear Information System (INIS)

Martin, C.M.

1978-01-01

A thermal analysis was performed on the LOFT modular drag-disc turbine transducer (MDTT) modular assembly. The purpose of this analysis was to determine the maximum temperature difference between the MDTT shroud and end cap during a LOCE. This temperature difference is needed for stress analysis of the MDTT endcap to fairing welds. The thermal analysis was done using TRIPLE, a three dimensional finite element code. A three dimensional model of the MDTT was made and transient temperature solutions were found for the different MDTT locations. The fluid temperature transients used for the solutions at all locations were from RELAP4 predictions of the LOFT L2-4 test which is considered the most severe temperature transient. Results of these calculations show the maximum temperature difference is 92 0 C (165 0 F) and occurs in the intact loop cold leg. This value and those found at other locations, are evaluated from the best available RELAP predicted temperatures during a nuclear LOCE

Nonlinear Transient Thermal Analysis by the Force-Derivative Method

Science.gov (United States)

Balakrishnan, Narayani V.; Hou, Gene

1997-01-01

High-speed vehicles such as the Space Shuttle Orbiter must withstand severe aerodynamic heating during reentry through the atmosphere. The Shuttle skin and substructure are constructed primarily of aluminum, which must be protected during reentry with a thermal protection system (TPS) from being overheated beyond the allowable temperature limit, so that the structural integrity is maintained for subsequent flights. High-temperature reusable surface insulation (HRSI), a popular choice of passive insulation system, typically absorbs the incoming radiative or convective heat at its surface and then re-radiates most of it to the atmosphere while conducting the smallest amount possible to the structure by virtue of its low diffusivity. In order to ensure a successful thermal performance of the Shuttle under a prescribed reentry flight profile, a preflight reentry heating thermal analysis of the Shuttle must be done. The surface temperature profile, the transient response of the HRSI interior, and the structural temperatures are all required to evaluate the functioning of the HRSI. Transient temperature distributions which identify the regions of high temperature gradients, are also required to compute the thermal loads for a structural thermal stress analysis. Furthermore, a nonlinear analysis is necessary to account for the temperature-dependent thermal properties of the HRSI as well as to model radiation losses.

FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Sakagami, M.

1984-01-01

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

Fast thermal transients on valve

International Nuclear Information System (INIS)

Ferjancic, M.; Stok, B.; Halilovic, M.; Koc, P.; Mole, N.; Otrin, Z.; Kotar, A.

2007-01-01

One of the regulatory body methods to supervise nuclear safety of a nuclear power plant is a review of plant modifications and evaluation of their impact on plant operating experience. The Slovenian Nuclear Safety Administration (SNSA) licensed in April 2003 the use of leak-before-break (LBB) methodology in the Krsko NPP for the primary loop including surge line and connecting pipelines with minimal diameter of 6 inch. The SNSA decision based also on fracture mechanics analyses that include direct pipe failure mechanisms such as water hammer, creep damage, erosion and corrosion, fatigue and environmental conditions over the entire life of the plant. The evaluation of the operating transients pointed out, that presumed loadings, used for the LBB analysis, did not incorporate all the fast thermal transients data. For that purpose the SNSA requested Faculty of Mechanical Engineering (FS) in Ljubljana to perform additional analyses. The results of the analysis shall confirm the validity of the LBB analysis. (author)

The influence of core material on transient thermal impedances in transformers

International Nuclear Information System (INIS)

Górecki, K; Górski, K

2016-01-01

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

TRANTHAC-1: transient thermal-hydraulic analysis code for HTGR core of multi-channel model

International Nuclear Information System (INIS)

Sato, Sadao; Miyamoto, Yoshiaki

1980-08-01

The computer program TRANTHAC-1 is for predicting thermal-hydraulic transient behavior in HTGR's core of pin-in-block type fuel elements, taking into consideration of the core flow distribution. The program treats a multi-channel model, each single channel representing the respective column composed of fuel elements. The fuel columns are grouped in flow control regions; each region is provided with an orifice assembly. In the region, all channels are of the same shape except one channel. Core heat is removed by downward flow of the control through the channel. In any transients, for given time-dependent power, total core flow, inlet coolant temperature and coolant pressure, the thermal response of the core can be determined. In the respective channels, the heat conduction in radial and axial direction are represented. And the temperature distribution in each channel with the components is calculated. The model and usage of the program are described. The program is written in FORTRAN-IV for computer FACOM 230-75 and it is composed of about 4,000 cards. The required core memory is about 75 kilowords. (author)

Fatigue evaluation of piping connections under thermal transients

International Nuclear Information System (INIS)

Aquino, C.T.E. de; Maneschy, J.E.

1993-01-01

In designing nuclear power plant piping, thermal transients, caused by non-steady operation conditions, should be considered. These events may reduce considerably the lifetime of the pipes, creating the necessity of using structural elements designed in such a way to minimize the acting thermal stresses. Typical examples of the usage of these elements are the connections between pipes of small and large diameters, in which it is usually used a weldolet. Nevertheless, in some situations, the thermal stresses caused by the transients are greater than the allowable limits, being, in this case, an alternative for best results, the introduction of a special fitting replacing the weldolet. Such a fitting is designed in a way to permit a better distribution of the stresses, reducing its maximum value to acceptable levels. This paper intends to present a fatigue evaluation of a connection, using the above mentioned fitting, when subjected to a load expressed in terms of a step thermal gradient, varying from 263 deg to 40 deg C. Two different methodologies are used in this analysis: (a) Determination of the temperature distribution from the heat transfer equations for piping, being the stresses calculated according to ASME III NB-3600. (b) Thermal and stress analyses using axisymmetric elements, according to the rules presented at ASME III NB-3200. In the first case, named simplified analysis, the computer code used is the PIPESTRESS, while in the second case, the ANSYS program was adopted

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.

Computer-aided methods of determining thyristor thermal transients

International Nuclear Information System (INIS)

Lu, E.; Bronner, G.

1988-08-01

An accurate tracing of the thyristor thermal response is investigated. This paper offers several alternatives for thermal modeling and analysis by using an electrical circuit analog: topological method, convolution integral method, etc. These methods are adaptable to numerical solutions and well suited to the use of the digital computer. The thermal analysis of thyristors was performed for the 1000 MVA converter system at the Princeton Plasma Physics Laboratory. Transient thermal impedance curves for individual thyristors in a given cooling arrangement were known from measurements and from manufacturer's data. The analysis pertains to almost any loading case, and the results are obtained in a numerical or a graphical format. 6 refs., 9 figs

Thermal transient analysis applied to horizontal wells

Energy Technology Data Exchange (ETDEWEB)

Duong, A.N. [Society of Petroleum Engineers, Canadian Section, Calgary, AB (Canada)]|[ConocoPhillips Canada Resources Corp., Calgary, AB (Canada)

2008-10-15

Steam assisted gravity drainage (SAGD) is a thermal recovery process used to recover bitumen and heavy oil. This paper presented a newly developed model to estimate cooling time and formation thermal diffusivity by using a thermal transient analysis along the horizontal wellbore under a steam heating process. This radial conduction heating model provides information on the heat influx distribution along a horizontal wellbore or elongated steam chamber, and is therefore important for determining the effectiveness of the heating process in the start-up phase in SAGD. Net heat flux estimation in the target formation during start-up can be difficult to measure because of uncertainties regarding heat loss in the vertical section; steam quality along the horizontal segment; distribution of steam along the wellbore; operational conditions; and additional effects of convection heating. The newly presented model can be considered analogous to pressure transient analysis of a buildup after a constant pressure drawdown. The model is based on an assumption of an infinite-acting system. This paper also proposed a new concept of a heating ring to measure the heat storage in the heated bitumen at the time of testing. Field observations were used to demonstrate how the model can be used to save heat energy, conserve steam and enhance bitumen recovery. 18 refs., 14 figs., 2 appendices.

Thermal-hydraulic analysis of PWR cores in transient condition

International Nuclear Information System (INIS)

Silva Galetti, M.R. da.

1984-01-01

A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt

Simulation of Thermal Transients using CSMP

International Nuclear Information System (INIS)

Konuk, A.A.

1981-01-01

A mathematical model has been developed to simulate thermal transientes for the Hellum Loop of the 'Instituto de Pesquisas Energeticas e Nuleares', Sao Paulo. The model is based on the energy equation applied to the various components of the loop. The non-linear system of first order ordinary differential equation and algebraic equations has been solved using IBM'S 'System/360-Continuous System Modeling Program-CSMP'. The model has been tested satisfactory with experimental results. (Author) [pt

Analysis of transient thermal response in the outlet plenum of an LMFBR

International Nuclear Information System (INIS)

Yang, J.W.

1976-05-01

A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the upper outlet plenum of an LMFBR. The one-dimensional turbulent jet flow equations were solved to determine the maximum penetration of the core flow. The maximum penetration is used as the criterion for dividing the sodium region into two mixing zones. The lumped-parameter model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of bypass flow into the plenum. Numerical calculations were performed for two cases. The first case corresponds to a normal scram followed by flow coast-down. The second case represents the double-ended pipe rupture at the inlet of cold leg followed by reactor scram. The results indicate that effects of flow stratification, chimney height, metal heat capacity and bypass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas does not play any significant role on sodium temperature. This two-zone mixing model will be a part of the thermohydraulic transient code SSC

Modeling and analysis of thermal-hydraulic response of uranium-aluminum reactor fuel plates under transient heatup conditions

Energy Technology Data Exchange (ETDEWEB)

Navarro-Valenti, S.; Kim, S.H.; Georgevich, V. [Oak Ridge National Lab., TN (United States)] [and others

1995-09-01

The purpose of this paper is to describe the analysis performed to predict the thermal behavior of fuel miniplates under rapid transient heatup conditions. The possibility of explosive boiling was considered, and it was concluded that the heating rates are not large enough for explosive boiling to occur. However, transient boiling effects were pronounced. Because of the complexity of transient pool boiling and the unavailability of experimental data for the situations studied, an approximation was made that predicted the data very well within the uncertainties present. If pool boiling from the miniplates had been assumed to be steady during the heating pulse, the experimental data would have been greatly overestimated. This fact demonstrates the importance of considering the transient nature of heat transfer in the analysis of reactivity excursion accidents. An additional contribution of the present work is that it provided data on highly subcooled steady nulceate boiling from the cooling portion of the thermocouple traces.

Numerical Calculation of Transient Thermal Characteristics in Gas-Insulated Transmission Lines

Directory of Open Access Journals (Sweden)

Hongtao Li

2013-11-01

Full Text Available For further knowledge of the thermal characteristics in gas-insulated transmission lines (GILs installed above ground, a finite-element model coupling fluid field and thermal field is established, in which the corresponding assumptions and boundary conditions are given.  Transient temperature rise processes of the GIL under the conditions of variable ambient temperature, wind velocity and solar radiation are respectively investigated. Equivalent surface convective heat transfer coefficient and heat flux boundary conditions are updated in the analysis process. Unlike the traditional finite element methods (FEM, the variability of the thermal properties with temperature is considered. The calculation results are validated by the tests results reported in the literature. The conclusion provides method and theory basis for the knowledge of transient temperature rise characteristics of GILs in open environment.

Transient, heat-induced thermal resistance in the small intestine of mouse

International Nuclear Information System (INIS)

Hume, S.P.; Marigold, J.C.L.

1980-01-01

Heat-induced thermal resistance has been investigated in mouse jejunum by assaying crypt survival 24 h after treatment. Hyperthermia was achieved by immersing an exteriorized loop of intestine in a bath of Krebs-Ringer solution. Two approaches have been used. In the first, thermal survival curves were obtained following single hyperthermal treatments at temperatures in the range 42 to 44 0 C. Transient thermal resistance, inducted by a plateau in the crypt survival curve, developed during heating at temperatures around 42.5 0 C after 60 to 80 min. In the second series of experiments, a priming heat treatment (40.0, 41.0, 41.5, or 42.0 0 C for 60 min) was followed at varying intervals by a test treatment at 43.0 0 C. A transient resistance to the second treatment was induced, the extent and time of development being dependent upon the priming treatment. Crypt survival curves for thermally resistant intestine showed an increase in thermal D 0 and a decrease in n compared with curves from previously unheated intestine

Evaluation of LWR fuel rod behavior under operational transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Hiramoto, K.; Maru, A.

1984-01-01

To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

Transient cases analyses of the TRIGA IPR-R1 using thermal hydraulic and neutron kinetic coupled codes

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Scari, Maria E., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

2015-07-01

Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)

Theoretical basis for a transient thermal elastic-plastic stress analysis of nuclear reactor fuel elements

International Nuclear Information System (INIS)

Hsu, T.R.; Bertels, A.W.M.; Banerjee, S.; Harrison, W.C.

1976-07-01

This report presents the theoretical basis for a transient thermal elastic-plastic stress analysis of a nuclear reactor fuel element subject to severe transient thermo-mechanical loading. A finite element formulation is used for both the non-linear stress analysis and thermal analysis. These two major components are linked together to form an integrated program capable of predicting fuel element transient behaviour in two dimensions. Specific case studies are presented to illustrate capabilities of the analysis. (author)

Analysis of transient and hysteresis behavior of cross-flow heat exchangers under variable fluid mass flow rate for data center cooling applications

International Nuclear Information System (INIS)

Gao, Tianyi; Murray, Bruce; Sammakia, Bahgat

2015-01-01

Effective thermal management of data centers is an important aspect of reducing the energy required for the reliable operation of data processing and communications equipment. Liquid and hybrid (air/liquid) cooling approaches are becoming more widely used in today's large and complex data center facilities. Examples of these approaches include rear door heat exchangers, in-row and overhead coolers and direct liquid cooled servers. Heat exchangers are primary components of liquid and hybrid cooling systems, and the effectiveness of a heat exchanger strongly influences the thermal performance of a cooling system. Characterizing and modeling the dynamic behavior of heat exchangers is important for the design of cooling systems, especially for control strategies to improve energy efficiency. In this study, a dynamic thermal model is solved numerically in order to predict the transient response of an unmixed–unmixed crossflow heat exchanger, of the type that is widely used in data center cooling equipment. The transient response to step and ramp changes in the mass flow rate of both the hot and cold fluid is investigated. Five model parameters are varied over specific ranges to characterize the transient performance. The parameter range investigated is based on available heat exchanger data. The thermal response to the magnitude, time period and initial and final conditions of the transient input functions is studied in detail. Also, the hysteresis associated with the fluid mass flow rate variation is investigated. The modeling results and performance data are used to analyze specific dynamic performance of heat exchangers used in practical data center cooling applications. - Highlights: • The transient performance of a crossflow heat exchanger was modeled and studied. • This study provides design information for data center thermal management. • The time constant metric was used to study the impacts of many variable inputs. • The hysteresis behavior

Apparatus and method for transient thermal infrared spectrometry

Science.gov (United States)

McClelland, John F.; Jones, Roger W.

1991-12-03

A method and apparatus for enabling analysis of a material (16, 42) by applying a cooling medium (20, 54) to cool a thin surface layer portion of the material and to transiently generate a temperature differential between the thin surface layer portion and the lower portion of the material sufficient to alter the thermal infrared emission spectrum of the material from the black-body thermal infrared emission spectrum of the material. The altered thermal infrared emission spectrum of the material is detected by a spectrometer/detector (28, 50) while the altered thermal infrared emission spectrum is sufficiently free of self-absorption by the material of the emitted infrared radiation. The detection is effected prior to the temperature differential propagating into the lower portion of the material to an extent such that the altered thermal infrared emission spectrum is no longer sufficiently free of self-absorption by the material of emitted infrared radiation, so that the detected altered thermal infrared emission spectrum is indicative of the characteristics relating to the molecular composition of the material.

The OECD/NEA/NSC PBMR 400 MW coupled neutronics thermal hydraulics transient benchmark: transient results - 290

International Nuclear Information System (INIS)

Strydom, G.; Reitsma, F.; Ngeleka, P.T.; Ivanov, K.N.

2010-01-01

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated, and code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MWth design and a representative set of transient exercises are defined as an OECD benchmark. The scope of the benchmark is to establish a series of well defined multi-dimensional computational benchmark problems with a common given set of cross sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. This paper describes the current status of the benchmark project and shows the results for the six transient exercises, consisting of three Loss of Cooling Accidents, two Control Rod Withdrawal transients, a power load-follow transient, and a Helium over-cooling Accident. The participants' results are compared using a statistical method and possible areas of future code improvement are identified. (authors)

Study of the behavior of thermal shield support system for the French CPO series plants

International Nuclear Information System (INIS)

Bellet, S.; Roux, P.; Bhandari, D.R.; Schwirian, R.E.; Yu, C.; Matarazzo, J.C.; Singleton, N.R.

1996-01-01

Degradation/failure of thermal shield support system in PWRs has been observed in the US as well as in foreign plants. In almost all the cases, remedial actions were put in place at very high economic costs to the utilities only after the failures had occurred. This paper presents the results of a comprehensive study to predict the long term behavior of a thermal shield support system due to flow-induced vibratory loads and thermal transients. Excellent agreement from the system finite model between the measured plant test data on the barrel/thermal shield beam and shell mode frequencies and the flexure strains confirms the basic structural behavior and physics of the flow induced vibrations. Loads and stresses on the support bolts and the flexures were determined to predict the fatigue life of the components

Experimental Setup with Transient Behavior of Fuel Cladding of SFR

Energy Technology Data Exchange (ETDEWEB)

Shin, Sang Hun; Kim, Jun Hwan; Kim, June-Hyung; Ryu, Woo Seog; Park, Sang Gyu; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Nowadays, in Korea, advanced cladding such as FC92 is developed and its transient behaviors are required for the safety analysis of SFR. Design and safety analyses of sodium-cooled fast reactor (SFR) require understanding fuel pin responses to a wide range of off-normal events. In a loss-of-flow (LOF) or transient over-power (TOP), the temperature of the cladding is rapidly increased above its steady-state service temperature. Transient tests have been performed in sections of fuel pin cladding and a large data base has been established for austenitic stainless steel such as 20% cold-worked 316 SS and ferritic/martensitic steels such as HT9. This paper summarizes the technical status of transient testing facilities and their results. Previous researches showed the transient behaviors of HT9 cladding. For the safety analyses in SFR in Korea, simulated transient tests with newly developed FC92 as well as HT9 cladding are being carried out.

Simple method of calculating the transient thermal performance of composite material and its applicable condition

Institute of Scientific and Technical Information of China (English)

张寅平; 梁新刚; 江忆; 狄洪发; 宁志军

2000-01-01

Degree of mixing of composite material is defined and the condition of using the effective thermal diffusivity for calculating the transient thermal performance of composite material is studied. The analytical result shows that for a prescribed precision of temperature, there is a condition under which the transient temperature distribution in composite material can be calculated by using the effective thermal diffusivity. As illustration, for the composite material whose temperatures of both ends are constant, the condition is presented and the factors affecting the relative error of calculated temperature of composite materials by using effective thermal diffusivity are discussed.

Transient thermal creep of nuclear reactor pressure vessel type concretes

International Nuclear Information System (INIS)

Khoury, G.A.

1983-01-01

The immediate aim of the research was to study the transient thermal strain behaviour of four AGR type nuclear reactor concretes during first time heating in an unsealed condition to 600 deg. C. The work being also relevant to applications of fire exposed concrete structures. The programme was, however, expanded to serve a second more theoretical purpose, namely the further investigation of the strain development of unsealed concrete under constant, transient and cyclic thermal states in particular and the effect of elevated temperatures on concrete in general. The range of materials investigated included seven different concretes and three types of cement paste. Limestone, basalt, gravel and lightweight aggregates were employed as well as OPC and SRC cements. Cement replacements included pfa and slag. Test variables comprised two rates of heating (0.2 and 1 deg. C/minute), three initial moisture contents (moist as cast, air-dry and oven dry at 105 deg. C), two curing regimes (bulk of tests represented mass cured concrete), five stress levels (0, 10, 20, 30 and a few tests at 60% of the cold strength), two thermal cycles and levels of test temperature up to 720 deg. C. Supplementary, dilatometry, TGA and DTA tests were performed at CERL on individual samples of aggregate and cement paste which helped towards explaining the observed trends in the concretes. A simple formula was developed which relates the elastic thermal stresses generated from radial temperature gradients to the solution obtained from the transient heat conduction equation. Thermal stresses can, therefore, be minimized by reductions in the radius of the specimen and the rate of heating The results were confirmed by finite element analysis which indicate( tensile stresses in the central region and compressive stresses near the surf ace during heating which are reversed during cooling. It is shown that the temperature gradients, pore pressures and tensile thermal stresses during both heating and

Predication of skin temperature and thermal comfort under two-way transient environments.

Science.gov (United States)

Zhou, Xin; Xiong, Jing; Lian, Zhiwei

2017-12-01

In this study, three transient environmental conditions consisting of one high-temperature phase within two low-temperature phases were developed, thus creating a temperature rise followed by a temperature fall. Twenty-four subjects (including 12 males and 12 females) were recruited and they underwent all three test scenarios. Skin temperature on seven body parts were measured during the whole period of the experiment. Besides, thermal sensation was investigated at specific moments by questionnaires. Thermal sensation models including PMV model, Fiala model and the Chinese model were applied to predict subjects' thermal sensation with comparisons carried out among them. Results show that most predicated thermal sensation by Chinese model lies within the range of 0.5 scale of the observed sensation vote, and it agrees best with the observed thermal sensation in transient thermal environment than PMV and Fiala model. Further studies should be carried out to improve performance of Chinese model for temperature alterations between "very hot" to "hot" environment, for prediction error in the temperature-fall situation of C5 (37-32°C) was over 0.5 scale. Copyright © 2017 Elsevier Ltd. All rights reserved.

The impact of transient thermal loads on beryllium as plasma facing material

Energy Technology Data Exchange (ETDEWEB)

Spilker, Benjamin Christof

2017-01-24

The rising global energy consumption requires a broad research and development approach in the field of energy technology. Besides renewables, nuclear fusion promises an efficient, CO{sub 2} free, no long-term radioactive waste producing, and safe energy source using only deuterium and lithium as primary resources, which are widely abundant. However, several technical challenges have to be overcome before a nuclear fusion power plant can be built. For this purpose, the experimental reactor ITER is currently under construction in France. ITER is intended to demonstrate the scientific and technological feasibility of net energy generation via nuclear fusion. The most heavily loaded components inside a fusion reactor, which are directly facing the fusion plasma, have to be armoured with well suited materials, which need to be able to withstand the high thermal and particle loads for an economically reasonable lifetime. For ITER, beryllium is chosen as plasma facing material for the largest fraction of the inner vacuum vessel, the so called first wall. Tungsten will be applied in the bottom region of the vacuum vessel, the so called divertor, which acts as the exhaust system of the machine. The choice of beryllium as plasma facing material was driven by its outstanding advantages, e.g. the low atomic number assures that eroded wall material does not strongly decrease the fusion plasma performance, while it combines a high thermal conductivity with low chemical sputtering characteristics. However, the relatively low melting temperature of beryllium of 1287 C comprises the risk of amour damage by melting during transient plasma events, such as edge localized modes or plasma disruptions. Even when mitigated, these events put tremendous power densities in the GW m{sup -2} range with durations in the ms scale onto the plasma facing materials. Hence, the performance of the ITER reference beryllium grade S-65 under transient thermal loads was studied within this work. Thereby

The impact of transient thermal loads on beryllium as plasma facing material

International Nuclear Information System (INIS)

Spilker, Benjamin Christof

2017-01-01

The rising global energy consumption requires a broad research and development approach in the field of energy technology. Besides renewables, nuclear fusion promises an efficient, CO_2 free, no long-term radioactive waste producing, and safe energy source using only deuterium and lithium as primary resources, which are widely abundant. However, several technical challenges have to be overcome before a nuclear fusion power plant can be built. For this purpose, the experimental reactor ITER is currently under construction in France. ITER is intended to demonstrate the scientific and technological feasibility of net energy generation via nuclear fusion. The most heavily loaded components inside a fusion reactor, which are directly facing the fusion plasma, have to be armoured with well suited materials, which need to be able to withstand the high thermal and particle loads for an economically reasonable lifetime. For ITER, beryllium is chosen as plasma facing material for the largest fraction of the inner vacuum vessel, the so called first wall. Tungsten will be applied in the bottom region of the vacuum vessel, the so called divertor, which acts as the exhaust system of the machine. The choice of beryllium as plasma facing material was driven by its outstanding advantages, e.g. the low atomic number assures that eroded wall material does not strongly decrease the fusion plasma performance, while it combines a high thermal conductivity with low chemical sputtering characteristics. However, the relatively low melting temperature of beryllium of 1287 C comprises the risk of amour damage by melting during transient plasma events, such as edge localized modes or plasma disruptions. Even when mitigated, these events put tremendous power densities in the GW m"-"2 range with durations in the ms scale onto the plasma facing materials. Hence, the performance of the ITER reference beryllium grade S-65 under transient thermal loads was studied within this work. Thereby, the

Suppression of thermal transients in advanced LIGO interferometers using CO2 laser preheating

Science.gov (United States)

Jaberian Hamedan, V.; Zhao, C.; Ju, L.; Blair, C.; Blair, D. G.

2018-06-01

In high optical power interferometric gravitational wave detectors, such as Advanced LIGO, the thermal effects due to optical absorption in the mirror coatings and the slow thermal response of fused silica substrate cause time dependent changes in the mirror profile. After locking, high optical power builds up in the arm cavities. Absorption induced heating causes optical cavity transverse mode frequencies to drift over a period of hours, relative to the fundamental mode. At high optical power this can cause time dependent transient parametric instability, which can lead to interferometer disfunction. In this paper, we model the use of CO2 laser heating designed to enable the interferometer to be maintained in a thermal condition such that transient changes in the mirrors are greatly reduced. This can minimize transient parametric instability and compensate dark port power fluctuations. Modeling results are presented for both single compensation where a CO2 laser acting on one test mass per cavity, and double compensation using one CO2 laser for each test mass. Using parameters of the LIGO Hanford Observatory X-arm as an example, single compensation allows the maximum mode frequency shift to be limited to 6% of its uncompensated value. However, single compensation causes transient degradation of the contrast defect. Double compensation minimise contrast defect degradation and reduces transients to less than 1% if the CO2 laser spot is positioned within 2 mm of the cavity beam position.

Fracture mechanics analysis of reactor pressure vessel under pressurized thermal shock - The effect of elastic-plastic behavior and stainless steel cladding -

International Nuclear Information System (INIS)

Joo, Jae Hwang; Kang, Ki Ju; Jhung, Myung Jo

2002-01-01

Performed here is an assessment study for deterministic fracture mechanics analysis of a pressurized thermal shock (PTS). The PTS event means an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. The problems consisting of two transients and 10 cracks are solved and maximum stress intensity factors and maximum allowable nil-ductility reference temperatures are calculated. Their results are compared each other to address the general characteristics between transients, crack types and analysis methods. The effects of elastic-plastic material behavior and clad coating on the inner surface are explored

Thermal Management of Transient Power Spikes in Electronics - Phase Change Energy Storage or Copper Heat Sinks?

OpenAIRE

Krishnan, S.; Garimella, S V

2004-01-01

A transient thermal analysis is performed to investigate thermal control of power semiconductors using phase change materials, and to compare the performance of this approach to that of copper heat sinks. Both the melting of the phase change material under a transient power spike input, as well as the resolidification process, are considered. Phase change materials of different kinds (paraffin waxes and metallic alloys) are considered, with and without the use of thermal conductivity enhancer...

Simultaneous Determination of Thermal Conductivity and Thermal Diffusivity of Food and Agricultural Materials Using a Transient Plane-Source Method

Science.gov (United States)

Thermal conductivity and thermal diffusivity are two important physical properties essential for designing any food engineering processes. Recently a new transient plane-source method was developed to measure a variety of materials, but its application in foods has not been documented. Therefore, ...

Transient thermal stress problem for a circumferentially cracked hollow cylinder

Science.gov (United States)

Nied, H. F.; Erdogan, F.

1982-01-01

The transient thermal stress problem for a hollow elasticity cylinder containing an internal circumferential edge crack is considered. It is assumed that the problem is axisymmetric with regard to the crack geometry and the loading, and that the inertia effects are negligible. The problem is solved for a cylinder which is suddenly cooled from inside. First the transient temperature and stress distributions in an uncracked cylinder are calculated. By using the equal and opposite of this thermal stress as the crack surface traction in the isothermal cylinder the crack problem is then solved and the stress intensity factor is calculated. The numerical results are obtained as a function of the Fourier number tD/b(2) representing the time for various inner-to-outer radius ratios and relative crack depths, where D and b are respectively the coefficient of diffusivity and the outer radius of the cylinder.

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Transient thermal stresses in an orthotropic finite rectangular plate due to arbitrary surface heat-generations

International Nuclear Information System (INIS)

Sugano, Y.

1980-01-01

The transient thermal stresses in an orthotropic finite rectangular plate due to arbitrary surface heat-generations on two edges are studied by means of the Airy stress function. The purposes of this paper are to present a method of determing the transient thermal stresses in an orthographic rectangular plate with four edges of distinct thermal boundary condition of the third kind which exactly satisfy the traction-free conditions of shear stress over all boundaries including four corners of the plate, and to consider the effects of the anisotropies of material properties and the convective heat transfer on the upper and lower surfaces on the thermal stress distribution. (orig.)

Transient thermal analysis of semiconductor diode lasers under pulsed operation

Science.gov (United States)

Veerabathran, G. K.; Sprengel, S.; Karl, S.; Andrejew, A.; Schmeiduch, H.; Amann, M.-C.

2017-02-01

Self-heating in semiconductor lasers is often assumed negligible during pulsed operation, provided the pulses are `short'. However, there is no consensus on the upper limit of pulse width for a given device to avoid-self heating. In this paper, we present an experimental and theoretical analysis of the effect of pulse width on laser characteristics. First, a measurement method is introduced to study thermal transients of edge-emitting lasers during pulsed operation. This method can also be applied to lasers that do not operate in continuous-wave mode. Secondly, an analytical thermal model is presented which is used to fit the experimental data to extract important parameters for thermal analysis. Although commercial numerical tools are available for such transient analyses, this model is more suitable for parameter extraction due to its analytical nature. Thirdly, to validate this approach, it was used to study a GaSb-based inter-band laser and an InP-based quantum cascade laser (QCL). The maximum pulse-width for less than 5% error in the measured threshold currents was determined to be 200 and 25 ns for the GaSb-based laser and QCL, respectively.

Thermodynamic model of a thermal storage air conditioning system with dynamic behavior

Energy Technology Data Exchange (ETDEWEB)

Fleming, E; Wen, SY; Shi, L; da Silva, AK

2013-12-01

A thermodynamic model was developed to predict transient behavior of a thermal storage system, using phase change materials (PCMs), for a novel electric vehicle climate conditioning application. The main objectives of the paper are to consider the system's dynamic behavior, such as a dynamic air flow rate into the vehicle's cabin, and to characterize the transient heat transfer process between the thermal storage unit and the vehicle's cabin, while still maintaining accurate solution to the complex phase change heat transfer. The system studied consists of a heat transfer fluid circulating between either of the on-board hot and cold thermal storage units, which we refer to as thermal batteries, and a liquid-air heat exchanger that provides heat exchange with the incoming air to the vehicle cabin. Each thermal battery is a shell-and-tube configuration where a heat transfer fluid flows through parallel tubes, which are surrounded by PCM within a larger shell. The system model incorporates computationally inexpensive semianalytic solution to the conjugated laminar forced convection and phase change problem within the battery and accounts for airside heat exchange using the Number of Transfer Units (NTUs) method for the liquid-air heat exchanger. Using this approach, we are able to obtain an accurate solution to the complex heat transfer problem within the battery while also incorporating the impact of the airside heat transfer on the overall system performance. The implemented model was benchmarked against a numerical study for a melting process and against full system experimental data for solidification using paraffin wax as the PCM. Through modeling, we demonstrate the importance of capturing the airside heat exchange impact on system performance, and we investigate system response to dynamic operating conditions, e.g., air recirculation. (C) 2013 Elsevier Ltd. All rights reserved.

Review of behavior of mixed-oxide fuel elements in extended overpower transient tests in EBR-II

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.

1994-10-01

From a series of five tests conducted in EBR-II, a substantial data base has been established on the performance of mixed-oxide fuel elements in a liquid-metal-cooled reactor under slow-ramp transient overpower conditions. Each test contained 19 preirradiated fuel elements with varying design and prior operating histories. Elements with aggressive design features, such as high fuel smear density and/or thin cladding, were included to accentuate transient effects. The ramp rates were either 0.1 or 10% ΔP/P/s and the overpowers ranged between ∼60 and 100% of the elements' prior power ratings. Six elements breached during the tests, all with aggressive design parameters. The other elements, including all those with moderate design features for the reference or advanced long-life drivers for PNC's prototype fast reactor Monju, maintained their cladding integrity during the tests. Posttest examination results indicated that fuel/cladding mechanical interaction (FCMI) was the most significant mechanism causing the cladding strain and breach. In contrast, pressure loading from the fission gas in the element plenum was less important, even in high-burnup elements. During an overpower transient, FCMI arises from fuel/cladding differential thermal expansion, transient fuel swelling, and, significantly, the gas pressure in the sealed central cavity of elements with substantial centerline fuel melting. Fuel performance data from these tests, including cladding breaching margin and transient cladding strain, are correlatable with fuel-element design and operating parameters. These correlations are being incorporated into fuel-element behavior codes. At the two tested ramp rates, fuel element behavior appears to be insensitive to transient ramp rate and there appears to be no particular vulnerability to slow ramp transients as previously perceived

Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

International Nuclear Information System (INIS)

Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang

2005-01-01

Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

Transient Fuel Behavior and Failure Condition in the CABRI-2 Experiments

International Nuclear Information System (INIS)

Sato, Ikken; Lemoine, Francette; Struwe, Dankward

2004-01-01

In the CABRI-2 program, 12 tests were performed under various transient conditions covering a wide range of accident scenarios using two types of preirradiated fast breeder reactor (FBR) fuel pins with different smear densities and burnups. For each fuel, a nonfailure-transient test was performed, and it provided basic information such as fuel thermal condition, fuel swelling, and gas release. From the failure tests, information on failure mode, failure time, and axial location was obtained. Based on this information, failure conditions such as fuel enthalpy and cladding temperature were evaluated. These failure conditions were compared with the CABRI-1 tests in which different fuels as well as different transient conditions were used. This comparison, together with supporting information available from existing in-pile and out-of-pile experiments, allowed an effective understanding on failure mechanisms depending on fuel and transient conditions. It is concluded that pellet-cladding mechanical interaction (PCMI) due to fuel thermal expansion and fission-gas-induced swelling is playing an important role on mechanical clad loading especially with high smear density and low fuel-heating-rate conditions. At very high heating-rate conditions, there is no sufficient time to allow significant fuel swelling, so that cavity pressurization with fuel melting becomes the likely failure mechanism. Fuel smear density and fission-gas retention have a strong impact both on PCMI and cavity pressurization. Furthermore, pin failure is strongly dependent on cladding temperature, which plays an important role in the axial failure location. With the low smear-density fuel, considerable PCMI mitigation is possible leading to a high failure threshold as well as in-pin molten-fuel relocation along the central hole. However, even with the low smear density fuel, PCMI failure could take place with an elevated cladding-temperature condition. On the other hand, in case of a sufficiently long

An analysis of transient thermal properties for high power GaN-based laser diodes

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Min; Kim, Seungtaek; Kang, Sung Bok; Kim, Young Jin; Jeong, Hoon; Lee, Kyeongkyun; Kim, Jongseok [Korea Institute of Industrial Technology, 35-3 Hongcheon-Ri, Ipjang-Myeon, Cheonan, Chungnam 331-825 (Korea); Lee, Sangdon; Suh, Dongsik [QSI Co., Ltd., 315-9 Cheonheung-Ri, Sungger-Eup, Cheonan, Chungnam 330-836 (Korea); Yi, Jeong Hoon; Choi, Yoonho; Jung, Seok Gu; Noh, Minsoo [LG Electronics Advanced Research Institute, 16 Woomyeon-Dong, Seocho-Gu, Seoul 137-724 (Korea)

2010-07-15

Thermal properties of 405 nm GaN-based laser diodes were investigated by employing a transient heating response method based on the temperature dependence of diode forward voltage. Thermal resistances of materials consisting of packaged laser diodes were differentiated in transient thermal response curves at a current below threshold current. With a current above threshold current, no significant change in thermal resistances and difference between junction-up and junction-down laser diodes was observed at pulses shorter than 3 sec. From an analysis with long current injections, thermal resistance of a packaged laser diode with a junction-up bonding was {proportional_to}45 C/W which was higher than that of a junction-down bonded laser diode by {proportional_to}10 C/W. Further analyses based on parameters obtained from voltage recovery curves indicated that the time constant for cooling is directly related to the thermal resistance and thermal capacitance of a laser diode package. (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

International Nuclear Information System (INIS)

Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari

2001-03-01

JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

Assessments of the kinetic and dynamic transient behavior of sub-critical systems (ADS) in comparison to critical reactor systems

International Nuclear Information System (INIS)

Schikorr, W.M.

2001-01-01

The neutron kinetic and the reactor dynamic behavior of Accelerator Driven Systems (ADS) is significantly different from those of conventional power reactor systems currently in use for the production of power. It is the objective of this study to examine and to demonstrate the intrinsic differences of the kinetic and dynamic behavior of accelerator driven systems to typical plant transient initiators in comparison to the known, kinetic and dynamic behavior of critical thermal and fast reactor systems. It will be shown that in sub-critical assemblies, changes in reactivity or in the external neutron source strength lead to an asymptotic power level essentially described by the instantaneous power change (i.e. prompt jump). Shutdown of ADS operating at high levels of sub-criticality, (i.e. k eff ∼0.99), without the support of reactivity control systems (such as control or safety rods), may be problematic in case the ability of cooling of the core should be impaired (i.e. loss of coolant flow). In addition, the dynamic behavior of sub-critical systems to typical plant transients such as protected or unprotected loss of flow (LOF) or heat sink (LOH) transients are not necessarily substantially different from the plant dynamic behavior of critical systems if the reactivity feedback coefficients of the ADS design are unfavorable. As expected, the state of sub-criticality and the temperature feedback coefficients, such as Doppler and coolant temperature coefficient, play dominant roles in determining the course and direction of plant transients. Should the combination of these safety coefficients be very unfavorable, not much additional margin in safety may be gained by making a critical system only sub-critical (i.e. k eff ∼0.95). A careful optimization procedure between the selected operating level of sub-criticality, the safety reactivity coefficients and the possible need for additional reactivity control systems seems, therefore, advisable during the early

Investigation on in-vessel thermal transients in a fast breeder reactor

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Kasahara, Naoto

1999-01-01

Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of the phenomena in the design of the internal structures in an LMFBR plenum. To evaluate thermal stress characteristics for the inner barrel in a typical LMFBR upper plenum, numerical analysis was carried out with a multi-dimensional thermohydraulics code AQUA for a scram condition from full power operation conditions. Thereafter, thermal stress conditions for the inner barrel were evaluated by the use of a structural analysis code FINAS with the thermohydraulic results calculated by the AQUA code as boundary conditions. From the thermohydraulic analysis and the thermal stress analysis, the following results have been obtained. (1) A large axial temperature gradient was calculated at the region between the upper and lower flow holes located on the inner barrel. The axial position of the thermal stratification interface was fixed in the various circumferential directions. As for the comparison with a 40% operation condition, maximum temperature gradients at the lower flow hole region indicated a 2 times value of that in the 40% operation condition. (2) Transient thermal stratification phenomena were observed after 120 sec from the reactor scram in the numerical results. These tendencies on thermal stratification phenomena were sameness with the transient results from the 40% operation condition. (3) During the reactor trip from full power operation, large temperature gradient in both vertical and sectional direction are enforced around the lower flow hole, since there exists flow pass of low temperature sodium through this hole. As a result, the maximum thermal stress within 32.6 kg/mm 2 was predicted at the lower flow hole when considering stress concentration at the hole edge. (J.P.N.)

Effect of burnup on the response of stainless steel-clad mixed-oxide fuels to simulated thermal transients

International Nuclear Information System (INIS)

Fenske, G.R.; Badyopadhyay, G.

1981-01-01

Direct electrical heating experiments were performed on irradiated fuel to study the fuel and cladding response as a function of burnup during a slow thermal transient. The results indicated that the nature and extent of the fuel and cladding behavior depended on the quantity of fission gas retained in the fuel. Fission-gas-driven fuel ejection occurred as the molten cladding flowed down the stack exposing bare, radially unrestrained fuel. The fuel dispersion occurred in the absence of molten fuel and the amount of fuel ejected increased with increasing burnup. 31 refs

Transient subcritical crack-growth behavior in transformation-toughened ceramics

International Nuclear Information System (INIS)

Dauskardt, R.H.; Ritchie, R.O.; Carter, W.C.; Veirs, D.K.

1990-01-01

Transient subcritical crack-growth behavior following abrupt changes in the applied load are studied in transformation-toughened ceramics. A mechanics analysis is developed to model the transient nature of transformation shielding of the crack tip, K s , with subcritical crack extension following the applied load change. conditions for continued crack growth, crack growth followed by arrest, and no crack growth after the load change, are considered and related to the magnitude and sign of the applied load change and to materials properties such as the critical transformation stress. The analysis is found to provide similar trends in K s compared to values calculated from experimentally measured transformation zones in a transformation-toughened Mg-PSZ. In addition, accurate prediction of the post load-change transient crack-growth behavior is obtained using experimentally derived steady-state subcritical crack-growth relationships for cyclic fatigue in the same material

Thermal-hydraulics of the Loviisa reactor pressure vessel overcooling transients

International Nuclear Information System (INIS)

Tuomisto, Harri.

1987-06-01

In the Loviisa reactor pressure vessel safety analyses, the thermal-hydraulics of various overcooling transients has been evaluated to give pertinent initial data for fracture-mechanics calculations. The thermal-hydraulic simulations of the developed overcooling scenarios have been performed using best-estimate thermal-hydraulic computer codes. Experimental programs have been carried out to study phenomena related to natural circulation interruptions in the reactor coolant system. These experiments include buoyancy-induced phenomena such as thermal mixing and stratification of cold high-pressure safety injection water in the cold legs and the downcomer, and oscillations of the single-phase natural circulation. In the probabilistic pressurized thermal shock study, the Loviisa training simulator and the advanced system code RELAP5/MOD2 were utilized to simulate selected sequences. Flow stagnation cases were separately calculated with the REMIX computer program. The methods employed were assessed for these calculations against the plant data and own experiments

Numerical investigation of thermal behaviors in lithium-ion battery stack discharge

International Nuclear Information System (INIS)

Liu, Rui; Chen, Jixin; Xun, Jingzhi; Jiao, Kui; Du, Qing

2014-01-01

Highlights: • The thermal behaviors of a Li-ion battery stack have been investigated by modeling. • Parametric studies have been performed focusing on three different cooling materials. • Effects of discharge rate, ambient temperature and Reynolds number are examined. • General guidelines are proposed for the thermal management of a Li-ion battery stack. - Abstract: Thermal management is critically important to maintain the performance and prolong the lifetime of a lithium-ion (Li-ion) battery. In this paper, a two-dimensional and transient model has been developed for the thermal management of a 20-flat-plate-battery stack, followed by comprehensive numerical simulations to study the influences of ambient temperature, Reynolds number, and discharge rate on the temperature distribution in the stack with different cooling materials. The simulation results indicate that liquid cooling is generally more effective in reducing temperature compared to phase-change material, while the latter can lead to more homogeneous temperature distribution. Fast and deep discharge should be avoided, which generally yields high temperature beyond the acceptable range regardless of cooling materials. At low or even subzero ambient temperatures, air cooling is preferred over liquid cooling because heat needs to be retained rather than removed. Such difference becomes small when the ambient temperature increases to a mild level. The effects of Reynolds number are apparent in liquid cooling but negligible in air cooling. Choosing appropriate cooling material and strategy is particularly important in low ambient temperature and fast discharge cases. These findings improve the understanding of battery stack thermal behaviors and provide the general guidelines for thermal management system. The present model can also be used in developing control system to optimize battery stack thermal behaviors

Transient thermal sensation and comfort resulting from adjustment of clothing insulation

DEFF Research Database (Denmark)

Goto, Tomonobu; Toftum, Jørn; Fanger, Povl Ole

2003-01-01

This study investigated the transient effects on human thermal responses of clothing adjustments. Two different levels of activity were tested, and the temperature was set to result in a warm or cool thermal sensation at each activity level. The subjects (12 females and 12 males) wore identical...... uniforms and were asked to take off or don a part of the uniform after they had adapted to the experimental conditions for more than 20 minutes. The results showed that the thermal sensation votes responded immediately to the adjustment of clothing insulation and reached a new steady-state level within 5...

Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

International Nuclear Information System (INIS)

Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

2004-01-01

As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

Measurement of thermal conductivity of uranium metal using transient plane source technique

International Nuclear Information System (INIS)

Subramanian, G.G.S.; Bapuji, T.; Panneerselvam, G.; Antony, M.P.; Nagarajan, K.

2012-01-01

Thermo physical properties of fuel, cladding and structural materials play a significant role in the reactor operation. Thermal conductivity is one of the most important physical properties of the fuel which determines the maximum linear heat rating of the fuel in a reactor. As part of this study, the thermal conductivity of uranium metal was measured using a transient plane source (TPS) by Hot-disc method

Analysis of incoloy 800ht alloy tested in thermal transient conditions

International Nuclear Information System (INIS)

Velciu, L.; Meleg, T.; Nitu, A.; Popa, L.

2015-01-01

This paper investigated Incoloy 800 HT alloy after following thermal transient tests: fast heating rates (50° and 90°C/minute) up to 1,000°C, maintaining this temperature level (0 and 60 minutes), furnace-cooling until 220°C, and then air-cooling. This alloy is one of the candidate materials for construction of the steam generators of the future NPP reactors. The analysis consisted in metallographic examination and traction tests. The samples were investigated using the Olympus GX 71 optical microscope, the OPL microdurometer with automatic cycle and WALTER BAI traction device. The average grain size was determined by linear interception method. The micro hardness was calculated by the relationship from the device technical book. On the traction diagrams were obtained: strength resistance (Rm), elongation at rupture (A) and elastic modulus (E). The tested alloy was compared with the ''as received'' material, and the results showed a good behavior of this alloy in the presented conditions. (authors)

Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests

International Nuclear Information System (INIS)

Bauer, T.H.; Klickman, A.E.; Lo, R.K.; Rhodes, E.A.; Robinson, W.R.; Stanford, G.S.; Wright, A.E.

1986-01-01

TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors

Experimental studies on thermal hydraulic responses for transient operations of the SMART-P

International Nuclear Information System (INIS)

Choi, K.Y.; Park, H.S.; Cho, S.; Park, C.K.; Lee, S.J.; Song, C.H.; Chung, M.K.

2005-01-01

Full text of publication follows: Thermal hydraulic responses for transient operations of the SMART-P are experimentally investigated by using a integral effect test facility. This test facility (VISTA) has been constructed to simulate the SMART-P, which is a pilot plant of the SMART. The SMART-P is an advanced modular integral type pressurized water reactor (65 MWt) whose major RCS components, such as main coolant pumps, helical-coiled tube bundle steam generators and pressurizers, are contained in a reactor vessel. This integral design approach eliminates the large coolant loop piping, thus eliminates the occurrence of a large break LOCA. Passive Residual Heat Removal System (PRHRS) is installed to prevent overheating and over-pressurization of the primary system during accidental conditions. The PRHRS of the SMART-P removes the core decay heat by natural circulation of the two-phase fluid. The VISTA facility is a full height and 1/96 volume scaled test facility with respect to the SMART-P and will be used to understand the thermal-hydraulic responses following transients and finally to verify the system design of the SMART-P. The experimental data from the VISTA facility will be essential to system designers to resolve open issues relevant to the design of the SMART-P. The full functional control logics are implanted into the VISTA facility to cope with abnormal transients. The core of the facility can be selectively controlled by either a T-control or a T+N control method. The T-control method is a control method to adjust the core power according to the core exit coolant temperature and is designed to be used for high primary coolant flow conditions. On the other hand, the T+N control method is for low primary coolant flow conditions and it uses core exit temperature as well as core power itself as control inputs. The thermal hydraulic responses are carefully investigated according to different core control methods. Several experiments have been performed to

The effect of a transient thermal lens on the Nd:YVO4 laser output

International Nuclear Information System (INIS)

Yi, Jonghoon; Lee, Kangin; Kim, Youngjung; Kwon, Jinhyuk

2010-01-01

A Nd:YVO 4 laser was pumped by using a diode laser, which has maximum cw pump power of 1 W. The driving current of the diode laser was modulated to have a square waveform. The Nd:YVO 4 laser output power increased linearly and then saturated when the quasi-cw diode laser pulse was focused on the crystal. When the same diode laser pulse was applied on the crystal, transient thermal lensing in the Nd:YVO 4 crystal was monitored by using a probe beam in a non-lasing condition. The TEM 00 mode diameter of the laser was calculated as a function of the focal length of the thermal lens. The results indicated that transient thermal lensing in the crystal was the main cause of the temporally varying output.

Modeling of Transient Response of the Wickless Heat Pipes

International Nuclear Information System (INIS)

Hussien, A.K.A.

2013-01-01

Thermosyphons transient response for startup from ambient temperature to steady state until shutdown conditions, is considered a stringent necessity for applications such as electronic, solar, geothermal and even nuclear reactors safety systems. This typically returns to the need to keep the temperature within certain limits before reaching critical conditions. A simple network model is derived for describing the transient response of closed two-phase thermosyphon (CTPT) at startup and shutdown states. In addition, for predicting the effect of operational characteristics of water/copper closed two-phase thermosyphon such as thermal load, filling ratio, evaporator length, and thermosyphon tube diameter. The thermosyphons operation was considered a thermal network of various components with different thermal resistances and dynamic responses. The network model consists of six sub-models. These models are pure conduction in walls of evaporator, adiabatic and condenser, and convection in evaporator pool, evaporator film, and condenser film. So, an energy balance for each sub-model was done to estimate temperatures, heat transfer coefficients, thermal resistances, time constant, and other thermal characteristics that describe the required transient response of the closed two-phase thermosyphon. Governing equations of the transient thermosyphon behavior can be simplified into a set of first-order linear ordinary differential equations. The Runge-Kutta method can be used to obtain transient thermosyphon temperatures from these equations.

Thermodynamic model of a thermal storage air conditioning system with dynamic behavior

International Nuclear Information System (INIS)

Fleming, Evan; Wen, Shaoyi; Shi, Li; Silva, Alexandre K. da

2013-01-01

Highlights: • We developed an automotive thermal storage air conditioning system model. • The thermal storage unit utilizes phase change materials. • We use semi-analytic solution to the coupled phase change and forced convection. • We model the airside heat exchange using the NTU method. • The system model can incorporate dynamic inputs, e.g. variable inlet airflow. - Abstract: A thermodynamic model was developed to predict transient behavior of a thermal storage system, using phase change materials (PCMs), for a novel electric vehicle climate conditioning application. The main objectives of the paper are to consider the system’s dynamic behavior, such as a dynamic air flow rate into the vehicle’s cabin, and to characterize the transient heat transfer process between the thermal storage unit and the vehicle’s cabin, while still maintaining accurate solution to the complex phase change heat transfer. The system studied consists of a heat transfer fluid circulating between either of the on-board hot and cold thermal storage units, which we refer to as thermal batteries, and a liquid–air heat exchanger that provides heat exchange with the incoming air to the vehicle cabin. Each thermal battery is a shell-and-tube configuration where a heat transfer fluid flows through parallel tubes, which are surrounded by PCM within a larger shell. The system model incorporates computationally inexpensive semi-analytic solution to the conjugated laminar forced convection and phase change problem within the battery and accounts for airside heat exchange using the Number of Transfer Units (NTUs) method for the liquid–air heat exchanger. Using this approach, we are able to obtain an accurate solution to the complex heat transfer problem within the battery while also incorporating the impact of the airside heat transfer on the overall system performance. The implemented model was benchmarked against a numerical study for a melting process and against full system

Analytical transient analysis of Peltier device for laser thermal tuning

Science.gov (United States)

Sheikhnejad, Yahya; Vujicic, Zoran; Almeida, Álvaro J.; Bastos, Ricardo; Shahpari, Ali; Teixeira, António L.

2017-08-01

Recently, industrial trends strongly favor the concepts of high density, low power consumption and low cost applications of Datacom and Telecom pluggable transceiver modules. Hence, thermal management plays an important role, especially in the design of high-performance compact optical transceivers. Extensive care should be taken on wavelength drift for thermal tuning lasers using thermoelectric cooler and indeed, accurate expression is needed to describe transient characteristics of the Peltier device to achieve maximum controllability. In this study, the exact solution of governing equation is presented, considering Joule heating, heat conduction, heat flux of laser diode and thermoelectric effect in one dimension.

Fission gas behavior in mixed-oxide fuel during transient overpower

International Nuclear Information System (INIS)

Randklev, E.H.; Treibs, H.A.; Mastel, B.; Baldwin, D.L.

1979-01-01

Fission gas behavior can be important in determining fuel pin and core performance during a reactor transient. The results are presented of examinations characterizing the changes in microstructural distribution and retention of fission gas in fuel for a series of transient overpower (50 cents/s) tested mixed-oxide fuel pins and their steady state siblings

Realistic thermal transient margin analysis of 'MONJU' based on plant performance measurements. Reactor vessel outlet nozzle and evaporator feed water inlet tube sheet of the manual reactor plant trip

International Nuclear Information System (INIS)

Yamada, Fumiaki; Mori, Takero

2005-01-01

In order to develop technologies and achieve safe and stable operation of Monju' as well as realize optimized design and construction of safe and economically competitive fast breeder reactors, the authors are evaluating design approach applied to 'Monju' based on actually measured behavioral data during plant operations. This report uses actual measured characteristic data of 'Monju' during a plant trip test obtained at a commissioning stage with up to 40% power output and introduces plant thermal hydraulic behavior analysis in a representative thermal transient event, i.e. a manual plant trip. Thermal transient driven loads incurred by the reactor vessel outlet nozzle and by the evaporator feed water inlet tube sheet were further derived by structural analyses and were compared with the previously derived values in the design stage and with the limit values. Though the reactor vessel outlet nozzle was exposed to larger temperature change in the trip test than the analytical prediction, the newly calculated mechanical load was about 50% of the previous evaluation in the design stage. Also, the newly analyzed mechanical load incurred by the evaporator feed water inlet tube sheet in this event had a large margin against the limit value of cumulative damage cycle fraction, although the observed temperature disturbance in a steam blow test was wilder than the analytical prediction. Thus we concluded that the Monju' plant has an assured safety margin against thermal transient in plant trip events. (author)

Numerical simulation of time-dependent deformations under hygral and thermal transient conditions

International Nuclear Information System (INIS)

Roelfstra, P.E.

1987-01-01

Some basic concepts of numerical simulation of the formation of the microstructure of HCP are outlined. The aim is to replace arbitrary terms like aging by more realistic terms like bond density in the xerogel and bonds between hydrating particles of HCP. Actual state parameters such as temperature, humidity and degree of hydration can be determined under transient hygral and thermal conditions by solving numerically a series of appropriate coupled differential equations with given boundary conditions. Shrinkage of a composite structure without crack formation, based on calculated moisture distributions, has been determined with numerical concrete codes. The influence of crack formation, tensile strain-hardening and softening on the total deformation of a quasi-homogeneous drying material has been studied by means of model based on FEM. The difference between shrinkage without crack formation and shrinkage with crack formation can be quantified. Drying shrinkage and creep of concrete cannot be separated. The total deformation depends on the superimposed stress fields. Transient hygral deformation can be realistically predicted if the concept of point properties is applied rigorously. Transient thermal deformation has to be dealt with in the same way. (orig./HP)

Transient thermal analysis of cryocondensation pump for JET

International Nuclear Information System (INIS)

Baxi, C.B.; Obert, W.

1993-08-01

A cryopump with pumping speed of 50,000 1/sec is planned to be installed in the Joint European Torus (JET) as part of the pumped divertor. The purpose of this pump is to control the plasma impurities. The pump consists of a helium panel cooled by supercritical helium and a nitrogen shield cooled by liquid nitrogen. This paper presents the following transient thermal flow analysis for this cryopump: 1. Consequences of loss of torus vacuum on helium panel. 2. Cool down of the nitrogen shield form 300 K to 80 K

Characterization of various two-phase materials based on thermal conductivity using modified transient plane source method

Science.gov (United States)

Jayachandran, S.; Prithiviraajan, R. N.; Reddy, K. S.

2017-07-01

This paper presents the thermal conductivity of various two-phase materials using modified transient plane source (MTPS) technique. The values are determined by using commercially available C-Therm TCi apparatus. It is specially designed for testing of low to high thermal conductivity materials in the range of 0.02 to 100 Wm-1K-1 within a temperature range of 223-473 K. The results obtained for the two-phase materials (solids, powders and liquids) are having an accuracy better than 5%. The transient method is one of the easiest and less time consuming method to determine the thermal conductivity of the materials compared to steady state methods.

Transient behavior of redox flow battery connected to circuit based on global phase structure

Science.gov (United States)

Mannari, Toko; Hikihara, Takashi

A Redox Flow Battery (RFB) is one of the promising energy storage systems in power grid. An RFB has many advantages such as a quick response, a large capacity, and a scalability. Due to these advantages, an RFB can operate in mixed time scale. Actually, it has been demonstrated that an RFB can be used for load leveling, compensating sag, and smoothing the output of the renewable sources. An analysis on transient behaviors of an RFB is a key issue for these applications. An RFB is governed by electrical, chemical, and fluid dynamics. The hybrid structure makes the analysis difficult. To analyze transient behaviors of an RFB, the exact model is necessary. In this paper, we focus on a change in a concentration of ions in the electrolyte, and simulate the change with a model which is mainly based on chemical kinetics. The simulation results introduces transient behaviors of an RFB in a response to a load variation. There are found three kinds of typical transient behaviors including oscillations. As results, it is clarified that the complex transient behaviors, due to slow and fast dynamics in the system, arise by the quick response to load.

TRAC-PF1 analyses of potential pressurized-thermal-shock transients at a Combustion-Engineering PWR

International Nuclear Information System (INIS)

Koenig, J.E.; Spriggs, G.D.; Smith, R.C.

1984-01-01

Los Alamos is participating in a program to assess the risk of pressurized thermal shock (PTS) to a reactor vessel. Our role is to provide best-estimate thermal-hydraulic analyses of 12 postulated overcooling transients using TRAC-PF1. These transients are hypothetical and include multiple operator/equipment failures. Calvert Cliffs/Unit-1, a Combustion-Engineering plant, is the pressurized water reactor modeled for this study. The utility and the vendor supplied information for the comprehensive TRAC-PF1 model. Secondary and primary breaks from both hot-zero-power and full-power conditions were simulated for 7200 s (2 h). Low bulk temperatures and loop-flow stagnation while the system was at a high pressure were of particular interest for PTS analysis. Three transients produced primary temperatures below 405 K (270 0 F - the NRC screening criterion) with system repressurization. Six transients indicated flow stagnation would occur in one loop but not both. One transient showed flow stagnation might occur in both loops. Oak Ridge National Laboratory will do fracture-mechanics analysis using these TRAC-PF1 results and make the final determination of the risk of PTS

Transient failure behavior of HT9

International Nuclear Information System (INIS)

Huang, F.H.

1994-07-01

Alloy HT9 has-been chosen as candidate materials for fast and fusion reactor applications because the.material exhibits excellent resistance to void swelling. However, ferritic alloys are known to undergo a ductile-brittle transition as the test temperature is decreased. This inherent problem has limited their applications to reactor component materials subjected to low neutron exposures. Despite the ductile-brittle transition problem, results show that the materials exhibit superior resistance to fracture under very high neutron fluences at irradiation temperatures above 380C. Results also show that the transient behavior for HT9 cladding specimens taken from the fuel column region and cladding taken from outside the fuel column or unirradiated cladding are the same. HT9 cladding maintained its transient strength with irradiation to a fluence of 9 x 10 22 n/cm 2 (E > 0.1 MeV)

Effects of thermal stratification on transient free convective flow of a ...

Indian Academy of Sciences (India)

2016-09-22

Sep 22, 2016 ... as well as average skin friction and the rate of heat transfer of nanofluids are discussed and represented graphically. The results are found to be in good agreement with the existing results in literature. Keywords. Nanofluid; thermal stratification; transient; isothermal vertical plate. PACS Nos 44.20.+b; 47; 44.

An analysis of the transient's social behavior in the radiological emergency planning zone

International Nuclear Information System (INIS)

Bang, Sun Young; Lee, Gab Bock; Chung, Yang Geun; Lee, Jae Eun

2007-01-01

The purpose of this study is to analyze the social behavior, especially, the evacuation-related social behavior, of the transients in the radiological Emergency Planning Zone (EPZ) of nuclear power plants. So, the meaning and kinds of the evacuation and the significance of the Trip Generation Time (TGT) have been reviewed. The characteristics of the social behavior of the transient around Ulchin, Wolsong and Kori sites was analyzed through field surveys by using the questionnaire. The major findings of this research implications are as follows. First, for securing the safe evacuation, the alternatives to effectively provide the information on the evacuation warning may be prepared. Second, it is necessary to establish the education and training of transient's evacuation. Third, it is needed that the cause and background of the evacuation refusal are identified and the new response plan to secure transient's safety is prepared

Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

Energy Technology Data Exchange (ETDEWEB)

Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)

1997-07-01

This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

Transient Thermal Analyses of Passive Systems on SCEPTOR X-57

Science.gov (United States)

Chin, Jeffrey C.; Schnulo, Sydney L.; Smith, Andrew D.

2017-01-01

As efficiency, emissions, and noise become increasingly prominent considerations in aircraft design, turning to an electric propulsion system is a desirable solution. Achieving the intended benefits of distributed electric propulsion (DEP) requires thermally demanding high power systems, presenting a different set of challenges compared to traditional aircraft propulsion. The embedded nature of these heat sources often preclude the use of traditional thermal management systems in order to maximize performance, with less opportunity to exhaust waste heat to the surrounding environment. This paper summarizes the thermal analyses of X-57 vehicle subsystems that don't employ externally air-cooled heat sinks. The high-power battery, wires, high-lift motors, and aircraft outer surface are subjected to heat loads with stringent thermal constraints. The temperature of these components are tracked transiently, since they never reach a steady-state equilibrium. Through analysis and testing, this report demonstrates that properly characterizing the material properties is key to accurately modeling peak temperature of these systems, with less concern for spatial thermal gradients. Experimentally validated results show the thermal profile of these systems can be sufficiently estimated using reduced order approximations.

Transient forced convection with viscous dissipation to power-law fluids in thermal entrance region of circular ducts with constant wall heat flux

International Nuclear Information System (INIS)

Dehkordi, Asghar Molaei; Mohammadi, Ali Asghar

2009-01-01

A numerical investigation was conducted on the transient behavior of a hydrodynamically, fully developed, laminar flow of power-law fluids in the thermally developing entrance region of circular ducts taking into account the effect of viscous dissipation but neglecting the effect of axial conduction. In this regard, the unsteady state thermal energy equation was solved by using a finite difference method, whereas the steady state thermal energy equation without wall heat flux was solved analytically as the initial condition of the former. The effects of the power-law index and wall heat flux on the local Nusselt number and thermal entrance length were investigated. Moreover, the local Nusselt number of steady state conditions was correlated in terms of the power-law index and wall heat flux and compared with literature data, which were obtained by an analytic solution for Newtonian fluids. Furthermore, a relationship was proposed for the thermal entrance length

Eigenvalue solutions in finite element thermal transient problems

International Nuclear Information System (INIS)

Stoker, J.R.

1975-01-01

The eigenvalue economiser concept can be useful in solving large finite element transient heat flow problems in which the boundary heat transfer coefficients are constant. The usual economiser theory is equivalent to applying a unit thermal 'force' to each of a small sub-set of nodes on the finite element mesh, and then calculating sets of resulting steady state temperatures. Subsequently it is assumed that the required transient temperature distributions can be approximated by a linear combination of this comparatively small set of master temperatures. The accuracy of a reduced eigenvalue calculation depends upon a good choice of master nodes, which presupposes at least a little knowledge about what sort of shape is expected in the unknown temperature distributions. There are some instances, however, where a reasonably good idea exists of the required shapes, permitting a modification to the economiser process which leads to greater economy in the number of master temperatures. The suggested new approach is to use manually prescribed temperature distributions as the master distributions, rather than using temperatures resulting from unit thermal forces. Thus, with a little pre-knowledge one may write down a set of master distributions which, as a linear combination, can represent the required solution over the range of interest to a reasonable engineering accuracy, and using the minimum number of variables. The proposed modified eigenvalue economiser technique then uses the master distributions in an automatic way to arrive at the required solution. The technique is illustrated by some simple finite element examples

Transient behavior of interface state continuum at InP insulator-semiconductor interface

International Nuclear Information System (INIS)

Hasegawa, H.; Masuda, H.; He, L.; Luo, J.K.; Sawada, T.; Ohno, H.

1987-01-01

To clarify the drain current drift mechanism in InP MISFETs, an isothermal capacitance transient spectroscopy (ICTS) study of the interface state continuum is made on the anodic Al 2 O 3 /native oxide/ InP MIS system. Capture behavior is temperature-independent, non-exponential and extremely slow, whereas emission behavior is temperature- and bias- dependent, and is much faster. The observed behavior is explained quantitatively by the disorder induced gap state (DIGS) model, where states are distributed both in energy and in space. By comparing the transient behavior of interface states with the observed drift behavior of MISFETs, it is concluded that the electron capture by the DIGS continuum is responsible for the drain current drift of MISFETs. This led to a complete computer simulation of the observed current drift behavior

Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

International Nuclear Information System (INIS)

Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

1999-02-01

Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

Characteristic behavior of bubbles and slugs in transient two-phase flow using image-processing method

International Nuclear Information System (INIS)

Goto, Shoji; Ishizaki, Yasuo; Ohashi, Hirotada; Akiyama, Mamoru

1995-01-01

Simulation of transient two-phase flow has been performed by solving transient hydrodynamic equations. However, constitution relations used in this simulation are primarily based on steady-state experimental results. Thus it is important to understand the transient behavior of bubbles and slugs, in particular, transient behavior of the void fraction, the interfacial area and the flow pattern, to confirm the applicability of the present simulation method and to advance two-phase flow simulation further. The present study deals with measurement of transient two-phase flow. We have measured local and instantaneous void fractions using imaging techniques, and compared the experimental data with simulation results. (author)

Transient behavior of enrichment of tritium water in adsorption-distillation column

International Nuclear Information System (INIS)

Fukada, Satoshi

2006-01-01

Enrichment of tritium in an adsorption-distillation column was experimentally investigated under the two processes of simple distillation and total-reflux distillation. Adsorption of water on silica-gel pellets enhanced the total isotope separation factor in the water distillation column. The transient behavior of tritium enrichment was analyzed using material balance equations of tritium and water in each cell with a height corresponding to HETP. The experimental transient behavior was well simulated by the material balance equations with additional assumptions on vapor and liquid flow rates regardless of the different processes of simple distillation and total-reflux distillation. (author)

Transient Analysis and Design Improvement of a Gas Turbine Rotor Based on Thermal-Mechanical Method

Directory of Open Access Journals (Sweden)

Yang Liu

2018-01-01

Full Text Available The rotor is the core component of a gas turbine, and more than 80% of the failures in gas turbines occur in the rotor system, especially during the start-up period. Therefore, the safety assessment of the rotor during the start-up period is essential for the design of the gas turbine. In this paper, the transient equivalent stress of a gas turbine rotor under the cold start-up condition is investigated and the novel tie rod structure is introduced to reduce the equivalent stress. Firstly, a three-dimensional finite element model of the gas turbine rotor is built, and nonlinear contact behaviors such as friction are taken into account. Secondly, the convective heat transfer coefficients of the gas turbine rotor under the cold start-up condition are calculated using thermal dynamic theory. The transient analysis of the gas turbine rotor is conducted considering the thermal load, the centrifugal load, and the pretightening force. The temperature and stress distributions of the rotor under the cold start-up condition are shown in detail. In particular, the generation mechanism of maximum equivalent stress for tie rods and the change tendency of the pretightening force are illustrated in detail. The tie rod holes of the rear shaft and the turbine tie rod are the dangerous locations during the start-up period. Finally, a novel tie rod is proposed to reduce the maximum equivalent stress at the dangerous location. The maximum equivalent stress at this location is decreased by 15%. This paper provides some reference for the design of the gas turbine rotor.

Behavior of mixed-oxide fuel elements during an overpower transient

International Nuclear Information System (INIS)

Tsai, H.; Shikakura, S.

1993-01-01

A slow-ramp (0.1%/s), extended overpower (∼90%) transient test was conducted in EBR-II on 19 mixed-oxide fuel elements with conservative, moderate, and aggressive designs. Claddings for the elements were Type 316, D9, or PNC-316 stainless steel. Before the transient, the elements were preirradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions to burnups of 2.5-9.7 at%. Cladding integrity during the transient test was maintained by all fuel elements except one, which had experienced substantial overtemperature in the earlier stedy-state irradiation. Extensive centerline fuel melting occurred in all test elements. Significantly, this melting did not cause any elements to breach, although it did have a strong effect on the other aspects of fuel element behavior. (orig.)

Transient Diabetes Insipidus Following Thermal Burn; A Case Report and Literature Review.

Science.gov (United States)

Dash, Suvashis; Ghosh, Shibajyoti

2017-10-01

Diabetes insipidus is a disease charaterised by increased urine production and thrist. Neurogenic diabetes insipidus following head trauma,autoimmune disease and infection is quite common but diabetes insipidus following thermal burn injury is a rare complication.We should know about this complication as its management need a comprehensive approach for satisfactory outcome. Thermal burn can cause different complications in early post burn period like electrolyte imbalance, dehydration, acute renal failure, but diabetes insipidus is a very rare and unusual complication that may come across in thermal burn. We should be aware about this condition to prevent and treat mortality and morbidity in burn patients. We have reported a case of transient diabetes insipidus in a patient of thermal burn in early post burn period. Patient was treated accordingly, leading to complete recovery.

Proceedings of transient thermal-hydraulics and coupled vessel and piping system responses 1991

International Nuclear Information System (INIS)

Wang, G.Y.; Shin, Y.W.; Moody, F.J.

1991-01-01

This book reports on transient thermal-hydraulics and coupled vessel and piping system responses. Topics covered include: nuclear power plant containment designs; analysis of control rods; gate closure of hydraulic turbines; and shock wave solutions for steam water mixtures in piping systems

The simulation of transients in thermal plant. Part I: Mathematical model

International Nuclear Information System (INIS)

Morini, G.L.; Piva, S.

2007-01-01

This paper deals with the simulation of the transient behaviour of thermal plant with control systems. It is always more difficult for a designer to predict the effects on the plant of the control processes because of the increasing complexity of plants and control systems. The easiest way to obtain information about the dynamic behaviour of a thermal plant at the design-stage involves assessing the suitability of specific computer codes. To this end, the present work demonstrates that nowadays it is possible, by using the opportunities offered by some general purpose calculation systems, to obtain such significant information. It is described how a 'thermal-library' of customized blocks (one for each component of a thermal plant such as valves, boilers, and pumps) can be built and used, in an intuitive way, to study any kind of plant. As an example, the dynamic behaviour of a residential heating system will be shown in a companion paper, forming part II of the present article

Thermal performance behavior of a domestic hot water solar storage tank during consumption operation

International Nuclear Information System (INIS)

Dehghan, A.A.; Barzegar, A.

2011-01-01

Transient thermal performance behavior of a vertical storage tank of a domestic solar water heating system with a mantle heat exchanger has been investigated numerically in the discharge/consumption mode. It is assumed that the tank is initially stratified during its previous heat storing/charging operation. During the discharging period, the city cold water is fed at the bottom of the tank and hot water is extracted from its top outlet port for consumption. Meanwhile, the collector loop is assumed to be active. The conservation equations in the axis-symmetric cylindrical co-ordinate have been used and discretised by employing the finite volume method. The low Reynolds number (LRN) k - ω model is utilized for treating turbulence in the fluid. The influence of the tank Grashof number, the incoming cold fluid Reynolds number and the size of the inlet port of the heat storage tank on the transient thermal characteristics of the tank is investigated and discussed. It is found that for higher values of Grashof number, the pre-established thermal stratification is well preserved during the discharging operation mode. It is also noticed that in order to have a tank with a proper thermal performance and or have least mixing inside the tank during the consumption period, the tank inflow Reynolds number and or its inflow port diameter should be kept below certain values. In these cases, the storage tank is enabling to provide proper amount of hot water with a proper temperature for consumption purposes.

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Thermal states of coldest and hottest neutron stars in soft X-ray transients

OpenAIRE

Yakovlev, D. G.; Levenfish, K. P.; Potekhin, A. Y.; Gnedin, O. Y.; Chabrier, G.

2003-01-01

We calculate the thermal structure and quiescent thermal luminosity of accreting neutron stars (warmed by deep crustal heating in accreted matter) in soft X-ray transients (SXTs). We consider neutron stars with nucleon and hyperon cores and with accreted envelopes. It is assumed that an envelope has an outer helium layer (of variable depth) and deeper layers of heavier elements, either with iron or with much heavier nuclei (of atomic weight A > 100) on the top (Haensel & Zdunik 1990, 2003, as...

Thermal conductivity measurement of the He-ion implanted layer of W using transient thermoreflectance technique

Energy Technology Data Exchange (ETDEWEB)

Qu, Shilian; Li, Yuanfei [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Wang, Zhigang [Department of Electronic Engineering, Dalian University of Technology, Dalian 116024 (China); Jia, Yuzhen [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Science and Technology on Reactor Fuel and Materials Laboratory, Nuclear Power Institute of China, Chengdu 610213 (China); Li, Chun [School of Mechanical and Materials Engineering, North China University of Technology, Beijing 100144 (China); Xu, Ben; Chen, Wanqi [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Bai, Suyuan [School of Physics and Electronic Technology, Liaoning Normal University, Dalian 116029 (China); Huang, Zhengxing; Tang, Zhenan [Department of Electronic Engineering, Dalian University of Technology, Dalian 116024 (China); Liu, Wei, E-mail: liuw@mail.tsinghua.edu.cn [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China)

2017-02-15

Transient thermoreflectance method was applied on the thermal conductivity measurement of the surface damaged layer of He-implanted tungsten. Uniform damages tungsten surface layer was produced by multi-energy He-ion implantation with thickness of 450 nm. Result shows that the thermal conductivity is reduced by 90%. This technique was further applied on sample with holes on the surface, which was produced by the He-implanted at 2953 K. The thermal conductivity decreases to 3% from the bulk value.

Study of a spur gear dynamic behavior in transient regime

Science.gov (United States)

Khabou, M. T.; Bouchaala, N.; Chaari, F.; Fakhfakh, T.; Haddar, M.

2011-11-01

In this paper the dynamic behavior of a single stage spur gear reducer in transient regime is studied. Dynamic response of the single stage spur gear reducer is investigated at different rotating velocities. First, gear excitation is induced by the motor torque and load variation in addition to the fluctuation of meshing stiffness due to the variation of input rotational speed. Then, the dynamic response is computed using the Newmark method. After that, a parameter study is made on spur gear powered in the first place by an electric motor and in the second place by four strokes four cylinders diesel engine. Dynamic responses come to confirm a significant influence of the transient regime on the dynamic behavior of a gear set, particularly in the case of engine acyclism condition.

Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes

Energy Technology Data Exchange (ETDEWEB)

Kim, Yang Hoon; Han, Kee Soo; Moon, Bok Ja; Jang, Misuk [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

2014-05-15

In this presentation, simple thermal hydraulic behavior is analyzed using three codes to show the possibility of using alternative codes. We established three models of simple u-tube manometer using three different codes. CATHENA (Canadian Algorithm for Thermal hydraulic Network Analysis), RELAP (Reactor Excursion and Leak Analysis Program), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are used for this analysis. CATHENA and RELAP are widely used codes for the analysis of system behavior of CANDU and PWR. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. In this paper, the internal behavior of u-tube manometer was analyzed using 3 codes, CATHENA, RELAP and GOTHIC. The general transient behavior is similar among 3 codes. However, the behavior simulated using GOTHIC shows some different trend compared with the results from the other 2 codes at the end of the transient. It would be resulted from the use of different physical model in GOTHIC, which is specialized for the multi-phase thermal hydraulic behavior analysis of containment system unlike the other two codes.

Coupled transient thermo-fluid/thermal-stress analysis approach in a VTBM setting

International Nuclear Information System (INIS)

Ying, A.; Narula, M.; Zhang, H.; Abdou, M.

2008-01-01

A virtual test blanket module (VTBM) has been envisioned as a utility to aid in streamlining and optimizing the US ITER TBM design effort by providing an integrated multi-code, multi-physics modeling environment. Within this effort, an integrated simulation approach is being developed for TBM design calculations and performance evaluation. Particularly, integrated thermo-fluid/thermal-stress analysis is important for enabling TBM design and performance calculations. In this paper, procedures involved in transient coupled thermo-fluid/thermal-stress analysis are investigated. The established procedure is applied to study the impact of pulsed operational phenomenon on the thermal-stress response of the TBM first wall. A two-way coupling between the thermal strain and temperature field is also studied, in the context of a change in thermal conductivity of the beryllium pebble bed in a solid breeder blanket TBM due to thermal strain. The temperature field determines the thermal strain in beryllium, which in turn changes the temperature field. Iterative thermo-fluid/thermal strain calculations have been applied to both steady-state and pulsed operation conditions. All calculations have been carried out in three dimensions with representative MCAD models, including all the TBM components in their entirety

Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

Energy Technology Data Exchange (ETDEWEB)

Ebert, D.

1997-07-01

This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.

Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

International Nuclear Information System (INIS)

Ebert, D.

1997-07-01

This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts' meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes

Further elucidation of nanofluid thermal conductivity measurement using a transient hot-wire method apparatus

Science.gov (United States)

Yoo, Donghoon; Lee, Joohyun; Lee, Byeongchan; Kwon, Suyong; Koo, Junemo

2018-02-01

The Transient Hot-Wire Method (THWM) was developed to measure the absolute thermal conductivity of gases, liquids, melts, and solids with low uncertainty. The majority of nanofluid researchers used THWM to measure the thermal conductivity of test fluids. Several reasons have been suggested for the discrepancies in these types of measurements, including nanofluid generation, nanofluid stability, and measurement challenges. The details of the transient hot-wire method such as the test cell size, the temperature coefficient of resistance (TCR) and the sampling number are further investigated to improve the accuracy and consistency of the measurements of different researchers. It was observed that smaller test apparatuses were better because they can delay the onset of natural convection. TCR values of a coated platinum wire were measured and statistically analyzed to reduce the uncertainty in thermal conductivity measurements. For validation, ethylene glycol (EG) and water thermal conductivity were measured and analyzed in the temperature range between 280 and 310 K. Furthermore, a detailed statistical analysis was conducted for such measurements, and the results confirmed the minimum number of samples required to achieve the desired resolution and precision of the measurements. It is further proposed that researchers fully report the information related to their measurements to validate the measurements and to avoid future inconsistent nanofluid data.

Investigation of Thermal Interface Materials Using Phase-Sensitive Transient Thermoreflectance Technique: Preprint

Energy Technology Data Exchange (ETDEWEB)

Feng, X.; King, C.; DeVoto, D.; Mihalic, M.; Narumanchi, S.

2014-08-01

With increasing power density in electronics packages/modules, thermal resistances at multiple interfaces are a bottleneck to efficient heat removal from the package. In this work, the performance of thermal interface materials such as grease, thermoplastic adhesives and diffusion-bonded interfaces are characterized using the phase-sensitive transient thermoreflectance technique. A multi-layer heat conduction model was constructed and theoretical solutions were derived to obtain the relation between phase lag and the thermal/physical properties. This technique enables simultaneous extraction of the contact resistance and bulk thermal conductivity of the TIMs. With the measurements, the bulk thermal conductivity of Dow TC-5022 thermal grease (70 to 75 um bondline thickness) was 3 to 5 W/(m-K) and the contact resistance was 5 to 10 mm2-K/W. For the Btech thermoplastic material (45 to 80 μm bondline thickness), the bulk thermal conductivity was 20 to 50 W/(m-K) and the contact resistance was 2 to 5 mm2-K/W. Measurements were also conducted to quantify the thermal performance of diffusion-bonded interface for power electronics applications. Results with the diffusion-bonded sample showed that the interfacial thermal resistance is more than one order of magnitude lower than those of traditional TIMs, suggesting potential pathways to efficient thermal management.

Apparatus and method for transient thermal infrared spectrometry of flowable enclosed materials

Science.gov (United States)

McClelland, John F.; Jones, Roger W.

1993-03-02

A method and apparatus for enabling analysis of a flowable material enclosed in a transport system having an infrared transparent wall portion. A temperature differential is transiently generated between a thin surface layer portion of the material and a lower or deeper portion of the material sufficient to alter the thermal infrared emission spectrum of the material from the black-body thermal infrared emission spectrum of the material, and the altered thermal infrared emission spectrum is detected through the infrared transparent portion of the transport system while the altered thermal infrared emission spectrum is sufficiently free of self-absorption by the material of emitted infrared radiation. The detection is effected prior to the temperature differential propagating into the lower or deeper portion of the material to an extent such that the altered thermal infrared emission spectrum is no longer sufficiently free of self-absorption by the material of emitted infrared radiation. By such detection, the detected altered thermal infrared emission spectrum is indicative of characteristics relating to molecular composition of the material.

Transient thermal-hydraulic characteristics analysis software for PWR nuclear power systems

International Nuclear Information System (INIS)

Wu Yingwei; Zhuang Chengjun; Su Guanghui; Qiu Suizheng

2010-01-01

A point reactor neutron kinetics model, a two-phase drift-flow U-tube steam generator model, an advanced non-equilibrium three regions pressurizer model, and a passive emergency core decay heat-removed system model are adopted in the paper to develop the computerized analysis code for PWR transient thermal-hydraulic characteristics, by Compaq Visual Fortran 6.0 language. Visual input, real-time processing and dynamic visualization output are achieved by Microsoft Visual Studio. NET language. The reliability verification of the soft has been conducted by RELAP 5, and the verification results show that the software is with high calculation precision, high calculation speed, modern interface, luxuriant functions and strong operability. The software was applied to calculate the transient accident conditions for QSNP, and the analysis results are significant to the practical engineering applications. (authors)

Transient behavior of ASTRID with a gas power conversion system

Energy Technology Data Exchange (ETDEWEB)

Bertrand, F., E-mail: frederic.bertrand@cea.fr; Mauger, G.; Bensalah, M.; Gauthé, P.

2016-11-15

Highlights: • CATHARE2 transient calculations have been performed for ASTRID with a gas PCS. • The behavior of the reactor is close for gas and for water PCS in case of LOOP. • The gas PCS enables to cool the core for at least 10 h for pressurized transients. • The depressurization of the PCS induces an over-cooling for breaches on low pressure pipes. • The spurious opening of a by-pass line of the turbomachine can be controlled without scram. - Abstract: The present article is dedicated to preliminary transient studies carried out for the analysis of the system overall behavior of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) demonstrator developed in France by CEA and its industrial partners. ASTRID is foreseen to demonstrate the progress made in SFR technology at an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability. Among the innovative options, a gas power conversion systems (PCS) is envisaged. In this innovative PCS, the working gas is nitrogen whose flow rate delivers power to a turbine driving with the same shaft two compressors (low and high pressure) separated by an intercooler. The other part of the work delivered by the gas is used to drive the alternator that produces electricity. The main objective of such a PCS consists in avoiding physically the possibility of a sodium/water reaction with the secondary circuit but the impact of this PCS on the control of incidental and accidental transients has also been studied. The main purpose of the studies presented in the paper is to assess the dynamic behavior of ASTRID including a gas PCS with the CATHARE2 code. The first transient presented deals with a loss of off-site power and has been calculated for the gas PCS but also for a classical steam/water PCS for comparison purpose. Then typical transients of gas system have been investigated. Several families of

Transient behavior of ASTRID with a gas power conversion system

International Nuclear Information System (INIS)

Bertrand, F.; Mauger, G.; Bensalah, M.; Gauthé, P.

2016-01-01

Highlights: • CATHARE2 transient calculations have been performed for ASTRID with a gas PCS. • The behavior of the reactor is close for gas and for water PCS in case of LOOP. • The gas PCS enables to cool the core for at least 10 h for pressurized transients. • The depressurization of the PCS induces an over-cooling for breaches on low pressure pipes. • The spurious opening of a by-pass line of the turbomachine can be controlled without scram. - Abstract: The present article is dedicated to preliminary transient studies carried out for the analysis of the system overall behavior of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) demonstrator developed in France by CEA and its industrial partners. ASTRID is foreseen to demonstrate the progress made in SFR technology at an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability. Among the innovative options, a gas power conversion systems (PCS) is envisaged. In this innovative PCS, the working gas is nitrogen whose flow rate delivers power to a turbine driving with the same shaft two compressors (low and high pressure) separated by an intercooler. The other part of the work delivered by the gas is used to drive the alternator that produces electricity. The main objective of such a PCS consists in avoiding physically the possibility of a sodium/water reaction with the secondary circuit but the impact of this PCS on the control of incidental and accidental transients has also been studied. The main purpose of the studies presented in the paper is to assess the dynamic behavior of ASTRID including a gas PCS with the CATHARE2 code. The first transient presented deals with a loss of off-site power and has been calculated for the gas PCS but also for a classical steam/water PCS for comparison purpose. Then typical transients of gas system have been investigated. Several families of

Transient insomnia versus chronic insomnia: a comparison study of sleep-related psychological/behavioral characteristics.

Science.gov (United States)

Yang, Chien-Ming; Lin, Shih-Chun; Cheng, Chung-Ping

2013-10-01

Vulnerability to transient insomnia is regarded as a predisposing factor for chronic insomnia. However, most individuals with transient insomnia do not develop chronic insomnia. The current study investigated the differential contributing factors for these two conditions to further the understanding of this phenomenon. Chronic insomnia patients and normal sleepers with high and low vulnerability to transient insomnia completed measures of pre-sleep arousal, dysfunctional sleep beliefs, and sleep-related safety behaviors. Both cognitive and somatic pre-sleep arousals were identified as significant predictors for transient insomnia. Dysfunctional beliefs regarding worry about insomnia and cognitive arousal were predictors for chronic insomnia. Sleep-related safety behavior, although correlated with insomnia severity, was not a significant predictor for both conditions. Dysfunctional beliefs associated with worry and losing control over sleep are the most critical factors in differentiating chronic insomnia from transient insomnia. These factors should be addressed to help prevent individuals with high sleep vulnerability from developing chronic sleep disturbance. © 2013 Wiley Periodicals, Inc.

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

Energy Technology Data Exchange (ETDEWEB)

Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-05-15

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

International Nuclear Information System (INIS)

Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

2017-01-01

Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

Measurement of the thermal conductivity of liquid D2O by the transient hot-wire method

International Nuclear Information System (INIS)

Nagasaka, Y.; Hiraiwa, H.; Nagashima, A.

1990-01-01

The measurement of the thermal conductivity of liquid D 2 O (heavy water) started in 1951. Since then, many researchers have measured the thermal conductivity of heavy water mainly with the aid of steady-state methods such as the parallel plate method and the concentric cylinder method. It should be noted here that even in the case of pure H 2 O or D 2 O enclosed in metallic vessel for a couple of days, the electrical conductivity seems to be not low enough for precise transient hot-wire measurements. The purpose of this paper is to obtain precise thermal conductivity data of liquid D 2 O which can be the reference standard values by the transient hot-wire method. The temperature range covered was 4 degrees C to 80 degrees C with pressure up to 40 MPa and the experimental data have an estimated accuracy of ±0.5%

Design of a bolted flange subjected to severe nuclear system thermal transients - A case study

International Nuclear Information System (INIS)

Palmer, W.J.; Tomawski, R.J.; Ezekoye, L.I.; Lacey, M.L.

1986-01-01

Flange design standards recognize that flanged joints may develop leakage should they be exposed to severe thermal gradients and recommend that such operating conditions be avoided. In nuclear power plants, severe thermal transients may be encountered in many plant and system operating and test conditions. In such applications, conformance with standard design practice may not ensure a leak-tight joint. This paper describes the proper consideration of thermal effects on flanged joints and how that can lead to the development of a successful leak-tight design. Similar procedures may be applied generally to evaluate and upgrade flanged joints in thermal shock applications

Transient characteristics of thermal energy storage in an enclosure packed with MEPCM particles

International Nuclear Information System (INIS)

Siao, Yong-Hao; Yan, Wei-Mon; Lai, Chi-Ming

2015-01-01

The heat transfer characteristics of phase change materials have been of continuing interest of research due to various potential technical applications, such as the latent-heat thermal energy storage, thermal protection, as well as active/passive electronic cooling. In this work, the transient characteristics of thermal energy storage in a partitioned enclosure filled with microencapsulated phase change material (MEPCM) particles were investigated experimentally and numerically. To examine the different melting temperature effects, two different MEPCM particles are tested. The core phase change materials of the MEPCM are n-octadecane with melting temperature about T M  = 28 °C and 37 °C. The enclosure is partitioned and is differentially heated by the two horizontal isothermal surfaces, while the other vertical surfaces are considered thermally insulated. The studies have been undertaken for five sets of the hot and cold wall temperatures imposed across the enclosure. The consequents show that the numerical results are in agreement with the measured data. At the initial transient, the net energy storage in enclosure, Q net , increases with the time Fo. Finally, the Q net approaches quickly the steady state for the case with a higher temperature difference of T h  − T c . Additionally, higher dimensionless accumulated energy through the hot wall Q h and cold wall Q c is found for a case with higher hot wall temperature T h

Peach Bottom transient analysis with BWR TRACB02

International Nuclear Information System (INIS)

Alamgir, M.; Sutherland, W.A.

1984-01-01

TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

Transient thermal stress analysis of a near-edge elliptical defect in a semi-infinite plate subjected to a moving heat source

International Nuclear Information System (INIS)

Mingjong Wang; Weichung Wang

1994-01-01

In this paper, the maximum transient thermal stresses on the boundary of a near-edge elliptical defect in a semi-infinite thin plate were determined by the digital photoelastic technique, when the plate edge experiences a moving heat source. The relationships between the maximum transient thermal stresses and the size and inclination of the elliptical defect, the minimum distance from the elliptical defect to the plate edge as well as the speed of the moving heat source were also studied. Finally, by using a statistical analysis package, the variations of the maximum transient thermal stresses were then correlated with the time, the minimum distance between the edge and the elliptical defect, temperature difference, and speed of the moving heat source. (author)

Reactor thermal behaviors under kinetics parameters variations in fast reactivity insertion

Energy Technology Data Exchange (ETDEWEB)

Abou-El-Maaty, Talal [Reactors Department, Atomic Energy Authority, Cairo 13759 (Egypt)], E-mail: talal22969@yahoo.com; Abdelhady, Amr [Reactors Department, Atomic Energy Authority, Cairo 13759 (Egypt)

2009-03-15

The influences of variations in some of the kinetics parameters affecting the reactivity insertion are considered in this study, it has been accomplished in order to acquire knowledge about the role that kinetic parameters play in prompt critical transients from the safety point of view. The kinetics parameters variations are limited to the effective delayed neutron fraction ({beta}{sub eff}) and the prompt neutron generation time ({lambda}). The reactor thermal behaviors under the variations in effective delayed neutron fraction and prompt neutron generation time included, the reactor power, maximum fuel temperature, maximum clad temperature, maximum coolant temperature and the mass flux variations at the hot channel. The analysis is done for a typical swimming pool, plate type research reactor with low enriched uranium. The scram system is disabled during the accidents simulations. Calculations were done using PARET code. As a result of simulations, it is concluded that, the reactor (ETRR2) thermal behavior is considerably more sensitive to the variation in the effective delayed neutron fraction than to the variation in prompt neutron generation time and the fast reactivity insertion in both cases causes a flow expansion and contraction at the hot channel exit. The amplitude of the oscillated flow is a qualitatively increases with the decrease in both {beta}{sub eff} and {lambda}.

Rapid thermal transient in a reactor coolant channel

International Nuclear Information System (INIS)

Cherubini, A.

1986-01-01

This report deals with the problem of one-dimensional thermo-fluid-dynamics in a reactor coolant channel, with the aim of determining the evolution in time of the coolant (H*L2O), in one-and/or two-phase regimes, subjected to a great and rapid increase in heat flux (accident conditions). To this aim, the following are set out: a) the physical model used; b) the equations inherent in the above model; c) the numerical methods employed to solve them by means of a computer programme called CABO (CAnale BOllente). Next a typical problem of rapid thermal transient resolved by CABO is reported. The results obtained, expressed in form of graphs, are fully discussed. Finally comments on possible developments of CABO follow

PWR fuel behavior: lessons learned from LOFT

International Nuclear Information System (INIS)

Russell, M.L.

1981-01-01

A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

International Nuclear Information System (INIS)

Boulaich, Y.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Chakir, E.; El Bakkari, B.; El Younoussi, C.

2015-01-01

Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad

Momentum integral network method for thermal-hydraulic transient analysis

International Nuclear Information System (INIS)

Van Tuyle, G.J.

1983-01-01

A new momentum integral network method has been developed, and tested in the MINET computer code. The method was developed in order to facilitate the transient analysis of complex fluid flow and heat transfer networks, such as those found in the balance of plant of power generating facilities. The method employed in the MINET code is a major extension of a momentum integral method reported by Meyer. Meyer integrated the momentum equation over several linked nodes, called a segment, and used a segment average pressure, evaluated from the pressures at both ends. Nodal mass and energy conservation determined nodal flows and enthalpies, accounting for fluid compression and thermal expansion

CW 316 mechanical properties during thermal transients

International Nuclear Information System (INIS)

Cauvin, R.; Boutard, J.L.; Allegraud, G.

1984-06-01

During in pile incidents, the cladding can experience higher temperatures than the nominal one; it is necessary to know the mechanical properties of the cladding material during such thermal transients to predict the time and location of rupture. Two types of tests have been developed: first tensile (constant strain rate) tests after a heating at a constant rate and secondly constant load tests where heating is performed until rupture occurs. The tensile tests clearly show the role of the heating rate: the higher is the heating rate, the lower is the cold work recovery. Constant load tests were conducted with either uniaxial or biaxial (burst tests) loading. The same stress/failure temperature relation is found in both types of loading using the Von Mises equivalent stress. To predict failure, the Larson Miller parameter is not adequate, as well as all parameters based on a time/temperature equivalence. The yield stress measured in the two types of tests are very different probably due to a strain rate effect. Indeed the tensile tests are dynamic ones to avoid thermal recovery during the test duration, while the strain rate measured in constant load tests ranges only from 10 -5 s -1 to 10 -3 s -1 , being an increasing function of heating rate (ranging from 1 0 c/s to 100 0 c/s)

Binary nucleation kinetics. III. Transient behavior and time lags

International Nuclear Information System (INIS)

Wyslouzil, B.E.; Wilemski, G.

1996-01-01

Transient binary nucleation is more complex than unary because of the bidimensionality of the cluster formation kinetics. To investigate this problem qualitatively and quantitatively, we numerically solved the birth-death equations for vapor-to-liquid phase transitions. Our previous work showed that the customary saddle point and growth path approximations are almost always valid in steady state gas phase nucleation and only fail if the nucleated solution phase is significantly nonideal. Now, we demonstrate that in its early transient stages, binary nucleation rarely, if ever, occurs via the saddle point. This affects not only the number of particles forming but their composition and may be important for nucleation in glasses and other condensed mixtures for which time scales are very long. Before reaching the state of saddle point nucleation, most binary systems pass through a temporary stage in which the region of maximum flux extends over a ridge on the free energy surface. When ridge crossing nucleation is the steady state solution, it thus arises quite naturally as an arrested intermediate state that normally occurs in the development of saddle point nucleation. While the time dependent and steady state distributions of the fluxes and concentrations for each binary system are strongly influenced by the gas composition and species impingement rates, the ratio of nonequilibrium to equilibrium concentrations has a quasiuniversal behavior that is determined primarily by the thermodynamic properties of the liquid mixture. To test our quantitive results of the transient behavior, we directly calculated the time lag for the saddle point flux and compared it with the available analytical predictions. Although the analytical results overestimate the time lag by factors of 1.2-5, they should be adequate for purposes of planning experiments. We also found that the behavior of the saddle point time lag can indicate when steady state ridge crossing nucleation will occur

TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

International Nuclear Information System (INIS)

Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman

1988-01-01

TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory

TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

Energy Technology Data Exchange (ETDEWEB)

Muzhang, Peng; Quan, Zhang; Guoli, Wang; Yuman, Zhang

1988-03-01

TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory.

Thermal-hydraulic behaviors of vapor-liquid interface due to arrival of a pressure wave

Energy Technology Data Exchange (ETDEWEB)

Inoue, Akira; Fujii, Yoshifumi; Matsuzaki, Mitsuo [Tokyo Institute of Technology (Japan)

1995-09-01

In the vapor explosion, a pressure wave (shock wave) plays a fundamental role for triggering, propagation and enhancement of the explosion. Energy of the explosion is related to the magnitude of heat transfer rate from hot liquid to cold volatile one. This is related to an increasing rate of interface area and to an amount of transient heat flux between the liquids. In this study, the characteristics of transient heat transfer and behaviors of vapor film both on the platinum tube and on the hot melt tin drop, under same boundary conditions have been investigated. It is considered that there exists a fundamental mechanism of the explosion in the initial expansion process of the hot liquid drop immediately after arrival of pressure wave. The growth rate of the vapor film is much faster on the hot liquid than that on the solid surface. Two kinds of roughness were observed, one due to the Taylor instability, by rapid growth of the explosion bubble, and another, nucleation sites were observed at the vapor-liquid interface. Based on detailed observation of early stage interface behaviors after arrival of a pressure wave, the thermal fragmentation mechanism is proposed.

Finite element study of a HDR-RPV-section including a nozzle under thermal shock transient

Energy Technology Data Exchange (ETDEWEB)

Roos, E [Stuttgart Univ. (Germany); Katzenmeier, G [Forschungszentrum Juelich GmbH (Germany); Wanner, R; Mercier, O [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1988-12-31

This document presents a finite element study of a reactor pressure vessel section under thermal stresses. The strength properties of the vessel walls are studied as well as cracks due to the thermo-shock transient. (TEC). 6 refs.

Transient thermal stresses due to a zonal heat source moving back and forth over the surface on an infinite plate

International Nuclear Information System (INIS)

Sumi, N.; Hetnarski, R.B.

1989-01-01

A solution is given for the transient thermal stresses due to a zonal heat source moving back and forth with a constant angular frequency over the surface of an infinite elastic plate. The transient temperature distribution is obtained by using the complex Fourier and Laplace transforms, and the associated thermal stresses are obtained by means of the thermoelastic displacement potential and the Galerkin function. Graphical representations for the solution in dimensionless terms are included in this paper. (orig.)

Development of neutronics and thermal hydraulics coupled code – SAC-RIT for plate type fuel and its application to reactivity initiated transient analysis

International Nuclear Information System (INIS)

Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.

2013-01-01

Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code

Study on transient hydrogen behavior and effect on passive containment cooling system of the advanced PWR

International Nuclear Information System (INIS)

Wang Yan

2014-01-01

A certain amount of hydrogen will be generated due to zirconium-steam reaction or molten corium concrete interaction during severe accidents in the pressurized water reactor (PWR). The generated hydrogen releases into the containment, and the formed flammable mixture might cause deflagration or detonation to produce high thermal and pressure loads on the containment, which may threaten the integrity of the containment. The non-condensable hydrogen in containment may also reduce the steam condensation on the containment surface to affect the performance of the passive containment cooling system (PCCS). To study the transient hydrogen behavior in containment with the PCCS performance during the accidents is significant for the further study on the PCCS design and the hydrogen risk mitigation. In this paper, a new developed PCCS analysis code with self-reliance intellectual property rights, which had been validated by comparison on the transients in the containment during the design basis accidents with other developed PCCS analysis code, is brief introduced and used for the transient simulation in the containment under a postulated small break LOCA of cold-leg. The results show that the hydrogen will flow upwards with the coolant released from the break and spread in the containment by convection and diffusion, and it results in the increase of the pressure in the containment due to reducing the heat removal capacity of the PCCS. (author)

THEBES: a thermal hydraulic code for the calculation of transient two phase flow in bundle geometry

International Nuclear Information System (INIS)

Camous, F.

1983-01-01

The three dimensional thermal hydraulic code THEBES, capable to calculate transient boiling of sodium in rod bundles is described here. THEBES, derived from the transient single phase code SABRE-2A, was developed in CADARACHE by the SIES to analyse the SCARABEE N loss of flow experiments. This paper also presents the results of tests which were performed against various types of experiments: (1) transient boiling in a 7 pin bundle simulating a partial blockage at the bottom of a subassembly (rapid transient SCARABEE 7.2 experiment), (2) transient boiling in a 7 pin bundle simulating a coolant coast down (slow transient SCARABEE 7.3 experiment), (3) steady local and generalised boiling in a 19 pin bundle (GR 19 I experiment), (4) transient boiling in a 19 pin bundle simulating a coolant coast down (GR 19 I experiment), (5) steady local boiling in a 37 pin bundle with internal blockage (MOL 7C experiment). Excellent agreement was found between calculated and experimental results for these different situations. Our conclusion is that THEBES is able to calculate transient boiling of sodium in rod bundles in a quite satisfying way

Thermal ramp rate effects on mixed-oxide fuel swelling/gas release

International Nuclear Information System (INIS)

Hinman, C.A.; Randklev, E.H.

1979-01-01

Macroscopic swelling behavior of PNL-10 was compared to that of PNL-2 fuel and it was found that the swelling-threshold behavior is similar for similar thermal conditions. Transient fission gas release for the PNL-10 fuel is very similar to that observed for the PNL-2 fuel for similar thermal conditions

Damage behavior of REE-doped W-based material exposed to high-flux transient heat loads

International Nuclear Information System (INIS)

Shi, Jing; Luo, Lai–Ma; Lin, Jin–shan; Zan, Xiang; Zhu, Xiao–yong; Xu, Qiu; Wu, Yu–Cheng

2016-01-01

Pure W and W-Lu alloys were prepared by mechanical alloying (MA) and spark plasma sintering (SPS) technology. The performance and relevant damage mechanism of W-(0%, 2%, 5%, 10%) Lu alloys under transient heat loads were investigated using a laser beam heat load test to simulate the transient events in future nuclear fusion reactors. Scanning electron microscopy was used to observe the morphologies of the damaged surfaces and energy dispersive X-ray spectroscopy was used to conduct composition analysis. Damages to the surface such as cracks, pits, melting layers, Lu-rich droplets, and thermal ablation were observed. A mass of dense fuzz-like nanoparticles formed on the outer region of the laser-exposed area. Recrystallization, grain growth, increased surface roughness, and material erosion were also observed. W-Lu samples with low Lu content demonstrated better thermal performance than pure W, and the degree of damage significantly deteriorated under repetitive transient heat loads.

Computational model for transient studies of IRIS pressurizer behavior

International Nuclear Information System (INIS)

Rives Sanz, R.; Montesino Otero, M.E.; Gonzalez Mantecon, J.; Rojas Mazaira, L.

2014-01-01

International Reactor Innovative and Secure (IRIS) excels other Small Modular Reactor (SMR) designs due to its innovative characteristics regarding safety. IRIS integral pressurizer makes the design of larger pressurizer system than the conventional PWR, without any additional cost. The IRIS pressurizer volume of steam can provide enough margins to avoid spray requirement to mitigate in-surge transient. The aim of the present research is to model the IRIS pressurizer's dynamic using the commercial finite volume Computational Fluid Dynamic code CFX 14. A symmetric tridimensional model equivalent to 1/8 of the total geometry was adopted to reduce mesh size and minimize processing time. The model considers the coexistence of three phases: liquid, steam, and vapor bubbles in liquid volume. Additionally, it takes into account the heat losses between the pressurizer and primary circuit. The relationships for interfacial mass, energy, and momentum transport are programmed and incorporated into CFX by using expressions in CFX Command Language (CCL) format. Moreover, several additional variables are defined for improving the convergence and allow monitoring of boron dilution sequences and condensation-evaporation rate in different control volumes. For transient states a non - equilibrium stratification in the pressurizer is considered. This paper discusses the model developed and the behavior of the system for representative transients sequences such as the in/out-surge transients and boron dilution sequences. The results of analyzed transients of IRIS can be applied to the design of pressurizer internal structures and components. (author)

Mechanical behavior of fast reactor fuel pin cladding subjected to simulated overpower transients

International Nuclear Information System (INIS)

Johnson, G.D.; Hunter, C.W.

1978-06-01

Cladding mechanical property data for analysis and prediction of fuel pin transient behavior were obtained under experimental conditions in which the temperature ramps of reactor transients were simulated. All cladding specimens were 20% CW Type 316 stainless steel and were cut from EBR-II irradiated fuel pins. It was determined that irradiation degraded the cladding ductility and failure strength. Specimens that had been adjacent to the fuel exhibited the poorest properties. Correlations were developed to describe the effect of neutron fluence on the mechanical behavior of the cladding. Metallographic examinations were conducted to characterize the failure mode and to establish the nature of internal and external surface corrosion. Various mechanisms for the fuel adjacency effect were examined and results for helium concentration profiles were presented. Results from the simulated transient tests were compared with TREAT test results

Thermal conductivity of wood ash diatomite composites using the transient hot strip method

International Nuclear Information System (INIS)

Muia, L.M.; Gaitho, F.

2003-08-01

The transient Hot Strip method (THS) was used to determine the thermal conductivities of pure Wood Ash (WA), two kinds of diatomite i.e., DB and DF, and their composites. The effects of grain size and temperature on the thermal conductivities of the three systems and their composites were also determined. The lowest thermal conductivities of 0.02x10 -2 Wm -1 K -1 for wood ash and ∼ 3x10 -2 Wm -1 K -1 for the diatomites are found in the particle size range 60 -80μm. The thermal conductivities of the various composites range between 1.3x10 -3 and 6.8x10 -2 Wm -1 K -1 . These values are a factor of 10 lower than those of the pure materials. The thermal conductivity of the three composites is independent of temperature in the range 26-350 deg. C, in contrast to those pure materials which increase with temperature. Generally, the thermal conductivites of the pure materials which increase as their porosity or moisture contents are increased. (author)

Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

International Nuclear Information System (INIS)

Henderson, J.M.; Wood, S.A.; Knight, D.D.

1981-01-01

The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting

Lumped-Element Dynamic Electro-Thermal model of a superconducting magnet

Science.gov (United States)

Ravaioli, E.; Auchmann, B.; Maciejewski, M.; ten Kate, H. H. J.; Verweij, A. P.

2016-12-01

Modeling accurately electro-thermal transients occurring in a superconducting magnet is challenging. The behavior of the magnet is the result of complex phenomena occurring in distinct physical domains (electrical, magnetic and thermal) at very different spatial and time scales. Combined multi-domain effects significantly affect the dynamic behavior of the system and are to be taken into account in a coherent and consistent model. A new methodology for developing a Lumped-Element Dynamic Electro-Thermal (LEDET) model of a superconducting magnet is presented. This model includes non-linear dynamic effects such as the dependence of the magnet's differential self-inductance on the presence of inter-filament and inter-strand coupling currents in the conductor. These effects are usually not taken into account because superconducting magnets are primarily operated in stationary conditions. However, they often have significant impact on magnet performance, particularly when the magnet is subject to high ramp rates. Following the LEDET method, the complex interdependence between the electro-magnetic and thermal domains can be modeled with three sub-networks of lumped-elements, reproducing the electrical transient in the main magnet circuit, the thermal transient in the coil cross-section, and the electro-magnetic transient of the inter-filament and inter-strand coupling currents in the superconductor. The same simulation environment can simultaneously model macroscopic electrical transients and phenomena at the level of superconducting strands. The model developed is a very useful tool for reproducing and predicting the performance of conventional quench protection systems based on energy extraction and quench heaters, and of the innovative CLIQ protection system as well.

Transient atomic behavior and surface kinetics of GaN

International Nuclear Information System (INIS)

Moseley, Michael; Billingsley, Daniel; Henderson, Walter; Trybus, Elaissa; Doolittle, W. Alan

2009-01-01

An in-depth model for the transient behavior of metal atoms adsorbed on the surface of GaN is developed. This model is developed by qualitatively analyzing transient reflection high energy electron diffraction (RHEED) signals, which were recorded for a variety of growth conditions of GaN grown by molecular-beam epitaxy (MBE) using metal-modulated epitaxy (MME). Details such as the initial desorption of a nitrogen adlayer and the formation of the Ga monolayer, bilayer, and droplets are monitored using RHEED and related to Ga flux and shutter cycles. The suggested model increases the understanding of the surface kinetics of GaN, provides an indirect method of monitoring the kinetic evolution of these surfaces, and introduces a novel method of in situ growth rate determination.

Transient atomic behavior and surface kinetics of GaN

Science.gov (United States)

Moseley, Michael; Billingsley, Daniel; Henderson, Walter; Trybus, Elaissa; Doolittle, W. Alan

2009-07-01

An in-depth model for the transient behavior of metal atoms adsorbed on the surface of GaN is developed. This model is developed by qualitatively analyzing transient reflection high energy electron diffraction (RHEED) signals, which were recorded for a variety of growth conditions of GaN grown by molecular-beam epitaxy (MBE) using metal-modulated epitaxy (MME). Details such as the initial desorption of a nitrogen adlayer and the formation of the Ga monolayer, bilayer, and droplets are monitored using RHEED and related to Ga flux and shutter cycles. The suggested model increases the understanding of the surface kinetics of GaN, provides an indirect method of monitoring the kinetic evolution of these surfaces, and introduces a novel method of in situ growth rate determination.

Oxidation and thermal behavior of Jatropha curcas biodiesel ...

African Journals Online (AJOL)

The thermal and oxidation behavior is also affected adversely by the container metal. The present paper is dealing with the study of oxidation and thermal behavior of JCB with respect to different metal contents. It was found that influence of metal was detrimental to thermal and oxidation stability. Even small concentrations ...

Lifetime improvement of sheathed thermocouples for use in high-temperature and thermal transient operations

International Nuclear Information System (INIS)

McCulloch, R.W.; Clift, J.H.

1982-01-01

Premature failure of small-diameter, magnesium-oxide-insulated sheathed thermocouples occurred when they were placed within nuclear fuel rod simulators (FRSs) to measure high temperatures and to follow severe thermal transients encountered during simulation of nuclear reactor accidents in Oak Ridge National Laboratory (ORNL) thermal-hydraulic test facilities. Investigation of thermally cycled thermocouples yielded three criteria for improvement of thermocouple lifetime: (1) reduction of oxygen impurities prior to and during their fabrication, (2) refinement of thermoelement grain size during their fabrication, and (3) elimination of prestrain prior to use above their recrystallization temperature. The first and third criteria were satisfied by improved techniques of thermocouple assembly and by a recovery anneal prior to thermocouple use

Thermal analysis of a one-element PWR spent fuel shipping cask

International Nuclear Information System (INIS)

Fields, S.R.

1979-06-01

The transient thermal behavior of a typical one-element PWR spent fuel shipping cask, following a hypothetical accident and fire, has been simulated. The objectives of the study were to determine the transient behavior of the cask and its spent fuel primary coolant through the pressure relief system and possible fuel pin clad failure due to overheating following loss of coolant. 15 figures, 7 tables

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)

Modeling transient thermal hydraulic behavior of a thermionic fuel element for nuclear space reactors

International Nuclear Information System (INIS)

Al-Kheliewi, A.S.; Klein, A.C.

1994-01-01

A transient code (TFETC) for determining the temperature distribution throughout the radial and axial positions of a thermionic fuel element (TFE) during changes in operating conditions has been successfully developed and tested. A fully implicit method is used to solve the system of equations for temperatures at each time step. Presently, TFETC has the ability to handle the following transients: startup, loss of flow accidents, and shutdown. The code has been applied to the startup of the ATI single cell configuration which appears to start up and shut down in an orderly and reasonable fashion. No unexpected transient features were observed. The TFE also appears to function robustly under loss of flow accident conditions. It appears hat sufficient time is available to shut the reactor down safely without melting point the fuel. The model shows that during a complete loss of flow accident (without shutdown) the coolant reaches its boiling point in approximately 35 seconds. The fuel may exceed its melting point after this time as the NaK coolant will boil if the reactor is not shut down. For 1/2, 1/3, and 1/4 pump failures, the fuel temperatures never exceed the fuel melting point even if the reactor is not shut down

Importance of thermal nonequilibrium considerations for the simulation of nuclear reactor LOCA transients

International Nuclear Information System (INIS)

Fischer, S.R.; Nelson, R.A.; Sullivan, L.H.

1980-01-01

The purpose of this paper is to show the importance of considering thermal nonequilibrium effects in computer simulations of the refill and reflood portions of pressurized water reactor (PWR) loss-of-coolnat accident (LOCA) transients. Although RELAP4 assumes thermodynamic equilibrium between phases, models that account for the nonequilibrium phenomena associated with the mixing of subcooled emergency cooling water with steam and the superheating of vapor in the presence of liquid droplets have recently been incorporated into the code. Code calculated results, both with and without these new models, have been compared with experimental test data to assess the importance of including thermal nonequilibrium phenomena in computer code simulations

Analysis of transient permeation behavior of hydrogen isotope caused by abrupt temperature change of first wall and blanket wall material

International Nuclear Information System (INIS)

Yamawaki, Michio; Tanaka, Satoru; Kiyoshi, Tsukasa

1989-01-01

To obtain further information on the transient permeation behavior of hydrogen isotopes as caused by an abrupt temperature change, numerical calculations were carried out for two typical metals, nickel and vanadium. Deuterium permeation through nickel is analyzed as a typical case of bulk-diffusion-limited permeation. Its transient behavior changed dramatically according to the specimen thickness. The transient behavior, in general, is separated into two parts, initial and latter period behaviors. Conditions which cause such a separation were evaluated. Evaluation of the hydrogen diffusivity and solubility by an analysis of transient curves of hydrogen permeation was carried out. The transient behavior of simultaneous gas- and ion-driven hydrogen permeation through vanadium was also analyzed. Overshooting of the hydrogen permeation rate appears with an abrupt temperature increase. Increasing the impinging ion flux causes the overshooting peak to become sharper, and also reduces the change of the steady-state permeation rate to be attained after the temperature change compared with the initial value. (orig.)

Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

International Nuclear Information System (INIS)

Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.

1991-09-01

The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

Transient Thermal Analysis of 3-D Integrated Circuits Packages by the DGTD Method

KAUST Repository

Li, Ping

2017-03-11

Since accurate thermal analysis plays a critical role in the thermal design and management of the 3-D system-level integration, in this paper, a discontinuous Galerkin time-domain (DGTD) algorithm is proposed to achieve this purpose. Such as the parabolic partial differential equation (PDE), the transient thermal equation cannot be directly solved by the DGTD method. To address this issue, the heat flux, as an auxiliary variable, is introduced to reduce the Laplace operator to a divergence operator. The resulting PDE is hyperbolic, which can be further written into a conservative form. By properly choosing the definition of the numerical flux used for the information exchange between neighboring elements, the hyperbolic thermal PDE can be solved by the DGTD together with the auxiliary differential equation. The proposed algorithm is a kind of element-level domain decomposition method, which is suitable to deal with multiscale geometries in 3-D integrated systems. To verify the accuracy and robustness of the developed DGTD algorithm, several representative examples are benchmarked.

Transient Thermal Response of Lightweight Cementitious Composites Made with Polyurethane Foam Waste

Science.gov (United States)

Kismi, M.; Poullain, P.; Mounanga, P.

2012-07-01

The development of low-cost lightweight aggregate (LWA) mortars and concretes presents many advantages, especially in terms of lightness and thermal insulation performances of structures. Low-cost LWA mainly comes from the recovery of vegetal or plastic wastes. This article focuses on the characterization of the thermal conductivity of innovative lightweight cementitious composites made with fine particles of rigid polyurethane (PU) foam waste. Five mortars were prepared with various mass substitution rates of cement with PU-foam particles. Their thermal conductivity was measured with two transient methods: the heating-film method and the hot-disk method. The incorporation of PU-foam particles causes a reduction of up to 18 % of the mortar density, accompanied by a significant improvement of the thermal insulating performance. The effect of segregation on the thermal properties of LWA mortars due to the differences of density among the cementitious matrix, sand, and LWA has also been quantified. The application of the hot-disk method reveals a gradient of thermal conductivity along the thickness of the specimens, which could be explained by a non-uniform repartition of fine PU-foam particles and mineral aggregates within the mortars. The results show a spatial variation of the thermal conductivity of the LWA mortars, ranging from 9 % to 19 %. However, this variation remains close to or even lower than that observed on a normal weight aggregate mortar. Finally, a self-consistent approach is proposed to estimate the thermal conductivity of PU-foam cement-based composites.

Effect of microscale gaseous thermal conduction on the thermal behavior of a buckled microbridge

International Nuclear Information System (INIS)

Wang Jiaqi; Tang Zhenan; Li Jinfeng; Zhang Fengtian

2008-01-01

A microbridge is a basic micro-electro-mechanical systems (MEMS) device and has great potential for application in microsensors and microactuators. The thermal behavior of a microbridge is important for designing a microbridge-based thermal microsensor or microactuator. To study the thermal behavior of a microbridge consisting of Si 3 N 4 and polysilicon with a 2 µm suspended gap between the substrate and the microbridge while the microbridge is heated by an electrical current fed through the polysilicon, a microbridge model is developed to correlate theoretically the input current and the temperature distribution under the buckling conditions, especially considering the effects of the microscale gaseous thermal conduction due to the microbridge buckling. The calculated results show that the buckling of the microbridge changes the microscale gaseous thermal conduction, and thus greatly affects the thermal behavior of the microbridge. We also evaluate the effects of initial buckling on the temperature distribution of the microbridge. The experimental results show that buckling should be taken into account if the buckling is large. Therefore, the variation in gaseous thermal conduction and the suspended gap height caused by the buckling should be considered in the design of such thermomechanical microsensors and microactuators, which requires more accurate thermal behavior

Transient redistribution of intragranular fission gas in irradiated mixed oxide

International Nuclear Information System (INIS)

Hinman, C.A.; Randklev, E.H.

1981-01-01

Safety analyses for an LMFBR require a knowledge of the fuel and fission gas behavior under transient conditions. Analyses of microstructural data derived from transiently heated, irradiated, mixed oxide fuel specimens have allowed the calculation of the degree of nonequilibrium of intragranular bubbles formed during the transient. It is hypothesized that the observed over-pressurization of the intragranular bubbles mechanically loads the fuel within the grain, leading to a stress gradient derived force upon near-grain-surface bubbles, driving them preferentially to the grain boundaries. Using existing models for forced diffusion it can be estimated that the stress derived forces on bubbles are within the same magnitude, and possibly greater, than the forces derived from the thermal gradient

Thermal Reactor Safety

Energy Technology Data Exchange (ETDEWEB)

1980-06-01

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

Thermal Reactor Safety

International Nuclear Information System (INIS)

1980-06-01

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

Thermal performance of a PCM thermal storage unit

Energy Technology Data Exchange (ETDEWEB)

Liu Ming; Bruno, Frank; Saman, Wasim [Sustainable Energy Centre, Inst. for Sustainable Systems and Technologies, Univ. of South Australia, Mawson Lakes, Adelaide (Australia)

2008-07-01

The thermal performance of a PCM thermal storage unit (TSU) is studied numerically and experimentally. The TSU under analysis consists of several flat slabs of phase change material (PCM) with melting temperature of -26.7 C. Liquid heat transfer fluid (HTF) passes between the slabs to charge and discharge the storage unit. A one dimensional mathematical model was employed to analyze the transient thermal behavior of the storage unit during the melting and freezing processes. The model takes into consideration the temperature variations in the wall along the flow direction of the HTF. The paper compares the experimental and numerical simulation results in terms of HTF outlet temperatures during the melting period. (orig.)

Lumped-parameter fuel rod model for rapid thermal transients

International Nuclear Information System (INIS)

Perkins, K.R.; Ramshaw, J.D.

1975-07-01

The thermal behavior of fuel rods during simulated accident conditions is extremely sensitive to the heat transfer coefficient which is, in turn, very sensitive to the cladding surface temperature and the fluid conditions. The development of a semianalytical, lumped-parameter fuel rod model which is intended to provide accurate calculations, in a minimum amount of computer time, of the thermal response of fuel rods during a simulated loss-of-coolant accident is described. The results show good agreement with calculations from a comprehensive fuel-rod code (FRAP-T) currently in use at Aerojet Nuclear Company

Numerical Model and Experimental Analysis of the Thermal Behavior of Electric Radiant Heating Panels

Directory of Open Access Journals (Sweden)

Giovanni Ferrarini

2018-01-01

Full Text Available Electric radiant heating panels are frequently selected during the design phase of residential and industrial heating systems, especially for retrofit of existing buildings, as an alternative to other common heating systems, such as radiators or air conditioners. The possibility of saving living and working space and the ease of installation are the main advantages of electric radiant solutions. This paper investigates the thermal performance of a typical electric radiant panel. A climatic room was equipped with temperature sensors and heat flow meters to perform a steady state experimental analysis. For the dynamic behavior, a mathematical model was created and compared to a thermographic measurement procedure. The results showed for the steady state an efficiency of energy transformation close to one, while in a transient thermal regime the time constant to reach the steady state condition was slightly faster than the typical ones of hydronic systems.

Transient plane source (tps) sensors for simultaneous measurements of thermal conductivity and thermal diffusivity of insulators, fluids and conductors

Science.gov (United States)

Maqsood, Asghari; Anis-ur-Rehman, M.

2013-12-01

Thermal conductivity and thermal diffusivity are two important physical properties for designing any food engineering processes1. The knowledge of thermal properties of the elements, compounds and different materials in many industrial applications is a requirement for their final functionality. Transient plane source (tps) sensors are reported2 to be useful for the simultaneous measurement of thermal conductivity, thermal diffusivity and volumetric heat capacity of insulators, conductor liquids3 and high-TC superconductors4. The tps-sensor consists of a resistive element in the shape of double spiral made of 10 micrometer thick Ni-foils covered on both sides with 25 micrometer thick Kapton. This sensor acts both as a heat source and a resistance thermometer for recording the time dependent temperature increase. From the knowledge of the temperature co-efficient of the metal spiral, the temperature increase of the sensor can be determined precisely by placing the sensor in between two surfaces of the same material under test. This temperature increase is then related to the thermal conductivity, thermal diffusivity and volumetric heat capacity by simple relations2,5. The tps-sensor has been used to measure thermal conductivities from 0.001 Wm-1K-1to 600 Wm-1K-1 and temperature ranges covered from 77K- 1000K. This talk gives the design, advantages and limitations of the tpl-sensor along with its applications to the measurementof thermal properties in a variety of materials.

Transient plane source (tps) sensors for simultaneous measurements of thermal conductivity and thermal diffusivity of insulators, fluids and conductors

International Nuclear Information System (INIS)

Maqsood, Asghari; Anis-ur-Rehman, M

2013-01-01

Thermal conductivity and thermal diffusivity are two important physical properties for designing any food engineering processes 1 . The knowledge of thermal properties of the elements, compounds and different materials in many industrial applications is a requirement for their final functionality. Transient plane source (tps) sensors are reported 2 to be useful for the simultaneous measurement of thermal conductivity, thermal diffusivity and volumetric heat capacity of insulators, conductor liquids 3 and high-T C superconductors 4 . The tps-sensor consists of a resistive element in the shape of double spiral made of 10 micrometer thick Ni-foils covered on both sides with 25 micrometer thick Kapton. This sensor acts both as a heat source and a resistance thermometer for recording the time dependent temperature increase. From the knowledge of the temperature co-efficient of the metal spiral, the temperature increase of the sensor can be determined precisely by placing the sensor in between two surfaces of the same material under test. This temperature increase is then related to the thermal conductivity, thermal diffusivity and volumetric heat capacity by simple relations 2,5 . The tps-sensor has been used to measure thermal conductivities from 0.001 Wm −1 K −1 to 600 Wm −1 K −1 and temperature ranges covered from 77K– 1000K. This talk gives the design, advantages and limitations of the tpl-sensor along with its applications to the measurementof thermal properties in a variety of materials

Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

2015-05-15

Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

LOFT transient thermal analysis for 10 inch primary coolant blowdown piping weld

International Nuclear Information System (INIS)

Howell, S.K.

1978-01-01

A flaw in a weld in the 10 inch primary coolant blowdown piping was discovered by LOFT personnel. As a result of this, a thermal analysis and fracture mechanics analysis was requested by LOFT personnel. The weld and pipe section were analyzed for a complete thermal cycle, heatup and Loss of Coolant Experiment (LOCE), using COUPLE/MOD2, a two-dimensional finite element heat conduction code. The finite element representation used in this analysis was generated by the Applied Mechanics Branch. The record of nodal temperatures for the entire transient was written on tape VSN=T9N054, and has been forwarded to the Applied Mechanics Branch for use in their mechanical analysis. Specific details and assumptions used in this analysis are found in appropriate sections of this report

Overcooling transient selection and thermal hydraulic analyses of the Loviisa PTS assessments

Energy Technology Data Exchange (ETDEWEB)

Tuomisto, H [IVO Power Engineering Ltd, Vantaa (Finland)

1997-09-01

This paper describes transients selection and thermal hydraulic analyses of various PTS assessment studies performed for the pressure vessels of the Loviisa WWER-reactors. Deterministic analyses have been performed in various stages of the PTS studies and they have always made the formal basis for design and licensing of the reactor pressure vessel. The integrated, probabilistic PTS study was carried out to give an overview of the severity of all different PTS sequences, and give a quantitative estimate of the importance of the PTS issues in relation to the overall safety of the plant. Later, the sequences including external flooding of the pressure vessels were added to the PTS assessment. Thermal recovery annealing of the Loviisa 1 reactor pressure vessel took place during refuelling outage in 1996. (author). 10 refs, 4 figs, 3 tabs.

Thermal Analysis of Cold Vacuum Drying (CVD) of Spent Nuclear Fuel (SNF)

International Nuclear Information System (INIS)

PIEPHO, M.G.

2000-01-01

The thermal analysis examined transient thermal and chemical behavior of the Multi-Canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with N Reactor spent fuel. This analysis provides the basis for the MCO thermal behavior at the CVD Facility in support of the safety basis documentation

Thermal fluid mixing behavior during medium break LOCA in evaluation of pressurized thermal shock

International Nuclear Information System (INIS)

Jung, Jae Won; Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung

1998-01-01

Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. the applicability of RELAP5 code to analyze the thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of thermal stratification is investigated using Theofanous's empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing

Transient analysis for resolving safety issues

International Nuclear Information System (INIS)

Chao, J.; Layman, W.

1987-01-01

The Nuclear Safety Analysis Center (NSAC) has a Generic Safety Analysis Program to help resolve high priority generic safety issues. This paper describes several high priority safety issues considered at NSAC and how they were resolved by transient analysis using thermal hydraulics and neutronics codes. These issues are pressurized thermal shock (PTS), anticipated transients without scram (ATWS), steam generator tube rupture (SGTR), and reactivity transients in light of the Chernobyl accident

Thermal behavior in the transition region between nucleate and film boiling

International Nuclear Information System (INIS)

Adiutori, E.F.

1991-01-01

The prediction of post Critical Heat Flux (CHF) behavior is complicated by the highly nonlinear thermal behavior of boiling interfaces--ie by the nonlinear nature of the boiling curve. Nonlinearity in the boiling curve can and does cause thermal instability, resulting in temperature discontinuities. Thus the prediction of post CHF behavior requires the analysis of thermal stability. This in turn requires an accurate description of thermal behavior in transition boiling. This paper determines thermal behavior in transition boiling by analysis of literature data. It also describes design features which improve post CHF performance and are reported in the literature

Fast reactor fuel pin behavior analyses in a LOF type transient event

International Nuclear Information System (INIS)

Mizuno, Tomoyasu; Koyama, Shin-ichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

2013-06-01

In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, fuel pin behavior calculations were made using the fast reactor fuel pin performance code CEDAR. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross and Stoute type gap conductance model and constant gap conductance model, which is used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross and Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross and Stoute type gap conductance model which is thought to be realistic. (author)

Power transients of Ghana research reactor-1 using PARET/ANL thermal hydraulic code

International Nuclear Information System (INIS)

Ampomah-Amoaka, E.; Akaho, E.H.K.; Anim-Sampong, S.; Nyarko, B.J.B.

2010-01-01

PARET/ANL(Version 7.3 of 2007) thermal-hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1.The reactivities inserted were 2.1mk and 4mk.The peak power of 5.81kW was obtained for 2.1 mk insertion whereas the peak power for 4mk insertion of reactivity was 92.32kW.These results compare closely with experiments and theoretical studies conducted previously.

Thermal analysis of cold vacuum drying of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Piepho, M.G.

1998-07-20

The thermal analysis examined transient thermal and chemical behavior of the Multi canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with Mark IV N, Reactor spent fuel in four fuel baskets and one scrap basket. This analysis provides the basis for the MCO thermal behavior at the CVD Facility for its Phase 2 Safety Analysis Report (revision 4).

Casting thermal simulation

International Nuclear Information System (INIS)

Shamsuddin bin Sulaiman

1994-01-01

The whole of this study is concerned with process simulation in casting processes. This study describes the application of the finite element method as an aid to simulating the thermal design of a high pressure die casting die by analysing the cooling transients in the casting cycle. Two types of investigation were carried out to model the linear and non-linear cooling behavior with consideration of a thermal interface effect. The simulated cooling for different stages were presented in temperature contour form. These illustrate the successful application of the Finite Element Method to model the process and they illustrate the significance of the thermal interface at low pressure

Mapping Thermal Habitat of Ectotherms Based on Behavioral Thermoregulation in a Controlled Thermal Environment

Science.gov (United States)

Fei, T.; Skidmore, A.; Liu, Y.

2012-07-01

Thermal environment is especially important to ectotherm because a lot of physiological functions rely on the body temperature such as thermoregulation. The so-called behavioural thermoregulation function made use of the heterogeneity of the thermal properties within an individual's habitat to sustain the animal's physiological processes. This function links the spatial utilization and distribution of individual ectotherm with the thermal properties of habitat (thermal habitat). In this study we modelled the relationship between the two by a spatial explicit model that simulates the movements of a lizard in a controlled environment. The model incorporates a lizard's transient body temperatures with a cellular automaton algorithm as a way to link the physiology knowledge of the animal with the spatial utilization of its microhabitat. On a larger spatial scale, 'thermal roughness' of the habitat was defined and used to predict the habitat occupancy of the target species. The results showed the habitat occupancy can be modelled by the cellular automaton based algorithm at a smaller scale, and can be modelled by the thermal roughness index at a larger scale.

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

International Nuclear Information System (INIS)

Reitsma, Frederik

2007-01-01

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

Thermal fluid mixing behavior during medium break LOCA in evaluation of pressurized thermal shock

Energy Technology Data Exchange (ETDEWEB)

Jung, Jae Won; Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

1998-12-31

Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze the thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of thermal stratification is investigated using Theofanous`s empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing. 6 refs., 8 figs. (Author)

RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

International Nuclear Information System (INIS)

1976-06-01

A discussion is presented of the use of the RELAP4/MOD5 computer program in simulating the thermal-hydraulic behavior of light-water reactor systems when subjected to postulated transients such as a LOCA, pump failure, or nuclear excursion. The volume is divided into main sections which cover: (1) program description, (2) input data, (3) problem initialization, (4) user guidelines, (5) output discussion, (6) source program description, (7) implementation requirements, (8) data files, (9) description of PLOTR4M, (10) description of STH20, (11) summary flowchart, (12) sample problems, (13) problem definition, and (14) problem input

Transient behavior of a flare-associated solar wind. I - Gas dynamics in a radial open field region

Science.gov (United States)

Nagai, F.

1984-01-01

A numerical investigation is conducted into the way in which a solar wind model initially satisfying both steady state and energy balance conditions is disturbed and deformed, under the assumption of heating that correspoonds to the energy release of solar flares of an importance value of approximately 1 which occur in radial open field regions. Flare-associated solar wind transient behavior is modeled for 1-8 solar radii. The coronal temperature around the heat source region rises, and a large thermal conductive flux flows inward to the chromosphere and outward to interplanetary space along field lines. The speed of the front of expanding chromospheric material generated by the impingement of the conduction front on the upper chromosphere exceeds the local sound velocity in a few minutes and eventually exceeds 100 million cm/sec.

CEDNBR: a computer code for transient thermal margin analysis of a reactor core

International Nuclear Information System (INIS)

Shesler, A.T.; Lehmann, C.R.

1976-09-01

The report describes the CEDNBR computer code. This code was developed for the transient thermal analysis of a pressurized water reactor core or a critical heat flux test. Included are the code structure, conservation equations, and correlations utilized by CEDNBR. The methods of modelling a reactor core and hot channel and a CHF test are presented. Comparisons of CEDNBR calculations are made with both empirical pressure loss data and simulated loss of flow test data. The code solves the one-dimensional conservation of mass, energy, and momentum equations and the equation of state for the fluid for either steady-state or transient conditions. Tabular time dependent functions of inlet temperatures, pressure, mass velocity, axial heat flux distributions, normalized heat flux, radial peaking factors, and incremental mixing factors are required input to the code. Transient effects are included in the calculation of enthalpy rise and fluid properties. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by applying a Critical Heat Flux (CHF) correlation to the computed local fluid properties. A code user's guide is provided for preparing input to the code. In addition, descriptions of the sub-routines used by CEDNBR are given

Thermal behavior of horizontally mixed surfaces on Mars

Science.gov (United States)

Putzig, Nathaniel E.; Mellon, Michael T.

2007-11-01

Current methods for deriving thermal inertia from spacecraft observations of planetary brightness temperature generally assume that surface properties are uniform for any given observation or co-located set of observations. As a result of this assumption and the nonlinear relationship between temperature and thermal inertia, sub-pixel horizontal heterogeneity may yield different apparent thermal inertia at different times of day or seasons. We examine the effects of horizontal heterogeneity on Mars by modeling the thermal behavior of various idealized mixed surfaces containing differing proportions of either dust, sand, duricrust, and rock or slope facets at different angles and azimuths. Latitudinal effects on mixed-surface thermal behavior are also investigated. We find large (several 100 J m -2 K -1 s -1/2) diurnal and seasonal variations in apparent thermal inertia even for small (˜10%) admixtures of materials with moderately contrasting thermal properties or slope angles. Together with similar results for layered surfaces [Mellon, M.T., Putzig, N.E., 2007. Lunar Planet. Sci. XXXVIII. Abstract 2184], this work shows that the effects of heterogeneity on the thermal behavior of the martian surface are substantial and may be expected to result in large variations in apparent thermal inertia as derived from spacecraft instruments. While our results caution against the over-interpretation of thermal inertia taken from median or average maps or derived from single temperature measurements, they also suggest the possibility of using a suite of apparent thermal inertia values derived from single observations over a range of times of day and seasons to constrain the heterogeneity of the martian surface.

Thermal sensations and comfort investigations in transient conditions in tropical office.

Science.gov (United States)

Dahlan, Nur Dalilah; Gital, Yakubu Yau

2016-05-01

The study was done to identify affective and sensory responses observed as a result of hysteresis effects in transient thermal conditions consisting of warm-neutral and neutral - warm performed in a quasi-experiment setting. Air-conditioned building interiors in hot-humid areas have resulted in thermal discomfort and health risks for people moving into and out of buildings. Reports have shown that the instantaneous change in air temperature can cause abrupt thermoregulation responses. Thermal sensation vote (TSV) and thermal comfort vote (TCV) assessments as a consequence of moving through spaces with distinct thermal conditions were conducted in an existing single-story office in a hot-humid microclimate, maintained at an air temperature 24 °C (± 0.5), relative humidity 51% (± 7), air velocity 0.5 m/s (± 0.5), and mean radiant temperature (MRT) 26.6 °C (± 1.2). The measured office is connected to a veranda that showed the following semi-outdoor temperatures: air temperature 35 °C (± 2.1), relative humidity 43% (± 7), air velocity 0.4 m/s (± 0.4), and MRT 36.4 °C (± 2.9). Subjective assessments from 36 college-aged participants consisting of thermal sensations, preferences and comfort votes were correlated against a steady state predicted mean vote (PMV) model. Local skin temperatures on the forehead and dorsal left hand were included to observe physiological responses due to thermal transition. TSV for veranda-office transition showed that no significant means difference with TSV office-veranda transition were found. However, TCV collected from warm-neutral (-0.24, ± 1.2) and neutral-warm (-0.72, ± 1.3) conditions revealed statistically significant mean differences (p thermal transition after travel from warm-neutral-warm conditions did not replicate the hysteresis effects of brief, slightly cool, thermal sensations found in previous laboratory experiments. These findings also indicate that PMV is an acceptable alternative to predict thermal

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

International Nuclear Information System (INIS)

Chvetsov, I.; Volkov, A.

2000-01-01

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

Consideration of loading conditions initiated by thermal transients in PWR pressure vessels

International Nuclear Information System (INIS)

Azodi; Glahn; Kersting; Schulz; Jansky.

1983-01-01

This report describes the present state of PWR-plants in the Federal Republic of Germany with respect to - the design of the primary pressure boundary - the analysis of thermal transients and resulting loads - the material conditions and neutron fluence - the requirements for protection against fast fracture. The experimental and analytical research and development programs are delineated together with some foreign R and D programs. It is shown that the parameters investigated (loading condition, crack shape and orientation etc.) cover a broad range. Extensive analytical investigations are emphasized. (orig./RW) [de

Strain and thermally induced magnetic dynamics and spin current in magnetic insulators subject to transient optical grating

Science.gov (United States)

Wang, Xi-Guang; Chotorlishvili, Levan; Berakdar, Jamal

2017-07-01

We analyze the magnetic dynamics and particularlythe spin current in an open-circuit ferromagnetic insulator irradiated by two intense, phase-locked laser pulses. The interference of the laser beams generates a transient optical grating and a transient spatio-temporal temperature distribution. Both effects lead to elastic and heat waves at the surface and into the bulk of the sample. The strain induced spin current as well as the thermally induced magnonic spin current are evaluated numerically on the basis of micromagnetic simulations using solutions of the heat equation. We observe that the thermo-elastically induced magnonic spin current propagates on a distance larger than the characteristic size of thermal profile, an effect useful for applications in remote detection of spin caloritronics phenomena. Our findings point out that exploiting strain adds a new twist to heat-assisted magnetic switching and spin-current generation for spintronic applications.

Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

International Nuclear Information System (INIS)

Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

2004-01-01

An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect

Thermal reactor safety

International Nuclear Information System (INIS)

1980-06-01

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport

Thermal reactor safety

Energy Technology Data Exchange (ETDEWEB)

1980-06-01

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Transient thermal response of a packed bed for energy storage unit utilizing phase change material: experimental and numerical study

International Nuclear Information System (INIS)

Bemansour, A.

2006-01-01

The present work concerns the numerical and experimental study of the transient response of a packed bed latent heat thermal energy storage system. Experiments were carried out to measures the transient temperature distributions inside a cylindrical bed, which is randomly packed with spheres having uniform sizes and encapsulated the paraffin wax as a phase change material (PCM), with air as a working fluid. A two-dimensional separate phases formulation is used to develop a numerical analysis of the transient response of the bed, considering the influence of both axial and radial thermal dispersion. The fluid energy equation was transformed by finite difference approximation and solved by alternating direction implicit scheme, while the PCM energy equation was solved using fully explicit scheme. This analysis can be applied for both charging and recovery modes and a broad range of Reynolds numbers. Measurements of both fluid and PCM temperature were conducted at different axial and radial positions and at different operating parameters. Experimental measurements of temperature distribution compare favorably with the numerical results over a broad range of Reynolds numbers.(Author)

Fission product transport and behavior during two postulated loss of flow transients in the air

International Nuclear Information System (INIS)

Adams, J.P.; Carboneau, M.L.

1991-01-01

This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10 -5 and 10 -7 per reactor year for LCP15 and LPP9, respectively

RETRAN-02: a program for transient thermal-hydraulic analysis of complex fluid-flow systems. Volume 4. Applications

International Nuclear Information System (INIS)

Peterson, C.E.; Gose, G.C.; McFadden, J.H.

1983-01-01

RETRAN-02 represents a significant achievement in the development of a versatile and reliable computer program for use in best estimate transient thermal-hydraulic analysis of light water reactor systems. The RETRAN-02 computer program is an extension of the RETRAN-01 program designed to provide analysis capabilities for 1) BWR and PWR transients, 2) small break loss of coolant accidents, 3) balance of plant modeling, and 4) anticipated transients without scram, while maintaining the analysis capabilities of the predecessor code. The RETRAN-02 computer code is constructed in a semimodular and dynamic dimensioned form where additions to the code can be easily carried out as new and improved models are developed. This report (the fourth of a five volume computer code manual) describes the verification and validation of RETRAN-02

Investigation of transient dynamics of capillary assisted particle assembly yield

Energy Technology Data Exchange (ETDEWEB)

Virganavičius, D. [Institute of Materials Science, Kaunas University of Technology, K. Baršausko St. 59, Kaunas LT-51423 (Lithuania); Laboratory of Micro- and Nanotechnology, Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); Juodėnas, M. [Institute of Materials Science, Kaunas University of Technology, K. Baršausko St. 59, Kaunas LT-51423 (Lithuania); Tamulevičius, T., E-mail: tomas.tamulevicius@ktu.lt [Institute of Materials Science, Kaunas University of Technology, K. Baršausko St. 59, Kaunas LT-51423 (Lithuania); Department of Physics, Kaunas University of Technology, Studentų St. 50, Kaunas LT-51368 (Lithuania); Schift, H. [Laboratory of Micro- and Nanotechnology, Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); Tamulevičius, S. [Institute of Materials Science, Kaunas University of Technology, K. Baršausko St. 59, Kaunas LT-51423 (Lithuania); Department of Physics, Kaunas University of Technology, Studentų St. 50, Kaunas LT-51368 (Lithuania)

2017-06-01

Highlights: • Regular particles arrays were assembled by capillary force assisted deposition. • Deposition yield dynamics was investigated at different thermal velocity regimes. • Yield transient behavior was approximated with logistic function. • Pattern density influence for switching behavior was assessed. - Abstract: In this paper, the transient behavior of the particle assembly yield dynamics when switching from low yield to high yield deposition at different velocity and thermal regimes is investigated. Capillary force assisted particle assembly (CAPA) using colloidal suspension of green fluorescent 270 nm diameter polystyrene beads was performed on patterned poly (dimethyl siloxane) substrates using a custom-built deposition setup. Two types of patterns with different trapping site densities were used to assess CAPA process dynamics and the influence of pattern density and geometry on the deposition yield transitions. Closely packed 300 nm diameter circular pits ordered in hexagonal arrangement with 300 nm pitch, and 2 × 2 mm{sup 2} square pits with 2 μm spacing were used. 2-D regular structures of the deposited particles were investigated by means of optical fluorescence and scanning electron microscopy. The fluorescence micrographs were analyzed using a custom algorithm enabling to identify particles and calculate efficiency of the deposition performed at different regimes. Relationship between the spatial distribution of particles in transition zone and ambient conditions was evaluated and quantified by approximation of the yield profile with a logistic function.

SAFE: A computer code for the steady-state and transient thermal analysis of LMR fuel elements

International Nuclear Information System (INIS)

Hayes, S.L.

1993-12-01

SAFE is a computer code developed for both the steady-state and transient thermal analysis of single LMR fuel elements. The code employs a two-dimensional control-volume based finite difference methodology with fully implicit time marching to calculate the temperatures throughout a fuel element and its associated coolant channel for both the steady-state and transient events. The code makes no structural calculations or predictions whatsoever. It does, however, accept as input structural parameters within the fuel such as the distributions of porosity and fuel composition, as well as heat generation, to allow a thermal analysis to be performed on a user-specified fuel structure. The code was developed with ease of use in mind. An interactive input file generator and material property correlations internal to the code are available to expedite analyses using SAFE. This report serves as a complete design description of the code as well as a user's manual. A sample calculation made with SAFE is included to highlight some of the code's features. Complete input and output files for the sample problem are provided

Anisotropic Thermal Behavior of Silicone Polymer, DC 745

Energy Technology Data Exchange (ETDEWEB)

Adams, Jillian Cathleen [Univ. of Oregon, Eugene, OR (United States). Dept. of Chemistry; Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Torres, Joseph Angelo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Volz, Heather Michelle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gallegos, Jennifer Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Yang, Dali [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-09-02

In material applications, it is important to understand how polymeric materials behave in the various environments they may encounter. One factor governing polymer behavior is processing history. Differences in fabrication will result in parts with varied or even unintended properties. In this work, the thermal expansion behavior of silicone DC 745 is studied. Thermomechanical analysis (TMA) is used to determine changes in sample dimension resulting from changes in temperature. This technique can measure thermal events such as the linear coefficient of thermal expansion (CTE), melting, glass transitions, cure shrinkage, and internal relaxations. Using a thermomechanical analyzer (Q400 TMA), it is determined that DC 745 expands anisotropically when heated. This means that the material has a different CTE depending upon which direction is being measured. In this study, TMA experiments were designed in order to confirm anisotropic thermal behavior in multiple DC 745 samples of various ages and lots. TMA parameters such as temperature ramp rate, preload force, and temperature range were optimized in order to ensure the most accurate and useful data. A better understanding of the thermal expansion of DC 745 will allow for more accurate modeling of systems using this material.

Ballooning of CANDU pressure tube in local thermal transients

International Nuclear Information System (INIS)

Mihalache, Maria; Ionescu, Viorel

2008-01-01

In certain LOCA scenarios for the CANDU fuel channel, the ballooning of the pressure tube and contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator takes place through the contact area. If the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. In INR-Pitesti the DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after occurrence of the contact between the two tubes. The code contains few models: thermal creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This paper gives a DELOCA code description and the fuel channel behaviour analysis, in transient temperature conditions of the pressure tube, using the materials properties, time and temperature dependencies of these properties as obtained in the different laboratories of the world and in the INR - Pitesti in the last years. DELOCA computer code simulated the fuel channel response to the constant heating rates of inside pressure tube surface. The paper presents contact temperature and time dependencies on the heating rate, and the appropriate fitting functions. (authors)

Transient Mass and Thermal Transport during Methane Adsorption into the Metal-Organic Framework HKUST-1.

Science.gov (United States)

Babaei, Hasan; McGaughey, Alan J H; Wilmer, Christopher E

2018-01-24

Methane adsorption into the metal-organic framework (MOF) HKUST-1 and the resulting heat generation and dissipation are investigated using molecular dynamics simulations. Transient simulations reveal that thermal transport in the MOF occurs two orders of magnitude faster than gas diffusion. A large thermal resistance at the MOF-gas interface (equivalent to 127 nm of bulk HKUST-1), however, prevents fast release of the generated heat. The mass transport resistance at the MOF-gas interface is equivalent to 1 nm of bulk HKUST-1 and does not present a bottleneck in the adsorption process. These results provide important insights into the application of MOFs for gas storage applications.

Enhancing the ABAQUS Thermomechanics Code to Simulate Multidimensional Steady and Transient Fuel Rod Behavior

Energy Technology Data Exchange (ETDEWEB)

Williamson, R.L.; Knoll, D.A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-3855 (United States)

2009-06-15

Important aspects of fuel rod behavior, for example pellet-clad interaction (PCI), fuel fracture, and non-axisymmetric cooling and oxide formation, are inherently 3-D. Current fuel rod simulation codes typically approximate such behavior using a quasi 2D (or 1.5D) approach and, often, separate codes must be used for steady and transient (or accident) conditions. Notable exceptions are the EPRI propriety code FALCON which is 2D and can be applied to steady or transient operation, and TOUTATIS which is 3D. Recent studies have indicated the need for multidimensional fuel rod simulation capability, particularly for accurate predictions of PCI. The Idaho National Laboratory (INL) is currently developing next-generation capability to model nuclear fuel performance. The goal is to develop a 2D/3D computer code (BISON) which solves the fully coupled thermomechanics equations, includes multi-physics constitutive behavior for both fuel and cladding materials, and is designed for efficient use on highly parallel computers. To provide guidance and a proto-typing environment for this effort, plus provide the INL with near-term fuel modeling capability, the commercially available ABAQUS thermomechanics software has been enhanced to include the fuel behavior phenomena necessary to afford a practical fuel performance simulation capability. This paper details the enhancements which have been implemented in ABAQUS to date, and provides results of a multi-pellet fuel problem which demonstrates the new capability. ABAQUS employs modern finite element methods to solve the nonlinear thermomechanics equations in 1, 2, or 3-D, using linear or quadratic elements. The temperature and displacement fields are solved in a fully-coupled fashion, using sophisticated iteration and time integration error control. The code includes robust contact algorithms, essential for computing multidimensional pellet-pellet or pellet-clad interaction. Extensive constitutive models are available, including

Temporal neural network for the identification of nuclear power plant transients

International Nuclear Information System (INIS)

Uluyol, O.; Ragheb, M.

1993-01-01

In this paper a layered spatiotemporal neural network is proposed for the identification of nuclear power plant transients. The developed layered spatiotemporal network is inspired by the formal avalanche structure developed by S. Grossberg and offers advantages compared with the stationary pattern approach using the perceptron paradigm. Each layer in the network is trained to recognize a separate time-dependent accident scenario. Within each scenario, the temporal behavior of the relevant parameters such as pressurizer pressure, pressurizer water volume, cold and hot legs temperatures, vessel flow, and power, are considered. Numerical cases are considered where the proposed methodology is applied to two nuclear power plant anticipated transient scenarios: the Station Blackout and the Anticipated Transient without Scram transients in a pressurized water reactor . The transient signatures used were generated by modeling the accidents using RELAP5/MOD2, a best-estimate thermal-hydraulics numerical code. The ability of the proposed layered spatiotemporal network to operate at different noise levels is investigated. Its incorporation within an Insightful Algorithm and Anticipatory Systems context for identifying and in predicting the course of nuclear transients is discussed

DORT-TD/THERMIX solutions for the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

International Nuclear Information System (INIS)

Tyobeka, Bismark; Pautz, Andreas; Ivanov, Kostadin

2008-01-01

In new reactor designs that are still under review such as the PBMR, not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400 MW OECD/NEA/NSC coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate the transient scenarios in the above-mentioned benchmark problem. Steady-state calculations results are compared with selected participants' results as well as transient models in which the diffusion and transport theory solutions of the same code system are directly compared. Several sensitivity studies are also shown in order to determine how much the change in certain parameters influences the overall behaviour of a given transient. It is shown in this paper that DORT-TD/THERMIX is a versatile tool which can be deployed for design and safety analyses of high temperature reactors of pebble-bed type. (authors)

Development of system for automatic measurement of transient photocurrent and thermally stimulated current

Directory of Open Access Journals (Sweden)

Asdrubal Antonio Ramirez Botero

2017-01-01

Full Text Available This paper presents details of the design and implementation of a system for measuring of thermally stimulated current (TSC and transient photocurrent (Iph, developed using the Virtual Instrumentation concept. For that we have used National Instrument hardware and the LabView® package as software. The system is controlled by a virtual instrument (VI which includes facilities to perform measurements of photocurrent keeping the temperature of the sample and the pressure of the chamber of measurement controlled as well as  real time display of the Iph vs t and TSC vs T curves. The system was tested by performing transient photocurrent and TSC measurements on CH3NH3PbI3 thin films that are generally used as absorbent layer of solar cells. This type of characterization is very useful to get information of the  trapping and recombination processes that affect the transport properties of the devices.

Molybdenum peroxo complex. Structure and thermal behavior

Energy Technology Data Exchange (ETDEWEB)

Segawa, Koichi; Ooga, Katsumi; Kurusu, Yasuhiko

1984-10-01

The molybdenum peroxide (Mo-y) prepared by oxidation of molybdenum metal with hydrogen peroxide has been studied to determine its structure and thermal behavior. Temperature programmed decomposition has been used to study the thermal stability of Mo-y. Two distinct peaks, I and II, of decomposition processes are discernible in Mo-y. Peak I corresponds to the elimination of water of crystallization and peak II to the decomposition of a peroxide ion of Mo-y. IR and UV examinations support the results of the thermal analysis. The IR band at 931 cm/sup -1/ and the UV band at 381 nm show the same thermal behavior. Both bands are attributable to the peroxide ion of Mo-y. Spectroscopic studies show that Mo-y has the tetrahedral coordination derived from the single molybdenum complex, which has double bond oxygens attached to Mo atom and has a symmetric type of peroxide ion with one water of crystallization.

Thermal expansion behavior in fabricated cellular structures

International Nuclear Information System (INIS)

Oruganti, R.K.; Ghosh, A.K.; Mazumder, J.

2004-01-01

Thermal expansion behavior of cellular structures is of interest in applications where undesirable deformation and failure are caused by thermal expansion mismatch. This report describes the role of processing-induced effects and metallurgical aspects of melt-processed cellular structures, such as a bi-material structure designed to contract on heating, as well as uni-material structures of regular and stochastic topology. This bi-material structure utilized the principle of internal geometric constraints to alter the expansion behavior of the internal ligaments to create overall contraction of the structure. Homogenization design method was used to design the structure, and fabrication was by direct metal deposition by laser melting of powder in another part of a joint effort. The degree of porosity and grain size in the fabricated structure are characterized and related to the laser deposition parameters. The structure was found to contract upon heating over a short range of temperature subsequent to which normal expansion ensued. Also examined in this report are uni-material cellular structures, in which internal constraints arise from residual stress variations caused by the fabrication process, and thereby alter their expansion characteristics. A simple analysis of thermal strain of this material supports the observed thermal expansion behavior

Transient thermal stresses in a circular cylinder with constrained ends

International Nuclear Information System (INIS)

Goshima, Takahito; Miyao, Kaju

1986-01-01

This paker deals with the transient thermal stresses in a finite circular cylinder constrained at both end surfaces and subjected to axisymmetric temperature distribution on the lateral surface. The thermoelastic problem is formulated in terms of a thermoelastic displacement potential and three harmonic stress functions. Numerical calculations are carried out for the case of the uniform temperature distribution on the lateral surface. The stress distributions on the constrained end and the free suface are shown graphically, and the singularity in stresses appearing at the circumferencial edge is considered. Moreover, the approximate solution based upon the plane strain theory is introduced in order to compare the rigorous one, and it is considered how the length of the cylinder and the time proceeds affect on the accuracy of the approximation. (author)

Transient thermal protection of film covering circular aperture by sublimation and weak decomposition

Energy Technology Data Exchange (ETDEWEB)

Havstad, Mark A.; Miles, Robin R.; Hsieh, Henry, E-mail: hsieh6@llnl.gov

2015-03-15

Highlights: • Precise sublimating layers can provide protection in transient thermal environments. • Sensitivity analysis shows that the uncertainty in properties has modest influence. • It is likely that methane layers are a good choice for IFE targets. - Abstract: Unwanted heating of sensitive surfaces in harsh thermal environments can be prevented by precise application of sacrificial materials such as sublimation layers and pyrolyzing films. The use of sublimation for the protection of circular polyimide membranes subjected to brief (∼100 ms) heating by infrared radiation and hot (6000 K) inert gas convection is analyzed. Selection of sublimation material and sublimation layer and membrane thickness is considered with emphasis on providing sufficient thermal protection yet negligible unwanted material remaining at the end of a specified heating period. Though the analysis here is general, the motivation is protection of the polyimide films covering the laser entrance holes on IFE (inertial fusion energy) hohlraums being injected into the hot gas (xenon) protecting IFE reactor chambers. Both one and two dimensional thermal models are used to develop a robust thermal concept. Sensitivity analyses (SA) methods are exercised to show where the design may be vulnerable and which input parameters have the greatest effect on performance and likelihood of success. For the design and conditions considered, methane sublimating layers are probably preferred over xenon or pentane.

Development of the containment transient analysis code for the passive reactor

Energy Technology Data Exchange (ETDEWEB)

Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-05-01

This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.

The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

International Nuclear Information System (INIS)

Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

2006-01-01

The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

Multi-scale modeling of the mechanical behavior of polycrystalline ice under transient creep.

OpenAIRE

Suquet , Pierre; Moulinec , Hervé; Castelnau , O.; Montagnat , Maurine; Lahellec , Noël; Grennerat , Fanny; Duval , Paul; Brenner , Renald

2012-01-01

International audience; Ice is a challenging material for understanding the overall behavior of polycrystalline materials and more specifically the coupling between elastic and viscous effects during transient creep. At the single crystal level, ice is an hexagonal material with a rather weak elastic anisotropy but with a strong viscoplastic anisotropy. The strain-stress curve of ice single crystals shows a softening behavior depending on the strain-rate. The strong viscous anisotropy of ice ...

Neutron and thermo - hydraulic model of a reactivity transient in a nuclear power plant fuel element

International Nuclear Information System (INIS)

Oliva, Jose de Jesus Rivero

2012-01-01

A reactivity transient without reactor scram was modeled and calculated using analytical expressions for the space distributions of the temperature fields, combined with discrete numerical calculations for the time dependences of thermal power and temperatures. The transient analysis covered the time dependencies of reactivity, global thermal power, fuel heat flux and temperatures in fuel, cladding and cooling water. The model was implemented in Microsoft Office Excel, dividing the Excel file in several separated worksheets for input data, initial steady-state calculations, calculation of parameters non-depending on eigenvalues, eigenvalues determination, calculation of parameters depending on eigenvalues, transient calculation and graphical representation of intermediate and final results. The results show how the thermal power reaches a new equilibrium state due to the negative reactivity feedback derived from the fuel temperature increment. Nevertheless, the reactor mean power increases 40% during the first second and, in the hottest channel, the maximum fuel temperature goes to a significantly high value, slightly above 2100 deg C, after 8 seconds of transient. Consequently, the results confirm that certain degree of fuel damage could be expected in case of a reactor scram failure. Once the basic model has being established the scope of accidents for future analyses can be extended, modifying the nuclear power behavior (reactivity) during transient and the boundary conditions for coolant temperature. A more complex model is underway for an annular fuel element. (author)

The simulation of CANDU fuel channel behavior in thermal transient conditions

International Nuclear Information System (INIS)

Mihalache, M.; Roth, M.; Radu, V.; Dumitrescu, I.

2005-01-01

In certain LOCA conditions into the CANDU fuel channel, is possible the ballooning of the pressure tube and the contact with the calandria tube. After the contact moment, a radial heat transfer to the moderator through the contact area is occurs. When the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. Thus, the fuel channel could lose its integrity. This paper present a computer code, DELOCA, developed in INR, which simulate the transient thermo-mechanical behaviour of CANDU fuel channel before and after contact. The code contains few models: alloy creep, heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. It was verified step by step by Contact1 and Cathena codes. In this paper, the results obtained at different temperature increasing rates are presented. Also, the contact moment for a RIH 5% postulated accident was presented. The input data was furnished by the Cathena thermo-hydraulic code. (author)

Thermal sensation and comfort models for non-uniform and transient environments: Part III: whole-body sensation and comfort

OpenAIRE

Zhang, Hui; Arens, Edward; Huizenga, Charlie; Han, Taeyoung

2009-01-01

A three-part series presents the development of models for predicting the local thermal sensation (Part I) and local thermal comfort (Part II) of different parts of the human body, and also the whole-body sensation and comfort (Part III) that result from combinations of local sensation and comfort. The models apply to sedentary activities in a range of environments: uniform and non-uniform, stable and transient. They are based on diverse findings from the literature and from body-part-specifi...

Severe transient tests on operation steam generators: Analysis of the fluid structure dynamic thermal interaction

International Nuclear Information System (INIS)

Billon, F.; David, J.; Procaccia, H.

1983-01-01

The operating efficiency of steam generators (S.G.s) and their structural integrity depend on the design configurations of the feedwater spray within the S.G., and on the operating procedure. To check the merit of some design modifications, and to verify the fluid-structure interaction with a view to preserve the S.G.s integrity during severe operating transients, a special instrumentation that admits the determination of the instantaneous thermal hydraulic characteristics of the flow in the secondary water and the S.G. tube sheet, has been installed by EDF on one steam generator of Tricastin unit 1 power plant. In parallel, FRAMATOME has developped a computer code, TEMPTRON, that allows the calculations of the thermal loads and the consequent stresses in the most sollicited zones of the steam generator during transient operation of the plant. This code divides the S.G. into three parts: - the first concerns the S.G.s region above the downcomer, zone where the mixing between hot water and cold feedwater occurs, - the second is the downcomer itself which is divided into n segments, - the third concerns the tube sheet zone which is also divided into n segments. The most severe transient test performed is the auxiliary cold feedwater injection into the steam generator during a hot standby of the plant: two levels of flow rate have been realised: 55 and 110 m 3 /h of 42 0 C feedwater. The tests have shown that if the cold feedwater injection occurs when the steam generator water level is below feedwater ring, the lowest fluid temperature reached at tube sheet inlet is about 230 0 C. (orig.)

Challenges in mechanical modeling of SFR fuel rod transient behavior

Energy Technology Data Exchange (ETDEWEB)

Feria, F.; Herranz, L. E.

2013-07-01

Modeling of SFR fuel rod mechanical behavior under transient conditions entails the development of a creep law to predict cladding viscoplastic strain. In this regard, this work is focused on defining a proper clad creep law structure as the basis to set a suitable model under SFR off-normal conditions as transient overpower and loss of fluid. To do so, a review of in-codes clad creep models has been done by using SAS-SFR, SCANAIR and ASTEC. The proposed creep model has been structured in two parts: viscoplastic behaviour before the failure (primary and secondary creep) and the failure due to viscoplastic collapse (tertiary creep). In order to model the first part, Norton creep law has been proposed as a conservative option. An irradiation hardening factor should be included for best estimate calculations. The recommendation for the second part is to apply a failure criterion based on strain limit or rupture time, which allows achieving conservative results.

Systems of independent Markov components and their transient behavior

International Nuclear Information System (INIS)

Keilson, J.

1975-01-01

The transient behavior of redundant systems of independent, repairable, memoryless (two state) components is studied. Four failure times for such systems are considered, each an exit time from the set of working states for initial system conditions of interest: the failure time from the perfect state, the post-recovery exit time, the ergodic exit time, and the quasi-stationary exit time. The structure of these failure time distributions and their interrelation are discussed and asymptotic estimates and bounds for their expectations are presented. When such systems have high reliability, the failure time distributions are approximately exponential and asymptotically equivalent

The dynamic behavior of the SUPER-PHENIX reactor under unprotected transient

International Nuclear Information System (INIS)

Gouriou, A.; Francillon, E.; Kayser, G.; Malenfer, G.; Languille, A.

1982-01-01

Due to design changes and progress on the knowledge of feed-back effects, a reactualization of the dynamic behavior of SUPER-PHENIX under unprotected transients was undertaken. We present the main data on feed-back characteristics and the results of dynamic calculations. With the present state of knowledge, the former conclusion is confirmed: the dynamic evolution is very slow and no irreversible phenomena happen in the short term

Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

Science.gov (United States)

Jefferies, K. S.; Tew, R. C.

1974-01-01

A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

Transient thermal, hydraulic, and mechanical analysis of a counter flow offset strip fin intermediate heat exchanger using an effective porous media approach

Science.gov (United States)

Urquiza, Eugenio

This work presents a comprehensive thermal hydraulic analysis of a compact heat exchanger using offset strip fins. The thermal hydraulics analysis in this work is followed by a finite element analysis (FEA) to predict the mechanical stresses experienced by an intermediate heat exchanger (IHX) during steady-state operation and selected flow transients. In particular, the scenario analyzed involves a gas-to-liquid IHX operating between high pressure helium and liquid or molten salt. In order to estimate the stresses in compact heat exchangers a comprehensive thermal and hydraulic analysis is needed. Compact heat exchangers require very small flow channels and fins to achieve high heat transfer rates and thermal effectiveness. However, studying such small features computationally contributes little to the understanding of component level phenomena and requires prohibitive computational effort using computational fluid dynamics (CFD). To address this issue, the analysis developed here uses an effective porous media (EPM) approach; this greatly reduces the computation time and produces results with the appropriate resolution [1]. This EPM fluid dynamics and heat transfer computational code has been named the Compact Heat Exchanger Explicit Thermal and Hydraulics (CHEETAH) code. CHEETAH solves for the two-dimensional steady-state and transient temperature and flow distributions in the IHX including the complicating effects of temperature-dependent fluid thermo-physical properties. Temperature- and pressure-dependent fluid properties are evaluated by CHEETAH and the thermal effectiveness of the IHX is also calculated. Furthermore, the temperature distribution can then be imported into a finite element analysis (FEA) code for mechanical stress analysis using the EPM methods developed earlier by the University of California, Berkeley, for global and local stress analysis [2]. These simulation tools will also allow the heat exchanger design to be improved through an

Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

International Nuclear Information System (INIS)

Grotz, S.; Ghoniem, N.M.

1986-02-01

The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250 0 C)

Thermal Behavior of Tacca leontopetaloides Starch-Based Biopolymer

Directory of Open Access Journals (Sweden)

Nurul Shuhada Mohd Makhtar

2013-01-01

Full Text Available Starch is used whenever there is a need for natural elastic properties combined with low cost of production. However, the hydrophilic properties in structural starch will decrease the thermal performance of formulated starch polymer. Therefore, the effect of glycerol, palm olein, and crude palm oil (CPO, as plasticizers, on the thermal behavior of Tacca leontopetaloides starch incorporated with natural rubber in biopolymer production was investigated in this paper. Four different formulations were performed and represented by TPE1, TPE2, TPE3, and TPE4. The compositions were produced by using two-roll mill compounding. The sheets obtained were cut into small sizes prior to thermal testing. The addition of glycerol shows higher enthalpy of diffusion in which made the material easily can be degraded, leaving to an amount of 6.6% of residue. Blending of CPO with starch (TPE3 had a higher thermal resistance towards high temperature up to 310°C and the thermal behavior of TPE2 only gave a moderate performance compared with other TPEs.

Steady- and transient-state analyses of fully ceramic microencapsulated fuel loaded reactor core via two-temperature homogenized thermal-conductivity model

International Nuclear Information System (INIS)

Lee, Yoonhee; Cho, Nam Zin

2015-01-01

Highlights: • Fully ceramic microencapsulated fuel-loaded core is analyzed via a two-temperature homogenized thermal-conductivity model. • The model is compared to harmonic- and volumetric-average thermal conductivity models. • The three thermal analysis models show ∼100 pcm differences in the k eff eigenvalue. • The three thermal analysis models show more than 70 K differences in the maximum temperature. • There occur more than 3 times differences in the maximum power for a control rod ejection accident. - Abstract: Fully ceramic microencapsulated (FCM) fuel, a type of accident-tolerant fuel (ATF), consists of TRISO particles randomly dispersed in a SiC matrix. In this study, for a thermal analysis of the FCM fuel with such a high heterogeneity, a two-temperature homogenized thermal-conductivity model was applied by the authors. This model provides separate temperatures for the fuel-kernels and the SiC matrix. It also provides more realistic temperature profiles than those of harmonic- and volumetric-average thermal conductivity models, which are used for thermal analysis of a fuel element in VHTRs having a composition similar to the FCM fuel, because such models are unable to provide the fuel-kernel and graphite matrix temperatures separately. In this study, coupled with a neutron diffusion model, a FCM fuel-loaded reactor core is analyzed via a two-temperature homogenized thermal-conductivity model at steady- and transient-states. The results are compared to those from harmonic- and volumetric-average thermal conductivity models, i.e., we compare k eff eigenvalues, power distributions, and temperature profiles in the hottest single-channel at steady-state. At transient-state, we compare total powers, reactivity, and maximum temperatures in the hottest single-channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized thermal

Transient heat transfer in a directly-irradiated solar chemical reactor for the thermal dissociation of ZnO

International Nuclear Information System (INIS)

Mueller, R.; Lipinski, W.; Steinfeld, A.

2008-01-01

A numerical and experimental investigation is carried out in a solar thermochemical reactor for the thermal dissociation of ZnO at 2000 K using concentrated solar energy. The reactor consists of a cavity-receiver lined with ZnO particles and directly exposed to high-flux irradiation. A transient heat transfer model is formulated to link the rate of radiation, convection, and conduction heat transfer to the reaction kinetics. The radiosity and Monte Carlo methods are applied to obtain the distribution of net radiative fluxes at the internal surfaces of the reactor cavity and at the surface of the ZnO bed. Validation is accomplished in terms of the calculated and measured transient temperature profiles and chemical reaction rates

Thermal behavior of asphalt cements

International Nuclear Information System (INIS)

Claudy, P.M.; Letoffe, J.M.; Martin, D.; Planche, J.P.

1998-01-01

Asphalt cements are highly complex mixtures of hydrocarbon molecules whose thermal behavior is of prime importance for petroleum and road industry. From DSC, the determination of several thermal properties of asphalts is given, e.g. glass-transition temperature and crystallized fraction content.The dissolution of a pure n-paraffin C n H 2n+2 in an asphalt, as seen by DSC, should be a single peak. For 20 g of these glasses change with time and temperature. The formation of the crystallized phases is superposed to the enthalpic relaxation of the glasses, making a kinetic study very difficult. (Copyright (c) 1998 Elsevier Science B.V., Amsterdam. All rights reserved.)

Application of sensitivity analysis to a simplified coupled neutronic thermal-hydraulics transient in a fast reactor using Adjoint techniques

International Nuclear Information System (INIS)

Gilli, L.; Lathouwers, D.; Kloosterman, J.L.; Van der Hagen, T.H.J.J.

2011-01-01

In this paper a method to perform sensitivity analysis for a simplified multi-physics problem is presented. The method is based on the Adjoint Sensitivity Analysis Procedure which is used to apply first order perturbation theory to linear and nonlinear problems using adjoint techniques. The multi-physics problem considered includes a neutronic, a thermo-kinetics, and a thermal-hydraulics part and it is used to model the time dependent behavior of a sodium cooled fast reactor. The adjoint procedure is applied to calculate the sensitivity coefficients with respect to the kinetic parameters of the problem for two reference transients using two different model responses, the results obtained are then compared with the values given by a direct sampling of the forward nonlinear problem. Our first results show that, thanks to modern numerical techniques, the procedure is relatively easy to implement and provides good estimation for most perturbations, making the method appealing for more detailed problems. (author)

Advanced Instrumentation for Transient Reactor Testing

Energy Technology Data Exchange (ETDEWEB)

Corradini, Michael L.; Anderson, Mark; Imel, George; Blue, Tom; Roberts, Jeremy; Davis, Kurt

2018-01-31

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

Numerical simulation of the transient thermal-hydraulic behaviour of the ITER blanket cooling system under the draining operational procedure

Energy Technology Data Exchange (ETDEWEB)

Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Vallone, E., E-mail: eug.vallone@gmail.com [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy)

2015-10-15

Highlights: • ITER blanket cooling system hydraulic behaviour is studied under draining transient. • A computational approach based on the finite volume method has been followed. • Draining efficiency has been assessed in term of transient duration and residual water. • Transient duration ranges from ∼40 to 50 s, under the reference draining scenario. • Residual water is predicted to range from few tens of gram up to few kilograms. - Abstract: Within the framework of the research and development activities supported by the ITER Organization on the blanket system issues, an intense analysis campaign has been performed at the University of Palermo with the aim to investigate the thermal-hydraulic behaviour of the cooling system of a standard 20° sector of ITER blanket during the draining transient operational procedure. The analysis has been carried out following a theoretical-computational approach based on the finite volume method and adopting the RELAP5 system code. In a first phase, attention has been focused on the development and validation of the finite volume models of the cooling circuits of the most demanding modules belonging to the standard blanket sector. In later phase, attention has been put to the numerical simulation of the thermal-hydraulic transient behaviour of each cooling circuit during the draining operational procedure. The draining procedure efficiency has been assessed in terms of both transient duration and residual amount of coolant inside the circuit, observing that the former ranges typically between 40 and 120 s and the latter reaches at most ∼8 kg, in the case of the cooling circuit of twinned modules #6–7. Potential variations to operational parameters and/or to circuit lay-out have been proposed and investigated to optimize the circuit draining performances. In this paper, the set-up of the finite volume models is briefly described and the key results are summarized and critically discussed.

Transient thermal modeling of permafrost conditions in Southern Norway

Directory of Open Access Journals (Sweden)

S. Westermann

2013-04-01

Full Text Available Thermal modeling is a powerful tool to infer the temperature regime of the ground in permafrost areas. We present a transient permafrost model, CryoGrid 2, that calculates ground temperatures according to conductive heat transfer in the soil and in the snowpack. CryoGrid 2 is forced by operational air temperature and snow-depth products for potential permafrost areas in Southern Norway for the period 1958 to 2009 at 1 km2 spatial resolution. In total, an area of about 80 000 km2 is covered. The model results are validated against borehole temperatures, permafrost probability maps from "bottom temperature of snow" measurements and inventories of landforms indicative of permafrost occurrence. The validation demonstrates that CryoGrid 2 can reproduce the observed lower permafrost limit to within 100 m at all validation sites, while the agreement between simulated and measured borehole temperatures is within 1 K for most sites. The number of grid cells with simulated permafrost does not change significantly between the 1960s and 1990s. In the 2000s, a significant reduction of about 40% of the area with average 2 m ground temperatures below 0 °C is found, which mostly corresponds to degrading permafrost with still negative temperatures in deeper ground layers. The thermal conductivity of the snow is the largest source of uncertainty in CryoGrid 2, strongly affecting the simulated permafrost area. Finally, the prospects of employing CryoGrid 2 as an operational soil-temperature product for Norway are discussed.

Study of ATES thermal behavior using a steady flow model

Science.gov (United States)

Doughty, C.; Hellstroem, G.; Tsang, C. F.; Claesson, J.

1981-01-01

The thermal behavior of a single well aquifer thermal energy storage system in which buoyancy flow is neglected is studied. A dimensionless formulation of the energy transport equations for the aquifer system is presented, and the key dimensionless parameters are discussed. A simple numerical model is used to generate graphs showing the thermal behavior of the system as a function of these parameters. Some comparisons with field experiments are given to illustrate the use of the dimensionless groups and graphs.

Steady state and transient thermal-hydraulic characterization of full-scale ITER divertor plasma facing components

International Nuclear Information System (INIS)

Tincani, A.; Malavasi, A.; Ricapito, I.; Riccardi, B.; Di Maio, P.A.; Vella, G.

2007-01-01

In the frame of the activities related to ITER divertor R and D, ENEA CR Brasimone was charged by EFDA (European Fusion Design Agreement) to investigate the thermal-hydraulic behaviour of the full-scale divertor plasma facing components, i.e. Inner Vertical Target, Dome Liner and Outer Vertical Target, both in steady state and during draining and drying transient. More in detail, for each PFC, the first phase of the work is the steady state hydraulic characterization which consists of: - measurements of pressure drops at different temperatures; - determination of the velocity distribution in the internal channels; - check the possible insurgence of cavitation. The subsequent phase of the thermal-hydraulic characterization foresees a testing campaign of draining and drying procedure by means of a suitable gas flow. The objective of this experimental procedure is to eliminate in the most efficient way the residual amount of water after gravity discharge. In order to accomplish this experimental campaign a significant modification of CEF1 loop has been designed and realized. This paper presents, first of all, the experimental set-up, the agreed test matrix and the achieved results for both steady state and transient tests. Moreover, the level of the implementation of a predictive hydraulic model, based on RELAP 5 code, as well as its results are described, discussed and compared with the experimental ones. (orig.)

Fuel pin behavior under slow ramp-type transient-overpower conditions in the CABRI-FAST experiments

International Nuclear Information System (INIS)

Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, Jean

2009-01-01

In the CABRI-FAST experimental program, four in-pile tests were performed with slow power-ramp-type transient-overpower conditions (called hereafter as 'slow TOP') to study transient fuel pin behavior under inadvertent control rod withdrawal events in liquid metal cooled fast breeder reactors. Annular-pellet fuel pins were used in three tests, while a solid-pellet fuel pin was used in the other test. All of these pins were pre-irradiated in Phenix. The slow TOP test with a solid-pellet fuel pin was realized as a comparatory test against an existing test (E12) in the CABRI-2 program. In the CABRI-FAST test (BCF1), a power ramp rate of 3% Po/s was applied, while in the CABRI-2 test, 1% Po/s was adopted. Moreover, overpower condition was maintained for a few seconds beyond the observed pin failure in the BCF1 test. In spite of the different power ramp rates, evaluated fuel thermal conditions at the observed failure time are quite similar. The continued overpower condition in the BCF1 test resulted in gradual degradation of the pin structure providing information effective for evaluation of various accident scenarios. Three slow TOP tests with the annular fuel in the CABRI-FAST program resulted in no pin failure showing high failure threshold. Based on post-test examination data and a theoretical evaluation, it was concluded that intra-pin free spaces, such as central hole, macroscopic cracks and fuel-cladding gap effectively mitigated fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in case of very large amount of fuel melting. Furthermore, such cavity pressurization was effectively mitigated by a molten-fuel squirting into the upper blanket region pushing the blanket pellets upward. These CABRI FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database with various fuel and transient conditions. (author)

A theoretical prediction of critical heat flux in saturated pool boiling during power transients

International Nuclear Information System (INIS)

Pasamehmetoglu, K.O.; Nelson, R.A.; Gunnerson, F.S.

1987-01-01

Understanding and predicting critical heat flux (CHF) behavior during steady-state and transient conditions is of fundamental interest in the design, operation, and safety of boiling and two-phase flow devices. Presented within this paper are the results of a comprehensive theoretical study specifically conducted to model transient CHF behavior in saturated pool boiling. Thermal energy conduction within a heating element and its influence on the CHF are also discussed. The resultant theory provides new insight into the basic physics of the CHF phenomenon and indicates favorable agreement with the experimental data from cylindrical heaters with small radii. However, the flat-ribbon heater data compared poorly with the present theory, although the general trend was predicted. Finally, various factors that affect the discrepency between the data and the theory are listed

Diurnal Thermal Behavior of Photovoltaic Panel with Phase Change Materials under Different Weather Conditions

Directory of Open Access Journals (Sweden)

Jae-Han Lim

2017-12-01

Full Text Available The electric power generation efficiency of photovoltaic (PV panels depends on the solar irradiation flux and the operating temperature of the solar cell. To increase the power generation efficiency of a PV system, this study evaluated the feasibility of phase change materials (PCMs to reduce the temperature rise of solar cells operating under the climate in Seoul, Korea. For this purpose, two PCMs with different phase change characteristics were prepared and the phase change temperatures and thermal conductivities were compared. The diurnal thermal behavior of PV panels with PCMs under the Seoul climate was evaluated using a 2-D transient thermal analysis program. This paper discusses the heat flow characteristics though the PV cell with PCMs and the effects of the PCM types and macro-packed PCM (MPPCM methods on the operating temperatures under different weather conditions. Selection of the PCM type was more important than the MMPCM methods when PCMs were used to enhance the performance of PV panels and the mean operating temperature of PV cell and total heat flux from the surface could be reduced by increasing the heat transfer rate through the honeycomb grid steel container for PCMs. Considering the mean operating temperature reduction of 4 °C by PCM in this study, an efficiency improvement of approximately 2% can be estimated under the weather conditions of Seoul.

Modelling transient temperature distribution for injecting hot water through a well to an aquifer thermal energy storage system

Science.gov (United States)

Yang, Shaw-Yang; Yeh, Hund-Der; Li, Kuang-Yi

2010-10-01

Heat storage systems are usually used to store waste heat and solar energy. In this study, a mathematical model is developed to predict both the steady-state and transient temperature distributions of an aquifer thermal energy storage (ATES) system after hot water is injected through a well into a confined aquifer. The ATES has a confined aquifer bounded by aquicludes with different thermomechanical properties and geothermal gradients along the depth. Consider that the heat is transferred by conduction and forced convection within the aquifer and by conduction within the aquicludes. The dimensionless semi-analytical solutions of temperature distributions of the ATES system are developed using Laplace and Fourier transforms and their corresponding time-domain results are evaluated numerically by the modified Crump method. The steady-state solution is obtained from the transient solution through the final-value theorem. The effect of the heat transfer coefficient on aquiclude temperature distribution is appreciable only near the outer boundaries of the aquicludes. The present solutions are useful for estimating the temperature distribution of heat injection and the aquifer thermal capacity of ATES systems.

PRETTA：A COMPUTER PROGRAM FOR PWR PRESSURIZER’S TRANSIENT THERMODYNAMICS

Institute of Scientific and Technical Information of China (English)

阿谢德; 徐济鋆

2001-01-01

A computer program PRETTA “Pressurizer Transient Thermodynamics Analysis” was developed for the prediction of pressurizer under transient conditions. It is based on the solution of the conservation laws of heat and mass applied to the three separate and non equilibrium thermodynamic regions. In the program all of the important thermal-hydraulics phenomena occurring in the pressurizer: stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer have been considered. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented computer program will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant's pressurizer performance.

Transient thermal stresses in circular cylinder under intermittently sudden heat generation

International Nuclear Information System (INIS)

Sugano, Y.; Saito, K.; Takeuti, Y.

1975-01-01

The thermal stresses associated with the transient temperature distribution arising in a circular cylinder under intermittently changing sudden heat generation over a finite band and with heat loss to a surrounding medium on the remainder of the cylinder surface are exactly analysed. For the first time the temperature field in a circular cylinder under sudden heat generation over a finite band of the cylinder surface is determined by combined use of Fourier cosine, Laplace transforms in axial position and time, respectively. Secondly it is assumed that the temperature fields in a circular cylinder subjected to heat generation Qsub(i) (i=0, 1, 2, ...) independently over a finite band are given by T 0 (r,z,t), T 1 (r,z,t), T 2 (r,z,t),... respectively. Tsub(i)(r,z,t) indicates the temperature field before the i-th heat generation Qsub(i). The thermal stresses associated with the temperature field described above are analysed by using the Hoyle stress functions. Numerical calculations are carried out for the extensive case of the ratio of the heat-generating length to the diameter of cylinder. It is found that the time in which the maximum stresses occur on the cylinder surface does not depend on the heat-generating length-to-diameter ratio

Transient electro-thermal modeling of bipolar power semiconductor devices

CERN Document Server

Gachovska, Tanya Kirilova; Du, Bin

2013-01-01

This book presents physics-based electro-thermal models of bipolar power semiconductor devices including their packages, and describes their implementation in MATLAB and Simulink. It is a continuation of our first book Modeling of Bipolar Power Semiconductor Devices. The device electrical models are developed by subdividing the devices into different regions and the operations in each region, along with the interactions at the interfaces, are analyzed using the basic semiconductor physics equations that govern device behavior. The Fourier series solution is used to solve the ambipolar diffusio

Study on the Cross Plane Thermal Transport of Polycrystalline Molybdenum Nanofilms by Applying Picosecond Laser Transient Thermoreflectance Method

Directory of Open Access Journals (Sweden)

Tingting Miao

2014-01-01

Full Text Available Thin metal films are widely used as interconnecting wires and coatings in electronic devices and optical components. Reliable thermophysical properties of the films are required from the viewpoint of thermal management. The cross plane thermal transport of four polycrystalline molybdenum nanofilms with different thickness deposited on glass substrates has been studied by applying the picosecond laser transient thermoreflectance technique. The measurement is performed by applying both front pump-front probe and rear pump-front probe configurations with high quality signal. The determined cross plane thermal diffusivity of the Mo films greatly decreases compared to the corresponding bulk value and tends to increase as films become thicker, exhibiting significant size effect. The main mechanism responsible for the thermal diffusivity decrease of the present polycrystalline Mo nanofilms is the grain boundary scattering on the free electrons. Comparing the cross plane thermal diffusivity and inplane electrical conductivity indicates the anisotropy of the transport properties of the Mo films.

Transient behavior of superfluid turbulence in a large channel

International Nuclear Information System (INIS)

Schwarz, K.W.; Rozen, J.R.

1991-01-01

The transient behavior of superfluid turbulence is studied theoretically and experimentally with the aim of understanding the disagreement between vortex-tangle theory and past measurements of free vortex-tangle decay in superfluid 4 He. Scaling theory is extended and large-scale simulations based on the reconnecting-vortex model are carried out. These imply that the Vinen equation should be a reasonable approximation even for rather large transients, and predict definite values for the Vinen parameters. Direct measurements of the vortex-tangle response to a sudden change in the driving velocity are seen to be in reasonable agreement with these predictions. It is found, however, that when the vortex tangle is allowed to decay farther toward zero, it eventually crosses over into a state of anomalously slow decay, which appears to be that observed in previous experiments. We argue that this regime should be interpreted in terms of a coupled-turbulence state in which random superfluid and normal-fluid motion interacts with the vortex tangle, the whole system decaying self-consistently at a rate controlled by the normal-fluid viscosity. Several additional qualitative observations which may be relevant to the question of how the vortex tangle is initiated are also reported

Thermal behavior of natural zeolites

International Nuclear Information System (INIS)

Bish, D.L.

1993-01-01

Thermal behavior of natural zeolites impacts their application and identification and varies significantly from zeolite to zeolite. Zeolites evolve H 2 0 upon heating, but recent data show that distinct ''types'' of water (e.g., loosely bound or tightly bound zeolitic water) do not exist. Rather water is bound primarily to extra-framework cations with a continuum of energies, giving rise to pseudocontinuous loss of water accompanied by a dynamic interaction between remaining H 2 0 molecules and extra-framework cations. These interactions in the channels of zeolites give rise to dehydration dependent on the extra-framework cation, in addition to temperature and water vapor pressure. The dehydration reaction and the extra-framework cation also affect the thermal expansion/contraction. Most zeolites undergo dehydration-induced contractions that may be anisotropic, although minor thermal expansion can be seen with some zeolites. Such contractions can be partially or completely irreversible if they involve modifications of the tetrahedral framework and/or if rehydration is sluggish. Thermally induced structural modifications are also driven initially by dehydration and the concomitant contraction and migration of extra-framework cations. Contraction is accommodated by rotations of structural units and tetrahedral cation-oxygen linkages may break. Thermal reactions that involve breaking of tetrahedral cation-oxygen bonds markedly irreversible and may be kinetically limited, producing large differences between short- and long-term heating

Transient analysis on the SMART-P anticipated transients without scram

International Nuclear Information System (INIS)

Yang, S. H.; Bae, K. H.; Kim, H. C.; Zee, S. Q.

2005-01-01

Anticipated transients without scram (ATWS) are anticipated operational occurrences accompanied by a failure of an automatic reactor trip when required. Although the occurrence probability of the ATWS events is considerably low, these events can result in unacceptable consequences, i.e. the pressurization of the reactor coolant system (RCS) up to an unacceptable range and a core-melting situation. Therefore, the regulatory body requests the installation of a protection system against the ATWS events. According to the request, a diverse protection system (DPS) is installed in the SMART-P (System-integrated Modular Advanced ReacTor-Pilot). This paper presents the results of the transient analysis performed to identify the performance of the SMART-P against the ATWS. In the analysis, the TASS/SMR (Transients And Setpoint Simulation/Small and Medium Reactor) code is applied to identify the thermal hydraulic response of the RCS during the transients

Analytical modeling of pressure transient behavior for coalbed methane transport in anisotropic media

International Nuclear Information System (INIS)

Wang, Lei; Wang, Xiaodong

2014-01-01

Resulting from the nature of anisotropy of coal media, it is a meaningful work to evaluate pressure transient behavior and flow characteristics within coals. In this article, a complete analytical model called the elliptical flow model is established by combining the theory of elliptical flow in anisotropic media and Fick's laws about the diffusion of coalbed methane. To investigate pressure transient behavior, analytical solutions were first obtained through introducing a series of special functions (Mathieu functions), which are extremely complex and are hard to calculate. Thus, a computer program was developed to establish type curves, on which the effects of the parameters, including anisotropy coefficient, storage coefficient, transfer coefficient and rate constant, were analyzed in detail. Calculative results show that the existence of anisotropy would cause great pressure depletion. To validate new analytical solutions, previous results were used to compare with the new results. It is found that a better agreement between the solutions obtained in this work and the literature was achieved. Finally, a case study is used to explain the effects of the parameters, including rock total compressibility coefficient, coal medium porosity and anisotropic permeability, sorption time constant, Langmuir volume and fluid viscosity, on bottom-hole pressure behavior. It is necessary to coordinate these parameters so as to reduce the pressure depletion. (paper)

Learning from anticipated and abnormal plant transients

International Nuclear Information System (INIS)

Varnado, B.

1983-01-01

A report is given of the American Nuclear Society topical meeting on Anticipated and Abnormal Transients in Light Water Reactors held in Jackson, Wyoming in September 1983. Industry involvement in the evaluation of operating experience, human error contributions, transient management, thermal hydraulic modelling, the role of probabilistic risk assessment and the cost of transient incidents are discussed. (U.K.)

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Thermal fatigue behavior of thermal barrier coatings by air plasma spray

Energy Technology Data Exchange (ETDEWEB)

Lee, Han Sang; Kim, Eui Hyun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Lee, Jung Hyuk [Korea Plant Service and Engineering Co. Ltd., Incheon (Korea, Republic of)

2008-06-15

Effects of top coat morphology and thickness on thermal fatigue behavior of Thermal Barrier Coatings (TBC) were investigated in this study. Thermal fatigue tests were conducted on three coating specimens with different top coat morphology and thickness, and then the test data were compared via microstructures, cycles to failure, and fracture surfaces. In the air plasma spray specimens (APS1, APS2), top coat were 200 and 300 {mu}m respectively. The thickness of top coat was about 700 {mu}m in the Perpendicular Cracked Specimen (PCS). Under thermal fatigue condition at 1,100 .deg. C, the cycles to top coat failure of APS1, APS2, and PCS were 350, 560 and 480 cycles, respectively. The cracks were initiated at the interface of top coat and Thermally Grown Oxide (TGO) and propagated into TGO or top coat as the number of thermal fatigue cycles increased. For the PCS specimen, additive cracks were initiated and propagated at the starting points of perpendicular cracks in the top coat. Also, the thickness of TGO and the decrease of aluminium concentration in bond coat do not affect the cycles to failure.

A numerical thermal-hydraulic model to simulate the fast transients in a supercritical water channel subjected to sharp pressure variations

NARCIS (Netherlands)

Dutta, G.; Jiang, J.; Maitri, R.; Zhang, C.

2016-01-01

The present work demonstrates the extension of a thermal-hydraulic model, THRUST, with an objective to simulate the fast transient flow dynamics in a supercritical water channel of circular cross section. THRUST is a 1-D model which solves the nonlinearly coupled mass, axial momentum and energy

Peach Bottom Turbine Trip Simulations with RETRAN Using INER/TPC BWR Transient Analysis Method

International Nuclear Information System (INIS)

Kao Lainsu; Chiang, Show-Chyuan

2005-01-01

The work described in this paper is benchmark calculations of pressurization transient turbine trip tests performed at the Peach Bottom boiling water reactor (BWR). It is part of an overall effort in providing qualification basis for the INER/TPC BWR transient analysis method developed for the Kuosheng and Chinshan plants. The method primarily utilizes an advanced system thermal hydraulics code, RETRAN02/MOD5, for transient safety analyses. Since pressurization transients would result in a strong coupling effect between core neutronic and system thermal hydraulics responses, the INER/TPC method employs the one-dimensional kinetic model in RETRAN with a cross-section data library generated by the Studsvik-CMS code package for the transient calculations. The Peach Bottom Turbine Trip (PBTT) tests, including TT1, TT2, and TT3, have been successfully performed in the plant and assigned as standards commonly for licensing method qualifications for years. It is an essential requirement for licensing purposes to verify integral capabilities and accuracies of the codes and models of the INER/TPC method in simulating such pressurization transients. Specific Peach Bottom plant models, including both neutronics and thermal hydraulics, are developed using modeling approaches and experiences generally adopted in the INER/TPC method. Important model assumptions in RETRAN for the PBTT test simulations are described in this paper. Simulation calculations are performed with best-estimated initial and boundary conditions obtained from plant test measurements. The calculation results presented in this paper demonstrate that the INER/TPC method is capable of calculating accurately the core and system transient behaviors of the tests. Excellent agreement, both in trends and magnitudes between the RETRAN calculation results and the PBTT measurements, shows reliable qualifications of the codes/users/models involved in the method. The RETRAN calculated peak neutron fluxes of the PBTT

A basic research on the transient behavior for a metallic fuel FBR

International Nuclear Information System (INIS)

Baba, Mamoru; Hirano, Go; Kawada, Ken-ichi; Niwa, Hajime

1999-03-01

A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'A basic research on the transient behavior for a metallic fuel FBR'. The results are the following. (1) Target and Results of analysis: The accident initiator considered is a LOF accident without scram. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transition phase. It is shown that the moderator is quite elective to mitigate the accident at the initiation phase. However, it is necessary to analyze the transition phase to know if the re-criticality is totally avoided after the initiation phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large scale material movement in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics, and examined

Coupled Cryogenic Thermal and Electrical Models for Transient Analysis of Superconducting Power Devices with Integrated Cryogenic Systems

Science.gov (United States)

Satyanarayana, S.; Indrakanti, S.; Kim, J.; Kim, C.; Pamidi, S.

2017-12-01

Benefits of an integrated high temperature superconducting (HTS) power system and the associated cryogenic systems on board an electric ship or aircraft are discussed. A versatile modelling methodology developed to assess the cryogenic thermal behavior of the integrated system with multiple HTS devices and the various potential configurations are introduced. The utility and effectiveness of the developed modelling methodology is demonstrated using a case study involving a hypothetical system including an HTS propulsion motor, an HTS generator and an HTS power cable cooled by an integrated cryogenic helium circulation system. Using the methodology, multiple configurations are studied. The required total cooling power and the ability to maintain each HTS device at the required operating temperatures are considered for each configuration and the trade-offs are discussed for each configuration. Transient analysis of temperature evolution in the cryogenic helium circulation loop in case of a system failure is carried out to arrive at the required critical response time. The analysis was also performed for a similar liquid nitrogen circulation for an isobaric condition and the cooling capacity ratio is used to compare the relative merits of the two cryogens.

Anisotropic thermal expansion behaviors of copper matrix in β-eucryptite/copper composite

International Nuclear Information System (INIS)

Wang Lidong; Xue Zongwei; Qiao Yingjie; Fei, W.D.

2012-01-01

Highlights: ► The thermal expansion behaviors of Cu matrix were studied by in situ XRD. ► The expansion of Cu{1 1 1} plane is linear, that of Cu{2 0 0} is nonlinear. ► The anisotropic thermal expansion of Cu is related to the twinning of Cu matrix. ► The twinning of Cu matrix makes the CTE of the composite increasing. - Abstract: A β-eucryptite/copper composite was fabricated by spark plasma sintering process. The thermal expansion behaviors of Cu matrix of the composite were studied by in situ X-ray diffraction during heating process. The results show that Cu matrix exhibits anisotropic thermal expansion behaviors for different crystallographic directions, the expansion of Cu{1 1 1} plane is linear in the temperature range from 20 °C to 300 °C and the expansion of Cu{2 0 0} is nonlinear with a inflection at about 180 °C. The microstructures of Cu matrix before and after thermal expansion testing were investigated using transmission electronic microscope. The anisotropic thermal expansion behavior is related to the deformation twinning formed in the matrix during heating process. At the same time, the deformation twinning of Cu matrix makes the average coefficient of thermal expansion of the composite increase.

SCANAIR: A transient fuel performance code

International Nuclear Information System (INIS)

Moal, Alain; Georgenthum, Vincent; Marchand, Olivier

2014-01-01

Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under

SCANAIR: A transient fuel performance code

Energy Technology Data Exchange (ETDEWEB)

Moal, Alain, E-mail: alain.moal@irsn.fr; Georgenthum, Vincent; Marchand, Olivier

2014-12-15

Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under

A mathematical model for the simulation of thermal transients in the water loop of IPEN

International Nuclear Information System (INIS)

Pontedeiro, A.C.

1980-01-01

A mathematical model for simulation of thermal transients in the water loop at the Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo, Brasil, is developed. The model is based on energy equations applied to the components of the experimental water loop. The non-linear system of first order diferencial equations and of non-linear algebraic equations obtained through the utilization of the IBM 'System/360-Continous System Modeling Program' (CSMP) is resolved. An optimization of the running time of the computer is made and a typical simulation of the water loop is executed. (Author) [pt

Investigation of typicality of non-nuclear rod and fuel-clad gap effect during reflood phase, and development of a FEM thermal transient analysis code HETFEM

International Nuclear Information System (INIS)

Sudoh, Takashi

1981-06-01

The objective of this study are: 1) Evaluate the capability of the electrical heater for simulating the fuel rod during the reflood phase, and 2) To investigate the effect of the clad-fuel gap in the fuel rod on the clad thermal response during the reflood phase. A computer code HETFEM which is the two dimensional transient thermal conductivity analysis code utilized a finite element method is developed for analysing thermal responses of heater and fuel rod. The two kinds of electrical heaters and a fuel rod are calculated with simple boundary conditions. 1) direct heater (former JAERI reflood test heater), 2) indirect heater (FLECHT test heater), 3) fuel rod (15 x 15 type in Westinghouse PWR). The comparison of the clad temperature responses shows the quench time is influenced by the thermal diffusivity and gap conductance. In the conclusion, the ELECHT heater shows atypicality in the clad temperature response and heat releasing rate. But the direct heater responses are similar to those of the fuel rod. For the gap effect on the fuel rod behavior, the lower gap conductance causes sooner quench and less heat releasing rate. This calculation is not considered the precursory cooling which is affected by heat releasing rate at near and below the quench front. Therefore two dimensional calculation with heat transfer related to the local fluid conditions will be needed. (author)

Transient Heat Conduction

DEFF Research Database (Denmark)

Rode, Carsten

1998-01-01

Analytical theory of transient heat conduction.Fourier's law. General heat conducation equation. Thermal diffusivity. Biot and Fourier numbers. Lumped analysis and time constant. Semi-infinite body: fixed surface temperature, convective heat transfer at the surface, or constant surface heat flux...

Transient analysis for Laguna Verde nuclear power plant

International Nuclear Information System (INIS)

Ramos Pablos, J.C. et.al.

1991-01-01

Relationship between transients analysis and safety of Laguna Verde nuclear power plant is described a general panorama of safety thermal limits of a nuclear station, as well as transients classification and events simulation codes are exposed. Activities of a group of transients analysis of electrical research institute are also mentioned (Author)

Thermal-mechanical aspects for radioactive waste storage into underground caverns

International Nuclear Information System (INIS)

Vieira, Alvaro.

1985-12-01

The thermal and mechanical behaviors of rock mass by analytical models, considering transient effects of the heat generation from radioactive wastes, are analysed. The models were applied to Brazilian gneissic type of rock, considering the usual design of vitrified waste cylinders individually installed into conveniently spaced holes. (M.C.K.) [pt

Visual investigation of transient fuel behavior under a rapid heating condition

International Nuclear Information System (INIS)

Saito, Shinzo

1981-10-01

An in-reactor experimental research on fuel behavior under reactivity initiated accident (RIA) conditions is being conducted in the Nuclear Safety Research Reactor (NSRR). The optical system in which a non-browning lens periscope is directly installed in the test section was successfully developed for photographing transient fuel behavior. Several phenomena which had never been revealed before were observed in the slow motion pictures taken in the NSRR experiments which were performed in the water and air environments. As for incipient failure mechanism for an unirradiated fuel rod under RIA conditions, brittle fracture of the cladding during quenching is dominant. However, a split cracking possibly occurs during even red hot state of the cladding. It is considered that the crack is generated by the local internal pressure increase at the specified region blocked up due to the melting of the cladding inner surface. The film boiling is unexpectablly violent specially in the early stage of the transient, and film thickness becomes 5 -- 6 mm at maximum. The observed thick vapor film can not be explained by the conventional theory, but the effect of hydrogen which is produced by Zircaloy-water reaction is reasonably explained to form thick film in the report. The molten fuel was expelled from the cladding in the experiment which was performed in an air environment. The expelled fuel fragmented due to possibly initial motion effect, not mechanical collision effect, because Weber number is smaller than the critical value. (author)

Analysis of the thermal hydraulics and core degradation behavior in the PHEBUS-FPT1 test train with impact/SAMPSON code

International Nuclear Information System (INIS)

Terada, Masafumi; Ikeda, Takashi; Nakahara, Katsuhiko; Shirakawa, Noriyuki; Horie, Hideki; Katsuragi, Kazuyuki; Yamagishi, Makoto; Ito, Takahiro

2003-01-01

As one of the verification studies of SAMPSON code, PHEBUS-FPT1, which is authorized as the International Standard Problem-46, was analyzed about the in-core phenomena with four modules, the molten core relocation analysis (MCRA) module, the fuel rod heat up analysis (FRHA) module, the fission product release analysis (FPRA) module, and the analysis control module (ACM) of SAMPSON. This paper describes the analysis of thermal hydraulics and core degradation behavior in the test train. Two-dimensional version of MCRA models the whole structure of the test train in the cylindrical system, including the fuel bundle and the shroud. FRHA models eighteen irradiated fuel rods, two fresh fuel rods, and one control rod in the center of the bundle. FRHA evaluates the transient behavior of fuel rods and releases failed fuel components to MCRA. MCRA evaluates the fluid dynamics of steam and debris considering the thermal and fluid mechanical interaction between them, and at the same time the thermal interaction between gas/debris and shroud material. By the phase change model of MCRA, molten debris forms debris pool and a part of them possibly freezes on fuel rods or shroud surface, then forms crust. This combination of modules of SAMPSON was proved to be capable for modeling the PHEBUS-FPT1 in-core phenomena sufficiently. The analysis has shown sufficient agreement with test results regarding to steam flow rates at the outlet, reproducing its reduction due to hydrogen generation, steam and shroud temperature, and debris relocation behavior. (author)

Transient analysis of a thermal storage unit involving a phase change material

Science.gov (United States)

Griggs, E. I.; Pitts, D. R.; Humphries, W. R.

1974-01-01

The transient response of a single cell of a typical phase change material type thermal capacitor has been modeled using numerical conductive heat transfer techniques. The cell consists of a base plate, an insulated top, and two vertical walls (fins) forming a two-dimensional cavity filled with a phase change material. Both explicit and implicit numerical formulations are outlined. A mixed explicit-implicit scheme which treats the fin implicity while treating the phase change material explicitly is discussed. A band algorithmic scheme is used to reduce computer storage requirements for the implicit approach while retaining a relatively fine grid. All formulations are presented in dimensionless form thereby enabling application to geometrically similar problems. Typical parametric results are graphically presented for the case of melting with constant heat input to the base of the cell.

Thermal-structural response of EBR-II major components under reactor operational transients

International Nuclear Information System (INIS)

Chang, L.K.; Lee, M.J.

1983-01-01

Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

GAPCON-THERMAL-3 code description

International Nuclear Information System (INIS)

Lanning, D.D.; Mohr, C.L.; Panisko, F.E.; Stewart, K.B.

1978-01-01

GAPCON-3 is a computer program that predicts the thermal and mechanical behavior of an operating fuel rod during its normal lifetime. The code calculates temperatures, dimensions, stresses, and strains for the fuel and the cladding in both the radial and axial directions for each step of the user specified power history. The method of weighted residuals is for the steady state temperature calculation, and is combined with a finite difference approximation of the time derivative for transient conditions. The stress strain analysis employs an iterative axisymmetric finite element procedure that includes plasticity and creep for normal and pellet-clad mechanical interaction loads. GAPCON-3 can solve steady state and operational transient problems. Comparisons of GAPCON-3 predictions to both closed form analytical solutions and actual inpile instrumented fuel rod data have demonstrated the ability of the code to calculate fuel rod behavior. GAPCON-3 features a restart capability and an associated plot package unavailable in previous GAPCON series codes

GAPCON-THERMAL-3 code description

Energy Technology Data Exchange (ETDEWEB)

Lanning, D.D.; Mohr, C.L.; Panisko, F.E.; Stewart, K.B.

1978-01-01

GAPCON-3 is a computer program that predicts the thermal and mechanical behavior of an operating fuel rod during its normal lifetime. The code calculates temperatures, dimensions, stresses, and strains for the fuel and the cladding in both the radial and axial directions for each step of the user specified power history. The method of weighted residuals is for the steady state temperature calculation, and is combined with a finite difference approximation of the time derivative for transient conditions. The stress strain analysis employs an iterative axisymmetric finite element procedure that includes plasticity and creep for normal and pellet-clad mechanical interaction loads. GAPCON-3 can solve steady state and operational transient problems. Comparisons of GAPCON-3 predictions to both closed form analytical solutions and actual inpile instrumented fuel rod data have demonstrated the ability of the code to calculate fuel rod behavior. GAPCON-3 features a restart capability and an associated plot package unavailable in previous GAPCON series codes.

Simplified Transient Hot-Wire Method for Effective Thermal Conductivity Measurement in Geo Materials: Microstructure and Saturation Effect

Directory of Open Access Journals (Sweden)

B. Merckx

2012-01-01

Full Text Available The thermal conductivity measurement by a simplified transient hot-wire technique is applied to geomaterials in order to show the relationships which can exist between effective thermal conductivity, texture, and moisture of the materials. After a validation of the used “one hot-wire” technique in water, toluene, and glass-bead assemblages, the investigations were performed (1 in glass-bead assemblages of different diameters in dried, water, and acetone-saturated states in order to observe the role of grain sizes and saturation on the effective thermal conductivity, (2 in a compacted earth brick at different moisture states, and (3 in a lime-hemp concrete during 110 days following its manufacture. The lime-hemp concrete allows the measurements during the setting, desiccation and carbonation steps. The recorded Δ/ln( diagrams allow the calculation of one effective thermal conductivity in the continuous and homogeneous fluids and two effective thermal conductivities in the heterogeneous solids. The first one measured in the short time acquisitions (<1 s mainly depends on the contact between the wire and grains and thus microtexture and hydrated state of the material. The second one, measured for longer time acquisitions, characterizes the mean effective thermal conductivity of the material.

Evaluation of transient natural circulation behavior during accident in low power/shutdown condition of YGN units 3/4

International Nuclear Information System (INIS)

Bang, Young Seok; Kim, Kap; Seul, Kwang Won; Kim, Hho Jung

1997-01-01

A transient natural circulation behavior during a LOCA at hot-standby operation is evaluated for YGN Units 3/4. The plant initial condition is determined within the EOP limitation as suitable to hot-standby mode and the transient scenario is prepared as relevant to evaluation of transient natural circulation. A 0.4% cold leg break with loss of off-site power is calculated with RELAP5/MOD3.2, whose predictability has been verified for SBLOCA natural circulation test, S-NC-8B. Through one hour transient analysis, it is found that the plant has its own decay heat removal capability by natural circulation following a LOCA at hot-standby mode. Additional calculation is performed to investigate an effect of HPSI flow on natural circulation

Transient three-dimensional thermal-hydraulic analysis of nuclear reactor fuel rod arrays: general equations and numerical scheme

International Nuclear Information System (INIS)

Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.; Hughes, E.D.; Solbrig, C.W.

1975-11-01

A mathematical model and a numerical solution scheme for thermal-hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media

Simulation of protected and unprotected loss of flow transients in a WWER-1000 reactor based on the Drift-Flux model

Energy Technology Data Exchange (ETDEWEB)

Baghban, Ghonche [Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of). Nuclear Science and Technology Research Inst.; Shayesteh, Mohsen [Imam Hussein Univ., Tehran (Iran, Islamic Republic of). Dept. of Physics; Bahonar, Majid [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

2017-03-15

In view of the importance of studying coolant transient behavior in a nuclear reactor, this work is devoted to the thermal-hydraulic analysis of protected and unprotected loss of flow transients in a WWER-1000 reactor. A series of corresponding mathematical and physical models based on the four-equation Drift-Flux model has been applied. Based on a multi-channel approach, the core has been divided into different regions. Each region has different characteristics as represented in a single fuel pin with its associated coolant channel. Appropriate initial and boundary conditions have been considered and two situations of tripping four and two primary pumps in a protected core in addition to situation of tripping all four pumps in an unprotected core have been analyzed. For each transient, a full range of thermal-hydraulic parameters has been obtained. For verification of the proposed model, the results have been compared with those of the RELAP5/MOD3 and Bushehr nuclear power plant Final Safety Analysis Report (FSAR). A good agreement between results has been attained for the aforementioned transients.

Overpower transient in the first wall cooling system of NET/ITER

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1993-09-01

The overpower transient from a plasma power excursion. The overpower transient considered in this report results from a postulated linear increase of the plasma power from the nominal generated power to four times this nominal power in 30 s. The Next European Torus (NET) design or the International Thermonuclear Experimental Reactor (ITER) design will be cooled by a number of separate cooling systems. The most important cooling systems are: The first wall cooling system, the blanket cooling system, the divertor cooling system, and the shield cooling system. In this report, the thermal-hydraulic analysis of the above-mentioned overpower transient will be presented for the first wall cooling system of NET/ITER. During overpower transients, the fusion power will increase to less than four times the nominal power. For this reason, the overpower transient considered in this report is the worst case scenario. The analysis of the thermal-hydraulic system behaviour during the considered overpower transient has been performed for a coolant temperature of 333 K (60 C) in the first wall inlet manifolds and 433 K (160 C) in the first wall outlet manifolds. The analysis has been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analysis, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall. (orig.)

The PARET code and the analysis of the SPERT I transients

Energy Technology Data Exchange (ETDEWEB)

Woodruff, William L [Argonne National Laboratory, Argonne (United States)

1983-09-01

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients.

The PARET code and the analysis of the SPERT I transients

International Nuclear Information System (INIS)

Woodruff, William L.

1983-01-01

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients

Studies of implicit and explicit solution techniques in transient thermal analysis of structures

International Nuclear Information System (INIS)

Adelman, H.M.; Haftka, R.T.; Robinson, J.C.

1982-08-01

Studies aimed at an increase in the efficiency of calculating transient temperature fields in complex aerospace vehicle structures are reported. The advantages and disadvantages of explicit and implicit algorithms are discussed and a promising set of implicit algorithms with variable time steps, known as GEARIB, is described. Test problems, used for evaluating and comparing various algorithms, are discussed and finite element models of the configurations are described. These problems include a coarse model of the Space Shuttle wing, an insulated frame test article, a metallic panel for a thermal protection system, and detailed models of sections of the Space Shuttle wing. Results generally indicate a preference for implicit over explicit algorithms for transient structural heat transfer problems when the governing equations are stiff (typical of many practical problems such as insulated metal structures). The effects on algorithm performance of different models of an insulated cylinder are demonstrated. The stiffness of the problem is highly sensitive to modeling details and careful modeling can reduce the stiffness of the equations to the extent that explicit methods may become the best choice. Preliminary applications of a mixed implicit-explicit algorithm and operator splitting techniques for speeding up the solution of the algebraic equations are also described

Studies of implicit and explicit solution techniques in transient thermal analysis of structures

Science.gov (United States)

Adelman, H. M.; Haftka, R. T.; Robinson, J. C.

1982-01-01

Studies aimed at an increase in the efficiency of calculating transient temperature fields in complex aerospace vehicle structures are reported. The advantages and disadvantages of explicit and implicit algorithms are discussed and a promising set of implicit algorithms with variable time steps, known as GEARIB, is described. Test problems, used for evaluating and comparing various algorithms, are discussed and finite element models of the configurations are described. These problems include a coarse model of the Space Shuttle wing, an insulated frame tst article, a metallic panel for a thermal protection system, and detailed models of sections of the Space Shuttle wing. Results generally indicate a preference for implicit over explicit algorithms for transient structural heat transfer problems when the governing equations are stiff (typical of many practical problems such as insulated metal structures). The effects on algorithm performance of different models of an insulated cylinder are demonstrated. The stiffness of the problem is highly sensitive to modeling details and careful modeling can reduce the stiffness of the equations to the extent that explicit methods may become the best choice. Preliminary applications of a mixed implicit-explicit algorithm and operator splitting techniques for speeding up the solution of the algebraic equations are also described.

Advanced methods for BWR transient and stability analysis

Energy Technology Data Exchange (ETDEWEB)

Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)

2008-07-01

The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

Advanced methods for BWR transient and stability analysis

International Nuclear Information System (INIS)

Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

2008-01-01

The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

NALAP: an LMFBR system transient code

International Nuclear Information System (INIS)

Martin, B.A.; Agrawal, A.K.; Albright, D.C.; Epel, L.G.; Maise, G.

1975-07-01

NALAP is a LMFBR system transient code. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic response of sodium cooled fast breeder reactors when subjected to postulated accidents such as a massive pipe break as well as a variety of other upset conditions that do not disrupt the system geometry. Various components of the plant are represented by control volumes. These control volumes are connected by junctions some of which may be leak or fill junctions. The fluid flow equations are modeled as compressible, single-stream flow with momentum flux in one dimension. The transient response is computed by integrating the thermal-hydraulic conservation equations from user-initialized operating conditions by an implicit numerical scheme. Point kinetics approximation is used to represent the time dependent heat generation in the reactor core

Modelling of Thermal Behavior of Borehole Heat Exchangers of Geothermal Heat Pump Heating Systems

Directory of Open Access Journals (Sweden)

Gornov V.F.

2016-01-01

Full Text Available This article reports results of comparing the accuracy of the software package “INSOLAR.GSHP.12”, modeling non-steady thermal behavior of geothermal heat pump heating systems (GHCS and of the similar model “conventional” using finite difference methods for solving spatial non-steady problems of heat conductivity. The software package is based on the method of formulating mathematical models of thermal behavior of ground low-grade heat collection systems developed by INSOLAR group of companies. Equations of mathematical model of spatial non-steady thermal behavior of ground mass of low-grade heat collection system obtained by the developed method have been solved analytically that significantly reduced computing time spent by the software complex “INSOLAR.GSHP.12” for calculations. The method allows to turn aside difficulties associated with information uncertainty of mathematical models of the ground thermal behavior and approximation of external factors affecting the ground. Use of experimentally obtained information about the ground natural thermal behavior in the software package allows to partially take into account the whole complex of factors (such as availability of groundwater, their velocity and thermal behavior, structure and arrangement of ground layers, the Earth’s thermal background, precipitation, phase transformations of moisture in the pore space, and more, significantly influencing the formation of thermal behavior of the ground mass of a low-grade geothermal heat collection system. Numerical experiments presented in the article confirmed the high convergence of the results obtained through the software package “INSOLAR.GSHP.12” with solutions obtained by conventional finite-difference methods.

Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results

International Nuclear Information System (INIS)

Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A.; Beelman, R.J.

1987-11-01

The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs

Effects of load and thermal histories on mechanical behavior of materials; Proceedings of the Symposium, Denver, CO, Feb. 25, 26, 1987

Energy Technology Data Exchange (ETDEWEB)

Liaw, P.K.; Nicholas, T.

1987-01-01

This volume includes topics on fatigue crack propagation; isothermal and thermal-mechanical fatigue; and microstructure, fracture, and damage. Papers are presented on transients in fatigue crack growth, elevated-temperature fatigue crack propagation, the role of crack closure in crack retardation in P/M and I/M aluminum alloys, the acoustic interrogation of fatigue overload effects, and the effects of frequency and environment on crack growth in Inconel 718. Special attention is given to isothermal fatigue failure mechanisms in low-tin lead-based solder, the stress and strain controlled low-cycle fatigue of Pb-Sn solder for electronic packaging applications, load sequence effects on the deformation of isolated microplastic grains, and thermal fatigue of stainless steel. Other papers are on the influence of thermal aging on the creep crack growth behavior of a Cr-Mo steel, the effect of cyclic loading on the fracture toughness of a modified 4340 steel, and the effects of hot rolling condition and boron microalloying on phase transformation and microstructure in niobium-bearing interstitial free steel.

FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code

International Nuclear Information System (INIS)

Valtonen, Keijo; Hamalainen, Anitta; Cunningham, Mitchel E.

2002-01-01

The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.

FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code

Energy Technology Data Exchange (ETDEWEB)

Valtonen, Keijo (Radiation and Nuclear Safety Authority, Finland); Hamalainen, Anitta (VTT Energy, Finland); Cunningham, Mitchel E.(BATTELLE (PACIFIC NW LAB))

2002-05-01

The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.

Investigation of Thermal and Vacuum Transients on the LHC Prototype Magnet String

CERN Document Server

Cruikshank, P; Riddone, G; Tavian, L

1996-01-01

The prototype magnet string, described in a companion paper, is a full-scale working model of a 50-m length of the future Large Hadron Collider (LHC), CERN's new accelerator project, which will use high-field superconducting magnets operating below 2 K in superfluid helium. As such, it provides an excellent test bed for practising standard operating modes of LHC insulation vacuum and cryogenics, as well as for experimentally assessing accidental behaviour and failure modes, and thus verifying design calculations. We present experimental investigation of insulation vacuum pumpdown, magnet forced-flow cooldown and warmup, and evolution of residual vacuum pressures and temperatures in natural warmup, as well as catastrophic loss of insulation vacuum. In all these transient modes, experimental results are compared with simulated behaviour, using a non-linear, one-dimensional thermal model of the magnet string.

Pressurized-thermal-shock technology

International Nuclear Information System (INIS)

Dickson, T.L.

1991-01-01

It was recognized at the time the original Issues on Pressurized Thermal Shock (IPTS) studies were conducted that distinct vertical plumes of cooling water form beneath the cold leg inlet nozzles during those particular transients that exhibit fluid/thermal stratification. The formation of these plumes (referred to as thermal streaming) induces a time-dependent circumferential temperature variation on the inner surface of the Reactor Pressure Vessel (RPV) wall that creates an axial stress component. This axial stress component is in addition to the axial stress components induced by time-dependent radial temperature variation through the wall thickness and the time-dependent pressure transient. This additional axial stress component will result in a larger axial stress resultant that results in a larger stress-intensity factor acting on circumferential flaws, thus reducing the fracture margin for circumferential flaws. Although this was recognized at the time of the original IPTS study, the contribution appeared to be relatively small; therefore, it was neglected. The original IPTS studies were performed with OCA-P, a computer program developed at ORNL to analyze the cleavage fracture response of a nuclear RPV subjected to PTS loading. OCA-P is a one-dimensional (1-D) finite-element code that analyzes the stresses and stress-intensity factors (axial and tangential) resulting from the pressure and the radial temperature variation through the wall thickness only. The HSST Program is investigating the potential effects of thermal-streaming-induced stresses in circumferential welds on the reactor vessel PTS analyses. The initial phase of this investigation focused on an evaluation of the available thermal-hydraulic data and analyses results. The objective for the initial phase of the investigation is to evaluate thermal-streaming behavior under conditions relevant to the operation of U.S. PWRs and chracterize any predicted thermal-streaming plumes

Comparison of 'system thermal-hydraulics-3 dimensional reactor kinetics' coupled calculations using the MARS 1D and 3D modules and the MASTER code

International Nuclear Information System (INIS)

Jung, J. J.; Joo, H. K.; Lee, W. J.; Ji, S. K.; Jung, B. D.

2002-01-01

KAERI has developed the coupled 'system thermal-hydraulics - 3 dimensional reactor kinetics' code, MARS/MASTER since 1998. However, there is a limitation in the existing MARS/MASTER code; that is, to perform the coupled calculations using MARS/MASTER, we have to utilize the hydrodynamic model and the heat structure model of the MARS '3D module'. In some transients, reactor kinetics behavior is strongly multi-dimensional, but core thermal-hydraulic behavior remains in one-dimensional manner. For efficient analysis of such transients, we coupled the MARS 1D module with MASTER. The new feature has been assessed by the 'OECD NEA Main Steam Line Break (MSLB) benchmark exercise III' simulations

Thermal behavior of spatial structures under solar irradiation

International Nuclear Information System (INIS)

Liu, Hongbo; Liao, Xiangwei; Chen, Zhihua; Zhang, Qian

2015-01-01

The temperature, particularly the non-uniform temperature under solar irradiation, is the main load for large-span steel structures. Due the shortage of in-site temperature test in previous studies, an in-site test was conducted on the large-span steel structures under solar irradiation, which was covered by glass roof and light roof, to gain insight into the temperature distribution of steel members under glass roof or light roof. A numerical method also was presented and verified to forecast the temperature of steel member under glass roof or light roof. Based on the on-site measurement and numerical analyses conducted, the following conclusions were obtained: 1) a remarkable temperature difference exists between the steel member under glass roof and that under light roof, 2) solar irradiation has a significant effect on the temperature distribution and thermal behavior of large-span spatial structures, 3) negative thermal load is the controlling factor for member stress, and the positive thermal load is the controlling factor for nodal displacement. - Highlights: • Temperature was measured for a steel structures under glass roof and light roof. • Temperature simulation method was presented and verified. • The thermal behavior of steel structures under glass or light roof was presented

Modelling of the thermal parameters of high-power linear laser-diode arrays. Two-dimensional transient model

International Nuclear Information System (INIS)

Bezotosnyi, V V; Kumykov, Kh Kh

1998-01-01

A two-dimensional transient thermal model of an injection laser is developed. This model makes it possible to analyse the temperature profiles in pulsed and cw stripe lasers with an arbitrary width of the stripe contact, and also in linear laser-diode arrays. This can be done for any durations and repetition rates of the pump pulses. The model can also be applied to two-dimensional laser-diode arrays operating quasicontinuously. An analysis is reported of the influence of various structural parameters of a diode array on the thermal regime of a single laser. The temperature distributions along the cavity axis are investigated for different variants of mounting a crystal on a heat sink. It is found that the temperature drop along the cavity length in cw and quasi-cw laser diodes may exceed 20%. (lasers)

Transient response of rotating laminated functionally graded cylindrical shells in thermal environment

International Nuclear Information System (INIS)

Malekzadeh, P.; Heydarpour, Y.; Haghighi, M.R. Golbahar; Vaghefi, M.

2012-01-01

Based on the elasticity theory, the transient analysis of dynamically pressurized rotating multi-layered functionally graded (FG) cylindrical shells in thermal environment is presented. The variations of the field variables across the shell thickness are accurately modeled by dividing the shell into a set of co-axial mathematical layers in the radial direction. The initial thermo-mechanical stresses are obtained by solving the thermoelastic equilibrium equations. The differential quadrature method and Newmark's time integration scheme are employed to discretize the obtained governing equations of each mathematical layer. After performing the convergence and comparison studies, parametric studies for two common types of FG sandwich shells, namely, the shell with homogeneous inner/outer layers and FG core and the shell with FG inner/outer layers and homogeneous core are carried out. The influences of the temperature dependence of material properties, material graded index, the convective heat transfer coefficient, the angular velocity, the boundary condition and the geometrical parameters (length and thickness to outer radius ratios) on the dynamic response of the FG shells are investigated. Highlights: ► As a first endeavor, transient analysis of rotating laminated functionally graded cylinders. ► Employing an elasticity based discrete layer-differential quadrature method. ► Evaluating and including the initial thermo-mechanical stresses accurately. ► Considering the temperature-dependence of the material properties. ► Presenting some new results, which can be used as benchmark solution for future works.

Improved numerical algorithm and experimental validation of a system thermal-hydraulic/CFD coupling method for multi-scale transient simulations of pool-type reactors

International Nuclear Information System (INIS)

Toti, A.; Vierendeels, J.; Belloni, F.

2017-01-01

Highlights: • A system thermal-hydraulic/CFD coupling methodology is proposed for high-fidelity transient flow analyses. • The method is based on domain decomposition and implicit numerical scheme. • A novel interface Quasi-Newton algorithm is implemented to improve stability and convergence rate. • Preliminary validation analyses on the TALL-3D experiment. - Abstract: The paper describes the development and validation of a coupling methodology between the best-estimate system thermal-hydraulic code RELAP5-3D and the CFD code FLUENT, conceived for high fidelity plant-scale safety analyses of pool-type reactors. The computational tool is developed to assess the impact of three-dimensional phenomena occurring in accidental transients such as loss of flow (LOF) in the research reactor MYRRHA, currently in the design phase at the Belgian Nuclear Research Centre, SCK• CEN. A partitioned, implicit domain decomposition coupling algorithm is implemented, in which the coupled domains exchange thermal-hydraulics variables at coupling boundary interfaces. Numerical stability and interface convergence rates are improved by a novel interface Quasi-Newton algorithm, which is compared in this paper with previously tested numerical schemes. The developed computational method has been assessed for validation purposes against the experiment performed at the test facility TALL-3D, operated by the Royal Institute of Technology (KTH) in Sweden. This paper details the results of the simulation of a loss of forced convection test, showing the capability of the developed methodology to predict transients influenced by local three-dimensional phenomena.

Application of RELAP5/MOD3.3 to Calculate Thermal Hydraulic Behavior of the Pressurizer Safety Valve Performance Test Facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Hyun; Kim, Young Ae; Oh, Seung Jong; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

2007-10-15

The increase of the acceptance tolerance of Pressurizer Safety Valve (PSV) test is vital for the safe operation of nuclear power plants because the frequent tests may make the valves decrepit and become a cause of leak. Recently, Korea Hydro and Nuclear Power Company (KHNP) is building a PSV performance test facility to provide the technical background data for the relaxation of the acceptance tolerance of PSV including the valve pop-up characteristics and the loop seal dynamics (if the plant has the loop seal in the upstream of PSV). The discharge piping and supports must be designed to withstand severe transient hydrodynamic loads when the safety valve actuates. The evaluation of hydrodynamic loads is a two-step process: first the thermal hydraulic behavior in the piping must be defined, and then the hydrodynamic loads are calculated from the thermal hydraulic parameters such as pressure and mass flow. The hydrodynamic loads are used as input to the structural analysis.

The use of computer-generated color graphic images for transient thermal analysis. [for hypersonic aircraft

Science.gov (United States)

Edwards, C. L. W.; Meissner, F. T.; Hall, J. B.

1979-01-01

Color computer graphics techniques were investigated as a means of rapidly scanning and interpreting large sets of transient heating data. The data presented were generated to support the conceptual design of a heat-sink thermal protection system (TPS) for a hypersonic research airplane. Color-coded vector and raster displays of the numerical geometry used in the heating calculations were employed to analyze skin thicknesses and surface temperatures of the heat-sink TPS under a variety of trajectory flight profiles. Both vector and raster displays proved to be effective means for rapidly identifying heat-sink mass concentrations, regions of high heating, and potentially adverse thermal gradients. The color-coded (raster) surface displays are a very efficient means for displaying surface-temperature and heating histories, and thereby the more stringent design requirements can quickly be identified. The related hardware and software developments required to implement both the vector and the raster displays for this application are also discussed.

Transient thermal effects in Alpine permafrost

Directory of Open Access Journals (Sweden)

J. Noetzli

2009-04-01

Full Text Available In high mountain areas, permafrost is important because it influences the occurrence of natural hazards, because it has to be considered in construction practices, and because it is sensitive to climate change. The assessment of its distribution and evolution is challenging because of highly variable conditions at and below the surface, steep topography and varying climatic conditions. This paper presents a systematic investigation of effects of topography and climate variability that are important for subsurface temperatures in Alpine bedrock permafrost. We studied the effects of both, past and projected future ground surface temperature variations on the basis of numerical experimentation with simplified mountain topography in order to demonstrate the principal effects. The modeling approach applied combines a distributed surface energy balance model and a three-dimensional subsurface heat conduction scheme. Results show that the past climate variations that essentially influence present-day permafrost temperatures at depth of the idealized mountains are the last glacial period and the major fluctuations in the past millennium. Transient effects from projected future warming, however, are likely larger than those from past climate conditions because larger temperature changes at the surface occur in shorter time periods. We further demonstrate the accelerating influence of multi-lateral warming in steep and complex topography for a temperature signal entering the subsurface as compared to the situation in flat areas. The effects of varying and uncertain material properties (i.e., thermal properties, porosity, and freezing characteristics on the subsurface temperature field were examined in sensitivity studies. A considerable influence of latent heat due to water in low-porosity bedrock was only shown for simulations over time periods of decades to centuries. At the end, the model was applied to the topographic setting of the Matterhorn

RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part I: Theory and description of model capabilities

Science.gov (United States)

Raffray, A. René; Federici, Gianfranco

1997-04-01

RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under 'slow' high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case.

RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Pt. I. Theory and description of model capabilities

International Nuclear Information System (INIS)

Raffray, A.R.; Federici, G.

1997-01-01

For pt.II see ibid., p.101-30, 1997. RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under 'slow' high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case. (orig.)

Design and Development of Embedded System for the Measurement of Thermal Conductivity of Liquids by Transient Hot Wire Method

Directory of Open Access Journals (Sweden)

Nagamani GOSALA

2011-06-01

Full Text Available Thermal conductivity of polymers is an important property for both polymer applications and processing industry. The successful application of thermal insulating fluids in the last several years has demonstrated that such fluids can effectively control the heat loss. Understanding and controlling the thermal environment for oilfield operations has been a concern and research topic. As a consequence of this trend, there is huge demand for new methods of instrumentation to evaluate the performance of material properties and characterization. The main aim of the present study is the development of hardware and software for measuring the thermal conductivity of liquids using transient hot wire method. Because of the relatively short experimental times and large amounts of parametric data involved in the measurement process, embedded control of the measurement is essential. The experimental implementation requires a suitable temperature sensing, automatic control, data acquisition, and data analysis systems accomplished using an embedded system that has been built around the ARM LPC 2103 mixed signal controller.

Thermal Behavior of Doubly-Fed Induction Generator Wind Turbine System during Balanced Grid Fault

DEFF Research Database (Denmark)

Zhou, Dao; Blaabjerg, Frede; Lau, Mogens

2014-01-01

Ride-through capabilities of the doubly-fed induction generator (DFIG) during grid fault have been studied a lot. However, the thermal performance of the power device during this transient period is seldom investigated. In this paper, the dynamic model for the DFIG and the influence of the rotor...

Transient Analysis and Dosimetry of the Tokaimura Criticality Incident

International Nuclear Information System (INIS)

Pain, Christopher C.; Oliveira, Cassiano R.E. de; Goddard, Antony J. H.; Eaton, Matthew D.; Gundry, Sarah; Umpleby, Adrian P.

2003-01-01

This paper describes research on the application of the finite element transient criticality (FETCH) code to modeling and neutron dosimetry of the Tokaimura criticality incident. FETCH has been developed to model criticality transients in single and multiphase media and is applied here to fissile solution transient criticality. Since the initial transient behavior has different time scales and physics to the longer transient behavior, the transient modeling is divided into two parts: modeling the initial transient over a time scale of seconds in which radiolytic gases and free-surface sloshing play an important role in the transient - this provides information about the dose to workers; and modeling the long-term transient behavior following the initial transient that has a time scale over hours.The neutron dosimetry of worker A who received the largest dose during the Tokaimura criticality incident is also investigated here. This dose was received mainly in the first few seconds of the ensuing nuclear criticality transient. In addition to the multiorgan dosimetry of worker A, this work provides a method of helping to evaluate the yield in the initial phase of the criticality incident; it also shows how kinetic simulations can be calibrated so that they can be applied to investigate the physics behind the incident

Pseudo-Bond Graph model for the analysis of the thermal behavior of buildings

Directory of Open Access Journals (Sweden)

Merabtine Abdelatif

2013-01-01

Full Text Available In this work, a simplified graphical modeling tool, which in some extent can be considered in halfway between detailed physical and Data driven dynamic models, has been developed. This model is based on Bond Graphs approach. This approach has the potential to display explicitly the nature of power in a building system, such as a phenomenon of storage, processing and dissipating energy such as Heating, Ventilation and Air-Conditioning (HVAC systems. This paper represents the developed models of the two transient heat conduction problems corresponding to the most practical cases in building envelope, such as the heat transfer through vertical walls, roofs and slabs. The validation procedure consists of comparing the results obtained with this model with analytical solution. It has shown very good agreement between measured data and Bond Graphs model simulation. The Bond Graphs technique is then used to model the building dynamic thermal behavior over a single zone building structure and compared with a set of experimental data. An evaluation of indoor temperature was carried out in order to check our Bond Graphs model.

Modeling of District Heating Networks for the Purpose of Operational Optimization with Thermal Energy Storage

Science.gov (United States)

Leśko, Michał; Bujalski, Wojciech

2017-12-01

The aim of this document is to present the topic of modeling district heating systems in order to enable optimization of their operation, with special focus on thermal energy storage in the pipelines. Two mathematical models for simulation of transient behavior of district heating networks have been described, and their results have been compared in a case study. The operational optimization in a DH system, especially if this system is supplied from a combined heat and power plant, is a difficult and complicated task. Finding a global financial optimum requires considering long periods of time and including thermal energy storage possibilities into consideration. One of the most interesting options for thermal energy storage is utilization of thermal inertia of the network itself. This approach requires no additional investment, while providing significant possibilities for heat load shifting. It is not feasible to use full topological models of the networks, comprising thousands of substations and network sections, for the purpose of operational optimization with thermal energy storage, because such models require long calculation times. In order to optimize planned thermal energy storage actions, it is necessary to model the transient behavior of the network in a very simple way - allowing for fast and reliable calculations. Two approaches to building such models have been presented. Both have been tested by comparing the results of simulation of the behavior of the same network. The characteristic features, advantages and disadvantages of both kinds of models have been identified. The results can prove useful for district heating system operators in the near future.

The assessment of RELAP5/MOD2 based on pressurizer transient experiments

International Nuclear Information System (INIS)

Xue Hanjun; Tanrikut, A.; Menzel, R.

1992-03-01

Two typical experiments have been performed in Chinese test facility under full pressure load corresponding to typical PWRs, 1) dynamic behavior of pressurizer due to relief valve operations (Case-I) is extremely important in transients and accident conditions regarding depressurization of PWR primary system; 2) Outsurge/Insurge operation is one of the transient which is often encountered and experienced in pressurizer systems due to pressure transients in primary system of PWRs. The simulation capability of RELAP5/MOD2 is good in comparison to experimental results. The physical models (such as interface model, stratification model), playing a major role in such simulation, seems to be realistic. The effect of realistic valve modeling in depressurization simulation is extremely important. Sufficient data for relief valve (the dynamic characteristics of valve) play a major role. The time dependent junction model and the trip valve model with a reduced discharge coefficient of 0.2 give better predictions in agreement with the experiment data while the trip valve models with discharge coefficient 1.0 yield overdepressurization. The simulation of outsurge/insurge transient yields satisfactory results. The thermal non-equilibrium model is important with respect to simulation of complicated physical phenomena in outsurge/insurge transient but has a negligible effect upon the depressurization simulation. (orig./HP)

Radio and white-light observations of coronal transients

International Nuclear Information System (INIS)

Dulk, G.A.

1980-01-01

Optical, radio and X-ray evidence of violent mass motions in the corona has existed for some years but only recently have the form, nature, frequency and implication of the transients become obvious. The author reviews the observed properties of coronal transients, concentrating on the white-light and radio manifestations. The classification according to speeds seems to be meaningful, with the slow transients having thermal emissions at radio wavelengths and the fast ones non-thermal. The possible mechanisms involved in the radio bursts are discussed and the estimates of various forms of energy are reviewed. It appears that the magnetic energy transported from the Sun by the transient exceeds that of any other form, and that magnetic forces dominate in the dynamics of the motions. The conversion of magnetic energy into mechanical energy, by expansion of the fields, provides a possible driving force for the coronal and interplanetary shock waves. (Auth.)

Development and assessment of a sub-channel code applicable for trans-critical transient of SCWR

International Nuclear Information System (INIS)

Liu, X.J.; Yang, T.; Cheng, X.

2013-01-01

Highlights: • A new sub-channel code COBRA-SC for SCWR is developed. • Pseudo two-phase method is employed to realize trans-critical transient calculation. • Good suitability of COBRA-SC is demonstrated by preliminary assessment. • The calculation results of COBRA-SC agree well with ATHLET code. -- Abstract: In the last few years, extensive R and D activities have been launched covering various aspects of supercritical water-cooled reactor (SCWR), especially the thermal-hydraulic analysis. Sub-channel code plays an indispensable role to predict the detail thermal-hydraulic behavior of the SCWR fuel assembly. This paper develops a new version of sub-channel code COBRA-SC based on the previous COBRA-IV code. The supercritical water property and heat transfer/pressure drop correlations under supercritical pressure are implemented to this code. Moreover, in order to simulate the trans-critical transient (the pressure undergo a decrease from the supercritical pressure to the subcritical pressure), pseudo two-phase method is employed in COBRA-SC code. This work is completed by introduction of a virtual two-phase region near the pseudo-critical line. A smooth transition of void fraction can be realized. In addition, several heat transfer correlations right underneath the critical point are introduced into this code to capture the heat transfer behavior during the trans-critical transient. Some experimental data from simple geometry, e.g. the single tube, small rod bundle, is used to validate and evaluate this new developed COBRA-SC code. The predicted results show a good agreement with the experimental data, demonstrating good feasibility of this code for SCWR condition. A code to code comparison between COBRA-SC and ATHLET for a blowdown transient of a small fuel assembly is also presented and discussed in this paper

A novel approach to model the transient behavior of solid-oxide fuel cell stacks

Science.gov (United States)

Menon, Vikram; Janardhanan, Vinod M.; Tischer, Steffen; Deutschmann, Olaf

2012-09-01

This paper presents a novel approach to model the transient behavior of solid-oxide fuel cell (SOFC) stacks in two and three dimensions. A hierarchical model is developed by decoupling the temperature of the solid phase from the fluid phase. The solution of the temperature field is considered as an elliptic problem, while each channel within the stack is modeled as a marching problem. This paper presents the numerical model and cluster algorithm for coupling between the solid phase and fluid phase. For demonstration purposes, results are presented for a stack operated on pre-reformed hydrocarbon fuel. Transient response to load changes is studied by introducing step changes in cell potential and current. Furthermore, the effect of boundary conditions and stack materials on response time and internal temperature distribution is investigated.

Intermediate size inducer pump - structural analysis and transient deformation studies

International Nuclear Information System (INIS)

Cheng, T.K.; Nishizaka, J.N.

1979-05-01

This report summarizes the structural and thermal transient deformation analysis of the Intermediate Size Inducer Pump. The analyses were performed in accordance to the requirements of N266ST310001, the specification for the ISIP. Results of stress analysis indicate that the thermal transient stress and strain are within the stress strain limits of RDT standard F9-4 which was used as a guide

Mathematical model validation of a thermal architecture system connecting east/west radiators by flight data

International Nuclear Information System (INIS)

Torres, Alejandro; Mishkinis, Donatas; Kaya, Tarik

2014-01-01

A novel satellite thermal architecture connecting the east and west radiators of a geostationary telecommunication satellite via loop heat pipes (LHPs) is flight tested on board the satellite Hispasat 1E. The LHP operating temperature is regulated by using pressure regulating valves (PRVs). The flight data demonstrated the successful operation of the proposed concept. A transient numerical model specifically developed for the design of this system satisfactorily simulated the flight data. The validated mathematical model can be used to design and analyze the thermal behavior of more complex architectures. - Highlights: •A novel spacecraft thermal control architecture is presented. •The east–west radiators of a GEO communications satellite are connected using LHPs. •A transient mathematical model is validated with flight data. •The space flight data proved successful in-orbit operation of the novel architecture. •The model can be used to design/analyze LHP based complex thermal architectures

Thermal hydraulic and neutronic interaction in the rotating bed reactor

International Nuclear Information System (INIS)

Lee, C.C.

1986-01-01

Power transient characteristics in a rotating fluidized bed reactor (RBR) are investigated theoretically. A propellant flow perturbation is assumed to occur in an initially equilibrium state of the core. Transfer functions representing quasi-one-dimensional mutual feedback between thermal hydraulics and neutronics are developed and analyzed in the frequency domain. Neutronic responses are determined by Fermi-age theory for slowing down of fast neutrons and diffusion theory for thermal neutron distribution. Neutron leakage through the exhaust nozzle is accounted for by applying diffuse view factors similar to those applied in radiative heat transfer. The bed expansion behavior is described by a kinematic wave equation derived from the continuity of the gas phase. The drift flux approach is used to determine the yield fractions in the equilibrium bed. Thermal responses of fuel are evaluated by dividing it into several volume-averaged zones to better account for the transient effects over single zone models. Sample calculations are undertaken for the various operation conditions and design parameters of the RBR based on 250 MW/sub t/, 1000 MW/sub t/, and 5000 MW/sub t/ power reactors. The results show that power transients are dependent on the parametric changes of optical thickness and view factors

Transient thermal effect, nonlinear refraction and nonlinear absorption properties of graphene oxide sheets in dispersion.

Science.gov (United States)

Zhang, Xiao-Liang; Liu, Zhi-Bo; Li, Xiao-Chun; Ma, Qiang; Chen, Xu-Dong; Tian, Jian-Guo; Xu, Yan-Fei; Chen, Yong-Sheng

2013-03-25

The nonlinear refraction (NLR) properties of graphene oxide (GO) in N, N-Dimethylformamide (DMF) was studied in nanosecond, picosecond and femtosecond time regimes by Z-scan technique. Results show that the dispersion of GO in DMF exhibits negative NLR properties in nanosecond time regime, which is mainly attributed to transient thermal effect in the dispersion. The dispersion also exhibits negative NLR in picosecond and femtosecond time regimes, which are arising from sp(2)- hybridized carbon domains and sp(3)- hybridized matrix in GO sheets. To illustrate the relations between NLR and nonlinear absorption (NLA), NLA properties of the dispersion were also studied in nanosecond, picosecond and femtosecond time regimes.

Effect of Thermal Environment on the Mechanical Behaviors of Building Marble

Directory of Open Access Journals (Sweden)

Haijian Su

2018-01-01

Full Text Available High temperature and thermal environment can influence the mechanical properties of building materials worked in the civil engineering, for example, concrete, building rock, and steel. This paper examines standard cylindrical building marble specimens (Φ50 × 100 mm that were treated with high temperatures in two different thermal environments: vacuum (VE and airiness (AE. Uniaxial compression tests were also carried out on those specimens after heat treatment to study the effect that the thermal environment has on mechanical behaviors. With an increase in temperature, the mechanical behavior of marble in this study indicates a critical temperature of 600°C. Both the peak stress and elasticity modulus were larger for the VE than they were for the AE. The thermal environment has an obvious influence on the mechanical properties, especially at temperatures of 450∼750°C. The failure mode of marble specimens under uniaxial compression is mainly affected by the thermal environment at 600°C.

Transient analysis of a U-tube natural circulation steam generator

Energy Technology Data Exchange (ETDEWEB)

Gaikwad, A J; Kumar, Rajesh; Bhadra, Anu; Chakraborty, G; Venkat Raj, V [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

A computer code has been developed, for transient thermal-hydraulic analysis of proposed 500 MWe PHWR steam generator. The transient behaviour of a nuclear power plant is very much dependent on the steam generator performance, as it provides a thermal linkage between the primary and secondary systems. Study of dynamics of steam generator is essential for over all power plant dynamics as well as design of control systems for steam generator. A mathematical model has been developed for the simulation of thermal-hydraulic phenomena in a U tube natural circulation steam generator. Fluid model is based on one dimensional, nonlinear, single fluid conservation equations of mass, momentum, energy and equation of state. This model includes coupled two phase flow heat transfer and natural circulation. The model accounts for both compressibility and thermal expansion effects. The process simulation and results obtained for transients such as step change in load and total loss of feed water are presented. (author). 5 refs., 7 figs.

Oxide fuel pin transient performance analysis and design with the TEMECH code

International Nuclear Information System (INIS)

Bard, F.E.; Dutt, S.P.; Hinman, C.A.; Hunter, C.W.; Pitner, A.L.

1986-01-01

The TEMECH code is a fast-running, thermal-mechanical-hydraulic, analytical program used to evaluate the transient performance of LMR oxide fuel pins. The code calculates pin deformation and failure probability due to fuel-cladding differential thermal expansion, expansion of fuel upon melting, and fission gas pressurization. The mechanistic fuel model in the code accounts for fuel cracking, crack closure, porosity decrease, and the temperature dependence of fuel creep through the course of the transient. Modeling emphasis has been placed on results obtained from Fuel Cladding Transient Test (FCTT) testing, Transient Fuel Deformation (TFD) tests and TREAT integral fuel pin experiments

Transient magnetoviscosity of dilute ferrofluids

International Nuclear Information System (INIS)

Soto-Aquino, Denisse; Rinaldi, Carlos

2011-01-01

The magnetic field induced change in the viscosity of a ferrofluid, commonly known as the magnetoviscous effect and parameterized through the magnetoviscosity, is one of the most interesting and practically relevant aspects of ferrofluid phenomena. Although the steady state behavior of ferrofluids under conditions of applied constant magnetic fields has received considerable attention, comparatively little attention has been given to the transient response of the magnetoviscosity to changes in the applied magnetic field or rate of shear deformation. Such transient response can provide further insight into the dynamics of ferrofluids and find practical application in the design of devices that take advantage of the magnetoviscous effect and inevitably must deal with changes in the applied magnetic field and deformation. In this contribution Brownian dynamics simulations and a simple model based on the ferrohydrodynamics equations are applied to explore the dependence of the transient magnetoviscosity for two cases: (I) a ferrofluid in a constant shear flow wherein the magnetic field is suddenly turned on, and (II) a ferrofluid in a constant magnetic field wherein the shear flow is suddenly started. Both simulations and analysis show that the transient approach to a steady state magnetoviscosity can be either monotonic or oscillatory depending on the relative magnitudes of the applied magnetic field and shear rate. - Research Highlights: →Rotational Brownian dynamics simulations were used to study the transient behavior of the magnetoviscosity of ferrofluids. →Damped and oscillatory approach to steady state magnetoviscosity was observed for step changes in shear rate and magnetic field. →A model based on the ferrohydrodynamics equations qualitatively captured the damped and oscillatory features of the transient response →The transient behavior is due to the interplay of hydrodynamic, magnetic, and Brownian torques on the suspended particles.

BEHAVIOR OF THERMAL SPRAY COATINGS AGAINST HYDROGEN ATTACK

OpenAIRE

Vargas, Fabio; Latorre, Guillermo; Uribe, Iván

2003-01-01

The behavior of nickel and chrome alloys applied as thermal spray coatings to be used as protection against embrittlement by hydrogen is studied. Coatings were applied on a carbon steel substrate, under conditions that allow obtain different crystalline structures and porosity levels, in order to determine the effect of these variables on the hydrogen permeation kinetics and as a protection means against embrittlement caused this element. In order to establish behaviors as barriers and protec...

Timing of transients: quantifying reaching times and transient behavior in complex systems

Science.gov (United States)

Kittel, Tim; Heitzig, Jobst; Webster, Kevin; Kurths, Jürgen

2017-08-01

In dynamical systems, one may ask how long it takes for a trajectory to reach the attractor, i.e. how long it spends in the transient phase. Although for a single trajectory the mathematically precise answer may be infinity, it still makes sense to compare different trajectories and quantify which of them approaches the attractor earlier. In this article, we categorize several problems of quantifying such transient times. To treat them, we propose two metrics, area under distance curve and regularized reaching time, that capture two complementary aspects of transient dynamics. The first, area under distance curve, is the distance of the trajectory to the attractor integrated over time. It measures which trajectories are ‘reluctant’, i.e. stay distant from the attractor for long, or ‘eager’ to approach it right away. Regularized reaching time, on the other hand, quantifies the additional time (positive or negative) that a trajectory starting at a chosen initial condition needs to approach the attractor as compared to some reference trajectory. A positive or negative value means that it approaches the attractor by this much ‘earlier’ or ‘later’ than the reference, respectively. We demonstrated their substantial potential for application with multiple paradigmatic examples uncovering new features.

Simulation of the impact of wind power on the transient fault behavior of the Nordic power system

DEFF Research Database (Denmark)

Jauch, Clemens; Sørensen, Poul Ejnar; Norheim, Ian

2007-01-01

influences the post-fault behavior of the Nordic power system. It is concluded that an increasing level of wind power penetration leads to stronger system oscillations in case of fixed speed wind turbines. It is found that fixed speed wind turbines that merely ride through transient faults have negative......In this paper the effect of wind power on the transient fault behavior of the Nordic power system is investigated. The Nordic power system is the interconnected power system of the countries Norway, Sweden, Finland and Denmark. For the purpose of these investigations the wind turbines installed...... and connected in eastern Denmark are taken as study case. The current and future wind power situation in eastern Denmark is modeled and short circuit faults in the system simulated. The simulations yield information on (i) how the faults impact on the wind turbines and (ii) how the response of the wind turbines...

Steady-State and Transient Analysis for Design Validation of SMART-ITL Secondary System

Energy Technology Data Exchange (ETDEWEB)

Yun, Eunkoo; Bae, Hwang; Ryu, Sung Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

SMART can prevent large-break loss of coolant accident (LBLOCA) inherently. SMART-ITL is an experimental simulation facility designed to perform integral effect tests for the SMART plant. In terms of the secondary system of SMART-ITL, the design has been simplified from that of reference plant by replacing several components, such as expansion device and condenser, with an appropriate device to be functional as the alternatives. In this paper, in order to understand the operational characteristics as well as design concept, the secondary system of SMRAT-ITL is analyzed in steady-state and transient aspects, and the results are compared with relevant experimental results. This study focuses on the understanding of thermal-hydraulic behavior of SMART-ITL secondary system, which is simplified from that of reference plant. To identify the behaviors of the secondary system, the steady-state and transient analysis were conducted based on experimental results. In steady-state analysis, the results clearly showed that the system pressure is related to the temperature of condensation tank which varies depending on mixture enthalpy. In transient analysis, the dynamic behavior during heat-up process has been investigated. The results reveal that we can reasonably assume the fluid filled in TK-CD-01 be in a saturated condition. The results showed that the design of SMART-ITL secondary system is appropriate, and the system is being properly operated to match the design intent.

Development of THYDE-HTGR: computer code for transient thermal-hydraulics of high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Hirano, Masashi; Hada, Kazuhiko

1990-04-01

The THYDE-HTGR code has been developed for transient thermal-hydraulic analyses of high-temperature gas-cooled reactors, based on the THYDE-W code. THYDE-W is a code developed at JAERI for the simulation of Light Water Reactor plant dynamics during various types of transients including loss-of-coolant accidents. THYDE-HTGR solves the conservation equations of mass, momentum and energy for compressible gas, or single-phase or two-phase flow. The major code modification from THYDE-W is to treat helium loops as well as water loops. In parallel to this, modification has been made for the neutron kinetics to be applicable to helium-cooled graphite-moderated reactors, for the heat transfer models to be applicable to various types of heat exchangers, and so forth. In order to assess the validity of the modifications, analyses of some of the experiments conducted at the High Temperature Test Loop of ERANS have been performed. In this report, the models applied in THYDE-HTGR are described focusing on the present modifications and the results from the assessment calculations are presented. (author)

A COMETHE version with transient capability

International Nuclear Information System (INIS)

Vliet, J. van; Lebon, G.; Mathieu, P.

1980-01-01

A version of the COMETHE code is under development to simulate transient situations. This paper focuses on some aspects of the transient heat transfer models. Initially the coupling between transient heat transfer and other thermomechanical models is discussed. An estimation of the thermal characteristic times shows that the cladding temperatures are often in quasi-steady state. In order to reduce the computing time, calculations are therefore switched from a transient to a quasi-static numerical procedure as soon as such a quasi-equilibrium is detected. The temperature calculation is performed by use of the Lebon-Lambermont restricted variational principle, with piecewise polynoms as trial functions. The method has been checked by comparison with some exact results and yields good agreement for transient as well as for quasi-static situations. This method therefore provides a valuable tool for the simulation of the transient behaviour of nuclear reactor fuel rods. (orig.)

Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Asaga, T.; Shikakura, S.

1991-02-01

The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab

CFD analysis of thermal-hydraulic behavior in SCWR typical flow channels

International Nuclear Information System (INIS)

Gu, H.Y.; Cheng, X.; Yang, Y.H.

2008-01-01

Investigations on thermal-hydraulic behavior in SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding and ability to predict the heat transfer behavior of supercritical water. In this paper, CFD analysis is carried out to study the flow and heat transfer behavior of supercritical water in sub-channels of both square and triangular rod bundles. Effect of various parameters, e.g. thermal boundary conditions and pitch-to-diameter ratio on the thermal-hydraulic behavior is investigated. Two boundary conditions, i.e., constant heat flux at the outer surface of cladding and constant heat density in the fuel pin are applied. The results show that the structure of the secondary flow mainly depends on the rod bundle configuration as well as the pitch-to-diameter ratio, whereas, the amplitude of the secondary flow is affected by the thermal boundary conditions, as well. The secondary flow is much stronger in a square lattice than that in a triangular lattice. The turbulence behavior is similar in both square and triangular lattices. The dependence of the amplitude of the turbulent velocity fluctuation across the gap on Reynolds number becomes prominent in both lattices as the pitch-to-diameter ratio increases. The effect of thermal boundary conditions on turbulent velocity fluctuation is negligibly small. For both lattices with small pitch-to-diameter ratios (P/D < 1.3), the mixing coefficient is about 0.022. Both secondary flow and turbulent mixing show unusual behavior in the vicinity of the pseudo-critical point. Further investigation is needed. A strong circumferential non-uniformity of wall temperature and heat transfer is observed in tight lattices at constant heat flux boundary conditions, especially in square lattices. In the case with constant heat density of fuel pin, the circumferential conductive heat transfer significantly reduces the non-uniformity of circumferential

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

Energy Technology Data Exchange (ETDEWEB)

Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm² and temporary heat flux limit of 600 W/cm². This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

Impacts of transient heat transfer modeling on prediction of advanced cladding fracture during LWR LBLOCA

Energy Technology Data Exchange (ETDEWEB)

Lee, Youho, E-mail: euo@kaist.ac.kr; Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

2016-03-15

Highlights: • Use of constant heat transfer coefficient for fracture analysis is not sound. • On-time heat transfer coefficient should be used for thermal fracture prediction. • ∼90% of the actual fracture stresses were predicted with the on-time transient h. • Thermal-hydraulic codes can be used to better predict brittle cladding fracture. • Effects of surface oxides on thermal shock fracture should be accounted by h. - Abstract: This study presents the importance of coherency in modeling thermal-hydraulics and mechanical behavior of a solid for an advanced prediction of cladding thermal shock fracture. In water quenching, a solid experiences dynamic heat transfer rate evolutions with phase changes of the fluid over a short quenching period. Yet, such a dynamic change of heat transfer rates has been overlooked in the analysis of thermal shock fracture. In this study, we are presenting quantitative evidence against the prevailing use of a constant heat transfer coefficient for thermal shock fracture analysis in water. We conclude that no single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials. The presented results show a remarkable stress prediction improvement up to 80–90% of the actual stress with the use of the surface temperature dependent heat transfer coefficient. For thermal shock fracture analysis of brittle fuel cladding such as oxidized zirconium-based alloy or silicon carbide during LWR reflood, transient subchannel heat transfer coefficients obtained from a thermal-hydraulics code should be used as input for stress analysis. Such efforts will lead to a fundamental improvement in thermal shock fracture predictability over the current experimental empiricism for cladding fracture analysis during reflood.

FORTRAN routines for calculating water thermodynamic properties for use in transient thermal-hydraulics codes

International Nuclear Information System (INIS)

Green, C.

1979-12-01

A set of FORTRAN subroutines is described for calculating water thermodynamic properties. These were written for use in a transient thermal-hydraulics program, where speed of execution is paramount. The choice of which subroutines to optimise depends on the primary variables in the thermal-hydraulics code. In this particular case the subroutine which has been optimised is the one which calculates pressure and specific enthalpy given the specific volume and the specific internal energy. Another two subroutines are described which complete a self-consistent set. These calculate the specific volume and the temperature given the pressure and the specific enthalpy, and the specific enthalpy and the specific volume given the pressure and the temperature (or the quality). The accuracy is high near the saturation lines, typically less than 1% relative error, and decreases as the fluid becomes more subcooled in the liquid region or more superheated in the steam region. This behaviour is inherent in the method which uses quantities defined on the saturation lines and assumes that certain derivatives are constant for excursions away from these saturation lines. The accuracy and speed of the subroutines are discussed in detail in this report. (author)

Transient thermal analysis for radioactive liquid mixing operations in a large-scaled tank

International Nuclear Information System (INIS)

Lee, S. Y.; Smith, F. G. III

2014-01-01

A transient heat balance model was developed to assess the impact of a Submersible Mixer Pump (SMP) on radioactive liquid temperature during the process of waste mixing and removal for the high-level radioactive materials stored in Savannah River Site (SRS) tanks. The model results will be mainly used to determine the SMP design impacts on the waste tank temperature during operations and to develop a specification for a new SMP design to replace existing longshaft mixer pumps used during waste removal. The present model was benchmarked against the test data obtained by the tank measurement to examine the quantitative thermal response of the tank and to establish the reference conditions of the operating variables under no SMP operation. The results showed that the model predictions agreed with the test data of the waste temperatures within about 10%

Single- and two-phase flow modeling for coupled neutronics / thermal-hydraulics transient analysis of advanced sodium-cooled fast reactors

International Nuclear Information System (INIS)

Chenu, A.

2011-10-01

Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the

Crack growth in first wall made of reduced activation ferritic steel by transient creep due to long pulse operation

International Nuclear Information System (INIS)

Honda, T.; Kudo, Y.; Hatano, T.; Kikuchi, K.; Nishimura, T.; Saito, M.

2003-01-01

The long pulse operation is assumed in ITER and future reactor. If the first wall has a defect, the crack may be propagated by cyclic thermal loads. In addition, flattop of more than 300 s during plasma burning is expected in ITER, so the crack propagation behavior will depend on the operation duration period. This study deals with the crack propagation behavior on F82H under high thermal load cycles. The high heat flux tests were performed under three types of duration periods to investigate creep fatigue behavior. To clarify the crack growth mechanism and the effects of transient creep, three-dimensional analyses were performed. It was concluded that the creep effect during the operation duration period enlarges stress intensity factor K in the cooling period and that consequently, the crack propagation length was increased

Positron beam studies of transients in semiconductors

International Nuclear Information System (INIS)

Beling, C.D.; Ling, C.C.; Cheung, C.K.; Naik, P.S.; Zhang, J.D.; Fung, S.

2006-01-01

Vacancy-sensing positron deep level transient spectroscopy (PDLTS) is a positron beam-based technique that seeks to provide information on the electronic ionization levels of vacancy defects probed by the positron through the monitoring of thermal transients. The experimental discoveries leading to the concept of vacancy-sensing PDLTS are first reviewed. The major problem associated with this technique is discussed, namely the strong electric fields establish in the near surface region of the sample during the thermal transient which tend to sweep positrons into the contact with negligible defect trapping. New simulations are presented which suggest that under certain conditions a sufficient fraction of positrons may be trapped into ionizing defects rendering PDLTS technique workable. Some suggestions are made for techniques that might avoid the problematic electric field problem, such as optical-PDLTS where deep levels are populated using light and the use of high forward bias currents for trap filling

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

International Nuclear Information System (INIS)

2007-01-01

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

Thermal stratification in the pressurizer

International Nuclear Information System (INIS)

Baik, S.J.; Lee, K.W.; Ro, T.S.

2001-01-01

The thermal stratification in the pressurizer due to the insurge from the hot leg to the pressurizer has been studied. The insurge flow of the cold water into the pressurizer takes place during the heatup/cooldown and the normal or abnormal transients during power operation. The pressurizer vessel can undergo significant thermal fatigue usage caused by insurges and outsurges. Two-dimensional axisymmetric transient analysis for the thermal stratification in the pressurizer is performed using the computational fluid dynamics code, FLUENT, to get the velocity and temperature distribution. Parametric study has been carried out to investigate the effect of the inlet velocity and the temperature difference between the hot leg and the pressurizer on the thermal stratification. The results show that the insurge flow of cold water into the pressurizer does not mix well with hot water, and the cold water remains only in the lower portion of the pressurizer, which leads to the thermal stratification in the pressurizer. The thermal load on the pressurizer due to the thermal stratification or the cyclic thermal transient should be examined with respect to the mechanical integrity and this study can serve the design data for the stress analysis. (authors)

Estimation of apparent kinetic parameters of polymer pyrolysis with complex thermal degradation behavior

International Nuclear Information System (INIS)

Srimachai, Taranee; Anantawaraskul, Siripon

2010-01-01

Full text: Thermal degradation behavior during polymer pyrolysis can typically be described using three apparent kinetic parameters (i.e., pre-exponential factor, activation energy, and reaction order). Several efficient techniques have been developed to estimate these apparent kinetic parameters for simple thermal degradation behavior (i.e., single apparent pyrolysis reaction). Unfortunately, these techniques cannot be directly extended to the case of polymer pyrolysis with complex thermal degradation behavior (i.e., multiple concurrent reactions forming single or multiple DTG peaks). In this work, we proposed a deconvolution method to determine the number of apparent reactions and estimate three apparent kinetic parameters and contribution of each reaction for polymer pyrolysis with complex thermal degradation behavior. The proposed technique was validated with the model and experimental pyrolysis data of several polymer blends with known compositions. The results showed that (1) the number of reaction and (2) three apparent kinetic parameters and contribution of each reaction can be estimated reasonably. The simulated DTG curves with estimated parameters also agree well with experimental DTG curves. (author)

Thermal-mechanical analysis for a viscoelastoplastic model by finite element method

International Nuclear Information System (INIS)

Vaz, L.E.; Vaz, R.O.E.

1989-01-01

The aim of this work is to present a formulation and a computer program which permits the study of problems involving the influence of the temperature on the mechanical behavior of a viscoelastoplastic material. The thermo-mechanical analysis is carried out in two steps. The first step performs the transient thermal analysis. The second step uses the time-history of the temperature distribution that results on the first step, for the transient stress analysis. The program treat plane and axi-symmetrical problems. As an application of the formulation the quenching of a cylinder of metal is examined. (author)

Effect of thermal transients on the hardness of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Hobson, D.O.

1976-06-01

This study is directed toward the determination of the effects of annealing cycles with rapid heating rates, short hold times at specific temperatures, and rapid cool-down rates on the hardness of Zircaloy fuel cladding. These rapid annealing cycles are designed to provide preliminary annealing behavior data on Loss-of-Fluid-Test Reactor cladding samples. Information has been obtained on (1) the time dependence of the hardness as a function of annealing temperature, and (2) a correlation of single- and multitransient annealing relationships. Both single- and triple-cycle transients were used; four hold times at each of five maximum temperatures comprised the data set (each portion of the triple-cycle experiments had isothermal hold times equal to one-third of their analogous single-cycle times). It was found that there was little difference in the hardness response between single- and triple-cycle transients for a given total hold time at a particular temperature. Test temperatures range from 1000 to 1400 0 F (538 to 760 0 C) and hold times from 5 to 135 sec. The 1100 0 F (593 0 C) level was found to be the transition level for hardness changes, with shorter times (5 and 15 sec) effecting little or no hardness decrease and the longer times (45 and 135 sec) producing partially and fully annealed material, respectively. Temperatures equal to or greater than 1300 0 F (704 0 C) resulted in fully annealed material for all hold times. The 1000 0 F (538 0 C) tests produced no measurable softening

Thermal and kinetic behaviors of biomass and plastic wastes in co-pyrolysis

International Nuclear Information System (INIS)

Çepelioğullar, Özge; Pütün, Ayşe E.

2013-01-01

Graphical abstract: - Highlights: • Co-pyrolysis of biomass together with the plastic wastes in thermogravimetric analyzer. • Investigations into thermal and kinetic behaviors at high temperature regions. • Determination of the kinetic parameters. - Abstract: In this study, co-pyrolysis characteristics and kinetics of biomass-plastic blends were investigated. Cotton stalk, hazelnut shell, sunflower residue, and arid land plant Euphorbia rigida, were blended in definite ratio (1:1, w/w) with polyvinyl chloride (PVC) and polyethylene terephthalate (PET). Experiments were conducted with a heating rate of 10 °C min −1 from room temperature to 800 °C in the presence of N 2 atmosphere with a flow rate of 100 cm 3 min −1 . After thermal decomposition in TGA, a kinetic analysis was performed to fit thermogravimetric data and a detailed discussion of co-pyrolysis mechanism was achieved. Experimental results demonstrated that the structural differences between biomass and plastics directly affect their thermal decomposition behaviors. Biomass pyrolysis generally based on three main steps while plastic material’s pyrolysis mechanism resulted in two steps for PET and three steps for PVC. Also, the required activation energies needed to achieve the thermal degradation for plastic were found higher than the biomass materials. In addition, it can be concluded that the evaluation of plastic materials together with biomass created significant changes not only for the thermal behaviors but also for the kinetic behaviors

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

International Nuclear Information System (INIS)

Daw, J.E.; Knudson, D.L.; Villard, J.F.; Liothin, J.; Destouches, C.; Rempe, J.L.; Matheron, P.; Lambert, T.

2015-01-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

Energy Technology Data Exchange (ETDEWEB)

Daw, J.E.; Knudson, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415, (United States); Villard, J.F.; Liothin, J.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Rempe, J.L. [Rempe and Associates, LLC, Idaho Falls, ID, 83404 (United States); Matheron, P. [CEA, DEN, DEC, Uranium Fuels Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Lambert, T. [CEA, DEN, DEC, Innovative Fuel Design and Irradiation Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France)

2015-07-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were

TORT-TD/ATTICA3D: a coupled neutron transport and thermal hydraulics code system for 3-D transient analysis of gas cooled high temperature reactors

International Nuclear Information System (INIS)

Lapins, J.; Seubert, A.; Buck, M.; Bader, J.; Laurien, E.

2011-01-01

Comprehensive safety studies of high temperature gas cooled reactors (HTR) require full three dimensional coupled treatments of both neutron kinetics and thermal-hydraulics. In a common effort, GRS and IKE developed the coupled code system TORT-TD/ATTICA3D for pebble bed type HTR that connects the 3-D transient discrete-ordinates transport code TORT-TD with the 3-D porous medium thermal-hydraulics code ATTICA3D. In this paper, the physical models and calculation capabilities of TORT-TD and ATTICA3D are presented, focusing on model improvements in ATTICA3D and extensions made in TORT-TD related to HTR application. For first applications, the OECD/NEA/NSC PBMR-400 benchmark has been chosen. Results obtained with TORT-TD/ATTICA3D will be shown for transient exercises, e.g. control rod withdrawal and a control rod ejection. Results are compared to other benchmark participants' solutions with special focus on fuel temperature modelling features of ATTICA3D. The provided “grey-curtain” nuclear cross section libraries have been used. First results on 3-D effects during a control rod withdrawal transient will be presented. (author)

A heat transfer model for the analysis of transient heating of the slab in a direct-fired walking beam type reheating furnace

Energy Technology Data Exchange (ETDEWEB)

Kim, Man Young [School of Mechanical and Aerospace Systems Engineering, Research Center of Industrial Technology, Chonbuk National University, 664-14 Duckjin-Dong, Duckjin-Gu, Jeonju, Chonbuk 561-756 (Korea)

2007-09-15

A mathematical heat transfer model for the prediction of heat flux on the slab surface and temperature distribution in the slab has been developed by considering the thermal radiation in the furnace chamber and transient heat conduction governing equations in the slab, respectively. The furnace is modeled as radiating medium with spatially varying temperature and constant absorption coefficient. The steel slabs are moved on the next fixed beam by the walking beam after being heated up through the non-firing, charging, preheating, heating, and soaking zones in the furnace. Radiative heat flux calculated from the radiative heat exchange within the furnace modeled using the FVM by considering the effect of furnace wall, slab, and combustion gases is introduced as the boundary condition of the transient conduction equation of the slab. Heat transfer characteristics and temperature behavior of the slab is investigated by changing such parameters as absorption coefficient and emissivity of the slab. Comparison with the experimental work show that the present heat transfer model works well for the prediction of thermal behavior of the slab in the reheating furnace. (author)

Stress analysis in pipelines submitted to internal pressure - and temperature transients

International Nuclear Information System (INIS)

Mansur, T.R.

1981-08-01

Experimental determination of the structural behaviour of a thermal-hydraulic loop, when submitted to simultaneous fast change of pressure and temperature, was performed. For this, electrical strain-gages were positioned at some critical points in order to measure the deformation conditions of the structure. The study of the kinetics of the deformation revealed the presence of important transient stresses, mainly from thermal origin. After this transient behaviour, the structure is submitted to a thermal stress, which is shown to be strongly dependent on the degree of restraint of the structure. (Author) [pt

Thermal stress mitigation by Active Thermal Control

DEFF Research Database (Denmark)

Soldati, Alessandro; Dossena, Fabrizio; Pietrini, Giorgio

2017-01-01

This work proposes an Active Thermal Control (ATC) of power switches. Leveraging on the fact that thermal stress has wide impact on the system reliability, controlling thermal transients is supposed to lengthen the lifetime of electronic conversion systems. Indeed in some environments...... results of control schemes are presented, together with evaluation of the proposed loss models. Experimental proof of the ability of the proposed control to reduce thermal swing and related stress on the device is presented, too....

Thermal fatigue. Materials modelling

International Nuclear Information System (INIS)

Siegele, D.; Fingerhuth, J.; Mrovec, M.

2012-01-01

In the framework of the ongoing joint research project 'Thermal Fatigue - Basics of the system-, outflow- and material-characteristics of piping under thermal fatigue' funded by the German Federal Ministry of Education and Research (BMBF) fundamental numerical and experimental investigations on the material behavior under transient thermal-mechanical stress conditions (high cycle fatigue V HCF and low cycle fatigue - LCF) are carried out. The primary objective of the research is the further development of simulation methods applied in safety evaluations of nuclear power plant components. In this context the modeling of crack initiation and growth inside the material structure induced by varying thermal loads are of particular interest. Therefore, three scientific working groups organized in three sub-projects of the joint research project are dealing with numerical modeling and simulation at different levels ranging from atomistic to micromechanics and continuum mechanics, and in addition corresponding experimental data for the validation of the numerical results and identification of the parameters of the associated material models are provided. The present contribution is focused on the development and experimental validation of material models and methods to characterize the damage evolution and the life cycle assessment as a result of thermal cyclic loading. The individual purposes of the subprojects are as following: - Material characterization, Influence of temperature and surface roughness on fatigue endurances, biaxial thermo-mechanical behavior, experiments on structural behavior of cruciform specimens and scatter band analysis (IfW Darmstadt) - Life cycle assessment with micromechanical material models (MPA Stuttgart) - Life cycle assessment with atomistic and damage-mechanical material models associated with material tests under thermal fatigue (Fraunhofer IWM, Freiburg) - Simulation of fatigue crack growth, opening and closure of a short crack under

ELOCA: fuel element behaviour during high temperature transients

International Nuclear Information System (INIS)

Sills, H.E.

1979-03-01

The ELOCA computer code was developed to simulate the uniform thermal-mechanical behaviour of a fuel element during high-temperature transients such as a loss-of-coolant accident (LOCA). Primary emphasis is on the diametral expansion of the fuel sheath. The model assumed is a single UO2/zircaloy-clad element with axisymmetric properties. Physical effects considered by the code are fuel expansion, cracking and melting; variation, during the transient, of internal gas pressure; changing fuel/sheath heat transfer; thermal, elastic and plastic sheath deformation (anisotropic); Zr/H 2 O chemical reaction effects; and beryllium-assisted crack penetration of the sheath. (author)

State of the art of CATHARE model for transient safety analysis of ASTRID SFR

International Nuclear Information System (INIS)

Lavastre, R.; Conti, A.; Marsault, Ph.; Chenaud, M.S.; Tosello, A.

2014-01-01

Within the framework of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), the conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves enhancing the general design in order to : - increase the safety margins for all unprotected-loss-of-flow (ULOF) and unprotected-loss-of-heat-sink (ULOHS) transients, - identify the need for additional safety devices that would complement core natural behavior so that temperature criteria on coolant, core and primary circuit structures can remain under the safety criteria. For this purpose, the use of CATHARE system code has been very important from the early stage of design in order to ensure a feedback for design teams to improve behavior during unprotected transients. Until 2012, CATHARE ULOxx transient calculations have been used mainly to compare different core designs. They contributed to lead to the choice of CFV core (axially heterogeneous core with an upper sodium plenum employed to achieve a negative sodium void reactivity worth). Meanwhile, models for an accurate core description and transients have been developed in CATHARE to improve the calculations towards best estimate calculations for safety analysis. This paper therefore presents these main developments in core modeling achieved for the 2 past years. For instance, we will focus on the way of dealing with fuel assemblies that have to be grouped together in the CATHARE code to form a channel with similar neutronic physics and thermal-hydraulics characteristics. We will also explain the way we deal with heterogeneity of fuel pin to obtain the accurate fuel temperature along the axis and to take into account pellet-cladding gap state. These two points have a great importance on feedback effects linked to the fuel, mainly the Doppler effect. The paper will finally introduce the upcoming improvements that are under development nowadays

Prediction of the thermal behavior of a particle spherical fuel element using GITT

International Nuclear Information System (INIS)

Pessoa, C.V.; Oliveira, Claudio L. de; Jian, Su

2008-01-01

In this work, the transient and steady state heat conduction in a spherical fuel element of a pebble-bed high temperature were studied. This pebble element is composed by a particulate region with spherical inclusions, the fuel UO 2 particles, dispersed in a graphite matrix. A convective heat transfer by helium occurs on the outer surface of the fuel element. The two-energy equation model for the case of pure conduction was applied to this particulate spherical element, generating two macroscopic temperatures, respectively, of the inclusions and of the matrix. The transient analysis was carried out by using the Generalized Integral Transform Technique (GITT) that requires low computational efforts and allows a fast evaluation of the two macroscopic transient temperatures of the particulate region. The solution by GITT leads to a system of ordinary differential equations with the unknown transformed potentials. The mechanical properties (thermal conductivity and specific heat) of the materials were supposed not to depend on the temperature and to be uniform in each region. (author)

Thermal Gradient Cyclic Behavior of a Thermal/Environmental Barrier Coating System on SiC/SiC Ceramic Matrix Composites

Science.gov (United States)

Zhu, Dongming; Lee, Kang N.; Miller, Robert A.

2002-01-01

Thermal barrier and environmental barrier coatings (TBCs and EBCs) will play a crucial role in future advanced gas turbine engines because of their ability to significantly extend the temperature capability of the ceramic matrix composite (CMC) engine components in harsh combustion environments. In order to develop high performance, robust coating systems for effective thermal and environmental protection of the engine components, appropriate test approaches for evaluating the critical coating properties must be established. In this paper, a laser high-heat-flux, thermal gradient approach for testing the coatings will be described. Thermal cyclic behavior of plasma-sprayed coating systems, consisting of ZrO2-8wt%Y2O3 thermal barrier and NASA Enabling Propulsion Materials (EPM) Program developed mullite+BSAS/Si type environmental barrier coatings on SiC/SiC ceramic matrix composites, was investigated under thermal gradients using the laser heat-flux rig in conjunction with the furnace thermal cyclic tests in water-vapor environments. The coating sintering and interface damage were assessed by monitoring the real-time thermal conductivity changes during the laser heat-flux tests and by examining the microstructural changes after the tests. The coating failure mechanisms are discussed based on the cyclic test results and are correlated to the sintering, creep, and thermal stress behavior under simulated engine temperature and heat flux conditions.

Analysis of the thermal performance of heat pipe radiators

Science.gov (United States)

Boo, J. H.; Hartley, J. G.

1990-01-01

A comprehensive mathematical model and computational methodology are presented to obtain numerical solutions for the transient behavior of a heat pipe radiator in a space environment. The modeling is focused on a typical radiator panel having a long heat pipe at the center and two extended surfaces attached to opposing sides of the heat pipe shell in the condenser section. In the set of governing equations developed for the model, each region of the heat pipe - shell, liquid, and vapor - is thermally lumped to the extent possible, while the fin is lumped only in the direction normal to its surface. Convection is considered to be the only significant heat transfer mode in the vapor, and the evaporation and condensation velocity at the liquid-vapor interface is calculated from kinetic theory. A finite-difference numerical technique is used to predict the transient behavior of the entire radiator in response to changing loads.

Simultaneous measurement of thermal conductivity and heat capacity of bulk and thin film materials using frequency-dependent transient thermoreflectance method.

Science.gov (United States)

Liu, Jun; Zhu, Jie; Tian, Miao; Gu, Xiaokun; Schmidt, Aaron; Yang, Ronggui

2013-03-01

The increasing interest in the extraordinary thermal properties of nanostructures has led to the development of various measurement techniques. Transient thermoreflectance method has emerged as a reliable measurement technique for thermal conductivity of thin films. In this method, the determination of thermal conductivity usually relies much on the accuracy of heat capacity input. For new nanoscale materials with unknown or less-understood thermal properties, it is either questionable to assume bulk heat capacity for nanostructures or difficult to obtain the bulk form of those materials for a conventional heat capacity measurement. In this paper, we describe a technique for simultaneous measurement of thermal conductivity κ and volumetric heat capacity C of both bulk and thin film materials using frequency-dependent time-domain thermoreflectance (TDTR) signals. The heat transfer model is analyzed first to find how different combinations of κ and C determine the frequency-dependent TDTR signals. Simultaneous measurement of thermal conductivity and volumetric heat capacity is then demonstrated with bulk Si and thin film SiO2 samples using frequency-dependent TDTR measurement. This method is further testified by measuring both thermal conductivity and volumetric heat capacity of novel hybrid organic-inorganic thin films fabricated using the atomic∕molecular layer deposition. Simultaneous measurement of thermal conductivity and heat capacity can significantly shorten the development∕discovery cycle of novel materials.

Transient thermal hydraulic analysis of the IAEA 10 MW MTR reactor during Loss of Flow Accident to investigate the flow inversion

International Nuclear Information System (INIS)

AL-Yahia, Omar S.; Albati, Mohammad A.; Park, Jonghark; Chae, Heetaek; Jo, Daeseong

2013-01-01

Highlights: • Transient analyses of a slow and fast LOFA were investigated. • A reactor kinetic and thermal hydraulic coupled model was developed. • Based on force balance, the flow rate during flow inversion was determined. • Flow inversion in a hot channel occurred earlier than in an average channel. • Two temperature peaks were observed during both slow and fast LOFA. - Abstract: Transient analyses of the IAEA 10 MW MTR reactor are investigated during a fast and slow Loss of Flow Accident (LOFA) with a neutron kinetic and thermal hydraulic coupling model. A spatial-dependent thermal hydraulic technique is adopted for analyzing the local thermal hydraulic parameters and hotspot location during a flow inversion. The flow rate through the channel is determined in terms of a balance between driving and preventing forces. Friction and buoyancy forces act as resistance of the flow before a flow inversion while buoyancy force becomes the driving force after a flow inversion. By taking into account the buoyancy effect to determine the flow rate, the difference in the flow inversion time between hot and average channels is investigated: a flow inversion occurs earlier in the hot channel than in an average channel. Furthermore, the movement of the hotspot location before and after a flow inversion is investigated for a slow and fast LOFA. During a flow inversion, two temperature peaks are observed: (1) the first temperature peak is at the initiation of the LOFA, and (2) the second temperature peak is when a flow inversion occurs. The maximum temperature of the cladding is found at the second temperature peak for both LOFA analyses, and is lower than the saturation temperature

Mathematical Modeling and Numerical Analysis of Thermal Distribution in Arch Dams considering Solar Radiation Effect

Science.gov (United States)

Mirzabozorg, H.; Hariri-Ardebili, M. A.; Shirkhan, M.; Seyed-Kolbadi, S. M.

2014-01-01

The effect of solar radiation on thermal distribution in thin high arch dams is investigated. The differential equation governing thermal behavior of mass concrete in three-dimensional space is solved applying appropriate boundary conditions. Solar radiation is implemented considering the dam face direction relative to the sun, the slop relative to horizon, the region cloud cover, and the surrounding topography. It has been observed that solar radiation changes the surface temperature drastically and leads to nonuniform temperature distribution. Solar radiation effects should be considered in thermal transient analysis of thin arch dams. PMID:24695817

Mathematical modeling and numerical analysis of thermal distribution in arch dams considering solar radiation effect.

Science.gov (United States)

Mirzabozorg, H; Hariri-Ardebili, M A; Shirkhan, M; Seyed-Kolbadi, S M

2014-01-01

The effect of solar radiation on thermal distribution in thin high arch dams is investigated. The differential equation governing thermal behavior of mass concrete in three-dimensional space is solved applying appropriate boundary conditions. Solar radiation is implemented considering the dam face direction relative to the sun, the slop relative to horizon, the region cloud cover, and the surrounding topography. It has been observed that solar radiation changes the surface temperature drastically and leads to nonuniform temperature distribution. Solar radiation effects should be considered in thermal transient analysis of thin arch dams.

A plan for the modification and assessment of TRAC-PF1/MOD2 for use in analyzing CANDU 3 transient thermal-hydraulic phenomena

International Nuclear Information System (INIS)

Siebe, D.A.; Boyack, B.E.; Giguere, P.T.

1994-11-01

This report presents the results of the review and planning done for the United States Nuclear Regulatory Commission to identify the thermal-hydraulic phenomena that could occur in the CANDU 3 reactor design during transient conditions, plan modifications to the TRAC-PF1/MOD2 (TRAC) computer code needed to adequately predict CANDU 3 transient thermal-hydraulic phenomena, and identify an assessment program to verify the ability of TRAC, when modified, to predict these phenomena. This work builds on analyses and recommendations produced by the Idaho National Engineering Laboratory (INEL). To identify the thermal-hydraulic phenomena, a large-break loss-of-coolant accident simulation, performed as part of earlier work by INEL with an Atomic Energy of Canada, Limited (AECL) thermal-hydraulic computer code (CATHENA), was analyzed in detail. Other accident scenarios were examined for additional phenomena. A group of Los Alamos National Laboratory reactor thermal-hydraulics experts ranked the phenomena to produce a preliminary phenomena identification and ranking table (PIRT). The preliminary nature of the PIRT was a result of a lack of direct expertise with the unique processes and phenomena of the CANDU 3. Nonetheless, this PIRT provided an adequate foundation for planning a program of code modifications. We believe that this PIRT captured the most important phenomena and that refinements to the PIRT will mainly produce clarification of the relative importance (ranking) of phenomena. A plan for code modifications was developed based on this PIRT and on information about the modeling methodologies for CANDU-specific phenomena used in AECL codes. AECL thermal-hydraulic test facilities and programs were reviewed and the information used in developing an assessment plan to ensure that TRAC-PF1/MOD2, when modified, will adequately predict CANDU 3 phenomena

CENTAR code for extended nonlinear transient analysis of extraterrestrial reactor systems

International Nuclear Information System (INIS)

Nassersharif, B.; Peer, J.S.; DeHart, M.D.

1987-01-01

Current interest in the application of nuclear reactor-driven power systems to space missions has generated a need for a systems simulation code to model and analyze space reactor systems; such a code has been initiated at Texas A and M, and the first version is nearing completion; release was anticipated in the fall of 1987. This code, named CENTAR (Code for Extended Nonlinear Transient Analysis of Extraterrestrial Reactor Systems), is designed specifically for space systems and is highly vectorizable. CENTAR is composed of several specialized modules. A fluids module is used to model fluid behavior throughout the system. A wall heat transfer module models the heat transfer characteristics of all walls, insulation, and structure around the system. A fuel element thermal analysis module is used to predict the temperature behavior and heat transfer characteristics of the reactor fuel rods. A kinetics module uses a six-group point kinetics formulation to model reactivity feedback and control and the ANS 5.1 decay-heat curve to model shutdown decay-heat production. A pump module models the behavior of thermoelectric-electromagnetic pumps, and a heat exchanger module models not only thermal effects in thermoelectric heat exchangers, but also predicts electrical power production for a given configuration. Finally, an accumulator module models coolant expansion/contraction accumulators

Dynamic thermal characteristics of heat pipe via segmented thermal resistance model for electric vehicle battery cooling

Science.gov (United States)

Liu, Feifei; Lan, Fengchong; Chen, Jiqing

2016-07-01

Heat pipe cooling for battery thermal management systems (BTMSs) in electric vehicles (EVs) is growing due to its advantages of high cooling efficiency, compact structure and flexible geometry. Considering the transient conduction, phase change and uncertain thermal conditions in a heat pipe, it is challenging to obtain the dynamic thermal characteristics accurately in such complex heat and mass transfer process. In this paper, a ;segmented; thermal resistance model of a heat pipe is proposed based on thermal circuit method. The equivalent conductivities of different segments, viz. the evaporator and condenser of pipe, are used to determine their own thermal parameters and conditions integrated into the thermal model of battery for a complete three-dimensional (3D) computational fluid dynamics (CFD) simulation. The proposed ;segmented; model shows more precise than the ;non-segmented; model by the comparison of simulated and experimental temperature distribution and variation of an ultra-thin micro heat pipe (UMHP) battery pack, and has less calculation error to obtain dynamic thermal behavior for exact thermal design, management and control of heat pipe BTMSs. Using the ;segmented; model, the cooling effect of the UMHP pack with different natural/forced convection and arrangements is predicted, and the results correspond well to the tests.

Transient behaviour of small HTR for cogeneration

International Nuclear Information System (INIS)

Verkerk, E.C.; Van Heek, A.I.

2000-01-01

The Dutch market for combined generation of heat and power identifies a unit size of 40 MW thermal for the conceptual design of a nuclear cogeneration plant. The ACACIA system provides 14 MWe electricity combined with 17 t/h of high temperature steam (220 deg C, 10 bar) with a pebble-bed high temperature reactor directly coupled with a helium compressor and a helium turbine. The design of this small CHP unit that is used for industrial applications is mainly based on a pre-feasibility study in 1996, performed by a joint working group of five Dutch organisations, in which technical feasibility was shown. Thermal hydraulic and reactor physics analyses show favourable control characteristics during normal operation and a benign response to loss of helium coolant and loss of flow conditions. Throughout the response on these highly infrequent conditions, ample margin exists between the highest fuel temperatures and the temperature above which fuel degradation will occur. To come to quantitative statements about the ACACIA transient behaviour, a calculational coupling between the high temperature reactor core analysis code package PANTHER/DIREKT and the thermal hydraulic code RELAP5 for the energy conversion system has been made. This coupling offers a more realistic simulation of the entire system, since it removes the necessity of forcing boundary conditions on the simulation models at the data transfer points. In this paper, the models used for the dynamic components of the energy conversion system are described, and the results of the calculation for two operational transients in order to demonstrate the effects of the interaction between reactor core and its energy conversion system are shown. Several transient cases that are representative as operational transients for an HTR will be discussed, including one representing a load rejection case that shows the functioning of the control system, in particular the bypass valve. Another transient is a load following

The coupled code system DORT-TD/THERMIX and its application to the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

International Nuclear Information System (INIS)

Pautz, A.; Tyobeka, B.; Ivanov, K.

2009-01-01

In new reactor designs that are still under review such as the Pebble Bed Modular Reactor (PBMR), not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400MW OECD/NEA coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate both the steady-state as well as several transient scenarios in this benchmark problem. (orig.)

Assessment of the turbine trip transient in Cofrentes NPP with TRAC-BF1

International Nuclear Information System (INIS)

Castrillo, F.; Gomez, A.; Gallego, I.

1993-06-01

This report presents the results of the assessment of TRAC-BF1 (G1-J1) code with the model of C. N. Cofrentes for simulation of the transient originated by the manual trip of the main turbine. C. N. Cofrentes is a General Electric designed BWR/6 plant, with a nominal core thermal power of 2894 Mwt, in commercial operation since 1985, owned and operated by Hidroelectrica Espanola, S. A. The plant incorporates all the characteristics of BWR/6 reactors, with two turbine driven FW pumps. As a result of this assessment a model of C. N. Cofrentes has been developed for TRAC-BF1 that fairly reproduces operational transient behavior of the plant. A special purpose code was generated to obtain reactivity coefficients, as required by TRAC-BF1, from the 3D simulator

Applications of the thermit code to 3D thermal hydraulic analysis of LWR cores

International Nuclear Information System (INIS)

Reed, W.H.

1979-01-01

The THERMIT code calculates the three-dimensional transient thermal hydraulic behavior of light water reactor cores. Its two-fluid dynamics equations for two-phase flow offer improved physical modelling capability needed in the context of calculation coupled to neutron kinetics for feedback. The numerical fluid dynamics method was chosen for reliability over a wider range of transients. An improved heat transfer numerical method is presented which gives better numerical stability and accuracy. A number of example calculations are discussed which give an idea of the power and flexibility of the code

Thermal Analysis of a SHIELD Electromigration Test Structure

Energy Technology Data Exchange (ETDEWEB)

Benson, David A.; Bowman, Duane J.; Mitchell, Robert T.

1999-05-01

The steady state and transient thermal behavior of an electromigration test structure was analyzed. The test structure was a Sandia SHIELD (Self-stressing HIgh fregquency rELiability Device) electromigration test device manufactured by an outside vendor. This device has a high frequency oscillator circuit, a buffer circuit to isolate and drive the metal line to the tested (DUT), the DUT to be electromigrated itself, a metal resistance thermometry monitor, and a heater elment to temperature accelerate the electromigration effect.

Empirical Validation of Heat Transfer Performance Simulation of Graphite/PCM Concrete Materials for Thermally Activated Building System

Directory of Open Access Journals (Sweden)

Jin-Hee Song

2017-01-01

Full Text Available To increase the heat capacity in lightweight construction materials, a phase change material (PCM can be introduced to building elements. A thermally activated building system (TABS with graphite/PCM concrete hollow core slab is suggested as an energy-efficient technology to shift and reduce the peak thermal load in buildings. An evaluation of heat storage and dissipation characteristics of TABS in graphite/PCM concrete has been conducted using dynamic simulations, but empirical validation is necessary to acceptably predict the thermal behavior of graphite/PCM concrete. This study aimed to validate the thermal behavior of graphite/PCM concrete through a three-dimensional transient heat transfer simulation. The simulation results were compared to experimental results from previous studies of concrete and graphite/PCM concrete. The overall thermal behavior for both materials was found to be similar to experiment results. Limitations in the simulation modeling, which included determination of the indoor heat transfer coefficient, assumption of constant thermal conductivity with temperature, and assumption of specimen homogeneity, led to slight differences between the measured and simulated results.

Thermal Fatigue Behavior of Air-Plasma Sprayed Thermal Barrier Coating with Bond Coat Species in Cyclic Thermal Exposure

Directory of Open Access Journals (Sweden)

Ungyu Paik

2013-08-01

Full Text Available The effects of the bond coat species on the delamination or fracture behavior in thermal barrier coatings (TBCs was investigated using the yclic thermal fatigue and thermal-shock tests. The interface microstructures of each TBC showed a good condition without cracking or delamination after flame thermal fatigue (FTF for 1429 cycles. The TBC with the bond coat prepared by the air-plasma spray (APS method showed a good condition at the interface between the top and bond coats after cyclic furnace thermal fatigue (CFTF for 1429 cycles, whereas the TBCs with the bond coats prepared by the high-velocity oxygen fuel (HVOF and low-pressure plasma spray (LPPS methods showed a partial cracking (and/or delamination and a delamination after 780 cycles, respectively. The TBCs with the bond coats prepared by the APS, HVOF and LPPS methods were fully delaminated (>50% after 159, 36, and 46 cycles, respectively, during the thermal-shock tests. The TGO thickness in the TBCs was strongly dependent on the both exposure time and temperature difference tested. The hardness values were found to be increased only after the CFTF, and the TBC with the bond coat prepared by the APS showed the highest adhesive strength before and after the FTF.

Measurements of thermal parameters of solar modules

International Nuclear Information System (INIS)

Górecki, K; Krac, E

2016-01-01

In the paper the methods of measuring thermal parameters of photovoltaic panels - transient thermal impedance and the absorption factor of light-radiation are presented. The manner of realising these methods is described and the results of measurements of the considered thermal parameters of selected photovoltaic panels are presented. The influence of such selected factors as a type of the investigated panel and its mounting manner on transient thermal impedance of the considered panels is also discussed. (paper)

Lipophilic phytosterol derivatives: synthesis, thermal property and nanoemulsion behavior

DEFF Research Database (Denmark)

Panpipat, Worawan; Xu, Xuebing; Guo, Zheng

Phytosterols and their esters have been reported as a cholesterol lowering agent in human. However, natural phytosterols have a low solubility in both water and fat resulting in a poor absorption in intestine. To improve the intestinal absorption and bioavailability of phytosterols, conversion...... of phytosterols into enzyme-liable lipophilic derivatives, such as fatty acid esters was one of the possible strategies. Differences in molecular structures of modified phytosterols may result in the differences in their thermal and micelling behaviors. Therefore, the objectives of this study were to improve...... the productive yield of a series of -sitosteryl fatty acid esters (C2-C18) and to investigate the thermal property and nano-emulsion behaviors of those compounds. This work reported a novel approach to synthesize phytosterol (-sitosterol as a model) fatty acid ester by employing Candida antarctica lipase...

Thermally induced dispersion mechanisms for aluminum-based plate-type fuels under rapid transient energy deposition

International Nuclear Information System (INIS)

Georgevich, V.; Taleyarkham, R.P.; Navarro-Valenti, S.; Kim, S.H.

1995-01-01

A thermally induced dispersion model was developed to analyze for dispersive potential and determine onset of fuel plate dispersion for Al-based research and test reactor fuels. Effect of rapid energy deposition in a fuel plate was simulated. Several data types for Al-based fuels tested in the Nuclear Safety Research Reactor in Japan and in the Transient Reactor Test in Idaho were reviewed. Analyses of experiments show that onset of fuel dispersion is linked to a sharp rise in predicted strain rate, which futher coincides with onset of Al vaporization. Analysis also shows that Al oxidation and exothermal chemical reaction between the fuel and Al can significantly affect the energy deposition characteristics, and therefore dispersion onset connected with Al vaporization, and affect onset of vaporization

Isolation stress and chronic mild stress induced immobility in the defensive burying behavior and a transient increased ethanol intake in Wistar rats.

Science.gov (United States)

Vázquez-León, Priscila; Martínez-Mota, Lucía; Quevedo-Corona, Lucía; Miranda-Páez, Abraham

2017-09-01

Stress can be experienced with or without adverse effects, of which anxiety and depression are two of the most important due to the frequent comorbidity with alcohol abuse in humans. Historically, stress has been considered a cause of drug use, particularly alcohol abuse due to its anxiolytic effects. In the present work we exposed male Wistar rats to two different stress conditions: single housing (social isolation, SI), and chronic mild stress (CMS). We compared both stressed groups to group-housed rats and rats without CMS (GH) to allow the determination of a clear behavioral response profile related to their respective endocrine stress response and alcohol intake pattern. We found that SI and CMS, to a greater extent, induced short-lasting increased sucrose consumption, a transient increase in serum corticosterone level, high latency/immobility, and low burying behavior in the defensive burying behavior (DBB) test, and a transient increase in alcohol intake. Thus, the main conclusion was that stress caused by both SI and CMS induced immobility in the DBB test and, subsequently, induced a transient increased voluntary ethanol intake in Wistar rats with a free-choice home-cage drinking paradigm. Copyright © 2017 Elsevier Inc. All rights reserved.

TRAC analyses of severe overcooling transients for the Oconee-1 PWR

Energy Technology Data Exchange (ETDEWEB)

Ireland, J R [comp.

1985-05-01

This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs.

TRAC analyses of severe overcooling transients for the Oconee-1 PWR

International Nuclear Information System (INIS)

Ireland, J.R.

1985-05-01

This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs

Effect of Thermal Cycling on the Tensile Behavior of CF/AL Fiber Metal Laminates

Directory of Open Access Journals (Sweden)

Muhammad Farhan Noor

2017-09-01

Full Text Available The objective of this research work was to estimate the effect of thermal cycling on the tensile behavior of CARALL composites. Fiber metal laminates (FMLs, based on 2D woven carbon fabric and 2024-T3 Alclad aluminum alloy sheet, was manufactured by pressure molding technique followed by hand layup method. Before fabrication, aluminum sheets were anodized with phosphoric acid to produce micro porous alumina layer on surface. This micro-porous layer is beneficial to produce strong bonding between metal and fiber surfaces in FMLs. The effect of thermal cycling (-65 to +70ºC on the tensile behavior of Cf/Al based FML was studied. Tensile strength was increased after 10 thermal cycles, but it was slightly decreased to some extent after 30, and 50 thermal cycles. Tensile modulus also shown the similar behavior as that of tensile strength.

Transient two-phase flow

International Nuclear Information System (INIS)

Hsu, Y.Y.

1974-01-01

The following papers related to two-phase flow are summarized: current assumptions made in two-phase flow modeling; two-phase unsteady blowdown from pipes, flow pattern in Laval nozzle and two-phase flow dynamics; dependence of radial heat and momentum diffusion; transient behavior of the liquid film around the expanding gas slug in a vertical tube; flooding phenomena in BWR fuel bundles; and transient effects in bubble two-phase flow. (U.S.)

Influence of different temperatures on the thermal fatigue behavior and thermal stability of hot-work tool steel processed by a biomimetic couple laser technique

Science.gov (United States)

Meng, Chao; Zhou, Hong; Zhou, Ying; Gao, Ming; Tong, Xin; Cong, Dalong; Wang, Chuanwei; Chang, Fang; Ren, Luquan

2014-04-01

Three kinds of biomimetic non-smooth shapes (spot-shape, striation-shape and reticulation-shape) were fabricated on the surface of H13 hot-work tool steel by laser. We investigated the thermal fatigue behavior of biomimetic non-smooth samples with three kinds of shapes at different thermal cycle temperature. Moreover, the evolution of microstructure, as well as the variations of hardness of laser affected area and matrix were studied and compared. The results showed that biomimetic non-smooth samples had better thermal fatigue behavior compared to the untreated samples at different thermal cycle temperatures. For a given maximal temperature, the biomimetic non-smooth sample with reticulation-shape had the optimum thermal fatigue behavior, than with striation-shape which was better than that with the spot-shape. The microstructure observations indicated that at different thermal cycle temperatures the coarsening degrees of microstructures of laser affected area were different and the microstructures of laser affected area were still finer than that of the untreated samples. Although the resistance to thermal cycling softening of laser affected area was lower than that of the untreated sample, laser affected area had higher microhardness than the untreated sample at different thermal cycle temperature.

Impact of the surface quality on the thermal shock performance of beryllium armor tiles for first wall applications

Energy Technology Data Exchange (ETDEWEB)

Spilker, B., E-mail: b.spilker@fz-juelich.de; Linke, J.; Pintsuk, G.; Wirtz, M.

2016-11-01

Highlights: • Different surface qualities of S-65 beryllium are tested under high heat flux conditions. • After 1000 thermal shocks, the loaded area exhibits a crucial destruction. • Stress accelerated grain boundary oxidation/dynamic embrittlement effects are linked to the thermal shock performance of beryllium. • Thermally induced cracks form between 1 and 10 pulses and grow wider and deeper between 10 and 100 pulses. • Thermally induced cracks form and propagate independently from surface grooves and the surface quality. - Abstract: Beryllium will be applied as first wall armor material in ITER. The armor has to sustain high steady state and transient power fluxes. For transient events like edge localized modes, these transient power fluxes rise up to 1.0 GW m{sup −2} with a duration of 0.5–0.75 ms in the divertor region and a significant fraction of this power flux is deposited on the first wall as well. In the present work, the reference beryllium grade for the ITER first wall application S-65 was prepared with various surface conditions and subjected to transient power fluxes (thermal shocks) with ITER relevant loading parameters. After 1000 thermal shocks, a crucial destruction of the entire loaded area was observed and linked to the stress accelerated grain boundary oxidation (SAGBO)/dynamic embrittlement (DE) effect. Furthermore, the study revealed that the majority of the thermally induced cracks formed between 1 and 10 pulses and then grew wider and deeper with increasing pulse number. The surface quality did not influence the cracking behavior of beryllium in any detectable way. However, the polished surface demonstrated the highest resistance against the observed crucial destruction mechanism.

Analytical expression of the thermal stresses in a vessel or pipe with cladding submitted to any thermal transient

International Nuclear Information System (INIS)

Marie, Stephane

2004-01-01

This article proposes an extension of the known analytical solution for the temperature and stresses in the event of a linear shock in a pipe containing a fluid. The intention is to propose a simple solution for any variation of the temperature in the fluid and to cover the influence of cladding on the inner surface. The approach consists of breaking down the fluid temperature variation into a succession of linear shocks. Using the linear shock resolution approach, it is possible to propose a simple analytical solution, using the same constant (Biot number B, etc.). The proposed solution is compared with finite element analysis: the solution is found to be reliable for any thermal shock or cyclic variation of fluid temperature, and can even replicate the transient regime. The following stage has made it possible to account for the effect of cladding on the inner surface of the piping on temperature distribution. The second part gives analytical expressions for the elastic stresses due to the temperature field alone

Thermal analysis of continuous and patterned multilayer films in the presence of a nanoscale hot spot

Science.gov (United States)

Juang, Jia-Yang; Zheng, Jinglin

2016-10-01

Thermal responses of multilayer films play essential roles in state-of-the-art electronic systems, such as photo/micro-electronic devices, data storage systems, and silicon-on-insulator transistors. In this paper, we focus on the thermal aspects of multilayer films in the presence of a nanoscale hot spot induced by near field laser heating. The problem is set up in the scenario of heat assisted magnetic recording (HAMR), the next-generation technology to overcome the data storage density limit imposed by superparamagnetism. We characterized thermal responses of both continuous and patterned multilayer media films using transient thermal modeling. We observed that material configurations, in particular, the thermal barriers at the material layer interfaces crucially impact the temperature field hence play a key role in determining the hot spot geometry, transient response and power consumption. With a representative generic media model, we further explored the possibility of optimizing thermal performances by designing layers of heat sink and thermal barrier. The modeling approach demonstrates an effective way to characterize thermal behaviors of micro and nano-scale electronic devices with multilayer thin film structures. The insights into the thermal transport scheme will be critical for design and operations of such electronic devices.

Experimental data showing the thermal behavior of a flat roof with phase change material.

Science.gov (United States)

Tokuç, Ayça; Başaran, Tahsin; Yesügey, S Cengiz

2015-12-01

The selection and configuration of building materials for optimal energy efficiency in a building require some assumptions and models for the thermal behavior of the utilized materials. Although the models for many materials can be considered acceptable for simulation and calculation purposes, the work for modeling the real time behavior of phase change materials is still under development. The data given in this article shows the thermal behavior of a flat roof element with a phase change material (PCM) layer. The temperature and energy given to and taken from the building element are reported. In addition the solid-liquid behavior of the PCM is tracked through images. The resulting thermal behavior of the phase change material is discussed and simulated in [1] A. Tokuç, T. Başaran, S.C. Yesügey, An experimental and numerical investigation on the use of phase change materials in building elements: the case of a flat roof in Istanbul, Build. Energy, vol. 102, 2015, pp. 91-104.

Simulation of the impact of wind power on the transient fault behavior of the Nordic power system

Energy Technology Data Exchange (ETDEWEB)

Jauch, Clemens; Soerensen, Poul [Risoe National Laboratory, Wind Energy Department, P.O. Box 49, DK-4000 Roskilde (Denmark); Norheim, Ian [SINTEF Energy Research, The department of Energy Systems, Sem Saelands Vei 11, NO-7463 Trondheim (Norway); Rasmussen, Carsten [Elkraft System, 2750 Ballerup (Denmark)

2007-02-15

In this paper the effect of wind power on the transient fault behavior of the Nordic power system is investigated. The Nordic power system is the interconnected power system of the countries Norway, Sweden, Finland and Denmark. For the purpose of these investigations the wind turbines installed and connected in eastern Denmark are taken as study case. The current and future wind power situation in eastern Denmark is modeled and short circuit faults in the system simulated. The simulations yield information on (i) how the faults impact on the wind turbines and (ii) how the response of the wind turbines influences the post-fault behavior of the Nordic power system. It is concluded that an increasing level of wind power penetration leads to stronger system oscillations in case of fixed speed wind turbines. It is found that fixed speed wind turbines that merely ride through transient faults have negative impacts on the dynamic response of the system. These negative impacts can be mitigated though, if sophisticated wind turbine control is applied. (author)

Dispersal, behavioral responses and thermal adaptation in Musca domestica

DEFF Research Database (Denmark)

Kjaersgaard, Anders; Blackenhorn, Wolf U.; Pertoldi, Cino

were obtained with flies held for several generations in a laboratory common garden setting, therefore we suggest that exposure to and avoidance of high temperatures under natural conditions has been an important selective agent causing the suggested adaptive differentiation between the populations.......Behavioral traits can have great impact on an organism’s ability to cope with or avoidance of thermal stress, and are therefore of evolutionary importance for thermal adaptation. We compared the morphology, heat resistance, locomotor (walking and flying) activity and flight performance of three...

Uncertainty and sensitivity analysis for the modeling of transients with interaction of thermal hydraulics and neutron kinetics

International Nuclear Information System (INIS)

Soeren Kliem; Siegfried Mittag; Siegfried Langenbuch

2005-01-01

Full text of publication follows: The transition from the application of conservative models to the use of best-estimate models raises the question about the uncertainty of the obtained results. This question becomes especially important, if the best-estimate models should be used for safety analyses in the field of nuclear engineering. Different methodologies were developed to assess the uncertainty of the calculation results of computer simulation codes. One of them is the methodology developed by Gesellschaft fuer Anlagenund Reaktorsicherheit (GRS) which uses the statistical code package SUSA. In the past, this methodology was applied to the calculation results of the advanced thermal hydraulic system code ATHLET. In the frame of the recently finished EU FP5 funded research project VALCO, that methodology was extended and successfully applied to different coupled code systems, including the uncertainty analysis for neutronics. These code systems consist of a thermal hydraulic system code and a 3D neutron kinetic core model. One of the code systems applied was ATHLET coupled with the Rossendorf kinetics code DYN3D. Two real transients at NPPs with VVER-type reactors documented within the VALCO project were selected for analyses. One was the load drop of one of two turbines to house load level at the Loviisa-1 NPP (VVER-440), the second was a test with the switching-off of one of two main feed water pumps at the VVER-1000 Balakovo-4 NPP. The current paper is dedicated to the different steps of the use and implementation of the GRS methodology to coupled code systems and to the assessment of the results obtained by the DYN3D/ATHLET code. Based on the relevant physical processes in both transients, lists of possible sources of uncertainties were compiled. They are specific for the two transients. Besides control parameters like control rod movement and thermal hydraulic parameters like secondary side pressure, mass flow rates, pressurizer sprayer and heater

A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector - 15271

International Nuclear Information System (INIS)

Saas, L.; Le Tellier, R.; Bajard, S.

2015-01-01

In this document, we present a simplified geometrical model (0D model) for both the in-core corium propagation transient and the characterization of the mode of corium transfer from the core to the vessel. A degraded core with a formed corium pool is used as an initial state. This initial state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate...). During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4

Thermal sensation and comfort with transient metabolic rates

DEFF Research Database (Denmark)

Goto, Tomonobu; Toftum, Jørn; Dear, R. d.

2002-01-01

This study investigated the effect on thermal perceptions and preferences of controlled metabolic excursions of various intensities (20%, 40%, 60% relative work load) and durations (3-30 min) imposed on subjects that alternated between sedentary activity and exercise on a treadmill. The thermal...... environment was held constant at a temperature corresponding to PMV=0 at sedentary activity. Even low activity changes of short duration (1 min at 20% relative work load) affected thermal perceptions. However, after circa 15 min of constant activity, subjective thermal responses approximated the steady...

Two-dimensional transient thermal analysis of a fuel rod by finite volume method

Energy Technology Data Exchange (ETDEWEB)

Costa, Rhayanne Yalle Negreiros; Silva, Mário Augusto Bezerra da; Lira, Carlos Alberto de Oliveira, E-mail: ryncosta@gmail.com, E-mail: mabs500@gmail.com, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

2017-07-01

One of the greatest concerns when studying a nuclear reactor is the warranty of safe temperature limits all over the system at all time. The preservation of core structure along with the constraint of radioactive material into a controlled system are the main focus during the operation of a reactor. The purpose of this paper is to present the temperature distribution for a nominal channel of the AP1000 reactor developed by Westinghouse Co. during steady-state and transient operations. In the analysis, the system was subjected to normal operation conditions and then to blockages of the coolant flow. The time necessary to achieve a new safe stationary stage (when it was possible) was presented. The methodology applied in this analysis was based on a two-dimensional survey accomplished by the application of Finite Volume Method (FVM). A steady solution is obtained and compared with an analytical analysis that disregard axial heat transport to determine its relevance. The results show the importance of axial heat transport consideration in this type of study. A transient analysis shows the behavior of the system when submitted to coolant blockage at channel's entrance. Three blockages were simulated (10%, 20% and 30%) and the results show that, for a nominal channel, the system can still be considerate safe (there's no bubble formation until that point). (author)

Transient waves in visco-elastic media

CERN Document Server

Ricker, Norman

1977-01-01

Developments in Solid Earth Geophysics 10: Transient Waves in Visco-Elastic Media deals with the propagation of transient elastic disturbances in visco-elastic media. More specifically, it explores the visco-elastic behavior of a medium, whether gaseous, liquid, or solid, for very-small-amplitude disturbances. This volume provides a historical overview of the theory of the propagation of elastic waves in solid bodies, along with seismic prospecting and the nature of seismograms. It also discusses the seismic experiments, the behavior of waves propagated in accordance with the Stokes wave

Effect of transient annealing on patterned CoFeB-based magnetic tunnel junctions

Energy Technology Data Exchange (ETDEWEB)

Wu, Kuo-Ming; Huang, Chao-Hsien; Lin, Shiao-Chi; Wu, Jong-Ching [Department of Physics and Taiwan SPIN Research Center, National Changhua University of Education, Changhua 50007 (China); Kao, Ming-Jer; Tsai, Ming-Jinn [Industrial Technology Research Institute, Hsinchu 31040 (China); Horng, Lance

2007-12-15

In this study, the transient annealing effect on the switching behavior of microstructured Co{sub 60}Fe{sub 20}B{sub 20}-based magnetic tunnel junctions has been studied through magnetoresistance measurements (R-H loop). Elliptical shape of devices with long/short axis of 4/2 micrometers was patterned out of sheet film stack of: Ta(20)/PtMn(15)/CoFeB(3)/Al(0.7)-oxide/CoFeB(2)/Ru(8)/Ta(40) (thickness unit in nanometers) after a conventional long time field cooling annealing. The transient annealing was then executed by sample loading into a furnace with pre-set temperatures ranging from 100 to 400 C for only 5 minutes in the absence of any external magnetic field. The vortex-like reverse of free layer in as-etched MTJ evidently changes to single-domain-like reverser after 200{proportional_to}250 C transient annealing. The magnetoresistance was found to increase with increasing annealing temperatures up to 265 C and then slowly decrease at higher annealing temperatures. The transient thermal annealing creates obvious efforts to repair magnetic properties of MTJ cell befor 265 C annealing and results in less damage at temperature of 350 C and 400 C. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Effect of transient annealing on patterned CoFeB-based magnetic tunnel junctions

International Nuclear Information System (INIS)

Wu, Kuo-Ming; Huang, Chao-Hsien; Lin, Shiao-Chi; Wu, Jong-Ching; Kao, Ming-Jer; Tsai, Ming-Jinn; Horng, Lance

2007-01-01

In this study, the transient annealing effect on the switching behavior of microstructured Co 60 Fe 20 B 20 -based magnetic tunnel junctions has been studied through magnetoresistance measurements (R-H loop). Elliptical shape of devices with long/short axis of 4/2 micrometers was patterned out of sheet film stack of: Ta(20)/PtMn(15)/CoFeB(3)/Al(0.7)-oxide/CoFeB(2)/Ru(8)/Ta(40) (thickness unit in nanometers) after a conventional long time field cooling annealing. The transient annealing was then executed by sample loading into a furnace with pre-set temperatures ranging from 100 to 400 C for only 5 minutes in the absence of any external magnetic field. The vortex-like reverse of free layer in as-etched MTJ evidently changes to single-domain-like reverser after 200∝250 C transient annealing. The magnetoresistance was found to increase with increasing annealing temperatures up to 265 C and then slowly decrease at higher annealing temperatures. The transient thermal annealing creates obvious efforts to repair magnetic properties of MTJ cell befor 265 C annealing and results in less damage at temperature of 350 C and 400 C. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Structural Design Optimization On Thermally Induced Vibration

International Nuclear Information System (INIS)

Gu, Yuanxian; Chen, Biaosong; Zhang, Hongwu; Zhao, Guozhong

2002-01-01

The numerical method of design optimization for structural thermally induced vibration is originally studied in this paper and implemented in application software JIFEX. The direct and adjoint methods of sensitivity analysis for thermal induced vibration coupled with both linear and nonlinear transient heat conduction is firstly proposed. Based on the finite element method, the structural linear dynamics is treated simultaneously with coupled linear and nonlinear transient heat structural linear dynamics is treated simultaneously with coupled linear and nonlinear transient heat conduction. In the thermal analysis model, the nonlinear heat conduction considered is result from the radiation and temperature-dependent materials. The sensitivity analysis of transient linear and nonlinear heat conduction is performed with the precise time integration method. And then, the sensitivity analysis of structural transient dynamics is performed by the Newmark method. Both the direct method and the adjoint method are employed to derive the sensitivity equations of thermal vibration, and there are two adjoint vectors of structure and heat conduction respectively. The coupling effect of heat conduction on thermal vibration in the sensitivity analysis is particularly investigated. With coupling sensitivity analysis, the optimization model is constructed and solved by the sequential linear programming or sequential quadratic programming algorithm. The methods proposed have been implemented in the application software JIFEX of structural design optimization, and numerical examples are given to illustrate the methods and usage of structural design optimization on thermally induced vibration

Safety analyses for transient behavior of plasma and in-vessel components during plasma abnormal events in fusion reactor

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

1997-01-01

Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6 s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several parametric studies are needed to comprehend the overpower transient including structure behavior, since many uncertainties are connected to the filed of the plasma physics. And, future work will need to discuss the burn control scenario considering confinement mode transition, system specifications, experimental plans and safety regulations, etc. to confirm the safety related to the plasma anomaly. (author)

Use of plant operating history to define transient loads

International Nuclear Information System (INIS)