The delay function in finite difference models for nuclear channels thermo-hydraulic transients
International Nuclear Information System (INIS)
Agazzi, A.
1977-01-01
The study of the thermo-hydraulic transients in a nuclear reactor core often requires a bi- or tri-dimensional mathematical simulation of a reactor channel. The equations involved are generally solved by means of finite-difference methods. The determination of the spatial mesh-width and the time interval is strongly conditioned by the necessity of a good accuracy in the description of the delay function which defines the transfer of thermal perturbations along the cooling channel. In this paper the effects of both space and time discretization on the delay function are considered and for the classical cases of inlet temperature step and ramp universal functions and diagrams are given in order to make possible the determination of optimal spatial mesh-width and time interval, once the requested accuracy of the model is fixed in advance
Pius, self-protective thermohydraulics transient without safety system intervention
International Nuclear Information System (INIS)
Fredell, J.; Bredolt, V.
1989-01-01
In this paper, the self-protective thermohydraulic feedback of the PIUS reactor system is illustrated by an in-depth discussion of one typical transient. The selected transient is an undetected total loss of feedwater in the complete absence of conventional safety system intervention. The reactor shuts itself down to residual power in two steps. First, the power decreases due to the strongly negative moderator temperature reactivity coefficient, and then a complete shutdown occurs by ingress of cold, highly borated water from the reactor pool. The transient is terminated without any harm to the fuel or paint systems
Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)
Energy Technology Data Exchange (ETDEWEB)
Agrawal, A.K.
1978-02-01
Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation.
Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)
International Nuclear Information System (INIS)
Agrawal, A.K.
1978-02-01
Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation
Dynamic thermo-hydraulic model of district cooling networks
International Nuclear Information System (INIS)
Oppelt, Thomas; Urbaneck, Thorsten; Gross, Ulrich; Platzer, Bernd
2016-01-01
Highlights: • A dynamic thermo-hydraulic model for district cooling networks is presented. • The thermal modelling is based on water segment tracking (Lagrangian approach). • Thus, numerical errors and balance inaccuracies are avoided. • Verification and validation studies proved the reliability of the model. - Abstract: In the present paper, the dynamic thermo-hydraulic model ISENA is presented which can be applied for answering different questions occurring in design and operation of district cooling networks—e.g. related to economic and energy efficiency. The network model consists of a quasistatic hydraulic model and a transient thermal model based on tracking water segments through the whole network (Lagrangian method). Applying this approach, numerical errors and balance inaccuracies can be avoided which leads to a higher quality of results compared to other network models. Verification and validation calculations are presented in order to show that ISENA provides reliable results and is suitable for practical application.
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)
International Nuclear Information System (INIS)
Bjoerndahl, Olof; Letzter, Adam; Marcinkiewicz, Jerzy; Segle, Peter
2007-03-01
Transient thermohydraulic events often control the design of piping systems in nuclear power plants. Water hammers due to valve closure, pressure transients caused by steam collapse and pipe break all result in structural loads that are characterised by a high frequency content. What also characterises these pressures/forces is the specific spatial and time dependence that is acting on the piping system and found in the wave propagation in the contained fluid. The aim with this project has been to develop recommendations for analysis of the stress response in piping systems subjected to thermohydraulic transients. Basis for this work is that the so called two-step-method is applied and that the structural response is calculated with modal superposition. Derived analysis criteria are based on the assumption that the associated volume strain energy in the wave propagation for the contained fluid may be well defined by a parameter, here called ε PN . The stress response in the piping system is assumed to be completely determined with certain accuracy for that part of the volume strain energy in the wave propagation associated with this parameter. A comprehensive work has been done to determine the accuracy in loadings calculated with RELAP5. Properties such as period elongation and associated spurious oscillations in the pressure wave transient have been investigated. Furthermore, has the characteristics of the artificial numerical damping in RELAP5 been identified. Based on desired accuracy of the thermohydraulic analysis together with knowledge about the duration of the thermohydraulic perturbation, the lowest upper frequency limit f Pipe , in the modal base that is required for the structure model is calculated. With perturbation is meant such as a valve closure. According to suggested criteria and with the upper frequency limit set, the essential parameters i) largest size of the elements in the structure model and ii) the largest applicable time step in the
Thermohydraulic tests in nuclear fuel model
International Nuclear Information System (INIS)
Ladeira, L.C.D.; Navarro, M.A.
1984-01-01
The main experimental works performed in the Thermohydraulics Laboratory of the NUCLEBRAS Nuclear Technology Development Center, in the field of thermofluodynamics are briefly described. These works include the performing of steady-state flow tests in single tube test sections, and the design and construction of a rod bundle test section, which will be also used for those kind of testes. Mention is made of the works to be performed in the near future, related to steady-state and transient flow tests. (Author) [pt
Development of a computer code for thermohydraulic analysis of a heated channel in transients
International Nuclear Information System (INIS)
Jafari, J.; Kazeminejad, H.; Davilu, H.
2004-01-01
This paper discusses the thermohydraulic analysis of a heated channel of a nuclear reactor in transients by a computer code that has been developed by the writer. The considered geometry is a channel of a nuclear reactor with cylindrical or planar fuel rods. The coolant is water and flows from the outer surface of the fuel rod. To model the heat transfer in the fuel rod, two dimensional time dependent conduction equations has been solved by combination of numerical methods, O rthogonal Collocation Method in radial direction and finite difference method in axial direction . For coolant modelling the single phase time dependent energy equation has been used and solved by finite difference method . The combination of the first module that solves the conduction in the fuel rod and a second one that solved the energy balance in the coolant region constitute the computer code (Thyc-1) to analysis thermohydraulic of a heated channel in transients. The Orthogonal collocation method maintains the accuracy and computing time of conventional finite difference methods, while the computer storage is reduced by a factor of two. The same problem has been modelled by RELAP5/M3 system code to asses the validity of the Thyc-1 code. The good agreement of the results qualifies the developed code
Development of real time visual evaluation system for sodium transient thermohydraulic experiments
International Nuclear Information System (INIS)
Tanigawa, Shingo
1990-01-01
A real time visual evaluation system, the Liquid Metal Visual Evaluation System (LIVES), has been developed for the Plant Dynamics Test Loop facility at O-arai Engineering Center. This facility is designed to provide sodium transient thermohydraulic experimental data not only in a fuel subassembly but also in a plant wide system simulating abnormal or accident conditions in liquid metal fast breeder reactors. Since liquid metal sodium is invisible, measurements to obtain experimental data are mainly conducted by numerous thermo couples installed at various locations in the test sections and the facility. The transient thermohydraulic phenomena are a result of complicated interactions among global and local scale three-dimensional phenomena, and short- and long-time scale phenomena. It is, therefore, difficult to grasp intuitively thermohydraulic behaviors and to observe accurately both temperature distribution and flow condition solely by digital data or various types of analog data in evaluating the experimental results. For effectively conducting sodium transient experiments and for making it possible to observe exactly thermohydraulic phenomena, the real time visualization technique for transient thermohydraulics has been developed using the latest Engineering Work Station. The system makes it possible to observe and compare instantly the experiment and analytical results while experiment or analysis is in progress. The results are shown by not only the time trend curves but also the graphic animations. This paper shows an outline of the system and sample applications of the system. (author)
International Nuclear Information System (INIS)
Mueller, R.G.
1987-06-01
Due to the strong influence of vapour bubbles on the nuclear chain reaction, an exact calculation of neutron physics and thermal hydraulics in light water reactors requires consideration of subcooled boiling. To this purpose, in the present study a dynamic model is derived from the time-dependent conservation equations. It contains new methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. Furthermore, it enables the complete two-phase flow region to be treated in a consistent manner. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement was reached. The results from the coupling of the new calculation model with a neutron kinetics program proved its suitability for the steady-state and transient calculation of reactor cores. (orig.) [de
Thermohydraulic modeling and simulation of breeder reactors
International Nuclear Information System (INIS)
Agrawal, A.K.; Khatib-Rahbar, M.; Curtis, R.T.; Hetrick, D.L.; Girijashankar, P.V.
1982-01-01
This paper deals with the modeling and simulation of system-wide transients in LMFBRs. Unprotected events (i.e., the presumption of failure of the plant protection system) leading to core-melt are not considered in this paper. The existing computational capabilities in the area of protected transients in the US are noted. Various physical and numerical approximations that are made in these codes are discussed. Finally, the future direction in the area of model verification and improvements is discussed
THYDE-P, PWR LOCA Thermohydraulic Transient Analysis
International Nuclear Information System (INIS)
Asahi, Yoshiro
2001-01-01
1 - Description of problem or function: THYDE-P1 analyzes the behaviour of LWR plants in response to various disturbances, including the thermal hydraulic transient following a break of the primary coolant pipe system, generally referred to as a loss-of-coolant-accident (LOCA). LOCA can be considered as the most critical condition for testing the methods and models for plant dynamics, since thermal hydraulic conditions in the system change drastically during the transient. THYDE-P is capable of a complete LOCA calculation from start to complete reflooding of the core by subcooled water. The program performs steady-state adjustment, which is complete in the sense that the steady state obtained is a set of exact solutions of all the transient equations without time derivatives, not only for plant hydraulics but also for all the other phenomena in the simulation of a PWR plant. THYDE-P2 contains among others the following improvements over THYDE-P1: (1) not only the mass and momentum equations but also the energy equation are included in the non-linear implicit scheme; (2) the valve model is implemented; (3) the relaxation equation for void fraction is theoretically derived; (4) vectorized programming is implemented; (5) both EM (evaluation mode) and BE (best estimate) calculations are possible. THYDE-W is an improved version of THYDE-P2 and contains the following additional features: (a) analysis of multiple number of disjoint loops is possible; (b) a control system simulation model is included; (c) the trip model has been improved; (d) heavy water is allowed as coolant; (e) the effect of drift flux is accounted for in the steady state calculation; (f) boron transport is included; (g) to obtain steady state loop heat balance, the option of adjusting the enthalpy distribution is prepared included in addition to that of adjusting heat exchanger areas; (h) to obtain steady state pressure distribution, three other options are prepared in addition to the original ones
Thermohydraulic model of WWER-1000 core
International Nuclear Information System (INIS)
Maroti, L.; Szabados, L.
1987-11-01
Safe and economic operation of the WWER-1000 type reactor requires more accurate calculation of the thermohydraulic processes than the one satisfactory for the 440 type cores. The high degree of accuracy is needed both for reactor physics calculations and for the determination of the operational safety limits of the core. The paper illustrates the most important differences between the 1000 and 440 type reactors and presents the main fields of the development work necessary to reach the required accuracy. A prediction for the capability of the computer programs after the proposed development is also given and some suggestions for the further improvement is outlined. (author) 7 refs
Modelling of thermohydraulic emergency core cooling phenomena
International Nuclear Information System (INIS)
Yadigaroglu, G.; Andreani, M.; Lewis, M.J.
1990-10-01
The codes used in the early seventies for safety analysis and licensing were based either on the homogeneous model of two-phase flow or on the so-called separate-flow models, which are mixture models accounting, however, for the difference in average velocity between the two phases. In both cases the behavior of the mixture is prescribed a priori as a function of local parameters such as the mass flux and the quality. The modern best-estimate codes used for analyzing LWR LOCA's and transients are often based on a two-fluid or 6-equation formulation of the conservation equations. In this case the conservation equations are written separately for each phase; the mixture is allowed to evolve on its own, governed by the interfacial exchanges of mass, momentum and energy between the phases. It is generally agreed that such relatively sophisticated 6-equation formulations of two-phase flow are necessary for the correct modelling of a number of phenomena and situations arising in LWR accidental situations. They are in particular indispensible for the analysis of stratified or countercurrent flows and of situations in which large departures from thermal and velocity equilibrium exist. This report will be devoted to a discussion of the need for, the capacity and the limitations of the two-phase flow models (with emphasis on the 6-equation formulations) in modelling these two-phase flow and heat transfer phenomena and/or different core cooling situations. 18 figs., 1 tab., 72 refs
International Nuclear Information System (INIS)
Wulff, W.
1977-01-01
A review is presented on the development of analyses and computer codes for the prediction of thermohydraulic transients in nuclear reactor systems. Models for the dynamics of two-phase mixtures are summarized. Principles of process, reactor component and reactor system modeling are presented, as well as the verification of these models by comparing predicted results with experimental data. Codes of major importance are described, which have recently been developed or are presently under development. The characteristics of these codes are presented in terms of governing equations, solution techniques and code structure. Current efforts and problems of code verification are discussed. A summary is presented of advances which are necessary for reducing the conservatism currently implied in reactor hydraulics codes for safety assessment
Thermo-Hydraulic Modelling of Buffer and Backfill
International Nuclear Information System (INIS)
Pintado, X.; Rautioaho, E.
2013-09-01
The temporal evolution of saturation, liquid pressure and temperature in the components of the engineered barrier system was studied using numerical methods. A set of laboratory tests was conducted to calibrate the parameters employed in the models. The modelling consisted of thermal, hydraulic and thermo-hydraulic analysis in which the significant thermo-hydraulic processes, parameters and features were identified. CODE B RIGHT was used for the finite element modelling and supplementary calculations were conducted with analytical methods. The main objective in this report is to improve understanding of the thermo-hydraulic processes and material properties that affect buffer behaviour in the Olkiluoto repository and to determine the parametric requirements of models for the accurate prediction of this behaviour. The analyses consisted of evaluating the influence of initial canister temperature and gaps in the buffer, and the role played by fractures and the rock mass located between fractures in supplying water for buffer and backfill saturation. In the thermo-hydraulic analysis, the primary processes examined were the effects of buffer drying near the canister on temperature evolution and the manner in which heat flow affects the buffer saturation process. Uncertainties in parameters and variations in the boundary conditions, modelling geometry and thermo-hydraulic phenomena were assessed with a sensitivity analysis. The material parameters, constitutive models, and assumptions made were carefully selected for all the modelling cases. The reference parameters selected for the simulations were compared and evaluated against laboratory measurements. The modelling results highlight the importance of understanding groundwater flow through the rock mass and from fractures in the rock in order to achieve reliable predictions regarding buffer saturation, since saturation times could range from a few years to tens of thousands of years depending on the hydrogeological
International Nuclear Information System (INIS)
Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko
2002-01-01
Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and
Simulation of thermohydraulic phenomena and model test for FBR
International Nuclear Information System (INIS)
Satoh, Kazuziro
1994-01-01
This paper summarizes the major thermohydraulic phenomena of FBRs and the conventional ways of their model tests, and introduces the recent findings regarding measurement technology and computational science. In the future commercial stage of FBRs, the design optimization will becomes important to improve economy and safety more and more. It is indispensable to use computational science to the plant design and safety evaluation. The most of the model tests will be replaced by the simulation analyses based on computational science. The measurement technology using ultrasonic and the numerical simulation with super parallel computing are considered to be the key technology to realize the design by analysis method. (author)
Proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic
International Nuclear Information System (INIS)
1986-01-01
The proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic - 6. ENFIR - allow to evaluate the present status of development in reactor physics and thermohydraulic fields. The mathematical models and methods for calculating neutronic of nuclear reactors, safety reactor analysis, measuring methods of neutronic parameters, computerized simulation of accidents, transients and thermohydraulic analysis are presented. (M.C.K.) [pt
Fine numerical modelling of thermohydraulic phenomena in EDF PWR reactors
International Nuclear Information System (INIS)
Boulot, F.
1993-01-01
Over the last 20 years, EDF has developed a family of 2D and 3D industrial thermohydraulics software to solve problems encountered in existing PWR power plants and to design new reactors for the future. The equations used in the models are the averaged Navier-Stokes and energy equations. A brief description is given of the four main codes developed for single-phase and two-phase water-steam flows, some of which use finite differences or finite volumes methods, while others make use of finite elements methods. An example of application is given for each code. (author). 4 figs., 4 refs
Trost, Nico; Jiménez, Javier; Imke, Uwe; Sanchez, Victor
2014-06-01
TWOPORFLOW is a thermo-hydraulic code based on a porous media approach to simulate single- and two-phase flow including boiling. It is under development at the Institute for Neutron Physics and Reactor Technology (INR) at KIT. The code features a 3D transient solution of the mass, momentum and energy conservation equations for two inter-penetrating fluids with a semi-implicit continuous Eulerian type solver. The application domain of TWOPORFLOW includes the flow in standard porous media and in structured porous media such as micro-channels and cores of nuclear power plants. In the latter case, the fluid domain is coupled to a fuel rod model, describing the heat flow inside the solid structure. In this work, detailed profiling tools have been utilized to determine the optimization potential of TWOPORFLOW. As a result, bottle-necks were identified and reduced in the most feasible way, leading for instance to an optimization of the water-steam property computation. Furthermore, an OpenMP implementation addressing the routines in charge of inter-phase momentum-, energy- and mass-coupling delivered good performance together with a high scalability on shared memory architectures. In contrast to that, the approach for distributed memory systems was to solve sub-problems resulting by the decomposition of the initial Cartesian geometry. Thread communication for the sub-problem boundary updates was accomplished by the Message Passing Interface (MPI) standard.
Numerical modeling of secondary side thermohydraulics of horizontal steam generator
Energy Technology Data Exchange (ETDEWEB)
Melikhov, V.I.; Melikhov, O.I.; Nigmatulin, B.I. [Research and Engineering Centre of LWR Nuclear Plants Safety, Moscow (Russian Federation)
1995-12-31
A mathematical model for the transient three-dimensional secondary side thermal hydraulics of the horizontal steam generator has been developed. The calculations of the steam generator PGV-1000 and PGV-4 nominal regimes and comparison of numerical and experimental results have been carried out. 7 refs.
Numerical modeling of secondary side thermohydraulics of horizontal steam generator
Energy Technology Data Exchange (ETDEWEB)
Melikhov, V I; Melikhov, O I; Nigmatulin, B I [Research and Engineering Centre of LWR Nuclear Plants Safety, Moscow (Russian Federation)
1996-12-31
A mathematical model for the transient three-dimensional secondary side thermal hydraulics of the horizontal steam generator has been developed. The calculations of the steam generator PGV-1000 and PGV-4 nominal regimes and comparison of numerical and experimental results have been carried out. 7 refs.
State space model extraction of thermohydraulic systems – Part I: A linear graph approach
International Nuclear Information System (INIS)
Uren, K.R.; Schoor, G. van
2013-01-01
Thermohydraulic simulation codes are increasingly making use of graphical design interfaces. The user can quickly and easily design a thermohydraulic system by placing symbols on the screen resembling system components. These components can then be connected to form a system representation. Such system models may then be used to obtain detailed simulations of the physical system. Usually this kind of simulation models are too complex and not ideal for control system design. Therefore, a need exists for automated techniques to extract lumped parameter models useful for control system design. The goal of this first paper, in a two part series, is to propose a method that utilises a graphical representation of a thermohydraulic system, and a lumped parameter modelling approach, to extract state space models. In this methodology each physical domain of the thermohydraulic system is represented by a linear graph. These linear graphs capture the interaction between all components within and across energy domains – hydraulic, thermal and mechanical. These linear graphs are analysed using a graph-theoretic approach to derive reduced order state space models. These models capture the dominant dynamics of the thermohydraulic system and are ideal for control system design purposes. The proposed state space model extraction method is demonstrated by considering a U-tube system. A non-linear state space model is extracted representing both the hydraulic and thermal domain dynamics of the system. The simulated state space model is compared with a Flownex ® model of the U-tube. Flownex ® is a validated systems thermal-fluid simulation software package. - Highlights: • A state space model extraction methodology based on graph-theoretic concepts. • An energy-based approach to consider multi-domain systems in a common framework. • Allow extraction of transparent (white-box) state space models automatically. • Reduced order models containing only independent state
TRACE/VALKIN: a neutronics-thermohydraulics coupled code to analyze strong 3D transients
Energy Technology Data Exchange (ETDEWEB)
Rafael Miro; Gumersindo Verdu; Ana Maria Sanchez [Chemical and Nuclear Engineering Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain); Damian Ginestar [Applied Mathematics Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain)
2005-07-01
Full text of publication follows: A nuclear reactor simulator consists mainly of two different blocks, which solve the models used for the basic physical phenomena taking place in the reactor. In this way, there is a neutronic module which simulates the neutron balance in the reactor core, and a thermal-hydraulics module, which simulates the heat transfer in the fuel, the heat transfer from the fuel to the water, and the different condensation and evaporation processes taking place in the reactor core and in the condenser systems. TRACE is a two-phase, two-fluid thermal-hydraulic reactor systems analysis code. The TRACE acronym stands for TRAC/RELAP Advanced Computational Engine, reflecting its ability to run both RELAP5 and TRAC legacy input models. It includes a three-dimensional kinetics module called PARCS for performing advanced analysis of coupled core thermal-hydraulic/kinetics problems. TRACE-VALKIN code is a new time domain analysis code to study transients in LWR reactors. This code uses the best estimate code TRACE to give account of the heat transfer and thermal-hydraulic processes, and a 3D neutronics module. This module has two options, the MODKIN option that makes use of a modal method based on the assumption that the neutronic flux can be approximately expanded in terms of the dominant lambda modes associated with a static configuration of the reactor core, and the NOKIN option that uses a one-step backward discretization of the neutron diffusion equation. The lambda modes are obtained using the Implicit Restarted Arnoldi approach or the Jacob-Davidson algorithm. To check the performance of the coupled code TRACE-VALKIN against complex 3D neutronic transients, using the cross-sections tables generated with the translator SIMTAB from SIMULATE to TRACE/VALKIN, the Cofrentes NPP SCRAM-61 transient is simulated. Cofrentes NPP is a General Electric BWR-6 design located in Valencia-land (Spain). It is in operation since 1985 and currently in its fifteenth
Energy Technology Data Exchange (ETDEWEB)
Huy, Ngo Quang; Khang, Ngo Phu; An, Tran Khac; Nghiem, Huynh Ton [Nuclear Research Inst., Da Lat (Viet Nam)
1994-10-01
Some experimental and theoretical thermal hydraulic characteristics of the Dalat Nuclear Research Reactor are presented, together with some general assessments, from a thermal hydraulic point of view, of its safety under transient conditions. (author). 3 refs., 9 figs., 7 tabs.
DRUFAN-01/MOD2, Transient Thermohydraulics of PWR Primary System LOCA
Energy Technology Data Exchange (ETDEWEB)
Burwell, M J; Lerchl, G; Steinhoff, F; Wolfert, K [Gesellschaft fuer Reaktorsicherheit (GRS) mbH, Forschungsgelaende, 8046 Garching (Germany)
1982-12-13
1 - Description of problem or function: DRUFAN is an advanced best estimate code for simulation of the transient thermal hydraulic behaviour during PWR-blowdown with large break size. 2 - Method of solution: The code is based on the lumped parameter approach and allows flexible control volume configurations. The physical model takes into account thermodynamic nonequilibrium. Using finite difference techniques a 1-dimensional representation of the discharge flow path including geometrical influences is possible. The physical model is based on separated field equations for liquid and vapour mass and overall field equations for energy and momentum. The mass transfer rates between phases during evaporation and condensation are based on correlations for the controlled growth and shrinkage of vapour bubbles or liquid droplets, respectively. A heat conductor model based on the energy transport equation is available for simulation of structures, electrical heater rods and fuel rods. For the heat transfer between solid structures and the fluid a comprehensive package of flow regime dependent heat transfer and critical heat flux correlations can be used. Simulation of components (valve, pressurizer, accumulator, pump, steam generator) is possible with functions or models. Power generation in solid structures may be simulated by an input time function, an electrical heater model or a neutron kinetics models. As a result of the lumped parameter approach a set of ordinary differential equations is obtained from the field equations. These equations, together with those resulting from the simulation of critical discharge flow near the outlet by a finite difference method, are solved by an explicit/implicit integration method with automatic time step, order and error control. The ordinary differential equations representing heat conductors are solved by an essentially implicit integration method. 3 - Restrictions on the complexity of the problem: - Vapour or liquid phase are
International Nuclear Information System (INIS)
Abbate, P.
1990-01-01
The CONVEC program developed for the thermohydraulic calculation under a natural convection regime for MTR type reactors is presented. The program is based on a stationary, one dimensional model of finite differences that allow to calculate the temperatures of cooler, cladding and fuel as well as the flow for a power level specified by the user. This model has been satisfactorily validated by a water cooling (liquid phase) and air system. (Author) [es
Development of a model for the primary system CAREM reactor's stationary thermohydraulic calculation
International Nuclear Information System (INIS)
Gaspar, C.; Abbate, P.
1990-01-01
The ESCAREM program oriented to CAREM reactors' stationary thermohydraulic calculation is presented. As CAREM gives variations in relation to models for BWR (Boiling Water Reactors)/PWR (Pressurized Water Reactors) reactors, it was decided to develop a suitable model which allows to calculate: a) if the Steam Generator design is adequate to transfer the power required; b) the circulation flow that occurs in the Primary System; c) the temperature at the entrance (cool branch) and d) the contribution of each component to the pressure drop in the circulation connection. Results were verified against manual calculations and alternative numerical models. An experimental validation at the Thermohydraulic Essays Laboratory is suggested. A parametric analysis series is presented on CAREM 25 reactor, demonstrating operating conditions, at different power levels, as well as the influence of different design aspects. (Author) [es
A benchmark for coupled thermohydraulics system/three-dimensional neutron kinetics core models
International Nuclear Information System (INIS)
Kliem, S.
1999-01-01
During the last years 3D neutron kinetics core models have been coupled to advanced thermohydraulics system codes. These coupled codes can be used for the analysis of the whole reactor system. Although the stand-alone versions of the 3D neutron kinetics core models and of the thermohydraulics system codes generally have a good verification and validation basis, there is a need for additional validation work. This especially concerns the interaction between the reactor core and the other components of a nuclear power plant (NPP). In the framework of the international 'Atomic Energy Research' (AER) association on VVER Reactor Physics and Reactor Safety, a benchmark for these code systems was defined. (orig.)
Thermohydraulics in rod bundles and critical heat flux in transient conditions in a tube
International Nuclear Information System (INIS)
Courtaud, M.; Roumy, R.
1975-01-01
After the determination of the scaling factor of Stevens's similitude for the pressure range of pressurized water vectors by comparison of critical heat flux data in from and in water, some examples of studies performed with freon are shown. The efficiency of the mixing vanes of spacer grids has been determined on the mixing phenomenon in single phase on critical heat flux. A calculation performed with the code FLICA using subchannel analysis on freon data transposed in water is in good agreement with the experiment. The influence of the number of spacer grids has been also shown. Critical heat fluxes have been determined in water at 140 bar in steady state and transient conditions on two tubular test sections. During the transient tests the flow rate was reduced by half in 0.5 seconds and the reincreased heat flux and inlet temperature remaining constant. These tests have shown the validity of the method which consists in using a critical heat flux correlation determined in steady state conditions applied with local transient conditions of enthalpy and mass velocity computed with the FLICA code [fr
FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients
International Nuclear Information System (INIS)
Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.
1984-01-01
1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry
A numerical method for a transient two-fluid model
International Nuclear Information System (INIS)
Le Coq, G.; Libmann, M.
1978-01-01
The transient boiling two-phase flow is studied. In nuclear reactors, the driving conditions for the transient boiling are a pump power decay or/and an increase in heating power. The physical model adopted for the two-phase flow is the two fluid model with the assumption that the vapor remains at saturation. The numerical method for solving the thermohydraulics problems is a shooting method, this method is highly implicit. A particular problem exists at the boiling and condensation front. A computer code using this numerical method allow the calculation of a transient boiling initiated by a steady state for a PWR or for a LMFBR
Solving linear systems in FLICA-4, thermohydraulic code for 3-D transient computations
International Nuclear Information System (INIS)
Allaire, G.
1995-01-01
FLICA-4 is a computer code, developed at the CEA (France), devoted to steady state and transient thermal-hydraulic analysis of nuclear reactor cores, for small size problems (around 100 mesh cells) as well as for large ones (more than 100000), on, either standard workstations or vector super-computers. As for time implicit codes, the largest time and memory consuming part of FLICA-4 is the routine dedicated to solve the linear system (the size of which is of the order of the number of cells). Therefore, the efficiency of the code is crucially influenced by the optimization of the algorithms used in assembling and solving linear systems: direct methods as the Gauss (or LU) decomposition for moderate size problems, iterative methods as the preconditioned conjugate gradient for large problems. 6 figs., 13 refs
PSH Transient Simulation Modeling
Energy Technology Data Exchange (ETDEWEB)
Muljadi, Eduard [National Renewable Energy Laboratory (NREL), Golden, CO (United States)
2017-12-21
PSH Transient Simulation Modeling presentation from the WPTO FY14 - FY16 Peer Review. Transient effects are an important consideration when designing a PSH system, yet numerical techniques for hydraulic transient analysis still need improvements for adjustable-speed (AS) reversible pump-turbine applications.
Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC
International Nuclear Information System (INIS)
Gomez Garcia-Torano, I.; Jimenez, G.
2013-01-01
The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.
Mathematical models of thermohydraulic disturbance sources in the NPP circuits
International Nuclear Information System (INIS)
Proskuryakov, K.N.
1999-01-01
Methods and means of diagnostics of equipment and processes at NPPs allowing one to substantially increase safety and economic efficiency of nuclear power plant operation are considered. Development of mathematical models, describing the occurrence and propagation of violations is conducted
Thermohydraulic modeling of nuclear thermal rockets: The KLAXON code
International Nuclear Information System (INIS)
Hall, M.L.; Rider, W.J.; Cappiello, M.W.
1992-01-01
The hydrogen flow from the storage tanks, through the reactor core, and out the nozzle of a Nuclear Thermal Rocket is an integral design consideration. To provide an analysis and design tool for this phenomenon, the KLAXON code is being developed. A shock-capturing numerical methodology is used to model the gas flow (the Harten, Lax, and van Leer method, as implemented by Einfeldt). Preliminary results of modeling the flow through the reactor core and nozzle are given in this paper
Elementary notions on one-dimensional flow thermohydraulic modelling
International Nuclear Information System (INIS)
Perrin, M.
1982-02-01
This paper is an overview of the notions of mathematical and simulation model applied to flows met in pipes and in several components of power plants (valves, pump, turbine). Finally, the results of a computer code based on the equations previously presented are given [fr
Tang, Jiajing; Yang, Xiaodong
2017-09-01
A novel thermo-hydraulic coupling model was proposed in this study to investigate the crater formation in electrical discharge machining (EDM). The temperature distribution of workpiece materials was included, and the crater formation process was explained from the perspective of hydrodynamic characteristics of the molten region. To better track the morphology of the crater and the movement of debris, the level-set method was introduced in this study. Simulation results showed that the crater appears shortly after the ignition of the discharge, and the molten material is removed by vaporizing in the initial stage, then by splashing at the following time. The driving force for the detachment of debris in the splashing removal stage comes from the extremely large pressure difference in the upper part of the molten region, and the morphology of the crater is also influenced by the shearing flow of molten material. It was found that the removal ratio of molten material is only about 7.63% under the studied conditions, leaving most to form the re-solidification layer on the surface of the crater. The size of the crater reaches the maximum at the end of discharge duration then experiences a slight reduction because of the reflux of molten material after the discharge. The results of single pulse discharge experiments showed that the morphologies and sizes between the simulation crater and actual crater are good at agreement, verifying the feasibility of the proposed thermo-hydraulic coupling model in explaining the mechanisms of crater formation in EDM.
EBaLM-THP - A neural network thermohydraulic prediction model of advanced nuclear system components
International Nuclear Information System (INIS)
Ridluan, Artit; Manic, Milos; Tokuhiro, Akira
2009-01-01
In lieu of the worldwide energy demand, economics and consensus concern regarding climate change, nuclear power - specifically near-term nuclear power plant designs are receiving increased engineering attention. However, as the nuclear industry is emerging from a lull in component modeling and analyses, optimization for example using ANN has received little research attention. This paper presents a neural network approach, EBaLM, based on a specific combination of two training algorithms, error-back propagation (EBP), and Levenberg-Marquardt (LM), applied to a problem of thermohydraulics predictions (THPs) of advanced nuclear heat exchangers (HXs). The suitability of the EBaLM-THP algorithm was tested on two different reference problems in thermohydraulic design analysis; that is, convective heat transfer of supercritical CO 2 through a single tube, and convective heat transfer through a printed circuit heat exchanger (PCHE) using CO 2 . Further, comparison of EBaLM-THP and a polynomial fitting approach was considered. Within the defined reference problems, the neural network approach generated good results in both cases, in spite of highly fluctuating trends in the dataset used. In fact, the neural network approach demonstrated cumulative measure of the error one to three orders of magnitude smaller than that produce via polynomial fitting of 10th order
International Nuclear Information System (INIS)
Delhaye, J.M.
2008-01-01
This scientific and technical handbook about PWR reactors thermohydraulics is the result of many years of teaching in the framework of the CEA-INSTN's atomic engineering training courses, in engineering schools and during continuing training activities. Its main goal is to present in a rigorous and pedagogical way the basic knowledge necessary for the understanding and modeling of single phase and two-phase thermohydraulic phenomena encountered during the design and operation of nuclear reactors. In particular, heat transfers in two-phase flows are presented in a detailed way. Most chapters include some nuclear engineering examples of application of the studied concepts, and some exercises aiming at mastering these concepts. Each example or exercise is accompanied by its detailed solution. Content: - thermohydraulic characteristics of reactors; - design and thermal dimensioning of reactors; - thermal engineering of the fuel element; - two-phase flow configurations in ducts; - recalls about single-phase flow equations; - basic equations for two-phase flows; - modeling of two-phase flows inside ducts; - pressure drops in ducts; - boiling and vapor condensation heat transfers; - two-phase flow instabilities in ducts; - two-phase flow blocking; thermohydraulics of naval propulsion reactors
International Nuclear Information System (INIS)
Kliem, S.
1998-01-01
The fifth dynamic benchmark was defined at seventh AER-Symposium, held in Hoernitz, Germany in 1997. It is the first benchmark for coupled thermohydraulic system/three-dimensional hexagonal neutron kinetic core models. In this benchmark the interaction between the components of a WWER-440 NPP with the reactor core has been investigated. The initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one control rod group stucking. This break causes an overcooling of the primary circuit. During this overcooling the scram reactivity is compensated and the scrammed reactor becomes re critical. The calculation was continued until the highly-borated water from the high pressure injection system terminated the power excursion. Each participant used own best-estimate nuclear cross section data. Only the initial subcriticality at the beginning of the transient was given. Solutions were received from Kurchatov Institute Russia with the code BIPR8/ATHLET, VTT Energy Finland with HEXTRAN/SMABRE, NRI Rez Czech Republic with DYN3/ATHLET, KFKI Budapest Hungary with KIKO3D/ATHLET and from FZR Germany with the code DYN3D/ATHLET.In this paper the results are compared. Beside the comparison of global results, the behaviour of several thermohydraulic and neutron kinetic parameters is presented to discuss the revealed differences between the solutions.(Authors)
Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD
Energy Technology Data Exchange (ETDEWEB)
Moreira, Uebert G.; Dominguez, Dany S. [Universidade Estadual de Santa Cruz (UESC), Ilh´eus, BA (Brazil). Programa de P´os-Graduacao em Modelagem Computacional em Ciencia e Tecnologia; Mazaira, Leorlen Y.R.; Lira, Carlos A.B.O. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Hernandez, Carlos R.G., E-mail: uebert.gmoreira@gmail.com, E-mail: dsdominguez@gmail.com, E-mail: leored1984@gmail.com, E-mail: cabol@ufpe.br, E-mail: cgh@instec.cu [Instituto Superior de Tecnologas y Ciencias Aplicadas (InSTEC), La Habana (Cuba)
2017-07-01
The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)
Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD
International Nuclear Information System (INIS)
Moreira, Uebert G.; Dominguez, Dany S.
2017-01-01
The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)
Energy Technology Data Exchange (ETDEWEB)
Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)
2015-05-15
NTR engines have continued as a main stream based on the mature technology. The typical core design of the NERVA derived engines uses hexagonal shaped fuel elements with circular cooling channels and structural tie-tube elements for supporting the fuel elements, housing moderator and regeneratively cooling the moderator. The state-of-the-art NTR designs mostly use a fast or epithermal neutron spectrum core utilizing a HEU fuel to make a high power reactor with small and simple core geometry. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The KANUTER-HEU and -LEU are the innovative and futuristic NTR engines to reduce the reactor size and to implement a LEU fuel in the reactor by using thermal neutron spectrum. The KANUTERs have some features in the reactor design such as the integrated fuel element and the regeneratively cooling channels to increase room for moderator and heat transfer in the core, and ensuing rocket performance. To study feasible design points in terms of thermo-hydraulics and to estimate rocket performance of the KANUTERs, the NSES is under development. The model of the NSES currently focuses on thermo-hydraulic analysis of the peculiar and complex EHTGR design during the propulsion mode in steady-state. The results indicate comparable performance for future applications, even though it uses the heavier LEU fuel. In future, the NSES will be modified to obtain temperature distribution of the entire reactor components and then more extensive design analysis of neutronics, thermohydraulics and their coupling will be conducted to validate design feasibility and to optimize the reactor design enhancing the rocket performance.
Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies
International Nuclear Information System (INIS)
Sorokin, G.A.; Avdeev, E.F.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P.
2000-01-01
An investigation into the reactor core accident cooling, which are associated with the power grow up or switch off circulation pumps in the event of the protective equipment comes into action, results in the problem of liquid metal boiling heat transfer. Considerable study has been given over the last 30 years to alkaline metal boiling including researches of heat transfer, boiling patterns, hydraulic resistance, crisis of heat transfer, initial heating up, boiling onset and instability of boiling. The results of these investigations have shown that the process of liquid metal boiling has substantial features in comparison with water boiling. Mathematical modeling of two phase flows in fast reactor fuel subassemblies have been developed intensively. Significant success has been achieved in formulation of two phase flow through the pin bundle and in their numerical realization. Currently a set of codes for thermohydraulic analysis of two phase flows in fast reactor subassembly have been developed with 3D macrotransfer governing equations. These codes are used for analysis of boiling onset and liquid metals boiling in fuel subassemblies during loss-of-coolant accidents, of warming up of reactor core, of blockage of some part of flow cross section in fuel subassembly. (author)
International Nuclear Information System (INIS)
Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.
1998-03-01
The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de
International Nuclear Information System (INIS)
Conti, C.F.S.; Silva Galetti, M.R. da.
1990-02-01
As Furnas intends to assume in the future the responsibility of performing Safety Analyses associated to Reload and Operation questions to Angra 1, it was figured out the necessity of qualifying its methodology by CNEN. The Methodology Qualification Process is based on guidelines proposed by CNEN at NT-DR-N o 02/87, where it was divided in four steps. This Technical Note aims to present the follow up of FURNAS Methodology Qualification Process and to bring it up to date in the areas of Core Physics (Neutronics), Core Thermal-Hydraulics, Fuel Rod Behaviour, Transient and Large Break Loss of Coolant Accident Analyses (LBLOCA). (author)
Assessment of TRAC-PD2 reflood core thermo-hydraulic model by CCTF Test C1-16
International Nuclear Information System (INIS)
Sugimoto, Jun
1982-11-01
The TRAC-PD2 reflood core thermo-hydraulic model was assessed by CCTF Test C1-16. The measured data were utilized as core boundary conditions in the TRAC calculations. The results indicate that the core inlet liquid temperature and the core heater rod temperatures are in reasonable agreement with data, but the pressure distribution in the core and water pool formation in the upper plenum are not in good agreement. The parametric effects of the droplet critical Weber number, the material properties of the heater rod, the noding of the upper plenum, and the minimum stable film boiling temperature are also discussed. (author)
Energy Technology Data Exchange (ETDEWEB)
Gamez, Abel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy Y.; Gonzalez, Daniel; Garcia, Carlos, E-mail: agamezgmf@gmail.com, E-mail: leored1984@gmail.com, E-mail: jrosales@instec.cu, E-mail: lcastro@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgr@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Oliveira, Carlos B. de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil)
2015-07-01
The high temperature gas cooled reactor (HTGR) is one of candidates of next generation of nuclear reactor according to IAEA report 2013. Evaluation of thermohydraulic performance and an experimental comparison results were proposed to the international research community. In this article, the tree dimensional CFD thermohydraulic modelation of steady state of HTR-10 modular reactor, using ANSYS CFX v14.0, has been done. Code-to-code and Code-to-experiment benchmark analyses, related to the testing program of the HTR-10 plant including steady state temperature distribution with the reactor at full power, were developed. The 3D real scale representation of reflector zone and fluid path flow inner and outer reflector blocks and cold helium cavity were carried out. The porous medium model was used to simulate the core zone in the reactor. The power distribution of the initial core published by IAEA-TECDOC-1694 obtained by Chief Scientific Investigators (CSIs) from China was used as heat sources in the core zone. (author)
LMFR core thermohydraulics: Status and prospects
International Nuclear Information System (INIS)
2000-06-01
One of the fundamental steps for a successful reactor core thermohydraulic design is the capability to predict, reliably and accurately, the temperature distribution in the core assemblies. A detailed knowledge of the assembly and fuel pin thermohydraulic behaviour in the steady state and transient conditions is an indispensable prerequisite to safe and stable operation of the reactor. Considerable experimental and theoretical studies on various aspects of LMFR core thermohydraulics are necessary to acquire such knowledge. During the last decade, there have been substantial advances in fast reactor core thermohydraulic design and operation in several countries with fast reactor programmes (notably in France, the Russian Federation, Japan, the United Kingdom, Germany and the United States of America). Chief among these has been the demonstration of reliable operation of reactor cores at a high burnup. During the last years, some additional countries such as China, India and the Republic of Korea have launched new fast reactor programmes. International exchange of information and experience on LMFR development including core thermohydraulic design is becoming of increasing importance to these countries. It is with this focus that the IAEA convened the Technical Committee on 'Methods and Codes for Calculations of Thermohydraulic Parameters for Fuel, Absorber Pins and Assemblies of LMFR's with Traditional and Burner Cores'. This meeting, attended by participants from seven countries, brought together a group of international experts to review and discuss the thermohydraulic advances and design approaches providing a reliable, safe and robust reactor core, as well as to exchange the experience accumulated in different countries of using the codes for thermohydraulic calculations and to discuss the issues requiring further research and development. A total of thirty technical papers presented covered theoretical and computational issues as well as experiments under
Directory of Open Access Journals (Sweden)
Farkas Istvan
2017-01-01
Full Text Available This paper focuses on the validation and applicability of CFD to simulate and analyze the thermo-hydraulic consequences of a main steam line break. Extensive validation data come from experiments performed using the Rossendorf coolant mixing model facility. For the calculation, the range of 9 to 12 million hexahe¬dral cells was constructed to capture all details in the interrogation domain in the system. The analysis was performed by running a time-dependent calculation, Detailed analyses were made at different cross-sections in the system to evaluate not only the value of the maximum and minimum temperature, but also the loca¬tion and the time at which it occurs during the transient which is considered to be indicator for the quality of mixing in the system. CFD and experimental results were qualitatively compared; mixing in the cold legs with emergency core cooling systems was overestimated. This could be explained by the sensitivity to the bound¬ary conditions. In the downcomer, the experiments displayed higher mixing: by our assumption this related to the dense measurement grid (they were not modelled. The temperature distribution in the core inlet plane agreed with the measurement results. Minor deviations were seen in the quantitative comparisons: the maximum temperature difference was 2ºC.
Papadimitriou, P.; Skorek, T.
THESUS is a thermohydraulic code for the calculation of steady state and transient processes of two-phase cryogenic flows. The physical model is based on four conservation equations with separate liquid and gas phase mass conservation equations. The thermohydraulic non-equilibrium is calculated by means of evaporation and condensation models. The mechanical non-equilibrium is modeled by a full-range drift-flux model. Also heat conduction in solid structures and heat exchange for the full spectrum of heat transfer regimes can be simulated. Test analyses of two-channel chilldown experiments and comparisons with the measured data have been performed.
International Nuclear Information System (INIS)
Kornienko, Y.; Kornienko, E.; Ninokata, H.
2001-01-01
One-dimensional mathematical models are extensively used in thermohydraulics assessment of Nuclear Power Plant (NPP) transients and accidents, because specifically 1-D system of the conservation laws allows to reduce computing time and required memory, especially in ''best estimate'' code calculations. This work is generalization of the well-known Zuber-Findley and Hancox-Nicoll methods for two-phase flow distribution parameters Cs taking into account the non-monotonous void fraction distribution in the transverse direction in terms of two superimposed monotonous profiles. The method is very useful in evaluating the saddle-shape void fraction profile effects. In this work two-phase flow distribution parameters Cs were developed for simple circular and rectangular pipes, and subchannel geometry in a rod bundle. Basic assumptions were power-mode approximations for describing the profiles of local volume flux density, phase velocity and temperature. The general analytical (quadrature) relationships for Cs were obtained and their 3-D illustrations are proposed. Also, we propose generalized formulation and simple approach to construct friction factor, heat and mass transfer coefficients within the gradient hypothesis and boundary layer assumptions. The contribution of momentum, heat and mass transfer as well as their sources and sinks in the channel cross-section are taken into account. In the same way, the friction factor, heat and mass transfer coefficients with the transversal and azimuthal variations being taken into account are proposed for subchannel geometry as well. (author)
Simulation Model of a Transient
DEFF Research Database (Denmark)
Jauch, Clemens; Sørensen, Poul; Bak-Jensen, Birgitte
2005-01-01
This paper describes the simulation model of a controller that enables an active-stall wind turbine to ride through transient faults. The simulated wind turbine is connected to a simple model of a power system. Certain fault scenarios are specified and the turbine shall be able to sustain operati...
Seminar on experimental thermohydraulics
International Nuclear Information System (INIS)
Katsaounis, A.
1979-06-01
Considerations on reactor safety are made. Problems related to the project and assembling of facilities for experiments in steady state and transient conditions are discussed. The advantages of using model fluids are finally, analysed. (Author) [pt
Thermohydraulic analysis of pressurized water reactors
International Nuclear Information System (INIS)
Veloso, M.A.
1980-01-01
The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt
Modelling of an ULOF transient in a sodium fast reactor
International Nuclear Information System (INIS)
Droin, Jean-Baptiste
2016-01-01
particular attention is paid to stabilize boiling occurrence which leads to minimize severe accident consequences.The phenomena occurring during the various ULOF transients are modelled in accordance to the level of details required to catch all the possible bifurcations of the transient. The tool coupled different (2D, 1D and 0D) models of thermics, thermo-hydraulics, core degradation (material melting and motions) and neutronics. The assumptions associated to these models are highlighted, discussed and validated. The physical tool capability of simulating the various realistic ULOF transients (without boiling, with stabilized boiling or flow excursion after boiling) is demonstrated by comparisons to experimental results (GR19, SCARABEE experiments) and to mechanistic simulations (CATHARE2 and SIMMER III).Parametric studies are then carried out on two variables: the fuel burn-up and the model of neutronic feedbacks. They underline the important influence of these parameters on the transient and the final core state. Finally, a preliminary sensitivity analysis (2000 simulations) is performed on 26 uncertain parameters (linked to initial core configuration, accident features, model uncertainties and radial nodalization). The variability of the final core state is underlined and quantified; only around 25% of cases lead to core degradation. The main influent parameters on transient phenomena are also identified, enabling to prioritize core design and safety studies. In the future, this tool will be used for safety-informed design and stability analyses of fast reactor systems, allowing to emphasize the main dominant phenomena and trends of significance for safety assessment. (author) [fr
Energy Technology Data Exchange (ETDEWEB)
Gomez Garcia-Torano, I.; Jimenez, G.
2013-07-01
The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.
THERMOSS: A thermohydraulic model of flow stagnation in a horizontal fuel channel
International Nuclear Information System (INIS)
Gulshani, P.; Caplan, M.Z.; Spinks, N.J.
1984-01-01
A model, called THERMOSS, is developed to compute the duration of stagnation in a CANDU reactor fuel channel with subcooled, stagnant initial conditions. The model solves, in closed form, the one dimensional, two-fluid conservation equations. In the computation of the duration of stagnation, the channel water level is an important intermediate variable because it determines the amount of steam production. A feature of the model is that water level is determined by a momentum balance between frictional pressure drop in the steam phase and hydrostatic head in the liquid phase. This is in contrast to an ealier model in which the level was determined from mass balance considerations. A satisfactory agreement between the predicted and experimentally observed channel water level and duration of stagnation is obtained. (orig.)
International Nuclear Information System (INIS)
Gashenko, I.V.; Melikhov, O.I.; Shmal, I.I.; Kouznetsov, V.D.
2001-01-01
The results of the calculational study using the RELAP5/MOD3.2 thermalhydraulic code performed on the influence of the heat losses to the ambient and the heat accumulated in the pipelines walls upon the evolution of the thermalhydraulic processes in the primary circuit of the integral test facility ISB-WWER when simulating the transients caused by the loss of the coolant are presented in the paper. (authors)
Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA
International Nuclear Information System (INIS)
Sugimoto, Jun
1989-01-01
This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)
Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall
International Nuclear Information System (INIS)
Kim, Byoung-Yoon; Ahn, Hee-Jae
2011-01-01
The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.
Modeling of Transients in an Enrichment Circuit
International Nuclear Information System (INIS)
Fernandino, Maria; Delmastro, Dario; Brasnarof, Daniel
2003-01-01
In the present work a mathematical model is presented in order to describe the dynamic behavior inside a closed enrichment loop, the latter representing a single stage of an uranium gaseous diffusion enrichment cascade.The analytical model is turned into a numerical model, and implemented through a computational code.Transients of two species separation were numerically analyzed, including setting times of each magnitude, behavior of each one of them during different transients, and redistribution of concentrations along the closed loop
Energy Technology Data Exchange (ETDEWEB)
Vianna Neto, Armando M.; Santos, Arnaldo M.; Mercon, Eduardo G. [TRANSPETRO - PETROBRAS Transportes, Rio de Janeiro, RJ (Brazil)
2003-07-01
This paper presents the development of an integrated simulation model, for the numerical calculation of thermal-hydraulic behaviors in the Brazilian southeast onshore gas pipeline flow system, remotely operated by TRANSPETRO's Gas Pipeline Control Centre (CCG). In its final application, this model is supposed to provide simulated results at the closer range to reality, in order to improve gas pipeline simulation studies and evaluations for the system in question. Considering the fact that numerical thermo-hydraulic simulation becomes the CCG's most important tool to analyze the boundary conditions to adjust the mentioned gas flow system, this paper seeks and takes aim to the optimization of the following prime attributions of a gas pipeline control centre: verification of system behaviors, face to some unit maintenance stop or procedure, programmed or not, or to some new gas outlet or inlet connection to the system; daily operational compatibility analysis between programmed and realized gas volumes; gas technical expedition and delivery analysis. Finally, all this work was idealized and carried out within the one-phase flow domain (dry gas) (author)
Library thermohydraulic components for training simulators
International Nuclear Information System (INIS)
Castelao Caruana, M. J.; Di Benedetto, A.; Pierini, J.P.
2013-01-01
The thermohydraulic components Library was modeled in MatLab/Simulink®. This library owns Pipe type components (pump, control valve and / or heaters), storage tanks (Open, Closed and Equilibrium Water Vapor-Air) and Heat Exchangers (Co-Current, Counter-Current and U-tubes). Each component can be attached to other components through the component library Header, in order to create a more complex thermal-hydraulic system which in turn can interact with other thermal-hydraulic systems. (author)
CAREM reactor thermohydraulic essays laboratory
International Nuclear Information System (INIS)
Horro, R.; Mazzi, R.; Rossini, A.
1990-01-01
The main characteristics, essays projected and the present state of the Thermohydraulic Essays Laboratory -under construction at present- prepared to meet the experimental needs resulting from a power reactor design of the CAREM type, are herein described. (Author) [es
A Semi-implicit Numerical Scheme for a Two-dimensional, Three-field Thermo-Hydraulic Modeling
International Nuclear Information System (INIS)
Hwang, Moonkyu; Jeong, Jaejoon
2007-07-01
The behavior of two-phase flow is modeled, depending on the purpose, by either homogeneous model, drift flux model, or separated flow model, Among these model, in the separated flow model, the behavior of each flow phase is modeled by its own governing equation, together with the interphase models which describe the thermal and mechanical interactions between the phases involved. In this study, a semi-implicit numerical scheme for two-dimensional, transient, two-fluid, three-field is derived. The work is an extension to the previous study for the staggered, semi-implicit numerical scheme in one-dimensional geometry (KAERI/TR-3239/2006). The two-dimensional extension is performed by specifying a relevant governing equation set and applying the related finite differencing method. The procedure for employing the semi-implicit scheme is also described in detail. Verifications are performed for a 2-dimensional vertical plate for a single-phase and two-phase flows. The calculations verify the mass and energy conservations. The symmetric flow behavior, for the verification problem, also confirms the momentum conservation of the numerical scheme
The TOPFLOW multi-purpose thermohydraulic test facility
International Nuclear Information System (INIS)
Schaffrath, Andreas; Kruessenberg, A.-K.; Weiss, F.-P.; Prasser, H.-M.
2002-01-01
The TOPFLOW (Transient Two Phase Flow Test Facility) multi-purpose thermohydraulic test facility is being built for studies of steady-state and transient flow phenomena in two-phase flows, and for the development and validation of the models contained in CFD (Computational Fluid Dynamics) codes. The facility is under construction at the Institute for Safety Research of the Rossendorf Research Center (FZR). It will be operated together with the Dresden Technical University and the Zittau/Goerlitz School for Technology, Economics and Social Studies within the framework of the Nuclear Technology Competence Preservation Program. TOPFLOW, with its test sections and its flexible concept, is available as an attractive facility also to users from all European countries. Experiments are planned in these fields, among others: - Transient two-phase flows in vertical and horizontal pipes and pipes of any inclination as well as in geometries typical of nuclear reactors (annulus, hot leg). - Boiling in large vessels and water pools (measurements of steam generation, 3D steam content distribution, turbulence, temperature stratification). - Test of passive components and safety systems. - Condensation in horizontal pipes in the absence and presence of non-condensable gases. The construction phase of TOPFLOW has been completed more or less on schedule. Experiments can be started after a commissioning phase in the 3rd quarter of 2002. (orig.) [de
International Nuclear Information System (INIS)
Tian Wenxi; Qiu Suizheng; Su Guanghui; Jia Dounan; Liu Xingmin; Zhang Jianwei
2007-01-01
A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules
DYNREL - the reference calculation (coupled code utilization on analysis of RIA-transient)
International Nuclear Information System (INIS)
Strmensky, C.; Darilek, P.
2003-01-01
DYNREL is coupled code, comprising DYN3D and RELAP5 programs. The coupled code has been developed during four years. Now DYNREL is tested on selected RIA and thermo-hydraulic transient calculations. This material describes some results from selected RIA transient calculation (initiated by control rod movement). DYNREL modelled the whole nuclear reactors. The core is modeled as 313 or 349 independent thermo-hydraulic channels with 10 or 20 axial layers. Thermo-hydraulic part contains about 700 components that covered the six loops' model of nuclear power plant in detail. The calculated results are compared with DYN3D/M3, DYN3D/H1.1 results (Authors)
International Nuclear Information System (INIS)
Uren, Kenneth Richard; Schoor, George van
2013-01-01
This second paper in a two part series presents the application of a developed state space model extraction methodology applied to a Brayton cycle-based PCU (power conversion unit) of a PBMR (pebble bed modular reactor). The goal is to investigate if the state space extraction methodology can cope with larger and more complex thermohydraulic systems. In Part I the state space model extraction methodology for the purpose of control was described in detail and a state space representation was extracted for a U-tube system to illustrate the concept. In this paper a 25th order nonlinear state space representation in terms of the different energy domains is extracted. This state space representation is solved and the responses of a number of important states are compared with results obtained from a PBMR PCU Flownex ® model. Flownex ® is a validated thermo fluid simulation software package. The results show that the state space model closely resembles the dynamics of the PBMR PCU. This kind of model may be used for nonlinear MIMO (multi-input, multi-output) type of control strategies. However, there is still a need for linear state space models since many control system design and analysis techniques require a linear state space model. This issue is also addressed in this paper by showing how a linear state space model can be derived from the extracted nonlinear state space model. The linearised state space model is also validated by comparing the state space model to an existing linear Simulink ® model of the PBMR PCU system. - Highlights: • State space model extraction of a pebble bed modular reactor PCU (power conversion unit). • A 25th order nonlinear time varying state space model is obtained. • Linearisation of a nonlinear state space model for use in power output control. • Non-minimum phase characteristic that is challenging in terms of control. • Models derived are useful for MIMO control strategies
Containment severe accident thermohydraulic phenomena
International Nuclear Information System (INIS)
Frid, W.
1991-08-01
This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)
International Nuclear Information System (INIS)
Konovalyuk, L.N.; Shevelev, D.V.; Kravchenko, V.G.
2003-01-01
PRZ model is proposed which allows taking into account in pressurizer convective heat- and mass transfer influence effects at the transients in VVER (PWR) Type Reactors case when calculations performed with using 1D thermohydraulic codes. The theoretical backgrounds are given to define the transients with the convective coolant instability in PRZ. The instability threshold is given for real PRZ geometry
Modeling transient radiation effects in power MOSFETS
International Nuclear Information System (INIS)
Hoffman, J.R.; Hall, W.E.; Dunn, D.E.
1987-01-01
Using standard device specifications and simple assumptions, the transient radiation response of VDMOS MOSFETs can be modeled in a standard circuit analysis program. The device model consists of a body diode, a parasitic bipolar transistor, and elements to simulate high-current reduced breakdown. The attached photocurrent model emulates response to any pulse shape and accounts for bias-dependent depletion regions. The model can be optimized to best fit available test data
Study of Transients in an Enrichment Closed Loop
International Nuclear Information System (INIS)
Fernandino, M.
2002-06-01
In the present thesis a mathematic model is presented in order to describe the dynamic behavior inside a closed enrichment loop, the latter representing a single stage of an uranium gaseous diffusion enrichment cascade.The analytical model is turned into a numerical model, and implemented through a computational code.For the verification of the model, measurements were taken in an experimental circuit using air as the process fluid.This circuit was instrumented so as to register its characteristic thermohydraulic variables.The measured transients were simulated, comparing the numerical results with the experimental measurements.A good agreement between the characteristic setting times and the thermohydraulic parameters evolution was observed.Besides, other transients of two species separation were numerically analyzed, including setting times of each magnitude, behavior of each one of them during different transients, and redistribution of concentrations
Modeling of environmentally induced transients within satellites
Stevens, N. John; Barbay, Gordon J.; Jones, Michael R.; Viswanathan, R.
1987-01-01
A technique is described that allows an estimation of possible spacecraft charging hazards. This technique, called SCREENS (spacecraft response to environments of space), utilizes the NASA charging analyzer program (NASCAP) to estimate the electrical stress locations and the charge stored in the dielectric coatings due to spacecraft encounter with a geomagnetic substorm environment. This information can then be used to determine the response of the spacecraft electrical system to a surface discharge by means of lumped element models. The coupling into the electronics is assumed to be due to magnetic linkage from the transient currents flowing as a result of the discharge transient. The behavior of a spinning spacecraft encountering a severe substorm is predicted using this technique. It is found that systems are potentially vulnerable to upset if transient signals enter through the ground lines.
Transient accelerating scalar models with exponential potentials
International Nuclear Information System (INIS)
Cui Wen-Ping; Zhang Yang; Fu Zheng-Wen
2013-01-01
We study a known class of scalar dark energy models in which the potential has an exponential term and the current accelerating era is transient. We find that, although a decelerating era will return in the future, when extrapolating the model back to earlier stages (z ≳ 4), scalar dark energy becomes dominant over matter. So these models do not have the desired tracking behavior, and the predicted transient period of acceleration cannot be adopted into the standard scenario of the Big Bang cosmology. When couplings between the scalar field and matter are introduced, the models still have the same problem; only the time when deceleration returns will be varied. To achieve re-deceleration, one has to turn to alternative models that are consistent with the standard Big Bang scenario.
A transient model to the thermal detonation
International Nuclear Information System (INIS)
Karachalios, K.
1987-04-01
The model calculates the escalation dynamics and the long time behavior of thermal detonation waves depending on the initial and boundary conditions (data of the premixture, ignition at a solid wall or at an open end, etc.). Especially, for a given mixture and a certain fragmentation behavior more than one stable steady-state cases resulted, depending on the applied ignition energy. Investigations showed a very good consistency between the transient model and a steady-state model which is based on the same physical description and includes an additional stability criterion. Also the influence of effects such as e.g. non-homogeneous coolant heating, spherical instead of plane wave propagation and inhomogeneities of the premixture on the development of the wave were investigated. Comparison calculations with large scale experiments showed that they can be well explained by means of the thermal detonation theory, especially considering the transient phase of the wave development. (orig./HP) [de
Several new thermo-hydraulic test facilities in NPIC
International Nuclear Information System (INIS)
Ye Shurong; Sun Yufa; Ji Fuyun; Zong Guifang; Guo Zhongchuan
1997-01-01
Several new thermo-hydraulic test facilities are under construction in Nuclear Power Institute of Chinese (NPIC) at Chengdu. These facilities include: 1. Nuclear Power Component Comprehensive Test Facility. 2. Reactor Hydraulic Modeling Test Facility. 3. Control Rod Drive Line Hydraulic Test Facility. 4. Large Scale Thermo-Hydraulic Test Facility. The construction of these facilities will make huge progress in the research and development capability of nuclear power technology in CHINA. The author will present a brief description of the design parameters flowchart and test program of these facilities
Experimental study and modelling of transient boiling
International Nuclear Information System (INIS)
Baudin, Nicolas
2015-01-01
A failure in the control system of the power of a nuclear reactor can lead to a Reactivity Initiated Accident in a nuclear power plant. Then, a power peak occurs in some fuel rods, high enough to lead to the coolant film boiling. It leads to an important increase of the temperature of the rod. The possible risk of the clad failure is a matter of interest for the Institut de Radioprotection et de Securite Nucleaire. The transient boiling heat transfer is not yet understood and modelled. An experimental set-up has been built at the Institut de Mecanique des Fluides de Toulouse (IMFT). Subcooled HFE-7000 flows vertically upward in a semi annulus test section. The inner half cylinder simulates the clad and is made of a stainless steel foil, heated by Joule effect. Its temperature is measured by an infrared camera, coupled with a high speed camera for the visualization of the flow topology. The whole boiling curve is studied in steady state and transient regimes: convection, onset of boiling, nucleate boiling, critical heat flux, film boiling and rewetting. The steady state heat transfers are well modelled by literature correlations. Models are suggested for the transient heat flux: the convection and nucleate boiling evolutions are self-similar during a power step. This observation allows to model more complex evolutions, as temperature ramps. The transient Hsu model well represents the onset of nucleate boiling. When the intensity of the power step increases, the film boiling begins at the same temperature but with an increasing heat flux. For power ramps, the critical heat flux decreases while the corresponding temperature increases with the heating rate. When the wall is heated, the film boiling heat transfer is higher than in steady state but it is not understood. A two-fluid model well simulates the cooling film boiling and the rewetting. (author)
Aeroelastic Modeling of a Nozzle Startup Transient
Wang, Ten-See; Zhao, Xiang; Zhang, Sijun; Chen, Yen-Sen
2014-01-01
Lateral nozzle forces are known to cause severe structural damage to any new rocket engine in development during test. While three-dimensional, transient, turbulent, chemically reacting computational fluid dynamics methodology has been demonstrated to capture major side load physics with rigid nozzles, hot-fire tests often show nozzle structure deformation during major side load events, leading to structural damages if structural strengthening measures were not taken. The modeling picture is incomplete without the capability to address the two-way responses between the structure and fluid. The objective of this study is to develop a tightly coupled aeroelastic modeling algorithm by implementing the necessary structural dynamics component into an anchored computational fluid dynamics methodology. The computational fluid dynamics component is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, while the computational structural dynamics component is developed under the framework of modal analysis. Transient aeroelastic nozzle startup analyses at sea level were performed, and the computed transient nozzle fluid-structure interaction physics presented,
The modelling of BLEVE fireball transients
Energy Technology Data Exchange (ETDEWEB)
Shield, S.R. [Shell Research Ltd., Chester (United Kingdom). Thornton Research Centre
1995-12-31
An existing physically based BLEVE fireball model has been developed to predict the shape of the transient heat pulse to a receiver, to model ``cold`` BLEVEs, and to assess the consequences to structures and people. This has been achieved by finding a correlation to predict the size of a characteristic liquid drop within the flashing cloud. These drops burn out to predict the time to fireball break-up and extinction, and their height is tracked to find the rise of the fireball, agreement with small and large scale data being satisfactory. The model is semi-empirical in that, when the level of pre-heating is low, a heat balance cannot predict the development of fireball temperature with time since the fuel is pyrolising. Also, the modelling of the transition from ``cold`` BLEVE fireball to pool fire is a cautious best estimate. Using both traditional and more recent models, the prediction of pain, burns and fatality have been incorporated into the model. Whatever criterion for a safety distance is used, the effect of modelling the transient is to reduce the calculated safety distances by up to a factor of two, which brings fatality predictions for real incidents much more in line with the historical record. (author)
Thermohydraulic analysis for power increase of IEAR-1 reactor
International Nuclear Information System (INIS)
Umbehaun, Pedro E.; Bastos, Jose L.F.
1996-01-01
In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30
ALARM, Thermohydraulics of BWR with Jet Pumps During LOCA
International Nuclear Information System (INIS)
Araya, F.; Akimoto, M.
1985-01-01
1 - Nature of physical problem solved: ALARM-B2 which is an improved version of ALARM-B1 is a computer program to analyze thermo-hydraulic phenomena of BWR during a blowdown period under a large-break loss-of-coolant accident condition with special emphasis on the heat transfer phenomena in the core region. 2 - Method of solution: A so called volume-junction method is used to present fluid conservations. The primary system is divided into a number of special elements called 'control-volumes'. The system of partial differential equations describing fluid conservations for a stream-tube are integrated over a number of control volumes. The resulting set of simultaneous differential equations that is based on the assumptions of one-dimensional, homogeneous and thermal- equilibrium flow is linearized and solved for a small time increment by a simple explicit numerical technique. The one-dimensional heat conduction equations describing temperature profiles within solid material are written in finite difference forms which are linearized and solved by the Crank-Nicholson implicit method. In order to simulate the blowdown heat transfer phenomena, the code has correlation packages for heat transfer coefficient and critical heat flux. The heat generation in the core is given by a point reactor kinetics model with six groups of delayed neutrons and decay of eleven groups of fission products and actinides. The solution technique of the reactor kinetics is based on the Runge-Kutta method. ALARM-B2 has the models to simulate various components incorporated in BWRs such as jet pumps, recirculation pumps, steam separators, valves, and so on. The discharge and injection systems are modeled by leak and fill systems, respectively. 3 - Restrictions on the complexity of the problem: As this has been developed to simulate a blowdown thermo-hydraulic transient during a large break LOCA, users must pay attention when applying the code to any medium or small break LOCAs or to later phases
Thermonuclear model for high energy transients
International Nuclear Information System (INIS)
Woosley, S.E.
1982-01-01
The thermonuclear model for x- and γ-ray bursts is discussed. Different regimes of nuclear burning are reviewed, each appropriate to a given range of (steady state) accretion rate. Accretion rates in the range 10 -14 to 10 -8 Msub solar y -1 all appear capable of producing x-ray transients of various durations and intervals. Modifications introduced by radiatively driven mass loss, the thermal inertia of the envelope, different burning mechanisms, and two-dimensional considerations are discussed as are difficulties encountered when the thermonuclear model is confronted with observations of rapidly recurrent bursts (less than or equal to 10 min), and super-Eddington luminosities and temperatures. Results from a numerical simulation of a combined hydrogen-helium runaway initiated at pycnonuclear density are presented for the first time. The thermonuclear model for γ-ray bursts is also reviewed and updated, particularly with regard to the breakdown of the steady state hypothesis employed in previous work. Solely on the basis of nuclear instability, γ-ray bursts of various types appear possible for a very broad variety of accretion rates (approx. 10 -17 to approx. 10 -11 Msub solar y -1 ) although other considerations may restrict this range. The thermonuclear model appears capable of yielding a great diversity of high energy transient phenomena for various accretion rates, magnetic field configurations, and neutron star envelope histories
Energy Technology Data Exchange (ETDEWEB)
Farahani, Aref Zarnooshe [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch; Yousefpour, Faramarz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Hoseyni, Seyed Mohsen [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Basic Sciences; Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Young Researchers and Elite Club
2017-07-15
Development of a steady-state model is the first step in nuclear safety analysis. The developed model should be qualitatively analyzed first, then a sensitivity analysis is required on the number of nodes for models of different systems to ensure the reliability of the obtained results. This contribution aims to show through sensitivity analysis, the independence of modeling results to the number of nodes in a qualified MELCOR model for a Westinghouse type pressurized power plant. For this purpose, and to minimize user error, the nuclear analysis software, SNAP, is employed. Different sensitivity cases were developed by modification of the existing model and refinement of the nodes for the simulated systems including steam generators, reactor coolant system and also reactor core and its connecting flow paths. By comparing the obtained results to those of the original model no significant difference is observed which is indicative of the model independence to the finer nodes.
Modelling structural systems for transient response analysis
International Nuclear Information System (INIS)
Melosh, R.J.
1975-01-01
This paper introduces and reports success of a direct means of determining the time periods in which a structural system behaves as a linear system. Numerical results are based on post fracture transient analyses of simplified nuclear piping systems. Knowledge of the linear response ranges will lead to improved analysis-test correlation and more efficient analyses. It permits direct use of data from physical tests in analysis and simplication of the analytical model and interpretation of its behavior. The paper presents a procedure for deducing linearity based on transient responses. Given the forcing functions and responses of discrete points of the system at various times, the process produces evidence of linearity and quantifies an adequate set of equations of motion. Results of use of the process with linear and nonlinear analyses of piping systems with damping illustrate its success. Results cover the application to data from mathematical system responses. The process is successfull with mathematical models. In loading ranges in which all modes are excited, eight digit accuracy of predictions are obtained from the equations of motion deduced. Small changes (less than 0.01%) in the norm of the transfer matrices are produced by manipulation errors for linear systems yielding evidence that nonlinearity is easily distinguished. Significant changes (greater than five %) are coincident with relatively large norms of the equilibrium correction vector in nonlinear analyses. The paper shows that deducing linearity and, when admissible, quantifying linear equations of motion from transient response data for piping systems can be achieved with accuracy comparable to that of response data
International Nuclear Information System (INIS)
Belem, J.A.T.
1993-09-01
The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs
MODELLING OF NON-ROAD TRANSIENT CYCLE
Directory of Open Access Journals (Sweden)
Martin Kotus
2013-12-01
Full Text Available The paper describes the modeling of NRTC (Non-Road Transient Cycle test procedure based on previously measured characteristics of fuel consumption, carbon monoxide (CO, carbon dioxide (CO2, hydrocarbons (HC, nitrogen oxides (NOx and particulates (PM production. It makes possible to compare the current technical condition of an internal combustion engine of an agricultural tractor with its previous state or other tractor’s engine. Based on measured characteristics, it is also possible to model any other cycle without further measurements (NRSC test procedure, cycle for specific conditions – mountain tractor, etc.. The result may thus contribute to improving the environment by reducing the production of harmful substances emitted into the air and save money due to reduced fuel consumption.
International Nuclear Information System (INIS)
Bobkov, V.P.; Vinogradov, V.N.; Efanov, A.D.; Sergeev, V.V.; Smogalev, I.P.
2003-01-01
The results of verifying the model for calculating the heat exchange crisis in the uniformly heated rod bundles, realized in the calculation code of the improved evaluation KORSAR, are presented. The model for calculating the critical heat fluxes in this code is based on the tabular method. The experimental data bank of the Branch base center of the thermophysical data GNTs RF - FEhI for the rod bundles, structurally similar to the WWER fuel assemblies, was used by the verification within the wide range of parameters: pressure from 0.11 up to 20 MPa and mass velocity from 5- up to 5000 kg/(m 2 s) [ru
Wertelaers, P
2010-01-01
The current design foresees a central heat exchanger followed by a controlled post heater, for all ECAL. We discuss the scheme and try to assess its performance, from a Barrel viewpoint. This is based on computational work. The coolant transfer pipes play an essential role in building a dynamical model. After some studies on the behaviour of the cooling circuit itself, a strong yet simple controller is proposed. Then, the system with feedback control is scrutinized, with emphasis on disturbance rejection. The most relevant disturbances are cooling ripple, pipe heat attack, and electronics’ switching.
Modeling photocurrent transients in organic solar cells
International Nuclear Information System (INIS)
Hwang, I; Greenham, N C
2008-01-01
We investigate the transient photocurrents of organic photovoltaic devices in response to a sharp turn-on of illumination, by numerical modeling of the drift-diffusion equations. We show that the photocurrent turn-on dynamics are determined not only by the transport dynamics of free charges, but also by the time required for the population of geminate charge pairs to reach its steady-state value. The dissociation probability of a geminate charge pair is found to be a key parameter in determining the device performance, not only by controlling the efficiency at low intensities, but also in determining the fate of charge pairs formed by bimolecular recombination at high intensities. Bimolecular recombination is shown to reduce the turn-on time at high intensities, since the typical distance traveled by a charge pair is reduced.
International Nuclear Information System (INIS)
Lee, Yong Bum; Jeong, Hae Yong; Cho, Chung Ho; Kwon, Young Min; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong; Hahn, Do Hee
2009-01-01
The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect
International Nuclear Information System (INIS)
Serbanel, M.; Catana, A.
2001-01-01
This report presents a comparative analysis of the behaviour of primary circuit during a LOCA 20% RIH accident for two types of reactor core, namely, normally charged, i.e., with clusters of 37 rods and charged with clusters of 43 rods, respectively. This type of accident was chosen since Canadian analyses showed that the associated transient regime stress the fuel elements. The void reactivity as a function of coolant average density was calibrated for a reference regime (LOCA 20% RIH) so that the results of the model be able to reproduce the average distribution in the reference transient regime. The computation makes use of CERBERUS and FIREBIRD codes externally coupled by files. The void reactivity of the hot pencil was obtained this way. An extremely conservative hypothesis was used, namely that the momentary power of the cluster hosting the pencil is the maximal power over the cluster for the corresponding half reactor core. To carry out this work the following steps were covered: 1. The scenario for the LOCA 20% RIH accident was worked out and the input data corresponding to the thermohydraulic and neutronic modules, for the complex model and the 37 rod clusters, were checked; 2. The input data corresponding to the thermohydraulic module for the complex model and the 43 rod cluster were checked; 3. The kinetic parameters corresponding to the 37 rod cluster were computed; 4. The kinetic parameters corresponding to the 43 rod cluster were computed and the file for the input data in the neutronic module was built; 5. A sub-routine for writing files with the thermohydraulic and neutronic quantities, in a format adequate to the other programs, was implemented; 6. The two transient regimes considered were implemented and the archives containing the quantities were built ;7. The results obtained were analyzed. The conclusion of this work is that in case of LOCA 20% RIH accident the 43 bar clusters have a better behaviour than the 37 bar clusters
Thermonuclear model for x-ray transients
International Nuclear Information System (INIS)
Wallace, R.K.; Woosley, S.E.; Weaver, T.A.
1982-01-01
The thermonuclear evolution of a 1.41 M sub solar neutron star accreting both solar and metal-deficient mixtures of hydrogen, helium, and heavy elements at rates ranging from about 10 -11 to 10 -10 M sub solar per year is examined using a one-dimensional numerical model. The metal deficient compositions may result either from placement of the neutron star in a binary system with a Population II red giant or from gravitational settling of heavy ions in the accreted material. For such accretion rates and metallicities, hydrogen burning, mediated by the β-limited CNO cycle, is stable and leads to the accumulation of a thick helium layer with mass 10 23 to 10 25 g and temperature 0.7 less than or equal to T 8 less than or equal to 1.2. Helium ignition occurs under extremely degenerate circumstances and is catastrophically violent. In the lower t helium shells this runaway is propagated as a convective deflagration, for the thicker layers a detonation front is set up which steepens into a strong relativistic shock wave in the neutron star envelope. In all models greatly super-Eddington luminosities in the outer layers of the neutron star lead to a sustained epoch of radiatively driven mass loss. Observationally, such models may correspond to rapid x-ray transients. The hopeless prospect for constructing a one-dimensional model for γ-ray bursts without magnetic field confinement is discussed and uncertainties pointed out in the strong screening correction for helium burning reaction
Cable system transients theory, modeling and simulation
Ametani, Akihiro; Nagaoka, Naoto
2015-01-01
A systematic and comprehensive introduction to electromagnetic transient in cable systems, written by the internationally renowned pioneer in this field Presents a systematic and comprehensive introduction to electromagnetic transient in cable systems Written by the internationally renowned pioneer in the field Thorough coverage of the state of the art on the topic, presented in a well-organized, logical style, from fundamentals and practical applications A companion website is available
Point kinetics improvements to evaluate three-dimensional effects in transients calculation
International Nuclear Information System (INIS)
Castellotti, U.
1987-01-01
A calculation method, which considers the flux axial perturbations in the parameters related to the reactivity within a point kinetics model, is described. The method considered uses axial factors of consideration which act on the thermohydraulic variables included in the reactivity calculation. The PUMA three-dimensional code as reference model for the comparisons, is used. The limitations inherent to the reactivity balance of the point models used in the transients calculation, are given. (Author)
Modelling and transient simulation of water flow in pipelines using WANDA Transient software
Directory of Open Access Journals (Sweden)
P.U. Akpan
2017-09-01
Full Text Available Pressure transients in conduits such as pipelines are unsteady flow conditions caused by a sudden change in the flow velocity. These conditions might cause damage to the pipelines and its fittings if the extreme pressure (high or low is experienced within the pipeline. In order to avoid this occurrence, engineers usually carry out pressure transient analysis in the hydraulic design phase of pipeline network systems. Modelling and simulation of transients in pipelines is an acceptable and cost effective method of assessing this problem and finding technical solutions. This research predicts the pressure surge for different flow conditions in two different pipeline systems using WANDA Transient simulation software. Computer models were set-up in WANDA Transient for two different systems namely; the Graze experiment (miniature system and a simple main water riser system based on some initial laboratory data and system parameters. The initial laboratory data and system parameters were used for all the simulations. Results obtained from the computer model simulations compared favourably with the experimental results at Polytropic index of 1.2.
Energy Technology Data Exchange (ETDEWEB)
Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.
1998-03-01
The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum
Confirmatory simulation of safety and operational transients in LMFBR systems
International Nuclear Information System (INIS)
Guppy, J.G.; Agrawal, A.K.
1978-01-01
Operational and safety transients that may originate anywhere in an LMFBR system must be adequately simulated to assist in safety evaluation and plant design efforts. This paper describes an advanced thermohydraulic transient code, the Super System Code (SSC), that may be used for confirmatory safety evaluations of plant wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions, and presents results obtained with SSC illustrating the degree of modelling detail present in the code as well as the computing efficiency. (author)
International Nuclear Information System (INIS)
Grange, J.L.; Caremoli, C.; Eddi, M.
1988-01-01
This paper presents improvements performed on SICLE numerical model in order to analyse the condensation front that occurs in the moisture separator reheaters (MSR) of nuclear power plants. Modifications of SICLE numerical model architecture and a fine modelling of reheater have allowed to correctly simulate the MSR thermohydraulic behaviour during a severe transient (plant islanding) [fr
Basic researches on thermo-hydraulic non-equilibrium phenomena related to nuclear reactor safety
International Nuclear Information System (INIS)
Sakurai, Akira; Kataoka, Isao; Aritomi, Masanori.
1989-01-01
A review was made of recent developments of fundamental researches on thermo-hydraulic non-equilibrium phenomena related to light water reactor safety, in relation to problems to be solved for the improvement of safety analysis codes. As for the problems related to flow con ditions, fundamental researches on basic conservation equations and constitutive equations for transient two-phase flow were reviewed. Regarding to the problems related to thermal non-equilibrium phenomena, fundamental researches on film boiling in pool and forced convection, transient boiling heat transfer and flow behavior caused by pressure transients were reviewed. (author)
Analytic models for fuel pin transient performance
International Nuclear Information System (INIS)
Bard, F.E.; Fox, G.L.; Washburn, D.F.; Hanson, J.E.
1976-09-01
HEDL's ability to analyze various mechanisms that operate within a fuel pin has progressed substantially through development of codes such as PECTCLAD, which solves cladding response, and DSTRESS, which solves fuel response. The PECTCLAD results show good correlation with a variety of mechanical tests on cladding material and also demonstrate the significance of cladding strength when applying the life fraction rule. The DSTRESS results have shown that fuel deforms sufficiently during overpower transient tests that available volumes are filled, whether in the form of a central cavity or start-up cracks
Development of a computer code for transients simulation in PWR type reactors
International Nuclear Information System (INIS)
Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de
1981-01-01
A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt
Thermohydraulic relationships for advanced water cooled reactors
International Nuclear Information System (INIS)
2001-04-01
This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes
Thermohydraulic relationships for advanced water cooled reactors
Energy Technology Data Exchange (ETDEWEB)
NONE
2001-04-01
This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes
Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3
Energy Technology Data Exchange (ETDEWEB)
Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)
1985-07-01
This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)
Complete wind farm electromagnetic transient modelling for grid integration studies
International Nuclear Information System (INIS)
Zubia, I.; Ostolaza, X.; Susperregui, A.; Tapia, G.
2009-01-01
This paper presents a modelling methodology to analyse the impact of wind farms in surrounding networks. Based on the transient modelling of the asynchronous generator, the multi-machine model of a wind farm composed of N generators is developed. The model incorporates step-up power transformers, distribution lines and surrounding loads up to their connection to the power network. This model allows the simulation of symmetric and asymmetric short-circuits located in the distribution network and the analysis of transient stability of wind farms. It can be also used to study the islanding operation of wind farms
Computation of 3D thermohydraulics in partially blocked bundles during the reflood phase of a LOCA
International Nuclear Information System (INIS)
Cicero, G.M.; Briere, E.; Fornaciari, G.
1994-06-01
In Pressurized Water Reactors (PWR), ballooning of the fuel rod claddings may occur during a LOCA, since the fuel rod claddings are heated up, and the system pressure is low. The severe blockages that may result induce cross-flow diversion and three-dimensional effects on thermohydraulics in the core bundle, during the reflood phase. To improve the knowledge of these phenomena and their physical modelling in the code CATHARE, 3D computer codes are needed. In 1990, EDF has started up a development and validation program of the 3D THYC computer code to analyze the thermohydraulics of the flow during the reflood phase, in partially blocked bundles. The main objective is to calculate the temperatures of the rods above the quench front, when they are cooled by superheated steam with saturated droplets. First, this paper introduces the THYC model developed for reflood studies. Secondly, we report the first qualification results on a Flooding Experiments with Blocked Array (FEBA) test. Thirdly, we analyze the model predictions on a large break LOCA transient, in a 900 MW PWR 11x11 core area with a 3x3 central blockage. THYC simulates the transient in the bundle around and above the blockage, until the quench front enters the computational domain. Previously, a 1D CATHARE simulation gives the boundary conditions and, in the reactor core case, the deformation of the blocked fuel rods. The results analysis focused on the time evolution of the clad temperatures in the blocked and in the bypass region. In the FEBA test simulation, the main observations are properly predicted within the blockage. Temperatures are lower in blocked rod sleeves than in unblocked rod claddings since the steam gap reduces the power transmitted by the heater rod to the sleeve. In the core case, the model predicts the opposite result. Within the blockage, ballooned rod temperatures are higher than non-ballooned rod ones. We show by sensitivity studies that these behaviour difference between FEBA rods
Transient Heat Transfer Model for Car Body Primer Curing
D. Zabala; N. Sánchez; J. Pinto
2010-01-01
A transient heat transfer mathematical model for the prediction of temperature distribution in the car body during primer baking has been developed by considering the thermal radiation and convection in the furnace chamber and transient heat conduction governing equations in the car framework. The car cockpit is considered like a structure with six flat plates, four vertical plates representing the car doors and the rear and front panels. The other two flat plates are the...
Performance of neutron kinetics models for ADS transient analyses
International Nuclear Information System (INIS)
Rineiski, A.; Maschek, W.; Rimpault, G.
2002-01-01
Within the framework of the SIMMER code development, neutron kinetics models for simulating transients and hypothetical accidents in advanced reactor systems, in particular in Accelerator Driven Systems (ADSs), have been developed at FZK/IKET in cooperation with CE Cadarache. SIMMER is a fluid-dynamics/thermal-hydraulics code, coupled with a structure model and a space-, time- and energy-dependent neutronics module for analyzing transients and accidents. The advanced kinetics models have also been implemented into KIN3D, a module of the VARIANT/TGV code (stand-alone neutron kinetics) for broadening application and for testing and benchmarking. In the paper, a short review of the SIMMER and KIN3D neutron kinetics models is given. Some typical transients related to ADS perturbations are analyzed. The general models of SIMMER and KIN3D are compared with more simple techniques developed in the context of this work to get a better understanding of the specifics of transients in subcritical systems and to estimate the performance of different kinetics options. These comparisons may also help in elaborating new kinetics models and extending existing computation tools for ADS transient analyses. The traditional point-kinetics model may give rather inaccurate transient reaction rate distributions in an ADS even if the material configuration does not change significantly. This inaccuracy is not related to the problem of choosing a 'right' weighting function: the point-kinetics model with any weighting function cannot take into account pronounced flux shape variations related to possible significant changes in the criticality level or to fast beam trips. To improve the accuracy of the point-kinetics option for slow transients, we have introduced a correction factor technique. The related analyses give a better understanding of 'long-timescale' kinetics phenomena in the subcritical domain and help to evaluate the performance of the quasi-static scheme in a particular case. One
On the empirical relevance of the transient in opinion models
Energy Technology Data Exchange (ETDEWEB)
Banisch, Sven, E-mail: sven.banisch@universecity.d [Mathematical Physics, Physics Department, Bielefeld University, 33501 Bielefeld (Germany); Institute for Complexity Science (ICC), 1249-078 Lisbon (Portugal); Araujo, Tanya, E-mail: tanya@iseg.utl.p [Research Unit on Complexity in Economics (UECE), ISEG, TULisbon, 1249-078 Lisbon (Portugal); Institute for Complexity Science (ICC), 1249-078 Lisbon (Portugal)
2010-07-12
While the number and variety of models to explain opinion exchange dynamics is huge, attempts to justify the model results using empirical data are relatively rare. As linking to real data is essential for establishing model credibility, this Letter develops an empirical confirmation experiment by which an opinion model is related to real election data. The model is based on a representation of opinions as a vector of k bits. Individuals interact according to the principle that similarity leads to interaction and interaction leads to still more similarity. In the comparison to real data we concentrate on the transient opinion profiles that form during the dynamic process. An artificial election procedure is introduced which allows to relate transient opinion configurations to the electoral performance of candidates for which data are available. The election procedure based on the well-established principle of proximity voting is repeatedly performed during the transient period and remarkable statistical agreement with the empirical data is observed.
On the empirical relevance of the transient in opinion models
International Nuclear Information System (INIS)
Banisch, Sven; Araujo, Tanya
2010-01-01
While the number and variety of models to explain opinion exchange dynamics is huge, attempts to justify the model results using empirical data are relatively rare. As linking to real data is essential for establishing model credibility, this Letter develops an empirical confirmation experiment by which an opinion model is related to real election data. The model is based on a representation of opinions as a vector of k bits. Individuals interact according to the principle that similarity leads to interaction and interaction leads to still more similarity. In the comparison to real data we concentrate on the transient opinion profiles that form during the dynamic process. An artificial election procedure is introduced which allows to relate transient opinion configurations to the electoral performance of candidates for which data are available. The election procedure based on the well-established principle of proximity voting is repeatedly performed during the transient period and remarkable statistical agreement with the empirical data is observed.
Unification of three linear models for the transient visual system
Brinker, den A.C.
1989-01-01
Three different linear filters are considered as a model describing the experimentally determined triphasic impulse responses of discs. These impulse responses arc associated with the transient visual system. Each model reveals a different feature of the system. Unification of the models is
Thermo-Hydraulic behaviour of dual-channel superconducting Cable-In-Conduit Conductors for ITER
International Nuclear Information System (INIS)
Renard, B.
2006-09-01
In an effort to optimise the cryogenics of large superconducting coils for fusion applications (ITER), dual channel Cable-In-Conduit Conductors (CICC) are designed with a central channel spiral to provide low hydraulic resistance and faster helium circulation. The qualitative and economic rationale of the conductor central channel is here justified to limit the superconductor temperature increase, but brings more complexity to the conductor cooling characteristics. The pressure drop of spirals is experimentally evaluated in nitrogen and water and an explicit hydraulic friction model is proposed. Temperatures in the cable must be quantified to guarantee superconductor margin during coil operation under heat disturbance and set adequate inlet temperature. Analytical one-dimensional thermal models, in steady state and in transient, allow to better understand the thermal coupling of CICC central and annular channels. The measurement of a heat transfer characteristic space and time constants provides cross-checking experimental estimations of the internal thermal homogenization. A simple explicit model of global inter-channel heat exchange coefficient is proposed. The risk of thermosyphon between the two channels is considered since vertical portions of fusion coils are subject to gravity. The new hydraulic model, heat exchange model and gravitational risk ratio allow the thermohydraulic improvement of CICC central spirals. (author)
Model for transient creep of southeastern New Mexico rock salt
International Nuclear Information System (INIS)
Herrmann, W.; Wawersik, W.R.; Lauson, H.S.
1980-11-01
In a previous analysis, existing experimental data pertaining to creep tests on rock salt from the Salado formation of S.E. New Mexico were fitted to an exponential transient creep law. While very early time portions of creep strain histories were not fitted very well for tests at low temperatures and stresses, initial creep rates in particular generally being underestimated, the exponential creep law has the property that the transient creep strain approaches a finite limit with time, and is therefore desirable from a creep modelling point of view. In this report, an analysis of transient creep is made. It is found that exponential transient creep can be related to steady-state creep through a universal creep curve. The resultant description is convenient for creep analyses where very early time behavior is not important
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.
Modelling transient energy release from molten fuel coolant interaction debris
International Nuclear Information System (INIS)
Fletcher, D.F.
1984-05-01
A simple model of transient energy release in a Molten Fuel Coolant Interaction is presented. A distributed heat transfer model is used to examine the effect of heat transfer coefficient, time available for rapid energy heat transfer and particle size on transient energy release. The debris is assumed to have an Upper Limit Lognormal distribution. Model predictions are compared with results from the SUW series of experiments which used thermite-generated uranium dioxide molybdenum melts released below the surface of a pool of water. Uncertainties in the physical principles involved in the calculation of energy transfer rates are discussed. (author)
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)
Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility
International Nuclear Information System (INIS)
Coragem, Helio Boemer de Oliveira
1980-01-01
A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)
Deformation modeling and the strain transient dip test
International Nuclear Information System (INIS)
Jones, W.B.; Rohde, R.W.; Swearengen, J.C.
1980-01-01
Recent efforts in material deformation modeling reveal a trend toward unifying creep and plasticity with a single rate-dependent formulation. While such models can describe actual material deformation, most require a number of different experiments to generate model parameter information. Recently, however, a new model has been proposed in which most of the requisite constants may be found by examining creep transients brought about through abrupt changes in creep stress (strain transient dip test). The critical measurement in this test is the absence of a resolvable creep rate after a stress drop. As a consequence, the result is extraordinarily sensitive to strain resolution as well as machine mechanical response. This paper presents the design of a machine in which these spurious effects have been minimized and discusses the nature of the strain transient dip test using the example of aluminum. It is concluded that the strain transient dip test is not useful as the primary test for verifying any micromechanical model of deformation. Nevertheless, if a model can be developed which is verifiable by other experimentts, data from a dip test machine may be used to generate model parameters
Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)
Energy Technology Data Exchange (ETDEWEB)
Bradley K. Heath
2014-03-01
This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.
Transient analysis models for nuclear power plants
International Nuclear Information System (INIS)
Agapito, J.R.
1981-01-01
The modelling used for the simulation of the Angra-1 start-up reactor tests, using the RETRAN computer code is presented. Three tests are simulated: a)nuclear power plant trip from 100% of power; b)great power excursions tests and c)'load swing' tests.(E.G.) [pt
Magic Formula Tyre Model with Transient Properties
Pacejka, H.B.; Besselink, I.J.M.
1997-01-01
The tyre force and moment generating properties connected with the vehicle's horizontal motions are considered. Knowledge of tyre properties is necessary to properly design vehicle components and advanced control systems. For this purpose, mathematical models of the tyre are being used in vehicle
Directory of Open Access Journals (Sweden)
Li Zhang
2017-12-01
Full Text Available In recent years, with the rapid development of offshore wind turbines (WTs, the problem of lightning strikes has become more and more prominent. In order to reduce the failure rate caused by the transient overvoltage of lightning struck offshore WTs, the influencing factors and the response rules of transient overvoltage are analyzed. In this paper, a new integrated electromagnetic transient model of offshore WTs is established by using the numerical calculation method of the electromagnetic field first. Then, based on the lightning model and considering the impedance of the lightning channel, the transient overvoltage of lightning is analyzed. Last, the electromagnetic transient model of offshore WTs is simulated and analyzed by using the alternative transients program electro-magnetic transient program (ATP-EMTP software. The influence factors of lightning transient overvoltage are studied. The main influencing factors include the sea depth, the blade length, the tower height, the lightning flow parameters, the lightning strike point, and the blade rotation position. The simulation results show that the influencing factors mentioned above have different effects on the lightning transient overvoltage. The results of the study have some guiding significance for the design of the lightning protection of the engine room.
Modeling of the transient mobility in disordered organic semiconductors
Germs, W.C.; Van der Holst, J.M.M.; Van Mensfoort, S.L.M.; Bobbert, P.A.; Coehoorn, R.
2011-01-01
In non-steady-state experiments, the electrical response of devicesbased on disordered organic semiconductors often shows a large transient contribution due to relaxation of the out-of-equilibrium charge-carrier distribution. We have developed a model describing this process, based only on the
Transient Model of Hybrid Concentrated Photovoltaic with Thermoelectric Generator
DEFF Research Database (Denmark)
Mahmoudi Nezhad, Sajjad; Qing, Shaowei; Rezaniakolaei, Alireza
2017-01-01
Transient performance of a concentrated photovoltaic thermoelectric (CPV-TEG) hybrid system is modeled and investigated. A heat sink with water, as the working fluid has been implemented as the cold reservoir of the hybrid system to harvest the heat loss from CPV cell and to increase the efficiency...
The state of art of the methods for thermohydraulics design of LMFBR fuel elements
International Nuclear Information System (INIS)
Fernandez y Fernandez, E.; Carajilescov, P.
1981-09-01
The present (experimental and analytical) state of art of the methods for thermohydraulics design of LMFBR fuel elements is analyzed. A development program is suggested, in order to obtain a computer code for modelling the distribution of coolant enthalpy in reactor core. This computer code is in development. (Author) [pt
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.
Research activity on thermohydraulic problems of PWR reactors
International Nuclear Information System (INIS)
Szabados, L.
1976-06-01
The general review of the experimental and theoretical research works on thermohydraulic investigation of pressurized water type reactors being done in the Central Research Institute for Physics is given. The main results of the theoretical and theoretical-numerical research are summarized. The most important result of the past years is the construction of the High Pressure Water Cooled Loop (NVH) thermohydraulic loop. Another significant achievement was the development of the reactor thermohydraulic program system. (Sz.N.Z.)
A transient single particle model under FCI conditions
Institute of Scientific and Technical Information of China (English)
LI Xiao-Yan; SHANG Zhi; XU Ji-Jun
2005-01-01
The paper is focused on the coupling effect between film boiling heat transfer and evaporation drag around a hot-particle in cold liquid. Based on the continuity, momentum and energy equations of the vapor film, a transient two-dimensional single particle model has been established. This paper contains a detailed description of HPMC (High-temperature Particle Moving in Coolant) model for studying some aspects of the premixing stage of fuel-coolant interactions (FCIs). The transient process of high-temperature particles moving in coolant can be simulated. Comparisons between the experiment results and the calculations using HPMC model demonstrate that HPMC model achieves a good agreement in predicting the time-varying characteristic of high-temperature spheres moving in coolant.
Thermohydraulic calculations of PWR primary circuits
International Nuclear Information System (INIS)
Botelho, D.A.
1984-01-01
Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt
IBIS, FBR 3-D Steady-State and Kinetics with Thermohydraulic Feedback
International Nuclear Information System (INIS)
Konomura, Mamoru; Tada, Nobuo; Oka, Yoshiaki; An, Shigehiro
1987-01-01
1 - Description of program or function: The IBIS code performs steady state and kinetics calculations based on a three-dimensional nuclear diffusion kinetics with thermal hydraulic feedback. It can calculate the following values in hexagonal-Z geometry of a fast breeder reactor core through the progress of transient: (1) Net reactivity; (2) Total and group-wise delayed neutron fraction; (3) Group-wise delayed neutron precursor concentration; (4) Total power and energy; (5) Space dependent neutron flux in each energy group; (6) Space dependent temperature of each material; (7) Maximum temperature of each material and its location. 2 - Method of solution: The quasi-static method is adopted to solve the three-dimensional nuclear diffusion kinetics problem. The method is the same as employed in the code QX1. The shape function equation is solved with the finite difference treatment as used in the codes CITATION and HONEYCOMB. One-dimensional thermo-hydraulics is solved with a model similar to that given in the code SASLA. Sodium boiling can be taken into account. 3 - Restrictions on the complexity of the problem: The number of neutron energy groups is fixed to 3 groups in the present version of the code
International Nuclear Information System (INIS)
Nikonov, S.; Velkov, K.
2008-01-01
The ATHLET-BIPR-VVER coupled system code is applied for performing of safety analysis for different WWER reactors. During the last years its validation matrix is continuously being enlarged. The measurements performed during the commissioning phase of NPP Kalinin Unit 3 for the transient 'Switching-off of one Main Circulation Pump at nominal power' are very well documented and have a variety of recorded integral and local thermo-hydraulic and neutron-physic parameters including the measurements' errors. This data is being used for further validation of the coupled code system ATHLET-BIPR-VVER. In the paper are discussed the problems and our solutions by the correct interpretation of the measured thermocouples' records at NPP Kalinin-3 and the comparison with the predicted results by the coupled thermal-hydraulic/neutron-kinetic code ATHLET-BIPR-VVER. Of primary importance by such comparisons is the correct accounting of the fluid mixing process that take place in the surrounding of the measuring sensors and also the consideration of the time delay (inertia term) of the measuring devices. On the bases of previous experience and many simulations of the defined transient a method is discussed and proposed to consider correctly the inertia term of the thermocouples' measurements. The new modelling is implemented in the coupled system code ATHLET-BIPR-VVER for further validation. (Author)
Accelerating transient simulation of linear reduced order models.
Energy Technology Data Exchange (ETDEWEB)
Thornquist, Heidi K.; Mei, Ting; Keiter, Eric Richard; Bond, Brad
2011-10-01
Model order reduction (MOR) techniques have been used to facilitate the analysis of dynamical systems for many years. Although existing model reduction techniques are capable of providing huge speedups in the frequency domain analysis (i.e. AC response) of linear systems, such speedups are often not obtained when performing transient analysis on the systems, particularly when coupled with other circuit components. Reduced system size, which is the ostensible goal of MOR methods, is often insufficient to improve transient simulation speed on realistic circuit problems. It can be shown that making the correct reduced order model (ROM) implementation choices is crucial to the practical application of MOR methods. In this report we investigate methods for accelerating the simulation of circuits containing ROM blocks using the circuit simulator Xyce.
Runaway transient simulation of a model Kaplan turbine
Energy Technology Data Exchange (ETDEWEB)
Liu, S; Liu, D; Wu, Y [State Key Laboratory of Hydroscience and Engineering, Department of Thermal Eng., Tsinghua University, Beijing, 100084 (China); Zhou, D [Water Conservancy and Hydropower Eng., Hohai University, Nanjing. 210098 (China); Nishi, M, E-mail: liushuhong@tsinghua.edu.c [Kyushu Inst. Tech. Senior Academy, Kitakyushu, 804-8550 (Japan)
2010-08-15
The runaway transient is a typical transient process of a hydro power unit, where the rotational speed of a turbine runner rapidly increases up to the runaway speed under a working head as the guide vanes cannot be closed due to some reason at the load rejection. In the present paper, the characteristics of the runaway transient of a model Kaplan turbine having ns = 479(m-kW) is simulated by using a time-dependent CFD technique where equation of rotational motion of runner, continuity equation and unsteady RANS equations with RNG k-{epsilon} turbulence model are solved iteratively. In the calculation, unstructured mesh is used to the whole flow passage, which consists of several sub-domains: entrance, casing, stay vanes + guide vanes, guide section, runner and draft tube. And variable speed sliding mesh technique is used to exchange interface flow information between moving part and stationary part, and three-dimensional unstructured dynamic mesh technique is also adopted to ensure mesh quality. Two cases were treated in the simulation of runaway transient characteristics after load rejection: one is the rated operating condition as the initial condition, and the other is the condition at the maximum head. Regarding the runaway speed, the experimental speed is 1.45 times the initial speed and the calculation is 1.47 times the initial for the former case. In the latter case, the experiment and the calculation are 1.67 times and 1.69 times respectively. From these results, it is recognized that satisfactorily prediction will be possible by using the present numerical method. Further, numerical results show that the swirl in the draft-tube flow becomes stronger in the latter part of the transient process so that a vortex rope will occur in the draft tube and its precession will cause the pressure fluctuations which sometimes affect the stability of hydro power system considerably.
Runaway transient simulation of a model Kaplan turbine
Liu, S.; Zhou, D.; Liu, D.; Wu, Y.; Nishi, M.
2010-08-01
The runaway transient is a typical transient process of a hydro power unit, where the rotational speed of a turbine runner rapidly increases up to the runaway speed under a working head as the guide vanes cannot be closed due to some reason at the load rejection. In the present paper, the characteristics of the runaway transient of a model Kaplan turbine having ns = 479(m-kW) is simulated by using a time-dependent CFD technique where equation of rotational motion of runner, continuity equation and unsteady RANS equations with RNG k-epsilon turbulence model are solved iteratively. In the calculation, unstructured mesh is used to the whole flow passage, which consists of several sub-domains: entrance, casing, stay vanes + guide vanes, guide section, runner and draft tube. And variable speed sliding mesh technique is used to exchange interface flow information between moving part and stationary part, and three-dimensional unstructured dynamic mesh technique is also adopted to ensure mesh quality. Two cases were treated in the simulation of runaway transient characteristics after load rejection: one is the rated operating condition as the initial condition, and the other is the condition at the maximum head. Regarding the runaway speed, the experimental speed is 1.45 times the initial speed and the calculation is 1.47 times the initial for the former case. In the latter case, the experiment and the calculation are 1.67 times and 1.69 times respectively. From these results, it is recognized that satisfactorily prediction will be possible by using the present numerical method. Further, numerical results show that the swirl in the draft-tube flow becomes stronger in the latter part of the transient process so that a vortex rope will occur in the draft tube and its precession will cause the pressure fluctuations which sometimes affect the stability of hydro power system considerably.
Runaway transient simulation of a model Kaplan turbine
International Nuclear Information System (INIS)
Liu, S; Liu, D; Wu, Y; Zhou, D; Nishi, M
2010-01-01
The runaway transient is a typical transient process of a hydro power unit, where the rotational speed of a turbine runner rapidly increases up to the runaway speed under a working head as the guide vanes cannot be closed due to some reason at the load rejection. In the present paper, the characteristics of the runaway transient of a model Kaplan turbine having ns = 479(m-kW) is simulated by using a time-dependent CFD technique where equation of rotational motion of runner, continuity equation and unsteady RANS equations with RNG k-ε turbulence model are solved iteratively. In the calculation, unstructured mesh is used to the whole flow passage, which consists of several sub-domains: entrance, casing, stay vanes + guide vanes, guide section, runner and draft tube. And variable speed sliding mesh technique is used to exchange interface flow information between moving part and stationary part, and three-dimensional unstructured dynamic mesh technique is also adopted to ensure mesh quality. Two cases were treated in the simulation of runaway transient characteristics after load rejection: one is the rated operating condition as the initial condition, and the other is the condition at the maximum head. Regarding the runaway speed, the experimental speed is 1.45 times the initial speed and the calculation is 1.47 times the initial for the former case. In the latter case, the experiment and the calculation are 1.67 times and 1.69 times respectively. From these results, it is recognized that satisfactorily prediction will be possible by using the present numerical method. Further, numerical results show that the swirl in the draft-tube flow becomes stronger in the latter part of the transient process so that a vortex rope will occur in the draft tube and its precession will cause the pressure fluctuations which sometimes affect the stability of hydro power system considerably.
An overview of IPPE research on liquid metal fast reactor thermohydraulics
International Nuclear Information System (INIS)
Sorokin, A. P.; Efanov, A. D.; Zhukov, A. V.; Bogoslovskaia, G. P.
2003-01-01
The paper presents brief information on the most significant researches in the fields of liquid metal hydrodynamics and heat transfer performed in the State Scientific Center of Russian Federation 'Institute for Physics and Power Engineering' named after A.I.Leypunski applied to sodium-cooled fast reactors. Experimental methods for studying liquid metal thermohydraulics and applied measurement techniques are overviewed briefly in the paper. Some results of fundamental thermohydraulic investigations, such as quasi-universal character of velocity and temperature profile in liquid metals, if considered normally to the channel wall etc. are presented. Specific features of heat transfer in liquid metal cooled fuel subassembly are mentioned, among them there are: high level of coolant temperature; significant influence of an interchannel exchange on velocity and temperature distribution; an availability of contact thermal resistance; large azimuthal non-uniformity of velocity and temperature; 'conjugate' problem of heat transfer in combined geometry of fuel pin; an absence of stabilization of heat transfer in non-standard channels; an influence of non-uniform heat generation. Special attention is given to the temperature fields in fuel subassembly subjected to deformation because of radioactive swelling and creeping, as well as in case of blockage of a part of subassembly cross section. Some results of thermohydraulic investigation are demonstrated for intermediate heat exchangers, pressurized head collectors. Also the developed methods and codes of thermohydraulic calculations applied to fast reactor core are considered: subchannel approach, porous body model
Compositional Abstraction of PEPA Models for Transient Analysis
DEFF Research Database (Denmark)
Smith, Michael James Andrew
2010-01-01
- or interval - Markov chains allow us to aggregate states in such a way as to safely bound transient probabilities of the original Markov chain. Whilst we can apply this technique directly to a PEPA model, it requires us to obtain the CTMC of the model, whose state space may be too large to construct......Stochastic process algebras such as PEPA allow complex stochastic models to be described in a compositional way, but this leads to state space explosion problems. To combat this, there has been a great deal of work in developing techniques for abstracting Markov chains. In particular, abstract...
URGAP: A gap conductance model for transient conditions
International Nuclear Information System (INIS)
Lassmann, K.; Pazdera, F.
1983-01-01
A gap conductance model, URGAP, has been developed with contributions from solid, fluid and radiation heat transfer components. Model parameters are easily available, independent of different combinations of material surfaces. The model parameters were fitted to 388 data points under reactor conditions. For model verification, another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Materials data were carefully checked over a wide range of temperatures. Special attention was paid to the contact term for high temperatures. Thus, the model can be applied to transients. The URGAP model is being used successfully in several codes (e.g. URANUS, SSYST). (author)
Computer Models for IRIS Control System Transient Analysis
International Nuclear Information System (INIS)
Gary D Storrick; Bojan Petrovic; Luca Oriani
2007-01-01
This report presents results of the Westinghouse work performed under Task 3 of this Financial Assistance Award and it satisfies a Level 2 Milestone for the project. Task 3 of the collaborative effort between ORNL, Brazil and Westinghouse for the International Nuclear Energy Research Initiative entitled 'Development of Advanced Instrumentation and Control for an Integrated Primary System Reactor' focuses on developing computer models for transient analysis. This report summarizes the work performed under Task 3 on developing control system models. The present state of the IRIS plant design--such as the lack of a detailed secondary system or I and C system designs--makes finalizing models impossible at this time. However, this did not prevent making considerable progress. Westinghouse has several working models in use to further the IRIS design. We expect to continue modifying the models to incorporate the latest design information until the final IRIS unit becomes operational. Section 1.2 outlines the scope of this report. Section 2 describes the approaches we are using for non-safety transient models. It describes the need for non-safety transient analysis and the model characteristics needed to support those analyses. Section 3 presents the RELAP5 model. This is the highest-fidelity model used for benchmark evaluations. However, it is prohibitively slow for routine evaluations and additional lower-fidelity models have been developed. Section 4 discusses the current Matlab/Simulink model. This is a low-fidelity, high-speed model used to quickly evaluate and compare competing control and protection concepts. Section 5 describes the Modelica models developed by POLIMI and Westinghouse. The object-oriented Modelica language provides convenient mechanisms for developing models at several levels of detail. We have used this to develop a high-fidelity model for detailed analyses and a faster-running simplified model to help speed the I and C development process. Section
Evaluation of two-fluid and drift flux thermohydraulics in APROS code environment
International Nuclear Information System (INIS)
Miettinen, J.; Karppinen, I.; Haenninen, M.; Ylijoki, J.
1999-01-01
The characteristics of the thermohydraulic solutions in APROS are considered for the nuclear power plant modelling. The thermohydraulic model of the APROS plant analyzer includes three levels of solutions, homogeneous 3-equation model, 5-equation drift flux model and 6-equation two-fluid model. In practical modelling of versatile process systems different approaches are selected for different types of the power plant sections. The 3-equation model is used for turbines and auxiliary systems. The 5-equation model and 6-equation model are alternative models for main process sections of the primary and secondary sides. The 5-equation model has been typically selected for the real time applications and the 6-equation model for the safety analysis applications. The validation needs for both approaches are the same. Because the change of the solution mode is an easy task in APROS, the validation tasks are typically performed in parallel for 5-equation and 6-equation models. By calculating in parallel with both models systematic errors in solutions may be pointed out. The testing against both separate effects tests and integral tests is an essential part in the thermohydraulics. In different plant applications different physical features are important. The analysis requirements vary from one application to another. When nodalizations together with increased computer speed are growing up, the earlier validation cases may be insufficient. That is why the content of the code has to be known in detail. Such an expertise in the code development has to be gained that properties of the code against other thermohydraulics codes are known. (author)
Models for Type Ia Supernovae and Related Astrophysical Transients
Röpke, Friedrich K.; Sim, Stuart A.
2018-06-01
We give an overview of recent efforts to model Type Ia supernovae and related astrophysical transients resulting from thermonuclear explosions in white dwarfs. In particular we point out the challenges resulting from the multi-physics multi-scale nature of the problem and discuss possible numerical approaches to meet them in hydrodynamical explosion simulations and radiative transfer modeling. We give examples of how these methods are applied to several explosion scenarios that have been proposed to explain distinct subsets or, in some cases, the majority of the observed events. In case we comment on some of the successes and shortcoming of these scenarios and highlight important outstanding issues.
International Nuclear Information System (INIS)
Domijan, A.D. Jr.; Emami, M.V.
1990-01-01
This paper reports on a simulation of a MHO distance relay developed to study the effect of its operation under various system conditions. Simulation is accomplished using a state space approach and a modeling technique using ElectroMagnetic Transient Program (Transient Analysis of Control Systems). Furthermore, simulation results are compared with those obtained in another independent study as a control, to validate the results. A data code for the practical utilization of this simulation is given
Analysis of forced convective transient boiling by homogeneous model of two-phase flow
International Nuclear Information System (INIS)
Kataoka, Isao
1985-01-01
Transient forced convective boiling is of practical importance in relation to the accident analysis of nuclear reactor etc. For large length-to-diameter ratio, the transient boiling characteristics are predicted by transient two-phase flow calculations. Based on homogeneous model of two-phase flow, the transient forced convective boiling for power and flow transients are analysed. Analytical expressions of various parameters of transient two-phase flow have been obtained for several simple cases of power and flow transients. Based on these results, heat flux, velocity and time at transient CHF condition are predicted analytically for step and exponential power increases, and step, exponential and linear velocity decreases. The effects of various parameters on heat flux, velocity and time at transient CHF condition have been clarified. Numerical approach combined with analytical method is proposed for more complicated cases. Solution method for pressure transient are also described. (author)
Grounding line transient response in marine ice sheet models
Directory of Open Access Journals (Sweden)
A. S. Drouet
2013-03-01
Full Text Available Marine ice-sheet stability is mostly controlled by the dynamics of the grounding line, i.e. the junction between the grounded ice sheet and the floating ice shelf. Grounding line migration has been investigated within the framework of MISMIP (Marine Ice Sheet Model Intercomparison Project, which mainly aimed at investigating steady state solutions. Here we focus on transient behaviour, executing short-term simulations (200 yr of a steady ice sheet perturbed by the release of the buttressing restraint exerted by the ice shelf on the grounded ice upstream. The transient grounding line behaviour of four different flowline ice-sheet models has been compared. The models differ in the physics implemented (full Stokes and shallow shelf approximation, the numerical approach, as well as the grounding line treatment. Their overall response to the loss of buttressing is found to be broadly consistent in terms of grounding line position, rate of surface elevation change and surface velocity. However, still small differences appear for these latter variables, and they can lead to large discrepancies (> 100% observed in terms of ice sheet contribution to sea level when cumulated over time. Despite the recent important improvements of marine ice-sheet models in their ability to compute steady state configurations, our results question the capacity of these models to compute short-term reliable sea-level rise projections.
International Nuclear Information System (INIS)
Lazaro, A.; Ammirabile, L.; Martorell, S.
2014-01-01
The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)
Thermohydraulic stability coupled to the neutronic in a BWR
International Nuclear Information System (INIS)
Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.
2006-01-01
In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the Laguna Verde
Energy Technology Data Exchange (ETDEWEB)
Kostadinov, V; Petelin, S; Sarler, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)
1988-07-01
Nuclear safety analysis of NPP Krsko core operating at full power with 4% steam generator tubes plugged have been performed. Influence of individual parameters on core thermohydraulic performance have been evaluated. Using COBRA-III-C computer code we have analysed a core design (evaluation) model. The DNBR change was calculated as a consequence of 4% plugging. The influence of thermohydraulic parameters change on DNBR was analysed. (author)
Grounding modelling for transient overvoltage simulation in electric power transmission
International Nuclear Information System (INIS)
Moreno O, German; Valencia V, Jaime A; Villada, Fernando
1992-01-01
Grounding plays an important role in transmission line outages and consequently on electric energy transmission quality indexes. Fundamentals of an accurate modelling for transient behaviour analysis, particularly for the response of transmission lines to lightning, are presented. Also, a method to take into account the electromagnetic propagation guided by the grounding electrodes and finally to assess the grounding impedance in order to simulate the transmission line behaviour under lightning is presented. Analysis of impedance behaviour for diverse configurations and simulation results of over voltages on a real 220 kV line are presented to illustrate the capabilities of the method and of the computational program developed
Modeling and Analyzing Transient Military Air Traffic Control
2010-12-01
arrive and be serviced. In general, for n flights, the number of ways that flights can enter and leave the ATC is given by the nth Catalan number ...collection of information if it does not display a currently valid OMB control number . 1. REPORT DATE DEC 2010 2. REPORT TYPE 3. DATES COVERED 00-00...2010 to 00-00-2010 4. TITLE AND SUBTITLE Modeling and Analyzing Transient Military Air Traffic Control 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c
A transient model of a cesium-barium diode
International Nuclear Information System (INIS)
Luke, J.R.; El-Genk, M.S.
1995-01-01
In this work a transient model of a Cs-Ba diode is developed, and a series of experiments is performed using a diode equipped with Langmuir probes. The Langmuir probe data show that the electron energy distribution is non-Maxwellian at low discharge currents, indicating the presence of an electron beam from the emitter. Experimental results also showed that the plasma properties are non-homogeneous across the 1 mm diode gap; the electron temperature and plasma potential were higher near the emitter and the plasma density was higher near the collector. Experimental evidence is presented to show that the discharge contracts to a filament below the maximum thermal emission current
A flexible multipurpose model for normal and transient cell kinetics
International Nuclear Information System (INIS)
Toivonen, Harri.
1979-07-01
The internal hypothetical compartments within the different phases of the cell cycle have been adopted as the basis of models dealing with various specific problems in cell kinetics. This approach was found to be of more general validity, extending from expanding cell populations to complex maturation processes. The differential equations describing the system were solved with an effective, commercially available library subroutine. Special attention was devoted to analysis of transient and feedback kinetics of cell populations encountered in diverse environmental and exposure conditions, for instance in cases of wounding and radiation damage. (author)
Transient modeling of electrochemically assisted CO2 capture and release
DEFF Research Database (Denmark)
Singh, Shobhana; Stechel, Ellen B.; Buttry, Daniel A.
2017-01-01
to analyze the time-dependent behavior of CO2 capture and electro-migration transport across the cell length. Given high nonlinearity of the system, we used a finite element method (FEM) to numerically solve the coupled mass transport equations. The model describes the concentration profiles by taking......The present work aims to develop a model of a new electrochemical CO2 separation and release technology. We present a one-dimensional transient model of an electrochemical cell for point source CO2 capture and release, which mainly focuses on the simultaneous mass transport and complex chemical...... reactions associated with the separation process. For concreteness, we use an ionic liquid (IL) with 2 M thiolate anion (RS−) in 1 M disulfide (RSSR) as an electrolyte in the electrochemical cell to capture, transport and release CO2 under standard operating conditions. We computationally solved the model...
Heat-pipe transient model for space applications
International Nuclear Information System (INIS)
Tournier, J.; El-Genk, M.S.; Juhasz, A.J.
1991-01-01
A two-dimensional model is developed for simulating heat pipes transient performance following changes in the input/rejection power or in the evaporator/condenser temperatures. The model employs the complete form of governing equations and momentum and energy jump conditions at the liquid-vapor interface. Although the model is capable of handling both cylindrical and rectangular geometries, the results reported are for a circular heat pipe with liquid lithium as the working fluid. The model incorporates a variety of other working fluids, such as water, ammonia, potassium, sodium, and mercury, and offers combinations of isothermal, isoflux, convective and radiative heating/cooling conditions in the evaporator and condenser regions of the heat pipe. Results presented are for lithium heat pipes with exponential heating of the evaporator and isothermal cooling of the condenser
Transient behaviour of main coolant pump in nuclear power plants
International Nuclear Information System (INIS)
Delja, A.
1986-01-01
A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)
Physical modelling of a rapid boron dilution transient
Energy Technology Data Exchange (ETDEWEB)
Anderson, N.G.; Hemstroem, B.; Karlsson, R. [Vattenfall Utveckling AB, Aelvkarleby (Sweden); Jacobson, S. [Vattenfall AB, Ringhals, Vaeroebacka (Sweden)
1995-09-01
The analysis of boron dilution accidents in pressurised water reactors has traditionally assumed that mixing is instantaneous and complete everywhere, eliminating in this way the possibility of concentration inhomogeneities. Situations can nevertheless arise where a volume of coolant with a low boron concentration may eventually enter the core and generate a severe reactivity transient. The work presented in this paper deals with a category of Rapid Boron Dilution Events characterised by a rapid start of a Reactor Coolant Pump (RCP) with a plug of relatively unborated water present in the RCS pipe. Model tests have been made at Vattenfall Utveckling AB in a simplified 1:5 scale model of a Westinghouse PWR. Conductivity measurements are used to determine dimensionless boron concentration. The main purpose of this experimental work is to define an experimental benchmark against which a mathematical model can be tested. The final goal is to be able to numerically predict Boron Dilution Transients. This work has been performed as a part of a Co-operative Agreement with Electricite` de France (EDF).
Computer simulation of WWER - 440 normal and emergency transient operating conditions
International Nuclear Information System (INIS)
Izbeshesku, M.; Rajka, V.; Untaru, S.; Dumitresku, A.; Paneh, M.; Turku, I.
1976-01-01
Results of computer realization of a model for studying transient process in the nuclear system of vapour production at the WWER - 40 peactor nuclear power plant are presented. The first circuit model consists of a number of modules, corresponding to its main parts: for each module derived were the equations describing neutron and thermohydraulic parameters. The second circuit effect is taken into account by heat quantity accepted from a steam generator. The equations are mostly differential with constant coefficients. Coefficient values and initial values of physical quantities are evaluated according to the technical literature. Both manual and automatic operations are modelled [ru
FDTD modelling of induced polarization phenomena in transient electromagnetics
Commer, Michael; Petrov, Peter V.; Newman, Gregory A.
2017-04-01
The finite-difference time-domain scheme is augmented in order to treat the modelling of transient electromagnetic signals containing induced polarization effects from 3-D distributions of polarizable media. Compared to the non-dispersive problem, the discrete dispersive Maxwell system contains costly convolution operators. Key components to our solution for highly digitized model meshes are Debye decomposition and composite memory variables. We revert to the popular Cole-Cole model of dispersion to describe the frequency-dependent behaviour of electrical conductivity. Its inversely Laplace-transformed Debye decomposition results in a series of time convolutions between electric field and exponential decay functions, with the latter reflecting each Debye constituents' individual relaxation time. These function types in the discrete-time convolution allow for their substitution by memory variables, annihilating the otherwise prohibitive computing demands. Numerical examples demonstrate the efficiency and practicality of our algorithm.
Comparison of transient PCRV model test results with analysis
International Nuclear Information System (INIS)
Marchertas, A.H.; Belytschko, T.B.
1979-01-01
Comparisons are made of transient data derived from simple models of a reactor containment vessel with analytical solutions. This effort is a part of the ongoing process of development and testing of the DYNAPCON computer code. The test results used in these comparisons were obtained from scaled models of the British sodium cooled fast breeder program. The test structure is a scaled model of a cylindrically shaped reactor containment vessel made of concrete. This concrete vessel is prestressed axially by holddown bolts spanning the top and bottom slabs along the cylindrical walls, and is also prestressed circumferentially by a number of cables wrapped around the vessel. For test purposes this containment vessel is partially filled with water, which comes in direct contact with the vessel walls. The explosive charge is immersed in the pool of water and is centrally suspended from the top of the vessel. The load history was obtained from an ICECO analysis, using the equations of state for the source and the water. A detailed check of this solution was made to assure that the derived loading did provide the correct input. The DYNAPCON code was then used for the analysis of the prestressed concrete containment model. This analysis required the simulation of prestressing and the response of the model to the applied transient load. The calculations correctly predict the magnitudes of displacements of the PCRV model. In addition, the displacement time histories obtained from the calculations reproduce the general features of the experimental records: the period elongation and amplitude increase as compared to an elastic solution, and also the absence of permanent displacement. However, the period still underestimates the experiment, while the amplitude is generally somewhat large
Modeling transient streaming potentials in falling-head permeameter tests.
Malama, Bwalya; Revil, André
2014-01-01
We present transient streaming potential data collected during falling-head permeameter tests performed on samples of two sands with different physical and chemical properties. The objective of the work is to estimate hydraulic conductivity (K) and the electrokinetic coupling coefficient (Cl ) of the sand samples. A semi-empirical model based on the falling-head permeameter flow model and electrokinetic coupling is used to analyze the streaming potential data and to estimate K and Cl . The values of K estimated from head data are used to validate the streaming potential method. Estimates of K from streaming potential data closely match those obtained from the associated head data, with less than 10% deviation. The electrokinetic coupling coefficient was estimated from streaming potential vs. (1) time and (2) head data for both sands. The results indicate that, within limits of experimental error, the values of Cl estimated by the two methods are essentially the same. The results of this work demonstrate that a temporal record of the streaming potential response in falling-head permeameter tests can be used to estimate both K and Cl . They further indicate the potential for using transient streaming potential data as a proxy for hydraulic head in hydrogeology applications. © 2013, National Ground Water Association.
International Nuclear Information System (INIS)
Bujan, A.; Adamik, V.; Misak, J.
1986-01-01
A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)
International Nuclear Information System (INIS)
Daverio, H; Juanico, L
2000-01-01
Comparative analysis of thermohydraulic margins were studied of the CANDU 37 and CARA fuel bundles (FB) in Embalse power station with COBRA IV-HW code ., the geometry of the bundle laying on the channel was particularly modeled and discussing the results in comparison with former calculations with 1/6 simetry .The CARA design with enriched uranium (0.9 %) and extended burn up lets maintain the current thermohydraulic nominal margins , while compared with CANDU 37 rods FB enriched , the CARA design permits widely improve the current margins
Transient Heating and Thermomechanical Stress Modeling of Ceramic HEPA Filters
Energy Technology Data Exchange (ETDEWEB)
Bogle, Brandon [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kelly, James [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Haslam, Jeffrey [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2017-09-29
The purpose of this report is to showcase an initial finite-element analysis model of a ceramic High-Efficiency Particulate (HEPA) Air filter design. Next generation HEPA filter assemblies are being developed at LLNL to withstand high-temperature fire scenarios by use of ceramics and advanced materials. The filters are meant for use in radiological and nuclear facilities, and are required to survive 500°C fires over an hour duration. During such conditions, however, collecting data under varying parameters can be challenging; therefore, a Finite Element Analysis model of the filter was conducted using COMSOL ® Multiphysics to analyze the effects of fire. Finite Element Analysis (FEA) modelling offers several opportunities: researchers can quickly and easily consider impacts of potential design changes, material selection, and flow characterization on filter performance. Specifically, this model provides stress references for the sealant at high temperatures. Modeling of full filter assemblies was deemed inefficient given the computational requirements, so a section of three tubes from the assembly was modeled. The model looked at the transient heating and thermomechanical stress development during a 500°C air flow at 6 CFM. Significant stresses were found at the ceramic-metal interfaces of the filter, and conservative temperature profiles at locations of interest were plotted. The model can be used for the development of sealants that minimize stresses at the ceramic-metal interface. Further work on the model would include the full filter assembly and consider heat losses to make more accurate predictions.
Energy Technology Data Exchange (ETDEWEB)
Wiles, L.E.; McCann, R.A.
1981-09-01
The numerical modeling capability that has been developed at the Pacific Northwest Laboratory (PNL) for the prediction of the thermohydraulic performance of porous media reservoirs for compressed air energy storage (CAES) is described. The capability of the numerical models was demonstrated by application to a variety of parametric analyses and the support analyses for the CAES porous media field demonstration program. The demonstration site analyses include calculations for the displacement of aquifer water to develop the air storage zone, the potential for water coning, thermal development in the reservoir, and the dehydration of the near-wellbore region. Unique features of the demonstration site reservoir that affect the thermohydraulic performance are identified and contrasted against the predicted performance for conditions that would be considered more typical of a commercial CAES site.
International Nuclear Information System (INIS)
Zhukov, A.V.; Sorokin, A.P.
2000-01-01
The problems of numerical modeling of thermohydraulics in assembly of fuel elements of fast reactors with the partial blockage of cross-section under the coolant are considered. The information about existing codes constructed on use of subchannel technique and model of porous body are presented. The results of calculation obtained by these codes are presented. (author)
Modeling of Transient Nectar Flow in Hummingbird Tongues
Rico-Guevara, Alejandro; Fan, Tai-Hsi; Rubega, Margaret
2015-11-01
We demonstrate that hummingbirds do not pick up floral nectar via capillary action. The long believed capillary rise models were mistaken and unable to predict the dynamic nectar intake process. Instead, hummingbird's tongue acts as an elastic micropump. Nectar is drawn into the tongue grooves during elastic expansion after the grooves are squeezed flat by the beak. The new model is compared with experimental data from high-speed videos of 18 species and tens of individuals of wild hummingbirds. Self-similarity and transitions of short-to-long time behaviours have been resolved for the nectar flow driven by expansive filling. The transient dynamics is characterized by the relative contributions of negative excess pressure and the apparent area modulus of the tongue grooves.
Application of CFD methods in research of SCWR thermo-hydraulics
International Nuclear Information System (INIS)
Zeng Xiaokang; Li Yongliang; Yan Xiao; Xiao Zejun; Huang Yanping
2013-01-01
The CFD method has been an important tool in the research of SCWR thermo- hydraulics. Currently, the CFD methods uses commonly the subcritical turbulence models, which can not accurately simulate the gravity and thermal expansion acceleration effect, and CFD numerical method is not applicable when the heat flux is large. The paper summarizes the application status of the CFD methods in the research of SCWR thermo-hydraulics in RETH. (authors)
Neutronics and thermohydraulics of the reactor C.E.N.E. Part II
International Nuclear Information System (INIS)
Caro, R.
1976-01-01
In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs
Reactivity transient calculatios in research reactor
International Nuclear Information System (INIS)
Santos, R.S. dos
1986-01-01
A digital program for reactivity transient analysis in research reactor and cylindrical geometry was showed quite efficient when compared with methods and programs of the literature, as much in the solution of the neutron kinetics equation as in the thermohydraulic. An improvement in the representation of the feedback reactivity adopted on the program reduced markedly the computation time, with some accuracy. (Author) [pt
An Effective Distributed Model for Power System Transient Stability Analysis
Directory of Open Access Journals (Sweden)
MUTHU, B. M.
2011-08-01
Full Text Available The modern power systems consist of many interconnected synchronous generators having different inertia constants, connected with large transmission network and ever increasing demand for power exchange. The size of the power system grows exponentially due to increase in power demand. The data required for various power system applications have been stored in different formats in a heterogeneous environment. The power system applications themselves have been developed and deployed in different platforms and language paradigms. Interoperability between power system applications becomes a major issue because of the heterogeneous nature. The main aim of the paper is to develop a generalized distributed model for carrying out power system stability analysis. The more flexible and loosely coupled JAX-RPC model has been developed for representing transient stability analysis in large interconnected power systems. The proposed model includes Pre-Fault, During-Fault, Post-Fault and Swing Curve services which are accessible to the remote power system clients when the system is subjected to large disturbances. A generalized XML based model for data representation has also been proposed for exchanging data in order to enhance the interoperability between legacy power system applications. The performance measure, Round Trip Time (RTT is estimated for different power systems using the proposed JAX-RPC model and compared with the results obtained using traditional client-server and Java RMI models.
International Nuclear Information System (INIS)
Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.
1983-01-01
The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)
Transient heat conduction in a pebble fuel applying fractional model
International Nuclear Information System (INIS)
Gomez A, R.; Espinosa P, G.
2009-10-01
In this paper we presents the equation of thermal diffusion of temporary-fractional order in the one-dimensional space in spherical coordinates, with the objective to analyze the heat transference between the fuel and coolant in a fuel element of a Pebble Bed Modular Reactor. The pebble fuel is the heterogeneous system made by microsphere constitutes by U O, pyrolytic carbon and silicon carbide mixed with graphite. To describe the heat transfer phenomena in the pebble fuel we applied a constitutive law fractional (Non-Fourier) in order to analyze the behaviour transient of the temperature distribution in the pebble fuel with anomalous thermal diffusion effects a numerical model is developed. (Author)
A distributed parameter wire model for transient electrical discharges
International Nuclear Information System (INIS)
Maier, W.B. II; Kadish, A.; Sutherland, C.D.; Robiscoe, R.T.
1990-01-01
A model for freely propagating transient electrical discharges, such as lightning and punch-through arcs, is developed in this paper. We describe the electromagnetic fields by Maxwell's equations and we represent the interaction of electric fields with the medium to produce current by ∂J/∂t=ω 2 (E-E*J)/4π, where ω and E* are parameters characteristic of the medium, J≡current density, and J≡J/|J|. We illustrate the properties of this model for small-diameter, guided, cylindrically symmetric discharges. Analytic, numerical, and approximate solutions are given for special cases. The model describes, in a new and comprehensive fashion, certain macroscopic discharge properties, such as threshold behavior, quenching and reignition, path tortuosity, discharge termination with nonzero charge density remaining along the discharge path, and other experimentally observed discharge phenomena. Fields, current densities, and charge densities are quantitatively determined from given boundary and initial conditions. We suggest that many macroscopic discharge properties are properly explained by the model as electromagnetic phenomena, and we discuss extensions of the model to include chemistry, principally ionization and recombination
Fuel rod modelling during transients: The TOUTATIS code
International Nuclear Information System (INIS)
Bentejac, F.; Bourreau, S.; Brochard, J.; Hourdequin, N.; Lansiart, S.
2001-01-01
The TOUTATIS code is devoted to the PCI local phenomena simulation, in correlation with the METEOR code for the global behaviour of the fuel rod. More specifically, the TOUTATIS objective is to evaluate the mechanical constraints on the cladding during a power transient thus predicting its behaviour in term of stress corrosion cracking. Based upon the finite element computation code CASTEM 2000, TOUTATIS is a set of modules written in a macro language. The aim of this paper is to present both code modules: The axisymmetric bi-dimensional module, modeling a unique block pellet; The tri dimensional module modeling a radially fragmented pellet. Having shown the boundary conditions and the algorithms used, the application will be illustrated by: A short presentation of the bidimensional axisymmetric modeling performances as well as its limits; The enhancement due to the three dimensional modeling will be displayed by sensitivity studies to the geometry, in this case the pellet height/diameter ratio. Finally, we will show the easiness of the development inherent to the CASTEM 2000 system by depicting the process of a modeling enhancement by adding the possibility of an axial (horizontal) fissuration of the pellet. As conclusion, the future improvements planned for the code are depicted. (author)
Transient modelling of a natural circulation loop under variable pressure
International Nuclear Information System (INIS)
Vianna, Andre L.B.; Faccini, Jose L.H.; Su, Jian; Instituto de Engenharia Nuclear
2017-01-01
The objective of the present work is to model the transient operation of a natural circulation loop, which is one-tenth scale in height to a typical Passive Residual Heat Removal system (PRHR) of an Advanced Pressurized Water Nuclear Reactor and was designed to meet the single and two-phase flow similarity criteria to it. The loop consists of a core barrel with electrically heated rods, upper and lower plena interconnected by hot and cold pipe legs to a seven-tube shell heat exchanger of countercurrent design, and an expansion tank with a descending tube. A long transient characterized the loop operation, during which a phenomenon of self-pressurization, without self-regulation of the pressure, was experimentally observed. This represented a unique situation, named natural circulation under variable pressure (NCVP). The self-pressurization was originated in the air trapped in the expansion tank and compressed by the loop water dilatation, as it heated up during each experiment. The mathematical model, initially oriented to the single-phase flow, included the heat capacity of the structure and employed a cubic polynomial approximation for the density, in the buoyancy term calculation. The heater was modelled taking into account the different heat capacities of the heating elements and the heater walls. The heat exchanger was modelled considering the coolant heating, during the heat exchanging process. The self-pressurization was modelled as an isentropic compression of a perfect gas. The whole model was computationally implemented via a set of finite difference equations. The corresponding computational algorithm of solution was of the explicit, marching type, as for the time discretization, in an upwind scheme, regarding the space discretization. The computational program was implemented in MATLAB. Several experiments were carried out in the natural circulation loop, having the coolant flow rate and the heating power as control parameters. The variables used in the
Transient modelling of a natural circulation loop under variable pressure
Energy Technology Data Exchange (ETDEWEB)
Vianna, Andre L.B.; Faccini, Jose L.H.; Su, Jian, E-mail: avianna@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br, E-mail: faccini@ien.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Termo-Hidraulica Experimental
2017-07-01
The objective of the present work is to model the transient operation of a natural circulation loop, which is one-tenth scale in height to a typical Passive Residual Heat Removal system (PRHR) of an Advanced Pressurized Water Nuclear Reactor and was designed to meet the single and two-phase flow similarity criteria to it. The loop consists of a core barrel with electrically heated rods, upper and lower plena interconnected by hot and cold pipe legs to a seven-tube shell heat exchanger of countercurrent design, and an expansion tank with a descending tube. A long transient characterized the loop operation, during which a phenomenon of self-pressurization, without self-regulation of the pressure, was experimentally observed. This represented a unique situation, named natural circulation under variable pressure (NCVP). The self-pressurization was originated in the air trapped in the expansion tank and compressed by the loop water dilatation, as it heated up during each experiment. The mathematical model, initially oriented to the single-phase flow, included the heat capacity of the structure and employed a cubic polynomial approximation for the density, in the buoyancy term calculation. The heater was modelled taking into account the different heat capacities of the heating elements and the heater walls. The heat exchanger was modelled considering the coolant heating, during the heat exchanging process. The self-pressurization was modelled as an isentropic compression of a perfect gas. The whole model was computationally implemented via a set of finite difference equations. The corresponding computational algorithm of solution was of the explicit, marching type, as for the time discretization, in an upwind scheme, regarding the space discretization. The computational program was implemented in MATLAB. Several experiments were carried out in the natural circulation loop, having the coolant flow rate and the heating power as control parameters. The variables used in the
Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes
International Nuclear Information System (INIS)
Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.
2002-01-01
A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)
International Nuclear Information System (INIS)
2014-02-01
The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient
Three-dimensional reactor model for the Paks NPP full-scope simulator
International Nuclear Information System (INIS)
Gyori, C.; Hegyi, G.; Hozer, Z.; Kereszturi, A.; Maraczy, C.
1993-01-01
The reactor model includes thermohydraulic and neutron-physical components. The thermohydraulic model is based on the SMABRE code developed at the Technical Research Centre of Finland for the analysis of loss-of-coolant transients in PWRs. The fuel rod model will be replaced by a new software module providing a comprehensive description of the behavior of fuel rods during reactor transients and hypothetical accidents. The calculation is performed in four individual models: fuel rod temperature model, fuel rod internal pressure model, fuel rod deformation model and fuel rod failure model. In the neutron-physical model the core is calculated with nodes for all of the 349 fuel assemblies, and each assembly is calculated in ten layers. (Z.S.) 1 fig., 5 refs
Thermohydraulic behavior of liquid metal pool submitted to electronic bombardment
International Nuclear Information System (INIS)
Brun, Patrice
1998-01-01
This thesis deals with the thermohydraulics of liquid metal molten by an electron beam. We study the relationship between the liquid metal pool and the vapor rate. The aim is to find good conditions increasing the metal vapor rate. In first place, energy losses are identified. Mains are convection (buoyancy and thermo-capillary) strengthen by the deformation of the molten pool. The first action is to reduce the liquid interface deformation with a transient spot realized by scanning the electron beam. I find that in this case, the optimum vapor rate is obtained when the crossing time of the beam is smaller than characteristic time of formation of the cavity, but greater than the heating time of the surface. Secondly, I impose forces to change the morphology of the flow. Two actions are tried: magnetic field application and rotating motion of the crucible. External magnetic field application may reduce convective flow, by the creation of a magnetic brake. But in my experiment, magnetic field deteriorates electron beam before to be effective. Results obtained by the rotating motion of the crucible approve this choice to reduce energy losses and increase vapor rate. This growth of vapor rate is due to an expansion of the emitted vapor source and an increase of the central temperature of the molten pool. Nevertheless with the increase of the rotation velocity and after the optimum vapor rate, I note that the flow is not axisymmetric. My observation give to think about instabilities that are developed by baroclinic waves. The comparison of my works with the Eady's linear theory gives good results. (author) [fr
Three-dimensional transient electromagnetic modeling in the Laplace Domain
International Nuclear Information System (INIS)
Mizunaga, H.; Lee, Ki Ha; Kim, H.J.
1998-01-01
In modeling electromagnetic responses, Maxwell's equations in the frequency domain are popular and have been widely used (Nabighian, 1994; Newman and Alumbaugh, 1995; Smith, 1996, to list a few). Recently, electromagnetic modeling in the time domain using the finite difference (FDTD) method (Wang and Hohmann, 1993) has also been used to study transient electromagnetic interactions in the conductive medium. This paper presents a new technique to compute the electromagnetic response of three-dimensional (3-D) structures. The proposed new method is based on transforming Maxwell's equations to the Laplace domain. For each discrete Laplace variable, Maxwell's equations are discretized in 3-D using the staggered grid and the finite difference method (FDM). The resulting system of equations is then solved for the fields using the incomplete Cholesky conjugate gradient (ICCG) method. The new method is particularly effective in saving computer memory since all the operations are carried out in real numbers. For the same reason, the computing speed is faster than frequency domain modeling. The proposed approach can be an extremely useful tool in developing an inversion algorithm using the time domain data
Comparison of transient PCRV model test results with analysis
International Nuclear Information System (INIS)
Marchertas, A.H.; Belytschko, T.B.
1979-01-01
Comparisons are made of transient data derived from simple models of a reactor containment vessel with analytical solutions. This effort is a part of the ongoing process of development and testing of the DYNAPCON computer code. The test results used in these comparisons were obtained from scaled models of the British sodium cooled fast breeder program. The test structure is a scaled model of a cylindrically shaped reactor containment vessel made of concrete. This concrete vessel is prestressed axially by holddown bolts spanning the top and bottom slabs along the cylindrical walls, and is also prestressed circumferentially by a number of cables wrapped around the vessel. For test purposes this containment vessel is partially filled with water, which comes in direct contact with the vessel walls. The explosive charge is immersed in the pool of water and is centrally suspended from the top of the vessel. The tests are very similar to the series of tests made for the COVA experimental program, but the vessel here is the prestressed concrete container. (orig.)
Reduction of collisional-radiative models for transient, atomic plasmas
Abrantes, Richard June; Karagozian, Ann; Bilyeu, David; Le, Hai
2017-10-01
Interactions between plasmas and any radiation field, whether by lasers or plasma emissions, introduce many computational challenges. One of these computational challenges involves resolving the atomic physics, which can influence other physical phenomena in the radiated system. In this work, a collisional-radiative (CR) model with reduction capabilities is developed to capture the atomic physics at a reduced computational cost. Although the model is made with any element in mind, the model is currently supplemented by LANL's argon database, which includes the relevant collisional and radiative processes for all of the ionic stages. Using the detailed data set as the true solution, reduction mechanisms in the form of Boltzmann grouping, uniform grouping, and quasi-steady-state (QSS), are implemented to compare against the true solution. Effects on the transient plasma stemming from the grouping methods are compared. Distribution A: Approved for public release; unlimited distribution, PA (Public Affairs) Clearance Number 17449. This work was supported by the Air Force Office of Scientific Research (AFOSR), Grant Number 17RQCOR463 (Dr. Jason Marshall).
Atucha I nuclear power plant transients analysis
International Nuclear Information System (INIS)
Castano, J.; Schivo, M.
1987-01-01
A program for the transients simulation thermohydraulic calculation without loss of coolant (KWU-ENACE development) to evaluate Atucha I nuclear power plant behaviour is used. The program includes systems simulation and nuclear power plants control bonds with real parameters. The calculation results show a good agreement with the output 'protocol' of various transients of the nuclear power plant, keeping the error, in general, lesser than ± 10% from the variation of the nuclear power plant's state variables. (Author)
Experiment on thermohydraulics of simulated control rod
International Nuclear Information System (INIS)
Ogawa, Masuro; Ouchi, Mitsuo; Akino, Norio; Fujimura, Kaoru; Shiina, Yasuaki; Kawamura, Hiroshi
1984-10-01
A thermohydraulic study of a control rod channel is required for the core design of the Very High Temperature Gas Cooled Reactor (VHTR). A non-heating experiment with air flow was performed prior to heating experiment with helium flow. Experimental results on stability of flow, flow rate distribution and pressure drop of the control rod channel are reported. In a test section of the experimental apparatus, five simulated control subrods were suspended vertically in a circular duct. Their dimension was in coincide with those of the Detailed Disign (I) of the VHTR. Air of atomospheric pressure was used as a coolant gas, which flowed in inner and outer paths of the subrods. Total flow rate ranged from 0.0011 to 0.0062 kg/s. Flow rate distribution and pressure drop were obtained for various flow rates. Velocity fluctuation in the channel was also observed using a hot wire anemometer. From these experiments, it was found that the flow rate distribution was nearly the same as a disigned value and that turbulent and laminar flows were simultaneously realized in outer and inner paths respectively. These observations supported a feasibility of the present design. (author)
MODELING OF TRANSIENT HEAT TRANSFER IN FOAMED CONCRETE SLAB
Directory of Open Access Journals (Sweden)
MD AZREE OTHUMAN MYDIN
2013-06-01
Full Text Available This paper reports the basis of one-dimensional Finite Difference method to obtain thermal properties of foamed concrete in order to solve transient heat conduction problems in multi-layer panels. In addition, this paper also incorporates the implementation of the method and the validation of thermal properties model of foamed concrete. A one-dimensional finite difference heat conduction programme has been developed to envisage the temperature development through the thickness of the foamed concrete slab, based on an initial estimate of the thermal conductivity-temperature relationship as a function of porosity and radiation within the voids. The accuracy of the model was evaluated by comparing predicted and experimental temperature profiles obtained from small scale heat transfer test on foamed concrete slabs, so that the temperature history of the specimen calculated by the programme closely matches those recorded during the experiment. Using the thermal properties of foamed concrete, the validated heat transfer program predicts foamed concrete temperatures in close agreement with experimental results obtained from a number of high temperature tests. The proposed numerical and thermal properties are simple yet efficient and can be utilised to aid manufacturers to develop their products without having to conduct numerous large-scale fire tests.
Modeling transient heat transfer in nuclear waste repositories.
Yang, Shaw-Yang; Yeh, Hund-Der
2009-09-30
The heat of high-level nuclear waste may be generated and released from a canister at final disposal sites. The waste heat may affect the engineering properties of waste canisters, buffers, and backfill material in the emplacement tunnel and the host rock. This study addresses the problem of the heat generated from the waste canister and analyzes the heat distribution between the buffer and the host rock, which is considered as a radial two-layer heat flux problem. A conceptual model is first constructed for the heat conduction in a nuclear waste repository and then mathematical equations are formulated for modeling heat flow distribution at repository sites. The Laplace transforms are employed to develop a solution for the temperature distributions in the buffer and the host rock in the Laplace domain, which is numerically inverted to the time-domain solution using the modified Crump method. The transient temperature distributions for both the single- and multi-borehole cases are simulated in the hypothetical geological repositories of nuclear waste. The results show that the temperature distributions in the thermal field are significantly affected by the decay heat of the waste canister, the thermal properties of the buffer and the host rock, the disposal spacing, and the thickness of the host rock at a nuclear waste repository.
Three-Component Forward Modeling for Transient Electromagnetic Method
Directory of Open Access Journals (Sweden)
Bin Xiong
2010-01-01
Full Text Available In general, the time derivative of vertical magnetic field is considered only in the data interpretation of transient electromagnetic (TEM method. However, to survey in the complex geology structures, this conventional technique has begun gradually to be unsatisfied with the demand of field exploration. To improve the integrated interpretation precision of TEM, it is necessary to study the three-component forward modeling and inversion. In this paper, a three-component forward algorithm for 2.5D TEM based on the independent electric and magnetic field has been developed. The main advantage of the new scheme is that it can reduce the size of the global system matrix to the utmost extent, that is to say, the present is only one fourth of the conventional algorithm. In order to illustrate the feasibility and usefulness of the present algorithm, several typical geoelectric models of the TEM responses produced by loop sources at air-earth interface are presented. The results of the numerical experiments show that the computation speed of the present scheme is increased obviously and three-component interpretation can get the most out of the collected data, from which we can easily analyze or interpret the space characteristic of the abnormity object more comprehensively.
Thermo-hydraulic analysis of the generic equatorial port plug design
International Nuclear Information System (INIS)
Rodríguez, E.; Guirao, J.; Ordieres, J.; Cortizo, J.L.; Iglesias, S.
2012-01-01
Highlights: ► Thermo-hydraulic transient performance evaluation and optimization of the GEPP structure cooling/heating system under neutronic heating and baking conditions. ► The optimization of the GEPP box structure's cooling system includes positioning and minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions. - Abstract: The port-based ITER diagnostic systems are housed primarily in two locations, the equatorial and upper port plugs. The port plug structure provides confinement function, maintains ultra-high vacuum quality and the first confinement barrier for radioactive materials at the ports. The port plug structure design, from the ITER International Organisation (IO), is cooled and heated by pressurized water which flows through a series of gun-drilled water channels and water pipes. The cooling function is required to remove nuclear heating due to radiation during operation of ITER, while the heating function is intended to heat up uniformly the machine during baking condition. The work presented provides coupled thermo-hydraulic analysis and optimization of a Generic Equatorial Port Plug (GEPP) structure cooling and heating system. The optimization performed includes positioning, minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions.
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.
Mechanical modelling of transient- to- failure SFR fuel cladding
Energy Technology Data Exchange (ETDEWEB)
Feria, F.; Herranz, L. E.
2014-07-01
The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)
Perturbation analysis of transient population dynamics using matrix projection models
DEFF Research Database (Denmark)
Stott, Iain
2016-01-01
Non-stable populations exhibit short-term transient dynamics: size, growth and structure that are unlike predicted long-term asymptotic stable, stationary or equilibrium dynamics. Understanding transient dynamics of non-stable populations is important for designing effective population management...... these methods to know exactly what is being measured. Despite a wealth of existing methods, I identify some areas that would benefit from further development....
A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers
Energy Technology Data Exchange (ETDEWEB)
Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1997-12-31
A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)
Application of an estimation model to predict future transients at US nuclear power plants
International Nuclear Information System (INIS)
Hallbert, B.P.; Blackman, H.S.
1987-01-01
A model developed by R.A. Fisher was applied to a set of Licensee Event Reports (LERs) summarizing transient initiating events at US commercial nuclear power plants. The empirical Bayes model was examined to study the feasibility of estimating the number of categories of transients which have not yet occurred at nuclear power plants. An examination of the model's predictive ability using an existing sample of data provided support for use of the model to estimate future transients. The estimate indicates that an approximate fifteen percent increase in the number of categories of transient initiating events may be expected during the period 1983--1993, assuming a stable process of transients. Limitations of the model and other possible applications are discussed. 10 refs., 1 fig., 3 tabs
A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers
Energy Technology Data Exchange (ETDEWEB)
Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1998-12-31
A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)
Application of a qualified RETRAN model to plant transient evaluation support
International Nuclear Information System (INIS)
Sedano, P.G.; Mata, P.; Alcantud, F.; Serra, J.; Castrillo, F.
1989-01-01
This paper presents the applicability and usefulness of a complete and well qualified plant transient code and model to support in depth evaluation of anomalous plant transients. Analyses of several operational and abnormal transients that ocurred during the first three cycles of Cofrentes (BWR-6) NPP are presented. This application demonstrated the need of a very detailed and adjusted simulation of the control systems as well as the convenience of having as complete as possible data adquisition system. (orig.)
Application of a qualified RETRAN model to plant transient evaluation support
International Nuclear Information System (INIS)
Sedano, P.G.; Mata, P.; Alcantud, F.; Serra, J.
1989-01-01
This paper presents the applicability and usefulness of a complete and well qualified plant transient code and model to support in depth evaluation of anomalous plant transients. Analyses of several operational and abnormal transients occurred during the first three cycles of Cofrentes (BWR-6) NPP are presented. This application remarked the need of a very detailed and adjusted simulation of the control systems as well as the convenience of having an as complete as possible data acquisition system
Convergence analysis of neutronic/thermohydraulic coupling behavior of SCWR
International Nuclear Information System (INIS)
Liu, Shichang; Cai, Jiejin
2013-01-01
The neutronic/thermohydraulic coupling (N–T coupling) calculations play an important role in core design and stability analysis. The traditional iterative method is not applicable for some new reactors (such as supercritical water-cooled reactor) which have intense N–T coupling behavior. In this paper, the mathematical model of N–T coupling based on fixed point theory is established firstly, with the convergent criterion, which can show the real-time convergence situation of iteration. Secondly, the self-adaptive relaxation factor and corresponding algorithm are proposed. Thirdly, the convergence analysis of the method of self-adaptive relaxation factor and common relaxation iteration has been performed, based on three calculation examples of SCWR fuel assembly. The results show that the proposed algorithm can efficiently reduce the calculation time and be adapted to different coupling cases and different initial distribution. It is easy to program, providing convenience for reactor design and analysis. This research also provides the theoretical basis for further study of N–T coupling behavior of new reactors such as SCWR
Thermo-hydraulic design of earth-air heat exchangers
Energy Technology Data Exchange (ETDEWEB)
Paepe, M. de [Ghent University (Belgium). Department of Flow, Heat and Combustion Mechanics; Janssens, A. [Ghent University (Belgium). Department of Architecture and Urbanism
2003-05-01
Earth-air heat exchangers, also called ground tube heat exchangers, are an interesting technique to reduce energy consumption in a building. They can cool or heat the ventilation air, using cold or heat accumulated in the soil. Several papers have been published in which a design method is described. Most of them are based on a discretisation of the one-dimensional heat transfer problem in the tube. Three-dimensional complex models, solving conduction and moisture transport in the soil are also found. These methods are of high complexity and often not ready for use by designers. In this paper, a one-dimensional analytical method is used to analyse the influence of the design parameters of the heat exchanger on the thermo-hydraulic performance. A relation is derived for the specific pressure drop, linking thermal effectiveness with pressure drop of the air inside the tube. The relation is used to formulate a design method which can be used to determine the characteristic dimensions of the earth-air heat exchanger in such a way that optimal thermal effectiveness is reached with acceptable pressure loss. The choice of the characteristic dimensions, becomes thus independent of the soil and climatological conditions. This allows designers to choose the earth-air heat exchanger configuration with the best performance. (author)
Thermo-hydraulic design of earth-air heat exchangers
Energy Technology Data Exchange (ETDEWEB)
De Paepe, M. [Department of Flow, Heat and Combustion Mechanics, Ghent University, Ghent (Belgium); Janssens, A. [Department of Architecture and Urbanism, Ghent University, Ghent (Belgium)
2003-07-01
Earth-air heat exchangers, also called ground tube heat exchangers, are an interesting technique to reduce energy consumption in a building. They can cool or heat the ventilation air, using cold or heat accumulated in the soil. Several papers have been published in which a design method is described. Most of them are based on a discretisation of the one-dimensional heat transfer problem in the tube. Three-dimensional complex models, solving conduction and moisture transport in the soil are also found. These methods are of high complexity and often not ready for use by designers. In this paper, a one-dimensional analytical method is used to analyse the influence of the design parameters of the heat exchanger on the thermo-hydraulic performance. A relation is derived for the specific pressure drop, linking thermal effectiveness with pressure drop of the air inside the tube. The relation is used to formulate a design method which can be used to determine the characteristic dimensions of the earth-air heat exchanger in such a way that optimal thermal effectiveness is reached with acceptable pressure loss. The choice of the characteristic dimensions, becomes thus independent of the soil and climatological conditions. This allows designers to choose the earth-air heat exchanger configuration with the best performance. (author)
Thermohydraulic tests of 3x3-rod bundle maquette
International Nuclear Information System (INIS)
Ladeira, L.C.D.
1986-10-01
The results of a 3x3-rod bundle thermohydraulic research program, performed in the Thermohydraulics Laboratory of NUCLEBRAS' Nuclear Technology Development Center, are briefly described. This program included measurements of pressure drops in one and two-phase flows, heat transfer coefficients, mixing between interconnected subchannels in one-phase flow conditions and critical heat fluxes. The measurements covered the following parameter ranges: heat fluxes from zero to the critical values, pressure ranging from 1 to 15 ata, inlet temperature from 25 to 150 sup(0)C and flow rate from 20 to 300l/min. (author)
Observation and control system of the thermohydraulic assays laboratory
International Nuclear Information System (INIS)
Santome, D.; Hualde, R.
1990-01-01
The Thermohydraulic Assays Laboratory (L.E.T.) is an installation whose purpose will be the components testing and the CAREM-25 reactor thermohydraulic processes operation dynamics. This plant is located at Pilcaniyeu, province of Rio Negro. Part of the tests which will be carried out consist in the use of different control strategies. The control of the systems by digital processors (control by software) has been decided to proceed with a maximum flexibility and capacity to make changes in the algorithms. This work describes the design and implementation of a digital control system to command the three circuits of the installation. (Author) [es
Utilization of Relap 5 computer code for analyzing thermohydraulic projects
International Nuclear Information System (INIS)
Silva Filho, E.
1987-01-01
This work deals with the design of a scaled test facility of a typical pressurized water reactor plant of the 1300 MW (electric) class. A station blackout has been choosen to investigate the thermohydraulic behaviour of the the test facility in comparison to the reactor plant. The computer code RELAPS/MOD1 has been utilized to simulate the blackout and to compare the test facility behaviour with the reactor plant one. The results demonstrate similar thermohydraulic behaviours of the two systems. (author) [pt
Energy Technology Data Exchange (ETDEWEB)
Lazaro, A.; Ammirabile, L.; Martorell, S.
2014-07-01
The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)
Impact of Model Detail of Synchronous Machines on Real-time Transient Stability Assessment
DEFF Research Database (Denmark)
Weckesser, Johannes Tilman Gabriel; Jóhannsson, Hjörtur; Østergaard, Jacob
2013-01-01
In this paper, it is investigated how detailed the model of a synchronous machine needs to be in order to assess transient stability using a Single Machine Equivalent (SIME). The results will show how the stability mechanism and the stability assessment are affected by the model detail. In order...... of the machine models is varied. Analyses of the results suggest that a 4th-order model may be sufficient to represent synchronous machines in transient stability studies....
RAP-2A Computer code for transients analysis in fast reactors
International Nuclear Information System (INIS)
Iftode, I.; Popescu, C.; Turcu, I.; Biro, L.
1975-10-01
The RAP-2A computer code is designed for analyzing thermohydraulic transients and/or steady state problems for large LMFBR cores. Physical and mathematical models, main input-output data, the flow chart of the code and a sample problem are given. RAP-2A calculates the power and the thermoydraulic transients initiated by a flow or reactivity changes, from a normal operating state of the reactor up to core disassembly. In this analysis a representative fuel pin is considered: a one-group space-independent (point) kinetics model to describe the neutron kinetics and a one-dimensional model describing the heat transfer (radial in the fuel and axial in the coolant) are used. Mechanical deformations due to temperature gradient, pressure losses, fuel melting, etc., are also calculated. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer
The problems of thermohydraulics of prospective fast reactor concepts
International Nuclear Information System (INIS)
Sedov, A.A.
2000-01-01
In this report the main requirements to fast reactors in system of future multicomponent Nuclear Power with closed U-Pu fuel cycle are regarded. The peculiarities of different liquid-metal (sodium and lead-alloyed) coolants as well as the thermohydraulics problems of prospective fast reactors (FR) concepts are discussed. (author)
Systems for neutronic, thermohydraulic and shielding calculation in personal computers
International Nuclear Information System (INIS)
Villarino, E.A.; Abbate, P.; Lovotti, O.; Santini, M.
1990-01-01
The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author) [es
Villar, V. Ashley; Berger, Edo; Metzger, Brian D.; Guillochon, James
2017-11-01
The duration-luminosity phase space (DLPS) of optical transients is used, mostly heuristically, to compare various classes of transient events, to explore the origin of new transients, and to influence optical survey observing strategies. For example, several observational searches have been guided by intriguing voids and gaps in this phase space. However, we should ask, do we expect to find transients in these voids given our understanding of the various heating sources operating in astrophysical transients? In this work, we explore a broad range of theoretical models and empirical relations to generate optical light curves and to populate the DLPS. We explore transients powered by adiabatic expansion, radioactive decay, magnetar spin-down, and circumstellar interaction. For each heating source, we provide a concise summary of the basic physical processes, a physically motivated choice of model parameter ranges, an overall summary of the resulting light curves and their occupied range in the DLPS, and how the various model input parameters affect the light curves. We specifically explore the key voids discussed in the literature: the intermediate-luminosity gap between classical novae and supernovae, and short-duration transients (≲ 10 days). We find that few physical models lead to transients that occupy these voids. Moreover, we find that only relativistic expansion can produce fast and luminous transients, while for all other heating sources events with durations ≲ 10 days are dim ({M}{{R}}≳ -15 mag). Finally, we explore the detection potential of optical surveys (e.g., Large Synoptic Survey Telescope) in the DLPS and quantify the notion that short-duration and dim transients are exponentially more difficult to discover in untargeted surveys.
International Nuclear Information System (INIS)
Muramatu, Toshiharu; Yamaguchi, Akira
2004-01-01
A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. A numerical analysis was carried out with a multi-dimensional code AQUA to confirm an applicability to the evaluations for the in-vessel thermohydraulic phenomena using a 1/10 scaled water experiment simulating the large-scale fast breeder reactor in the feasibility studies. From the analysis, the following results were obtained. (1) In-vessel thermohydraulics characterized by a radiated flow pattern to the reactor vessel wall and a strong upward flow through a slit of the upper core structures were evaluated. These characteristics agreed approximately with the water experiment. (2) The upward velocity values at the slit agreed well with the experimental data under a condition of γ z = 0.3 and ξ z = 0.5, though overall evaluations of the in-vessel thermohydraulics were failed to predict quantitatively. (3) The AQUA code is applicable to the in-vessel thermohydraulics evaluations in the feasibility studies, though it is necessary to make further modifications of the calculational models for accurate evaluations. On the one hand, it was confirmed that calculated results for the 1/10 water experimental model and the 1/1 actual-scaled model agreed quantitatively for the in-vessel thermohydraulics characteristics indicated above. (author)
Pressure Transient Model of Water-Hydraulic Pipelines with Cavitation
Directory of Open Access Journals (Sweden)
Dan Jiang
2018-03-01
Full Text Available Transient pressure investigation of water-hydraulic pipelines is a challenge in the fluid transmission field, since the flow continuity equation and momentum equation are partial differential, and the vaporous cavitation has high dynamics; the frictional force caused by fluid viscosity is especially uncertain. In this study, due to the different transient pressure dynamics in upstream and downstream pipelines, the finite difference method (FDM is adopted to handle pressure transients with and without cavitation, as well as steady friction and frequency-dependent unsteady friction. Different from the traditional method of characteristics (MOC, the FDM is advantageous in terms of the simple and convenient computation. Furthermore, the mechanism of cavitation growth and collapse are captured both upstream and downstream of the water-hydraulic pipeline, i.e., the cavitation start time, the end time, the duration, the maximum volume, and the corresponding time points. By referring to the experimental results of two previous works, the comparative simulation results of two computation methods are verified in experimental water-hydraulic pipelines, which indicates that the finite difference method shows better data consistency than the MOC.
International Nuclear Information System (INIS)
2012-04-01
Most advanced nuclear power plant designs adopted several kinds of passive systems. Natural circulation is used as a key driving force for many passive systems and even for core heat removal during normal operation such as NuScale, CAREM, ESBWR and Indian AHWR designs. Simulation of natural circulation phenomena is very challenging since the driving force of it is weak compared to forced circulation and involves a coupling between primary system and containment for integral type reactor. The IAEA ICSP (International Collaborative Standard Problem) on 'Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents' was proposed within the CRP on 'Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems that utilize Natural Circulation'. Oregon State University (OSU) of USA offered to host this ICSP. This ICSP plans to conduct the following experiments and blind/open simulations with system codes: 1. Quasi-steady state operation with different core power levels: Conduct quasi-steady state operation with step-wise increase of core power level in order to observe single phase natural circulation flow according to power level. The experimental facility and operating conditions for an integral PWR will be used. 2. Thermo-hydraulic Coupling between Primary system and Containment: Conduct a loss of feedwater transient with subsequent ADS blowdown and long term cooling to determine the progression of a loss of feedwater transient by natural circulation through primary and containment systems. These tests would examine the blowdown phase as well as the long term cooling using sump natural circulation by coupling the primary to containment systems. This data could be used for the evaluation of system codes to determine if they model specific phenomena in an accurate manner. OSU completed planned two ICSP tests in July 2011 and real initial and boundary conditions measured from the
Energy Technology Data Exchange (ETDEWEB)
Massoud, M
1987-01-01
Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.
International Nuclear Information System (INIS)
Massoud, M.
1987-01-01
Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients
Investigation of transient models and performances for a doubly fed wind turbine under a grid fault
DEFF Research Database (Denmark)
Wang, M.; Zhao, B.; Li, H.
2011-01-01
fed induction generator (DFIG), the assessments of the impact on the electrical transient performances were investigated for the doubly fed wind turbine with different representations of wind turbine drive-train dynamics models, different initial operational conditions and different active crowbar...... crowbar on the transient performances of the doubly fed wind turbine were also investigated, with the possible reasonable trip time of crowbar. The investigation have shown that the transient performances are closely correlated with the wind turbine drive train models, initial operational conditions, key...
Wang, Weicheng; Natelson, Robert H.; Stikeleather, Larry F.; Roberts, William L.
2013-01-01
A chemical kinetic model has been developed for the transient stage of the continuous countercurrent hydrolysis of triglycerides to free fatty acids and glycerol. Departure functions and group contribution methods were applied to determine
Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors
Energy Technology Data Exchange (ETDEWEB)
Kalcheva, S.; Koonen, E. [SCK-CEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium); Olson, A. P. [RERTR Program, Nuclear Engineering Div., Argonne National Laboratory, Cass Avenue, Argonne, IL 60439 (United States)
2012-07-01
Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)
Linear and quadrature models for data from treshold measurements of the transient visual system
Brinker, den A.C.
1986-01-01
III this paper two models are considered for the transient visual system at threshold. One is a linear model and the other a model contain ing a quadrature element. Both models are commonly used on evidence from different experimental sources. It is shown that both models act in a similar fashion
PUMP: analog-hybrid reactor coolant hydraulic transient model
International Nuclear Information System (INIS)
Grandia, M.R.
1976-03-01
The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated
International Nuclear Information System (INIS)
Kroshilin, A.E.; Zhukavin, A.P.; Pryakhin, V.N.
1992-01-01
The mathematical model realized in the full-scale educational facility for NPP operator training is described. The RETACT computational complex providing real time process simulation for all regimes including the maximum credible accident is used for calculation of thermohydraulic parameters of the primary coolant circuits and steam generator under stationary and transient conditions. The two-velocity two-temperature model of one-dimensional steam-water flow containing uncondensed gases is realized in the program
Evaluation of core modeling effect on transients for multi-flow zone design of SFR
International Nuclear Information System (INIS)
Shin, Andong; Choi, Yong Won
2016-01-01
SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)
Evaluation of core modeling effect on transients for multi-flow zone design of SFR
Energy Technology Data Exchange (ETDEWEB)
Shin, Andong; Choi, Yong Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2016-10-15
SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)
DEFF Research Database (Denmark)
Gong, M.; Zhang, Y.; Weschler, Charles J.
2014-01-01
A transient model is developed to predict dermal absorption of gas-phase chemicals via direct air-to-skin-to-blood transport under non-steady-state conditions. It differs from published models in that it considers convective mass-transfer resistance in the boundary layer of air adjacent to the skin....... Results calculated with this transient model are in good agreement with the limited experimental results that are available for comparison. The sensitivity of the modeled estimates to key parameters is examined. The model is then used to estimate air-to-skin-to-blood absorption of six phthalate esters...... and less absorbed into blood than would a steady-state model. In the 7-day scenario, results calculated by the transient and steady-state models converge over a time period that varies between 3 and 4days for all but the largest phthalate (DEHP). Dermal intake is comparable to or larger than inhalation...
Directory of Open Access Journals (Sweden)
Yanli Xin
2016-12-01
Full Text Available This paper presents a comprehensive investigation on high frequency (HF switching transients due to energization of vacuum circuit breakers (VCBs in offshore wind farms (OWFs. This research not only concerns the modeling of main components in collector grids of an OWF for transient analysis (including VCBs, wind turbine transformers (WTTs, submarine cables, but also compares the effectiveness between several mainstream switching overvoltage (SOV protection methods and a new mitigation method called smart choke. In order to accurately reproduce such HF switching transients considering the current chopping, dielectric strength (DS recovery capability and HF quenching capability of VCBs, three models are developed, i.e., a user–defined VCB model, a HF transformer terminal model and a three-core (TC frequency dependent model of submarine cables, which are validated through simulations and compared with measurements. Based on the above models and a real OWF configuration, a simulation model is built and several typical switching transient cases are investigated to analyze the switching transient process and phenomena. Subsequently, according to the characteristics of overvoltages, appropriate parameters of SOV mitigation methods are determined to improve their effectiveness. Simulation results indicate that the user–defined VCB model can satisfactorily simulate prestrikes and the proposed component models display HF characteristics, which are consistent with onsite measurement behaviors. Moreover, the employed protection methods can suppress induced SOVs, which have a steep front, a high oscillation frequency and a high amplitude, among which the smart choke presents a preferable HF damping effect.
Application of a film flow model to predicting burnout under transient conditions
International Nuclear Information System (INIS)
Leslie, D.C.; Kirby, G.J.
1967-08-01
The film flow model developed previously has been generalised to transient situations by assuming that only convection is changed by the transient; evaporation, deposition and entrainment are assumed to be unaffected. A computer code TRABUT computes the time behaviour of the mass velocity and the quality by the method of characteristics, and then integrates the film flow equations along the same characteristics until the point of burn-out or zero film flow is reached. The time delay between the onset of a transient and burn-out has been computed both for flux and flow transients. These computations have been compared with those made using the standard local conditions hypothesis. The film flow model gives shorter delays in almost all cases, but the difference would not be detectable with present experimental techniques. (author)
New transient-flow modelling of a multiple-fractured horizontal well
International Nuclear Information System (INIS)
Jia, Yong-Lu; Wang, Ben-Cheng; Nie, Ren-Shi; Wang, Dan-Ling
2014-01-01
A new transient-flow modelling of a multiple-fractured horizontal well is presented. Compared to conventional modelling, the new modelling considered more practical physical conditions, such as various inclined angles for different fractures, different fracture intervals, different fracture lengths and partially penetrating fractures to formation. A kind of new mathematical method, including a three-dimensional eigenvalue and orthogonal transform, was created to deduce the exact analytical solutions of pressure transients for constant-rate production in real space. In order to consider a wellbore storage coefficient and skin factor, we used a Laplace-transform approach to convert the exact analytical solutions to the solutions in Laplace space. Then the numerical solutions of pressure transients in real space were gained using a Stehfest numerical inversion. Standard type curves were plotted to describe the transient-flow characteristics. Flow regimes were clearly identified from type curves. Furthermore, the differences between the new modelling and the conventional modelling in pressure transients were especially compared and discussed. Finally, an example application to show the accordance of the new modelling with real conditions was implemented. Our new modelling is different from, but more practical than, conventional modelling. (paper)
International Nuclear Information System (INIS)
Asad, Usman; Tjong, Jimi; Zheng, Ming
2014-01-01
Highlights: • Zero-dimensional EGR model for transient diesel combustion control. • Detailed analysis of EGR effects on intake, cylinder charge and exhaust properties. • Intake oxygen validated as an operating condition-independent measure of EGR. • Quantified EGR effectiveness in terms of NOx emission reduction. • Twin lambda sensor technique for estimation of EGR/in-cylinder parameters. - Abstract: The application of exhaust gas recirculation (EGR) during transient engine operation is a challenging task since small fluctuations in EGR may cause larger than acceptable spikes in NOx/soot emissions or deterioration in the combustion efficiency. Moreover, the intake charge dilution at any EGR ratio is a function of engine load and intake pressure, and typically changes during transient events. Therefore, the management of EGR during transient engine operation or advanced combustion cycles (that are inherently less stable) requires a fundamental understanding of the transient EGR behaviour and its impact on the intake charge development. In this work, a zero-dimensional EGR model is described to estimate the transient (cycle-by-cycle) progression of EGR and the time (engine cycles) required for its stabilization. The model response is tuned to a multi-cylinder engine by using an overall engine system time-constant and shown to effectively track the transient EGR changes. The impact of EGR on the actual air–fuel ratio of the cylinder charge is quantified by defining an in-cylinder excess-air ratio that accounts for the oxygen in the recycled exhaust gas. Furthermore, a twin lambda sensor (TLS) technique is implemented for tracking the intake dilution and in-cylinder excess-air ratio in real-time. The modelling and analysis results are validated against a wide range of engine operations, including transient and steady-state low temperature combustion tests
Simplified drive system models for power system transient studies in industrial plants
DEFF Research Database (Denmark)
Chen, Peiyuan; Sannino, Ambra
2007-01-01
In order to simulate industrial plants for different power system transient studies, simplified adjustable speed drive (ASD) models are needed. For power system transient studies such as assessing the voltage dip ride-through capability of ASDs, detailed representation of semiconductor valve...... switching can be avoided, thereby making possible to increase the time step of the simulation. In this paper, simplified ASD models are developed and compared with corresponding detailed models. The performance of the simplified models is assessed when increasing the simulation step as much as possible...
Transient Model Validation of Fixed-Speed Induction Generator Using Wind Farm Measurements
DEFF Research Database (Denmark)
Rogdakis, Georgios; Garcia-Valle, Rodrigo; Arana Aristi, Iván
2012-01-01
In this paper, an electromagnetic transient model for fixed-speed wind turbines equipped with induction generators is developed and implemented in PSCAD/EMTDC. The model is comprised by: an induction generator, aerodynamic rotor, and a two-mass representation of the shaft system. Model validation...
An integrated transient model for simulating the operation of natural gas transport systems
Pambour, Kwabena Addo; Bolado-Lavin, Ricardo; Dijkema, Gerard P. J.
This paper presents an integrated transient hydraulic model that describes the dynamic behavior of natural gas transport systems (GTS). The model includes sub models of the most important facilities comprising a GTS, such as pipelines, compressor stations, pressure reduction stations, underground
Models for transient analyses in advanced test reactors
International Nuclear Information System (INIS)
Gabrielli, Fabrizio
2011-01-01
Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are applied
Modification of the bubble rise model used in RELAP4/Mod5 computer code for transients analysis
International Nuclear Information System (INIS)
Scharfmann, E.
1981-01-01
To improve the separation phase and heat transfer models in RELAP4/MOD5 computer code, in order to make more realistic estimates of the thermohydraulic behavior of the core submitted to a loss-of-coolant accident, is the objective of this work. This research is directed to the accident analysis caused by small breaks in the primary circuit of PWR plants, where two-phase flow occurs most of the time. Calculation have been performed with the help of the original version of RELAP code, and the version containing the proposed modifications on this work. Comparing one results with the original ones, we arrive at the conclusion that our results show more conservative values of core pressure and coolant temperature, while the peak values of fuel temperature are not exceeded. (Author) [pt
Thermohydraulic verification during THTR steam generator commissioning
International Nuclear Information System (INIS)
Henry, C.; Elter, C.
1988-01-01
In one of the six THTR 300 steam generators thermocouples are installed inside the heat transfer tube bundles for measuring the gas and steam temperatures. Fluid temperature distribution measurements along and across the helix bundle have been recorded in its first months of operation over a load range of 40% up to 100% for steady state and transient conditions. Using these measurements as well as the rest of the operating instrumentation. the computer programs for the design of heat exchanger heat transfer areas are verified. The temperature measurements for steady state conditions are compared with predictions obtained in the design stage. In these codes. the heat transferred from the outside helium gas to the water/steam inside the tubes is determined in discrete steps along the heating surface by one- and two-phase heat transfer correlations. The degree of conformity between prediction and measurement is discussed and compared with more recent correlations. (author)
Numerical Modeling of Transient Heat Transfer in Longitudinal Fin
Directory of Open Access Journals (Sweden)
Farshad Panahizadeh
2017-11-01
Full Text Available The main objective of the present numerical study is to investigate the transient heat transfer in one kind of all-purpose longitudinal fin with the triangular profile. The lateral surface of the concerned fin and the tip of it are subjected to general situations included heat flux at the base and insulation on the tip. For this study developed a one dimensional in house code written by Fortran 90 programming language by using finite difference method with an implicit scheme in unsteady state condition. Generally, the result of this study in time variation state after 700 seconds is steady. The results also show the fin efficiency by increasing the time of study decreases due to a reduction in the total heat transfer which is happened in the fin. The grid independence study shows that for the number of nodes greater than 20 the result will not be changed and same as before. Finally, the result of Fortran code verified by commercial CFD code which relies on finite difference method and it was shown have a consistent agreement
On three-dimensional nuclear thermo-hydraulic computation techniques for ATR
International Nuclear Information System (INIS)
1997-08-01
The three-dimensional computation code for nuclear thermo-hydraulic combination core LAYMON-2A is used for the calculation of the power distribution and the control rod reactivity value of the ATR. This code possesses various functions which are required for planning the core operation such as the search function for critical boric acid concentration, and can do various simulation calculations such as core burning calculation. Further, the three-dimensional analysis code for xenon dynamic characteristics in the core LAYMON-2C, in which the dynamic characteristic equation of xenon-samarium was incorporated into the LAYMON-2A code can take the change with time lapse of xenon-samarium concentration accompanying the change of power level and power distribution into account, and it is used for the analysis of the spatial vibration characteristics of power and the regional power control characteristics due to xenon in the core. As to the LAYMON-2A, the computation flow, power distribution and thermo-hydraulic computation models, and critical search function are explained. As to the LAYMON-2C, the computation flow is described. The comparison of the calculated values by using the LAYMON-2A code and the operation data of the Fugen is reported. (K.I.)
BWR thermohydraulics simulation on the AD-10 peripheral processor
International Nuclear Information System (INIS)
Wulff, W.; Cheng, H.S.; Lekach, S.V.; Mallen, A.N.
1983-01-01
This presentation demonstrates the feasibility of simulating plant transients and severe abnormal transients in nuclear power plants at much faster than real-time computing speeds in a low-cost, dedicated, interactive minicomputer. This is achieved by implementing advanced modeling techniques in modern, special-purpose peripheral processors for high-speed system simulation. The results of this demonstration will impact safety analyses and parametric studies, studies on operator responses and control system failures and it will make possible the continuous on-line monitoring of plant performance and the detection and diagnosis of system or component failures
International Nuclear Information System (INIS)
An Chen; Su Jian
2011-01-01
Improved lumped parameter models were developed for the transient heat conduction in multi-layer composite slabs subjected to combined convective and radiative cooling. The improved lumped models were obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of three-layer composite slabs was analyzed to illustrate the applicability of the proposed lumped models, with respect to different values of the Biot numbers, the radiation-conduction parameter, the dimensionless thermal contact resistances, the dimensionless thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the higher order lumped model (H 1,1 /H 0,0 approximation) yielded significant improvement of average temperature prediction over the classical lumped model. In addition, the higher order (H 1,1 /H 0,0 ) model was applied to analyze the transient heat conduction problem of steel-concrete-steel sandwich plates. - Highlights: → Improved lumped models for convective-radiative cooling of multi-layer slabs were developed. → Two-point Hermite approximations for integrals were employed. → Significant improvement over classical lumped model was achieved. → The model can be applied to high Biot number and high radiation-conduction parameter. → Transient heat conduction in steel-concrete-steel sandwich pipes was analyzed as an example.
A simulation model for transient response of a gas separation module using a hollow fiber membrane
Energy Technology Data Exchange (ETDEWEB)
Sugiyama, Takahiko, E-mail: t-sugiyama@nucl.nagoya-u.ac.jp [Nagoya University, Fro-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Miyahara, Naoya [Nagoya University, Fro-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Tanaka, Masahiro [National Institute for Fusion Science, Oroshi-cho 322-6, Toki 509-5292 (Japan); Munakata, Kenzo [Akita University, Tegata Gakuen-cho 1-1, Akita-shi, Akita 010-8502 (Japan); Yamamoto, Ichiro [Nagoya University, Fro-cho, Chikusa-ku, Nagoya 464-8603 (Japan)
2011-10-15
A simulation model has been developed for transient response of a gas separation module using a hollow fiber membrane for the removal of tritium from the atmosphere of the confinement space. The mass transfer process such as sorption and desorption of gases at the surface of the dense layer and the porous support layer, diffusive transfer in the both layers are treated in the model. Sorption isotherm, mass transfer rate and permeance are estimated through step-wise transient response experiments. The present model represents well not only separation factors and recovery ratio at the steady state but also responses to the multi-step wise change in the sweep gas rate.
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)
Deshpande, K.; Zettergren, M. D.; Datta-Barua, S.
2017-12-01
Fluctuations in the Global Navigation Satellite Systems (GNSS) signals observed as amplitude and phase scintillations are produced by plasma density structures in the ionosphere. Phase scintillation events in particular occur due to structures at Fresnel scales, typically about 250 meters at ionospheric heights and GNSS frequency. Likely processes contributing to small-scale density structuring in auroral and polar regions include ionospheric gradient-drift instability (GDI) and Kelvin-Helmholtz instability (KHI), which result, generally, from magnetosphere-ionosphere interactions (e.g. reconnection) associated with cusp and auroral zone regions. Scintillation signals, ostensibly from either GDI or KHI, are frequently observed in the high latitude ionosphere and are potentially useful diagnostics of how energy from the transient forcing in the cusp or polar cap region cascades, via instabilities, to small scales. However, extracting quantitative details of instabilities leading to scintillation using GNSS data drastically benefits from both a model of the irregularities and a model of GNSS signal propagation through irregular media. This work uses a physics-based model of the generation of plasma density irregularities (GEMINI - Geospace Environment Model of Ion-Neutral Interactions) coupled to an ionospheric radio wave propagation model (SIGMA - Satellite-beacon Ionospheric-scintillation Global Model of the upper Atmosphere) to explore the cascade of density structures from medium to small (sub-kilometer) scales. Specifically, GEMINI-SIGMA is used to simulate expected scintillation from different instabilities during various stages of evolution to determine features of the scintillation that may be useful to studying ionospheric density structures. Furthermore we relate the instabilities producing GNSS scintillations to the transient space and time-dependent magnetospheric phenomena and further predict characteristics of scintillation in different geophysical
Steady-state and transient fission gas release and swelling model for LIFE-4
International Nuclear Information System (INIS)
Villalobos, A.; Liu, Y.Y.; Rest, J.
1984-06-01
The fuel-pin modeling code LIFE-4 and the mechanistic fission gas behavior model FASTGRASS have been coupled and verified against gas release data from mixed-oxide fuels which were transient tested in the TREAT reactor. Design of the interface between LIFE-4 and FASTGRASS is based on an earlier coupling between an LWR version of LIFE and the GRASS-SST code. Fission gas behavior can significantly affect steady-state and transient fuel performance. FASTGRASS treats fission gas release and swelling in an internally consistent manner and simultaneously includes all major mechanisms thought to influence fission gas behavior. The FASTGRASS steady-state and transient analysis has evolved through comparisons of code predictions with fission-gas release and swelling data from both in- and ex-reactor experiments. FASTGRASS was chosen over other fission-gas behavior models because of its availability, its compatibility with the LIFE-4 calculational framework, and its predictive capability
Modeling the release of E. coli D21g with transients in water content
Transients in water content are well known to mobilize colloids that are retained in the vadose zone. However, there is no consensus on the proper model formulation to simulate colloid release during drainage and imbibition. We present a model that relates colloid release to changes in the air-water...
Transient combustion modeling of an oscillating lean premixed methane/air flam
Withag, J.A.M.; Kok, Jacobus B.W.; Syed, Khawar
2009-01-01
The main objective of the present study is to demonstrate accurate low frequency transient turbulent combustion modeling. For accurate flame dynamics some improvements were made to the standard TFC combustion model for lean premixed combustion. With use of a 1D laminar flamelet code, predictions
Equilibrium and kinetic models for colloid release under transient solution chemistry conditions
We present continuum models to describe colloid release in the subsurface during transient physicochemical conditions. Our modeling approach relates the amount of colloid release to changes in the fraction of the solid surface area that contributes to retention. Equilibrium, kinetic, equilibrium and...
Modeling steady state and transient fission gas behaviour with the Karlsruhe code LAKU
International Nuclear Information System (INIS)
Vaeth, L.
1984-08-01
The programme LAKU models the behaviour of gaseous fission products in reactor fuel under steady state and transient conditions, including molten fuel. A presentation of the full model is given, starting with gas behaviour in the grains and on grain faces and including the treatment of release from porosity. The results of some recent calculations are presented. (orig.) [de
Mathematical modeling of an industrial firm in transient economy
Directory of Open Access Journals (Sweden)
I. G. Pospelov
2001-01-01
Full Text Available A behavior model of an industrial firm with a possibility of merchandising and purchasing of production on two channels, traditional and commercial, is given. The former is stable, but less profitable due to non-payments. The latter is profitable, but risky. The model describes different modes of firm operation depending on economic parameters. In such a model, firms have incentives to integrate in financial and industrial groups.
Transient dynamic and modeling parameter sensitivity analysis of 1D solid oxide fuel cell model
International Nuclear Information System (INIS)
Huangfu, Yigeng; Gao, Fei; Abbas-Turki, Abdeljalil; Bouquain, David; Miraoui, Abdellatif
2013-01-01
Highlights: • A multiphysics, 1D, dynamic SOFC model is developed. • The presented model is validated experimentally in eight different operating conditions. • Electrochemical and thermal dynamic transient time expressions are given in explicit forms. • Parameter sensitivity is discussed for different semi-empirical parameters in the model. - Abstract: In this paper, a multiphysics solid oxide fuel cell (SOFC) dynamic model is developed by using a one dimensional (1D) modeling approach. The dynamic effects of double layer capacitance on the electrochemical domain and the dynamic effect of thermal capacity on thermal domain are thoroughly considered. The 1D approach allows the model to predict the non-uniform distributions of current density, gas pressure and temperature in SOFC during its operation. The developed model has been experimentally validated, under different conditions of temperature and gas pressure. Based on the proposed model, the explicit time constant expressions for different dynamic phenomena in SOFC have been given and discussed in detail. A parameters sensitivity study has also been performed and discussed by using statistical Multi Parameter Sensitivity Analysis (MPSA) method, in order to investigate the impact of parameters on the modeling accuracy
A simplified model for tritium permeation transient predictions when trapping is active
International Nuclear Information System (INIS)
Longhurst, G.R.
1994-01-01
This report describes a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. Comparison calculations with the verified and validated TMAP4 transient code show good agreement. ((orig.))
A simplified model for tritium permeation transient predictions when trapping is active*1
Longhurst, G. R.
1994-09-01
This report describes a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. Comparison calculations with the verified and validated TMAP4 transient code show good agreement.
A simplified model for tritium permeation transient predictions when trapping is active
Energy Technology Data Exchange (ETDEWEB)
Longhurst, G.R. (Fusion Safety Program, Idaho National Engineering Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States))
1994-09-01
This report describes a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. Comparison calculations with the verified and validated TMAP4 transient code show good agreement. ((orig.))
International Nuclear Information System (INIS)
Chang Tiejun; Tian Mingzhen
2007-01-01
A previously developed numerical model based on Maxwell-Bloch equations was modified to simulate optical coherent transient and spectral hole burning processes with noisy laser sources. Random walk phase noise was simulated using laser-phase sequences generated numerically according to the normal distribution of the phase shift. The noise model was tested by comparing the simulated spectral hole burning effect with the analytical solution. The noise effects on a few typical optical coherence transient processes were investigated using this numerical tool. Flicker and random walk frequency noises were considered in accumulation process
Transient electro-thermal modeling of bipolar power semiconductor devices
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
Predictive modeling of transient storage and nutrient uptake: Implications for stream restoration
O'Connor, Ben L.; Hondzo, Miki; Harvey, Judson W.
2010-01-01
This study examined two key aspects of reactive transport modeling for stream restoration purposes: the accuracy of the nutrient spiraling and transient storage models for quantifying reach-scale nutrient uptake, and the ability to quantify transport parameters using measurements and scaling techniques in order to improve upon traditional conservative tracer fitting methods. Nitrate (NO3–) uptake rates inferred using the nutrient spiraling model underestimated the total NO3– mass loss by 82%, which was attributed to the exclusion of dispersion and transient storage. The transient storage model was more accurate with respect to the NO3– mass loss (±20%) and also demonstrated that uptake in the main channel was more significant than in storage zones. Conservative tracer fitting was unable to produce transport parameter estimates for a riffle-pool transition of the study reach, while forward modeling of solute transport using measured/scaled transport parameters matched conservative tracer breakthrough curves for all reaches. Additionally, solute exchange between the main channel and embayment surface storage zones was quantified using first-order theory. These results demonstrate that it is vital to account for transient storage in quantifying nutrient uptake, and the continued development of measurement/scaling techniques is needed for reactive transport modeling of streams with complex hydraulic and geomorphic conditions.
Predictive Modeling of Transient Storage and Nutrient Uptake: Implications for Stream Restoration
Energy Technology Data Exchange (ETDEWEB)
O’Connor, Ben L.; Hondzo, Miki; Harvey, Judson W.
2010-12-01
This study examined two key aspects of reactive transport modeling for stream restoration purposes: the accuracy of the nutrient spiraling and transient storage models for quantifying reach-scale nutrient uptake, and the ability to quantify transport parameters using measurements and scaling techniques in order to improve upon traditional conservative tracer fitting methods. Nitrate (NO-3)(NO3-) uptake rates inferred using the nutrient spiraling model underestimated the total NO-3NO3- mass loss by 82%, which was attributed to the exclusion of dispersion and transient storage. The transient storage model was more accurate with respect to the NO-3NO3- mass loss (±20%) and also demonstrated that uptake in the main channel was more significant than in storage zones. Conservative tracer fitting was unable to produce transport parameter estimates for a riffle-pool transition of the study reach, while forward modeling of solute transport using measured/scaled transport parameters matched conservative tracer breakthrough curves for all reaches. Additionally, solute exchange between the main channel and embayment surface storage zones was quantified using first-order theory. These results demonstrate that it is vital to account for transient storage in quantifying nutrient uptake, and the continued development of measurement/scaling techniques is needed for reactive transport modeling of streams with complex hydraulic and geomorphic conditions.
Unified fluid flow model for pressure transient analysis in naturally fractured media
International Nuclear Information System (INIS)
Babak, Petro; Azaiez, Jalel
2015-01-01
Naturally fractured reservoirs present special challenges for flow modeling with regards to their internal geometrical structure. The shape and distribution of matrix porous blocks and the geometry of fractures play key roles in the formulation of transient interporosity flow models. Although these models have been formulated for several typical geometries of the fracture networks, they appeared to be very dissimilar for different shapes of matrix blocks, and their analysis presents many technical challenges. The aim of this paper is to derive and analyze a unified approach to transient interporosity flow models for slightly compressible fluids that can be used for any matrix geometry and fracture network. A unified fractional differential transient interporosity flow model is derived using asymptotic analysis for singularly perturbed problems with small parameters arising from the assumption of a much smaller permeability of the matrix blocks compared to that of the fractures. This methodology allowed us to unify existing transient interporosity flow models formulated for different shapes of matrix blocks including bounded matrix blocks, unbounded matrix cylinders with any orthogonal crossection, and matrix slabs. The model is formulated using a fractional order diffusion equation for fluid pressure that involves Caputo derivative of order 1/2 with respect to time. Analysis of the unified fractional derivative model revealed that the surface area-to-volume ratio is the key parameter in the description of the flow through naturally fractured media. Expressions of this parameter are presented for matrix blocks of the same geometrical shape as well as combinations of different shapes with constant and random sizes. Numerical comparisons between the predictions of the unified model and those obtained from existing transient interporosity ones for matrix blocks in the form of slabs, spheres and cylinders are presented for linear, radial and spherical flow types for
Analytical model for transient fluid mixing in upper outlet plenum of an LMFBR
International Nuclear Information System (INIS)
Yang, J.W.; Agrawal, A.K.
1976-01-01
A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the outlet plenum of an LMFBR. The maximum penetration of core flow is used as the criterion for dividing the sodium region into two mixing zones. The 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 by-pass flow into the plenum. The results of numerical calculations indicate that effects of flow stratification, chimney height, metal heat capacity and by-pass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas do not play any significant role on sodium temperature
International Nuclear Information System (INIS)
March-Leuba, J.; Rey, J.M.
1992-01-01
This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of 'unexpected' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a 'new and improved' state of the art has emerged recently. (authors). 6 figs., 57 refs., 1 appendix
Experimental Verification of the Transient Model in an Enrichment Circuit
International Nuclear Information System (INIS)
Fernandino, Maria; Brasnarof, Daniel; Delmastro, Dario
2003-01-01
In the present work an experimental closed loop representing a single stage of an uranium gaseous diffusion enrichment cascade is described, loop that is used to experimentally validate an analytical model that describes the dynamics inside such a loop.The conditions established inside the experimental loop after a few working hours were reproduced by the analytical model, leaving the slower thermal phenomena taking place for future studies.Two kinds of perturbations were experimentally introduced: a change in the range of operation of one of the compressors and the addition of mass into the loop.Numerical and experimental results are compared and presented in this work. The analytical model proposed was verified against these two changes, with very good agreement in the time response and measured values.This analytical model allows us to determine the characteristic time response of the system
Confirmatory simulation of safety and operational transients in LMFBR systems
International Nuclear Information System (INIS)
Guppy, J.G.; Agrawal, A.K.
1978-01-01
Operational and safety transients (anticipated, unlikely, or extremely unlikely) that may originate anywhere in a liquid-metal fast breeder reactor (LMFBR) system must be adequately simulated to assist in safety evaluation and plant design efforts. An advanced thermohydraulic transient code, the Super System Code (SSC), is described that may be used for confirmatory safety evaluations of plant-wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions. Results obtained with SSC illustrating the degree of modeling detail present in the code as well as the computing efficiency are presented. A version of the SSC code, SSC-L, applicable to any loop-type LMFBR design, has been developed at Brookhaven. The scope of SSC-L is to enable the simulation of all plant-wide transients covered by Plant Protection System (PPS) action, including sodium pipe rupture and coastdown to natural circulation conditions. The computations are stopped when loss of core integrity (i.e., clad melting temperature exceeded) is indicated
Nuclear reactors transients identification and classification system
International Nuclear Information System (INIS)
Bianchi, Paulo Henrique
2008-01-01
This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)
1D thermonuclear model for x-ray transients
International Nuclear Information System (INIS)
Wallace, R.K.
1982-01-01
The thermonuclear evolution of a 1.41 M solar mass neutron star, with a radius of 14.3 km, accreting various mixtures of hydrogen, helium, and heavy elements at rates of 10 -11 to 10 -10 M solar mass/yr is examined, in conjunction with S.E. Woosley and T.A. Weaver, using a one-dimensional numerical model. We have ignored any effects due to general relativity or magnetic fields. Two cases shall be discussed. In both models, the accretion rate is such that the hydrogen shell burns to helium in steady state, with the hydrogen burning stabilized by the β-limited CNO cycle. A thick helium shell is produced, which is eventually ignited under extremely degenerate conditions, producing a thermonuclear runaway
BR2 reactor core steady state transient modeling
International Nuclear Information System (INIS)
Makarenko, A.; Petrova, T.
2000-01-01
A coupled neutronics/hydraulics/heat-conduction model of the BR2 reactor core is under development at SCK-CEN. The neutron transport phenomenon has been implemented as steady state and time dependent nodal diffusion. The non-linear heat conduction equation in-side fuel elements is solved with a time dependent finite element method. To allow coupling between functional modules and to simulate subcooled regimes, a simple single-phase hydraulics has been introduced, while the two-phase hydraulics is under development. Multiple tests, general benchmark cases as well as calculation/experiment comparisons demonstrated a good accuracy of both neutronic and thermal hydraulic models, numerical reliability and full code portability. A refinement methodology has been developed and tested for better neutronic representation in hexagonal geometry. Much effort is still needed to complete the development of an extended cross section library with kinetic data and two-phase flow representation. (author)
Model for transient simulation in a PWR steam circuit
International Nuclear Information System (INIS)
Mello, L.A. de.
1982-11-01
A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt
A large deviations approach to the transient of the Erlang loss model
Mandjes, M.R.H.; Ridder, Annemarie
2001-01-01
This paper deals with the transient behavior of the Erlang loss model. After scaling both arrival rate and number of trunks, an asymptotic analysis of the blocking probability is given. Apart from that, the most likely path to blocking is given. Compared to Shwartz and Weiss [Large Deviations for
Variable thickness transient ground-water flow model. Volume 1. Formulation
International Nuclear Information System (INIS)
Reisenauer, A.E.
1979-12-01
Mathematical formulation for the variable thickness transient (VTT) model of an aquifer system is presented. The basic assumptions are described. Specific data requirements for the physical parameters are discussed. The boundary definitions and solution techniques of the numerical formulation of the system of equations are presented
International Nuclear Information System (INIS)
Scharfmann, E.; Silva, D.E. da
1981-01-01
The modifications on the phase separation model and heat tranfer model in Relap4/Mod 5 computer code, in order to make more realistic estimates of the core thermohydraulic behavior submitted to a loss of coolant accident. This research is directed to the accident analysis caused by small breaks in the primary circuits of PWR plants, where two-phase flow occurs most of the time. Calculation have been performed with the help of the original version of Relap code, as well as the version containing the proposed modifications on this work. Comparing one results with the original ones, we arrive at the conclusion that our results show more conservative values of core pressure and coolant temperature, while the peak values of fuel temperature are not exceeded. (Author) [pt
A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector
Directory of Open Access Journals (Sweden)
Saas Laurent
2017-01-01
Full Text Available In the context of the simulation of the Severe Accidents (SA in Light Water Reactors (LWR, we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This 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.
Multifield methods for nuclear thermohydraulics problems
International Nuclear Information System (INIS)
Banerjee, S.
1987-01-01
The multifield model, in which separate sets of conservation equations are written for each phase, or clearly identifiable portions of a phase, is derived by averaging the local instantaneous equations. The closure relationships required to replace information lost in the averaging process are discussed. The mathematical structure of the model is considered and it is shown that application to a variety of problems in which the phases are well separated leads to good predictions of experimental data. For problems in which the phases are more closely coupled, the model is more difficult to apply correctly. However, careful consideration of interfield momentum and heat transfer is shown to give excellent results for some complex problems like density wave propagation in bubbly flows. The model in its present form is shown to be less useful for highly intermittent regimes like slug and churn flows. Data on a reflux condensation situation near the flooding point are discussed to indicate directions in which further work is required
A simple heat transfer model for a heat flux plate under transient conditions
International Nuclear Information System (INIS)
Ryan, L.; Dale, J.D.
1985-01-01
Heat flux plates are used for measuring rates of heat transfer through surfaces under steady state and transient conditions. Their usual construction is to have a resistive layer bounded by thermopiles and an exterior layer for protection. If properly designed and constructed a linear relationship between the thermopile generated voltage and heat flux results and calibration under steady state conditions is straight forward. Under transient conditions however the voltage output from a heat flux plate cannot instantaneously follow the heat flux because of the thermal capacitance of the plate and the resulting time lag. In order to properly interpret the output of a heat flux plate used under transient conditions a simple heat transfer model was constructed and tested. (author)
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
Modelling and transient stability of large wind farms
DEFF Research Database (Denmark)
Akhmatov, Vladislav; Knudsen, Hans; Nielsen, Arne Hejde
2003-01-01
by a physical model of grid-connected windmills. The windmill generators ate conventional induction generators and the wind farm is ac-connected to the power system. Improvements-of short-term voltage stability in case of failure events in the external power system are treated with use of conventional generator...... technology. This subject is treated as a parameter study with respect to the windmill electrical and mechanical parameters and with use of control strategies within the conventional generator technology. Stability improvements on the wind farm side of the connection point lead to significant reduction...
Modelling transient 3D multi-phase criticality in fluidised granular materials - the FETCH code
International Nuclear Information System (INIS)
Pain, C.C.; Gomes, J.L.M.A.; Eaton, M.D.; Ziver, A.K.; Umpleby, A.P.; Oliveira, C.R.E. de; Goddard, A.J.H.
2003-01-01
The development and application of a generic model for modelling criticality in fluidised granular materials is described within the Finite Element Transient Criticality (FETCH) code - which models criticality transients in spatial and temporal detail from fundamental principles, as far as is currently possible. The neutronics model in FETCH solves the neutron transport in full phase space with a spherical harmonics angle of travel representation, multi-group in neutron energy, Crank Nicholson based in time stepping, and finite elements in space. The fluids representation coupled with the neutronics model is a two-fluid-granular-temperature model, also finite element fased. A separate fluid is used to represent the liquid/vapour gas and the solid fuel particle phases, respectively. Particle-particle, particle-wall interactions are modelled using a kinetic theory approach on an analogy between the motion of gas molecules subject to binary collisions and granular flows. This model has been extensively validated by comparison with fluidised bed experimental results. Gas-fluidised beds involve particles that are often extremely agitated (measured by granular temperature) and can thus be viewed as a particularly demanding application of the two-fluid model. Liquid fluidised systems are of criticality interest, but these can become demanding with the production of gases (e.g. radiolytic and water vapour) and large fluid/particle velocities in energetic transients. We present results from a test transient model in which fissile material ( 239 Pu) is presented as spherical granules subsiding in water, located in a tank initially at constant temperature and at two alternative over-pressures in order to verify the theoretical model implemented in FETCH. (author)
TRACE/PARCS modelling of rips trip transients for Lungmen ABWR
Energy Technology Data Exchange (ETDEWEB)
Chang, C. Y. [Inst. of Nuclear Engineering and Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Lin, H. T.; Wang, J. R. [Inst. of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Shih, C. [Inst. of Nuclear Engineering and Science, Dept. of Engineering and System Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China)
2012-07-01
The objectives of this study are to examine the performances of the steady-state results calculated by the Lungmen TRACE/PARCS model compared to SIMULATE-3 code, as well as to use the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen TRACE/PARCS model. In this study, three power generation methods in TRACE were utilized to analyze the three reactor internal pumps (RIPs) trip transient for the purpose of validating the TRACE/PARCS model. In general, the comparisons show that the transient responses of key system parameters agree well with the FSAR results, including core power, core inlet flow, reactivity, etc. Further studies will be performed in the future using Lungmen TRACE/PARCS model. After the commercial operation of Lungmen nuclear power plant, TRACE/PARCS model will be verified. (authors)
Thermohydraulic design of the CAREM reactor's containment
International Nuclear Information System (INIS)
Abbate, P.
1990-01-01
The analysis made for the calculation of the temperature and pressure basic charges affecting the CAREM 25 reactor's contention system is presented. The case of a primary cooler loss, with simultaneous loss of electric supply, is analyzed. Different aspects of the numerical model used, are discussed. (Author) [es
The SABRE code for fuel rod cluster thermohydraulics
International Nuclear Information System (INIS)
Macdougall, J.D.; Lillington, J.N.
1984-01-01
This paper describes the capabilities of the SABRE code for the calculation of single phase and two phase fluid flow and temperature in fuel pin bundles, discusses the methods used in the modelling and solution of the problem, and presents some results including comparison with experiments. The SABRE code permits calculation of steady-state or transient, single or two phase flows and the geometrical options include general representation of grids, wire wraps, multiple blockages, bowed pins, etc. The derivation and solution of the difference equations is discussed. Emphasis is given to the derivation of the spatial differences in triangular subchannel geometry, and the use of central, upward or vector upwind schemes. The method of solution of the difference equations is described for both steady state and transient problems. Together with these topics we consider the problems involved in turbulence modelling and how it is implemented in SABRE. This includes supporting work with a fine scale curvilinear coordinate programme to provide turbulence source data. The problem of modelling boiling flows is discussed, with particular reference to the numerical problems caused by the rapid density change on boiling. The final part of the paper presents applications of the code to the analysis of blockage situations, the study of flow and power transients and analysis of natural circulation within clusters to demonstrate the scope of the code and compare with available experimental results. The comparisons include the calculation of a flow pressure drop characteristic of a boiling channel showing the Ledinegg instability, examples of overpower and flow rundown transients which lead to coolant boiling, and calculation of natural circulation within a rod cluster. (orig./GL)
Nonlinear Modeling of Forced Magnetic Reconnection with Transient Perturbations
Beidler, Matthew T.; Callen, James D.; Hegna, Chris C.; Sovinec, Carl R.
2017-10-01
Externally applied 3D magnetic fields in tokamaks can penetrate into the plasma and lead to forced magnetic reconnection, and hence magnetic islands, on resonant surfaces. Analytic theory has been reasonably successful in describing many aspects of this paradigm with regard to describing the time asymptotic-steady state. However, understanding the nonlinear evolution into a low-slip, field-penetrated state, especially how MHD events such as sawteeth and ELMs precipitate this transition, is in its early development. We present nonlinear computations employing the extended-MHD code NIMROD, building on previous work by incorporating a temporally varying external perturbation as a simple model for an MHD event that produces resonant magnetic signals. A parametric series of proof-of-principle computations and accompanying analytical theory characterize the transition into a mode-locked state with an emphasis on detailing the temporal evolution properties. Supported by DOE OFES Grants DE-FG02-92ER54139, DE-FG02-86ER53218, and the U.S. DOE FES Postdoctoral Research program administered by ORISE and managed by ORAU under DOE contract DE-SC0014664.
ITER transient consequences for material damage: modelling versus experiments
International Nuclear Information System (INIS)
Bazylev, B; Janeschitz, G; Landman, I; Pestchanyi, S; Loarte, A; Federici, G; Merola, M; Linke, J; Zhitlukhin, A; Podkovyrov, V; Klimov, N; Safronov, V
2007-01-01
Carbon-fibre composite (CFC) and tungsten macrobrush armours are foreseen as PFC for the ITER divertor. In ITER the main mechanisms of metallic armour damage remain surface melting and melt motion erosion. In the case of CFC armour, due to rather different heat conductivities of CFC fibres a noticeable erosion of the PAN bundles may occur at rather small heat loads. Experiments carried out in the plasma gun facilities QSPA-T for the ITER like edge localized mode (ELM) heat load also demonstrated significant erosion of the frontal and lateral brush edges. Numerical simulations of the CFC and tungsten (W) macrobrush target damage accounting for the heat loads at the face and lateral brush edges were carried out for QSPA-T conditions using the three-dimensional (3D) code PHEMOBRID. The modelling results of CFC damage are in a good qualitative and quantitative agreement with the experiments. Estimation of the droplet splashing caused by the Kelvin-Helmholtz (KH) instability was performed
STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY
Energy Technology Data Exchange (ETDEWEB)
Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart
2016-05-01
With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.
Testing the ontogenetic base for the transient model of inflorescence development.
Bull-Hereñu, Kester; Claßen-Bockhoff, Regine
2013-11-01
Current research in plant science has concentrated on revealing ontogenetic processes of key attributes in plant evolution. One recently discussed model is the 'transient model' successful in explaining some types of inflorescence architectures based on two main principles: the decline of the so called 'vegetativeness' (veg) factor and the transient nature of apical meristems in developing inflorescences. This study examines whether both principles find a concrete ontogenetic correlate in inflorescence development. To test the ontogenetic base of veg decline and the transient character of apical meristems the ontogeny of meristematic size in developing inflorescences was investigated under scanning electron microscopy. Early and late inflorescence meristems were measured and compared during inflorescence development in 13 eudicot species from 11 families. The initial size of the inflorescence meristem in closed inflorescences correlates with the number of nodes in the mature inflorescence. Conjunct compound inflorescences (panicles) show a constant decrease of meristematic size from early to late inflorescence meristems, while disjunct compound inflorescences present an enlargement by merging from early inflorescence meristems to late inflorescence meristems, implying a qualitative change of the apical meristems during ontogeny. Partial confirmation was found for the transient model for inflorescence architecture in the ontogeny: the initial size of the apical meristem in closed inflorescences is consistent with the postulated veg decline mechanism regulating the size of the inflorescence. However, the observed biphasic kinetics of the development of the apical meristem in compound racemes offers the primary explanation for their disjunct morphology, contrary to the putative exclusive transient mechanism in lateral axes as expected by the model.
Simplified distributed parameters BWR dynamic model for transient and stability analysis
International Nuclear Information System (INIS)
Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro
2006-01-01
This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR
DEFF Research Database (Denmark)
Mohammadi, Soma; Bojesen, Carsten
2015-01-01
the temperature in DH systems. The main focus is on modeling transient heat transfer in pipe networks regarding the time delays between the heat supply unit and the consumers, the heat loss in the pipe networks and the consumers’ dynamic heat loads. A pseudo-dynamic approach is adopted and also the implicit...... district heating networks [DHN] characteristics. This paper is presenting a new developed model, which reflects the thermo-dynamic behavior of DHN. It is designed for tree network topologies. The purpose of the model is to serve as a basis for applying a variety of scenarios towards lowering...... finite element method is applied to simulate transient temperature changes in pipe networks. The model is calculating time series data related to supply temperature to the DHN from heat production units, heat loads and return temperature related to each consumer to calculate dynamic temperature changes...
A novel approach to model the transient behavior of solid-oxide fuel cell stacks
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.
Energy Technology Data Exchange (ETDEWEB)
Crosby, Tamer, E-mail: tcrosby@ucla.edu; Ghoniem, Nasr M., E-mail: ghoniem@ucla.edu
2013-11-15
A combination of transient heating and bombardment by helium and hydrogen atoms has been experimentally proven to lead to severe surface and sub-surface damage. We developed a computational model to determine the relationship between the thermomechanical loading conditions and the onset of damage and failure of tungsten surfaces. The model is based on a thermoelasticity fracture damage approach that was developed using the phase field method. The model simulates the distribution of helium bubbles inside the grains and on grain boundaries using space-dependent rate theory. In addition, the model is coupled with a transient heat conduction analysis for temperature distributions inside the material. The results show the effects of helium bubbles on reducing tungsten surface energy. Further, a temperature gradient in the material equals to 10 K/μm, resulted in deep cracks propagating from the tungsten surface.
Development of a Transient Model of a Stirling-Based CHP System
Directory of Open Access Journals (Sweden)
Antón Cacabelos
2013-06-01
Full Text Available Although the Stirling engine was invented in 1816, this heat engine still continues to be investigated due to the variety of energy sources that can be used to power it (e.g., solar energy, fossil fuels, biomass, and geothermal energy. To study the performance of these machines, it is necessary to develop and simulate models under different operating conditions. In this paper, we present a one-dimensional dynamic model based on components from Trnsys: principally, a lumped mass and a heat exchanger. The resulting model is calibrated using GenOpt. Furthermore, the obtained model can be used to simulate the machine both under steady-state operation and during a transient response. The results provided by the simulations are compared with data measured in a Stirling engine that has been subjected to different operating conditions. This comparison shows good agreement, indicating that the model is an appropriate method for transient thermal simulations. This new proposed model requires few configuration parameters and is therefore easily adaptable to a wide range of commercial models of Stirling engines. A detailed analysis of the system results reveals that the power is directly related to the difference of temperatures between the hot and cold sources during the transient and steady-state processes.
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.
International Nuclear Information System (INIS)
Kamide, H.; Ieda, Y.; Toda, S.; Isozaki, T.; Sugawara, S.
1993-01-01
Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor core during natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)
Energy Technology Data Exchange (ETDEWEB)
Kamide, H; Ieda, Y; Toda, S; Isozaki, T; Sugawara, S [Reactor Engineering Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, Narita, O-arai, Ibaraki-ken (Japan)
1993-02-01
Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor coreduring natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)
Simulation model of a transient fault controller for an active-stall wind turbine
Energy Technology Data Exchange (ETDEWEB)
Jauch, C.; Soerensen, P.; Bak Jensen, B.
2005-01-01
This paper describes the simulation model of a controller that enables an active-stall wind turbine to ride through transient faults. The simulated wind turbine is connected to a simple model of a power system. Certain fault scenarios are specified and the turbine shall be able to sustain operation in case of such faults. The design of the controller is described and its performance assessed by simulations. The control strategies are explained and the behaviour of the turbine discussed. (author)
Validation and verification of the MTRPC thermohydraulic package
International Nuclear Information System (INIS)
Doval, Alicia
1998-01-01
The MTR P C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)
Thermohydraulic behaviour and heat transfer in the molten core
International Nuclear Information System (INIS)
Reineke, H.H.
1977-01-01
Increasing the application of nuclear reactors to produce electrical power extremely unprobable accidents should be investigated too. In the Federal Republic of Germany, a research program is performed for some years engaged in accidents at light water reactors in which the melting of the reactor core is presumed. A part of this program is to investigate the thermohydraulic and the heat transfer behavior in an accumulation of molten core material. The knowledge of these events is necessary to analyse the accident exactly. Further on the results of this work are of great importance to build a catcher for the molten core material. As a result of the decay heat the molten material is heated up and the density differences induce a free convection motion. In this work the thermohydraulic behavior and the distribution of the escaping heat fluxes for several accumulations of molten core material were determined. The numerical methods for solving the system of partial differential equation were used to develop computer codes, able to compute the average and local heat fluxes at the walls enclosing the molten core material and the inside increase of the temperature. The numerical computations were confirmed and verified by experimental investigations. In these investigations the molten core material was always assumed as a homogeneous fluid. In this case, the results could be reproduced by simple power laws
One component, volume heated, boiling pool thermohydraulics
International Nuclear Information System (INIS)
Bede, M.; Perret, C.; Pretrel, H.; Seiler, J.M.
1993-01-01
Prior work on boiling pools provided heat exchange correlations valid for bubbly flow with laminar or turbulent boundary layers. New experiments performed with water (SEBULON) and UO 2 (SCARABEE BF2) in a churn-turbulent flow configuration show unexpected heat flux distributions for which the maximum heat flux may be situated well below the pool surface. The origin of this behaviour is attributed to condensation effects, very unstable boundary layer flow and surface oscillation. A calculation model is discussed which permits to approach the experimental heat flux distribution with reasonable accuracy. (authors). 7 figs., 2 appendix., 14 refs
Limitations of transient power loads on DEMO and analysis of mitigation techniques
Energy Technology Data Exchange (ETDEWEB)
Maviglia, F., E-mail: francesco.maviglia@euro-fusion.org [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Consorzio CREATE, University Napoli Federico II – DIETI, 80125 Napoli (Italy); Federici, G. [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Strohmayer, G. [Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Wenninger, R. [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Bachmann, C. [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Albanese, R. [Consorzio CREATE, University Napoli Federico II – DIETI, 80125 Napoli (Italy); Ambrosino, R. [Consorzio CREATE University Napoli Parthenope, Naples (Italy); Li, M. [Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Loschiavo, V.P. [Consorzio CREATE, University Napoli Federico II – DIETI, 80125 Napoli (Italy); You, J.H. [Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Zani, L. [CEA, IRFM, F-13108 St Paul-Lez-Durance (France)
2016-11-01
Highlights: • A parametric thermo-hydraulic analysis of the candidate DEMO divertor is presented. • The operational space assessment is presented under static and transient heat loads. • Strike points sweeping is analyzed as a divertor power exhaust mitigation technique. • Results are presented on sweeping installed power required, AC losses and thermal fatigue. - Abstract: The present European standard DEMO divertor target technology is based on a water-cooled tungsten mono-block with a copper alloy heat sink. This paper presents the assessment of the operational space of this technology under static and transient heat loads. A transient thermo-hydraulic analysis was performed using the code RACLETTE, which allowed a broad parametric scan of the target geometry and coolant conditions. The limiting factors considered were the coolant critical heat flux (CHF), and the temperature limits of the materials. The second part of the work is devoted to the study of the plasma strike point sweeping as a mitigation technique for the divertor power exhaust. The RACLETTE code was used to evaluate the impact of a large range of sweeping frequencies and amplitudes. A reduced subset of cases, which complied with the constraints, was benchmarked with a 3D FEM model. A reduction of the heat flux to the coolant, up to a factor ∼4, and lower material temperatures were found for an incident heat flux in the range (15–30) MW/m{sup 2}. Finally, preliminary assessments were performed on the installed power required for the sweeping, the AC losses in the superconductors and thermal fatigue analysis. No evident show stoppers were found.
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 the effect of transient populations on epidemics in Washington DC.
Parikh, Nidhi; Youssef, Mina; Swarup, Samarth; Eubank, Stephen
2013-11-06
Large numbers of transients visit big cities, where they come into contact with many people at crowded areas. However, epidemiological studies have not paid much attention to the role of this subpopulation in disease spread. We evaluate the effect of transients on epidemics by extending a synthetic population model for the Washington DC metro area to include leisure and business travelers. A synthetic population is obtained by combining multiple data sources to build a detailed minute-by-minute simulation of population interaction resulting in a contact network. We simulate an influenza-like illness over the contact network to evaluate the effects of transients on the number of infected residents. We find that there are significantly more infections when transients are considered. Since much population mixing happens at major tourism locations, we evaluate two targeted interventions: closing museums and promoting healthy behavior (such as the use of hand sanitizers, covering coughs, etc.) at museums. Surprisingly, closing museums has no beneficial effect. However, promoting healthy behavior at the museums can both reduce and delay the epidemic peak. We analytically derive the reproductive number and perform stability analysis using an ODE-based model.
Modeling the effect of transient populations on epidemics in Washington DC
Parikh, Nidhi; Youssef, Mina; Swarup, Samarth; Eubank, Stephen
2013-11-01
Large numbers of transients visit big cities, where they come into contact with many people at crowded areas. However, epidemiological studies have not paid much attention to the role of this subpopulation in disease spread. We evaluate the effect of transients on epidemics by extending a synthetic population model for the Washington DC metro area to include leisure and business travelers. A synthetic population is obtained by combining multiple data sources to build a detailed minute-by-minute simulation of population interaction resulting in a contact network. We simulate an influenza-like illness over the contact network to evaluate the effects of transients on the number of infected residents. We find that there are significantly more infections when transients are considered. Since much population mixing happens at major tourism locations, we evaluate two targeted interventions: closing museums and promoting healthy behavior (such as the use of hand sanitizers, covering coughs, etc.) at museums. Surprisingly, closing museums has no beneficial effect. However, promoting healthy behavior at the museums can both reduce and delay the epidemic peak. We analytically derive the reproductive number and perform stability analysis using an ODE-based model.
Improved lumped parameter for annular fuel element thermohydraulic analysis
International Nuclear Information System (INIS)
Duarte, Juliana Pacheco; Su, Jian; Alvim, Antonio Carlos Marques
2011-01-01
Annular fuel elements have been intensively studied for the purpose of increasing power density in light water reactors (LWR). This paper presents an improved lumped parameter model for the dynamics of a LWR core with annular fuel elements, composed of three sub-models: the fuel dynamics model, the neutronics model, and the coolant energy balance model. The transient heat conduction in radial direction is analyzed through an improved lumped parameter formulation. The Hermite approximation for integration is used to obtain the average temperature of the fuel and cladding and also to obtain the average heat flux. The volumetric heat generation in fuel rods was obtained with the point kinetics equations with six delayed neutron groups. The equations for average temperature of fuel and cladding are solved along with the point kinetic equations, assuming linear reactivity and coolant temperature in cases of reactivity insertion. The analytical development of the model and the numeric solution of the ordinary differential equation system were obtained by using Mathematica 7.0. The dynamic behaviors for average temperatures of fuel, cladding and coolant in transient events as well as the reactor power were analyzed. (author)
Seymour, David C.; Martin, Michael A.; Nguyen, Huy H.; Greene, William D.
2005-01-01
The subject of mathematical modeling of the transient operation of liquid rocket engines is presented in overview form from the perspective of engineers working at the NASA Marshall Space Flight Center. The necessity of creating and utilizing accurate mathematical models as part of liquid rocket engine development process has become well established and is likely to increase in importance in the future. The issues of design considerations for transient operation, development testing, and failure scenario simulation are discussed. An overview of the derivation of the basic governing equations is presented along with a discussion of computational and numerical issues associated with the implementation of these equations in computer codes. Also, work in the field of generating usable fluid property tables is presented along with an overview of efforts to be undertaken in the future to improve the tools use for the mathematical modeling process.
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.
International Nuclear Information System (INIS)
Chen, Y.-S.; Chien, K.-H.; Wang, C.-C.; Hung, T.-C.; Pei, B.-S.
2006-01-01
The vapor chambers (flat plate heat pipes) have been applied on the electronic cooling recently. To satisfy the quick-response requirement of the industries, a simplified transient three-dimensional linear model has been developed and tested in this study. In the proposed model, the vapor is assumed as a single interface between the evaporator and condenser wicks, and this assumption enables the vapor chamber to be analyzed by being split into small control volumes. Comparing with the previous available results, the calculated transient responses have shown good agreements with the existing results. For further validation of the proposed model, a water-cooling experiment was conducted. In addition to the vapor chamber, the heating block is also taken into account in the simulation. It is found that the inclusion of the capacitance of heating block shows a better agreement with the measurements
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
Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model
Energy Technology Data Exchange (ETDEWEB)
Baudron, Anne-Marie, E-mail: anne-marie.baudron@cea.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex (France); Lautard, Jean-Jacques, E-mail: jean-jacques.lautard@cea.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex (France); Maday, Yvon, E-mail: maday@ann.jussieu.fr [Sorbonne Universités, UPMC Univ Paris 06, UMR 7598, Laboratoire Jacques-Louis Lions and Institut Universitaire de France, F-75005, Paris (France); Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); Brown Univ, Division of Applied Maths, Providence, RI (United States); Riahi, Mohamed Kamel, E-mail: riahi@cmap.polytechnique.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CMAP, Inria-Saclay and X-Ecole Polytechnique, Route de Saclay, 91128 Palaiseau Cedex (France); Salomon, Julien, E-mail: salomon@ceremade.dauphine.fr [CEREMADE, Univ Paris-Dauphine, Pl. du Mal. de Lattre de Tassigny, F-75016, Paris (France)
2014-12-15
In this paper we present a time-parallel algorithm for the 3D neutrons calculation of a transient model in a nuclear reactor core. The neutrons calculation consists in numerically solving the time dependent diffusion approximation equation, which is a simplified transport equation. The numerical resolution is done with finite elements method based on a tetrahedral meshing of the computational domain, representing the reactor core, and time discretization is achieved using a θ-scheme. The transient model presents moving control rods during the time of the reaction. Therefore, cross-sections (piecewise constants) are taken into account by interpolations with respect to the velocity of the control rods. The parallelism across the time is achieved by an adequate use of the parareal in time algorithm to the handled problem. This parallel method is a predictor corrector scheme that iteratively combines the use of two kinds of numerical propagators, one coarse and one fine. Our method is made efficient by means of a coarse solver defined with large time step and fixed position control rods model, while the fine propagator is assumed to be a high order numerical approximation of the full model. The parallel implementation of our method provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch–Maurer–Werner benchmark.
Research on Model-Based Fault Diagnosis for a Gas Turbine Based on Transient Performance
Directory of Open Access Journals (Sweden)
Detang Zeng
2018-01-01
Full Text Available It is essential to monitor and to diagnose faults in rotating machinery with a high thrust–weight ratio and complex structure for a variety of industrial applications, for which reliable signal measurements are required. However, the measured values consist of the true values of the parameters, the inertia of measurements, random errors and systematic errors. Such signals cannot reflect the true performance state and the health state of rotating machinery accurately. High-quality, steady-state measurements are necessary for most current diagnostic methods. Unfortunately, it is hard to obtain these kinds of measurements for most rotating machinery. Diagnosis based on transient performance is a useful tool that can potentially solve this problem. A model-based fault diagnosis method for gas turbines based on transient performance is proposed in this paper. The fault diagnosis consists of a dynamic simulation model, a diagnostic scheme, and an optimization algorithm. A high-accuracy, nonlinear, dynamic gas turbine model using a modular modeling method is presented that involves thermophysical properties, a component characteristic chart, and system inertial. The startup process is simulated using this model. The consistency between the simulation results and the field operation data shows the validity of the model and the advantages of transient accumulated deviation. In addition, a diagnostic scheme is designed to fulfill this process. Finally, cuckoo search is selected to solve the optimization problem in fault diagnosis. Comparative diagnostic results for a gas turbine before and after washing indicate the improved effectiveness and accuracy of the proposed method of using data from transient processes, compared with traditional methods using data from the steady state.
Modeling of boron control during power transients in a pressurized water reactor
International Nuclear Information System (INIS)
Mathieu, P.; Distexhe, E.
1986-01-01
Accurate control instructions in a reactor control aid computer are included in order to realize the boron makeup throughput, which is required to obtain the boron concentration in the primary coolant loop, predicted by a neutronic code. A modeling of the transfer function between the makeup and the primary loop is proposed. The chemical and volumetric control system, the pressurizer, and the primary loop are modeled as instantaneous diffusion cells. The pipes are modeled as time lag lines. The model provides the unstationary boron distributions in the different elements of the setup. A numerical code is developed to calculate the time evolutions of the makeup throughput during power transients
A model for transient analysis of a multiple-medium confinement filter system
International Nuclear Information System (INIS)
Hyder, M.L.; Ellison, P.G.; Leonard, M.T.; Louie, D.L.Y.; Donbroski, E.L.; Wagner, K.C.
1990-01-01
A computational model is described that calculates the transient behavior of aerosol and vapor (adsorption) filter compartments such as those used in the Savannah River Site (SRS) production reactor confinement system. The principal application of the model is in the analysis of confinement response to hypothetical severe (core melt) accidents. Under these conditions, aerosol and radio-iodine deposition on filter compartments may be substantial. Attendant filter degradation mechanisms are modeled. Sample calculations are included to illustrate model performance. 6 refs., 14 figs., 1 tab
Coupling a Transient Solvent Extraction Module with the Separations and Safeguards Performance Model
Energy Technology Data Exchange (ETDEWEB)
de Almeida, Valmor F [ORNL; Birdwell Jr, Joseph F [ORNL; DePaoli, David W [ORNL; Gauld, Ian C [ORNL
2009-10-01
A past difficulty in safeguards design for reprocessing plants is that no code existed for analysis and evaluation of the design. A number of codes have been developed in the past, but many are dated, and no single code is able to cover all aspects of materials accountancy, process monitoring, and diversion scenario analysis. The purpose of this work was to integrate a transient solvent extraction simulation module developed at Oak Ridge National Laboratory, with the SSPM Separations and Safeguards Performance Model, developed at Sandia National Laboratory, as a first step toward creating a more versatile design and evaluation tool. The SSPM was designed for materials accountancy and process monitoring analyses, but previous versions of the code have included limited detail on the chemical processes, including chemical separations. The transient solvent extraction model is based on the ORNL SEPHIS code approach to consider solute build up in a bank of contactors in the PUREX process. Combined, these capabilities yield a much more robust transient separations and safeguards model for evaluating safeguards system design. This coupling and the initial results are presented. In addition, some observations toward further enhancement of separations and safeguards modeling based on this effort are provided, including: items to be addressed in integrating legacy codes, additional improvements needed for a fully functional solvent extraction module, and recommendations for future integration of other chemical process modules.
Energy Technology Data Exchange (ETDEWEB)
DePaoli, David W. (Oak Ridge National Laboratory, Oak Ridge, TN); Birdwell, Joseph F. (Oak Ridge National Laboratory, Oak Ridge, TN); Gauld, Ian C. (Oak Ridge National Laboratory, Oak Ridge, TN); Cipiti, Benjamin B.; de Almeida, Valmor F. (Oak Ridge National Laboratory, Oak Ridge, TN)
2009-10-01
A number of codes have been developed in the past for safeguards analysis, but many are dated, and no single code is able to cover all aspects of materials accountancy, process monitoring, and diversion scenario analysis. The purpose of this work was to integrate a transient solvent extraction simulation module developed at Oak Ridge National Laboratory, with the Separations and Safeguards Performance Model (SSPM), developed at Sandia National Laboratory, as a first step toward creating a more versatile design and evaluation tool. The SSPM was designed for materials accountancy and process monitoring analyses, but previous versions of the code have included limited detail on the chemical processes, including chemical separations. The transient solvent extraction model is based on the ORNL SEPHIS code approach to consider solute build up in a bank of contactors in the PUREX process. Combined, these capabilities yield a more robust transient separations and safeguards model for evaluating safeguards system design. This coupling and initial results are presented. In addition, some observations toward further enhancement of separations and safeguards modeling based on this effort are provided, including: items to be addressed in integrating legacy codes, additional improvements needed for a fully functional solvent extraction module, and recommendations for future integration of other chemical process modules.
Transient modeling of an air conditioner with a rapid cycling compressor and multi-indoor units
International Nuclear Information System (INIS)
Zhang Weijiang; Zhang Chunlu
2011-01-01
Rapid cycling the compressor is an alternative of the variable speed compressor to modulate the capacity of refrigeration systems for the purpose of energy saving at part-load conditions. The multi-evaporator air conditioner combined with the rapid cycling compressor brings difficulties in control design because of the sophisticated system physics and dynamics. In this paper the transient model of a multi-split air conditioner with a digital scroll compressor is developed for predicting the system transients under performance modulations. The predicted cycling dynamics are in good agreement with the experimental data. Based on the validated model, the impact of compressor idle power and cycle period to the part load performance is discussed.
Transient modeling of an air conditioner with a rapid cycling compressor and multi-indoor units
Energy Technology Data Exchange (ETDEWEB)
Zhang, Wei-Jiang [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Zhang, Chun-Lu [College of Mechanical Engineering, Tongji University, 4800 Cao An Highway, Shanghai 201804 (China)
2011-01-15
Rapid cycling the compressor is an alternative of the variable speed compressor to modulate the capacity of refrigeration systems for the purpose of energy saving at part-load conditions. The multi-evaporator air conditioner combined with the rapid cycling compressor brings difficulties in control design because of the sophisticated system physics and dynamics. In this paper the transient model of a multi-split air conditioner with a digital scroll compressor is developed for predicting the system transients under performance modulations. The predicted cycling dynamics are in good agreement with the experimental data. Based on the validated model, the impact of compressor idle power and cycle period to the part load performance is discussed. (author)
Transient modeling of an air conditioner with a rapid cycling compressor and multi-indoor units
Energy Technology Data Exchange (ETDEWEB)
Zhang Weijiang [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Zhang Chunlu, E-mail: chunlu.zhang@carrier.utc.co [College of Mechanical Engineering, Tongji University, 4800 Cao An Highway, Shanghai 201804 (China)
2011-01-15
Rapid cycling the compressor is an alternative of the variable speed compressor to modulate the capacity of refrigeration systems for the purpose of energy saving at part-load conditions. The multi-evaporator air conditioner combined with the rapid cycling compressor brings difficulties in control design because of the sophisticated system physics and dynamics. In this paper the transient model of a multi-split air conditioner with a digital scroll compressor is developed for predicting the system transients under performance modulations. The predicted cycling dynamics are in good agreement with the experimental data. Based on the validated model, the impact of compressor idle power and cycle period to the part load performance is discussed.
Winnerless competition principle and prediction of the transient dynamics in a Lotka-Volterra model
Afraimovich, Valentin; Tristan, Irma; Huerta, Ramon; Rabinovich, Mikhail I.
2008-12-01
Predicting the evolution of multispecies ecological systems is an intriguing problem. A sufficiently complex model with the necessary predicting power requires solutions that are structurally stable. Small variations of the system parameters should not qualitatively perturb its solutions. When one is interested in just asymptotic results of evolution (as time goes to infinity), then the problem has a straightforward mathematical image involving simple attractors (fixed points or limit cycles) of a dynamical system. However, for an accurate prediction of evolution, the analysis of transient solutions is critical. In this paper, in the framework of the traditional Lotka-Volterra model (generalized in some sense), we show that the transient solution representing multispecies sequential competition can be reproducible and predictable with high probability.
Directory of Open Access Journals (Sweden)
Yan-jie Ni
2016-04-01
Full Text Available A 30 mm electrothermal-chemical (ETC gun experimental system is employed to research the burning rate characteristics of 4/7 high-nitrogen solid propellant. Enhanced gas generation rates (EGGR of propellants during and after electrical discharges are verified in the experiments. A modified 0D internal ballistic model is established to simulate the ETC launch. According to the measured pressure and electrical parameters, a transient burning rate law including the influence of EGGR coefficient by electric power and pressure gradient (dp/dt is added into the model. The EGGR coefficient of 4/7 high-nitrogen solid propellant is equal to 0.005 MW−1. Both simulated breech pressure and projectile muzzle velocity accord with the experimental results well. Compared with Woodley's modified burning rate law, the breech pressure curves acquired by the transient burning rate law are more consistent with test results. Based on the parameters calculated in the model, the relationship among propellant burning rate, pressure gradient (dp/dt and electric power is analyzed. Depending on the transient burning rate law and experimental data, the burning of solid propellant under the condition of plasma is described more accurately.
Fabrication, characterization, and modeling of a biodegradable battery for transient electronics
Edupuganti, Vineet; Solanki, Raj
2016-12-01
Traditionally, emphasis has been placed on durable, long-lasting electronics. However, electronics that are meant to intentionally degrade over time can actually have significant practical applications. Biodegradable, or transient, electronics would open up opportunities in the field of medical implants, where the need for surgical removal of devices could be eliminated. Environmental sensors and, eventually, consumer electronics would also greatly benefit from this technology. An essential component of transient electronics is the battery, which serves as a biodegradable power source. This work involves the fabrication, characterization, and modeling of a magnesium-based biodegradable battery. Galvanostatic discharge tests show that an anode material of magnesium alloy AZ31 extends battery lifetime by over six times, as compared to pure magnesium. With AZ31, the maximum power and capacity of the fabricated device are 67 μW and 5.2 mAh, respectively, though the anode area is just 0.8 cm2. The development of an equivalent circuit model provided insight into the battery's behavior by extracting fitting parameters from experimental data. The model can accurately simulate device behavior, taking into account its intentional degradation. The size of the device and the power it produces are in accordance with typical levels for low-power transient systems.
Development of the MARS input model for Kori nuclear units 1 transient analyzer
International Nuclear Information System (INIS)
Hwang, M.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Jeong, J. J.
2004-11-01
KAERI has been developing the 'NSSS transient analyzer' based on best-estimate codes for Kori Nuclear Units 1 plants. The MARS and RETRAN codes have been used as the best-estimate codes for the NSSS transient analyzer. Among these codes, the MARS code is adopted for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. So it is necessary to develop the MARS input model for Kori Nuclear Units 1 plants. This report includes the input model (hydrodynamic component and heat structure models) requirements and the calculation note for the MARS input data generation for Kori Nuclear Units 1 plant analyzer (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Kori Nuclear Units 1
Energy Technology Data Exchange (ETDEWEB)
Coragem, Helio Boemer de Oliveira
1980-07-01
A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)
Energy Technology Data Exchange (ETDEWEB)
Coragem, Helio Boemer de Oliveira
1980-07-01
A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)
Nonlinear time-domain cochlear model for transient stimulation and human otoacoustic emission
DEFF Research Database (Denmark)
Verhulst, Sarah; Dau, Torsten; Shera, Christopher A.
2012-01-01
This paper describes the implementation and performance of a nonlinear time-domain model of the cochlea for transient stimulation and human otoacoustic emission generation. The nonlinearity simulates compressive growth of measured basilar-membrane impulse responses. The model accounts...... for reflection and distortion-source otoacoustic emissions (OAEs) and simulates spontaneous OAEs through manipulation of the middle-ear reflectance. The model was calibrated using human psychoacoustical and otoacoustic tuning parameters. It can be used to investigate time-dependent properties of cochlear...
FAST: a combined NOC and transient fuel performance model using a commercial FEM environment
Energy Technology Data Exchange (ETDEWEB)
Prudil, A.; Bell, J.; Oussoren, A.; Chan, P. [Royal Military College of Canada, Kingston, ON (Canada); Lewis, B. [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)
2014-07-01
The Fuel And Sheath modelling Tool (FAST) is a combined normal operating conditions (NOC) and transient fuel performance code developed on the COMSOL Multiphysics finite-element platform. The FAST code has demonstrated excellent performance in proof of concept validation tests against experimental data and comparison to the ELESIM, ELESTRES and ELOCA fuel performance codes. In this paper we discuss ongoing efforts to expand the capabilities of the code to include multiple pellet geometries, model stress-corrosion cracking phenomena and modelling of advanced fuels composed of mixed oxides of thorium, uranium, and plutonium for the Canadian Supercritical Water Reactor (SCWR). (author)
Bifurcation structures and transient chaos in a four-dimensional Chua model
Energy Technology Data Exchange (ETDEWEB)
Hoff, Anderson, E-mail: hoffande@gmail.com; Silva, Denilson T. da; Manchein, Cesar, E-mail: cesar.manchein@udesc.br; Albuquerque, Holokx A., E-mail: holokx.albuquerque@udesc.br
2014-01-10
A four-dimensional four-parameter Chua model with cubic nonlinearity is studied applying numerical continuation and numerical solutions methods. Regarding numerical solution methods, its dynamics is characterized on Lyapunov and isoperiodic diagrams and regarding numerical continuation method, the bifurcation curves are obtained. Combining both methods the bifurcation structures of the model were obtained with the possibility to describe the shrimp-shaped domains and their endoskeletons. We study the effect of a parameter that controls the dimension of the system leading the model to present transient chaos with its corresponding basin of attraction being riddled.
Mathematical modelling of plant transients in the PWR for simulator purposes
International Nuclear Information System (INIS)
Hartel, K.
1984-01-01
This chapter presents the results of the testing of anticipated and abnormal plant transients in pressurized water reactors (PWRs) of the type WWER 440 by means of the numerical simulation of 32 different, stationary and nonstationary, operational regimes. Topics considered include the formation of the PWR mathematical model, the physical approximation of the reactor core, the structure of the reactor core model, a mathematical approximation of the reactor model, the selection of numerical methods, and a computerized simulation system. The necessity of a PWR simulator in Czechoslovakia is justified by the present status and the outlook for the further development of the Czechoslovak nuclear power complex
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
Transient stratification modelling of a corium pool in a LWR vessel lower head
International Nuclear Information System (INIS)
Le Tellier, R.; Saas, L.; Bajard, S.
2015-01-01
Highlights: • A kinetic stratification model is proposed for the simulation of the in-vessel corium behaviour during a LWR severe accident. • The different associated “modes” of vessel failure by thermal focusing effect are highlighted and discussed. • A sensitivity study for a 1650 MWe GenIII PWR is presented with this model in order to illustrate the associated R&D issues. - Abstract: In the context of light water reactor severe accidents analysis, this paper is focused on one key parameter of in-vessel corium phenomenology: the immiscible phases stratification and its impact on the heat flux distribution at the corium pool lateral boundary with the so-called focusing effect related to a “thin” top metal phase and the potential vessel failure at that point. More particularly, based on the limited knowledge of the stratification transient phenomenon derived from the MASCA-RCW experiment, a basic model is proposed that can be used for corium in lower head sensitivity analyses. It has been implemented in the PROCOR platform developed at CEA Cadarache. A short parametric study on a simple hypothetical transient is presented in order to highlight the different focusing effect “modes” that can be encountered based on this in-vessel corium pool model. An early mode may occur during the formation of the top metal layer while two other modes may appear later during the thinning of this top metal layer because of thermochemically induced mass transfers. Some associated relevant parameters (model or scenario-dependent) and modelling issues are mentioned and illustrated with some results of a Monte-Carlo based sensitivity calculation on the transient behaviour of the corium in the lower head of a 1650 MWe GenIII PWR. Within the limiting modelling hypotheses, the thermal modelling of the steel layer for small (centimetre) heights and the mass diffusivity (limited in this case to the uranium diffusivity in the oxidic layer) are main sensitive parameters
International Nuclear Information System (INIS)
Kuczera, B.; Royl, P.
1975-01-01
The CAPRI-2 code system for analysis of hypothetical core disruptive accidents in LMFBRs has recently been coupled with the transient deformation model BREDA-2. The new code system determines thermal and mechanical loads under transient conditions for both, fresh and irradiated fuel and cladding, taking into account fuel restructuring as well as effects from fission gas and fuel and clad swelling. The system has been used for analysis of mild uncontrolled overpower transients in the SNR-300 to predict failure, and to initialize and calculate subsequent fuel coolant interaction (FCI). Thirteen channels have been coupled by point kinetics for the whole core analysis. Three different failure mechanisms and their influence on accident sequence have been investigated: clad melt-through; clad burst caused by internal pressure build-up; clad straining due to differential thermal expansion between fuel and clad cylinders. The results of these analyses show that each failure mechanism will lead to rather different failure and accident sequences. There is still a lack of experimental data from which failure thresholds can be derived. To get better predictions from the applied models an improved understanding of fission release and its relation to fuel porosity also some better experimental data on fluence and temperature dependent rupture strains of the cladding material should be available
Mountain scale modeling of transient, coupled gas flow, heat transfer and carbon-14 migration
International Nuclear Information System (INIS)
Lu, Ning; Ross, B.
1993-01-01
We simulate mountain-scale coupled heat transfer and gas flow at Yucca Mountain. A coupled rock-gas flow and heat transfer model, TGIF2, is used to simulate mountain-scale two-dimensional transient heat transfer and gas flow. The model is first verified against an analytical solution for the problem of an infinite horizontal layer of fluid heated from below. Our numerical results match very well with the analytical solution. Then, we obtain transient temperature and gas flow distributions inside the mountain. These distributions are used by a transient semianalytical particle tracker to obtain carbon-14 travel times for particles starting at different locations within the repository. Assuming that the repository is filled with 30-year-old waste at an initial areal power density of 57 kw/acre, we find that repository temperatures remain above 60 degrees C for more than 10,000 years. Carbon-14 travel times to the surface are mostly less than 1000 years, for particles starting at any time within the first 10,000 years
Energy Technology Data Exchange (ETDEWEB)
Caro, R
1976-07-01
In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.
Transient response of a five-region nonequilibrium real-time pressurizer model
International Nuclear Information System (INIS)
Fakory, M.R.; Seifaee, F.
1987-01-01
Recent accidents at nuclear power plants in the US and abroad have prompted accurate analysis and simulation of the plant systems and the training of reactor operators on plant-specific simulators that are equipped with the simulation models. Consequently, several models for real-time and off-time simulation of nuclear reactor systems, with various levels of accuracy and simulation fidelity, have been introduced. Experience with power plant simulation demonstrates that in order to realistically predict and simulate reactor responses during unanticipated transients, it is necessary to equip the simulation model with a multielement pressurizer model. The objective of this paper is to present the results of a five-region drift-flux-based pressurizer model, which has been developed for integration with real-time training simulators. A comparison between the plant data and the results of the nonequilibrium pressurizer model demonstrates that the model is well capable of close simulation of dynamic behavior of the pressurizer system
The application of model with lumped parameters for transient condition analyses of NPP
International Nuclear Information System (INIS)
Stankovic, B.; Stevanovic, V.
1985-01-01
The transient behaviour of NPP Krsko during the accident of pressurizer spray valve stuck open has been simulated y lumped parameters model of the PWR coolant system components, developed at the faculty of Mechanical Engineering, University of Belgrade. The elementary volumes which are characterised by the process and state parameters, and by junctions which are characterised by the geometrical and flow parameters are basic structure of physical model. The process parameters obtained by the model RESI, show qualitative agreement with the measured valves, in a degree in which the actions of reactor safety engineered system and emergency core cooling system are adequately modelled; in spite of the elementary physical model structure and only the modelling of thermal process in reactor core and equilibrium conditions of pressurizer and steam generator. The pressurizer pressure and liquid level predicted by the non-equilibrium pressurizer model SOP show good agreement until the HIPS (high pressure pumps) is activated. (author)
Effects of thermohydraulics on clad ballooning, flow blockage and coolability in a LOCA
International Nuclear Information System (INIS)
Erbacher, F.J.; Neitzel, H.J.; Wiehr, K.
1983-01-01
Thermohydraulic boundary conditions have a dominating effect on clad ballooning, flow blockage and coolability: Increasing heat transfer to the fluid decreases the total circumferential strain; Countercurrent flow in a combined injection leads to a relatively small flow blockage; Burst claddings exhibit premature quenching. Differences in the test results obtained in several countries are mainly due to different thermohydraulic test conditions; all test data are consistent with the understanding elaborated within the REBEKA program. Core coolability in a LOCA can be maintained. (author)
Modelling the transient behaviour of pulsed current tungsten-inert-gas weldpools
Wu, C. S.; Zheng, W.; Wu, L.
1999-01-01
A three-dimensional model is established to simulate the pulsed current tungsten-inert-gas (TIG) welding process. The goal is to analyse the cyclic variation of fluid flow and heat transfer in weldpools under periodic arc heat input. To this end, an algorithm, which is capable of handling the transience, nonlinearity, multiphase and strong coupling encountered in this work, is developed. The numerical simulations demonstrate the transient behaviour of weldpools under pulsed current. Experimental data are compared with numerical results to show the effectiveness of the developed model.
Directory of Open Access Journals (Sweden)
Orie Sakamoto
2016-01-01
Full Text Available In remote site power systems with small diesel generators, weak distribution feeders with diesel generators may suffer from voltage and power fluctuations due to misfiring of the engine cylinder. An electromagnetic transient (EMT program named XTAP is considered to be useful to analyze these phenomena. In this study, a new diesel generator model with example fluctuating engine torque has been developed using XTAP for analyses of small power systems with those diesel engines. The configuration and verification results of the developed model are presented in the paper.
Design base transient analysis using the real-time nuclear reactor simulator model
International Nuclear Information System (INIS)
Tien, K.K.; Yakura, S.J.; Morin, J.P.; Gregory, M.V.
1987-01-01
A real-time simulation model has been developed to describe the dynamic response of all major systems in a nuclear process reactor. The model consists of a detailed representation of all hydraulic components in the external coolant circulating loops consisting of piping, valves, pumps and heat exchangers. The reactor core is described by a three-dimensional neutron kinetics model with detailed representation of assembly coolant and moderator thermal hydraulics. The models have been developed to support a real-time training simulator, therefore, they reproduce system parameters characteristic of steady state normal operation with high precision. The system responses for postulated severe transients such as large pipe breaks, loss of pumping power, piping leaks, malfunctions in control rod insertion, and emergency injection of neutron absorber are calculated to be in good agreement with reference safety analyses. Restrictions were imposed by the requirement that the resulting code be able to run in real-time with sufficient spare time to allow interfacing with secondary systems and simulator hardware. Due to hardware set-up and real plant instrumentation, simplifications due to symmetry were not allowed. The resulting code represents a coarse-node engineering model in which the level of detail has been tailored to the available computing power of a present generation super-minicomputer. Results for several significant transients, as calculated by the real-time model, are compared both to actual plant data and to results generated by fine-mesh analysis codes
Design base transient analysis using the real-time nuclear reactor simulator model
International Nuclear Information System (INIS)
Tien, K.K.; Yakura, S.J.; Morin, J.P.; Gregory, M.V.
1987-01-01
A real-time simulation model has been developed to describe the dynamic response of all major systems in a nuclear process reactor. The model consists of a detailed representation of all hydraulic components in the external coolant circulating loops consisting of piping, valves, pumps and heat exchangers. The reactor core is described by a three-dimensional neutron kinetics model with detailed representation of assembly coolant and mode-rator thermal hydraulics. The models have been developed to support a real-time training simulator, therefore, they reproduce system parameters characteristic of steady state normal operation with high precision. The system responses for postulated severe transients such as large pipe breaks, loss of pumping power, piping leaks, malfunctions in control rod insertion, and emergency injection of neutron absorber are calculated to be in good agreement with reference safety analyses. Restrictions were imposed by the requirement that the resulting code be able to run in real-time with sufficient spare time to allow interfacing with secondary systems and simulator hardware. Due to hardware set-up and real plant instrumentation, simplifications due to symmetry were not allowed. The resulting code represents a coarse-node engineering model in which the level of detail has been tailored to the available computing power of a present generation super-minicomputer. Results for several significant transients, as calculated by the real-time model, are compared both to actual plant data and to results generated by fine-mesh analysis codes
Wang, Weicheng
2013-11-01
A chemical kinetic model has been developed for the transient stage of the continuous countercurrent hydrolysis of triglycerides to free fatty acids and glycerol. Departure functions and group contribution methods were applied to determine the equilibrium constants of the four reversible reactions in the kinetic model. Continuous countercurrent hydrolysis of canola oil in subcritical water was conducted experimentally in a lab-scale reactor over a range of temperatures and the concentrations of all neutral components were quantified. Several of the rate constants in the model were obtained by modeling this experimental data, with the remaining determined from calculated equilibrium constants. Some reactions not included in the present, or previous, hydrolysis modeling efforts were identified from glycerolysis kinetic studies and may explain the slight discrepancy between model and experiment. The rate constants determined in this paper indicate that diglycerides in the feedstock accelerate the transition from "emulsive hydrolysis" to "rapid hydrolysis". © 2013 Elsevier Ltd.
RETRAN nonequilibrium two-phase flow model for operational transient analyses
International Nuclear Information System (INIS)
Paulsen, M.P.; Hughes, E.D.
1982-01-01
The field balance equations, flow-field models, and equation of state for a nonequilibrium two-phase flow model for RETRAN are given. The differential field balance model equations are: (1) conservation of mixture mass; (2) conservation of vapor mass; (3) balance of mixture momentum; (4) a dynamic-slip model for the velocity difference; and (5) conservation of mixture energy. The equation of state is formulated such that the liquid phase may be subcooled, saturated, or superheated. The vapor phase is constrained to be at the saturation state. The dynamic-slip model includes wall-to-phase and interphase momentum exchanges. A mechanistic vapor generation model is used to describe vapor production under bulk subcooling conditions. The speed of sound for the mixture under nonequilibrium conditions is obtained from the equation of state formulation. The steady-state and transient solution methods are described
Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling
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.
Integrated tokamak modelling with the fast-ion Fokker–Planck solver adapted for transient analyses
International Nuclear Information System (INIS)
Toma, M; Hamamatsu, K; Hayashi, N; Honda, M; Ide, S
2015-01-01
Integrated tokamak modelling that enables the simulation of an entire discharge period is indispensable for designing advanced tokamak plasmas. For this purpose, we extend the integrated code TOPICS to make it more suitable for transient analyses in the fast-ion part. The fast-ion Fokker–Planck solver is integrated into TOPICS at the same level as the bulk transport solver so that the time evolutions of the fast ion and the bulk plasma are consistent with each other as well as with the equilibrium magnetic field. The fast-ion solver simultaneously handles neutral beam-injected ions and alpha particles. Parallelisation of the fast-ion solver in addition to its computational lightness owing to a dimensional reduction in the phase space enables transient analyses for long periods in the order of tens of seconds. The fast-ion Fokker–Planck calculation is compared and confirmed to be in good agreement with an orbit following a Monte Carlo calculation. The integrated code is applied to ramp-up simulations for JT-60SA and ITER to confirm its capability and effectiveness in transient analyses. In the integrated simulations, the coupled evolution of the fast ions, plasma profiles, and equilibrium magnetic fields are presented. In addition, the electric acceleration effect on fast ions is shown and discussed. (paper)
Development of the MARS input model for Ulchin 1/2 transient analyzer
International Nuclear Information System (INIS)
Jeong, J. J.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Chung, B. D.; Hwang, M.
2003-03-01
KAERI has been developing the NSSS transient analyzer based on best-estimate codes for Ulchin 1/2 plants. The MARS and RETRAN code are used as the best-estimate codes for the NSSS transient analyzer. Among the two codes, the MARS code is to be used for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. This report includes the input model requirements and the calculation note for the Ulchin 1/2 MARS input data generation (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Ulchin 1/2
Development of the MARS input model for Ulchin 3/4 transient analyzer
International Nuclear Information System (INIS)
Jeong, J. J.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Hwang, M. G.
2003-12-01
KAERI has been developing the NSSS transient analyzer based on best-estimate codes.The MARS and RETRAN code are adopted as the best-estimate codes for the NSSS transient analyzer. Among these two codes, the MARS code is to be used for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. This report includes the MARS input model requirements and the calculation note for the MARS input data generation (see the Appendix) for Ulchin 3/4 plant analyzer. In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Ulchin 3/4
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
A model for the calculation of vent clearing transients in pressure suppression systems
International Nuclear Information System (INIS)
Brosche, D.
1975-01-01
For the layout of a pressure suppression system of a light water cooled reactor (boiling water reactor) it is important to know the time dependent behavior of the vent clearing transient after a loss-of-coolant accident for two main reasons: time of the end of the vent clearing transient influences strongly the pressure and temperature maxima in the drywell and wetwell. Time-dependent behavior of the vent clearing transient influences pressure loads in the condensation pool of the wetwell and therefore pressure induced stresses to the structure. The time-dependent behavior of the water masses in the vent pipes and wetwell are described by the basic equations for a nonstationary incompressible friction flow: momentum equation, continuity equation and a correlation for the variation of the state of the gas volume in the wetwell above the water level. After many algebraic operations and integrations along the flow path, a single ordinary nonlinear differential equation for the variations of the water levels in the vent pipes and wetwell is obtained. Therefore the time-dependent velocities and accelerations of the water levels and the moment of the end clearing transient are known. The time-dependent pressure behavior in the drywell, geometrical conditions, initial submergence depth of the vent pipes and different friction and pressure loss factors are presented. The theoretical model has been tested at corresponding experiments performed at a full scale 1/48 segment of the Humboldt Bay pressure suppression containment in the USA and at the pressure suppression containment at the Marviken nuclear power station in Sweden. All these comparisons have shown good agreement between theory and experiment
Energy Technology Data Exchange (ETDEWEB)
Belem, J A.T.
1993-09-01
The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs.
A transient multi-scale model for direct methanol fuel cells
International Nuclear Information System (INIS)
Jahnke, T.; Zago, M.; Casalegno, A.; Bessler, W.G.; Latz, A.
2017-01-01
The DMFC is a promising option for backup power systems and for the power supply of portable devices. However, from the modeling point of view liquid-feed DMFC are challenging systems due to the complex electrochemistry, the inherent two-phase transport and the effect of methanol crossover. In this paper we present a physical 1D cell model to describe the relevant processes for DMFC performance ranging from electrochemistry on the surface of the catalyst up to transport on the cell level. A two-phase flow model is implemented describing the transport in gas diffusion layer and catalyst layer at the anode side. Electrochemistry is described by elementary steps for the reactions occurring at anode and cathode, including adsorbed intermediate species on the platinum and ruthenium surfaces. Furthermore, a detailed membrane model including methanol crossover is employed. The model is validated using polarization curves, methanol crossover measurements and impedance spectra. It permits to analyze both steady-state and transient behavior with a high level of predictive capabilities. Steady-state simulations are used to investigate the open circuit voltage as well as the overpotentials of anode, cathode and electrolyte. Finally, the transient behavior after current interruption is studied in detail.
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
DEFF Research Database (Denmark)
Yang, Jingwei; Zhang, Ning; Kang, Chongqing
2017-01-01
The rapid deployment of gas-fired generating units makes the power system more vulnerable to failures in the natural gas system. To reduce the risk of gas system failure and to guarantee the security of power system operation, it is necessary to take the security constraints of natural gas...... accurately, they are hard to be embedded into the power system scheduling model, which consists of algebraic equations and inequations. This paper addresses this dilemma by proposing an algebraic transient model of natural gas network which is similar to the branch-node model of power network. Based...... pipelines into account in the day-ahead power generation scheduling model. However, the minute- and hour-level dynamic characteristics of gas systems prevents an accurate decision-making simply with the steady-state gas flow model. Although the partial differential equations depict the dynamics of gas flow...
International Nuclear Information System (INIS)
Ho, J.C.
2004-01-01
Among those theories to interpret the PWR iodine spiking behaviors, the most accepted concept is based on steam formation and condensation in damaged fuel rods. Due to the complex nature of the phenomenon, a comprehensive model of the iodine behavior has not yet been successfully developed. In 1992 a new empirical model was introduced to establish a correlation with the operating parameters. The comparison results of the predicted iodine-131 equivalent activity value with the operating radiochemistry database was off by 23%. This paper presents an improved model. Although it is still an empirical model which also gives a first order estimation of the peak iodine spiking magnitude, the deviation between prediction and measurement was reduced to ∼7%. It is believed that this improved model can be used for better prediction and control of the iodine spiking magnitude resulted from failed fuel rods during power transients or plant shutdown. (author)
Energy Technology Data Exchange (ETDEWEB)
Vazquez, M.; Martin-Fuertes, F.
2013-07-01
As a complement to the attached code 3D neutron-CIEMAT thermohydraulic added a module to simulate transient. Temporary kinetics is resolved by factoring flow in a spatial part and another storm. MCNP provides the reactivity and updated spatial function and COBRA-IV calculates the temperature distribution. Temporary dependence of amplitude is calculated using time delayed neutron Kinetic equations. As an example of application, examines a transient loss of flow in MYRRHA, a lead-cooled experimental reactor.
Transient non-isothermal model of a polymer electrolyte fuel cell
Energy Technology Data Exchange (ETDEWEB)
Shah, A.A. [Queen' s-RMC Fuel Cell Research Centre, 945 Princess Street, Kingston, Ont. K7L 5L9 (Canada); Kim, G.-S.; Harvey, D. [Ballard Power Systems, 4343 North Fraser Way, Burnaby, BC V5J 5J9 (Canada); Sui, P.C. [Institute for Integrated Energy Systems, University of Victoria, Victoria, BC V8W 3P6 (Canada)
2007-01-01
In this paper we present a one-dimensional transient model for the membrane electrode assembly of a polymer-electrolyte fuel cell. In earlier work we established a framework to describe the water balance in a steady-state, non-isothermal cathode model that explicitly included an agglomerate catalyst layer component. This paper extends that work in several directions, explicitly incorporating components of the anode, including a micro-porous layer, and accounting for electronic potential variations, gas convection and time dependance. The inclusion of temperature effects, which are vital to the correct description of condensation and evaporation, is new to transient modelling. Several examples of the modelling results are given in the form of potentiostatic sweeps and compared to experimental results. Excellent qualitative agreement is demonstrated, particularly in regard to the phenomenon of hysteresis, a manifestation of the sensitive response of the system to the presence of water. Results pertaining to pore size, contact angle and the presence of a micro-porous layer are presented and future work is discussed. (author)
Directory of Open Access Journals (Sweden)
Oramus Piotr
2015-09-01
Full Text Available Electric arc is a complex phenomenon occurring during the current interruption process in the power system. Therefore performing digital simulations is often necessary to analyse transient conditions in power system during switching operations. This paper deals with the electric arc modelling and its implementation in simulation software for transient analyses during switching conditions in power system. Cassie, Cassie-Mayr as well as Schwarz-Avdonin equations describing the behaviour of the electric arc during the current interruption process have been implemented in EMTP-ATP simulation software and presented in this paper. The models developed have been used for transient simulations to analyse impact of the particular model and its parameters on Transient Recovery Voltage in different switching scenarios: during shunt reactor switching-off as well as during capacitor bank current switching-off. The selected simulation cases represent typical practical scenarios for inductive and capacitive currents breaking, respectively.
International Nuclear Information System (INIS)
Alekseev, A.; Arslanova, D.; Belov, A.; Belyakov, V.; Gapionok, E.; Gornikel, I.; Gribov, Y.; Ioki, K.; Kukhtin, V.; Lamzin, E.; Sugihara, M.; Sychevsky, S.; Terasawa, A.; Utin, Y.
2013-01-01
A set of detailed computational models are reviewed that covers integrally the system “vacuum vessel (VV), cryostat, and thermal shields (TS)” to study transient electromagnetics (EMs) in the ITER machine. The models have been developed in the course of activities requested and supervised by the ITER Organization. EM analysis is enabled for all ITER operational scenarios. The input data are derived from results of DINA code simulations. The external EM fields are modeled accurate to the input data description. The known magnetic shell approach can be effectively applied to simulate thin-walled structures of the ITER machine. Using an integral–differential formulation, a single unknown is determined within the shells in terms of the vector electric potential taken only at the nodes of a finite-element (FE) mesh of the conducting structures. As a result, the FE mesh encompasses only the system “VV + Cryostat + TS”. The 3D model requires much higher computational resources as compared to a shell model based on the equivalent approximation. The shell models have been developed for all principal conducting structures in the system “VV + Cryostat + TS” including regular ports and neutral beam ports. The structures are described in details in accordance with the latest design. The models have also been applied for simulations of EM transients in components of diagnostic systems and cryopumps and estimation of the 3D effects of the ITER structures on the plasma performance. The developed models have been elaborated and applied for the last 15 years to support the ITER design activities. The finalization of the ITER VV design enables this set of models to be considered ready to use in plasma-physics computations and the development of ITER simulators
Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark
Energy Technology Data Exchange (ETDEWEB)
Elina Syrjaelahti; Anitta Haemaelaeinen [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)
2005-07-01
Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)
Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark
International Nuclear Information System (INIS)
Elina Syrjaelahti; Anitta Haemaelaeinen
2005-01-01
Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)
International Nuclear Information System (INIS)
Limnios, N.
1983-01-01
The APACHE code (Automatic Analysis of Failures of Hydraulic and Thermohydraulic Circuits more particularly of Water) situates in an important program of computer codes development in the field of studies on reliability and safety of systems in nuclear power plants. APACHE is an automatic generation code of failure pattern and of their effects. After a presentation of the theoretical basis, the methodological principles of the theory of networks are developed. Then, the model of the code is developed: model of individual behavior of each classical model component of normal behavior and model of failure pattern with specifications. The global model of hydraulic systems and the resolution systems are then developed. More particularly, some aspects of the theory of graphs, and the algorithms developed for the automatic construction of the equation systems and especially the algorithm of the research of meshes are presented. The computer aspect of the code and the programming of the code with its limits and some specifications are described. The practical aspect of utilization is finally presented [fr
International Nuclear Information System (INIS)
Wang Jianyu; Zhang Shenru; Min Yuanyou
1994-01-01
To estimate the performance of the passive containment cooling system (PCCS) of the AC-600 nuclear power plant, the PCCSAC code is developed currently by the jointed efforts between Tsinghua University and NPIC. Different features on the passive behavior of the system and the main components of the containment are considered in the code which is needed by the further AC-600 R and D Program. With a brief description of the AC-600 passive containment cooling system and components, the main thermohydraulic models and numerical scheme used in the PCCSAC code are introduced and the selected results of the verification and the prediction for the performance of the AC-600 passive containment cooling system under LOCA and a steam line break accident are presented to preliminarily demonstrate the applicability and reliability of the PCCSAC model. The current PCCSAC model is conservative and a further 2-D PCCSAC version is under consideration in addition to provide the database for models from some tests associated with the components and systems unique to AC-600 nuclear power plant to meet the requirement of the more realistic modelization for the AC-600 passive containment cooling system. (author)
Comparison of empirical transport models with transient transport experiments in LHD
International Nuclear Information System (INIS)
Yakovlev, Mikhail; Inagaki, Shigeru; Ida, Katsumi
2004-01-01
A study of the electron transport in helical plasma of Large Helical Device (LHD) has been performed using a perturbation to an equilibrium state. The periodic perturbation in plasma is induced by on-axis Electron Cyclotron Heating (ECH) modulated signal for different temperatures of plasma electron. The experimental data are compared with results from simulation within framework of the diffusive model with additional convective term. The convection heat flux is introduced to describe the heat propagation in LHD. It has been shown that the dynamic plasma heat diffusivity coefficient χ e estimated from the transient analysis becomes larger with increasing electron temperature in LHD plasma. (author)
Application of transient ignition model to multi-canister (MCO) accident analysis
International Nuclear Information System (INIS)
Kummerer, M.
1996-01-01
The potential for ignition of spent nuclear fuel in a Multi-Canister Overpack (MCO) is examined. A transient model is applied to calculate the highest ambient gas temperature outside an MCO wall tube or shipping cask for which a stable temperature condition exists. This integral analysis couples reaction kinetics with a description of the MCO configuration, heat and mass transfer, and fission product phenomena. It thereby allows ignition theory to be applied to various complex scenarios, including MCO water loss accidents and dry MCO air ingression
Modeling startup and shutdown transient of the microlinear piezo drive via ANSYS
Azin, A. V.; Bogdanov, E. P.; Rikkonen, S. V.; Ponomarev, S. V.; Khramtsov, A. M.
2017-02-01
The article describes the construction-design of the micro linear piezo drive intended for a peripheral cord tensioner in the reflecting surface shape regulator system for large-sized transformable spacecraft antenna reflectors. The research target -the development method of modeling startup and shutdown transient of the micro linear piezo drive. This method is based on application software package ANSYS. The method embraces a detailed description of the calculation stages to determine the operating characteristics of the designed piezo drive. Based on the numerical solutions, the time characteristics of the designed piezo drive are determined.
DEFF Research Database (Denmark)
Zhu, Huayang; Ricote, Sandrine; Coors, W. Grover
2014-01-01
the computational implementation of a Nernst–Planck–Poisson (NPP) model to represent and interpret conductivity-relaxation measurements. Defect surface chemistry is represented with both equilibrium and finite-rate kinetic models. The experiments and the models are capable of representing relaxations from strongly......A model-based approach is used to interpret equilibrium and transient conductivity measurements for 10% gadolinium-doped ceria: Ce0.9Gd0.1O1.95 − δ (GDC10). The measurements were carried out by AC impedance spectroscopy on slender extruded GDC10 rods. Although equilibrium conductivity measurements...... provide sufficient information from which to derive material properties, it is found that uniquely establishing properties is difficult. Augmenting equilibrium measurements with conductivity relaxation significantly improves the evaluation of needed physical properties. This paper develops and applies...
Directory of Open Access Journals (Sweden)
Babak. B. Naghshine
2017-02-01
Full Text Available Laser processing is one of the most popular small-scale patterning methods and has many applications in semiconductor device fabrication and biomedical engineering. Numerical modelling of this process can be used for better understanding of the process, optimization, and predicting the quality of the final product. An accurate 3D model is presented here for short laser pulses that can predict the ablation depth and temperature distribution on any section of the material in a minimal amount of time. In this transient model, variations of thermal properties, plasma shielding, and phase change are considered. Ablation depth was measured using a 3D optical profiler. Calculated depths are in good agreement with measured values on laser treated titanium surfaces. The proposed model can be applied to a wide range of materials and laser systems.
International Nuclear Information System (INIS)
Grundmann, U.; Kliem, S.
2003-12-01
various calculations for phase II. So, differences of the power and void distribution in single fuel assemblies can be explained by the differences in the neutron kinetic and thermo-hydraulic models. Comparisons between ATHLET/DYN3D (parallel coupling) and ATHLET/QUABOX-CUBBOX (internal coupling) show only small deviations for spatially averaged parameters. The deviations in the local parameters can be explained mainly by the differences in the modeling of the reactor core (reduced number of coolant channels, disregard of the ADF and a different two-phase model in the calculations with ATHLET/QUABOX-CUBBOX). The calculations for the extreme scenarios in phase III demonstrate the applicability of the coupled code ATHLET/DYN3D to transients with conditions beyond the experiment. (orig.) [de
Development of an Aeroelastic Modeling Capability for Transient Nozzle Side Load Analysis
Wang, Ten-See; Zhao, Xiang; Zhang, Sijun; Chen, Yen-Sen
2013-01-01
Lateral nozzle forces are known to cause severe structural damage to any new rocket engine in development during test. While three-dimensional, transient, turbulent, chemically reacting computational fluid dynamics methodology has been demonstrated to capture major side load physics with rigid nozzles, hot-fire tests often show nozzle structure deformation during major side load events, leading to structural damages if structural strengthening measures were not taken. The modeling picture is incomplete without the capability to address the two-way responses between the structure and fluid. The objective of this study is to develop a coupled aeroelastic modeling capability by implementing the necessary structural dynamics component into an anchored computational fluid dynamics methodology. The computational fluid dynamics component is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, while the computational structural dynamics component is developed in the framework of modal analysis. Transient aeroelastic nozzle startup analyses of the Block I Space Shuttle Main Engine at sea level were performed. The computed results from the aeroelastic nozzle modeling are presented.
International Nuclear Information System (INIS)
Kim, Jong Tae; Kim, Sang Baik; Kim, Hee Dong
2002-03-01
LILAC-meltpool has been developed to study thermo-hydraulic behavior of molten pool and thermal behavior of vessel wall during severe accident. To validate LILAC-meltpool code several two and three dimensional thermo-hydraulic problems were selected and solved. The benchmark problems have experimental results or verified numerical results. Through the validation it was found that LILAC-meltpool reproduces very accurate numerical results. Two-layered semicircular pool was solved to study thermal and hydraulic characteristics of pool stratification. The LAVA experiment using alumina/ferrite molten pool was calculated and compared with computed results. Cooling of alumina/ferrite two-layered pool was affected by stratification. In the numerical results temperature of vessel inner was highest at a location below the interface. Crust was developed from upper surface and lower outer surface, but in the area near the interface corium simulant existed as molten state for long time. LAVA-4 experiment was studied using gap-cooling model in LILAC-meltpool code. Temperature increase of LAVA vessel after alumina melt relocation was strongly dependent on gap formation mechanism. Calculated cooling rates of the vessel were very similar to experimental results. For LAVA experiments which do not have heat generation coolant penetrates easily into a gap and it is found that gap-cooling is very effective for cooling of vessel, but it is thought that coolant penetration could be limited near upper part of gap because of decay heat and high temperature of corium crust
Modeling Transients and Designing a Passive Safety System for a Nuclear Thermal Rocket Using Relap5
Khatry, Jivan
Long-term high payload missions necessitate the need for nuclear space propulsion. Several nuclear reactor types were investigated by the Nuclear Engine for Rocket Vehicle Application (NERVA) program of National Aeronautics and Space Administration (NASA). Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. A NERVA design known as the Pewee I was selected for this purpose. The following transients were run: (i) modeling of corrosion-induced blockages on the peripheral fuel element coolant channels and their impact on radiation heat transfer in the core, and (ii) modeling of loss-of-flow-accidents (LOFAs) and their impact on radiation heat transfer in the core. For part (i), the radiation heat transfer rate of blocked channels increases while their neighbors' decreases. For part (ii), the core radiation heat transfer rate increases while the flow rate through the rocket system is decreased. However, the radiation heat transfer decreased while there was a complete LOFA. In this situation, the peripheral fuel element coolant channels handle the majority of the radiation heat transfer. Recognizing the LOFA as the most severe design basis accident, a passive safety system was designed in order to respond to such a transient. This design utilizes the already existing tie rod tubes and connects them to a radiator in a closed loop. Hence, this is basically a secondary loop. The size of the core is unchanged. During normal steady-state operation, this secondary loop keeps the moderator cool. Results show that the safety system is able to remove the decay heat and prevent the fuel elements from melting, in response to a LOFA and subsequent SCRAM.
International Nuclear Information System (INIS)
Stoller, R.E.; Grossbeck, M.L.; Mansur, L.K.
1990-01-01
A theoretical model has been developed using the reaction rate theory of radiation effects to explain experimental results that showed higher than expected values of irradiation creep at low temperatures in the Oak Ridge Research Reactor. The customary assumption that the point defect concentrations are at steady state was not made; rather, the time dependence of the vacancy and interstitial concentrations and the creep rate were explicitly calculated. For temperatures below about 100 to 200 degree C, the time required for the vacancy concentration to reach steady state exceeds the duration of the experiment. For example, if materials parameters typical of austenitic stainless steel are used, the calculated vacancy transient dose at 100 degree C is about 100 dpa. At 550 degree C this transient is over by 10 -8 dpa. During the time that the vacancy population remains lower than its steady state value, dislocation climb is increased since defects of primarily one type are being absorbed. Using the time-dependent point defect concentrations, the dislocation climb velocity has been calculated as a function of time and a climb-enabled glide creep model had been invoked. The extended transient time for the vacancies leads to high creep rates at low temperatures. In agreement with the experimental observations, a minimum in the temperature dependence of creep is predicted at a temperature between 50 and 350 degree C. The temperature at which the minimum occurs decreases as the irradiation dose increases. Predicted values of creep at 8 dpa are in good agreement with the results of the ORR-MFE-6J/7J experiment
Directory of Open Access Journals (Sweden)
Itissam ABUIZIAH
2014-03-01
Full Text Available When transient conditions (water hammer exist, the life expectancy of the system can be adversely impacted, resulting in pump and valve failures and catastrophic pipe rupture. Hence, transient control has become an essential requirement for ensuring safe operation of water pipeline systems. To protect the pipeline systems from transient effects, an accurate analysis and suitable protection devices should be used. This paper presents the problem of modeling and simulation of transient phenomena in hydraulic systems based on the characteristics method. Also, it provides the influence of using the protection devices to control the adverse effects due to excessive and low pressure occuring in the transient. We applied this model for two main pipeline systems: Valve and pump combined with a simple surge tank connected to reservoir. The results obtained by using this model indicate that the model is an efficient tool for water hammer analysis. Moreover, using a simple surge tank reduces the unfavorable effects of transients by reducing pressure fluctuations.
Directory of Open Access Journals (Sweden)
Andrzej Rusek
2008-01-01
Full Text Available The mathematical model of cylindrical linear induction motor (C-LIM fed via frequency converter is presented in the paper. The model was developed in order to analyze numerically the transient states. Problems concerning dynamics of ac-machines especially linear induction motor are presented in [1 – 7]. Development of C-LIM mathematical model is based on circuit method and analogy to rotary induction motor. The analogy between both: (a stator and rotor windings of rotary induction motor and (b winding of primary part of C-LIM (inductor and closed current circuits in external secondary part of C-LIM (race is taken into consideration. The equations of C-LIM mathematical model are presented as matrix together with equations expressing each vector separately. A computational analysis of selected transient states of C-LIM fed via frequency converter is presented in the paper. Two typical examples of C-LIM operation are considered for the analysis: (a starting the motor at various static loads and various synchronous velocities and (b reverse of the motor at the same operation conditions. Results of simulation are presented as transient responses including transient electromagnetic force, transient linear velocity and transient phase current.
Extended probit mortality model for zooplankton against transient change of PCO(2).
Sato, Toru; Watanabe, Yuji; Toyota, Koji; Ishizaka, Joji
2005-09-01
The direct injection of CO(2) in the deep ocean is a promising way to mitigate global warming. One of the uncertainties in this method, however, is its impact on marine organisms in the near field. Since the concentration of CO(2), which organisms experience in the ocean, changes with time, it is required to develop a biological impact model for the organisms against the unsteady change of CO(2) concentration. In general, the LC(50) concept is widely applied for testing a toxic agent for the acute mortality. Here, we regard the probit-transformed mortality as a linear function not only of the concentration of CO(2) but also of exposure time. A simple mathematical transform of the function gives a damage-accumulation mortality model for zooplankton. In this article, this model was validated by the mortality test of Metamphiascopsis hirsutus against the transient change of CO(2) concentration.
Energy Technology Data Exchange (ETDEWEB)
Karaagac, U.; Saad, H.; Mahseredjian, J. [Ecole Polytechnique de Montreal, Montreal, QC (Canada); Jensen, S.; Cai, L. [REpower Systems AG, Hamburg (Germany)
2012-07-01
The large number of switching elements in the modular multilevel converter (MMC) is a challenging problem for modeling the MMC-HVDC systems in electromagnetic transient type (EMT-type) programs. The modeling complexity increases even further when MMC-HVDC systems are used to integrate offshore wind farms (OWFs) with power electronics based wind energy converters, such as doubly-fed induction generators (DFIGs). This paper compares the computational performances of various combinations of MMC-HVDC and OWF models. Practical onshore ac fault scenarios are simulated for an OWF composed of DFIG type wind turbines and connected to a practical ac grid through a point-to-point MMC-HVDC system. (orig.)
SR calcium handling and calcium after-transients in a rabbit model of heart failure
Baartscheer, Antonius; Schumacher, Cees A.; Belterman, Charly N. W.; Coronel, Ruben; Fiolet, Jan W. T.
2003-01-01
Objective: After-depolarization associated arrhythmias are frequently observed in heart failure and associated with spontaneous calcium release from sarcoplasmic reticulum (SR), calcium after-transients. We hypothesize that disturbed SR calcium handling underlies calcium after-transients in heart
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)
Modelling of phenomena associated with high burnup fuel behaviour during overpower transients
International Nuclear Information System (INIS)
Sills, H.E.; Langman, V.J.; Iglesias, F.C.
1995-01-01
Phenomena of importance to the behaviour of high burnup fuel subjected to conditions of rapid overpower (i.e., LWR RIAs) include the change in cladding material properties due to irradiation, pellet-clad interaction (PCI) and 'rim' effects associated with the periphery of high burnup fuel. 'Rim' effects are postulated to be caused by changes in fuel morphology at high burnup. Typical discharge burnups for CANDU fuel are low compared to LWRs. Maximum linear ratings for CANDU fuel are higher than those for LWRs. However, under normal operating conditions, the Zircaloy-4 clad of the CANDU fuel is collapsed onto the fuel stack. Thus, the CANDU fuel performance codes model the transient behaviour of the fuel-to-clad interface and are capable of assessing the potential for pellet-clad mechanical interaction (PCMI) failures for a wide range of overpower conditions. This report provides a discussion of the modelling of the phenomena of importance to high burnup fuel behaviour during rapid overpower transients. (author)
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)
Equilibrium and kinetic models for colloid release under transient solution chemistry conditions.
Bradford, Scott A; Torkzaban, Saeed; Leij, Feike; Simunek, Jiri
2015-10-01
We present continuum models to describe colloid release in the subsurface during transient physicochemical conditions. Our modeling approach relates the amount of colloid release to changes in the fraction of the solid surface area that contributes to retention. Equilibrium, kinetic, equilibrium and kinetic, and two-site kinetic models were developed to describe various rates of colloid release. These models were subsequently applied to experimental colloid release datasets to investigate the influence of variations in ionic strength (IS), pH, cation exchange, colloid size, and water velocity on release. Various combinations of equilibrium and/or kinetic release models were needed to describe the experimental data depending on the transient conditions and colloid type. Release of Escherichia coli D21g was promoted by a decrease in solution IS and an increase in pH, similar to expected trends for a reduction in the secondary minimum and nanoscale chemical heterogeneity. The retention and release of 20nm carboxyl modified latex nanoparticles (NPs) were demonstrated to be more sensitive to the presence of Ca(2+) than D21g. Specifically, retention of NPs was greater than D21g in the presence of 2mM CaCl2 solution, and release of NPs only occurred after exchange of Ca(2+) by Na(+) and then a reduction in the solution IS. These findings highlight the limitations of conventional interaction energy calculations to describe colloid retention and release, and point to the need to consider other interactions (e.g., Born, steric, and/or hydration forces) and/or nanoscale heterogeneity. Temporal changes in the water velocity did not have a large influence on the release of D21g for the examined conditions. This insensitivity was likely due to factors that reduce the applied hydrodynamic torque and/or increase the resisting adhesive torque; e.g., macroscopic roughness and grain-grain contacts. Our analysis and models improve our understanding and ability to describe the amounts
Thermohydraulic design of saturated temperature capsule for IASCC irradiation test
International Nuclear Information System (INIS)
Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio
2002-10-01
An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)
Thermohydraulic design of saturated temperature capsule for IASCC irradiation test
Energy Technology Data Exchange (ETDEWEB)
Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others
2002-10-01
An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)
Latest developments for a computer aided thermohydraulic network
International Nuclear Information System (INIS)
Alemberti, A.; Graziosi, G.; Mini, G.; Susco, M.
1999-01-01
Thermohydraulic networks are I-D systems characterized by a small number of basic components (pumps, valves, heat exchangers, etc) connected by pipes and limited spatially by a defined number of boundary conditions (tanks, atmosphere, etc). The network system is simulated by the well known computer program RELAPS/mod3. Information concerning the network geometry component behaviour, initial and boundary conditions are usually supplied to the RELAPS code using an ASCII input file by means of 'input cards'. CATNET (Computer Aided Thermalhydraulic NETwork) is a graphically user interface that, under specific user guidelines which completely define its range of applicability, permits a very high level of standardization and simplification of the RELAPS/mod3 input deck development process as well as of the output processing. The characteristics of the components (pipes, valves, pumps etc), defining the network system can be entered through CATNET. The CATNET interface is provided by special functions to compute form losses in the most typical bending and branching configurations. When the input of all system components is ready, CATNET is able to generate the RELAPS/mod3 input file. Finally, by means of CATNET, the RELAPS/mod3 code can be run and its output results can be transformed to an intuitive display form. The paper presents an example of application of the CATNET interface as well as the latest developments which greatly simplified the work of the users and allowed to reduce the possibility of input errors. (authors)
Thermo-hydraulic consequence of pressure suppression containment vessel during blowdown, 2
International Nuclear Information System (INIS)
Aya, Izuo; Nariai, Hideki; Kobayashi, Michiyuki
1980-01-01
As a part of the safety research works for the integral-type marine reactor, an analytical code SUPPAC-2V was developed to simulate the thermo-hydraulic consequence of a pressure suppression containment system during blowdown and the code was applied to the Model Experimental Facility of the Safety of Integral Type Marine Reactors (explained already in Part 1). SUPPAC-2V is much different from existing codes in the following points. A nonhomogeneous model for the gaseous region in the drywell, a new correlation for condensing heat transfer coefficient at drywell wall based on existing data and approximation of air bubbles in wetwell water by one dimensional bubble rising model are adopted in this code. In comparing calculational results with experimental results, values of predominant input parameters were evaluated and discussed. Moreover, the new code was applied also to the NSR-7 marine reactor, conceptually designed at the Shipbuilding Research Association in Japan, of which suppression system had been already analysed by CONTEMPT-PS. (author)
Variable thickness transient ground-water flow model. Volume 3. Program listings
International Nuclear Information System (INIS)
Reisenauer, A.E.
1979-12-01
The Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) Program is developing and applying the methodology for assessing the far-field, long-term post-closure safety of deep geologic nuclear waste repositories. AEGIS is being performed by Pacific Northwest Laboratory (PNL) under contract with the Office of Nuclear Waste Isolation (OWNI) for the Department of Energy (DOE). One task within AEGIS is the development of methodology for analysis of the consequences (water pathway) from loss of repository containment as defined by various release scenarios. Analysis of the long-term, far-field consequences of release scenarios requires the application of numerical codes which simulate the hydrologic systems, model the transport of released radionuclides through the hydrologic systems to the biosphere, and, where applicable, assess the radiological dose to humans. Hydrologic and transport models are available at several levels of complexity or sophistication. Model selection and use are determined by the quantity and quality of input data. Model development under AEGIS and related programs provides three levels of hydrologic models, two levels of transport models, and one level of dose models (with several separate models). This is the third of 3 volumes of the description of the VTT (Variable Thickness Transient) Groundwater Hydrologic Model - second level (intermediate complexity) two-dimensional saturated groundwater flow
Directory of Open Access Journals (Sweden)
Mojtaba Darabi
2016-06-01
Full Text Available Considering the fact that a large volume of iron reserve in the Sechahoon Iron Mine in Yazd Province has located under the water table, it is necessary to conduct a comprehensive study on water flow within the pit and its surroundings. The conceptual model of the aquifer was created using surface and underground geological information compared with water table data of the area of interest. In the data preparation stages, in order to create the numerical model, Logan and Lufran tests were studied to determine the hydrodynamic coefficients of the layers, precipitation and evaporation were investigated, and fractures and faults of the region, as a medium for flow channels in the hard formation, were also studied. The model was created in a transient state between 2000 and 2014. To validate its results, the water table was measured 4 times in the last 4 months of 2014. Considering the complexities in the heterogeneous fractured aquifer of the study area, numerical modeling results for the basin in a transient state present 90 percent correlation with field studies. Having investigated the water balance in the region, the boundary condition of the model was determined as the input water from the eastern south and the runoff water in the western north of the region. Since the general trend of faults in the area is north-south, variation in the water table is slight on north-south and intense on the east-west direction. On the other hand, due to the fact that the maximum flow is along the faults and fractures, the water table contour lines in different locations over the region are closed.
Thermohydraulics of emergency core cooling in light water reactors
International Nuclear Information System (INIS)
1989-10-01
This report, by a group of experts of the OECD-NEA Committee on the Safety of Nuclear Installations, reviews the current state-of-knowledge in the field of emergency core cooling (ECC) for design-basis, loss-of-coolant accidents (LOCA) and core uncover transients in pressurized- and boiling-water reactors. An overview of the LOCA scenarios and ECC phenomenology is provided for each type of reactor, together with a brief description of their ECC systems. Separate-effects and integral-test facilities, which contribute to understanding and assessing the phenomenology, are reviewed together with similarity and scaling compromises. All relevant LOCA phenomena are then brought together in the form of tables. Each phenomenon is weighted in terms of its importance to the course of a LOCA, and appraised for the adequacy of its data base and analytical modelling. This qualitative procedure focusses attention on the modelling requirements of dominant LOCA phenomena and the current capabilities of the two-fluid models in two-phase flows. This leads into the key issue with ECC: quantitative code assessment and the application of system codes to predict with a well defined uncertainty the behaviour of a nuclear power plant. This issue, the methodologies being developed for code assessment and the question of how good is good enough are discussed in detail. Some general conclusions and recommendations for future research activities are provided
Busy period analysis, rare events and transient behavior in fluid flow models
Directory of Open Access Journals (Sweden)
Søren Asmussen
1994-01-01
Full Text Available We consider a process {(Jt,Vt}t≥0 on E×[0,∞, such that {Jt} is a Markov process with finite state space E, and {Vt} has a linear drift ri on intervals where Jt=i and reflection at 0. Such a process arises as a fluid flow model of current interest in telecommunications engineering for the purpose of modeling ATM technology. We compute the mean of the busy period and related first passage times, show that the probability of buffer overflow within a busy cycle is approximately exponential, and give conditioned limit theorems for the busy cycle with implications for quick simulation. Further, various inequalities and approximations for transient behavior are given. Also explicit expressions for the Laplace transform of the busy period are found. Mathematically, the key tool is first passage probabilities and exponential change of measure for Markov additive processes.
International Nuclear Information System (INIS)
Chang Tiejun; Tian Mingzhen; Randall Babbitt, Wm.
2004-01-01
We present a theoretical model for optical coherent transient (OCT) processes based on Maxwell-Bloch equations for angled beam geometry. This geometry is critical in various OCT applications where the desired coherence outputs need to be spatially separated from the rest of the field. The model takes into account both the local interactions between inhomogeneously broadened two-level atoms and the laser fields, and the field propagation in optically thick media. Under the small-angle condition, the spatial dimensions transversing to the main propagation direction were treated with spatial Fourier transform to make the numerical computations for the practical settings confined within a reasonable time frame. The simulations for analog correlators and continuous processing based on stimulated photon echo have been performed using the simulator developed using the theory
Improved pump turbine transient behaviour prediction using a Thoma number-dependent hillchart model
International Nuclear Information System (INIS)
Manderla, M; Koutnik, J; Kiniger, K
2014-01-01
Water hammer phenomena are important issues for high head hydro power plants. Especially, if several reversible pump-turbines are connected to the same waterways there may be strong interactions between the hydraulic machines. The prediction and coverage of all relevant load cases is challenging and difficult using classical simulation models. On the basis of a recent pump-storage project, dynamic measurements motivate an improved modeling approach making use of the Thoma number dependency of the actual turbine behaviour. The proposed approach is validated for several transient scenarios and turns out to increase correlation between measurement and simulation results significantly. By applying a fully automated simulation procedure broad operating ranges can be covered which provides a consistent insight into critical load case scenarios. This finally allows the optimization of the closing strategy and hence the overall power plant performance
Improved pump turbine transient behaviour prediction using a Thoma number-dependent hillchart model
Manderla, M.; Kiniger, K.; Koutnik, J.
2014-03-01
Water hammer phenomena are important issues for high head hydro power plants. Especially, if several reversible pump-turbines are connected to the same waterways there may be strong interactions between the hydraulic machines. The prediction and coverage of all relevant load cases is challenging and difficult using classical simulation models. On the basis of a recent pump-storage project, dynamic measurements motivate an improved modeling approach making use of the Thoma number dependency of the actual turbine behaviour. The proposed approach is validated for several transient scenarios and turns out to increase correlation between measurement and simulation results significantly. By applying a fully automated simulation procedure broad operating ranges can be covered which provides a consistent insight into critical load case scenarios. This finally allows the optimization of the closing strategy and hence the overall power plant performance.
ERP-IV-A program for transient thermal-hydraulic analysis of PWR plant
International Nuclear Information System (INIS)
Dai Anguo; Tang Jiahuan; Qian Huifu; Gao Zhikang
1987-12-01
The author deal with the descriptions of physical model of transient process in PWR plant and the function of ERP-IV (ERR-IV Transient Thermo-Hydraulic Analysis Code). The code has been developed for safety analysis and design transient. The code is characterized by the multi-loop long-term, short term, wide-range plant simulation with the capability to analyze natural circulation condition. The description of ERP-IV includes following parts: reactor, primary coolant loops, pressurizer, steam generators, main steam system, turbine, feedwater system, steam dump, relive valves, and safety valves in secondary side, etc.. The code can use for accident analysis, such as loss of all A.C. power to power plant auxiliaries (a station blackout), loss of normal feedwater, loss of load, loss of condenser vacuum and other events causing a turbine trip, complete loss of forced reactor coolant flow, uncontrolled rod cluster control assembly bank withdrawal. It can also be used for accident analysis of the emergency and limiting conditions, such as feedwater line break and main steam line rupture. It can also be utilized as a tool for system design studies, component design, setpoint studies and design transition studies, etc
Thermo-hydraulic analysis of the cool-down of the EDIPO test facility
Lewandowska, Monika; Bagnasco, Maurizio
2011-09-01
The first cool-down of the EDIPO (European DIPOle) test facility is foreseen to take place in 2011 by means of the existing 1.2 kW cryoplant at EPFL-CRPP Villigen. In this work, the thermo-hydraulic analysis of the EDIPO cool-down is performed in order both to assess the its duration and to optimize the procedure. The cool-down is driven by the helium flowing in both the outer cooling channel and in the windings connected hydraulically in parallel. We take into account limitations due to the pressure drop in the cooling circuit and the refrigerator capacity as well as heat conduction in the iron yoke. Two schemes of the hydraulic cooling circuit in the EDIPO windings are studied (coils connected in series and coils connected in parallel). The analysis is performed by means of an analytical model complemented by and numerical model. The results indicate that the cool-down to 5 K can be achieved in about 12 days.
Ocean Heat and Carbon Uptake in Transient Climate Change: Identifying Model Uncertainty
Romanou, Anastasia; Marshall, John
2015-01-01
Global warming on decadal and centennial timescales is mediated and ameliorated by the oceansequestering heat and carbon into its interior. Transient climate change is a function of the efficiency by whichanthropogenic heat and carbon are transported away from the surface into the ocean interior (Hansen et al. 1985).Gregory and Mitchell (1997) and Raper et al. (2002) were the first to identify the importance of the ocean heat uptakeefficiency in transient climate change. Observational estimates (Schwartz 2012) and inferences from coupledatmosphere-ocean general circulation models (AOGCMs; Gregory and Forster 2008; Marotzke et al. 2015), suggest thatocean heat uptake efficiency on decadal timescales lies in the range 0.5-1.5 W/sq m/K and is thus comparable to theclimate feedback parameter (Murphy et al. 2009). Moreover, the ocean not only plays a key role in setting the timing ofwarming but also its regional patterns (Marshall et al. 2014), which is crucial to our understanding of regional climate,carbon and heat uptake, and sea-level change. This short communication is based on a presentation given by A.Romanou at a recent workshop, Oceans Carbon and Heat Uptake: Uncertainties and Metrics, co-hosted by US CLIVARand OCB. As briefly reviewed below, we have incomplete but growing knowledge of how ocean models used in climatechange projections sequester heat and carbon into the interior. To understand and thence reduce errors and biases inthe ocean component of coupled models, as well as elucidate the key mechanisms at work, in the final section we outlinea proposed model intercomparison project named FAFMIP. In FAFMIP, coupled integrations would be carried out withprescribed overrides of wind stress and freshwater and heat fluxes acting at the sea surface.
Fission gas behaviour modelling in plate fuel during a power transient
International Nuclear Information System (INIS)
Portier, S.
2003-01-01
This thesis is dedicated to the identification and modelization of the phenomena which are at the origin of the release of the fission gas formed in UO 2 plate fuels during the irradiation in a power transient. In the first experimental part, samples of plate fuels, irradiated at 36 GWj/tU, have been annealed to temperatures from 1100 C to 1500 C in a device that enabled the measurement of gas release in real time. At 1300 C, post-annealing observations demonstrated a link between the measured gas releases to a rapid formation of labyrinths at the grain surface. These labyrinths, which were formed by intergranular bubble interconnection, create release paths for the gas atoms which reach the grain surface. At this stage, the available experimental results (annealing and observations) were interpreted considering that it is the spreading of the gas atoms from the grains to the grain boundaries that is at the origin of the observed releases. This interpretation generates the hypothesis that a) at the end of the basic irradiation, the gas is at the atomic state and b) during the annealing, the spreading is reduced by the intragranular bubbles of the gas atoms. The last part of the work is dedicated to the modelization of the main phenomena at the origin of the gas release. The model developed, based on the model of the gas behaviour in MARGARET PWR, highlighted the great influence of the irradiation conditions on the gas distribution at the end of the irradiation and also its influence on the fission gas release during the power transient. (author) [fr
Analytical model for power plant condenser for transients and off-normal operating conditions
International Nuclear Information System (INIS)
Thangamani, I.; Dutta, Anu; Chakraborty, G.; Ghosh, A.K.
2006-11-01
A computer code for power plant condenser dynamic analysis has been developed based on a lumped parameter approach considering time dependent mass and energy conservation equations over the control volumes for the shell side as well as tube side fluids. Effects of heat transfer on condenser structure and hot well level transients were considered in the analysis. Suitable heat transfer coefficient recommended by various standards and codes were employed. The code was used to analyze the condenser performance during steady state as well as transient (load rejection or turbine trip) conditions. The condenser performance is predicted in terms of condenser back pressure, shell side steam temperature and tube side coolant exit temperature with respect to time. As a part of parametric studies, the effect of change in tube side coolant flow rate and inlet temperature was also studied. The analysis predicted that up to 47% of rated coolant flow rate on the tube side (for design conditions), the steam dumping can be continued without condenser isolation. The paper describes the detailed methodology adopted for the condenser modeling and presents the results obtained from the different parametric studies and code validation. (author)
Development of a simplified calculational model for the transient core bowing effect
International Nuclear Information System (INIS)
Yokoo, Takeshi
1997-01-01
A simplified method to analyze the transient core radial deformation has been developed based on a model that calculates the shape of a single representative fuel assembly on the outermost row. The plant transient code CERES has been revised utilizing this method so that a integrated calculational process for the core neutronics, thermal-hydraulics and deformation can be realized. Using CERES, the responses of a 1000MWe class pool type metal fuel FBR plant during a ULOF event are calculated. According to the results, it is clarified that a passive shutdown without coolant boiling is attainable by selecting appropriate values for major design parameters such as the gap width between load-pad and the pad material properties. The maximum coolant temperature during ULOF is found to be 790C when the above core load-pad gap is set to 0.05 mm, which can be regarded as the most likely valued. The temperature increases to 915C but is still lower than the boiling point when 40% of uncertainty is taken into account. (author)
International Nuclear Information System (INIS)
Magalhaes, Mardson Alencar de Sa; Lira, Carlos Alberto Brayner de Oliveira; Silva, Mario Augusto Bezerra da
2011-01-01
The IRIS project has significantly advanced in the last few years in response to a demand for a new generation reactor, that could fulfill the essential requirements for a future nuclear power plant: better economics, safety-by-design, low proliferation risk and environmental sustainability. IRIS reactor is a integral type PWR in which all primary components are arranged inside the pressure vessel. This configuration involves important changes in relation to a conventional PWR. These changes require several studies to comply with the safe operational limits for the reactor. In this paper, a study has been conducted to develop a dynamic model (named MODIRIS) for transient analysis, implemented in the MATLAB'S software SIMULINK, allowing the analysis of IRIS behavior by considering the neutron point kinetics for power production. The methodology is based on generating a set of differential equations of neutronic and thermal-hydraulic balances which describes the dynamics of the primary circuit, as well as a set of differential equations describing the dynamics of secondary circuit. The equations and initialization parameters at full power were into the SIMULINK and the code was validated by the confrontation with RELAP simulations for a transient of feedwater reduction in the steam generators. (author)
Directory of Open Access Journals (Sweden)
Xumei Wang
2015-01-01
Full Text Available Xueshuantong for Injection (Lyophilized (XST, a Chinese Materia Medica standardized product extracted from Panax notoginseng (Burk., is used extensively for the treatment of cerebrovascular diseases such as acutely cerebral infarction clinically in China. In the present study, we evaluated the acute and extended protective effects of XST in different rat cerebral ischemic model and explored its effect on peroxiredoxin (Prx 6-toll-like receptor (TLR 4 signaling pathway. We found that XST treatment for 3 days could significantly inhibit transient middle cerebral artery occlusion (MCAO induced infarct volume and swelling percent and regulate the mRNA expression of interleukin-1β (IL-1β, IL-17, IL-23p19, tumor necrosis factor-α (TNFα, and inducible nitric oxide synthase (iNOS in brain. Further study demonstrated that treatment with XST suppressed the protein expression of peroxiredoxin (Prx 6-toll-like receptor (TLR 4 and phosphorylation level of p38 and upregulated the phosphorylation level of STAT3. In permanent MCAO rats, XST could reduce the infarct volume and swelling percent. Moreover, our results revealed that XST treatment could increase the rats’ weight and improve a batch of functional outcomes. In conclusion, the present data suggested that XST could protect against ischemia injury in transient and permanent MCAO rats, which might be related to Prx6-TLR4 pathway.
Energy Technology Data Exchange (ETDEWEB)
Moulin, D J.F.
1993-09-10
In order to study the power transients effects on PWR fuel rod clad, ramp tests in a pressurized water loop, are carried out at OSIRIS reactor. The present thesis deals with the on-line control of the device, during power ramp and conditioning irradiation. Based on a convolution-type resolution of the kinetics equations, a dynamic compensation of the Silver self-powered neutron detector was developed. With this method, the uncertainty of the ramp end-point is lower than 1%, thus it is very suited for monitoring both transient, as well as steady state conditions. Furthermore, a thermohydraulic model of the irradiation device is described: heat transfer equations, including gamma heating in materials, are solved to obtain temperatures and thermal fluxes of steady states. Results from the model and temperature measurements of the coolant are used together for fuel power determination, in real time. The clad external temperature profile is also calculated and displayed, to improve the irradiation monitoring. (author), 51 refs., 12 annexes, 66 figs.
Directory of Open Access Journals (Sweden)
Guangtao Zhang
2015-01-01
Full Text Available In the field of hydropower station transient process simulation (HSTPS, characteristic graph-based iterative hydroturbine model (CGIHM has been widely used when large disturbance hydroturbine modeling is involved. However, by this model, iteration should be used to calculate speed and pressure, and slow convergence or no convergence problems may be encountered for some reasons like special characteristic graph profile, inappropriate iterative algorithm, or inappropriate interpolation algorithm, and so forth. Also, other conventional large disturbance hydroturbine models are of some disadvantages and difficult to be used widely in HSTPS. Therefore, to obtain an accurate simulation result, a simple method for hydroturbine modeling is proposed. By this method, both the initial operating point and the transfer coefficients of linear hydroturbine model keep changing during simulation. Hence, it can reflect the nonlinearity of the hydroturbine and be used for Francis turbine simulation under large disturbance condition. To validate the proposed method, both large disturbance and small disturbance simulations of a single hydrounit supplying a resistive, isolated load were conducted. It was shown that the simulation result is consistent with that of field test. Consequently, the proposed method is an attractive option for HSTPS involving Francis turbine modeling under large disturbance condition.
Development of a transient calculation model for a closed sodium natural circulation loop
International Nuclear Information System (INIS)
Chang, Won Pyo; Ha, Kwi Seok; Jeong, Hae Yong; Heo, Sun; Lee, Yong Bum
2003-09-01
A natural circulation loop has usually adopted for a Liquid Metal Reactor (LMR) because of its high reliability. Up-rating of the current KALIMER capacity requires an additional PDRC to the existing PVCS to remove its decay heat under an accident. As the system analysis code currently used for LMR in Korea does not feature a stand alone capability to simulate a closed natural circulation loop, it is not eligible to be applied to PDRC. To supplement its limitation, a steady state calculation model had been developed during the first phase, and development of the transient model has successively carried out to close the present study. The developed model will then be coupled with the system analysis code, SSC-K to assess a long term cooling for the new conceptual design. The incompressibility assumption of sodium which allows the circuit to be modeled with a single loop flow, makes the model greatly simplified comparing with LWR. Some thermal-hydraulic models developed during this study can be effectively applied to other system analysis codes which require such component models, and the present development will also contribute to establishment of a code system for the LMR analysis
Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments
Energy Technology Data Exchange (ETDEWEB)
Mendler, O J; Takeuchi, K; Young, M Y
1986-10-01
The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.
Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments
International Nuclear Information System (INIS)
Mendler, O.J.; Takeuchi, K.; Young, M.Y.
1986-10-01
The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results
Compressor Modeling for Transient Analysis of Supercritical CO2 Brayton Cycle by using MARS code
Energy Technology Data Exchange (ETDEWEB)
Park, Joo Hyun; Park, Hyun Sun; Kim, Tae Ho; Kwon, Jin Gyu [POSTECH, Pohang (Korea, Republic of); Bae, Sung Won; Cha, Jae Eun [KAERI, Daejeon (Korea, Republic of)
2016-05-15
In this study, SCIEL (Supercritical CO{sub 2} Integral Experimental Loop) was chosen as a reference loop and the MARS code was as the transient cycle analysis code. As a result, the compressor homologous curve was developed from the SCIEL experimental data and MARS analysis was performed and presented in the paper. The advantages attract SCO{sub 2}BC as a promising next generation power cycles. The high thermal efficiency comes from the operation of compressor near the critical point where the properties of SCO{sub 2}. The approaches to those of liquid phase, leading drastically lower the compression work loss. However, the advantage requires precise and smooth operation of the cycle near the critical point. However, it is one of the key technical challenges. The experimental data was steady state at compressor rotating speed of 25,000 rpm. The time, 3133 second, was starting point of steady state. Numerical solutions were well matched with the experimental data. The mass flow rate from the MARS analysis of approximately 0.7 kg/s was close to the experimental result of 0.9 kg/s. It is expected that the difference come from the measurement error in the experiment. In this study, the compressor model was developed and implemented in MARS to study the transient analysis of SCO{sub 2}BC in SCIEL. We obtained the homologous curves for the SCIEL compressor using experimental data and performed nodalization of the compressor model using MARS code. In conclusions, it was found that numerical solutions from the MARS model were well matched with experimental data.
Energy Technology Data Exchange (ETDEWEB)
Bianchi, Paulo Henrique
2008-07-01
This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)
Physical models and numerical methods of the reactor dynamic computer program RETRAN
International Nuclear Information System (INIS)
Kamelander, G.; Woloch, F.; Sdouz, G.; Koinig, H.
1984-03-01
This report describes the physical models and the numerical methods of the reactor dynamic code RETRAN simulating reactivity transients in Light-Water-Reactors. The neutron-physical part of RETRAN bases on the two-group-diffusion equations which are solved by discretization similar to the TWIGL-method. An exponential transformation is applied and the inner iterations are accelerated by a coarse-mesh-rebalancing procedure. The thermo-hydraulic model approximates the equation of state by a built-in steam-water-table and disposes of options for the calculation of heat-conduction coefficients and heat transfer coefficients. (Author) [de
3-dimensional thermohydraulic analysis of KALIMER reactor pool during unprotected accidents
Energy Technology Data Exchange (ETDEWEB)
Lee, Yong Bum; Hahn Do Hee
2003-01-01
During a normal reactor scram, the heat generation is reduced almost instantaneously while the coolant flow rate follows the pump coastdown. This mismatch between power and flow results in a situation where the core flow entering the hot pool is at a lower temperature than the temperature of the bulk pool sodium. This temperature difference leads to thermal stratification. Thermal stratification can occur in the hot pool region if the entering coolant is colder than the existing hot pool coolant and the flow momentum is not large enough to overcome the negative buoyancy force. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response. Hence, it is necessary to predict the pool coolant temperature distribution with sufficient accuracy to determine the inlet temperature conditions for the IHXs and its contribution to the net buoyancy head. Therefore, two-dimensional hot pool thermohydraulic model named HP2D has been developed. In this report code-to-code comparison analysis between HP2D and COMMIX-1AR/P has been performed in the case of steady-state and UTOP.
Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach
International Nuclear Information System (INIS)
Silva, Alexandro S.; Dominguez, Dany S.; Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G.; Lira, Carlos Alberto Brayner de Oliveira
2015-01-01
High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)
International Nuclear Information System (INIS)
Hampel, G.; Poss, G.; Frohlich, H.K.
1989-10-01
The objective of the project was to draw up an instrumentation plan for the French core melting programme PHEBUS FP. This instrumentation plan essentially was to include proven and reliable instruments for recording various thermohydraulic, aerosol and hydrogen phenomena. The candidate measuring methods, which are known mainly from reactor safety programmes, have been described and examined for their usefulness in PHEBUS. Each method and instrument has been described in detail under various aspects such as measuring principle, measuring range, technical design, evaluation model, calibration procedure, accuracy, previous experience, commercial availability, etc. Special attention has been paid to the behaviour of the measuring transducers when exposed to radiation. First, the performance of the instruments was compared with the requirements of PHEBUS. The results of this comparison served as the basis for a measuring concept in tabular form, giving the locations of the measurements, the measuring tasks, and the number and kind of instruments that are recommended. Redundancy and cost-benefit aspects have been taken into account in qualitative terms
Effect of thermohydraulic parameter on the flux distribution and the effective multiplication factor
International Nuclear Information System (INIS)
Mello, J.C.; Valladares, G.L.
1990-01-01
The influence of two thermohydraulics parameters; the coolant flow velocity along the reactor channels and the increase of the average water temperature through the core, on the thermal flux distribution and on the effective multiplication factor, was studied in a radioisotopes production reactor. The results show that, for a fixed values of the thermohydraulics parameters reffered above, there are limits for the reactor core volume reduction for each value of the V sub(mod)/V sub(comb) ratio. These thermohydraulics conditions determine the higher termal flux value in the flux-trap and the lower value of the reactor effective multiplication factor. It is also show that there is a V sub(mod)/V sub(comb) ratio value that correspond to the higher value of the lower effective multiplication factor. These results was interpreted and comment using fundamentals concepts and relations of reactor physics. (author)
Modeling the behavior of metallic fast reactor fuels during extended transients
International Nuclear Information System (INIS)
Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.
1993-01-01
Passive safety features in metal-fueled reactors utilizing the Integral Fast Reactor (IFR) fuel system make it possible to avoid core damage for extended time periods even when automatic scram system fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this intermediate time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements. (orig.)
Non-linear belt transient analysis. A hybrid model for numerical belt conveyor simulation
Energy Technology Data Exchange (ETDEWEB)
Harrison, A. [Scientific Solutions, Inc., Aurora, CO (United States)
2008-07-01
Frictional and rolling losses along a running conveyor are discussed due to their important influence on wave propagation during starting and stopping. Hybrid friction models allow belt rubber losses and material flexing to be included in the initial tension calculations prior to any dynamic analysis. Once running tensions are defined, a numerical integration method using non-linear stiffness gradients is used to generate transient forces during starting and stopping. A modified Euler integration technique is used to simulate the entire starting and stopping cycle in less than 0.1 seconds. The procedure enables a faster scrutiny of unforeseen conveyor design issues such as low belt tension zones and high forces at drives. (orig.)
Modeling the behavior of metallic fast reactor fuels during extended transients
International Nuclear Information System (INIS)
Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.
1992-01-01
Passive safety features in the metal-fueled Integral Fast Reactor (IFR) make it possible to avoid core damage for extended time periods even when automatic scram systems fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements
A diffusion model for two parallel queues with processor sharing: transient behavior and asymptotics
Directory of Open Access Journals (Sweden)
Charles Knessl
1999-01-01
Full Text Available We consider two identical, parallel M/M/1 queues. Both queues are fed by a Poisson arrival stream of rate λ and have service rates equal to μ. When both queues are non-empty, the two systems behave independently of each other. However, when one of the queues becomes empty, the corresponding server helps in the other queue. This is called head-of-the-line processor sharing. We study this model in the heavy traffic limit, where ρ=λ/μ→1. We formulate the heavy traffic diffusion approximation and explicitly compute the time-dependent probability of the diffusion approximation to the joint queue length process. We then evaluate the solution asymptotically for large values of space and/or time. This leads to simple expressions that show how the process achieves its stead state and other transient aspects.
International Nuclear Information System (INIS)
Dufour, C.; Toulemonde, M.; Paumier, E.; Lesellier de Chezelles, B.; Delignon, V.
1991-01-01
Latent tracks have been observed in amorphous semi-conductors after heavy ion irradiation in the electronic stopping power regime. A transient thermodynamic model is developed including energy diffusion on the electron gas and on the atomic lattice and energy exchange between these two systems. A set of two non linear differential equations is solved numerically in cylindrical geometry in order to predict the radii of the latent tracks observed in amorphous germanium and silicon. A good agreement is obtained for the two materials using the same set of input parameters for the energy diffusion on the electronic system and the same coupling constant for the energy exchange between electron and lattice atoms despite the large differences in the macroscopic lattice thermodynamical parameters of the two materials
A simplified model for equilibrium and transient swelling of thermo-responsive gels.
Drozdov, A D; deClaville Christiansen, J
2017-11-01
A simplified model is developed for the elastic response of thermo-responsive gels subjected to swelling under an arbitrary deformation with finite strains. The constitutive equations involve five adjustable parameters that are determined by fitting observations in equilibrium water uptake tests and T-jump transient tests on thin gel disks. Two scenarios for water release under heating are revealed by means of numerical simulation. When the final temperature in a T-jump test is below the volume-phase transition temperature, deswelling is characterized by smooth distribution of water molecules and small tensile stresses. When the final temperature exceeds the critical temperature, a gel disk is split into three regions (central part with a high concentration of water molecules and two domains near the boundaries with low water content) separated by sharp interfaces, whose propagation is accompanied by development of large (comparable with the elastic modulus) tensile stresses. Copyright © 2017 Elsevier Ltd. All rights reserved.
Modeling active region transient brightenings observed with X-ray telescope as multi-stranded loops
Energy Technology Data Exchange (ETDEWEB)
Kobelski, Adam R.; McKenzie, David E. [Department of Physics, P.O. Box 173840, Montana State University, Bozeman, MT 59717-3840 (United States); Donachie, Martin, E-mail: kobelski@solar.physics.montana.edu [University of Glasgow, Glasgow, G128QQ, Scotland (United Kingdom)
2014-05-10
Strong evidence exists that coronal loops as observed in extreme ultraviolet and soft X-rays may not be monolithic isotropic structures, but can often be more accurately modeled as bundles of independent strands. Modeling the observed active region transient brightenings (ARTBs) within this framework allows for the exploration of the energetic ramifications and characteristics of these stratified structures. Here we present a simple method of detecting and modeling ARTBs observed with the Hinode X-Ray Telescope (XRT) as groups of zero-dimensional strands, which allows us to probe parameter space to better understand the spatial and temporal dependence of strand heating in impulsively heated loops. This partially automated method can be used to analyze a large number of observations to gain a statistical insight into the parameters of coronal structures, including the number of heating events required in a given model to fit the observations. In this article, we present the methodology and demonstrate its use in detecting and modeling ARTBs in a sample data set from Hinode/XRT. These initial results show that, in general, multiple heating events are necessary to reproduce observed ARTBs, but the spatial dependence of these heating events cannot yet be established.
Modeling active region transient brightenings observed with X-ray telescope as multi-stranded loops
International Nuclear Information System (INIS)
Kobelski, Adam R.; McKenzie, David E.; Donachie, Martin
2014-01-01
Strong evidence exists that coronal loops as observed in extreme ultraviolet and soft X-rays may not be monolithic isotropic structures, but can often be more accurately modeled as bundles of independent strands. Modeling the observed active region transient brightenings (ARTBs) within this framework allows for the exploration of the energetic ramifications and characteristics of these stratified structures. Here we present a simple method of detecting and modeling ARTBs observed with the Hinode X-Ray Telescope (XRT) as groups of zero-dimensional strands, which allows us to probe parameter space to better understand the spatial and temporal dependence of strand heating in impulsively heated loops. This partially automated method can be used to analyze a large number of observations to gain a statistical insight into the parameters of coronal structures, including the number of heating events required in a given model to fit the observations. In this article, we present the methodology and demonstrate its use in detecting and modeling ARTBs in a sample data set from Hinode/XRT. These initial results show that, in general, multiple heating events are necessary to reproduce observed ARTBs, but the spatial dependence of these heating events cannot yet be established.
Two-fluid model for transient analysis of slug flow in oil wells
International Nuclear Information System (INIS)
Cazarez-Candia, O.; Benitez-Centeno, O.C.; Espinosa-Paredes, G.
2011-01-01
In this work it is presented a transient, one-dimensional, adiabatic model for slug flow simulation, which appears when liquid (mixture of oil and water) and gas flow simultaneously through pipes. The model is formed by space and time averaged conservation equations for mass, momentum and energy for each phase, the numerical solution is based on the finite difference technique in the implicit scheme. Velocity, pressure, volumetric fraction and temperature profiles for both phases were predicted for inclination angles from the horizontal to the vertical position (unified model) and ascendant flow. Predictions from the model were validated using field data and ten correlations commonly used in the oil industry. The effects of gas heating or cooling, due to compression and expansion processes, on the predictions and numerical stability, were studied. It was found that when these effects are taken into account, a good behavior of temperature predictions and numerical stability are obtained. The model presents deviations lower than 14% regarding field data and it presents better predictions than most of the correlations.
Two-fluid model for transient analysis of slug flow in oil wells
Energy Technology Data Exchange (ETDEWEB)
Cazarez-Candia, O., E-mail: ocazarez@imp.mx [Instituto Mexicano del Petroleo, Eje central Lazaro Cardenas No. 152, Col. San Bartolo Atepehuacan, Mexico D.F. 07730 (Mexico); Instituto Tecnologico de Zacatepec, Depto. de Metal-Mecanica, Calzada Tecnologico, No. 27, Zacatepec, Morelos 62780 (Mexico); Benitez-Centeno, O.C. [Centro Nacional de Investigacion y Desarrollo Tecnologico, Depto. de Mecanica, Interior Internado Palmira s/n, Col. Palmira, Cuernavaca, Morelos 62490 (Mexico); Espinosa-Paredes, G. [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av San Rafael Atlixco No 186, Col. Vicentina 55-534, Mexico D.F. 09340 (Mexico)
2011-06-15
In this work it is presented a transient, one-dimensional, adiabatic model for slug flow simulation, which appears when liquid (mixture of oil and water) and gas flow simultaneously through pipes. The model is formed by space and time averaged conservation equations for mass, momentum and energy for each phase, the numerical solution is based on the finite difference technique in the implicit scheme. Velocity, pressure, volumetric fraction and temperature profiles for both phases were predicted for inclination angles from the horizontal to the vertical position (unified model) and ascendant flow. Predictions from the model were validated using field data and ten correlations commonly used in the oil industry. The effects of gas heating or cooling, due to compression and expansion processes, on the predictions and numerical stability, were studied. It was found that when these effects are taken into account, a good behavior of temperature predictions and numerical stability are obtained. The model presents deviations lower than 14% regarding field data and it presents better predictions than most of the correlations.
International Nuclear Information System (INIS)
Zhang, Hao; Chen, Diyi; Xu, Beibei; Wang, Feifei
2015-01-01
Graphical abstract: Nonlinear dynamic transfer coefficients are introduced to the hydro-turbine governing system. In the process of load reject ion transient, the nonlinear dynamical behaviors of the system are studied in detail. - Highlights: • A novel mathematical model of a hydro-turbine governing system is established. • The process of load rejection transient is considered. • Nonlinear dynamic transfer coefficients are introduced to the system. • The bifurcation diagram with the variable t has better engineering significance. • The nonlinear dynamical behaviors of the system are studied in detail. - Abstract: This article pays attention to the mathematical modeling of a hydro-turbine governing system in the process of load rejection transient. As a pioneer work, the nonlinear dynamic transfer coefficients are introduced in a penstock system. Considering a generator system, a turbine system and a governor system, we present a novel nonlinear dynamical model of a hydro-turbine governing system. Fortunately, for the unchanged of PID parameters, we acquire the stable regions of the governing system in the process of load rejection transient by numerical simulations. Moreover, the nonlinear dynamic behaviors of the governing system are illustrated by bifurcation diagrams, Poincare maps, time waveforms and phase orbits. More importantly, these methods and analytic results will present theoretical groundwork for allowing a hydropower station in the process of load rejection transient
Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA
International Nuclear Information System (INIS)
Badulescu, A.; Groeneveld, D.C.
2000-01-01
Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)
International Nuclear Information System (INIS)
Iváncsy, T; Kiss, I; Tamus, Z Á; Szücs, L
2015-01-01
The lightning current generates time-varying magnetic field near the down-conductor and the down-conductors are mounted on the wall of the buildings where residential places might be situated. It is well known that the rapidly changing magnetic fields can generate dangerous eddy currents in the human body.The higher duration and gradient of the magnetic field can cause potentially life threatening cardiac stimulation. The coupling mechanism between the electromagnetic field and the human body is based on a well-known physical phenomena (e.g. Faradays law of induction). However, the calculation of the induced current is very complicated because the shape of the organs is complex and the determination of the material properties of living tissues is difficult, as well. Our previous study revealed that the cardiac stimulation is independent of the rising time of the lightning current and only the peak of the current counts.In this study, the authors introduce an improved model of the interaction of electromagnetic fields of lighting current near down-conductor and human body. Our previous models are based on the quasi stationer field calculations, the new improved model is a transient model. This is because the magnetic field around the down-conductor and in the human body can be determined more precisely, therefore the dangerous currents in the body can be estimated. (paper)
Freedom: a transient fission-product release model for radioactive and stable species
International Nuclear Information System (INIS)
Macdonald, L.D.; Lewis, B.J.; Iglesias, F.C.
1989-05-01
A microstructure-dependent fission-gas release and swelling model (FREEDOM) has been developed for UO 2 fuel. The model describes the transient release behaviour for both the radioactive and stable fission-product species. The model can be applied over the full range of operating conditions, as well as for accident conditions that result in high fuel temperatures. The model accounts for lattice diffusion and grain-boundary sweeping of fusion products to the grain boundaries, where the fission gases accumulate in grain-face bubbles as a result of vacancy diffusion. Release of fission-gas to the free void of the fuel element occurs through the interlinkage of bubbles and cracks on the grain boundaries. This treatment also accounts for radioactive chain decay and neutron-induced transmutation effects. These phenomena are described by mass balance equations which are numerically solved using a moving-boundary, finite-element method with mesh refinement. The effects of grain-face bubbles on fuel swelling and fuel thermal conductivity are included in the ELESIM fuel performance code. FREEDOM has an accuracy of better than 1% when assessed against an analytic solution for diffusional release. The code is being evaluated against a fuel performance database for stable gas release, and against sweep-gas and in-cell fission-product release experiments at Chalk River for active species
Simulating CRN derived erosion rates in a transient Andean catchment using the TTLEM model
Campforts, Benjamin; Vanacker, Veerle; Herman, Frédéric; Schwanghart, Wolfgang; Tenrorio Poma, Gustavo; Govers, Gerard
2017-04-01
Assessing the impact of mountain building and erosion on the earth surface is key to reconstruct and predict terrestrial landscape evolution. Landscape evolution models (LEMs) are an essential tool in this research effort as they allow to integrate our growing understanding of physical processes governing erosion and transport of mass across the surface. The recent development of several LEMs opens up new areas of research in landscape evolution. Here, we want to seize this opportunity by answering a fundamental research question: does a model designed to simulate landscape evolution over geological timescales allows to simulate spatially varying erosion rates at a millennial timescale? We selected the highly transient Paute catchment in the Southeastern Ecuadorian Andes as a study area. We found that our model (TTLEM) is capable to better explain the spatial patterns of ca. 30 Cosmogenic Radio Nuclide (CRN) derived catchment wide erosion rates in comparison to a classical, statistical approach. Thus, the use of process-based landscape evolution models may not only be of great help to understand long-term landscape evolution but also in understanding spatial and temporal variations in sediment fluxes at the millennial time scale.
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.
3-D Forward modeling of Induced Polarization Effects of Transient Electromagnetic Method
Wu, Y.; Ji, Y.; Guan, S.; Li, D.; Wang, A.
2017-12-01
In transient electromagnetic (TEM) detection, Induced polarization (IP) effects are so important that they cannot be ignored. The authors simulate the three-dimensional (3-D) induced polarization effects in time-domain directly by applying the finite-difference time-domain method (FDTD) based on Cole-Cole model. Due to the frequency dispersion characteristics of the electrical conductivity, the computations of convolution in the generalized Ohm's law of fractional order system makes the forward modeling particularly complicated. Firstly, we propose a method to approximate the fractional order function of Cole-Cole model using a lower order rational transfer function based on error minimum theory in the frequency domain. In this section, two auxiliary variables are introduced to transform nonlinear least square fitting problem of the fractional order system into a linear programming problem, thus avoiding having to solve a system of equations and nonlinear problems. Secondly, the time-domain expression of Cole-Cole model is obtained by using Inverse Laplace transform. Then, for the calculation of Ohm's law, we propose an e-index auxiliary equation of conductivity to transform the convolution to non-convolution integral; in this section, the trapezoid rule is applied to compute the integral. We then substitute the recursion equation into Maxwell's equations to derive the iterative equations of electromagnetic field using the FDTD method. Finally, we finish the stimulation of 3-D model and evaluate polarization parameters. The results are compared with those obtained from the digital filtering solution of the analytical equation in the homogeneous half space, as well as with the 3-D model results from the auxiliary ordinary differential equation method (ADE). Good agreements are obtained across the three methods. In terms of the 3-D model, the proposed method has higher efficiency and lower memory requirements as execution times and memory usage were reduced by 20
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)
Directory of Open Access Journals (Sweden)
U. Schneider
2009-01-01
Full Text Available The paper presents the structural application of a new thermal induced strain model for concrete – the TIS-Model. An advanced transient concrete model (ATCM is applied with the material model of the TIS-Model. The non-linear model comprises thermal strain, elastic strain, plastic strain and transient temperature strains, and load history modelling of restraint concrete structures subjected to fire.The calculations by finite element analysis (FEA were done using the SAFIR structural code. The FEA software was basically new with respect to the material modelling derived to use the new TIS-Model (as a transient model considers thermal induced strain. The equations of the ATCM consider a lot of capabilities, especially for considering irreversible effects of temperature on some material properties. By considering the load history during heating up, increasing load bearing capacity may be obtained due to higher stiffness of the concrete. With this model, it is possible to apply the thermal-physical behaviour of material laws for calculation of structures under extreme temperature conditions.A tunnel cross section designed and built by the cut and cover method is calculated with a tunnel fire curve. The results are compared with the results of a calculation with the model of the Eurocode 2 (EC2-Model. The effect of load history in highly loaded structures under fire load will be investigated.A comparison of this model with the ordinary calculation system of Eurocode 2 (EC2 shows that a better evaluation of the safety level was achieved with the new model. This opens a space for optimizing concrete structure design with transient temperature conditions up to 1000 °C.
Forward modeling transient brightenings and microflares around an active region observed with Hi-C
Energy Technology Data Exchange (ETDEWEB)
Kobelski, Adam R. [Now at National Radio Astronomy Observatory, P.O. Box 2, Green Bank, WV 24944 (United States); McKenzie, David E., E-mail: kobelski@solar.physics.montana.edu [Department of Physics, P.O. Box 173840, Montana State University, Bozeman, MT 59717-3840 (United States)
2014-10-20
Small-scale flare-like brightenings around active regions are among the smallest and most fundamental of energetic transient events in the corona, providing a testbed for models of heating and active region dynamics. In a previous study, we modeled a large collection of these microflares observed with Hinode/X-Ray Telescope (XRT) using EBTEL and found that they required multiple heating events, but could not distinguish between multiple heating events on a single strand, or multiple strands each experiencing a single heating event. We present here a similar study, but with extreme-ultraviolet data of Active Region 11520 from the High Resolution Coronal Imager (Hi-C) sounding rocket. Hi-C provides an order of magnitude improvement to the spatial resolution of XRT, and a cooler temperature sensitivity, which combine to provide significant improvements to our ability to detect and model microflare activity around active regions. We have found that at the spatial resolution of Hi-C (≈0.''3), the events occur much more frequently than expected (57 events detected, only 1 or 2 expected), and are most likely made from strands of the order of 100 km wide, each of which is impulsively heated with multiple heating events. These findings tend to support bursty reconnection as the cause of the energy release responsible for the brightenings.
Directory of Open Access Journals (Sweden)
Andreas Hackl
2016-12-01
Full Text Available Developing functions for advanced driver assistance systems requires very accurate tyre models, especially for the simulation of transient conditions. In the past, parametrisation of a given tyre model based on measurement data showed shortcomings, and the globally optimal solution obtained did not appear to be plausible. In this article, an optimisation strategy is presented, which is able to find plausible and physically feasible solutions by detecting many local outcomes. The firefly algorithm mimics the natural behaviour of fireflies, which use a kind of flashing light to communicate with other members. An algorithm simulating the intensity of the light of a single firefly, diminishing with increasing distances, is implicitly able to detect local solutions on its way to the best solution in the search space. This implicit clustering feature is stressed by an additional explicit clustering step, where local solutions are stored and terminally processed to obtain a large number of possible solutions. The enhanced firefly algorithm will be first applied to the well-known Rastrigin functions and then to the tyre parametrisation problem. It is shown that the firefly algorithm is qualified to find a high number of optimisation solutions, which is required for plausible parametrisation for the given tyre model.
Minshull, T. A.; Marín-Moreno, H.; Armstrong McKay, D. I.; Wilson, P. A.
2016-08-01
During the Paleocene-Eocene Thermal Maximum (PETM), the carbon isotopic signature (δ13C) of surface carbon-bearing phases decreased abruptly by at least 2.5 to 3.0‰. This carbon isotope excursion (CIE) has been attributed to widespread methane hydrate dissociation in response to rapid ocean warming. We ran a thermohydraulic modeling code to simulate hydrate dissociation due to ocean warming for various PETM scenarios. Our results show that hydrate dissociation in response to such warming can be rapid but suggest that methane release to the ocean is modest and delayed by hundreds to thousands of years after the onset of dissociation, limiting the potential for positive feedback from emission-induced warming. In all of our simulations at least half of the dissociated hydrate methane remains beneath the seabed, suggesting that the pre-PETM hydrate inventory needed to account for all of the CIE is at least double that required for isotopic mass balance.
International Nuclear Information System (INIS)
Costa, J.R.
1978-12-01
An analysis is done of the core behavior for a 1861 MW(th) pressurized water reactor with two coolant loops, during the blowdown phase of a double-ended cold leg rupture, between the main feedwater pump, and the pressure vessel. The analysis is done through a detailed thermohydraulic study of the hot pin channel with RELAP4/MOD 5 code, including the Evaluatin Model options. The problem is solved separately for two values of discharge coefficient (C sub(D)= 1,0 and 0,4). The results show that the maximum clad temperature is lower than the limit value for licensing purposes. Concerning clad material oxidation, the maximum value obtained is also under the limit of acceptance. (author) [pt
Experimental validation of Pu-Sm evolution model for CANDU-6 power transients
International Nuclear Information System (INIS)
Coutsiers, Eduardo E.; Pomerantz, Marcelo E.; Moreno, Carlos A.
2000-01-01
Development of a methodology to evaluate the reactivity produced by Pu-Sm transient, effect displayed after power transients. This methodology allows to predict the behavior of liquid zones with which the fine control of CANDU reactor power is made. With this information, it is easier to foresee the refueling demand after power movements. The comparison with experimental results showed good agreement. (author)
Ramadasan-Nair, Renjini; Gayathri, Narayanappa; Mishra, Sudha; Sunitha, Balaraju; Mythri, Rajeswara Babu; Nalini, Atchayaram; Subbannayya, Yashwanth; Harsha, Hindalahalli Chandregowda; Kolthur-Seetharam, Ullas; Bharath, Muchukunte Mukunda Srinivas
2014-01-01
Muscular dystrophies (MDs) and inflammatory myopathies (IMs) are debilitating skeletal muscle disorders characterized by common pathological events including myodegeneration and inflammation. However, an experimental model representing both muscle pathologies and displaying most of the distinctive markers has not been characterized. We investigated the cardiotoxin (CTX)-mediated transient acute mouse model of muscle degeneration and compared the cardinal features with human MDs and IMs. The CTX model displayed degeneration, apoptosis, inflammation, loss of sarcolemmal complexes, sarcolemmal disruption, and ultrastructural changes characteristic of human MDs and IMs. Cell death caused by CTX involved calcium influx and mitochondrial damage both in murine C2C12 muscle cells and in mice. Mitochondrial proteomic analysis at the initial phase of degeneration in the model detected lowered expression of 80 mitochondrial proteins including subunits of respiratory complexes, ATP machinery, fatty acid metabolism, and Krebs cycle, which further decreased in expression during the peak degenerative phase. The mass spectrometry (MS) data were supported by enzyme assays, Western blot, and histochemistry. The CTX model also displayed markers of oxidative stress and a lowered glutathione reduced/oxidized ratio (GSH/GSSG) similar to MDs, human myopathies, and neurogenic atrophies. MS analysis identified 6 unique oxidized proteins from Duchenne muscular dystrophy samples (n = 6) (versus controls; n = 6), including two mitochondrial proteins. Interestingly, these mitochondrial proteins were down-regulated in the CTX model thereby linking oxidative stress and mitochondrial dysfunction. We conclude that mitochondrial alterations and oxidative damage significantly contribute to CTX-mediated muscle pathology with implications for human muscle diseases. PMID:24220031
Comparison of thermohydraulic characteristics in the use of various coolants
International Nuclear Information System (INIS)
Muramatsu, Toshiharu; Suda, Kazunori; Yamaguchi, Akira
2000-11-01
Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiated based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1) There is no remarkable difference between liquid sodium and liquid Pb-Bi in characteristics of internal flows and free surface characteristics based on Fr number. (2) The AQUA-VOF code has a potential enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [Thermal Stratification Phenomena] (1) On-set position of thermal entrainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. On the other hand, the position in the case of CO 2 gas was shifted to upstream side with decreasing of Ri number. (2) Destruction speed of the thermal stratification interface was dependent on thermal diffusivity as fluid properties. Therefore it was concluded that an elimination method is necessary for the interface generated in CO 2 gas. [Thermal Striping Phenomena] (1) Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO 2 gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2) To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it is necessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phynomlena] (1) Fundamental behavior of the natural convection in various coolant follows buoyant jet
Energy Technology Data Exchange (ETDEWEB)
Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx
2006-07-01
In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the
A Thermo-Hydraulic Tool for Automatic Virtual Hazop Evaluation
Directory of Open Access Journals (Sweden)
Pugi L.
2014-12-01
Full Text Available Development of complex lubrication systems in the Oil&Gas industry has reached high levels of competitiveness in terms of requested performances and reliability. In particular, the use of HazOp (acronym of Hazard and Operability analysis represents a decisive factor to evaluate safety and reliability of plants. The HazOp analysis is a structured and systematic examination of a planned or existing operation in order to identify and evaluate problems that may represent risks to personnel or equipment. In particular, P&ID schemes (acronym of Piping and Instrument Diagram according to regulation in force ISO 14617 are used to evaluate the design of the plant in order to increase its safety and reliability in different operating conditions. The use of a simulation tool can drastically increase speed, efficiency and reliability of the design process. In this work, a tool, called TTH lib (acronym of Transient Thermal Hydraulic Library for the 1-D simulation of thermal hydraulic plants is presented. The proposed tool is applied to the analysis of safety relevant components of compressor and pumping units, such as lubrication circuits. Opposed to the known commercial products, TTH lib has been customized in order to ease simulation of complex interactions with digital logic components and plant controllers including their sensors and measurement systems. In particular, the proposed tool is optimized for fixed step execution and fast prototyping of Real Time code both for testing and production purposes. TTH lib can be used as a standard SimScape-Simulink library of components optimized and specifically designed in accordance with the P&ID definitions. Finally, an automatic code generation procedure has been developed, so TTH simulation models can be directly assembled from the P&ID schemes and technical documentation including detailed informations of sensor and measurement system.
International Nuclear Information System (INIS)
Hainoun, A.
1996-01-01
The ATHLET thermohydraulic code was developed at the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Society for Plant and Reactor Safety) to analyse leaks and transients for power reactors. In order to extend the code's range of application to the safety analysis of research reactors, a model was implemented permitting a description of the thermodynamic non-equilibrium effects in the subcooled boiling regime. The aim of the extension is, on one hand, to cover the thermohydraulic instability which is particularly characteristic of research reactors owing to their high power densities and low system pressures and, on the other hand, to provide a consideration of the influence of the steam formed in this boiling regime on the neutron balance. The model developed takes into consideration the competing evaporation and condensation effects in the subcooled boiling regime. It describes the bubble production rate at the superheated heating surfaces as well as the subsequent condensation of the bubbles in the subcooled core flow. The installed model is validated by the recalculation of two extensive series of experiments. In the first series the McMaster experiments on axial void distribution in the subcooled boiling regime are recalculated. The recalculation shows that the extended programme is capable of calculating the axial void distribution in the subcooled boiling regime with good agreement with the data. The second series deals with KFA experiments on thermohydraulic instability (flow excursion) in the subcooled boiling regime, comprising a broad parameter range of heat flow density, inlet temperature and channel width. Recalculation of this experimental series shows that the programme extension ensures simulation of thermohydraulic instability. (orig.)
THALES, Thermohydraulic LOCA Analysis of BWR and PWR
International Nuclear Information System (INIS)
ABE, Kiyoharu
1990-01-01
size. THALES has a lot of models some of which use the forward-explicit method for transient calculations. Even in cases where a more stable method is used in some models, interactions between models are solved with the forward - explicit method. If the time step sizes are too long, the calculated results will be inaccurate, especially for rapid transient
Implicit thermohydraulic coupling of two-phause flow calculations
International Nuclear Information System (INIS)
Lekach, S.; Kaufman, J.M.
1980-01-01
A numerical scheme that implicitly couples the hydraulic variables with thermal variables during a one or two-phase transient calculation in a one-dimensional pipe is presented. The transients are performed to achieve a steady-state condition. It is shown that by preserving the strong interdependence that exists between the hydraulic and thermal variables with an implicit flux treatment, it is possible to achieve a greater degree of numerical stability and in less computer time than with an explicit treatment. The method is slightly more complex but the large time step advantage more than pays for the overhead
International Nuclear Information System (INIS)
Ching, W-H; K H Leung, Michael; Leung, Dennis Y C
2009-01-01
Transient turbulent dispersion phenomena can be found in various practical problems, such as the accidental release of toxic chemical vapor and the airborne transmission of infectious droplets. Computational fluid dynamics (CFD) is an effective tool for analyzing such transient dispersion behaviors. However, the transient CFD analysis is often computationally expensive and time consuming. In the present study, a computationally efficient CFD-statistical hybrid modeling method has been developed for studying transient turbulent dispersion. In this method, the source emission is represented by emissions of many infinitesimal puffs. Statistical analysis is performed to obtain first the statistical properties of the puff trajectories and subsequently the most probable distribution of the puff trajectories that represent the macroscopic dispersion behaviors. In two case studies of ambient dispersion, the numerical modeling results obtained agree reasonably well with both experimental measurements and conventional k-ε modeling results published in the literature. More importantly, the proposed many-puff CFD-statistical hybrid modeling method effectively reduces the computational time by two orders of magnitude.
Two-dimensional computational modeling of high-speed transient flow in gun tunnel
Mohsen, A. M.; Yusoff, M. Z.; Hasini, H.; Al-Falahi, A.
2018-03-01
In this work, an axisymmetric numerical model was developed to investigate the transient flow inside a 7-meter-long free piston gun tunnel. The numerical solution of the gun tunnel was carried out using the commercial solver Fluent. The governing equations of mass, momentum, and energy were discretized using the finite volume method. The dynamic zone of the piston was modeled as a rigid body, and its motion was coupled with the hydrodynamic forces from the flow solution based on the six-degree-of-freedom solver. A comparison of the numerical data with the theoretical calculations and experimental measurements of a ground-based gun tunnel facility showed good agreement. The effects of parameters such as working gases and initial pressure ratio on the test conditions in the facility were examined. The pressure ratio ranged from 10 to 50, and gas combinations of air-air, helium-air, air-nitrogen, and air-CO2 were used. The results showed that steady nozzle reservoir conditions can be maintained for a longer duration when the initial conditions across the diaphragm are adjusted. It was also found that the gas combination of helium-air yielded the highest shock wave strength and speed, but a longer test time was achieved in the test section when using the CO2 test gas.
Directory of Open Access Journals (Sweden)
Diego Francisco Ledezma-Ramirez
2014-01-01
Full Text Available A theoretical control strategy for residual vibration control resulting from a shock pulse is studied. The semiactive control strategy is applied in a piecewise linear compound model and involves an on-off logic to connect and disconnect a secondary mass stiffness system from the primary isolation device, with the aim of providing high energy dissipation for lightly damped systems. The compound model is characterized by an energy dissipation mechanism due to the inelastic collision between the two masses and then viscous damping is introduced and its effects are analyzed. The objective of the simulations is to evaluate the transient vibration response in comparison to the results for a passive viscously damped single degree-of-freedom system considered as the benchmark or reference case. Similarly the decay in the compound system is associated with an equivalent decay rate or logarithmic decrement for direct comparison. It is found how the compound system provides improved isolation compared to the passive system, and the damping mechanisms are explained.
International Nuclear Information System (INIS)
Vermeul, Vince R; Cole, Charles R; Bergeron, Marcel P; Thorne, Paul D; Wurstner, Signe K
2001-01-01
The baseline three-dimensional transient inverse model for the estimation of site-wide scale flow parameters, including their uncertainties, using data on the transient behavior of the unconfined aquifer system over the entire historical period of Hanford operations, has been modified to account for the effects of basalt intercommunication between the Hanford unconfined aquifer and the underlying upper basalt confined aquifer. Both the baseline and alternative conceptual models (ACM-1) considered only the groundwater flow component and corresponding observational data in the 3-Dl transient inverse calibration efforts. Subsequent efforts will examine both groundwater flow and transport. Comparisons of goodness of fit measures and parameter estimation results for the ACM-1 transient inverse calibrated model with those from previous site-wide groundwater modeling efforts illustrate that the new 3-D transient inverse model approach will strengthen the technical defensibility of the final model(s) and provide the ability to incorporate uncertainty in predictions related to both conceptual model and parameter uncertainty
Thermohydraulic tests in the area of reactor safety done in CDTN
International Nuclear Information System (INIS)
Ladeira, L.C.D.
1990-01-01
The main experimental works performed in the last five years at the Thermohydraulics Laboratory of the Nuclear Technology Development Center, in the field of reactor safety are briefly described. This paper cover the performing and analysis of pressure drop, heat transfer and mixing tests in 3X3 rod bundle and rewetting tests in single tube section. (autor) [pt
Zarifakis, Marios; Coffey, William T.; Kalmykov, Yuri P.; Titov, Sergei V.
2017-06-01
An ever-increasing requirement to integrate greater amounts of electrical energy from renewable sources especially from wind turbines and solar photo-voltaic installations exists and recent experience in the island of Ireland demonstrates that this requirement influences the behaviour of conventional generating stations. One observation is the change in the electrical power output of synchronous generators following a transient disturbance especially their oscillatory behaviour accompanied by similar oscillatory behaviour of the grid frequency, both becoming more pronounced with reducing grid inertia. This behaviour cannot be reproduced with existing mathematical models indicating that an understanding of the behaviour of synchronous generators, subjected to various disturbances especially in a system with low inertia requires a new modelling technique. Thus two models of a generating station based on a double pendulum described by a system of coupled nonlinear differential equations and suitable for analysis of its stability corresponding to infinite or finite grid inertia are presented. Formal analytic solutions of the equations of motion are given and compared with numerical solutions. In particular the new finite grid model will allow one to identify limitations to the operational range of the synchronous generators used in conventional power generation and also to identify limits, such as the allowable Rate of Change of Frequency which is currently set to ± 0.5 Hz/s and is a major factor in describing the volatility of a grid as well as identifying requirements to the total inertia necessary, which is currently provided by conventional power generators only, thus allowing one to maximise the usage of grid connected non-synchronous generators, e.g., wind turbines and solar photo-voltaic installations.
Merlis, Timothy M.; Held, Isaac M.; Stenchikov, Georgiy L.; Zeng, Fanrong; Horowitz, Larry W.
2014-01-01
Coupled climate model simulations of volcanic eruptions and abrupt changes in CO2 concentration are compared in multiple realizations of the Geophysical Fluid Dynamics Laboratory Climate Model, version 2.1 (GFDL CM2.1). The change in global-mean surface temperature (GMST) is analyzed to determine whether a fast component of the climate sensitivity of relevance to the transient climate response (TCR; defined with the 1%yr-1 CO2-increase scenario) can be estimated from shorter-time-scale climate changes. The fast component of the climate sensitivity estimated from the response of the climate model to volcanic forcing is similar to that of the simulations forced by abrupt CO2 changes but is 5%-15% smaller than the TCR. In addition, the partition between the top-of-atmosphere radiative restoring and ocean heat uptake is similar across radiative forcing agents. The possible asymmetry between warming and cooling climate perturbations, which may affect the utility of volcanic eruptions for estimating the TCR, is assessed by comparing simulations of abrupt CO2 doubling to abrupt CO2 halving. There is slightly less (~5%) GMST change in 0.5 × CO2 simulations than in 2 × CO2 simulations on the short (~10 yr) time scales relevant to the fast component of the volcanic signal. However, inferring the TCR from volcanic eruptions is more sensitive to uncertainties from internal climate variability and the estimation procedure. The response of the GMST to volcanic eruptions is similar in GFDL CM2.1 and GFDL Climate Model, version 3 (CM3), even though the latter has a higher TCR associated with a multidecadal time scale in its response. This is consistent with the expectation that the fast component of the climate sensitivity inferred from volcanic eruptions is a lower bound for the TCR.
Merlis, Timothy M.
2014-10-01
Coupled climate model simulations of volcanic eruptions and abrupt changes in CO2 concentration are compared in multiple realizations of the Geophysical Fluid Dynamics Laboratory Climate Model, version 2.1 (GFDL CM2.1). The change in global-mean surface temperature (GMST) is analyzed to determine whether a fast component of the climate sensitivity of relevance to the transient climate response (TCR; defined with the 1%yr-1 CO2-increase scenario) can be estimated from shorter-time-scale climate changes. The fast component of the climate sensitivity estimated from the response of the climate model to volcanic forcing is similar to that of the simulations forced by abrupt CO2 changes but is 5%-15% smaller than the TCR. In addition, the partition between the top-of-atmosphere radiative restoring and ocean heat uptake is similar across radiative forcing agents. The possible asymmetry between warming and cooling climate perturbations, which may affect the utility of volcanic eruptions for estimating the TCR, is assessed by comparing simulations of abrupt CO2 doubling to abrupt CO2 halving. There is slightly less (~5%) GMST change in 0.5 × CO2 simulations than in 2 × CO2 simulations on the short (~10 yr) time scales relevant to the fast component of the volcanic signal. However, inferring the TCR from volcanic eruptions is more sensitive to uncertainties from internal climate variability and the estimation procedure. The response of the GMST to volcanic eruptions is similar in GFDL CM2.1 and GFDL Climate Model, version 3 (CM3), even though the latter has a higher TCR associated with a multidecadal time scale in its response. This is consistent with the expectation that the fast component of the climate sensitivity inferred from volcanic eruptions is a lower bound for the TCR.
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.
Directory of Open Access Journals (Sweden)
Nenad Marković
2016-01-01
Full Text Available The paper presents a new approach in the analysis of a transient state in a system where the feeding source is a transducer-IGBT inverter and load is introduced through the induction motor with its R-L parameters. Induction motors with different parameters of powers and power factors are tested. MATLAB simulation of the three-phase inverter that feeds the induction machine has replaced the missing lab equipment with which mathematical model of this system was verified. According to the selected parameters of the inverter and induction machine and through the simulation in the MATLAB program, the results are obtained in the form of diagrams that verify the model of a transient state of the induction machine operation when it operates as a motor which is presented as a variable R-L load. The transient process of the system three-phase bridge inverter whose active-inductive load is the induction machine in the conditions of the change of the load parameters is analyzed. The model of the transient process in the system formed by the inverter in PWM (Pulse Width Modulation converter and induction machine is developed in the time domain and phase coordinates.
International Nuclear Information System (INIS)
Gulshani, P.; So, C.B.
1986-10-01
In a number of postulated accident scenarios in a CANDU reactor, some of the horizontal fuel channels are predicted to experience periods of stratified channel coolant condition which can lead to a circumferential temperature gradient around the pressure tube. To study pressure tube strain and integrity under stratified flow channel conditions, it is, necessary to determine the pressure tube circumferential temperature distribution. This paper presents an algebraic model, called AMPTRACT (Algebraic Model for Pressure Tube TRAnsient Circumferential Temperature), developed to give the transient temperature distribution in a closed form. AMPTRACT models the following modes of heat transfer: radiation from the outermost elements to the pressure tube and from the pressure to calandria tube, convection between the fuel elements and the pressure tube and superheated steam, and circumferential conduction from the exposed to submerged part of the pressure tube. An iterative procedure is used to solve the mass and energy equations in closed form for axial steam and fuel-sheath transient temperature distributions. The one-dimensional conduction equation is then solved to obtain the pressure tube circumferential transient temperature distribution in a cosine series expansion. In the limit of large times and in the absence of convection and radiation to the calandria tube, the predicted pressure tube temperature distribution reduces identically to a parabolic profile. In this limit, however, radiation cannot be ignored because the temperatures are generally high. Convection and radiation tend to flatten the parabolic distribution
International Nuclear Information System (INIS)
Malone, R.C.; Pitcher, E.J.; Blackmon, M.L.; Puri, K.; Bourke, W.
1984-01-01
We examine the characteristics of stationary and transient eddies in the geopotential-height field as simulated by a spectral general circulation model. The model possessess a realistic, but smootheed, topography. Two simulations with perpetual January and July forcing by climatological sea surface temperatures, sea ice, and insolation were extended to 1200 days, of which the final 600 days were used for the results in this study. We find that the stationary waves are well simulated in both seasons in the Northern Hemisphere, where strong forcing by orography and land-sea thermal contrast exists. However, in the Southern Hemisphere, where no continents are present in midlatitudes, the stationary waves have smaller amplitude than that observed in both seasons. In both hemispheres, the transient eddies are well simulated in the winter season but are too weak in the summer season. The model fails to generate a sufficiently intense summertime midlatitude jet in either hemisphere, and this results in a low level of transient activity. The variance in the tropical troposphere is very well simulated. We examine the geographical distribution and vertical structure of the transient eddies. Fourier analysis in zonal wavenumber and temporal filtering are used to display the wavelength and frequency characteristics of the eddies
Deng, Baoqing; Si, Yinbing; Wang, Jia
2017-12-01
Transient storages may vary along the stream due to stream hydraulic conditions and the characteristics of storage. Analytical solutions of transient storage models in literature didn't cover the spatially non-uniform storage. A novel integral transform strategy is presented that simultaneously performs integral transforms to the concentrations in the stream and in storage zones by using the single set of eigenfunctions derived from the advection-diffusion equation of the stream. The semi-analytical solution of the multiple-zone transient storage model with the spatially non-uniform storage is obtained by applying the generalized integral transform technique to all partial differential equations in the multiple-zone transient storage model. The derived semi-analytical solution is validated against the field data in literature. Good agreement between the computed data and the field data is obtained. Some illustrative examples are formulated to demonstrate the applications of the present solution. It is shown that solute transport can be greatly affected by the variation of mass exchange coefficient and the ratio of cross-sectional areas. When the ratio of cross-sectional areas is big or the mass exchange coefficient is small, more reaches are recommended to calibrate the parameter.
The Coda of the Transient Response in a Sensitive Cochlea: A Computational Modeling Study.
Directory of Open Access Journals (Sweden)
Yizeng Li
2016-07-01
Full Text Available In a sensitive cochlea, the basilar membrane response to transient excitation of any kind-normal acoustic or artificial intracochlear excitation-consists of not only a primary impulse but also a coda of delayed secondary responses with varying amplitudes but similar spectral content around the characteristic frequency of the measurement location. The coda, sometimes referred to as echoes or ringing, has been described as a form of local, short term memory which may influence the ability of the auditory system to detect gaps in an acoustic stimulus such as speech. Depending on the individual cochlea, the temporal gap between the primary impulse and the following coda ranges from once to thrice the group delay of the primary impulse (the group delay of the primary impulse is on the order of a few hundred microseconds. The coda is physiologically vulnerable, disappearing when the cochlea is compromised even slightly. The multicomponent sensitive response is not yet completely understood. We use a physiologically-based, mathematical model to investigate (i the generation of the primary impulse response and the dependence of the group delay on the various stimulation methods, (ii the effect of spatial perturbations in the properties of mechanically sensitive ion channels on the generation and separation of delayed secondary responses. The model suggests that the presence of the secondary responses depends on the wavenumber content of a perturbation and the activity level of the cochlea. In addition, the model shows that the varying temporal gaps between adjacent coda seen in experiments depend on the individual profiles of perturbations. Implications for non-invasive cochlear diagnosis are also discussed.
Simard, Patrick Tremblay; Chesnaux, Romain; Rouleau, Alain; Daigneault, Réal; Cousineau, Pierre A.; Roy, Denis W.; Lambert, Mélanie; Poirier, Brigitte; Poignant-Molina, Léo
2015-08-01
Aquifer formations along the northern shore of the Saint-Lawrence River in Quebec (Canada) mainly consist of glacial and coastal deposits of variable thickness overlying Precambrian bedrock. These deposits are important because they provide the main water supply for many communities. As part of a continuing project aimed at developing an inventory of the groundwater resources in the Charlevoix and Haute-Côte-Nord (CHCN) regions of the province of Quebec in Canada, the central loop transient electromagnetic (TEM) method was used to map the principal hydrogeological environments in these regions. One-dimensional smooth inversion models of the TEM soundings have been used to construct two-dimensional electrical resistivity sections, which provided images for hydrogeological validation. Electrical contour lines of aquifer environments were compared against available well logs and Quaternary surface maps in order to interpret TEM soundings. A calibration table was achieved to represent common deposits and basements. The calibration table was then exported throughout the CHCN region. This paper presents three case studies; one in the Forestville site, another in the Les Escoumins site and the other in the Saint-Urbain site. These sites were selected as targets for geophysical surveys because of the general lack of local direct hydrogeological data related to them.
Energy Technology Data Exchange (ETDEWEB)
Baptista, Renan Martins [PETROBRAS, Rio de Janeiro, RJ (Brazil). Centro de Pesquisas. Div. de Explotacao]. E-mail: renan@cenpes.petrobras.com.br
2000-07-01
This paper describes a technical procedure to assess a software based leak detection system (LDS), by deciding between a simpler low cost, less effective product, having a fast installation and tuning, and a complex one with high cost and efficiency, which however takes a long time to be properly installed. This is a common decision among the pipeline operating companies, considering that the majority of the lines are short, with single phase liquid flow (which may include batches), basic communication system and instrumentation. Service companies offer realistic solutions for liquid flow, but usually designed to big pipeline networks, flowing multiple batches and allowing multiple fluid entrances and deliveries. Those solutions are sometimes impractical to short pipelines, due to its high cost, as well as long tuning procedures, complex instrumentation, communication and computer requirements. It is intended to approach here the best solution according to its cost. In a practical sense, it means to differentiate the various LDS techniques. Those techniques are available in a considerable number, and they are still spreading, according to the different scenarios. However, two most known and worldwide implemented techniques hold the majority of the market: the Compensated Volume Balance (CVB), which is less accurate, reliable and robust, but cheaper, simpler and faster to install, and the Real Time Transient Model (RTTM), which is very reliable, accurate and robust, but expensive and complex. This work will describe a way to define whether one can use or not a CVB in a pipeline. (author)
Ngau, Julie L.; Griffin, Peter B.; Plummer, James D.
2001-08-01
Recent work has indicated that the suppression of boron transient enhanced diffusion (TED) in carbon-rich Si is caused by nonequilibrium Si point defect concentrations, specifically the undersaturation of Si self-interstitials, that result from the coupled out-diffusion of carbon interstitials via the kick-out and Frank-Turnbull reactions. This study of boron TED reduction in Si1-x-yGexCy during 750 °C inert anneals has revealed that the use of an additional reaction that further reduces the Si self-interstitial concentration is necessary to describe accurately the time evolved diffusion behavior of boron. In this article, we present a comprehensive model which includes {311} defects, boron-interstitial clusters, a carbon kick-out reaction, a carbon Frank-Turnbull reaction, and a carbon interstitial-carbon substitutional (CiCs) pairing reaction that successfully simulates carbon suppression of boron TED at 750 °C for anneal times ranging from 10 s to 60 min.
International Nuclear Information System (INIS)
Ngau, Julie L.; Griffin, Peter B.; Plummer, James D.
2001-01-01
Recent work has indicated that the suppression of boron transient enhanced diffusion (TED) in carbon-rich Si is caused by nonequilibrium Si point defect concentrations, specifically the undersaturation of Si self-interstitials, that result from the coupled out-diffusion of carbon interstitials via the kick-out and Frank--Turnbull reactions. This study of boron TED reduction in Si 1-x-y Ge x C y during 750 o C inert anneals has revealed that the use of an additional reaction that further reduces the Si self-interstitial concentration is necessary to describe accurately the time evolved diffusion behavior of boron. In this article, we present a comprehensive model which includes {311} defects, boron-interstitial clusters, a carbon kick-out reaction, a carbon Frank--Turnbull reaction, and a carbon interstitial-carbon substitutional (C i C s ) pairing reaction that successfully simulates carbon suppression of boron TED at 750 o C for anneal times ranging from 10 s to 60 min. copyright 2001 American Institute of Physics
Low dimensional equivalence of core neutronics model and its application to transient analysis
International Nuclear Information System (INIS)
Song Hongbing; Zhao Fuyu
2015-01-01
Three-dimensional coupled neutronics thermal-hydraulics reactor analysis is time consuming and occupies huge memory. A one-dimensional model is preferable than the three one in nuclear system analysis, control system design and load following. In this paper, a corewide three dimensional to one dimensional equivalent method has been developed. On the basis of this method 1D axial few groups constants were obtained. The equivalent cross sections were calculated by general spatial homogenization while the transverse buckling was computed through an equivalence based on the 3D flux conservation. Three steady test cases were performed on one dimensional finite difference code ODTAC and the results were compared with TRIVAC-5. The comparison shows that the one dimensional axial power distribution computed by ODTAC correlates well with the three dimensional results calculated by TRIVAC-5. In this study, DRAGON-4 was used to generate the few-group constants of fuel assemblies and the reflector few-group parameters were calculated by WIMS-D4. These collapsed few-group constants were tabulated in a database sorted in ascending order of fuel temperature, coolant temperature and concentration of boric acid. Trilinear interpolation was adopted in cross sections feedback during the transient analysis. In this paper, G1 rod drop accident (RDA) and G1 rod ejection accident (REA) were performed on ODTAC and the computation results were consistent of the physical rules. (author)
Jazebi, Saeed
This thesis is a step forward toward achieving the final objective of creating a fully dual model for transformers including eddy currents and nonlinearities of the iron core using the fundamental electrical components already available in the EMTP-type programs. The model is effective for the study of the performance of transformers during power system transients. This is very important for transformer designers, because the insulation of transformers is determined with the overvoltages caused by lightning or switching operations. There are also internally induced transients that occur when a switch is actuated. For example switching actions for reconfiguration of distribution systems that offers economic advantages, or protective actions to clear faults and large short-circuit currents. Many of the smart grid concepts currently under development by many utilities rely heavily on switching to optimize resources that produce transients in the system. On the other hand, inrush currents produce mechanical forces which deform transformer windings and cause malfunction of the differential protection. Also, transformer performance under ferroresonance and geomagnetic induced currents are necessary to study. In this thesis, a physically consistent dual model applicable to single-phase two-winding transformers is proposed. First, the topology of a dual electrical equivalent circuit is obtained from the direct application of the principle of duality. Then, the model parameters are computed considering the variations of the transformer electromagnetic behavior under various operating conditions. Current modeling techniques use different topological models to represent diverse transient situations. The reversible model proposed in this thesis unifies the terminal and topological equivalent circuits. The model remains invariable for all low-frequency transients including deep saturation conditions driven from any of the two windings. The very high saturation region of the
Kammoun, Soulaymen; Sallem, Souhir; Ben Ali Kammoun, Mohamed
2017-11-01
The aim of this study is to enhance DFIG based Wind Energy Conversion Systems (WECS) dynamics during grid coupling. In this paper, a system modelling and a starting/coupling procedure for this generator to the grid are proposed. The proposed non-linear system is a variable structure system (VSS) and has two different states, before and after coupling. So, two different state models are given to the system to analyse transient stability during the coupling. The given model represents well the transient state of the machine, through which, a behaviour assessment of the generator before, during and after connection is given based on simulation results. For this, a 300 kW DFIG based wind generation system model was simulated on the Matlab/SIMULINK environment. We judge the proposed procedure to be practical, smooth and stability improved.
International Nuclear Information System (INIS)
Chandra, V.K.; Chandra, B.P.; Tiwari, M.; Baghel, R.N.; Ramrakhiani, M.
2012-01-01
expressions are derived for the transient EL intensity I, time t m and intensity I m corresponding to the peak of EL overshoot, total EL intensity I t and decay of the intensity of EL overshoot. In fact, the decay prior to the onset of EL overshoot is the decay of number of electrons moving in the electron transporting states. The ratio I m /I s decreases with increasing value of the applied pulse voltage because I m increases linearly with the amplitude of applied voltage pulse and I s increases nonlinearly and rapidly with the increasing amplitude of applied voltage pulse. The lifetime τ t of electrons at the interface decreases with increasing temperature whereby the dependence of τ t on temperature follows Arrhenius plot. This fact indicates that the detrapping involves thermally-assisted tunneling of electrons. Using the EL overshoot in bilayer OLEDs, the lifetime of the charge carriers at the interface, recombination time of charge carriers, decay time of the EL prior to onset of overshoot, and the time delay between the voltage pulse and onset time of the EL overshoot can be determined. The intense EL overshoot of nanosecond or shorter time duration may be useful in digital communication, and moreover, the EL overshoot gives important information about the processes involving injection, transport and recombination of charge carriers. The criteria for appearance of EL overshoot in bilayer OLEDs are explored. A good agreement is found between the theoretical and experimental results. - Highlights: ► Modeling is done for transient electroluminescence overshoot in bilayer OLEDs. ► Overshoot in spin-coated bilayer OLEDs is caused by the presence of an interfacial layer. ► Expressions are derived for the transient EL intensity and decay of EL in OLEDs. ► A good agreement is found between the theoretical and experimental results.
Modeling of transient ionizing radiation effects in bipolar devices at high dose-rates
International Nuclear Information System (INIS)
FJELDLY, T.A.; DENG, Y.; SHUR, M.S.; HJALMARSON, HAROLD P.; MUYSHONDT, ARNOLDO
2000-01-01
To optimally design circuits for operation at high intensities of ionizing radiation, and to accurately predict their a behavior under radiation, precise device models are needed that include both stationary and dynamic effects of such radiation. Depending on the type and intensity of the ionizing radiation, different degradation mechanisms, such as photoelectric effect, total dose effect, or single even upset might be dominant. In this paper, the authors consider the photoelectric effect associated with the generation of electron-hole pairs in the semiconductor. The effects of low radiation intensity on p-II diodes and bipolar junction transistors (BJTs) were described by low-injection theory in the classical paper by Wirth and Rogers. However, in BJTs compatible with modem integrated circuit technology, high-resistivity regions are often used to enhance device performance, either as a substrate or as an epitaxial layer such as the low-doped n-type collector region of the device. Using low-injection theory, the transient response of epitaxial BJTs was discussed by Florian et al., who mainly concentrated on the effects of the Hi-Lo (high doping - low doping) epilayer/substrate junction of the collector, and on geometrical effects of realistic devices. For devices with highly resistive regions, the assumption of low-level injection is often inappropriate, even at moderate radiation intensities, and a more complete theory for high-injection levels was needed. In the dynamic photocurrent model by Enlow and Alexander. p-n junctions exposed to high-intensity radiation were considered. In their work, the variation of the minority carrier lifetime with excess carrier density, and the effects of the ohmic electric field in the quasi-neutral (q-n) regions were included in a simplified manner. Later, Wunsch and Axness presented a more comprehensive model for the transient radiation response of p-n and p-i-n diode geometries. A stationary model for high-level injection in p
Identifying the optimal supply temperature in district heating networks - A modelling approach
DEFF Research Database (Denmark)
Mohammadi, Soma; Bojesen, Carsten
2014-01-01
of this study is to develop a model for thermo-hydraulic calculation of low temperature DH system. The modelling is performed with emphasis on transient heat transfer in pipe networks. The pseudo-dynamic approach is adopted to model the District Heating Network [DHN] behaviour which estimates the temperature...... dynamically while the flow and pressure are calculated on the basis of steady state conditions. The implicit finite element method is applied to simulate the transient temperature behaviour in the network. Pipe network heat losses, pressure drop in the network and return temperature to the plant...... are calculated in the developed model. The model will serve eventually as a basis to find out the optimal supply temperature in an existing DHN in later work. The modelling results are used as decision support for existing DHN; proposing possible modifications to operate at optimal supply temperature....
Forcada, J.; Royle, J. Andrew; Staniland, I.J.
2009-01-01
Correctly quantifying the impacts of rare apex marine predators is essential to ecosystem-based approaches to fisheries management, where harvesting must be sustainable for targeted species and their dependent predators. This requires modelling the uncertainty in such processes as predator life history, seasonal abundance and movement, size-based predation, energetic requirements, and prey vulnerability. We combined these uncertainties to evaluate the predatory impact of transient leopard seals on a community of mesopredators (seals and penguins) and their prey at South Georgia, and assess the implications for an ecosystem-based management. The mesopredators are highly dependent on Antarctic krill and icefish, which are targeted by regional fisheries. We used a state-space formulation to combine (1) a mark-recapture open-population model and individual identification data to assess seasonally variable leopard seal arrival and departure dates, numbers, and residency times; (2) a size-based bioenergetic model; and (3) a size-based prey choice model from a diet analysis. Our models indicated that prey choice and consumption reflected seasonal changes in leopard seal population size and structure, size-selective predation and prey vulnerability. A population of 104 (90?125) leopard seals, of which 64% were juveniles, consumed less than 2% of the Antarctic fur seal pup production of the area (50% of total ingested energy, IE), but ca. 12?16% of the local gentoo penguin population (20% IE). Antarctic krill (28% IE) were the only observed food of leopard seal pups and supplemented the diet of older individuals. Direct impacts on krill and fish were negligible, but the ?escapement? due to leopard seal predation on fur seal pups and penguins could be significant for the mackerel icefish fishery at South Georgia. These results suggest that: (1) rare apex predators like leopard seals may control, and may depend on, populations of mesopredators dependent on prey species
Transient modelling of heat loading of phase change material for energy storage
Directory of Open Access Journals (Sweden)
Asyraf W.M.
2017-01-01
Full Text Available As the development of solar energy is getting advance from time to time, the concentration solar technology also get the similar attention from the researchers all around the globe. This technology concentrate a large amount of energy into main spot. To collect all the available energy harvest from the solar panel, a thermal energy storage is required to convert the heat energy to one of the purpose such as electrical energy. With the idea of energy storage application that can be narrow down to commercial application such as cooking stove. Using latent heat type energy storage seem to be appropriate with the usage of phase change material (PCM that can release and absorb heat energy at nearly constant temperature by changing its state. Sodium nitrate (NaNO3 and potassium nitrate (KNO3 was selected to use as PCM in this project. This paper focus on the heat loading process and the melting process of the PCM in the energy storage using a computer simulation. The model of the energy storage was created as solid three dimensional modelling using computer aided software and the geometry size of it depend on how much it can apply to boil 1 kg of water in cooking application. The materials used in the tank, heat exchanger and the heat transfer fluid are stainless steel, copper and XCELTHERM MK1, respectively. The analysis was performed using a commercial simulation software in a transient state. The simulation run on different value of velocity but kept controlled under laminar state only, then the relationship of velocity and heat distribution was studied and the melting process of the PCM also has been analyzed. On the effect of heat transfer fluid velocity, the higher the velocity resulted in higher the rate of heat transfer. The comparison between the melting percentages of the PCMs under test conditions show that NaNO3 melts quite faster than KNO3.
Energy Technology Data Exchange (ETDEWEB)
Rian, Kjell Erik
2003-07-01
In numerical simulations of turbulent reacting compressible flows, artificial boundaries are needed to obtain a finite computational domain when an unbounded physical domain is given. Artificial boundaries which fluids are free to cross are called open boundaries. When calculating such flows, non-physical reflections at the open boundaries may occur. These reflections can pollute the solution severely, leading to inaccurate results, and the generation of spurious fluctuations may even cause the numerical simulation to diverge. Thus, a proper treatment of the open boundaries in numerical simulations of turbulent reacting compressible flows is required to obtain a reliable solution for realistic conditions. A local quasi-one-dimensional characteristic-based open-boundary treatment for the Favre-averaged governing equations for time-dependent three-dimensional multi-component turbulent reacting compressible flow is presented. A k-{epsilon} model for turbulent compressible flow and Magnussen's EDC model for turbulent combustion is included in the analysis. The notion of physical boundary conditions is incorporated in the method, and the conservation equations themselves are applied on the boundaries to complement the set of physical boundary conditions. A two-dimensional finite-difference-based computational fluid dynamics code featuring high-order accurate numerical schemes was developed for the numerical simulations. Transient numerical simulations of the well-known, one-dimensional shock-tube problem, a two-dimensional pressure-tower problem in a decaying turbulence field, and a two-dimensional turbulent reacting compressible flow problem have been performed. Flow- and combustion-generated pressure waves seem to be well treated by the non-reflecting subsonic open-boundary conditions. Limitations of the present open-boundary treatment are demonstrated and discussed. The simple and solid physical basis of the method makes it both favourable and relatively easy to
Calculation of coolant temperature sensitivity related to thermohydraulic parameters
International Nuclear Information System (INIS)
Silva, F.C. da; Andrade Lima, F.R. de
1985-01-01
It is verified the viability to apply the generalized Perturbation Theory (GPT) in the calculation of sensitivity for thermal-hydraulic problems. It was developed the TEMPERA code in FORTRAN-IV to transient calculations in the axial temperature distribution in a channel of PWR reactor and the associated importance function, as well as effects of variations of thermalhydraulic parameters in the coolant temperature. The results are compared with one which were obtained by direct calculation. (M.C.K.) [pt
International Nuclear Information System (INIS)
Choi, J.D.; Yoon, H.Y.; Um, K.S.; Kim, H.C.; Sim, S.K.
1997-01-01
Best estimate transient analysis code, TASS, has been developed for the normal and transient simulation of the Westinghouse and ABB-CE type PWRs. TASS thermal hydraulic model is based on the non-homogeneous, non-equilibrium two-phase continuity, energy and mixture momentum equations with constitutive relations for closure. Core neutronics model employs both the point kinetics and one-dimensional neutron diffusion model. Semi-implicit numerical scheme is used to solve the discretized finite difference equations. TASS one dimensional neutronics core model has been assessed through the reactivity induced transient analyses for the KORI-3, three loop Westinghouse PWR, and Younggwang-3 (YGN-3), two-loop ABB-CE PWR, nuclear power plants currently operating in Korea. The assessment showed that the TASS one dimensional neutronics core model can be applied for the Westinghouse and ABB-CE type PWRs to gain thermal margin which is necessary for a potential use of the high fuel burnup, extended fuel cycle, power upgrading and for the plant life extension
International Nuclear Information System (INIS)
Silva, Alice Cunha da; Su, Jian
2013-01-01
The High Temperature Gas cooled Reactor (HTGR) is a fourth generation thermal nuclear reactor, graphite-moderated and helium cooled. The HTGRs have important characteristics making essential the study of these reactors, as well as its fuel element. Examples of these are: high thermal efficiency,low operating costs and construction, passive safety attributes that allow implication of the respective plants. The Pebble Bed Modular Reactor (PBMR) is a HTGR with spherical fuel elements that named the reactor. This fuel element is composed by a particulate region with spherical inclusions, the fuel UO2 particles, dispersed in a graphite matrix and a convective heat transfer by Helium happens on the outer surface of the fuel element. In this work, the transient heat conduction in a spherical fuel element of a pebble-bed high temperature reactor was studied in a transient situation of combined convective and radiative cooling. Improved lumped parameter model was developed for the transient heat conduction in the two-layer composite sphere subjected to combined convective and radiative cooling. The improved lumped model was obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of the two-layer spherical fuel element was analyzed to illustrate the applicability of the proposed lumped model, with respect to die rent values of the Biot number, the radiation-conduction parameter, the dimensionless thermal contact resistance, the dimensionless inner diameter and coating thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the improved lumped model, with H2,1/H1,1/H0,0 approximation yielded significant improvement of average temperature prediction over the classical lumped model. (author)
A transient one-dimensional numerical model for kinetic Stirling engine
International Nuclear Information System (INIS)
Wang, Kai; Dubey, Swapnil; Choo, Fook Hoong; Duan, Fei
2016-01-01
Highlights: • A non-equilibrium thermal mode with considering loses is adopted in Stirling engine. • Good agreements are achieved for predicting various critical system parameters. • Differences between helium and hydrogen systems are highlighted and analyzed. • Pressure drop of helium system is much larger and more sensitive to frequency. - Abstract: A third-order numerical model based on one-dimensional computational fluid dynamics is developed for kinetic Stirling engines. Various loss mechanisms in Stirling engines, including gas spring hysteresis loss, shuttle loss, appendix displacer gap loss, gas leakage loss, finite speed loss, piston friction loss, pressure drop loss, heat conduction loss, mechanical loss and imperfect heat transfer, are considered and embedded into the basic control equations. The non-equilibrium thermal model is adopted for the regenerator to capture the oscillating features of the gas and solid temperatures. To improve the numerical stability and accuracy, the implicit second-order time difference scheme and the second-order upwind scheme are adopted for discretizing the time differential terms and convective terms, respectively. Experimental validations are then conducted on a beta-type Stirling engine with the extensive experimental data for diverse working conditions. The results show that the developed model has better accuracies than the previous second-order models. Good agreements are achieved for predicting various critical system parameters, including pressure-volume diagram, indicated power, brake power, indicated efficiency, brake efficiency and mechanical efficiency. In particular, both the experiments and simulations show that the Stirling engine charged with helium tends to have much lower optimal working frequencies and poorer performances compared to the hydrogen system. Based on the analyses of the losses, it reveals that the pressure drop in the flow channels plays a critical role in shaping the different
Vannametee, E.; Karssenberg, D.; Hendriks, M. R.; de Jong, S. M.; Bierkens, M. F. P.
2010-05-01
representation of the transient rainfall-runoff relations that exist in the artificial data set generated with the high-resolution model. The third step is to find the values of empirical parameters in the lumped forward model using the artificial dataset. For each scenario of the high-resolution model run, a set of lumped model parameters is determined with a fitting method using the corresponding time series of state variables and outputs retrieved from the database. Thus, the parameters in the lumped model can be estimated by using the artificial data set. The fourth step is to develop an approach to assign lumped model parameters based upon the properties of the geomorphological unit. This is done by finding relationships between the measurable physical properties of geomorphologic units (i.e. slope gradient, unit length, and regolith properties) and the lumped forward model parameters using multiple regression techniques. In this way, a set of lumped forward model parameters can be estimated as a function of morphology and physical properties of the geomorphologic units. The lumped forward model can then be applied to different geomorphologic units. Finally, the performance of the lumped forward model is evaluated; the outputs of the lumped forward model are compared with the results of the high-resolution model. Our results show that the lumped forward model gives the best estimates of total discharge volumes and peak discharges when rain intensities are not significantly larger than the infiltration capacities of the units and when the units are small with a flat gradient. Hydrograph shapes are fairly well reproduced for most cases except for flat and elongated units with large runoff volumes. The results of this study provide a first step towards developing low-dimensional models for large ungauged basins.
Eakins, D. E.; Thadhani, N. N.
2006-10-01
Instrumented Taylor anvil-on-rod impact tests have been conducted on oxygen-free electronic copper to validate the accuracy of current strength models for predicting transient states during dynamic deformation events. The experiments coupled the use of high-speed digital photography to record the transient deformation states and laser interferometry to monitor the sample back (free surface) velocity as a measure of the elastic/plastic wave propagation through the sample length. Numerical continuum dynamics simulations of the impact and plastic wave propagation employing the Johnson-Cook [Proceedings of the Seventh International Symposium on Ballistics, 1983, The Netherlands (Am. Def. Prep. Assoc. (ADPA)), pp. 541-547], Zerilli-Armstrong [J. Appl. Phys. C1, 1816 (1987)], and Steinberg-Guinan [J. Appl. Phys. 51, 1498 (1980)] constitutive equations were used to generate transient deformation profiles and the free surface velocity traces. While these simulations showed good correlation with the measured free surface velocity traces and the final deformed sample shape, varying degrees of deviations were observed between the photographed and calculated specimen profiles at intermediate deformation states. The results illustrate the usefulness of the instrumented Taylor anvil-on-rod impact technique for validating constitutive equations that can describe the path-dependent deformation response and can therefore predict the transient and final deformation states.
Thermophysics modeling of an infrared detector cryochamber for transient operational scenario
Singhal, Mayank; Singhal, Gaurav; Verma, Avinash C.; Kumar, Sushil; Singh, Manmohan
2016-05-01
An infrared detector (IR) is essentially a transducer capable of converting radiant energy in the infrared regime into a measurable form. The benefit of infrared radiation is that it facilitates viewing objects in dark or through obscured conditions by detecting the infrared energy emitted by them. One of the most significant applications of IR detector systems is for target acquisition and tracking of projectile systems. IR detectors also find widespread applications in the industry and commercial market. The performance of infrared detector is sensitive to temperatures and performs best when cooled to cryogenic temperatures in the range of nearly 120 K. However, the necessity to operate in such cryogenic regimes increases the complexity in the application of IR detectors. This entails a need for detailed thermophysics analysis to be able to determine the actual cooling load specific to the application and also due to its interaction with the environment. This will enable design of most appropriate cooling methodologies suitable for specific scenarios. The focus of the present work is to develop a robust thermo-physical numerical methodology for predicting IR cryochamber behavior under transient conditions, which is the most critical scenario, taking into account all relevant heat loads including radiation in its original form. The advantage of the developed code against existing commercial software (COMSOL, ANSYS, etc.), is that it is capable of handling gas conduction together with radiation terms effectively, employing a ubiquitous software such as MATLAB. Also, it requires much smaller computational resources and is significantly less time intensive. It provides physically correct results enabling thermal characterization of cryochamber geometry in conjunction with appropriate cooling methodology. The code has been subsequently validated experimentally as the observed cooling characteristics are found to be in close agreement with the results predicted using
SCANAIR a transient fuel performance code Part two: Assessment of modelling capabilities
Energy Technology Data Exchange (ETDEWEB)
Georgenthum, Vincent, E-mail: vincent.georgenthum@irsn.fr; Moal, Alain; Marchand, Olivier
2014-12-15
Highlights: • The SCANAIR code is devoted to the study of irradiated fuel rod behaviour during RIA. • The paper deals with the status of the code validation for PWR rods. • During the PCMI stage there is a good agreement between calculations and experiments. • The boiling crisis occurrence is rather well predicted. • The code assessment during the boiling crisis has still to be improved. - Abstract: In the frame of their research programmes on fuel safety, the French Institut de Radioprotection et de Sûreté Nucléaire develops the SCANAIR code devoted to the study of irradiated fuel rod behaviour during reactivity initiated accident. A first paper was focused on detailed modellings and code description. This second paper deals with the status of the code validation for pressurised water reactor rods performed thanks to the available experimental results. About 60 integral tests carried out in CABRI and NSRR experimental reactors and 24 separated tests performed in the PATRICIA facility (devoted to the thermal-hydraulics study) have been recalculated and compared to experimental data. During the first stage of the transient, the pellet clad mechanical interaction phase, there is a good agreement between calculations and experiments: the clad residual elongation and hoop strain of non failed tests but also the failure occurrence and failure enthalpy of failed tests are correctly calculated. After this first stage, the increase of cladding temperature can lead to the Departure from Nucleate Boiling. During the film boiling regime, the clad temperature can reach a very high temperature (>700 °C). If the boiling crisis occurrence is rather well predicted, the calculation of the clad temperature and the clad hoop strain during this stage have still to be improved.
Ding, Ming; Zhu, Qianlong
2016-01-01
Hardware protection and control action are two kinds of low voltage ride-through technical proposals widely used in a permanent magnet synchronous generator (PMSG). This paper proposes an innovative clustering concept for the equivalent modeling of a PMSG-based wind power plant (WPP), in which the impacts of both the chopper protection and the coordinated control of active and reactive powers are taken into account. First, the post-fault DC link voltage is selected as a concentrated expression of unit parameters, incoming wind and electrical distance to a fault point to reflect the transient characteristics of PMSGs. Next, we provide an effective method for calculating the post-fault DC link voltage based on the pre-fault wind energy and the terminal voltage dip. Third, PMSGs are divided into groups by analyzing the calculated DC link voltages without any clustering algorithm. Finally, PMSGs of the same group are equivalent as one rescaled PMSG to realize the transient equivalent modeling of the PMSG-based WPP. Using the DIgSILENT PowerFactory simulation platform, the efficiency and accuracy of the proposed equivalent model are tested against the traditional equivalent WPP and the detailed WPP. The simulation results show the proposed equivalent model can be used to analyze the offline electromechanical transients in power systems.
International Nuclear Information System (INIS)
Dubrovskij, E.V.; Vasil'ev, V.Ya.
2002-01-01
One introduces a technique to compare relatively thermohydraulic efficiency of heat transfer intensification in channels of heat exchange surfaces of any design types. It is shown that one should compare thermohydraulic efficiency of heat exchange intensification as to the thermal power of heat exchangers and pressure losses in channels with turbulators and in polished channels of heat exchange surfaces on the basis of dimensions of heat exchangers, their heat exchange surfaces and at similar (as to Re numbers) modes of coolant flow [ru
Transformer modeling for low- and mid-frequency electromagnetic transients simulation
Lambert, Mathieu
In this work, new models are developed for single-phase and three-phase shell-type transformers for the simulation of low-frequency transients, with the use of the coupled leakage model. This approach has the advantage that it avoids the use of fictitious windings to connect the leakage model to a topological core model, while giving the same response in short-circuit as the indefinite admittance matrix (BCTRAN) model. To further increase the model sophistication, it is proposed to divide windings into coils in the new models. However, short-circuit measurements between coils are never available. Therefore, a novel analytical method is elaborated for this purpose, which allows the calculation in 2-D of short-circuit inductances between coils of rectangular cross-section. The results of this new method are in agreement with the results obtained from the finite element method in 2-D. Furthermore, the assumption that the leakage field is approximately 2-D in shell-type transformers is validated with a 3-D simulation. The outcome of this method is used to calculate the self and mutual inductances between the coils of the coupled leakage model and the results are showing good correspondence with terminal short-circuit measurements. Typically, leakage inductances in transformers are calculated from short-circuit measurements and the magnetizing branch is calculated from no-load measurements, assuming that leakages are unimportant for the unloaded transformer and that magnetizing current is negligible during a short-circuit. While the core is assumed to have an infinite permeability to calculate short-circuit inductances, and it is a reasonable assumption since the core's magnetomotive force is negligible during a short-circuit, the same reasoning does not necessarily hold true for leakage fluxes in no-load conditions. This is because the core starts to saturate when the transformer is unloaded. To take this into account, a new analytical method is developed in this
Sensitivity calculation of the coolant temperature regarding the thermohydraulic parameters
International Nuclear Information System (INIS)
Andrade Lima, F.R. de; Silva, F.C. da; Thome Filho, Z.D.; Alvim, A.C.M.; Oliveira Barroso, A.C. de.
1985-01-01
It's studied the application of the Generalized Perturbation Theory (GPT) in the sensitivity calculation of thermalhydraulic problems, aiming at verifying the viability of the extension of the method. For this, the axial distribution, transient, of the coolant temperature in a PWR channel are considered. Perturbation expressions are developed using the GPT formalism, and a computer code (Tempera) is written, to calculate the channel temperature distribution and the associated importance function, as well as the effect of the thermalhydraulic parameters variations in the coolant temperature (sensitivity calculation). The results are compared with those from the direct calculation. (E.G.) [pt
Baum, Rex L.; Godt, Jonathan W.; Savage, William Z.
2010-01-01
Shallow rainfall-induced landslides commonly occur under conditions of transient infiltration into initially unsaturated soils. In an effort to predict the timing and location of such landslides, we developed a model of the infiltration process using a two-layer system that consists of an unsaturated zone above a saturated zone and implemented this model in a geographic information system (GIS) framework. The model links analytical solutions for transient, unsaturated, vertical infiltration above the water table to pressure-diffusion solutions for pressure changes below the water table. The solutions are coupled through a transient water table that rises as water accumulates at the base of the unsaturated zone. This scheme, though limited to simplified soil-water characteristics and moist initial conditions, greatly improves computational efficiency over numerical models in spatially distributed modeling applications. Pore pressures computed by these coupled models are subsequently used in one-dimensional slope-stability computations to estimate the timing and locations of slope failures. Applied over a digital landscape near Seattle, Washington, for an hourly rainfall history known to trigger shallow landslides, the model computes a factor of safety for each grid cell at any time during a rainstorm. The unsaturated layer attenuates and delays the rainfall-induced pore-pressure response of the model at depth, consistent with observations at an instrumented hillside near Edmonds, Washington. This attenuation results in realistic estimates of timing for the onset of slope instability (7 h earlier than observed landslides, on average). By considering the spatial distribution of physical properties, the model predicts the primary source areas of landslides.
A neural model for transient identification in dynamic processes with 'don't know' response
International Nuclear Information System (INIS)
Mol, Antonio C. de A.; Martinez, Aquilino S.; Schirru, Roberto
2003-01-01
This work presents an approach for neural network based transient identification which allows either dynamic identification or a 'don't know' response. The approach uses two 'jump' multilayer neural networks (NN) trained with the backpropagation algorithm. The 'jump' network is used because it is useful to dealing with very complex patterns, which is the case of the space of the state variables during some abnormal events. The first one is responsible for the dynamic identification. This NN uses, as input, a short set (in a moving time window) of recent measurements of each variable avoiding the necessity of using starting events. The other one is used to validate the instantaneous identification (from the first net) through the validation of each variable. This net is responsible for allowing the system to provide a 'don't know' response. In order to validate the method, a Nuclear Power Plant (NPP) transient identification problem comprising 15 postulated accidents, simulated for a pressurized water reactor (PWR), was proposed in the validation process it has been considered noisy data in order to evaluate the method robustness. Obtained results reveal the ability of the method in dealing with both dynamic identification of transients and correct 'don't know' response. Another important point studied in this work is that the system has shown to be independent of a trigger signal which indicates the beginning of the transient, thus making it robust in relation to this limitation
International Nuclear Information System (INIS)
Martin, L.; Saenz Tejada, P.
1993-01-01
When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)
Energy Technology Data Exchange (ETDEWEB)
Chandra, V.K. [Department of Electrical and Electronics Engineering, Chhatrapati Shivaji Institute of Technology, Shivaji Nagar, Kolihapuri, Durg 491001 (C.G.) (India); Chandra, B.P., E-mail: bpchandra4@yahoo.co.in [Department of Applied Physics, Ashoka Institute of Technology and Management, Rajnandgaon 491441 (C.G.) (India); Tiwari, M. [Department of Postgraduate Studies and Research in Physics and Electronics, Rani Durgavati University, Jabalpur 482001 (M.P.) (India); Baghel, R.N. [School of Studies in Physics and Astrophysics, Pt. Ravishankar Shukla University, Raipur 492010 (C.G.) (India); Ramrakhiani, M. [Department of Postgraduate Studies and Research in Physics and Electronics, Rani Durgavati University, Jabalpur 482001 (M.P.) (India)
2012-06-15
expressions are derived for the transient EL intensity I, time t{sub m} and intensity I{sub m} corresponding to the peak of EL overshoot, total EL intensity I{sub t} and decay of the intensity of EL overshoot. In fact, the decay prior to the onset of EL overshoot is the decay of number of electrons moving in the electron transporting states. The ratio I{sub m}/I{sub s} decreases with increasing value of the applied pulse voltage because I{sub m} increases linearly with the amplitude of applied voltage pulse and I{sub s} increases nonlinearly and rapidly with the increasing amplitude of applied voltage pulse. The lifetime {tau}{sub t} of electrons at the interface decreases with increasing temperature whereby the dependence of {tau}{sub t} on temperature follows Arrhenius plot. This fact indicates that the detrapping involves thermally-assisted tunneling of electrons. Using the EL overshoot in bilayer OLEDs, the lifetime of the charge carriers at the interface, recombination time of charge carriers, decay time of the EL prior to onset of overshoot, and the time delay between the voltage pulse and onset time of the EL overshoot can be determined. The intense EL overshoot of nanosecond or shorter time duration may be useful in digital communication, and moreover, the EL overshoot gives important information about the processes involving injection, transport and recombination of charge carriers. The criteria for appearance of EL overshoot in bilayer OLEDs are explored. A good agreement is found between the theoretical and experimental results. - Highlights: Black-Right-Pointing-Pointer Modeling is done for transient electroluminescence overshoot in bilayer OLEDs. Black-Right-Pointing-Pointer Overshoot in spin-coated bilayer OLEDs is caused by the presence of an interfacial layer. Black-Right-Pointing-Pointer Expressions are derived for the transient EL intensity and decay of EL in OLEDs. Black-Right-Pointing-Pointer A good agreement is found between the theoretical and
Energy Technology Data Exchange (ETDEWEB)
Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-08-01
This report explains the numerical methods and the set-up method of input data for a single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3 (Direct Numerical Simulation using a 3rd-order upwind scheme). The code was developed to simulate non-stationary temperature fluctuation phenomena related to thermal striping phenomena, developed at Power Reactor and Nuclear Fuel Development Corporation (PNC). The DINUS-3 code was characterized by the use of a third-order upwind scheme for convection terms in instantaneous Navier-Stokes and energy equations, and an adaptive control system based on the Fuzzy theory to control time step sizes. Author expect this report is very useful to utilize the DINUS-3 code for the evaluation of various non-stationary thermohydraulic phenomena in reactor applications. (author)
Influence of the outlet air temperature on the thermohydraulic behaviour of air coolers
Directory of Open Access Journals (Sweden)
Đorđević Emila M.
2003-01-01
Full Text Available The determination of the optimal process conditions for the operation of air coolers demands a detailed analysis of their thermohydraulic behaviour on the one hand, and the estimation of the operating costs, on the other. One of the main parameters of the thermohydraulic behaviour of this type of equipment, is the outlet air temperature. The influence of the outlet air temperature on the performance of air coolers (heat transfer coefficient overall heat transfer coefficient, required surface area for heat transfer air-side pressure drop, fan power consumption and sound pressure level was investigated in this study. All the computations, using AirCooler software [1], were applied to cooling of the process fluid and the condensation of a multicomponent vapour mixture on two industrial devices of known geometries.
Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis
International Nuclear Information System (INIS)
Arien, B.; Daniels, J.
1986-12-01
CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)
International Nuclear Information System (INIS)
Gerbeth, Gunter; Eckert, Sven; Stefani, Frank; Gundrum, Thomas
2013-01-01
DRESDYN: General features. DRESDYN: DREsden Sodium facility for DYNamo and thermohydraulic studies. A large-scale new infrastructure for liquid metal experiments. Features: • New building ~ 500 m 2 ; • Total sodium inventory: 12-15 tons; • Precession driven experiment with separate strong basement and containment for Argon flooding; • Big hall for SFR related experiments, including ISI, a sodium loop, X-ray lab; • Financing is given, construction will start soon in spring 2013; • First experiments 2015 (hopefully...)
Preliminary study of the thermo-hydraulic behaviour of the binary breeder reactor
International Nuclear Information System (INIS)
Silveira Luz, M. da; Ferreira, W.J.
1984-06-01
Continuing the development of the Binary Breeder Reactor, its physical configuration and the advantages of differents types of spacers are analysed. In order to simulate the thermo-hydraulic behaviour and obtain data for a preliminary evaluation of the core geometry, the COBRA III C code was used to study the effects of the lenght and diameter of the fuel element, the coolant inlet temperature, the system pressure, helicoidal pitch and the pitch to diameter ratio. (Author) [pt
Energy Technology Data Exchange (ETDEWEB)
Pradhan, Santosh K., E-mail: santosh@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai 400094 (India); Obaidurrahman, K. [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai 400094 (India); Iyer, Kannan N. [Department of Mechanical Engineering, IIT Bombay, Mumbai 400076 (India); Gaikwad, Avinash J. [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai 400094 (India)
2016-04-15
Highlights: • A multi-point kinetics model is developed for RELAP5 system thermal hydraulics code. • Model is validated against extensive 3D kinetics code. • RELAP5 multi-point kinetics formulation is used to investigate critical break for LOCA in PHWR. - Abstract: Point kinetics approach in system code RELAP5 limits its use for many of the reactivity induced transients, which involve asymmetric core behaviour. Development of fully coupled 3D core kinetics code with system thermal-hydraulics is the ultimate requirement in this regard; however coupling and validation of 3D kinetics module with system code is cumbersome and it also requires access to source code. An intermediate approach with multi-point kinetics is appropriate and relatively easy to implement for analysis of several asymmetric transients for large cores. Multi-point kinetics formulation is based on dividing the entire core into several regions and solving ODEs describing kinetics in each region. These regions are interconnected by spatial coupling coefficients which are estimated from diffusion theory approximation. This model offers an advantage that associated ordinary differential equations (ODEs) governing multi-point kinetics formulation can be solved using numerical methods to the desired level of accuracy and thus allows formulation based on user defined control variables, i.e., without disturbing the source code and hence also avoiding associated coupling issues. Euler's method has been used in the present formulation to solve several coupled ODEs internally at each time step. The results have been verified against inbuilt point-kinetics models of RELAP5 and validated against 3D kinetics code TRIKIN. The model was used to identify the critical break in RIH of a typical large PHWR core. The neutronic asymmetry produced in the core due to the system induced transient was effectively handled by the multi-point kinetics model overcoming the limitation of in-built point kinetics model
Directory of Open Access Journals (Sweden)
Der-Sheng Chan
2010-04-01
Full Text Available Most of the voltage losses of proton exchange membrane fuel cells (PEMFC are due to the sluggish kinetics of oxygen reduction on the cathode and the low oxygen diffusion rate inside the flooded cathode. To simulate the transient flooding in the cathode of a PEMFC, a transient model was developed. This model includes the material conservation of oxygen, vapor, water inside the gas diffusion layer (GDL and micro-porous layer (MPL, and the electrode kinetics in the cathode catalyst layer (CL. The variation of hydrophobicity of each layer generated a wicking effect that moves water from one layer to the other. Since the GDL, MPL, and CL are made of composite materials with different hydrophilic and hydrophobic properties, a linear function of saturation was used to calculate the wetting contact angle of these composite materials. The balance among capillary force, gas/liquid pressure, and velocity of water in each layer was considered. Therefore, the dynamic behavior of PEMFC, with saturation transportation taken into account, was obtained in this study. A step change of the cell voltage was used to illustrate the transient phenomena of output current, water movement, and diffusion of oxygen and water vapor across the entire cathode.
International Nuclear Information System (INIS)
Martin, M.; Shen, T.; Thadhani, N.N.
2008-01-01
Instrumented anvil-on-rod impact experiments were performed to access the applicability of this approach for validating a constitutive strength model for dynamic, transient-state deformation and elastic-plastic wave interactions in vanadium, 21-6-9 stainless steel, titanium, and Ti-6Al-4V. In addition to soft-catching the impacted rod-shaped samples, their transient deformation states were captured by high-speed imaging, and velocity interferometry was used to record the sample back (free) surface velocity and monitor elastic-plastic wave interactions. Simulations utilizing AUTODYN-2D hydrocode with Steinberg-Guinan constitutive equation were used to generate simulated free surface velocity traces and final/transient deformation profiles for comparisons with experiments. The simulations were observed to under-predict the radial strain for bcc vanadium and fcc steel, but over-predict the radial strain for hcp titanium and Ti-6Al-4V. The correlations illustrate the applicability of the instrumented anvil-on-rod impact test as a method for providing robust model validation based on the entire deformation event, and not just the final deformed state
Directory of Open Access Journals (Sweden)
Viknash Shagar
2018-03-01
Full Text Available This paper aims to understand how the common phenomenon of fluctuations in propulsion and service load demand contribute to frequency transients in hybrid electric ship power systems. These fluctuations arise mainly due to changes in sea conditions resulting in significant variations in the propulsion load demand of ships. This leads to poor power quality for the power system that can potentially cause hazardous conditions such as blackout on board the ship. Effects of these fluctuations are analysed using a hybrid electric ship power system model and a proposed Model Predictive Control (MPC strategy to prevent propagation of transients from the propellers into the shipboard power system. A battery energy storage system, which is directly connected to the DC-link of the frequency converter, is used as the smoothing element. Case studies that involve propulsion and service load changes have been carried out to investigate the efficacy of the proposed solution. Simulation results show that the proposed solution with energy storage and MPC is able to contain frequency transients in the shipboard power system within the permissible levels stipulated by the relevant power quality standards. These findings will help ship builders and operators to consider using battery energy storage systems controlled by advanced control techniques such as MPC to improve the power quality on board ships.
Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters
International Nuclear Information System (INIS)
Wiles, L.E.; McCann, R.A.
1981-09-01
This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables
International Nuclear Information System (INIS)
Lin, J.; Bartal, Y.; Uhrig, R.E.
1995-01-01
The importance of automatic diagnostic systems for nuclear power plants (NPPs) has been discussed in numerous studies, and various such systems have been proposed. None of those systems were designed to predict the severity of the diagnosed scenario. A classification and severity prediction system for NPP transients is developed. The system is based on nearest neighbors modeling, which is optimized using genetic algorithms. The optimization process is used to determine the most important variables for each of the transient types analyzed. An enhanced version of the genetic algorithms is used in which a local downhill search is performed to further increase the accuracy achieved. The genetic algorithms search was implemented on a massively parallel supercomputer, the KSR1-64, to perform the analysis in a reasonable time. The data for this study were supplied by the high-fidelity simulator of the San Onofre unit 1 pressurized water reactor
Shu, Hongying; Wang, Lin; Watmough, James
2014-01-01
Sustained and transient oscillations are frequently observed in clinical data for immune responses in viral infections such as human immunodeficiency virus, hepatitis B virus, and hepatitis C virus. To account for these oscillations, we incorporate the time lag needed for the expansion of immune cells into an immunosuppressive infection model. It is shown that the delayed antiviral immune response can induce sustained periodic oscillations, transient oscillations and even sustained aperiodic oscillations (chaos). Both local and global Hopf bifurcation theorems are applied to show the existence of periodic solutions, which are illustrated by bifurcation diagrams and numerical simulations. Two types of bistability are shown to be possible: (i) a stable equilibrium can coexist with another stable equilibrium, and (ii) a stable equilibrium can coexist with a stable periodic solution.
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.
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.)
Energy Technology Data Exchange (ETDEWEB)
Pouget, V.; Lapuyade, H.; Lewis, D.; Deval, Y.; Fouillat, P. [Bordeaux-1 Univ., IXL, 33 - Talence (France); Sarger, L. [Bordeaux-1 Univ., CPMOH, 33 - Talence (France)
1999-07-01
A new SPICE model of irradiated MOSFET taking into account the real response of the 4 electrodes is proposed. The component that has been simulated is an NMOS transistor issued from the AMS BiCMOS 0.8 {mu}m technology. A comparison between SPICE-generated transients and PISCES device simulation demonstrates the accuracy benefits when used in complex electronic architectures. This model could be used when designing electronic circuits able to sustain hardening due to SEE (single event effect), it will be an efficient complement to the physical simulations.
Sprague, Mark W; Luczkovich, Joseph J
2016-01-01
This finite-difference time domain (FDTD) model for sound propagation in very shallow water uses pressure and velocity grids with both 3-dimensional Cartesian and 2-dimensional cylindrical implementations. Parameters, including water and sediment properties, can vary in each dimension. Steady-state and transient signals from discrete and distributed sources, such as the surface of a vibrating pile, can be used. The cylindrical implementation uses less computation but requires axial symmetry. The Cartesian implementation allows asymmetry. FDTD calculations compare well with those of a split-step parabolic equation. Applications include modeling the propagation of individual fish sounds, fish aggregation sounds, and distributed sources.
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...
Multi-scale modeling of the mechanical behavior of polycrystalline ice under transient creep.
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 ...