GRASS-SST, Fission Products Gas Release and Fuel Swelling in Steady-State and Transients
International Nuclear Information System (INIS)
Zawadzki, S.
2001-01-01
1 - Description of program or function: GRASS-SST is a comprehensive, mechanistic model for the prediction of fission-gas behaviour in UO 2 -base fuels during steady-state and transient conditions. GRASS-SST treats fission-gas release and fuel swelling on an equal basis and simultaneously treats all major mechanisms that influence fission-gas behaviour. Models are included for intra- and inter-granular fission-gas bubble behaviour as well as a mechanistic description of the role of grain-edge inter-linked porosity on fission-gas release and swelling. GRASS-SST calculations include the effects of gas production from fissioning uranium atoms, bubble nucleation, a realistic equation of state for xenon, lattice bubble diffusivities based on experimental observations, bubble migration, bubble coalescence, re-solution, temperature and temperature gradients, inter-linked porosity, and fission-gas interaction with structural defects (dislocations and grain boundaries) on both the distribution of fission-gas within the fuel and on the amount of fission-gas released from the fuel. GRASS-SST includes the effects of the degree of nonequilibrium in the UO 2 lattice on fission-gas bubble mobility and bubble coalescence and also accounts for the observed formation of grain-surface channels. GRASS-SST also includes mechanistic models for grain-growth/grain boundary sweeping and for the behaviour of fission gas during liquefaction/dissolution and fuel melting conditions. 2 - Method of solution: A system of coupled equations for the evolution of the fission-gas bubble-size distributions in the lattice, on dislocations, on grain faces, and grain edges is derived based on the GRASS-SST models. Given a set of operating conditions, GRASS-SST calculates the bubble radii for the size classes of bubbles under consideration using a realistic equation of state for xenon as well as a generalised capillary relation. 3 - Restrictions on the complexity of the problem: Maxima of : 1 axial section
PWR systems transient analysis
International Nuclear Information System (INIS)
Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.
1985-01-01
Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents
International Nuclear Information System (INIS)
Saha, P.
1984-01-01
This chapter reviews the papers on the pressurized water reactor (PWR) and boiling water reactor (BWR) transient analyses given at the American Nuclear Society Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors. Most of the papers were based on the systems calculations performed using the TRAC-PWR, RELAP5 and RETRAN codes. The status of the nuclear industry in the code applications area is discussed. It is concluded that even though comprehensive computer codes are available for plant transient analysis, there is still a need to exercise engineering judgment, simpler tools and even hand calculations to supplement these codes
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
Transient analysis of multicavity klystrons
International Nuclear Information System (INIS)
Lavine, T.L.; Miller, R.H.; Morton, P.L.; Ruth, R.D.
1988-09-01
We describe a model for analytic analysis of transients in multicavity klystron output power and phase. Cavities are modeled as resonant circuits, while bunching of the beam is modeled using linear space-charge wave theory. Our analysis has been implemented in a computer program which we use in designing multicavity klystrons with stable output power and phase. We present as examples transient analysis of a relativistic klystron using a magnetic pulse compression modulator, and of a conventional klystron designed to use phase shifting techniques for RF pulse compression. 4 refs., 4 figs
Transient analysis of DTT rakes
International Nuclear Information System (INIS)
Kamath, P.S.; Lahey, R.T. Jr.
1981-01-01
This paper presents an analytical model for the determination of the cross-sectionally averaged transient mass flux of a two-phase fluid flowing in a conduit instrumented by a Drag-Disk Turbine Transducer (DTT) Rake and a multibeam gamma densitometer. Parametric studies indicate that for a typical blowdown transient, dynamic effects such as rotor inertia can be important for the turbine-meter. In contrast, for the drag-disk, a frequency response analysis showed that the quasisteady solution is valid below a forcing frequency of about 10 Hz, which is faster than the time scale normally encountered during blowdowns. The model showed reasonably good agreement with full scale transient rake data, where the flow regimes were mostly homogeneous or stratified, thus indicating that the model is suitable for the analysis of a DTT rake. (orig.)
Reactor operational transient analysis
International Nuclear Information System (INIS)
Shin, W.K.; Chae, S.K.; Han, K.I.; Yang, K.S.; Chung, H. D.; Kim, H.G.; Moon, H.J.; Ryu, Y.H.
1983-01-01
To build up efficient capability of safety review and inspection for the nuclear power plants, four area of studies have performed as follows: 1) In order to search the most optimized operating method during load follow operating schemes, automatic control and normal control, are compared each other under the CAOC condition. The analysis performed by DDID code has shown that the reactor has to be controlled by the operator manually during load follow operation. 2) Through the sensitivity analysis by COBRA code, the operating parameters, such as coolant pressure, flow rate, inlet temperature, and power distribution are shown to be important to the determination of DNBR. Expecially, inlet temperature of primary coolant system is appeared as the most senstive parameter on DNBR. 3) FRAPCON code is adapted to study the sensitivity of several operational parameters on the mechanical properties of reactor fuel rod. 4) The calculations procedure which is required to be obtained the neutron fluence at the reactor vessel and the spectrum at the surveillance capsule is established. The results of computation are conpared with those of FSAR and SWRI report and proved its applicability to reactor surveillance program. (Author)
Transient analysis for resolving safety issues
International Nuclear Information System (INIS)
Chao, J.; Layman, W.
1987-01-01
The Nuclear Safety Analysis Center (NSAC) has a Generic Safety Analysis Program to help resolve high priority generic safety issues. This paper describes several high priority safety issues considered at NSAC and how they were resolved by transient analysis using thermal hydraulics and neutronics codes. These issues are pressurized thermal shock (PTS), anticipated transients without scram (ATWS), steam generator tube rupture (SGTR), and reactivity transients in light of the Chernobyl accident
Transient analysis capabilities at ABB-CE
International Nuclear Information System (INIS)
Kling, C.L.
1992-01-01
The transient capabilities at ABB-Combustion Engineering (ABB-CE) Nuclear Power are a function of the computer hardware and related network used, the computer software that has evolved over the years, and the commercial technical exchange agreements with other related organizations and customers. ABB-CEA is changing from a mainframe/personal computer network to a distributed workstation/personal computer local area network. The paper discusses computer hardware, mainframe computing, personal computers, mainframe/personal computer networks, workstations, transient analysis computer software, design/operation transient analysis codes, safety (licensed) analysis codes, cooperation with ABB-Atom, and customer support
Transient thermal analysis of Vega launcher structures
Energy Technology Data Exchange (ETDEWEB)
Gori, F. [University of Rome ' Tor Vergata' , Rome (Italy); De Stefanis, M. [Thales Alenia Space Italia, Rome (Italy); Worek, W.M. [University of Illinois at Chicago, Chicago (United States)], E-mail: wworek@uic.edu; Minkowycz, W.J. [University of Illinois at Chicago, Chicago (United States)
2008-12-15
A transient thermal analysis is carried out to verify the base cover thermal protection system of Vega 2nd stage Solid Rocket Motor (SRM) and the flange coupling of the inter-stage 2/3. The analysis is performed with a finite element code. The work has developed suitable numerical Fortran subroutines to assign radiation and convection boundary conditions. The thermal behaviour of the structures is presented.
Transient analysis for Laguna Verde nuclear power plant
International Nuclear Information System (INIS)
Ramos Pablos, J.C. et.al.
1991-01-01
Relationship between transients analysis and safety of Laguna Verde nuclear power plant is described a general panorama of safety thermal limits of a nuclear station, as well as transients classification and events simulation codes are exposed. Activities of a group of transients analysis of electrical research institute are also mentioned (Author)
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
Transient analysis of intermittent multijet sprays
Energy Technology Data Exchange (ETDEWEB)
Panao, Miguel R.O.; Moreira, Antonio Luis N. [Universidade Tecnica de Lisboa, IN, Center for Innovation, Technology and Policy Research, Instituto Superior Tecnico, Lisboa (Portugal); Durao, Diamantino G. [Universidade Lusiada, Lisboa (Portugal)
2012-07-15
This paper analyzes the transient characteristics of intermittent sprays produced by the single-point impact of multiple cylindrical jets. The aim is to perform a transient analysis of the intermittent atomization process to study the effect of varying the number of impinging jets in the hydrodynamic mechanisms of droplet formation. The results evidence that hydrodynamic mechanisms underlying the physics of ligament fragmentation in 2-impinging jets sprays also apply to sprays produced with more than 2 jets during the main period of injection. Ligaments detaching from the liquid sheet, as well as from its bounding rim, have been identified and associated with distinct droplet clusters, which become more evident as the number of impinging jets increases. Droplets produced by detached ligaments constitute the main spray, and their axial velocity becomes more uniformly distributed with 4-impinging jets because of a delayed ligament fragmentation. Multijet spray dispersion patterns are geometric depending on the number of impinging jets. Finally, an analysis on the Weber number of droplets suggests that multijet sprays are more likely to deposit on interposed surfaces, thus becoming a promising and competitive atomization solution for improving spray cooling. (orig.)
Thermal transient analysis applied to horizontal wells
Energy Technology Data Exchange (ETDEWEB)
Duong, A.N. [Society of Petroleum Engineers, Canadian Section, Calgary, AB (Canada)]|[ConocoPhillips Canada Resources Corp., Calgary, AB (Canada)
2008-10-15
Steam assisted gravity drainage (SAGD) is a thermal recovery process used to recover bitumen and heavy oil. This paper presented a newly developed model to estimate cooling time and formation thermal diffusivity by using a thermal transient analysis along the horizontal wellbore under a steam heating process. This radial conduction heating model provides information on the heat influx distribution along a horizontal wellbore or elongated steam chamber, and is therefore important for determining the effectiveness of the heating process in the start-up phase in SAGD. Net heat flux estimation in the target formation during start-up can be difficult to measure because of uncertainties regarding heat loss in the vertical section; steam quality along the horizontal segment; distribution of steam along the wellbore; operational conditions; and additional effects of convection heating. The newly presented model can be considered analogous to pressure transient analysis of a buildup after a constant pressure drawdown. The model is based on an assumption of an infinite-acting system. This paper also proposed a new concept of a heating ring to measure the heat storage in the heated bitumen at the time of testing. Field observations were used to demonstrate how the model can be used to save heat energy, conserve steam and enhance bitumen recovery. 18 refs., 14 figs., 2 appendices.
Transient Analysis of a Magnetic Heat Pump
Schroeder, E. A.
1985-01-01
An experimental heat pump that uses a rare earth element as the refrigerant is modeled using NASTRAN. The refrigerant is a ferromagnetic metal whose temperature rises when a magnetic field is applied and falls when the magnetic field is removed. The heat pump is used as a refrigerator to remove heat from a reservoir and discharge it through a heat exchanger. In the NASTRAN model the components modeled are represented by one-dimensional ROD elements. Heat flow in the solids and fluid are analyzed. The problem is mildly nonlinear since the heat capacity of the refrigerant is temperature-dependent. One simulation run consists of a series of transient analyses, each representing one stroke of the heat pump. An auxiliary program was written that uses the results of one NASTRAN analysis to generate data for the next NASTRAN analysis.
Analysis of transient signals by Wavelet transform
International Nuclear Information System (INIS)
Penha, Rosani Libardi da; Silva, Aucyone A. da; Ting, Daniel K.S.; Oliveira Neto, Jose Messias de
2000-01-01
The objective of this work is to apply the Wavelet Transform in transient signals. The Wavelet technique can outline the short time events that are not easily detected using traditional techniques. In this work, the Wavelet Transform is compared with Fourier Transform, by using simulated data and rotor rig data. This data contain known transients. The wavelet could follow all the transients, what do not happen to the Fourier techniques. (author)
Transient analysis on the SMART-P anticipated transients without scram
International Nuclear Information System (INIS)
Yang, S. H.; Bae, K. H.; Kim, H. C.; Zee, S. Q.
2005-01-01
Anticipated transients without scram (ATWS) are anticipated operational occurrences accompanied by a failure of an automatic reactor trip when required. Although the occurrence probability of the ATWS events is considerably low, these events can result in unacceptable consequences, i.e. the pressurization of the reactor coolant system (RCS) up to an unacceptable range and a core-melting situation. Therefore, the regulatory body requests the installation of a protection system against the ATWS events. According to the request, a diverse protection system (DPS) is installed in the SMART-P (System-integrated Modular Advanced ReacTor-Pilot). This paper presents the results of the transient analysis performed to identify the performance of the SMART-P against the ATWS. In the analysis, the TASS/SMR (Transients And Setpoint Simulation/Small and Medium Reactor) code is applied to identify the thermal hydraulic response of the RCS during the transients
Fuel cladding mechanical properties for transient analysis
International Nuclear Information System (INIS)
Johnson, G.D.; Hunter, C.W.; Hanson, J.E.
1976-01-01
Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s (5.6 0 K/s) and 200 0 F/s (111 0 K/s) to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence
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)
Thermal-hydraulic analysis of PWR cores in transient condition
International Nuclear Information System (INIS)
Silva Galetti, M.R. da.
1984-01-01
A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt
Quantum-corrected transient analysis of plasmonic nanostructures
Uysal, Ismail Enes; Ulku, Huseyin Arda; Sajjad, Muhammad; Singh, Nirpendra; Schwingenschlö gl, Udo; Bagci, Hakan
2017-01-01
A time domain surface integral equation (TD-SIE) solver is developed for quantum-corrected analysis of transient electromagnetic field interactions on plasmonic nanostructures with sub-nanometer gaps. “Quantum correction” introduces an auxiliary
Intermediate size inducer pump - structural analysis and transient deformation studies
International Nuclear Information System (INIS)
Cheng, T.K.; Nishizaka, J.N.
1979-05-01
This report summarizes the structural and thermal transient deformation analysis of the Intermediate Size Inducer Pump. The analyses were performed in accordance to the requirements of N266ST310001, the specification for the ISIP. Results of stress analysis indicate that the thermal transient stress and strain are within the stress strain limits of RDT standard F9-4 which was used as a guide
Transient analysis for PWR reactor core using neural networks predictors
International Nuclear Information System (INIS)
Gueray, B.S.
2001-01-01
In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions
Transient analysis of the IRIS reactor
International Nuclear Information System (INIS)
Bajs, T.; Oriani, L.; Ricotti, M.E.; Barroso, A.C.
2002-01-01
An international consortium of industry, laboratory, university and utility establishments, led by Westinghouse, is developing a modular, integral, light water cooled, small to medium power reactor, the International Reactor Innovative and Secure (IRIS). IRIS features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). Given the large number of organizations involved in the IRIS design, the RELAP5/MOD 3.3 code has been selected as the main system code. A nodalization of the reference IRIS design has been developed with a basic set of protective functions and controls. Engineered Safety Features of the concept are being also implemented, and in particular the Emergency Heat Removal System that is used for safety grade decay heat removal and in the small break LOCA response of IRIS (Large break LOCAs are eliminated in IRIS by the adoption of the Integral layout) This paper discusses developed model and transient behavior of the system for representative transient sequences.(author)
Lightning transient analysis in wind turbine blades
DEFF Research Database (Denmark)
Candela Garolera, Anna; Holbøll, Joachim; Madsen, Søren Find
2013-01-01
The transient behavior of lightning surges in the lightning protection system of wind turbine blades has been investigated in this paper. The study is based on PSCAD models consisting of electric equivalent circuits with lumped and distributed parameters involving different lightning current...... waveforms. The aim of the PSCAD simulations is to study the voltages induced by the lightning current in the blade that may cause internal arcing. With this purpose, the phenomenon of current reflections in the lightning down conductor of the blade and the electromagnetic coupling between the down conductor...... and other internal conductive elements of the blade is studied. Finally, several methods to prevent internal arcing are discussed in order to improve the lightning protection of the blade....
Alternatives Analysis for the Resumption of Transient Testing Program
Energy Technology Data Exchange (ETDEWEB)
Lee Nelson
2013-11-01
An alternatives analysis was performed for resumption of transient testing. The analysis considered eleven alternatives – including both US international facilities. A screening process was used to identify two viable alternatives from the original eleven. In addition, the alternatives analysis includes a no action alternative as required by the National Environmental Policy Act (NEPA). The alternatives considered in this analysis included: 1. Restart the Transient Reactor Test Facility (TREAT) 2. Modify the Annular Core Research Reactor (ACRR) which includes construction of a new hot cell and installation of a new hodoscope. 3. No Action
Transient analysis of house load operation for LNPP
International Nuclear Information System (INIS)
Shi Junying; Zheng Bin
2000-01-01
The author analysis the transient of house load operation for Ling'ao Nuclear Power Plant by using the methods of dynamic simulation and closed loops of primary and secondary system. The transient of house load operation from 100% FP is the most severe that can occur on the unit in normal operation because it causes immediately shedding of 95% of turbine load and requires the unit to operate steadily at reduced power. The results show that the transient can be successful both at beginning of core life and manual house load operation. However, more attentions must be paid to automatic house load operation caused by grid fault at toward end of core life because the success of the transient could be threatened by the actuation of the protection of high flux and high flux rate
Transient analysis of multifailure conditions by using PWR plant simulator
International Nuclear Information System (INIS)
Morisaki, Hidetoshi; Yokobayashi, Masao.
1984-11-01
This report describes results of the analysis of abnormal transients caused by multifailures using a PWR plant simulator. The simulator is based on an existing 822MWe power plant with 3 loops, and designed to cover wide range of plant operation from cold shutdown to full power at the end of life. Various malfunctions to simulate abnormal conditions caused by equipment failures are provided. In this report, features of abnormal transients caused by concurrence of malfunctions are discussed. The abnormal conditions studied are leak of primary coolant, loss of charging and feedwater flows, and control systems failure. From the results, it was observed that transient responses caused by some of the malfunctions are almost same as the addition of behaviors caused by each single malfunction. Therefore, it can be said that kinds of malfunctions which are concurrent may be estimated from transient characteristics of each single malfunction. (author)
APR1400 Locked Rotor Transient Analysis using KNAP
International Nuclear Information System (INIS)
Lee, Dong-Hyuk; Kim, Yo-Han; Ha, Sang Jun
2007-01-01
KEPRI (Korea Electric Power Research Institute) has developed safety analysis methodology for non-LOCA (Loss Of Coolant Accident) analysis of OPR1000 (Optimized Power Reactor 1000, formerly KSNP). The new methodology, named KNAP (Korea Non-LOCA Analysis Package), uses RETRAN as the main system analysis code for most transients. For locked rotor transient DNBR analysis, UNICORN-TM code is used. UNICORN-TM is the unified code of RETRAN, MASTER and TORC. The UNICORN-TM has 1-D and 3-D neutron kinetics calculation capability. For locked rotor DNBR analysis, 1-D neutron kinetics is used. In this paper, we apply KNAP methodology to APR1400 (Advanced Power Reactor 1400) locked rotor analysis and compare the results with those in the APR1400 SSAR(Standard Safety Analysis Report). The locked rotor transient is one of the 'decrease in reactor coolant system flow rate' events and the results are typically described in the chapter 15.3.3 of SAR (Safety Analysis Report). In this study, to confirm the applicability of the KNAP methodology and code system to APR1400, locked rotor transient is analyzed using UNICORN-TM code and the results are compared with those from APR1400 SSAR
APR1400 Locked Rotor Transient Analysis using KNAP
Energy Technology Data Exchange (ETDEWEB)
Lee, Dong-Hyuk; Kim, Yo-Han; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)
2007-07-01
KEPRI (Korea Electric Power Research Institute) has developed safety analysis methodology for non-LOCA (Loss Of Coolant Accident) analysis of OPR1000 (Optimized Power Reactor 1000, formerly KSNP). The new methodology, named KNAP (Korea Non-LOCA Analysis Package), uses RETRAN as the main system analysis code for most transients. For locked rotor transient DNBR analysis, UNICORN-TM code is used. UNICORN-TM is the unified code of RETRAN, MASTER and TORC. The UNICORN-TM has 1-D and 3-D neutron kinetics calculation capability. For locked rotor DNBR analysis, 1-D neutron kinetics is used. In this paper, we apply KNAP methodology to APR1400 (Advanced Power Reactor 1400) locked rotor analysis and compare the results with those in the APR1400 SSAR(Standard Safety Analysis Report). The locked rotor transient is one of the 'decrease in reactor coolant system flow rate' events and the results are typically described in the chapter 15.3.3 of SAR (Safety Analysis Report). In this study, to confirm the applicability of the KNAP methodology and code system to APR1400, locked rotor transient is analyzed using UNICORN-TM code and the results are compared with those from APR1400 SSAR.
A fast reactor transient analysis methodology for personal computers
International Nuclear Information System (INIS)
Ott, K.O.
1993-01-01
A simplified model for a liquid-metal-cooled reactor (LMR) transient analysis, in which point kinetics as well as lumped descriptions of the heat transfer equations in all components are applied, is converted from a differential into an integral formulation. All 30 differential balance equations are implicitly solved in terms of convolution integrals. The prompt jump approximation is applied as the strong negative feedback effectively keeps the net reactivity well below prompt critical. After implicit finite differencing of the convolution integrals, the kinetics equation assumes a new form, i.e., the quadratic dynamics equation. In this integral formulation, the initial value problem of typical LMR transients can be solved with large item steps (initially 1 s, later up to 256 s). This then makes transient problems amenable to a treatment on personal computer. The resulting mathematical model forms the basis for the GW-BASIC program LMR transient calculation (LTC) program. The LTC program has also been converted to QuickBASIC. The running time for a 10-h transient overpower transient is then ∼40 to 10 s, depending on the hardware version (286, 386, or 486 with math coprocessors)
Transient electromagnetic analysis in tokamaks using TYPHOON code
International Nuclear Information System (INIS)
Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.
1996-01-01
The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)
Code Coupling for Multi-Dimensional Core Transient Analysis
International Nuclear Information System (INIS)
Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il
2015-01-01
After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident
Code Coupling for Multi-Dimensional Core Transient Analysis
Energy Technology Data Exchange (ETDEWEB)
Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)
2015-05-15
After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.
Analysis of piping response to thermal and operational transients
International Nuclear Information System (INIS)
Wang, C.Y.
1987-01-01
The reactor piping system is an extremely complex three-dimensional structure. Maintaining its structural integrity is essential to the safe operation of the reactor and the steam-supply system. In the safety analysis, various transient loads can be imposed on the piping which may cause plastic deformation and possible damage to the system, including those generated from hydrodynamic wave propagations, thermal and operational transients, as well as the seismic events. At Argonne National Laboratory (ANL), a three-dimensional (3-D) piping code, SHAPS, aimed for short-duration transients due to wave propagation, has been developed. Since 1984, the development work has been shifted to the long-duration accidents originating from the thermal and operational transient. As a result, a new version of the code, SHAPS-2, is being established. This paper describes many features related to this later development. To analyze piping response generated from thermal and operational transients, a 3-D implicit finite element algorithm has been developed for calculating the hoop, flexural, axial, and torsional deformations induced by the thermomechanical loads. The analysis appropriately accounts for stresses arising from the temperature dependence of the elastic material properties, the thermal expansion of the materials, and the changes in the temperature-dependent yield surface. Thermal softening, failure, strain rate, creep, and stress ratching can also be considered
Taipower's transient analysis methodology for pressurized water reactors
International Nuclear Information System (INIS)
Huang, Pinghue
1998-01-01
The methodology presented in this paper is a part of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors' developed by the Taiwan Power Company (TPC) and the Institute of Nuclear Energy Research. This methodology utilizes four computer codes developed or sponsored by Electric Power Research institute: system transient analysis code RETRAN-02, core thermal-hydraulic analysis code COBRAIIIC, three-dimensional spatial kinetics code ARROTTA, and fuel rod evaluation code FREY. Each of the computer codes was extensively validated. Analysis methods and modeling techniques were conservatively established for each application using a systematic evaluation with the assistance of sensitivity studies. The qualification results and analysis methods were documented in detail in TPC topical reports. The topical reports for COBRAIIIC, ARROTTA. and FREY have been reviewed and approved by the Atomic Energy Council (ABC). TPC 's in-house transient methodology have been successfully applied to provide valuable support for many operational issues and plant improvements for TPC's Maanshan Units I and 2. Major applications include the removal of the resistance temperature detector bypass system, the relaxation of the hot-full-power moderator temperature coefficient design criteria imposed by the ROCAEC due to a concern on Anticipated Transient Without Scram, the reduction of boron injection tank concentration and the elimination of the heat tracing, and the reduction of' reactor coolant system flow. (author)
Lumped thermal capacitance analysis of transient heat conduction ...
African Journals Online (AJOL)
Lumped thermal capacitance analysis has been undertaken to investigate the transient temperature variations, associated induced thermal stress distributions, and the structural integrity of Ghana Research Reactor-1 (GHAR R-1) vessel after 15 years of operation. The beltline configuration of the cylindrical vessel of the ...
Transient stability analysis of a distribution network with distributed generators
Xyngi, I.; Ishchenko, A.; Popov, M.; Sluis, van der L.
2009-01-01
This letter describes the transient stability analysis of a 10-kV distribution network with wind generators, microturbines, and CHP plants. The network being modeled in Matlab/Simulink takes into account detailed dynamic models of the generators. Fault simulations at various locations are
Verification and validation of COBRA-SFS transient analysis capability
International Nuclear Information System (INIS)
Rector, D.R.; Michener, T.E.; Cuta, J.M.
1998-05-01
This report provides documentation of the verification and validation testing of the transient capability in the COBRA-SFS code, and is organized into three main sections. The primary documentation of the code was published in September 1995, with the release of COBRA-SFS, Cycle 2. The validation and verification supporting the release and licensing of COBRA-SFS was based solely on steady-state applications, even though the appropriate transient terms have been included in the conservation equations from the first cycle. Section 2.0, COBRA-SFS Code Description, presents a capsule description of the code, and a summary of the conservation equations solved to obtain the flow and temperature fields within a cask or assembly model. This section repeats in abbreviated form the code description presented in the primary documentation (Michener et al. 1995), and is meant to serve as a quick reference, rather than independent documentation of all code features and capabilities. Section 3.0, Transient Capability Verification, presents a set of comparisons between code calculations and analytical solutions for selected heat transfer and fluid flow problems. Section 4.0, Transient Capability Validation, presents comparisons between code calculations and experimental data obtained in spent fuel storage cask tests. Based on the comparisons presented in Sections 2.0 and 3.0, conclusions and recommendations for application of COBRA-SFS to transient analysis are presented in Section 5.0
Application of transient analysis methodology to heat exchanger performance monitoring
International Nuclear Information System (INIS)
Rampall, I.; Soler, A.I.; Singh, K.P.; Scott, B.H.
1994-01-01
A transient testing technique is developed to evaluate the thermal performance of industrial scale heat exchangers. A Galerkin-based numerical method with a choice of spectral basis elements to account for spatial temperature variations in heat exchangers is developed to solve the transient heat exchanger model equations. Testing a heat exchanger in the transient state may be the only viable alternative where conventional steady state testing procedures are impossible or infeasible. For example, this methodology is particularly suited to the determination of fouling levels in component cooling water system heat exchangers in nuclear power plants. The heat load on these so-called component coolers under steady state conditions is too small to permit meaningful testing. An adequate heat load develops immediately after a reactor shutdown when the exchanger inlet temperatures are highly time-dependent. The application of the analysis methodology is illustrated herein with reference to an in-situ transient testing carried out at a nuclear power plant. The method, however, is applicable to any transient testing application
Analysis of operator's behaviour under accidental transients
International Nuclear Information System (INIS)
Llory, M.; Lemaitre, D.; Griffon-Fouco, C.; Meslin, B.
1992-01-01
Since 1979, EDF has been conducting intensive test campaigns on full-scale PWR simulators in order to study and improve the operators behaviour under incident as well as accident conditions. This paper presents some results obtained during tests carried out in 1986 on the P4 (1300 MWe power plant series) simulators of the Paluel Training Center. These results essentially concern the observed deviations, the diagnosis and the safety engineer's role. They are compared with the results of previous tests on 900 MWe unit simulators. The test organization and methodology, the result analysis methods and the biases introduced by this kind of test are also discussed. (author). 7 refs, 1 fig., 6 figs
Transient and fuel performance analysis with VTT's coupled code system
International Nuclear Information System (INIS)
Daavittila, A.; Hamalainen, A.; Raty, H.
2005-01-01
VTT (technical research center of Finland) maintains and further develops a comprehensive safety analysis code system ranging from the basic neutronic libraries to 3-dimensional transient analysis and fuel behaviour analysis codes. The code system is based on various types of couplings between the relevant physical phenomena. The main tools for analyses of reactor transients are presently the 3-dimensional reactor dynamics code HEXTRAN for cores with a hexagonal fuel assembly geometry and TRAB-3D for cores with a quadratic fuel assembly geometry. HEXTRAN has been applied to safety analyses of VVER type reactors since early 1990's. TRAB-3D is the latest addition to the code system, and has been applied to BWR and PWR analyses in recent years. In this paper it is shown that TRAB-3D has calculated accurately the power distribution during the Olkiluoto-1 load rejection test. The results from the 3-dimensional analysis can be used as boundary conditions for more detailed fuel rod analysis. For this purpose a general flow model GENFLO, developed at VTT, has been coupled with USNRC's FRAPTRAN fuel accident behaviour model. The example case for FRAPTRAN-GENFLO is for an ATWS at a BWR plant. The basis for the analysis is an oscillation incident in the Olkiluoto-1 BWR during reactor startup on February 22, 1987. It is shown that the new coupled code FRAPTRAN/GENFLO is quite a promising tool that can handle flow situations and give a detailed analysis of reactor transients
Transient flow analysis of integrated valve opening process
Energy Technology Data Exchange (ETDEWEB)
Sun, Xinming; Qin, Benke; Bo, Hanliang, E-mail: bohl@tsinghua.edu.cn; Xu, Xingxing
2017-03-15
Highlights: • The control rod hydraulic driving system (CRHDS) is a new type of built-in control rod drive technology and the integrated valve (IV) is the key control component. • The transient flow experiment induced by IV is conducted and the test results are analyzed to get its working mechanism. • The theoretical model of IV opening process is established and applied to get the changing rule of the transient flow characteristic parameters. - Abstract: The control rod hydraulic driving system (CRHDS) is a new type of built-in control rod drive technology and the IV is the key control component. The working principle of integrated valve (IV) is analyzed and the IV hydraulic experiment is conducted. There is transient flow phenomenon in the valve opening process. The theoretical model of IV opening process is established by the loop system control equations and boundary conditions. The valve opening boundary condition equation is established based on the IV three dimensional flow field analysis results and the dynamic analysis of the valve core movement. The model calculation results are in good agreement with the experimental results. On this basis, the model is used to analyze the transient flow under high temperature condition. The peak pressure head is consistent with the one under room temperature and the pressure fluctuation period is longer than the one under room temperature. Furthermore, the changing rule of pressure transients with the fluid and loop structure parameters is analyzed. The peak pressure increases with the flow rate and the peak pressure decreases with the increase of the valve opening time. The pressure fluctuation period increases with the loop pipe length and the fluctuation amplitude remains largely unchanged under different equilibrium pressure conditions. The research results lay the base for the vibration reduction analysis of the CRHDS.
Advanced methods for BWR transient and stability analysis
Energy Technology Data Exchange (ETDEWEB)
Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)
2008-07-01
The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)
Advanced methods for BWR transient and stability analysis
International Nuclear Information System (INIS)
Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.
2008-01-01
The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)
Thermal analysis of LOFT modular DTT for LOCE transient
International Nuclear Information System (INIS)
Martin, C.M.
1978-01-01
A thermal analysis was performed on the LOFT modular drag-disc turbine transducer (MDTT) modular assembly. The purpose of this analysis was to determine the maximum temperature difference between the MDTT shroud and end cap during a LOCE. This temperature difference is needed for stress analysis of the MDTT endcap to fairing welds. The thermal analysis was done using TRIPLE, a three dimensional finite element code. A three dimensional model of the MDTT was made and transient temperature solutions were found for the different MDTT locations. The fluid temperature transients used for the solutions at all locations were from RELAP4 predictions of the LOFT L2-4 test which is considered the most severe temperature transient. Results of these calculations show the maximum temperature difference is 92 0 C (165 0 F) and occurs in the intact loop cold leg. This value and those found at other locations, are evaluated from the best available RELAP predicted temperatures during a nuclear LOCE
Analysis of short-term reactor cavity transient
International Nuclear Information System (INIS)
Cheng, T.C.; Fischer, S.R.
1981-01-01
Following the transient of a hypothetical loss-of-coolant accident (LOCA) in a nuclear reactor, peak pressures are reached within the first 0.03 s at different locations inside the reactor cavity. Due to the complicated multidimensional nature of the reactor cavity, the short-term analysis of the LOCA transient cannot be performed by using traditional containment codes, such as CONTEMPT. The advanced containment code, BEACON/MOD3, developed at the Idaho National Engineering Laboratory (INEL), can be adapted for such analysis. This code provides Eulerian, one and two-dimensional, nonhomogeneous, nonequilibrium flow modeling as well as lumped parameter, homogeneous, equilibrium flow modeling for the solution of two-component, two-phase flow problems. The purpose of this paper is to demonstrate the capability of the BEACON code to analyze complex containment geometry such as a reactor cavity
RELAP5/MOD2: for PWR transient analysis
International Nuclear Information System (INIS)
Ransom, V.H.
1983-01-01
RELAP5 is a light water reactor system transient simulation code for use in nuclear plant safety analysis. Development of a new version, RELAP5/MOD2, has been completed and will be released to the United States Nuclear Regulatory Commission during September of 1983. The new and improved modeling capability of RELAP5/MOD2 is described and some developmental assessment results are presented. The future plans for extension to severe accident modeling are briefly discussed
Transient Dynamics Analysis of The Reachstacker Speader Based On ANSYS
Directory of Open Access Journals (Sweden)
Shu Yu Feng
2016-01-01
Full Text Available Reachstacker is an indispensable handling machinery, it will inevitably lead to unbalanced force at the job site. This paper does transient dynamics analysis for the spreader mechanism, which is one of the most significance key components. We get dynamic response of the spreader in lifting instant, results not only provide a reference for designers to understand the mechanical characteristics of spreader comprehensively, but also bedding for the future research.
International Nuclear Information System (INIS)
Kinoshita, Hidetaka; Kaminaga, Masanori; Hino, Ryutaro
2000-02-01
In order to promote the Neutron Science Project of JAERI, the design of a 5MW-spallation target system is in progress with the purpose of producing a practical neutron application while at the same time adhering to the highest levels of safety. To establish the safety of the target system, it is important to understand the transient behaviors during anticipated operational events of the system, and to design the safety protection systems for the safe termination of the transients. This report presents the analytical results of transient behaviors in the mercury experimental loop using mercury properties. At first, the analytical pressure distributions were compared with experimental data measured with the mercury experimental loop. The modeling data were modified to reproduce the actual pressure distributions of the mercury experimental loop. Then a loss of forced convection and a loss of coolant accident were analyzed. In the case of the pump trip, the transient analysis was conducted using two types of mercury pumps, the mechanical type pump with moment of inertia, and the electrical-magnetic type pump without moment of inertia. The results show there was no clear difference in the two analyses, since the mercury had a large inertia, which was 13.5 times that of the water. Moreover, in the case of a pipe rupture at the pump exit, a moderate pressure decrease was confirmed when a small breakage area existed in which the coolant flowed out gradually. Based on these results, it was appeared that the transient fluctuation of pressure in the mercury loop would not become large and accidents would have to be detected by small fluctuations in pressure. Based on these analyses, we plan to conduct a simulation test to verify the RELAP5 code, and then the analysis of a full-scale mercury system will be performed. (author)
Enhanced Severe Transient Analysis for Prevention Technical Program Plan
Energy Technology Data Exchange (ETDEWEB)
Gougar, Hans [Idaho National Lab. (INL), Idaho Falls, ID (United States)
2014-09-01
This document outlines the development of a high fidelity, best estimate nuclear power plant severe transient simulation capability that will complement or enhance the integral system codes historically used for licensing and analysis of severe accidents. As with other tools in the Risk Informed Safety Margin Characterization (RISMC) Toolkit, the ultimate user of Enhanced Severe Transient Analysis and Prevention (ESTAP) capability is the plant decision-maker; the deliverable to that customer is a modern, simulation-based safety analysis capability, applicable to a much broader class of safety issues than is traditional Light Water Reactor (LWR) licensing analysis. Currently, the RISMC pathway’s major emphasis is placed on developing RELAP-7, a next-generation safety analysis code, and on showing how to use RELAP-7 to analyze margin from a modern point of view: that is, by characterizing margin in terms of the probabilistic spectra of the “loads” applied to systems, structures, and components (SSCs), and the “capacity” of those SSCs to resist those loads without failing. The first objective of the ESTAP task, and the focus of one task of this effort, is to augment RELAP-7 analyses with user-selected multi-dimensional, multi-phase models of specific plant components to simulate complex phenomena that may lead to, or exacerbate, severe transients and core damage. Such phenomena include: coolant crossflow between PWR assemblies during a severe reactivity transient, stratified single or two-phase coolant flow in primary coolant piping, inhomogeneous mixing of emergency coolant water or boric acid with hot primary coolant, and water hammer. These are well-documented phenomena associated with plant transients but that are generally not captured in system codes. They are, however, generally limited to specific components, structures, and operating conditions. The second ESTAP task is to similarly augment a severe (post-core damage) accident integral analyses code
Abnormal transient analysis by using PWR plant simulator, (2)
International Nuclear Information System (INIS)
Naitoh, Akira; Murakami, Yoshimitsu; Yokobayashi, Masao.
1983-06-01
This report describes results of abnormal transient analysis by using a PWR plant simulator. The simulator is based on an existing 822MWe power plant with 3 loops, and designed to cover wide range of plant operation from cold shutdown to full power at EOL. In the simulator, malfunctions are provided for abnormal conditions of equipment failures, and in this report, 17 malfunctions for secondary system and 4 malfunctions for nuclear instrumentation systems were simulated. The abnormal conditions are turbine and generator trip, failure of condenser, feedwater system and valve and detector failures of pressure and water level. Fathermore, failure of nuclear instrumentations are involved such as source range channel, intermediate range channel and audio counter. Transient behaviors caused by added malfunctions were reasonable and detail information of dynamic characteristics for turbine-condenser system were obtained. (author)
TRAWA, a transient analysis code for water reactions
International Nuclear Information System (INIS)
Rajamaeki, M.
1976-06-01
TRAWA is a transient analysis code for water reactors. It solves the two-group neutron diffusion equations simultaneously with the heat conduction equations and the two-phase hydraulic equations for one or more channels. At most one-dimensional submodels are used. Neither thermal nor hydraulic mixing appear between channels. Doppler, coolant density, coolant temperature, and soluble poison density feedbacks due to the thermohydraulics of the channels are described by using polynomial expansions for the group constants. The hydraulic circuit outside the reactor core consists of by-pass channel and risers with two-phase flow and of pump lines with incompressible flow. Nontrivial implicit methods are employed in the discretization of the equations to allow for sparse spatial mesh and flexible choice of time steps. Various transients can be calculated by applying external disturbances. The code is extensively supplied by input and output capabilities. TRAWA is written in FORTRAN V for UNIVAC 1108 computer. (author)
A faster reactor transient analysis methodology for PCs
International Nuclear Information System (INIS)
Ott, K.O.
1991-10-01
The simplified ANL model for LMR transient analysis, in which point kinetics as well as lumped descriptions of the heat transfer equations in all components are applied, is converted from a differential into an integral formulation. All differential balance equations are implicitly solved in terms of convolution integrals. The prompt jump approximation is applied as the strong negative feedback effectively keeps the net reactivity well below prompt critical. After implicit finite differencing of the convolution integrals, the kinetics equation assumes the form of a quadratic equation, the ''quadratic dynamics equation.'' This model forms the basis for GW-BASIC program, LTC, for LMR Transient Calculation program, which can effectively be run on a PC. The GW-BASIC version of the LTC program is described in detail in Volume 2 of this report
DYNAVAC: a transient-vacuum-network analysis code
International Nuclear Information System (INIS)
Deis, G.A.
1980-01-01
This report discusses the structure and use of the program DYNAVAC, a new transient-vacuum-network analysis code implemented on the NMFECC CDC-7600 computer. DYNAVAC solves for the transient pressures in a network of up to twenty lumped volumes, interconnected in any configuration by specified conductances. Each volume can have an internal gas source, a pumping speed, and any initial pressure. The gas-source rates can vary with time in any piecewise-linear manner, and up to twenty different time variations can be included in a single problem. In addition, the pumping speed in each volume can vary with the total gas pumped in the volume, thus simulating the saturation of surface pumping. This report is intended to be both a general description and a user's manual for DYNAVAC
Thermomechanical CSM analysis of a superheater tube in transient state
Taler, Dawid; Madejski, Paweł
2011-12-01
The paper presents a thermomechanical computational solid mechanics analysis (CSM) of a pipe "double omega", used in the steam superheaters in circulating fluidized bed (CFB) boilers. The complex cross-section shape of the "double omega" tubes requires more precise analysis in order to prevent from failure as a result of the excessive temperature and thermal stresses. The results have been obtained using the finite volume method for transient state of superheater. The calculation was carried out for the section of pipe made of low-alloy steel.
Fission gas induced fuel swelling in low and medium burnup fuel during high temperature transients
International Nuclear Information System (INIS)
Vinjamuri, K.
1980-01-01
The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results
Energy Technology Data Exchange (ETDEWEB)
Vinjamuri, K.
1980-01-01
The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results.
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 transients in the SRP test pile
International Nuclear Information System (INIS)
Church, J.P.
1976-11-01
Analysis of the hypothetical upper limit accident in the Savannah River Test Pile showed that the offsite thyroid dose from fission product release would be -3 of the 10-CFR-100 guideline dose for 95 percent of measured meteorological conditions. Offsite whole body dose would be negligible. The Test Pile was modified to limit the length of test piece that can be charged to the pile. These modifications reduce the potential offsite dose to -5 of the regulatory guidelines. Assessment of Test Pile safety included calculations of transients initiated by a variety of reactivity additions that were either terminated or not terminated by safety systems. Reactivity addition mechanisms considered were abnormally driving control rods out of the pile and charging abnormal test pieces into the pile. The transients were evaluated in the adiabatic approximation in which three-dimensional calculations of static flux shapes and reactivity were superimposed on point reactor kinetics calculations. Negative reactivity feedback effects appropriate for the pile and the temperature dependence of material properties, such as specific heat and thermal conductivity, were included. The results show that, for the worst initiators, safety systems can prevent the temperature rise from exceeding 1 0 C anywhere in the Test Pile. If the safety systems do not function, the pile temperatures will increase until the transient is ended by the inherent negative reactivity effects, including the melting of some fuel
TRAB, a transient analysis program for BWR. Part 1
International Nuclear Information System (INIS)
Rajamaeki, Markku.
1980-03-01
TRAB is a transient analysis program for BWR. The present report describes its principles. The program has been developed from TRAWA-program. It models the interior of the pressure vessel and related subsystems of BWR viz. reactor core, recirculation loop including the upper part of the vessel, recirculation pumps, incoming and outgoing flow systems, and control and protection systems. Concerning core phenomena and all flow channel hydraulics the submodels are one-dimensional of main features. The geometry is very flexible. The program has been made particularly to simulate various reactivity transients, but it is applicable more generally to reactor incidents and accidents in which no flow reversal or no emptying of the circuit must occur below the water level. The program is extensively supplied by input and output capabilities. The user can act upon the simulation of a transient by defining external disturbances, scheduled timevariations for any system variable, by modeling new subsystems, which are representable with ordinary linear differential equations, and by defining relations of functional form between system variables. The run of the program can be saved and restarted. (author)
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)
Transient thermal performance analysis of micro heat pipes
International Nuclear Information System (INIS)
Liu, Xiangdong; Chen, Yongping
2013-01-01
A theoretical analysis of transient fluid flow and heat transfer in a triangular micro heat pipes (MHP) has been conducted to study the thermal response characteristics. By introducing the system identification theory, the quantitative evaluation of the MHP's transient thermal performance is realized. The results indicate that the evaporation and condensation processes are both extended into the adiabatic section. During the start-up process, the capillary radius along axial direction of MHP decreases drastically while the liquid velocity increases quickly at the early transient stage and an approximately linear decrease in wall temperature arises along the axial direction. The MHP behaves as a first-order LTI control system with the constant input power as the 'step input' and the evaporator wall temperature as the 'output'. Two corresponding evaluation criteria derived from the control theory, time constant and temperature constant, are able to quantitatively evaluate the thermal response speed and temperature level of MHP under start-up, which show that a larger triangular groove's hydraulic diameter within 0.18–0.42 mm is able to accelerate the start-up and decrease the start-up temperature level of MHP. Additionally, the MHP starts up fastest using the fluid of ethanol and most slowly using the working fluid of methanol, and the start-up temperature reaches maximum level for acetone and minimum level for the methanol. -- Highlights: • Transient thermal response of micro heat pipe is simulated by an improved model. • Control theory is introduced to quantify the thermal response of micro heat pipe. • Evaluation criteria are proposed to represent thermal response of micro heat pipe. • Effects of groove dimensions and working fluids on start-up of micro heat pipe are evaluated
Seismic transient analysis of a containment vessel with penetrations
International Nuclear Information System (INIS)
Dahlke, H.J.; Weiner, E.O.
1979-12-01
A linear transient analysis of the FFTF containment vessel was conducted with STAGS to justify the load levels used for the seismic qualification testing of the heating and ventiliation valve operators. The modeling consists of a thin axisymmetric shell for the containment vessel with four penetrations characterized by linear and rotational inertias as well as attachment characteristics to the shell. Motions considered are horizontal, rocking and vertical input to the base, and the solution is carried out by direct integration. Results show that the test levels and the approximate analyses considered are conservative. Response spectra for some containment vessel penetrations applicable to the model are presented
Analysis of the linear induction motor in transient operation
Energy Technology Data Exchange (ETDEWEB)
Gentile, G; Rotondale, N; Scarano, M
1987-05-01
The paper deals with the analysis of a bilateral linear induction motor in transient operation. We have considered an impressed voltage one-dimensional model which takes into account end effects. The real winding distribution of the armature has been represented as a lumped parameters system. By using the space vectors methodology, the partial differential equation of the sheet is solved bythe variable separation method. Therefore it's possible to arrange a system of ordinary differential equations where the unknown quantities are the space vectors of the air-gap flux density and sheet currents. Finally, we have analyzed the characteristic quantities for a no-load starting of small power motors.
Transient thermal analysis of cryocondensation pump for JET
International Nuclear Information System (INIS)
Baxi, C.B.; Obert, W.
1993-08-01
A cryopump with pumping speed of 50,000 1/sec is planned to be installed in the Joint European Torus (JET) as part of the pumped divertor. The purpose of this pump is to control the plasma impurities. The pump consists of a helium panel cooled by supercritical helium and a nitrogen shield cooled by liquid nitrogen. This paper presents the following transient thermal flow analysis for this cryopump: 1. Consequences of loss of torus vacuum on helium panel. 2. Cool down of the nitrogen shield form 300 K to 80 K
Momentum integral network method for thermal-hydraulic transient analysis
International Nuclear Information System (INIS)
Van Tuyle, G.J.
1983-01-01
A new momentum integral network method has been developed, and tested in the MINET computer code. The method was developed in order to facilitate the transient analysis of complex fluid flow and heat transfer networks, such as those found in the balance of plant of power generating facilities. The method employed in the MINET code is a major extension of a momentum integral method reported by Meyer. Meyer integrated the momentum equation over several linked nodes, called a segment, and used a segment average pressure, evaluated from the pressures at both ends. Nodal mass and energy conservation determined nodal flows and enthalpies, accounting for fluid compression and thermal expansion
Failure analysis of carbide fuels under transient overpower (TOP) conditions
International Nuclear Information System (INIS)
Nguyen, D.H.
1980-06-01
The failure of carbide fuels in the Fast Test Reactor (FTR) under Transient Overpower (TOP) conditions has been examined. The Beginning-of-Cycle Four (BOC-4) all-oxide base case, at $.50/sec ramp rate was selected as the reference case. A coupling between the advanced fuel performance code UNCLE-T and HCDA Code MELT-IIIA was necessary for the analysis. UNCLE-T was used to determine cladding failure and fuel preconditioning which served as initial conditions for MELT-III calculations. MELT-IIIA determined the time of molten fuel ejection from fuel pin
Transient dynamic and inelastic analysis of shells of revolution
International Nuclear Information System (INIS)
Svalbonas, V.
1975-01-01
Advances in the limits of structural use in the aerospace and nuclear power industries over the past years have increased the requirements upon the applicable analytical computer programs to include accurate capabilities for inelastic and transient dynamic analyses. In many minds, however, this advanced capability is unequivocally linked with the large scale, general purpose, finite element programs. This idea is also combined with the view that, therefore, such analyses are prohibitively expensive and should be relegated to the 'last resort' classification. While this, in the general sense, may indeed be the case, if however, the user needs only to analyze structures falling into limited categories, he may find that a variety of smaller special purpose programs are available, which do not put an undue strain upon his resources. One such structural category is shells of revolution. This survey of programs will concentrate upon the analytical tools which have been developed predominantly for shells of revolution. The survey will be subdivided into three parts: a) consideration of programs for transient dynamic analysis, b) consideration of programs for inelastic analysis, and finally, c) consideration of programs capable of dynamic plasticity analysis. In each part, programs based upon finite difference, finite element, and numerical integration methods will be considered. The programs will be compared on the basis of analytical capabilities, and ease of idealization and use. In each part of the survey sample problems will be utilized to exemplify the state-of-the-art. (orig.) [de
Nonlinear Transient Thermal Analysis by the Force-Derivative Method
Balakrishnan, Narayani V.; Hou, Gene
1997-01-01
High-speed vehicles such as the Space Shuttle Orbiter must withstand severe aerodynamic heating during reentry through the atmosphere. The Shuttle skin and substructure are constructed primarily of aluminum, which must be protected during reentry with a thermal protection system (TPS) from being overheated beyond the allowable temperature limit, so that the structural integrity is maintained for subsequent flights. High-temperature reusable surface insulation (HRSI), a popular choice of passive insulation system, typically absorbs the incoming radiative or convective heat at its surface and then re-radiates most of it to the atmosphere while conducting the smallest amount possible to the structure by virtue of its low diffusivity. In order to ensure a successful thermal performance of the Shuttle under a prescribed reentry flight profile, a preflight reentry heating thermal analysis of the Shuttle must be done. The surface temperature profile, the transient response of the HRSI interior, and the structural temperatures are all required to evaluate the functioning of the HRSI. Transient temperature distributions which identify the regions of high temperature gradients, are also required to compute the thermal loads for a structural thermal stress analysis. Furthermore, a nonlinear analysis is necessary to account for the temperature-dependent thermal properties of the HRSI as well as to model radiation losses.
Probabilistic finite elements for transient analysis in nonlinear continua
Liu, W. K.; Belytschko, T.; Mani, A.
1985-01-01
The probabilistic finite element method (PFEM), which is a combination of finite element methods and second-moment analysis, is formulated for linear and nonlinear continua with inhomogeneous random fields. Analogous to the discretization of the displacement field in finite element methods, the random field is also discretized. The formulation is simplified by transforming the correlated variables to a set of uncorrelated variables through an eigenvalue orthogonalization. Furthermore, it is shown that a reduced set of the uncorrelated variables is sufficient for the second-moment analysis. Based on the linear formulation of the PFEM, the method is then extended to transient analysis in nonlinear continua. The accuracy and efficiency of the method is demonstrated by application to a one-dimensional, elastic/plastic wave propagation problem. The moments calculated compare favorably with those obtained by Monte Carlo simulation. Also, the procedure is amenable to implementation in deterministic FEM based computer programs.
Transient thermal analysis of semiconductor diode lasers under pulsed operation
Veerabathran, G. K.; Sprengel, S.; Karl, S.; Andrejew, A.; Schmeiduch, H.; Amann, M.-C.
2017-02-01
Self-heating in semiconductor lasers is often assumed negligible during pulsed operation, provided the pulses are `short'. However, there is no consensus on the upper limit of pulse width for a given device to avoid-self heating. In this paper, we present an experimental and theoretical analysis of the effect of pulse width on laser characteristics. First, a measurement method is introduced to study thermal transients of edge-emitting lasers during pulsed operation. This method can also be applied to lasers that do not operate in continuous-wave mode. Secondly, an analytical thermal model is presented which is used to fit the experimental data to extract important parameters for thermal analysis. Although commercial numerical tools are available for such transient analyses, this model is more suitable for parameter extraction due to its analytical nature. Thirdly, to validate this approach, it was used to study a GaSb-based inter-band laser and an InP-based quantum cascade laser (QCL). The maximum pulse-width for less than 5% error in the measured threshold currents was determined to be 200 and 25 ns for the GaSb-based laser and QCL, respectively.
Qualitative diagnosis for transients analysis on nuclear reactors
International Nuclear Information System (INIS)
Lorre, J.P.; Dorlet, E.; Evrard, J.M.
1995-01-01
One of the major aims of an intelligent monitoring system, is the supervision task which assist the operator in understanding what occurs on a process. Failures hypotheses must be located and the inferring process must be explained. This paper demonstrate a second generation expert system (SEXTANT) decided to the transients analysis on PWR nuclear reactors. This system detects failures by simulating the process with a numerical model. A diagnosis module uses an even graph built from a causal graph model of the plant to generate hypotheses, and a numerical model to validate these hypotheses. Hypotheses are stored into scenarios which are concurrent possible interpretations of the process evolution. The approach is illustrated by an application for the analysis of the house load operation on a pressurized water reactor. (authors). 9 refs., 10 figs
Intelligent simulations for on-line transient analysis
International Nuclear Information System (INIS)
Hassberger, J.A.; Lee, J.C.
1987-01-01
A unique combination of simulation, parameter estimation and expert systems technology is applied to the problem of diagnosing nuclear power plant transients. Knowledge-based reasoning is ued to monitor plant data and hypothesize about the status of the plant. Fuzzy logic is employed as the inferencing mechanism and an implication scheme based on observations is developed and employed to handle scenarios involving competing failures. Hypothesis testing is performed by simulating the behavior of faulted components using numerical models. A filter has been developed for systematically adjusting key model parameters to force agreement between simulations and actual plant data. Pattern recognition is employed as a decision analysis technique for choosing among several hypotheses based on simulation results. An artificial Intelligence framework based on a critical functions approach is used to deal with the complexity of a nuclear plant system. Detailed simulation results of various nuclear power plant accident scenarios are presented to demonstrate the performance and robustness properties of the diagnostic algorithm developed. The system is shown to be successful in diagnosing and identifying fault parameters for a normal reactor scram, loss-of-feedwater (LOFW) and small loss-of-coolant (LOCA) transients occurring together in a scenario similar to the accident at Three Mile Island
Transient Analysis and Dosimetry of the Tokaimura Criticality Incident
International Nuclear Information System (INIS)
Pain, Christopher C.; Oliveira, Cassiano R.E. de; Goddard, Antony J. H.; Eaton, Matthew D.; Gundry, Sarah; Umpleby, Adrian P.
2003-01-01
This paper describes research on the application of the finite element transient criticality (FETCH) code to modeling and neutron dosimetry of the Tokaimura criticality incident. FETCH has been developed to model criticality transients in single and multiphase media and is applied here to fissile solution transient criticality. Since the initial transient behavior has different time scales and physics to the longer transient behavior, the transient modeling is divided into two parts: modeling the initial transient over a time scale of seconds in which radiolytic gases and free-surface sloshing play an important role in the transient - this provides information about the dose to workers; and modeling the long-term transient behavior following the initial transient that has a time scale over hours.The neutron dosimetry of worker A who received the largest dose during the Tokaimura criticality incident is also investigated here. This dose was received mainly in the first few seconds of the ensuing nuclear criticality transient. In addition to the multiorgan dosimetry of worker A, this work provides a method of helping to evaluate the yield in the initial phase of the criticality incident; it also shows how kinetic simulations can be calibrated so that they can be applied to investigate the physics behind the incident
International Nuclear Information System (INIS)
Hsu, T.R.; Bertels, A.W.M.; Banerjee, S.; Harrison, W.C.
1976-07-01
This report presents the theoretical basis for a transient thermal elastic-plastic stress analysis of a nuclear reactor fuel element subject to severe transient thermo-mechanical loading. A finite element formulation is used for both the non-linear stress analysis and thermal analysis. These two major components are linked together to form an integrated program capable of predicting fuel element transient behaviour in two dimensions. Specific case studies are presented to illustrate capabilities of the analysis. (author)
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
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
FAST: An advanced code system for fast reactor transient analysis
International Nuclear Information System (INIS)
Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh
2005-01-01
One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems
TRAB - A transient analysis program for BWR. Part 2
International Nuclear Information System (INIS)
Raety, H.; Rajamaeki, M.
1991-05-01
TRAB is a transient analysis code for BWRs developed at the Technical Research Centre of Finland. It models the phenomena in the interior of the BWR pressure vessel and in related subsystems. The core model of TRAB can be used separately for LWR modelling. For PWR modelling the core model of TRAB is connected to circuit model SMABRE to form the SMATRA code. This report is a user's manual and documents the structure, contents and preparation of input for TRAB. The structure of TRAB input is very flexible, featuring input groups and subgroups identified with keywords and given in any order as well as data items in free format, freely mixed with explanatory texts. Users interface of the code can be used for modelling within input: through normal input it is possible to create new submodels. These may be functional or tabulated dependencies of the code variables, different types of delays, or ordinary linear differential equations
Tightly coupled transient analysis of EBR-II
International Nuclear Information System (INIS)
Makowitz, H.; Lehto, W.K.; Sackett, J.I.
1988-01-01
A Tightly Coupled transient analysis system for the Experimental Breeder Reactor-II (EBR-II) is currently being tested. The system consists of a faster than real time high fidelity reactor simulation, advanced graphics displays, expert system coupling, and real time data coupling via the EBR-II data acquisition system to and from the plant and the control system. The base, first generation software has been developed and is presently being tested. Various subsystem couplings and the total system integration are being checked out. This system should enhance the diagnostic and prognostic capability of EBR-II in the near term and provide automatic control during startup and power maneuvering in the future, as well as serve as a testbed for new control system development for advanced reactors
Quantum-corrected transient analysis of plasmonic nanostructures
Uysal, Ismail Enes
2017-03-08
A time domain surface integral equation (TD-SIE) solver is developed for quantum-corrected analysis of transient electromagnetic field interactions on plasmonic nanostructures with sub-nanometer gaps. “Quantum correction” introduces an auxiliary tunnel to support the current path that is generated by electrons tunneled between the nanostructures. The permittivity of the auxiliary tunnel and the nanostructures is obtained from density functional theory (DFT) computations. Electromagnetic field interactions on the combined structure (nanostructures plus auxiliary tunnel connecting them) are computed using a TD-SIE solver. Time domain samples of the permittivity and the Green function required by this solver are obtained from their frequency domain samples (generated from DFT computations) using a semi-analytical method. Accuracy and applicability of the resulting quantum-corrected solver scheme are demonstrated via numerical examples.
Transient analysis of intercalation electrodes for parameter estimation
Devan, Sheba
An essential part of integrating batteries as power sources in any application, be it a large scale automotive application or a small scale portable application, is an efficient Battery Management System (BMS). The combination of a battery with the microprocessor based BMS (called "smart battery") helps prolong the life of the battery by operating in the optimal regime and provides accurate information regarding the battery to the end user. The main purposes of BMS are cell protection, monitoring and control, and communication between different components. These purposes are fulfilled by tracking the change in the parameters of the intercalation electrodes in the batteries. Consequently, the functions of the BMS should be prompt, which requires the methodology of extracting the parameters to be efficient in time. The traditional transient techniques applied so far may not be suitable due to reasons such as the inability to apply these techniques when the battery is under operation, long experimental time, etc. The primary aim of this research work is to design a fast, accurate and reliable technique that can be used to extract parameter values of the intercalation electrodes. A methodology based on analysis of the short time response to a sinusoidal input perturbation, in the time domain is demonstrated using a porous electrode model for an intercalation electrode. It is shown that the parameters associated with the interfacial processes occurring in the electrode can be determined rapidly, within a few milliseconds, by measuring the response in the transient region. The short time analysis in the time domain is then extended to a single particle model that involves bulk diffusion in the solid phase in addition to interfacial processes. A systematic procedure for sequential parameter estimation using sensitivity analysis is described. Further, the short time response and the input perturbation are transformed into the frequency domain using Fast Fourier Transform
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)
Steady State and Transient Analysis of Induction Motor Driving a ...
African Journals Online (AJOL)
The importance of using a digital computer in studying the performance of Induction machine under steady and transient states is presented with computer results which show the transient behaviour of 3-phase machine during balanced and unbalanced conditions. The computer simulation for these operating conditions is ...
Analysis of transient fuel failure mechanisms: selected ANL programs
International Nuclear Information System (INIS)
Deitrich, L.W.
1975-01-01
Analytical programs at Argonne National Laboratory related to fuel pin failure mechanisms in fast-reactor accident transients are described. The studies include transient fuel pin mechanics, mechanics of unclad fuel, and mechanical effects concerning potential fuel failure propagation. (U.S.).
Transient Analysis Needs for Generation IV Reactor Concepts
International Nuclear Information System (INIS)
Siefken, L.J.; Harvego, E.A.; Coryell, E.W.; Davis, C.B.
2002-01-01
The importance of nuclear energy as a vital and strategic resource in the U. S. and world's energy supply mix has led to an initiative, termed Generation IV by the U.S. Department of Energy (DOE), to develop and demonstrate new and improved reactor technologies. These new Generation IV reactor concepts are expected to be substantially improved over the current generation of reactors with respect to economics, safety, proliferation resistance and waste characteristics. Although a number of light water reactor concepts have been proposed as Generation IV candidates, the majority of proposed designs have fundamentally different characteristics than the current generation of commercial LWRs operating in the U.S. and other countries. This paper presents the results of a review of these new reactor technologies and defines the transient analyses required to support the evaluation and future development of the Generation IV concepts. The ultimate objective of this work is to identify and develop new capabilities needed by INEEL to support DOE's Generation IV initiative. In particular, the focus of this study is on needed extensions or enhancements to SCDAP/RELAP5/3D code. This code and the RELAP5-3D code from which it evolved are the primary analysis tools used by the INEEL and others for the analysis of design-basis and beyond-design-basis accidents in current generation light water reactors. (authors)
Deterministic and Probabilistic Analysis against Anticipated Transient Without Scram
International Nuclear Information System (INIS)
Choi, Sun Mi; Kim, Ji Hwan; Seok, Ho
2016-01-01
An Anticipated Transient Without Scram (ATWS) is an Anticipated Operational Occurrences (AOOs) accompanied by a failure of the reactor trip when required. By a suitable combination of inherent characteristics and diverse systems, the reactor design needs to reduce the probability of the ATWS and to limit any Core Damage and prevent loss of integrity of the reactor coolant pressure boundary if it happens. This study focuses on the deterministic analysis for the ATWS events with respect to Reactor Coolant System (RCS) over-pressure and fuel integrity for the EU-APR. Additionally, this report presents the Probabilistic Safety Assessment (PSA) reflecting those diverse systems. The analysis performed for the ATWS event indicates that the NSSS could be reached to controlled and safe state due to the addition of boron into the core via the EBS pump flow upon the EBAS by DPS. Decay heat is removed through MSADVs and the auxiliary feedwater. During the ATWS event, RCS pressure boundary is maintained by the operation of primary and secondary safety valves. Consequently, the acceptance criteria were satisfied by installing DPS and EBS in addition to the inherent safety characteristics
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.
Deterministic and Probabilistic Analysis against Anticipated Transient Without Scram
Energy Technology Data Exchange (ETDEWEB)
Choi, Sun Mi; Kim, Ji Hwan [KHNP Central Research Institute, Daejeon (Korea, Republic of); Seok, Ho [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)
2016-10-15
An Anticipated Transient Without Scram (ATWS) is an Anticipated Operational Occurrences (AOOs) accompanied by a failure of the reactor trip when required. By a suitable combination of inherent characteristics and diverse systems, the reactor design needs to reduce the probability of the ATWS and to limit any Core Damage and prevent loss of integrity of the reactor coolant pressure boundary if it happens. This study focuses on the deterministic analysis for the ATWS events with respect to Reactor Coolant System (RCS) over-pressure and fuel integrity for the EU-APR. Additionally, this report presents the Probabilistic Safety Assessment (PSA) reflecting those diverse systems. The analysis performed for the ATWS event indicates that the NSSS could be reached to controlled and safe state due to the addition of boron into the core via the EBS pump flow upon the EBAS by DPS. Decay heat is removed through MSADVs and the auxiliary feedwater. During the ATWS event, RCS pressure boundary is maintained by the operation of primary and secondary safety valves. Consequently, the acceptance criteria were satisfied by installing DPS and EBS in addition to the inherent safety characteristics.
An analysis of transients in the PWR downcomer
International Nuclear Information System (INIS)
Jovanovic, A.
1981-01-01
The paper deals with the problem of determining non-stationary temperature field in the downcomer of a PWR type reactor. For this purpose, an analytical model has been developed. The model covers five components of (PWR - Krsko) downcomer: the core-barrel, floor between the core-barrel and the thermal shield, the thermal shield, flow between the thermal shield and the reactor vessel wall, the reactor vessel wall. The model includes internal heat generation in metal structures. The governing equations of the model have been written in the finite difference explicit form. The system of resulting algebraic equations was solved bu Gauss-Seidel method, using a modular computer code. Several characteristic transients were examined (step and continuous change of fluid temperature at the inlet nozzle). Also, an analysis of main parameters (heat transfer coefficient and flow rate) has been performed. The model is intended to be used as basics for further development of a more realistic model that could be used for practical safety analysis. (author)
Boom or bust? A comparative analysis of transient population dynamics in plants
DEFF Research Database (Denmark)
Stott, Iain; Franco, Miguel; Carslake, David
2010-01-01
researchers as further possible effectors of complicated dynamics. Previously published methods of transient analysis have tended to require knowledge of initial population structure. However, this has been overcome by the recent development of the parametric Kreiss bound (which describes how large...... a population must become before reaching its maximum possible transient amplification following a disturbance) and the extension of this and other transient indices to simultaneously describe both amplified and attenuated transient dynamics. We apply the Kreiss bound and other transient indices to a data base...... worrying artefact of basic model parameterization. Synthesis. Transient indices describe how big or how small plant populations can get, en route to long-term stable rates of increase or decline. The patterns we found in the potential for transient dynamics, across many species of plants, suggest...
The Dynamic Monte Carlo Method for Transient Analysis of Nuclear Reactors
Sjenitzer, B.L.
2013-01-01
In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing
Yamashita, Hideo; Nakamae, Eihachiro; Namera, Akihiro; Cingoski, Vlatko; Kitamura, Hideo
1998-01-01
This paper deals with design improvements on graded insulation of power transformers using transient electric field analysis and a visualization technique. The calculation method for transient electric field analysis inside a power transformer impressed with impulse voltage is presented: Initially, the concentrated electric network for the power transformer is concentrated by dividing transformer windings into several blocks and by computing the electric circuit parameters.
Shaking of reinforced concrete structures subjected to transient dynamic analysis
International Nuclear Information System (INIS)
Rouzaud, Christophe
2015-01-01
In the design of nuclear engineering structures security and safety present a crucial aspect. Civil engineering design and the qualification of materials to dynamic loads must consider the accelerations which they undergo. These accelerations could integrate seismic activity and shaking movements consecutive to aircraft impact with higher cut-off frequency. Current methodologies for assessing this shock are based on transient analyses using classical finite element method associated with explicit numerical schemes or projection on modal basis, often linear. In both cases, to represent in meaningful way a medium-frequency content, it should implement a mesh refinement which is hardly compatible with the size of models of the civil engineering structures. In order to extend industrial methodologies used and to allow a better representation of the behavior of the structure in medium-frequency, an approach coupling a temporal and non-linear analysis for shock area with a frequency approach to treatment of shaking with VTCR (Variational Theory of Complex Rays) has been used. The aim is to use the computational efficiency of the implemented strategy, including medium frequency to describe the nuclear structures to aircraft impact. (author)
Network thermodynamic approach compartmental analysis. Na+ transients in frog skin.
Mikulecky, D C; Huf, E G; Thomas, S R
1979-01-01
We introduce a general network thermodynamic method for compartmental analysis which uses a compartmental model of sodium flows through frog skin as an illustrative example (Huf and Howell, 1974a). We use network thermodynamics (Mikulecky et al., 1977b) to formulate the problem, and a circuit simulation program (ASTEC 2, SPICE2, or PCAP) for computation. In this way, the compartment concentrations and net fluxes between compartments are readily obtained for a set of experimental conditions involving a square-wave pulse of labeled sodium at the outer surface of the skin. Qualitative features of the influx at the outer surface correlate very well with those observed for the short circuit current under another similar set of conditions by Morel and LeBlanc (1975). In related work, the compartmental model is used as a basis for simulation of the short circuit current and sodium flows simultaneously using a two-port network (Mikulecky et al., 1977a, and Mikulecky et al., A network thermodynamic model for short circuit current transients in frog skin. Manuscript in preparation; Gary-Bobo et al., 1978). The network approach lends itself to computation of classic compartmental problems in a simple manner using circuit simulation programs (Chua and Lin, 1975), and it further extends the compartmental models to more complicated situations involving coupled flows and non-linearities such as concentration dependencies, chemical reaction kinetics, etc.
Numerical Analysis on Transient of Steam-gas Pressurizer
International Nuclear Information System (INIS)
Kim, Jong-Won; Lee, Yeon-Gun; Park, Goon-Cherl
2008-01-01
In nuclear reactors, various pressurizers are adopted to satisfy their characteristics and uses. The additional active systems such as heater, pressurizer cooler, spray and insulator are essential for a steam or a gas pressurizer. With a steam-gas pressurizer, additional systems are not required due to the use of steam and non-condensable gas as pressure-buffering materials. The steam-gas pressurizer in integrated small reactors experiences very complicated thermal-hydraulic phenomena. To ensure the integrity of this pressurizer type, the analysis on the transient behavior of the steam-gas pressure is indispensable. For this purpose, the steam-gas pressurizer model is introduced to predict the accurate system pressure. The proposed model includes bulk flashing, rainout, inter-region heat and mass transfer and wall condensation with non-condensable gas. However, the ideal gas law is not applied because of significant interaction at high pressure between steam and non-condensable gas. The results obtained from this proposed model agree with those from pressurizer tests. (authors)
Analysis of the FFTF primary pipe rupture transients
International Nuclear Information System (INIS)
Perkins, K.R.; Bari, R.A.; Chen, L.C.; Albright, D.C.
1979-01-01
The response of the Fast Flux Test Facility (FFTF) to hypothetical ruptures of the high pressure primary piping has been analyzed using two LMFBR plant systems codes, namely IANUS and DEMO. Comparisons of the average channel temperatures predicted by the two codes show good agreement for identical transients. However, the hot channel temperatures predicted by DEMO are about 60K higher than the corresponding IANUS predictions for severe transients. This difference is attributed to the dynamic hot channel factors employed in DEMO which discount the thermal inertia of the duct walls for rapid transients. DEMO also predicts more severe transients for hot-leg ruptures in FFTF than previously reported analyses for the CRBR
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....
DEFF Research Database (Denmark)
Wang, Yun; Wu, Qiuwei
2014-01-01
This paper analysis the electromagnetic transient response characteristics of DFIG under symmetrical and asymmetrical cascading grid fault conditions considering phaseangel jump of grid. On deriving the dynamic equations of the DFIG with considering multiple constraints on balanced and unbalanced...... conditions, phase angel jumps, interval of cascading fault, electromagnetic transient characteristics, the principle of the DFIG response under cascading voltage fault can be extract. The influence of grid angel jump on the transient characteristic of DFIG is analyzed and electromagnetic response...
A fast reactor transient analysis methodology for PCs
International Nuclear Information System (INIS)
Ott, K.O.
1991-10-01
This Manual describes a PC program for LMR Transient Calculations, LTC, written in GW-BASIC. It calculates the power and temperature trajectories for unscrammed TOP and LOHS transients. The LOF transient treatment is not operational in the GW-BASIC program because of storage limitations. The corresponding mathematical model, which allows a rapid treatment of the kinetics and the various feedback effects, is described in Ref. 1. It is briefly reviewed in Sec. 1. The program structure is outlined in Sec. 2, followed by a more detailed description in Sec. 3. Computational details are presented in Appendix A. A complete listing of the GW-BASIC program is given in Appendix B. Appendix C shows input-echo and output for a TOP sample problem, and Appendix D is a Glossary of all quantities used in the LTC program. The limitations of the GW-BASIC storage (to about 60K) are removed if it is run within Quick-BASIC. This then allows the extension of this program to treat LOF transients. Running LTC in Quick-BASIC permits also larger ''Dimensions'' for TOP and LOHS transients
DEFF Research Database (Denmark)
Li, H.; Zhao, B.; Yang, C.
2011-01-01
based on normal form theory is proposed. The transient models of the wind turbine generation system including the flexible drive train model are derived based on the direct transient stability estimation method. A method of critical clearing time (CCT) calculation is developed for the transient......Increasing levels of wind energy in modern electrical power system is initiating a need for accurate analysis and estimation of transient stability of wind turbine generation systems. This paper investigates the transient behaviors and possible direct methods for transient stability evaluation...... of a grid-connected wind turbine with squirrel cage induction generator (SCIG). Firstly, by using an equivalent lump mass method, a three-mass wind turbine equivalent model is proposed considering both the blades and the shaft flexibility of the wind turbine drive train system. Combined with the detailed...
Preliminary analysis of typical transients in fusion driven subcritical system (FDS-I)
International Nuclear Information System (INIS)
Bai Yunqing; Ke Yan; Wu Yican
2007-01-01
The potential safety characteristic is expected as one of the advantages of fusion-driven subcritical system (FDS-I) for the transmutation and incineration of nuclear waste compared with the critical reactor. Transients of the FDS-I may occur due to the perturbation of external neutron source, the failure of functional device, and the occurrence of the uncontrolled event. As typical transient scenarios, the following cases were analyzed: unprotected plasma overpower (UPOP), unprotected loss of flow (ULOF), unprotected transient overpower (UTOP). The transient analyses for the FDS-I were performed with a coupled two-dimensional thermal-hydraulics and neutronics transient analysis code NTC2D. The negative feedback of reactivity is the interesting safety feature of FDS-I as temperature increase, due to the fuel form of the circulating particle. The present simulation results showed that the current FDS-I design has a resistance against severe transient scenarios. (author)
Cernavoda unit2 recirculated cooling water system transient analysis
International Nuclear Information System (INIS)
Nita, I. P.; Pancef, R.
2015-01-01
The paper is an approach to calculate the response of Cernavoda NPP Unit 2 RCW System to transient regimes during normal and abnormal regimes. Then one started to analyse the system response to reactor trip on class III and IV of power, LOCA on class IV of power, LOCA on class III power, LOIA on class IV of power, and LOIA on class III power. Moreover, one analysed the system transient due to requirement of changeover of a RCW operating pump, planned and unplanned changeover. This is the first transient approach to this system that took in consideration all building of the system, obtaining a very large system model, with over 900 pipe, 4 pumps, 50 consumers, 21 control valves. The changeover procedure was required to be analysed in order to change the nominal operating mode for Unit 2, from current 2 pumps in operation to 3 pump operations during summer operating mode. (authors)
Analysis of the transient compressible vapor flow in heat pipes
Jang, J. H.; Faghri, A.; Chang, W. S.
1989-01-01
The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual vapor flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds and high mass flow rates are successfully predicted. The one-dimensional model also describes the vapor flow dynamics in cylindrical heat pipes at high temperatures.
Analysis of the transient compressible vapor flow in heat pipe
International Nuclear Information System (INIS)
Jang, J.H.; Faghri, A.; Chang, W.S.
1989-07-01
The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual vapor flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds and high mass flow rates are successfully predicted. The one-dimensional model also describes the vapor flow dynamics in cylindrical heat pipes at high temperatures
Analysis of the transient compressible vapor flow in heat pipe
Jang, Jong Hoon; Faghri, Amir; Chang, Won Soon
1989-01-01
The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds and high mass flow rates are successfully predicted. The one-dimensional model also describes the vapor flow dynamics in cylindrical heat pipes at high temperatures.
Directory of Open Access Journals (Sweden)
Huan-Feng Duan
2017-10-01
Full Text Available This paper investigates the impacts of non-uniformities of pipe diameter (i.e., an inhomogeneous cross-sectional area along pipelines on transient wave behavior and propagation in water supply pipelines. The multi-scale wave perturbation method is firstly used to derive analytical solutions for the amplitude evolution of transient pressure wave propagation in pipelines, considering regular and random variations of cross-sectional area, respectively. The analytical analysis is based on the one-dimensional (1D transient wave equation for pipe flow. Both derived results show that transient waves can be attenuated and scattered significantly along the longitudinal direction of the pipeline due to the regular and random non-uniformities of pipe diameter. The obtained analytical results are then validated by extensive 1D numerical simulations under different incident wave and non-uniform pipe conditions. The comparative results indicate that the derived analytical solutions are applicable and useful to describe the wave scattering effect in complex pipeline systems. Finally, the practical implications and influence of wave scattering effects on transient flow analysis and transient-based leak detection in urban water supply systems are discussed in the paper.
UNSUPERVISED TRANSIENT LIGHT CURVE ANALYSIS VIA HIERARCHICAL BAYESIAN INFERENCE
Energy Technology Data Exchange (ETDEWEB)
Sanders, N. E.; Soderberg, A. M. [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Betancourt, M., E-mail: nsanders@cfa.harvard.edu [Department of Statistics, University of Warwick, Coventry CV4 7AL (United Kingdom)
2015-02-10
Historically, light curve studies of supernovae (SNe) and other transient classes have focused on individual objects with copious and high signal-to-noise observations. In the nascent era of wide field transient searches, objects with detailed observations are decreasing as a fraction of the overall known SN population, and this strategy sacrifices the majority of the information contained in the data about the underlying population of transients. A population level modeling approach, simultaneously fitting all available observations of objects in a transient sub-class of interest, fully mines the data to infer the properties of the population and avoids certain systematic biases. We present a novel hierarchical Bayesian statistical model for population level modeling of transient light curves, and discuss its implementation using an efficient Hamiltonian Monte Carlo technique. As a test case, we apply this model to the Type IIP SN sample from the Pan-STARRS1 Medium Deep Survey, consisting of 18,837 photometric observations of 76 SNe, corresponding to a joint posterior distribution with 9176 parameters under our model. Our hierarchical model fits provide improved constraints on light curve parameters relevant to the physical properties of their progenitor stars relative to modeling individual light curves alone. Moreover, we directly evaluate the probability for occurrence rates of unseen light curve characteristics from the model hyperparameters, addressing observational biases in survey methodology. We view this modeling framework as an unsupervised machine learning technique with the ability to maximize scientific returns from data to be collected by future wide field transient searches like LSST.
UNSUPERVISED TRANSIENT LIGHT CURVE ANALYSIS VIA HIERARCHICAL BAYESIAN INFERENCE
International Nuclear Information System (INIS)
Sanders, N. E.; Soderberg, A. M.; Betancourt, M.
2015-01-01
Historically, light curve studies of supernovae (SNe) and other transient classes have focused on individual objects with copious and high signal-to-noise observations. In the nascent era of wide field transient searches, objects with detailed observations are decreasing as a fraction of the overall known SN population, and this strategy sacrifices the majority of the information contained in the data about the underlying population of transients. A population level modeling approach, simultaneously fitting all available observations of objects in a transient sub-class of interest, fully mines the data to infer the properties of the population and avoids certain systematic biases. We present a novel hierarchical Bayesian statistical model for population level modeling of transient light curves, and discuss its implementation using an efficient Hamiltonian Monte Carlo technique. As a test case, we apply this model to the Type IIP SN sample from the Pan-STARRS1 Medium Deep Survey, consisting of 18,837 photometric observations of 76 SNe, corresponding to a joint posterior distribution with 9176 parameters under our model. Our hierarchical model fits provide improved constraints on light curve parameters relevant to the physical properties of their progenitor stars relative to modeling individual light curves alone. Moreover, we directly evaluate the probability for occurrence rates of unseen light curve characteristics from the model hyperparameters, addressing observational biases in survey methodology. We view this modeling framework as an unsupervised machine learning technique with the ability to maximize scientific returns from data to be collected by future wide field transient searches like LSST
Peach Bottom transient analysis with BWR TRACB02
International Nuclear Information System (INIS)
Alamgir, M.; Sutherland, W.A.
1984-01-01
TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve
Numerical analysis of power system transients and dynamics
Ametani, Akihiro
2015-01-01
This book describes the three major power system transient and dynamics simulation tools based on a circuit-theory based approach which are most widely used all over the world (EMTP-ATP, EMTP-RV and EMTDC/PSCAD), together with other powerful simulation tools such as XTAP.
An analysis of power transients observed in SPERT I reactors
International Nuclear Information System (INIS)
Clancy, B.E.; Connolly, J.W.; Harrington, B.V.
1976-04-01
The analytical method described in Part I of this series has been applied to the calculation of spert I transients performed with higher initial moderator temperatures and also to those performed in a highly undermoderated core. Reasonable agreement has been obtained between calculated and experimental burst data. (author)
Vibrational Analysis of (SCN)2 and the Transient (SCN)2
DEFF Research Database (Denmark)
Jensen, N. H.; Wilbrandt, Robert Walter; Pagsberg, Palle Bjørn
1979-01-01
The vibrational spectra of thiocyanogen and the transient radical anion (SCN)2− are interpreted in detail through molecular orbital and normal coordinate calculations. The results support the assignment of (SCN)2− to the anion of thiocyanogen and indicate a substantial weakening of the S–S and C......≡N bonds in going from the parent molecule to its radical anion....
Transient analysis of a variable speed rotary compressor
International Nuclear Information System (INIS)
Park, Youn Cheol
2010-01-01
A transient simulation model of a rolling piston type rotary compressor is developed to predict the dynamic characteristics of a variable speed compressor. The model is based on the principles of conservation, real gas equations, kinematics of the crankshaft and roller, mass flow loss due to leakage, and heat transfer. For the computer simulation of the compressor, the experimental data were obtained from motor performance tests at various operating frequencies. Using the developed model, re-expansion loss, friction loss, mass flow loss and heat transfer loss is estimated as a function of the crankshaft speed in a variable speed compressor. In addition, the compressor efficiency and energy losses are predicted at various compressor-operating frequencies. Since the transient state of the compressor strongly depends on the system, the developed model is combined with a transient system simulation program to get transient variations of the compression process in the system. Motor efficiency, mechanical efficiency, motor torque and volumetric efficiency are calculated with respect to variation of the driving frequency in a rotary compressor.
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
Development of three dimensional transient analysis code STTA for SCWR core
International Nuclear Information System (INIS)
Wang, Lianjie; Zhao, Wenbo; Chen, Bingde; Yao, Dong; Yang, Ping
2015-01-01
Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation
EURDYN, Nonlinear Transient Analysis of Structure with Dynamic Loads
International Nuclear Information System (INIS)
Donea, J.; Giuliani, S.; Halleux, J.P.
1987-01-01
1 - Description of program or function: The EURDYN computer codes are under development at JRC-Ispra since 1973 for the simulation of non- linear dynamic response of fast-reactor components submitted to impulsive loading due to abnormal working conditions. They are thus mainly used in reactor safety analysis but can apply to other fields. Indeed the codes compute the elasto-plastic transient response of 2-D and thin 3-D structures submitted to fast dynamic loading generated by explosions, impacts... and represented by time dependent pressures, concentrated loads and prescribed displacements, or by initial speeds. Two releases of the structural computer codes EURDYN 01 (2-D beams and triangles and axisymmetric conical shells and triangular tori), 02 (axisymmetric and 2-D quadratic iso-parametric elements) and 03 (triangular plate elements) have already been produced in 1976(1) and 1980(2). They include material (elasto-plasticity using the classical flow theory approach) and geometrical (large displacements and rotations treated by a co-rotational technique) nonlinearities. The present version (Release 3) has been completed mid-1982 and is documented in EUR 8357 EN. The new features of Release 3, as compared to the former ones, roughly consist in: - full large strain capability for 9-node iso-parametric elements (EURDYN 02), - generalized array dimensions, - introduction of the radial return algorithm for elasto-plastic material modelling, - extension of the energy check facility to the case of prescribed displacements, - possible interface to a post-processing package including time plot facilities (TPLOT). The theoretical aspects can be found in refs. 2,4,5,6,7,8. 2 - Method of solution: - Finite element space discretization. - Explicit time integration. - Lumped masses. - EURDYN 01: 2-D co-rotational formulation including constant strain triangles (plane or axisymmetric), beams and conical shells, this last element being particularly useful for the study of thin
Transient analysis for a system with a tilted disc check valve
International Nuclear Information System (INIS)
Jeung, Jaesik; Lee, Kyukwang; Cho, Daegwan
2014-01-01
Check valves are used to prevent reverse flow conditions in a variety of systems in nuclear power plants. When a check valve is closed by a reverse flow, the transient load can jeopardize the structural integrity on the piping system and its supports. It may also damage intended function of the in-line components even though the severity of the load differs and depends strongly on types of the check valves. To incorporate the transient load in the piping system, it is very important to properly predict the system response to transients such as a check valve closure accompanied by pump trip and to evaluate the system transient. The one-dimensional transient simulation codes such as the RELAP5/MOD3.3 and TRACE were used. There has not been a single model that integrates the two codes to handle the behavior of a tilted disc check valve, which is designed to mitigate check valve slams by shorting the travel of the disc. In this paper a model is presented to predict the dynamic motion of a tilted disc check valve in the transient simulation using the RELAP5/MOD3.3 code and the model is incorporated in a system transient analysis using control variables of the code. In addition, transient analysis for Essential Service Water (ESW) system is performed using the proposed model and the associated load is evaluated for the system. (author)
Transient Dynamic Mechanical Analysis of Resilin-based Elastomeric Hydrogels
Li, Linqing; Kiick, Kristi
2014-04-01
The outstanding high-frequency properties of emerging resilin-like polypeptides (RLPs) have motivated their development for vocal fold tissue regeneration and other applications. Recombinant RLP hydrogels show efficient gelation, tunable mechanical properties, and display excellent extensibility, but little has been reported about their transient mechanical properties. In this manuscript, we describe the transient mechanical behavior of new RLP hydrogels investigated via both sinusoidal oscillatory shear deformation and uniaxial tensile testing. Oscillatory stress relaxation and creep experiments confirm that RLP-based hydrogels display significantly reduced stress relaxation and improved strain recovery compared to PEG-based control hydrogels. Uniaxial tensile testing confirms the negligible hysteresis, reversible elasticity and superior resilience (up to 98%) of hydrated RLP hydrogels, with Young’s modulus values that compare favorably with those previously reported for resilin and that mimic the tensile properties of the vocal fold ligament at low strain (engineering applications, of a range of RLP hydrogels.
Transient space-time surface waves characterization using Gabor analysis
Energy Technology Data Exchange (ETDEWEB)
Martinez, L; Wilkie-Chancellier, N; Caplain, E [Universite de Cergy Pontoise, ENS Cachan, UMR CNRS 8029, Laboratoire Systemes et Applications des Techniques de l' Information et de l' Energie (SATIE), 5 mail Gay-Lussac, F 9500 Cergy-Pontoise (France); Glorieux, C; Sarens, B, E-mail: nicolas.wilkie-chancellier@u-cergy.f [Katholieke Universiteit Leuven, Laboratorium voor Akoestiek en Thermische Fysica (LATF), Celestijnenlaan 200D, B-3001 Leuven (Belgium)
2009-11-01
Laser ultrasonics allow the observation of transient surface waves along their propagation media and their interaction with encountered objects like cracks, holes, borders. In order to characterize and localize these transient aspects in the Space-Time-Wave number-Frequency domains, the 1D, 2D and 3D Gabor transforms are presented. The Gabor transform enables the identification of several properties of the local wavefronts such as their shape, wavelength, frequency, attenuation, group velocity and the full conversion sequence along propagation. The ability of local properties identification by Gabor transform is illustrated by two experimental studies: Lamb waves generated by an annular source on a circular quartz and Lamb wave interaction with a fluid droplet. In both cases, results obtained with Gabor transform enable ones to identify the observed local waves.
Dynamic remedial action scheme using online transient stability analysis
Shrestha, Arun
Economic pressure and environmental factors have forced the modern power systems to operate closer to their stability limits. However, maintaining transient stability is a fundamental requirement for the operation of interconnected power systems. In North America, power systems are planned and operated to withstand the loss of any single or multiple elements without violating North American Electric Reliability Corporation (NERC) system performance criteria. For a contingency resulting in the loss of multiple elements (Category C), emergency transient stability controls may be necessary to stabilize the power system. Emergency control is designed to sense abnormal conditions and subsequently take pre-determined remedial actions to prevent instability. Commonly known as either Remedial Action Schemes (RAS) or as Special/System Protection Schemes (SPS), these emergency control approaches have been extensively adopted by utilities. RAS are designed to address specific problems, e.g. to increase power transfer, to provide reactive support, to address generator instability, to limit thermal overloads, etc. Possible remedial actions include generator tripping, load shedding, capacitor and reactor switching, static VAR control, etc. Among various RAS types, generation shedding is the most effective and widely used emergency control means for maintaining system stability. In this dissertation, an optimal power flow (OPF)-based generation-shedding RAS is proposed. This scheme uses online transient stability calculation and generator cost function to determine appropriate remedial actions. For transient stability calculation, SIngle Machine Equivalent (SIME) technique is used, which reduces the multimachine power system model to a One-Machine Infinite Bus (OMIB) equivalent and identifies critical machines. Unlike conventional RAS, which are designed using offline simulations, online stability calculations make the proposed RAS dynamic and adapting to any power system
Transient Response Analysis of Metropolis Learning in Games
Jaleel, Hassan
2017-10-19
The objective of this work is to provide a qualitative description of the transient properties of stochastic learning dynamics like adaptive play, log-linear learning, and Metropolis learning. The solution concept used in these learning dynamics for potential games is that of stochastic stability, which is based on the stationary distribution of the reversible Markov chain representing the learning process. However, time to converge to a stochastically stable state is exponential in the inverse of noise, which limits the use of stochastic stability as an effective solution concept for these dynamics. We propose a complete solution concept that qualitatively describes the state of the system at all times. The proposed concept is prevalent in control systems literature where a solution to a linear or a non-linear system has two parts, transient response and steady state response. Stochastic stability provides the steady state response of stochastic learning rules. In this work, we study its transient properties. Starting from an initial condition, we identify the subsets of the state space called cycles that have small hitting times and long exit times. Over the long time scales, we provide a description of how the distributions over joint action profiles transition from one cycle to another till it reaches the globally optimal state.
Transient Response Analysis of Metropolis Learning in Games
Jaleel, Hassan; Shamma, Jeff S.
2017-01-01
The objective of this work is to provide a qualitative description of the transient properties of stochastic learning dynamics like adaptive play, log-linear learning, and Metropolis learning. The solution concept used in these learning dynamics for potential games is that of stochastic stability, which is based on the stationary distribution of the reversible Markov chain representing the learning process. However, time to converge to a stochastically stable state is exponential in the inverse of noise, which limits the use of stochastic stability as an effective solution concept for these dynamics. We propose a complete solution concept that qualitatively describes the state of the system at all times. The proposed concept is prevalent in control systems literature where a solution to a linear or a non-linear system has two parts, transient response and steady state response. Stochastic stability provides the steady state response of stochastic learning rules. In this work, we study its transient properties. Starting from an initial condition, we identify the subsets of the state space called cycles that have small hitting times and long exit times. Over the long time scales, we provide a description of how the distributions over joint action profiles transition from one cycle to another till it reaches the globally optimal state.
Analysis of the Mannshan Unit 2 full load rejection transient
International Nuclear Information System (INIS)
Kang, J.C.; Pei, B.S.; Yu, G.P.; Yuann, R.Y.
1987-01-01
Mannshan Unit 2 is a Westinghouse three-loop pressurized water reactor with a rated core power of 2775 MW(thermal) and a rated core flow of 4702 kg/s. Before full power operation, a planned net load rejection was performed during the startup test by opening the main transformer highside breakers. The generator power rapidly reduced to station load. All 16 steam dump valves immediately popped open, and control bank-D rods automatically stepped in as the temperature difference T/sub avg/ - T/sub ref/ reached a programmed 2.8 0 C. Nuclear power decreased smoothly as control rods were inserted into the core. The pressurizer pressure and liquid levels also dropped. Neither safety injection nor reactor trip occurred during this transient. The test was done to verify that the whole system would function properly under a transient to keep the reactor from scramming and that the vessel integrity would also be protected. In this study, which is the preliminary stage of RELAP5/MOD2 transient simulation of the Mannshan PWR plants, system thermal-hydraulic response is tested first and isolated from the neutronic effects. The variation of core power versus time curve was extracted from the power test data to serve as a time varying boundary condition. The comparison of the analytical results of four major parameters (pressurizer pressure, average temperature of the core, steam dump flow rate, and feedwater flow rate) from RELAP5/MOD2 and the power test data is illustrated
Transient analysis of LMFBR reinforced/prestressed concrete containment
International Nuclear Information System (INIS)
Marchertas, A.H.; Belytschko, T.B.; Bazant, Z.P.
1979-01-01
The use of prestressed concrete reactor vessels (PCRVs) for LMFBR containment creates a need for analytical methods for treating the transient response of such structures, for LMFBR containments must be capable of sustaining the dynamic effects which arise in a hypothetical core disruptive accident (HCDA). These analyses require several unique features: a model of concrete which includes tensile cracking, a methodology for representing the prestressing tendons and for simulating the prestressing operation, and an efficient computational tool for treating the transient response. Furthermore, for the sake of convenience, all of these features should be available in a single computer code. For the purpose of treating the transient response, a finite element program with explicit time integration was chosen. The use of explicit time integration has the advantage that it can easily treat the complicated constitutive model which arises from the considerations of concrete cracking and it can handle the slip between reinforcing tendons and the concrete through the use of the well known sliding interface options. However, explicit time integration programs are usually not well suited to the simulation of static processes such as prestressing. Nevertheless, explicit time integration programs can handle static processes through the introduction of damping by what is known as a dynamic relaxation procedure. For this reason, the dynamic relaxation procedure was refined through the introduction of lumped mass, viscous damping. This provision made the prestressing operation of the concrete structures by means of the explicit formulation rather convenient. (orig.)
Peptidomics Analysis of Transient Regeneration in the Neonatal Mouse Heart.
Fan, Yi; Zhang, Qijun; Li, Hua; Cheng, Zijie; Li, Xing; Chen, Yumei; Shen, Yahui; Wang, Liansheng; Song, Guixian; Qian, Lingmei
2017-09-01
Neonatal mouse hearts have completely regenerative capability after birth, but the ability to regenerate rapidly lost after 7 days, the mechanism has not been clarified. Previous studies have shown that mRNA profile of adult mouse changed greatly compared to neonatal mouse. So far, there is no research of peptidomics related to heart regeneration. In order to explore the changes of proteins, enzymes, and peptides related to the transient regeneration, we used comparative petidomics technique to compare the endogenous peptides in the mouse heart of postnatal 1 and 7 days. In final, we identified 236 differentially expressed peptides, 169 of which were upregulated and 67 were downregulated in the postnatal 1 day heart, and also predicted 36 functional peptides associated with transient regeneration. The predicted 36 candidate peptides are located in the important domains of precursor proteins and/or contain the post-transcriptional modification (PTM) sites, which are involved in the biological processes of cardiac development, cardiac muscle disease, cell proliferation, necrosis, and apoptosis. In conclusion, for the first time, we compared the peptidomics profiles of neonatal heart between postnatal 1 day and postnatal 7 day. This study provides a new direction and an important basis for the mechanism research of transient regeneration in neonatal heart. J. Cell. Biochem. 118: 2828-2840, 2017. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.
Mitigation method of thermal transient stress by thermalhydraulic-structure total analysis
International Nuclear Information System (INIS)
Kasahara, Naoto; Jinbo, Masakazu; Hosogai, Hiromi
2003-01-01
This study proposes a rational evaluation and mitigation method of thermal transient loads in fast reactor components by utilizing relationships among plant system parameters and stresses induced by thermal transients of plants. A thermalhydraulic-structure total analysis procedure helps us to grasp relationship among system parameters and thermal stresses. Furthermore, it enables mitigation of thermal transient loads by adjusting system parameters. In order to overcome huge computations, a thermalhydraulic-structure total analysis code and the Design of Experiments methodology are utilized. The efficiency of the proposed mitigation method is validated through thermal stress evaluation of an intermediate heat exchanger in Japanese demonstration fast reactor. (author)
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...
Transient analysis of a bunched beam free electron laser
International Nuclear Information System (INIS)
Wang, J.M.; Yu, L.H.
1985-01-01
The problem of the bunched beam operation of a free electron laser was studied. Assuming the electron beam to be initially monoenergetic, the Maxwell-Vlasov equations describing the system reduce to a third order partial differential equation for the envelope of the emitted light. The Green's function corresponding to an arbitrary shape of the electron bunch, which describes the transient behavior of the system, is obtained. The Green's function was used to discuss the start up problem as well as the power output and the power specrum of a self-amplified spontaneous emission
Analytical transient analysis of Peltier device for laser thermal tuning
Sheikhnejad, Yahya; Vujicic, Zoran; Almeida, Álvaro J.; Bastos, Ricardo; Shahpari, Ali; Teixeira, António L.
2017-08-01
Recently, industrial trends strongly favor the concepts of high density, low power consumption and low cost applications of Datacom and Telecom pluggable transceiver modules. Hence, thermal management plays an important role, especially in the design of high-performance compact optical transceivers. Extensive care should be taken on wavelength drift for thermal tuning lasers using thermoelectric cooler and indeed, accurate expression is needed to describe transient characteristics of the Peltier device to achieve maximum controllability. In this study, the exact solution of governing equation is presented, considering Joule heating, heat conduction, heat flux of laser diode and thermoelectric effect in one dimension.
Analysis of transient phenomena in hydroelectric generation plants
Energy Technology Data Exchange (ETDEWEB)
Calendray, J.F.; Ilhat, D.; Planchard, J.; Lauro, J.F.; Velo, C.
1986-01-01
The construction in recent years of a number of pumping power transfer plants and overequipment of existing hydraulic systems required Electricite de France to acquire a program to simulate the transient states in the most complex systems. A computation tool - the Belier code - was therefore developed to calculate pressures and flows in any point of a water system which can include Francis and Pelton turbines, valves, vents, etc. After a brief review of the computation methods used, a number of recent plants designed using this program are described and comparisons with measurements on site are given.
Analysis of cofrentes abnormal plant transients with RETRAN-02 and RETRAN-03
International Nuclear Information System (INIS)
Mata, P.; Sedano, P.G.; Serra, J.
1992-01-01
In this paper the applicability and usefulness of a complete and well-qualified plant transient code and model to support in-depth evaluation of anomalous plant transients are described. The qualified best-estimate RETRAN-02 model for the Cofrentes nuclear power plant (a boiling water reactor with an uprated power of 2952 MW) has been updated for RETRAN-03 using algebraic slip and one-dimensional kinetics. This model has been used in the analysis of recent abnormal plant transients at Cofrentes, including a partial control rod insertion at 92% power, a turbine trip at 67% power with reactor vessel overfill, and reactor instabilities during startup
Tool for Turbine Engine Closed-Loop Transient Analysis (TTECTrA) Users' Guide
Csank, Jeffrey T.; Zinnecker, Alicia M.
2014-01-01
The tool for turbine engine closed-loop transient analysis (TTECTrA) is a semi-automated control design tool for subsonic aircraft engine simulations. At a specific flight condition, TTECTrA produces a basic controller designed to meet user-defined goals and containing only the fundamental limiters that affect the transient performance of the engine. The purpose of this tool is to provide the user a preliminary estimate of the transient performance of an engine model without the need to design a full nonlinear controller.
The development of the fuel rod transient performance analysis code FTPAC
International Nuclear Information System (INIS)
Han Zhijie; Ji Songtao
2014-01-01
Fuel rod behavior, especially the integrity of cladding, played an important role in fuel safety research during reactor transient and hypothetical accidents conditions. In order to study fuel rod performance under transient accidents, FTPAC (Fuel Transient Performance Analysis Code) has been developed for simulating light water reactor fuel rod transient behavior when power or coolant boundary conditions are rapidly changing. It is composed of temperature, mechanical deformation, cladding oxidation and gas pressure model. The assessment was performed by comparing FTPAC code analysis result to experiments data and FRAPTRAN code calculations. Comparison shows that, the FTPAC gives reasonable agreement in temperature, deformation and gas pressure prediction. And the application of slip coefficient is more suitable for simulating the sliding between pellet and cladding when the gap is closed. (authors)
LMFBR system-wide transient analysis: the state of the art and US validation needs
International Nuclear Information System (INIS)
Khatib-Rahbar, M.; Guppy, J.G.; Cerbone, R.J.
1982-01-01
This paper summarizes the computational capabilities in the area of liquid metal fast breeder reactor (LMFBR) system-wide transient analysis in the United States, identifies various numerical and physical approximations, the degree of empiricism, range of applicability, model verification and experimental needs for a wide class of protected transients, in particular, natural circulation shutdown heat removal for both loop- and pool-type plants
Development of a computer code for Dalat research reactor transient analysis
International Nuclear Information System (INIS)
Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong
2003-01-01
DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)
Developing and investigating a pure Monte-Carlo module for transient neutron transport analysis
International Nuclear Information System (INIS)
Mylonakis, Antonios G.; Varvayanni, M.; Grigoriadis, D.G.E.; Catsaros, N.
2017-01-01
Highlights: • Development and investigation of a Monte-Carlo module for transient neutronic analysis. • A transient module developed on the open-source Monte-Carlo static code OpenMC. • Treatment of delayed neutrons is inserted. • Simulation of precursors’ decay process is performed. • Transient analysis of simplified test-cases. - Abstract: In the field of computational reactor physics, Monte-Carlo methodology is extensively used in the analysis of static problems while the transient behavior of the reactor core is mostly analyzed using deterministic algorithms. However, deterministic algorithms make use of various approximations mainly in the geometric and energetic domain that may induce inaccuracy. Therefore, Monte-Carlo methodology which generally does not require significant approximations seems to be an attractive candidate tool for the analysis of transient phenomena. One of the most important constraints towards this direction is the significant computational cost; however since nowadays the available computational resources are continuously increasing, the potential use of the Monte-Carlo methodology in the field of reactor core transient analysis seems feasible. So far, very few attempts to employ Monte-Carlo methodology to transient analysis have been reported. Even more, most of those few attempts make use of several approximations, showing the existence of an “open” research field of great interest. It is obvious that comparing to static Monte-Carlo, a straight-forward physical treatment of a transient problem requires the temporal evolution of the simulated neutrons; but this is not adequate. In order to be able to properly analyze transient reactor core phenomena, the proper simulation of delayed neutrons together with other essential extensions and modifications is necessary. This work is actually the first step towards the development of a tool that could serve as a platform for research and development on this interesting but also
ANO-2 turbine trip transient test analysis using MMS
International Nuclear Information System (INIS)
Jain, P.K.; Divakaruni, S.M.
1984-01-01
The data from the turbine trip transient tests conducted at the Arkansas Nuclear One-Unit 2 was used as one of the benchmark cases for validating the Modular Modeling System (MMS) Code, developed by the Electric Power Research Institute (EPRI). The data was used first to validate the modules in stand-alone simulation tests and then in a Nuclear Steam Supply system integral tests. This paper presents the results from the MMS simulation effort and compares the code generated results with the plant data as well as RETRAN results. In general, MMS simulation results compare very well with the plant data. The code calculations for the hot and cold leg temperatures, primary system pressure and the pressurizer level are very good compared to RETRAN; however, MMS results for steam generator level compare reasonably well only with RETRAN calculations
Analysis of pump's shaft torsional vibrations in transient conditions
International Nuclear Information System (INIS)
Pasqualini, G.R.; Cauquelin, C.
1989-01-01
When the voltage is applied to an induction motor, the currents in the stator's phases are subject to a transient period. It is consequently also the case for the torques. A method to calculate the torque in the case of an induction motor with deep bars is presented. A model is proposed to represent the squirrel cage. It allows to take into account the fact the currents are not sinusoidal and that, in this case, the rotor's winding cannot be represented by only one resistance and once reactance. The electrical model is completed by a mechanical model for the shaftline. The calculation is realized for the start up of an reactor coolant pump. A comparison is made between the results given by the new model, by the classical model and by tests
Implicit analysis of the transient water flow with dissolved air
Directory of Open Access Journals (Sweden)
J. Twyman
2018-01-01
Full Text Available The implicit finite-difference method (IFDM for solving a system that transports water with dissolved air using a fixed (or variable rectangular space-time mesh defined by the specified time step method is applied. The air content in the fluid modifies both the wave speed and the Courant number, which makes it inconvenient to apply the traditional Method of Characteristics (MOC and other explicit schemes due to their impossibility to simulate the changes in magnitude, shape and frequency of the pressures train. The conclusion is that the IFDM delivers an accurate and stable solution, with a good adjustment level with respect to a classical case reported in the literature, being a valid alternative for the transient solution in systems that transport water with dissolved air.
NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR
Directory of Open Access Journals (Sweden)
Surian Pinem
2016-01-01
Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.
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)
Analysis of ventilation systems subjected to explosive transients: far-field analysis
International Nuclear Information System (INIS)
Tang, P.K.; Andrae, R.W.; Bolstad, J.W.; Duerre, K.H.; Gregory, W.S.
1981-11-01
Progress in developing a far-field explosion simulation computer code is outlined. The term far-field implies that this computer code is suitable for modeling explosive transients in ventilation systems that are far removed from the explosive event and are rather insensitive to the particular characteristics of the explosive event. This type of analysis is useful when little detailed information is available and the explosive event is described parametrically. The code retains all the features of the TVENT code and allows completely compressible flow with inertia and choking effects. Problems that illustrate the capabilities and limitations of the code are described
Gas-core reactor power transient analysis. Final report
International Nuclear Information System (INIS)
Kascak, A.F.
1972-01-01
The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)
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.)
Preliminary analysis of the transient overpower accident for CRBRP. Final report
International Nuclear Information System (INIS)
Kastenberg, W.E.; Frank, M.V.
1975-07-01
A preliminary analysis of the transient overpower accident for the Clinch River Breeder Reactor Plant (CRBRP) is presented. Several uncertainties in the analysis and the estimation of ramp rates during the transition to disassembly are discussed. The major conclusions are summarized
Uncertainty and sensitivity analysis applied to coupled code calculations for a VVER plant transient
International Nuclear Information System (INIS)
Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K. D.
2004-01-01
The development of coupled codes, combining thermal-hydraulic system codes and 3D neutron kinetics, is an important step to perform best-estimate plant transient calculations. It is generally agreed that the application of best-estimate methods should be supplemented by an uncertainty and sensitivity analysis to quantify the uncertainty of the results. The paper presents results from the application of the GRS uncertainty and sensitivity method for a VVER-440 plant transient, which was already studied earlier for the validation of coupled codes. For this application, the main steps of the uncertainty method are described. Typical results of the method applied to the analysis of the plant transient by several working groups using different coupled codes are presented and discussed The results demonstrate the capability of an uncertainty and sensitivity analysis. (authors)
Validation of the probabilistic approach for the analysis of PWR transients
International Nuclear Information System (INIS)
Amesz, J.; Francocci, G.F.; Clarotti, C.
1978-01-01
This paper reviews the pilot study at present being carried out on the validation of probabilistic methodology with real data coming from the operational records of the PWR power station at Obrigheim (KWO, Germany) operating since 1969. The aim of this analysis is to validate the a priori predictions of reactor transients performed by a probabilistic methodology, with the posteriori analysis of transients that actually occurred at a power station. Two levels of validation have been distinguished: (a) validation of the rate of occurrence of initiating events; (b) validation of the transient-parameter amplitude (i.e., overpressure) caused by the above mentioned initiating events. The paper describes the a priori calculations performed using a fault-tree analysis by means of a probabilistic code (SALP 3) and event-trees coupled with a PWR system deterministic computer code (LOOP 7). Finally the principle results of these analyses are presented and critically reviewed
Transient analysis of the new Cold Source at the FRM-II
International Nuclear Information System (INIS)
Gutsmiedl, E.; Posselt, H.; Scheuer, A.
2003-01-01
The new Cold Source (CNS) at the FRM-II research reactor is completely installed. This paper reports on the results of the transient analysis in the design status for this facility for producing cold neutrons for neutron experiments, the implementation of the results in the design of the mechanical components, the measurements at the cold tests and the comparison with the data of the transient analysis. The important load cases are fixed in the system description and the design data sheet of the CNS. A transient analysis was done with the computer program ESATAN, the nodal configuration was identical with the planned system of the CNS and the boundary conditions were chosen so, that conservative results can be expected. The following transients of the load cases in the piping system behind the inpile part 1) normal storage of D 2 at the hydride storage vessel 2) breakdown of cooling system of the CNS and transfer of D 2 to the buffer tank 3) rapid charge of D 2 to the buffer tank with break of the insulation vacuum and flooding of Neon 4) reloading of the D 2 from the buffer tank to the D 2 hydride storage vessel were calculated. Additionally the temperature distribution for these transients in the connecting flanges of the systems to the inpile part were analysed. The temperature distributions in the flange region were take into account for the strength calculation of the flange construction. The chosen construction shows allowable values and a leak tight flange connection for the load cases. The piping system was designed to the lowest expected temperatures. The load cases in the moderator tank were take into account in the stress analysis and the fatigue analysis of the vacuum vessel and the moderator vessel. The results shows allowable stresses. The results shows that a transient analysis is necessary and helpful for good design of the CNS. (author)
The importance of transient analysis in the light water reactor licensing procedure
International Nuclear Information System (INIS)
Izouierdo, J.M.; Villadoniga, J.I.
1979-01-01
The basic principles of the Nuclear Regulation are developed in the first part of this report. The achievement of the safety objective by establishing protections -that prevent or reduce the barriers failure- is analyzed. An iterative method for the definition of the systems and components safety design bases is proposed, analyzing the role of Technical Specifications in this process. The second part shows how this methodology can be used in the case of the first barrier: the fuel cladding. The safety criteria, transient clasification problems, transient analysis and its relation with surveillance and protection systems, and the role of such analysis in fuel protection design verification are discused. (author)
TRANSPA: a code for transient thermal analysis of a single fuel pin
International Nuclear Information System (INIS)
Prenger, F.C.
1985-02-01
An analytical model (TRANSPA) for the transient thermal analysis of a single uranium carbide fuel pin was developed. This model uses thermal boundary conditions obtained from COBRA-WC output and calculates the transient thermal response of a single fuel pin to changes in internal power generation, coolant flowrate, or fuel pin physical configuration. The model uses the MITAS finite difference thermal analyzer. MITAS provides the means to input separate conductance models through the use of a user subroutine input capability. The model is a lumped-mass representation of the fuel pin using 26 nodes and 42 conductors. Run time for each transient analysis is approximately one minute of central processor time on the NOS operating system
Transient Safety Analysis of Fast Spectrum TRU Burning LWRs with Internal Blankets
Energy Technology Data Exchange (ETDEWEB)
Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Zazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hill, Bob [Argonne National Lab. (ANL), Argonne, IL (United States)
2015-01-31
The objective of this proposal was to perform a detailed transient safety analysis of the Resource-Renewable BWR (RBWR) core designs using the U.S. NRC TRACE/PARCS code system. This project involved the same joint team that has performed the RBWR design evaluation for EPRI and therefore be able to leverage that previous work. And because of their extensive experience with fast spectrum reactors and parfait core designs, ANL was also part the project team. The principal outcome of this project was the development of a state-of-the-art transient analysis capability for GEN-IV reactors based on Monte Carlo generated cross sections and the US NRC coupled code system TRACE/PARCS, and a state-of-the-art coupled code assessment of the transient safety performance of the RBWR.
Identification of speech transients using variable frame rate analysis and wavelet packets.
Rasetshwane, Daniel M; Boston, J Robert; Li, Ching-Chung
2006-01-01
Speech transients are important cues for identifying and discriminating speech sounds. Yoo et al. and Tantibundhit et al. were successful in identifying speech transients and, emphasizing them, improving the intelligibility of speech in noise. However, their methods are computationally intensive and unsuitable for real-time applications. This paper presents a method to identify and emphasize speech transients that combines subband decomposition by the wavelet packet transform with variable frame rate (VFR) analysis and unvoiced consonant detection. The VFR analysis is applied to each wavelet packet to define a transitivity function that describes the extent to which the wavelet coefficients of that packet are changing. Unvoiced consonant detection is used to identify unvoiced consonant intervals and the transitivity function is amplified during these intervals. The wavelet coefficients are multiplied by the transitivity function for that packet, amplifying the coefficients localized at times when they are changing and attenuating coefficients at times when they are steady. Inverse transform of the modified wavelet packet coefficients produces a signal corresponding to speech transients similar to the transients identified by Yoo et al. and Tantibundhit et al. A preliminary implementation of the algorithm runs more efficiently.
Analysis of reactivity transient for the DIDO type research reactors using RELAP5
International Nuclear Information System (INIS)
Adorni, M.; Bousbia-Salah, A.; D'Auria, F.; Nabbi, R.
2005-01-01
Recent availability of high performance computers and computational methods together with the continuing increase in operational experience imposes revising some operational constrains and conservative safety margins. The application of Best-Estimate (BE) method constitutes a real necessity in the safety and design analysis and allows getting more realistic simulation of the processes taking place during the steady state operation and transients. In comparison to the conservative approaches, the application of Best-Estimate methods results in the mitigation of the constraining limits in design and operation. This paper presents the results of the application of the RELAP5/Mod3.3 system thermal-hydraulic code to the German FRJ-2 research reactor for a reactivity transient, which has been analyzed in the past using the verified system code CATHENA [1], [2], [3]. The work mainly aims checking the capability of RELAP5 [4] for research reactor transient analysis by the comparison of the results of the two codes and including modeling basis and analytical approaches. According to the existing references RELAP5 applications are concentrated on the transient analysis of nuclear power systems. The considered case consists of a simulation related to a hypothetical fast reactivity transient, which is assumed to be caused by the failure of one shutdown arm. The case has been chosen due to the importance of the models for the precise description of the complex phenomenon of subcooled boiling and two phase flow taking place during the transient. For this purpose, the fuel element assembly was modeled in detail according to design data. The primary circuit was included in the whole model in order to consider the interaction with individual fuel elements with core. In general the results of the two codes are in agreement and comparable during the initial phase of the transient. After reaching the flow regime with fully developed nucleate boiling and two phase flow RELAP5 exhibits
Transient Analysis of Monopile Foundations Partially Embedded in Liquefied Soil
DEFF Research Database (Denmark)
Barari, Amin; Bayat, Mehdi; Meysam, Saadati
2015-01-01
Lagrangian Analysis of Continua (FLAC), which captured the fundamental mechanisms of the monopiles in saturated granular soil. The effects of inertia and the kinematic flow of soil are investigated separately, to highlight the importance of considering the combined effect of these phenomena on the seismic...
Peach Bottom Turbine Trip Simulations with RETRAN Using INER/TPC BWR Transient Analysis Method
International Nuclear Information System (INIS)
Kao Lainsu; Chiang, Show-Chyuan
2005-01-01
The work described in this paper is benchmark calculations of pressurization transient turbine trip tests performed at the Peach Bottom boiling water reactor (BWR). It is part of an overall effort in providing qualification basis for the INER/TPC BWR transient analysis method developed for the Kuosheng and Chinshan plants. The method primarily utilizes an advanced system thermal hydraulics code, RETRAN02/MOD5, for transient safety analyses. Since pressurization transients would result in a strong coupling effect between core neutronic and system thermal hydraulics responses, the INER/TPC method employs the one-dimensional kinetic model in RETRAN with a cross-section data library generated by the Studsvik-CMS code package for the transient calculations. The Peach Bottom Turbine Trip (PBTT) tests, including TT1, TT2, and TT3, have been successfully performed in the plant and assigned as standards commonly for licensing method qualifications for years. It is an essential requirement for licensing purposes to verify integral capabilities and accuracies of the codes and models of the INER/TPC method in simulating such pressurization transients. Specific Peach Bottom plant models, including both neutronics and thermal hydraulics, are developed using modeling approaches and experiences generally adopted in the INER/TPC method. Important model assumptions in RETRAN for the PBTT test simulations are described in this paper. Simulation calculations are performed with best-estimated initial and boundary conditions obtained from plant test measurements. The calculation results presented in this paper demonstrate that the INER/TPC method is capable of calculating accurately the core and system transient behaviors of the tests. Excellent agreement, both in trends and magnitudes between the RETRAN calculation results and the PBTT measurements, shows reliable qualifications of the codes/users/models involved in the method. The RETRAN calculated peak neutron fluxes of the PBTT
Transient Voltage Stability Analysis and Improvement of A Network with different HVDC Systems
DEFF Research Database (Denmark)
Liu, Yan; Chen, Zhe
2011-01-01
This paper presents transient voltage stability analysis of an AC system with multi-infeed HVDC links including a traditional LCC HVDC link and a VSC HVDC link. It is found that the voltage supporting capability of the VSC-HVDC link is significantly influenced by the tie-line distance between the...
Mattos, A Z; Mattos, A A
Many different non-invasive methods have been studied with the purpose of staging liver fibrosis. The objective of this study was verifying if transient elastography is superior to aspartate aminotransferase to platelet ratio index for staging fibrosis in patients with chronic hepatitis C. A systematic review with meta-analysis of studies which evaluated both non-invasive tests and used biopsy as the reference standard was performed. A random-effects model was used, anticipating heterogeneity among studies. Diagnostic odds ratio was the main effect measure, and summary receiver operating characteristic curves were created. A sensitivity analysis was planned, in which the meta-analysis would be repeated excluding each study at a time. Eight studies were included in the meta-analysis. Regarding the prediction of significant fibrosis, transient elastography and aspartate aminotransferase to platelet ratio index had diagnostic odds ratios of 11.70 (95% confidence interval = 7.13-19.21) and 8.56 (95% confidence interval = 4.90-14.94) respectively. Concerning the prediction of cirrhosis, transient elastography and aspartate aminotransferase to platelet ratio index had diagnostic odds ratios of 66.49 (95% confidence interval = 23.71-186.48) and 7.47 (95% confidence interval = 4.88-11.43) respectively. In conclusion, there was no evidence of significant superiority of transient elastography over aspartate aminotransferase to platelet ratio index regarding the prediction of significant fibrosis, but the former proved to be better than the latter concerning prediction of cirrhosis.
International Nuclear Information System (INIS)
Guerreiro, J.N.C.; Loula, A.F.D.
1988-12-01
The mixed Petrov-Galerkin finite element formulation is applied to transiente and steady state creep problems. Numerical analysis has shown additional stability of this method compared to classical Galerkin formulations. The accuracy of the new formulation is confirmed in some representative examples of two dimensional and axisymmetric problems. (author) [pt
The PARET code and the analysis of the SPERT I transients
Energy Technology Data Exchange (ETDEWEB)
Woodruff, William L [Argonne National Laboratory, Argonne (United States)
1983-09-01
The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients.
Availability analysis of a turbocharged diesel engine operating under transient load conditions
International Nuclear Information System (INIS)
Rakopoulos, C.D.; Giakoumis, E.G.
2004-01-01
A computer analysis is developed for studying the energy and availability performance of a turbocharged diesel engine, operating under transient load conditions. The model incorporates many novel features for the simulation of transient operation, such as detailed analysis of mechanical friction, separate consideration for the processes of each cylinder during a cycle ('multi-cylinder' model) and mathematical modeling of the fuel pump. This model has been validated against experimental data taken from a turbocharged diesel engine, located at the authors' laboratory and operated under transient conditions. The availability terms for the diesel engine and its subsystems are analyzed, i.e. cylinder for both the open and closed parts of the cycle, inlet and exhaust manifolds, turbocharger and aftercooler. The present analysis reveals, via multiple diagrams, how the availability properties of the diesel engine and its subsystems develop during the evolution of the engine cycles, assessing the importance of each property. In particular the irreversibilities term, which is absent from any analysis based solely on the first-law of thermodynamics, is given in detail as regards transient response as well as the rate and cumulative terms during a cycle, revealing the magnitude of contribution of all the subsystems to the total availability destruction
Theory of lifetime measurements with the scanning electron microscope: transient analysis
Kuiken, H.K.
1976-01-01
A transient analysis of an SEM experiment is given with the purpose of determining directly the lifetime of minority carriers in a semiconductor material. The injection takes place below a surface normal to the junction and expressions are derived for the current-decay which ensues when the electron
The PARET code and the analysis of the SPERT I transients
International Nuclear Information System (INIS)
Woodruff, William L.
1983-01-01
The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients
Transient analysis of blowdown thrust force under PWR LOCA
International Nuclear Information System (INIS)
Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni
1982-10-01
The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces obtained by Navier-Stokes momentum equation about a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break. (5) The blowdown thrust force in the analysis greatly depends on the selection of the exit pressure. (author)
RELAP5/MOD2 Overview and Developmental. Assessment Results from TMl-1 Plant Transient Analysis
International Nuclear Information System (INIS)
Lin, J. C.; Tsai, C. C.; Ransom, V. H.; Johnsen, G. W.
2013-01-01
RELAP5/MOD2 is a new version of the RELAP5 thermal-hydraulic computer code containing improved modeling features that provide a generic capability for pressurized water reactor transient simulation. The objective of this paper is to provide code users with an overview of the code and to report developmental assessment results obtained from a Three Mile Island Unit One plant transient analysis. The assessment shows that the injection of highly sub-cooled water into a high-pressure primary coolant system does not cause unphysical results or pose a problem for RELAP5/MOD2. (author)
Analysis of transient heat conduction in a PWR fuel rod by an improved lumped parameter approach
Energy Technology Data Exchange (ETDEWEB)
Dourado, Eneida Regina G. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Cotta, Renato M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Mecanica; Jian, Su, E-mail: eneidadourado@gmail.com, E-mail: sujian@nuclear.ufrj.br, E-mail: cotta@mecanica.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear
2017-07-01
This paper aims to analyze transient heat conduction in a nuclear fuel rod by an improved lumped parameter approach. One-dimensional transient heat conduction is considered, with the circumferential symmetry assumed and the axial conduction neglected. The thermal conductivity and specific heat in the fuel pellet are considered temperature dependent, while the thermophysical properties of the cladding are considered constant. Hermite approximation for integration is used to obtain the average temperature and heat flux in the radial direction. Significant improvement over the classical lumped parameter formulation has been achieved. The proposed model can be also used in dynamic analysis of PWR and nuclear power plant simulators. (author)
Analysis of transient heat conduction in a PWR fuel rod by an improved lumped parameter approach
International Nuclear Information System (INIS)
Dourado, Eneida Regina G.; Cotta, Renato M.; Jian, Su
2017-01-01
This paper aims to analyze transient heat conduction in a nuclear fuel rod by an improved lumped parameter approach. One-dimensional transient heat conduction is considered, with the circumferential symmetry assumed and the axial conduction neglected. The thermal conductivity and specific heat in the fuel pellet are considered temperature dependent, while the thermophysical properties of the cladding are considered constant. Hermite approximation for integration is used to obtain the average temperature and heat flux in the radial direction. Significant improvement over the classical lumped parameter formulation has been achieved. The proposed model can be also used in dynamic analysis of PWR and nuclear power plant simulators. (author)
York, B. J.; Sinha, N.; Dash, S. M.; Hosangadi, A.; Kenzakowski, D. C.; Lee, R. A.
1992-07-01
The analysis of steady and transient aerodynamic/propulsive/plume flowfield interactions utilizing several state-of-the-art computer codes (PARCH, CRAFT, and SCHAFT) is discussed. These codes have been extended to include advanced turbulence models, generalized thermochemistry, and multiphase nonequilibrium capabilities. Several specialized versions of these codes have been developed for specific applications. This paper presents a brief overview of these codes followed by selected cases demonstrating steady and transient analyses of conventional as well as advanced missile systems. Areas requiring upgrades include turbulence modeling in a highly compressible environment and the treatment of particulates in general. Recent progress in these areas are highlighted.
Inverse Transient Analysis for Classification of Wall Thickness Variations in Pipelines
Directory of Open Access Journals (Sweden)
Jeffrey Tuck
2013-12-01
Full Text Available Analysis of transient fluid pressure signals has been investigated as an alternative method of fault detection in pipeline systems and has shown promise in both laboratory and field trials. The advantage of the method is that it can potentially provide a fast and cost effective means of locating faults such as leaks, blockages and pipeline wall degradation within a pipeline while the system remains fully operational. The only requirement is that high speed pressure sensors are placed in contact with the fluid. Further development of the method requires detailed numerical models and enhanced understanding of transient flow within a pipeline where variations in pipeline condition and geometry occur. One such variation commonly encountered is the degradation or thinning of pipe walls, which can increase the susceptible of a pipeline to leak development. This paper aims to improve transient-based fault detection methods by investigating how changes in pipe wall thickness will affect the transient behaviour of a system; this is done through the analysis of laboratory experiments. The laboratory experiments are carried out on a stainless steel pipeline of constant outside diameter, into which a pipe section of variable wall thickness is inserted. In order to detect the location and severity of these changes in wall conditions within the laboratory system an inverse transient analysis procedure is employed which considers independent variations in wavespeed and diameter. Inverse transient analyses are carried out using a genetic algorithm optimisation routine to match the response from a one-dimensional method of characteristics transient model to the experimental time domain pressure responses. The accuracy of the detection technique is evaluated and benefits associated with various simplifying assumptions and simulation run times are investigated. It is found that for the case investigated, changes in the wavespeed and nominal diameter of the
Inverse Transient Analysis for Classification of Wall Thickness Variations in Pipelines
Tuck, Jeffrey; Lee, Pedro
2013-01-01
Analysis of transient fluid pressure signals has been investigated as an alternative method of fault detection in pipeline systems and has shown promise in both laboratory and field trials. The advantage of the method is that it can potentially provide a fast and cost effective means of locating faults such as leaks, blockages and pipeline wall degradation within a pipeline while the system remains fully operational. The only requirement is that high speed pressure sensors are placed in contact with the fluid. Further development of the method requires detailed numerical models and enhanced understanding of transient flow within a pipeline where variations in pipeline condition and geometry occur. One such variation commonly encountered is the degradation or thinning of pipe walls, which can increase the susceptible of a pipeline to leak development. This paper aims to improve transient-based fault detection methods by investigating how changes in pipe wall thickness will affect the transient behaviour of a system; this is done through the analysis of laboratory experiments. The laboratory experiments are carried out on a stainless steel pipeline of constant outside diameter, into which a pipe section of variable wall thickness is inserted. In order to detect the location and severity of these changes in wall conditions within the laboratory system an inverse transient analysis procedure is employed which considers independent variations in wavespeed and diameter. Inverse transient analyses are carried out using a genetic algorithm optimisation routine to match the response from a one-dimensional method of characteristics transient model to the experimental time domain pressure responses. The accuracy of the detection technique is evaluated and benefits associated with various simplifying assumptions and simulation run times are investigated. It is found that for the case investigated, changes in the wavespeed and nominal diameter of the pipeline are both important
Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis
Hoogenboom, J. Eduard; Sjenitzer, Bart L.
2014-06-01
To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.
Fuel element thermo-mechanical analysis during transient events using the FMS and FETMA codes
International Nuclear Information System (INIS)
Hernandez Lopez Hector; Hernandez Martinez Jose Luis; Ortiz Villafuerte Javier
2005-01-01
In the Instituto Nacional de Investigaciones Nucleares of Mexico, the Fuel Management System (FMS) software package has been used for long time to simulate the operation of a BWR nuclear power plant in steady state, as well as in transient events. To evaluate the fuel element thermo-mechanical performance during transient events, an interface between the FMS codes and our own Fuel Element Thermo Mechanical Analysis (FETMA) code is currently being developed and implemented. In this work, the results of the thermo-mechanical behavior of fuel rods in the hot channel during the simulation of transient events of a BWR nuclear power plant are shown. The transient events considered for this work are a load rejection and a feedwater control failure, which among the most important events that can occur in a BWR. The results showed that conditions leading to fuel rod failure at no time appeared for both events. Also, it is shown that a transient due load rejection is more demanding on terms of safety that the failure of a controller of the feedwater. (authors)
Directory of Open Access Journals (Sweden)
Changjun Li
2017-12-01
Full Text Available In the future fusion devices, ELMs-induced transient heat flux may lead to the surface cracking of tungsten (W based plasma-facing materials (PFMs. In theory, the cracking is related to the material fracture toughness and the thermal stress-strain caused by transient heat flux. In this paper, a finite element model was successfully built to realize a theoretical semi infinite space. The temperature and stress-strain distribution as well as evolution of W during a single heating-cooling cycle of transient heat flux were simulated and analyzed. It showed that the generation of plastic deformation during the brittle temperature range between room temperature and DBTT (ductile to brittle transition temperature, ∼400 °C caused the cracking of W during the cooling phase. The cracking threshold for W under transient heat flux was successfully obtained by finite element analysis, to some extent, in consistent with the similar experimental results. Both the heat flux factors (FHF = P·t0.5 and the maximum surface temperatures at cracking thresholds were almost invariant for the transient heat fluxes with different pulse widths and temporal distributions. This method not only identified the theoretical conclusion but also obtained the detail values for W with actual temperature-dependent properties.
Analysis and computer simulation for transient flow in complex system of liquid piping
International Nuclear Information System (INIS)
Mitry, A.M.
1985-01-01
This paper is concerned with unsteady state analysis and development of a digital computer program, FLUTRAN, that performs a simulation of transient flow behavior in a complex system of liquid piping. The program calculates pressure and flow transients in the liquid filled piping system. The analytical model is based on the method of characteristics solution to the fluid hammer continuity and momentum equations. The equations are subject to wide variety of boundary conditions to take into account the effect of hydraulic devices. Water column separation is treated as a boundary condition with known head. Experimental tests are presented that exhibit transients induced by pump failure and valve closure in the McGuire Nuclear Station Low Level Intake Cooling Water System. Numerical simulation is conducted to compare theory with test data. Analytical and test data are shown to be in good agreement and provide validation of the model
Transient analysis of a U-tube natural circulation steam generator
Energy Technology Data Exchange (ETDEWEB)
Gaikwad, A J; Kumar, Rajesh; Bhadra, Anu; Chakraborty, G; Venkat Raj, V [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)
1994-06-01
A computer code has been developed, for transient thermal-hydraulic analysis of proposed 500 MWe PHWR steam generator. The transient behaviour of a nuclear power plant is very much dependent on the steam generator performance, as it provides a thermal linkage between the primary and secondary systems. Study of dynamics of steam generator is essential for over all power plant dynamics as well as design of control systems for steam generator. A mathematical model has been developed for the simulation of thermal-hydraulic phenomena in a U tube natural circulation steam generator. Fluid model is based on one dimensional, nonlinear, single fluid conservation equations of mass, momentum, energy and equation of state. This model includes coupled two phase flow heat transfer and natural circulation. The model accounts for both compressibility and thermal expansion effects. The process simulation and results obtained for transients such as step change in load and total loss of feed water are presented. (author). 5 refs., 7 figs.
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.
Transient analysis of ABWR reactor using a best estimate code
International Nuclear Information System (INIS)
Mizokami, S.; Kitamura, H.; Mototani, A.; Ono, H.
2004-01-01
Since the recirculation pumps are mounted internally within the ABWR, core flow will decrease rapidly in the event of a loss of their driving force. A rapid reduction in core flow may cause the onset of boiling transition (BT). Therefore, in order to prevent the onset of BT, a motor-generator (MG) set is added to the power supply system of the reactor internal pump (RIP). Recent studies, however, have shown that dryout within a fuel assembly over a short time period will result in only a small rise in fuel cladding temperature and thus does not pose a threat to fuel integrity. In response to this finding, the standards committee of the Atomic Energy Society of Japan (AESJ) has proposed a post-BT standard which incorporates a cladding temperature criterion. If it is assumed that the MG-set is not added to the RIP power supply system, the result of the safety analysis shows the onset of BT with a subsequent rise in fuel cladding temperature. Although BT occurs under the conservative assumptions of this safety analysis, a possibility exists that BT will not occur under actual operating conditions. The best estimate code TRACG was used to show that BT does not occur and that fuel integrity can be sufficiently maintained under actual conditions. (author)
Neutronics methods for transient and safety analysis of fast reactors
Energy Technology Data Exchange (ETDEWEB)
Marchetti, Marco
2017-07-01
Modeling the evolution of possible or postulated accidents in nuclear reactors is fundamental in designing safe systems. For the next generation of reactors, in particular fast reactors, fuel movement during an accident can, in principle, drive an energetic event. Such is the issue of recriticality. The thermal energy produced during these events will, possibly, be converted into mechanical energy by some mechanisms. For example, the nuclear heat deposited in the fuel could cause fuel vaporization and its subsequent expansion. This movement would accelerate the surrounding sodium: part of the initial energy in the fuel is thus converted into sodium kinetic energy. This mechanical energy will finally be absorbed, in some way or another, by the reactor vessel. Providing an accurate estimate for the maximum mechanical work that any accidental sequence can do onto the reactor vessel is an essential step in designing a reactor containment that would withstand any load generated by any accident. That would assure accident containment, without consequences for the general public. Fast reactor accident modeling is a complicated task. The outcome of an accident is determined by different physical phenomena, all acting at almost the same time. Safety analysts must track all these different phenomena. Multi-physics codes have been developed for this task. They must contain accurate models for fluid-dynamics, neutronics, and structures. This work has to do with neutronics modeling of such accidents. Past and recent analyses have been limited to the approximate description of the neutronic field, for example by using a rough description of the energy and/or of the angular dependence of the neutron flux. In this work, different neutronic solvers are selected and coupled into a general multi-physics code for fast reactor accident analysis. Performances of each of them is then assessed. Some emphasis has been put also in assessing the speed of these solvers for determining the
Noda, Taku
Nowadays, there is quite high demand for electromagnetic transient (EMT) analysis programs and real-time simulators for power systems. In addition to the conventional demand such as overvoltage, over-current and oscillation simulations, the new demand that includes simulations of power-electronics circuits and power quality is increasing. With this background, development groups of EMT programs and real-time simulators have made progress in terms of computational performance and user experience. In Japan, Central Research Institute of Electric Power Industry has newly developed an EMT analysis program called XTAP (eXpandable Transient Analysis Program). This article overviews these international and domestic development trends of EMT analysis programs and real-time simulators.
Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors
International Nuclear Information System (INIS)
Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan
1999-02-01
Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs
Analysis of core uncovery time in Kuosheng station blackout transient with MELCOR
International Nuclear Information System (INIS)
Wang, S.J.; Chien, C.S.
1996-01-01
The MELCOR code, developed by the Sandia National Laboratories, is capable of simulating severe accident phenomena of nuclear power plants. Core uncovery time is an important parameter in the probabilistic risk assessment. However, many MELCOR users do not generate the initial conditions in a station blackout (SBO) transient analysis. Thus, achieving reliable core uncovery time is difficult. The core uncovery time for the Kuosheng nuclear power plant during an SBO transient is analyzed. First, full-power steady-state conditions are generated with the application of a developed self-initialization algorithm. Then the response of the SBO transient up to core uncovery is simulated. The effects of key parameters including the initialization process and the reactor feed pump (RFP) coastdown time on the core uncovery time are analyzed. The initialization process is the most important parameter that affects the core uncovery time. Because SBO transient analysis, the correct initial conditions must be generated to achieve a reliable core uncovery time. The core uncovery time is also sensitive to the RFP coastdown time. A correct time constant is required
Soft error rate analysis methodology of multi-Pulse-single-event transients
International Nuclear Information System (INIS)
Zhou Bin; Huo Mingxue; Xiao Liyi
2012-01-01
As transistor feature size scales down, soft errors in combinational logic because of high-energy particle radiation is gaining more and more concerns. In this paper, a combinational logic soft error analysis methodology considering multi-pulse-single-event transients (MPSETs) and re-convergence with multi transient pulses is proposed. In the proposed approach, the voltage pulse produced at the standard cell output is approximated by a triangle waveform, and characterized by three parameters: pulse width, the transition time of the first edge, and the transition time of the second edge. As for the pulse with the amplitude being smaller than the supply voltage, the edge extension technique is proposed. Moreover, an efficient electrical masking model comprehensively considering transition time, delay, width and amplitude is proposed, and an approach using the transition times of two edges and pulse width to compute the amplitude of pulse is proposed. Finally, our proposed firstly-independently-propagating-secondly-mutually-interacting (FIP-SMI) is used to deal with more practical re-convergence gate with multi transient pulses. As for MPSETs, a random generation model of MPSETs is exploratively proposed. Compared to the estimates obtained using circuit level simulations by HSpice, our proposed soft error rate analysis algorithm has 10% errors in SER estimation with speed up of 300 when the single-pulse-single-event transient (SPSET) is considered. We have also demonstrated the runtime and SER decrease with the increment of P0 using designs from the ISCAS-85 benchmarks. (authors)
Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS
Energy Technology Data Exchange (ETDEWEB)
Barani, T.; Bruschi, E.; Pizzocri, D. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, I-20156 Milano (Italy); Pastore, G. [Fuel Modeling and Simulation Department, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Van Uffelen, P. [European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security, P.O. Box 2340, 76125 Karlsruhe (Germany); Williamson, R.L. [Fuel Modeling and Simulation Department, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Luzzi, L., E-mail: Lelio.Luzzi@polimi.it [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, I-20156 Milano (Italy)
2017-04-01
The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the sensitivity of results to the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release
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
Analysis of LOFT pressurizer spray and surge nozzles to include a 4500F step transient
International Nuclear Information System (INIS)
Nitzel, M.E.
1978-01-01
This report presents the analysis of the LOFT pressurizer spray and surge nozzles to include a 450 0 F step thermal transient. Previous analysis performed under subcontract by Basic Technology Incorporated was utilized where applicable. The SAASIII finite element computer program was used to determine stress distributions in the nozzles due to the step transient. Computer results were then incorporated in the necessary additional calculations to ascertain that stress limitations were not exceeded. The results of the analysis indicate that both the spray and surge nozzles will be within stress allowables prescribed by subsubarticle NB-3220 of the 1974 edition of the ASME Boiler and Pressure Vessel Code when subjected to currently known design, normal operating, upset, emergency, and faulted condition loads
International Nuclear Information System (INIS)
Rebollo, L.
1993-01-01
Union Fenosa, a utility company in Spain, has performed research on pressurized water reactor (PWR) safety with respect to the development of a best-estimate methodology for the analysis of anticipated transients without scram (ATWS), i.e., those anticipated transients for which failure of the reactor protection system is postulated. A scientific and technical approach is adopted with respect to the ATWS phenomenon as it affects a PWR, specifically the Zorita nuclear power plant, a single-loop Westinghouse-designed PWR in Spain. In this respect, an ATWS sequence analysis methodology based on published codes that is generically applicable to any PWR is proposed, which covers all the anticipated phenomena and defines the applicable acceptance criteria. The areas contemplated are cell neutron analysis, core thermal hydraulics, and plant dynamics, which are developed, qualified, and plant dynamics, which are developed, qualified, and validated by comparison with reference calculations and measurements obtained from integral or separate-effects tests
MINET: transient analysis of fluid-flow and heat-transfer networks
International Nuclear Information System (INIS)
Van Tuyle, G.J.; Guppy, J.G.; Nepsee, T.C.
1983-01-01
MINET, a computer code developed for the steady-state and transient analysis of fluid-flow and heat-transfer networks, is described. The code is based on a momentum integral network method, which offers significant computational advantages in the analysis of large systems, such as the balance of plant in a power-generating facility. An application is discussed in which MINET is coupled to the Super System Code (SSC), an advanced generic code for the transient analysis of loop- or pool-type LMFBR systems. In this application, the ability of the Clinch River Breeder Reactor Plant to operate in a natural circulation mode following an assumed loss of all electric power, was assessed. Results from the MINET portion of the calculations are compared against those generated independently by the Clinch River Project, using the DEMO code
Present status of numerical analysis on transient two-phase flow
International Nuclear Information System (INIS)
Akimoto, Masayuki; Hirano, Masashi; Nariai, Hideki.
1987-01-01
The Special Committee for Numerical Analysis of Thermal Flow has recently been established under the Japan Atomic Energy Association. Here, some methods currently used for numerical analysis of transient two-phase flow are described citing some information given in the first report of the above-mentioned committee. Many analytical models for transient two-phase flow have been proposed, each of which is designed to describe a flow by using differential equations associated with conservation of mass, momentum and energy in a continuous two-phase flow system together with constructive equations that represent transportation of mass, momentum and energy though a gas-liquid interface or between a liquid flow and the channel wall. The author has developed an analysis code, called MINCS, that serves for systematic examination of conservation equation and constructive equations for two-phase flow models. A one-dimensional, non-equilibrium two-liquid flow model that is used as the basic model for the code is described. Actual procedures for numerical analysis is shown and some problems concerning transient two-phase analysis are described. (Nogami, K.)
Analysis of steady state and transient two-phase flows in downwardly inclined lines
International Nuclear Information System (INIS)
Crawford, T.J.
1983-01-01
A study of steady-state and transient two-phase flows in downwardly inclined lines is described. Steady-state flow patterns maps are presented using Freon-113 as the working fluid to provide new high density vapors. These flow maps with high density vapor serve to significantly extend the investigations of steady-state downward two-phase flow patterns. Physical models developed which successfully predicted the onset or location of various flow pattern transitions. A new simplified criterion that would be useful to designers and experimenters is offered for the onset of dispersed flow. A new empirical holdup correlation and a new bubble diameter/flow rate correlation are also proposed. Flow transients in vertical downward lines were studied to investigate the possible formation of intermediate or spurious flow patterns that would not be seen at steady-state conditions. Void fraction behavior during the transients was modeled by using the dynamic slip equation from the transient analysis code RETRAN. Physical models of interfacial area were developed and compared with models and data from literature. There was satisfactory agreement between the models of the present study and the literature models and data. The concentration parameter of the drift flux model was evaluated for vertical downward flow. These new values of the flow dependent parameter were different from those previously proposed in the literature for use in upward flows, and made the drift flux model suitable for use in upward or downward flow lines
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
Transient analysis for alternating over-current characteristics of HTSC power transmission cable
Lim, S. H.; Hwang, S. D.
2006-10-01
In this paper, the transient analysis for the alternating over-current distribution in case that the over-current was applied for a high-TC superconducting (HTSC) power transmission cable was performed. The transient analysis for the alternating over-current characteristics of HTSC power transmission cable with multi-layer is required to estimate the redistribution of the over-current between its conducting layers and to protect the cable system from the over-current in case that the quench in one or two layers of the HTSC power cable happens. For its transient analysis, the resistance generation of the conducting layers for the alternating over-current was reflected on its equivalent circuit, based on the resistance equation obtained by applying discrete Fourier transform (DFT) for the voltage and the current waveforms of the HTSC tape, which comprises each layer of the HTSC power transmission cable. It was confirmed through the numerical analysis on its equivalent circuit that after the current redistribution from the outermost layer into the inner layers first happened, the fast current redistribution between the inner layers developed as the amplitude of the alternating over-current increased.
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
General purpose dynamic Monte Carlo with continuous energy for transient analysis
Energy Technology Data Exchange (ETDEWEB)
Sjenitzer, B. L.; Hoogenboom, J. E. [Delft Univ. of Technology, Dept. of Radiation, Radionuclide and Reactors, Mekelweg 15, 2629JB Delft (Netherlands)
2012-07-01
For safety assessments transient analysis is an important tool. It can predict maximum temperatures during regular reactor operation or during an accident scenario. Despite the fact that this kind of analysis is very important, the state of the art still uses rather crude methods, like diffusion theory and point-kinetics. For reference calculations it is preferable to use the Monte Carlo method. In this paper the dynamic Monte Carlo method is implemented in the general purpose Monte Carlo code Tripoli4. Also, the method is extended for use with continuous energy. The first results of Dynamic Tripoli demonstrate that this kind of calculation is indeed accurate and the results are achieved in a reasonable amount of time. With the method implemented in Tripoli it is now possible to do an exact transient calculation in arbitrary geometry. (authors)
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
Transient thermal-hydraulic characteristics analysis software for PWR nuclear power systems
International Nuclear Information System (INIS)
Wu Yingwei; Zhuang Chengjun; Su Guanghui; Qiu Suizheng
2010-01-01
A point reactor neutron kinetics model, a two-phase drift-flow U-tube steam generator model, an advanced non-equilibrium three regions pressurizer model, and a passive emergency core decay heat-removed system model are adopted in the paper to develop the computerized analysis code for PWR transient thermal-hydraulic characteristics, by Compaq Visual Fortran 6.0 language. Visual input, real-time processing and dynamic visualization output are achieved by Microsoft Visual Studio. NET language. The reliability verification of the soft has been conducted by RELAP 5, and the verification results show that the software is with high calculation precision, high calculation speed, modern interface, luxuriant functions and strong operability. The software was applied to calculate the transient accident conditions for QSNP, and the analysis results are significant to the practical engineering applications. (authors)
Transient pattern analysis for fault detection and diagnosis of HVAC systems
International Nuclear Information System (INIS)
Cho, Sung-Hwan; Yang, Hoon-Cheol; Zaheer-uddin, M.; Ahn, Byung-Cheon
2005-01-01
Modern building HVAC systems are complex and consist of a large number of interconnected sub-systems and components. In the event of a fault, it becomes very difficult for the operator to locate and isolate the faulty component in such large systems using conventional fault detection methods. In this study, transient pattern analysis is explored as a tool for fault detection and diagnosis of an HVAC system. Several tests involving different fault replications were conducted in an environmental chamber test facility. The results show that the evolution of fault residuals forms clear and distinct patterns that can be used to isolate faults. It was found that the time needed to reach steady state for a typical building HVAC system is at least 50-60 min. This means incorrect diagnosis of faults can happen during online monitoring if the transient pattern responses are not considered in the fault detection and diagnosis analysis
International Nuclear Information System (INIS)
Buckner, M.R.; Hostetler, D.E.; Anderson, M.M.; Dodds, H.L.
1977-01-01
GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering feedback. The general application of GRASS to the Savannah River reactors is discussed, and results are presented for the analyses of severla reactor transient calculations
Development of a system code for transient analysis in a HTGR
International Nuclear Information System (INIS)
Lee, Tae Beom
2004-02-01
A GAMMA (GAs Multi-component Multi-dimensional Analysis) code is developed for transient analysis and air ingress analysis in High Temperature Gas-cooled Reactors (HTGR). The PBMR of ESKOM is selected as a reference plant for the High Temperature Gas-cooled Reactor here, which uses a direct helium cycle and pebble fuel. Physical models included in GAMMA are the pebble conduction model, radiation heat transfer model, point kinetics model, decay heat model, and component models for break flow, valve, pump, cooler, power conversion unit model. The temperature distribution and the flow distribution of the PBMR are calculated for initial and accident core in the present study. In the accident analysis, typical design basis accident (DBA), including the load transient accident and depressurization accident into the system are selected and analyzed in detail. The predictions by GAMMA for PBMR at 100% power are compared with those by VSOP and PBR S IM. It turns out that the temperature in the upper region in the third channel predicted by GAMMA is about 62 .deg. C at maximum higher than that by VSOP, but is pretty close to that by PBR S IM. The center temperature of the fuel shows that that predicted by considering swelling effect is higher than that without swelling effect by about 10 .deg. C. The net efficiency of direct system is higher than that of indirect system due to an effect of the circulator power. The transient capability of GAMMA is validated through analytical solution and PBR S IM analyzing the depressurization (Loss Of Coolant Accident, LOCA) and load transient accident. After the LOCA the system pressure decreases dramatically from 8MPa to 0.4MPa within 2 sec. After the PI (Proportional-plus-Integral) controller senses that the power shaft is over the set-point of 3,600 rpm, the bypass valve makes shaft speed back to the set-point
The limiting events transient analysis by RETRAN02 and VIPRE01 for an ABWR
International Nuclear Information System (INIS)
Tsai Chiungwen; Shih Chunkuan; Wang Jongrong; Lin Haotzu; Jin Jiunan; Cheng Suchin
2009-01-01
This paper describes the transient analysis of generator load rejection (LR) and One Turbine Control Valve Closure (OTCVC) events for Lungmen nuclear power plant (LMNPP). According to the Critical Power Ratio (CPR) criterion, the Preliminary Safety Analysis Report (PSAR) concluded that LR and OTCVC are the first and second limiting events respectively. In addition, the fuel type is changed from GE12 to GE14 now. It's necessary to re-analyze these two events for safety consideration. In this study, to quantify the impact to reactor, the difference of initial critical power ratio (ICPR) and minimum critical power ratio (MCPR), ie. ΔCPR is calculated. The ΔCPRs of the LR and OTCVC events are calculated with the combination of RETRAN02 and VIPRE01 codes. In RETRAN02 calculation, a thermal-hydraulic model was prepared for the transient analysis. The data including upper plenum pressure, core inlet flow, normalized power, and axial power shapes during transient are furthermore submitted into VIPRE01 for ΔCPR calculation. In VIPRE01 calculation, there was a hot channel model built to simulate the hottest fuel bundle. Based on the thermal-hydraulic data from RETRAN02, the ΔCPRs are calculated by VIPRE01 hot channel model. Additionally, the different TCV control modes are considered to study the influence of different TCV closure curves on transient behavior. Meanwhile, sensitivity studies including different initial system pressure and different initial power/flow conditions are also considered. Based on this analysis, the maximum ΔCPRs for LR and OTCVC are 0.162 and 0.191 respectively. According CPR criterion, the result shows that the impact caused by OTCVC event leads to be larger than LR event. (author)
Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis
International Nuclear Information System (INIS)
Hoogenboom, J.E.
2013-01-01
To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branch-less collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires the coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail. (authors)
Analysis of transients for NPP with VVER-440 using the code SiTAP
International Nuclear Information System (INIS)
Kalinenko, V.
1994-06-01
The report contains analysis of transients ''Loop connection'' and ''Steam generator tube rupture'' for nuclear power plants (NPP) with VVER-440. To obtain more detailed information about NPP's dynamic characteristics, various variants of initial and boundary conditions are considerd. Calculation of these transients was performed using the SiTAP code developed at the Nuclear Safety Institute of the Russian Research Centre ''Kurchatov Institute''. SiTAP code is a multifunctional computer tool for fast analysis of transient and accidental processes of VVER type reactors for engineers working in the field of NPP dynamics. SiTAP can be used form comparative analysis of several variants of accident scenarios to find out the conditions leading to most serious consequences from a safety point of view. In such cases, additional analyses using best-estimate codes should be carried out. The results of SiTAP for a faulty loop connection leading to a boron dilution accident are intended to be used as boundary conditions for a more detailed anlaysis with the aid of the three-dimensional reactor core model DYN3D, developed in the Research Centre Rossendorf for the simulation of reactivity initiated accidents. (orig.)
Steady-State and Transient Analysis for Design Validation of SMART-ITL Secondary System
Energy Technology Data Exchange (ETDEWEB)
Yun, Eunkoo; Bae, Hwang; Ryu, Sung Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2016-10-15
SMART can prevent large-break loss of coolant accident (LBLOCA) inherently. SMART-ITL is an experimental simulation facility designed to perform integral effect tests for the SMART plant. In terms of the secondary system of SMART-ITL, the design has been simplified from that of reference plant by replacing several components, such as expansion device and condenser, with an appropriate device to be functional as the alternatives. In this paper, in order to understand the operational characteristics as well as design concept, the secondary system of SMRAT-ITL is analyzed in steady-state and transient aspects, and the results are compared with relevant experimental results. This study focuses on the understanding of thermal-hydraulic behavior of SMART-ITL secondary system, which is simplified from that of reference plant. To identify the behaviors of the secondary system, the steady-state and transient analysis were conducted based on experimental results. In steady-state analysis, the results clearly showed that the system pressure is related to the temperature of condensation tank which varies depending on mixture enthalpy. In transient analysis, the dynamic behavior during heat-up process has been investigated. The results reveal that we can reasonably assume the fluid filled in TK-CD-01 be in a saturated condition. The results showed that the design of SMART-ITL secondary system is appropriate, and the system is being properly operated to match the design intent.
An Efficient Topology-Based Algorithm for Transient Analysis of Power Grid
Yang, Lan
2015-08-10
In the design flow of integrated circuits, chip-level verification is an important step that sanity checks the performance is as expected. Power grid verification is one of the most expensive and time-consuming steps of chip-level verification, due to its extremely large size. Efficient power grid analysis technology is highly demanded as it saves computing resources and enables faster iteration. In this paper, a topology-base power grid transient analysis algorithm is proposed. Nodal analysis is adopted to analyze the topology which is mathematically equivalent to iteratively solving a positive semi-definite linear equation. The convergence of the method is proved.
Current interruption transients calculation
Peelo, David F
2014-01-01
Provides an original, detailed and practical description of current interruption transients, origins, and the circuits involved, and how they can be calculated Current Interruption Transients Calculationis a comprehensive resource for the understanding, calculation and analysis of the transient recovery voltages (TRVs) and related re-ignition or re-striking transients associated with fault current interruption and the switching of inductive and capacitive load currents in circuits. This book provides an original, detailed and practical description of current interruption transients, origins,
Nisa Khan, M
2017-09-20
We present measurement and analysis of color stability over time for two categories of white LED lamps based on their thermal management scheme, which also affects their transient lumen depreciation. We previously reported that lumen depreciation in LED lamps can be minimized by properly designing the heat sink configuration that allows lamps to reach a thermal equilibrium condition quickly. Although it is well known that lumen depreciation degrades color stability of white light since color coordinates vary with total lumen power by definition, quantification and characterization of color shifts based on thermal transient behavior have not been previously reported in literature for LED lamps. Here we provide experimental data and analysis of transient color shifts for two categories of household LED lamps (from a total of six lamps in two categories) and demonstrate that reaching thermal equilibrium more quickly provides better stability for color rendering, color temperature, and less deviation of color coordinates from the Planckian blackbody locus line, which are all very important characterization parameters of color for white light. We report for the first time that a lamp's color degradation from the turn-on time primarily depends on thermal transient behavior of the semiconductor LED chip, which experiences a wavelength shift as well as a decrease in its dominant wavelength peak value with time, which in turn degrades the phosphor conversion. For the first time, we also provide a comprehensive quantitative analysis that differentiates color degradation due to the heat rise in GaN/GaInN LED chips and subsequently the boards these chips are mounted on-from that caused by phosphor heating in a white LED module. Finally, we briefly discuss why there are some inevitable trade-offs between omnidirectionality and color and luminous output stability in current household LED lamps and what will help eliminate these trade-offs in future lamp designs.
Directory of Open Access Journals (Sweden)
Tao Li
Full Text Available Background and purpose: Transient elastography (TE has been shown to be a valuable tool for the prediction of large esophageal varices. However, the conclusions have not been always consistent throughout the different studies. Therefore, we performed a further meta-analysis in order to evaluate the diagnostic accuracy of transient elastography for the prediction of large esophageal varices. Methods: We performed a systematic literature search in PubMed, EMBASE, Web of Science, and CENTRAL in The Cochrane Library without time restriction. The strategy we used was "(fibroscan OR transient elastography OR stiffness AND esophageal varices". Accuracy measures such as pooled sensitivity, specificity, among others, were calculated using Meta-DiSc statistical software. Results: Twenty studies (2,994 patients were included in our meta-analysis. The values of pooled sensitivity, specificity, positive and negative likelihood ratios and diagnostic odds ratio were as follows: 0.81 (95% CI, 0.79-0.84, 0.71 (95% CI, 0.69-0.73, 2.63 (95% CI, 2.15-3.23, 0.27 (95% CI, 0.22-0.34 and 10.30 (95% CI, 7.33-14.47. The area under the receiver operating characteristics curve was 0.83. The Spearman correlation coefficient was 0.246 with a p-value of 0.296, indicating the absence of any significant threshold effects. In our subgroup analysis, the heterogeneity could be partially explained by the geographical origin of the study or etiology; or it could be partially explained blindingly, through the appropriate interval and cut-off value of the liver stiffness (LS. Conclusions: Transient elastography could be used as a valuable non-invasive screening tool for the prediction of large esophageal varices. However, since LS cut-off values vary throughout the different studies and significant heterogeneity also exists among them, we need more reasonable approaches or flow diagram in order to improve the operability of this technology.
International Nuclear Information System (INIS)
Barhen, J.; Bjerke, M.A.; Cacuci, D.G.; Mullins, C.B.; Wagschal, G.G.
1982-01-01
An advanced methodology for performing systematic uncertainty analysis of time-dependent nonlinear systems is presented. This methodology includes a capability for reducing uncertainties in system parameters and responses by using Bayesian inference techniques to consistently combine prior knowledge with additional experimental information. The determination of best estimates for the system parameters, for the responses, and for their respective covariances is treated as a time-dependent constrained minimization problem. Three alternative formalisms for solving this problem are developed. The two ''off-line'' formalisms, with and without ''foresight'' characteristics, require the generation of a complete sensitivity data base prior to performing the uncertainty analysis. The ''online'' formalism, in which uncertainty analysis is performed interactively with the system analysis code, is best suited for treatment of large-scale highly nonlinear time-dependent problems. This methodology is applied to the uncertainty analysis of a transient upflow of a high pressure water heat transfer experiment. For comparison, an uncertainty analysis using sensitivities computed by standard response surface techniques is also performed. The results of the analysis indicate the following. Major reduction of the discrepancies in the calculation/experiment ratios is achieved by using the new methodology. Incorporation of in-bundle measurements in the uncertainty analysis significantly reduces system uncertainties. Accuracy of sensitivities generated by response-surface techniques should be carefully assessed prior to using them as a basis for uncertainty analyses of transient reactor safety problems
International Nuclear Information System (INIS)
Shimada, Yoshio
2010-01-01
The purposes of the present study are firstly to investigate the status of practical use of electric transient analysis programs used in U.S. nuclear power plants, which has been extracted as good examples from the information analysis of overseas troubles, and secondly to select a program to be recommended for use in implementing electric transient analysis in domestic nuclear power plants. In addition, to promote its practical use, a selected electric transient analysis program was tested by simulating the transient response during a load sequence test of an emergency diesel generator (EDG) in a domestic representative nuclear plant to evaluate its simulation accuracy by comparing its result with the measured plant data. The results obtained are as follows: (1) In U.S. nuclear power plants, simulations using electric transient analysis programs, such as ETAP, EMPT, etc., are widely performed, which contributed to improve the plant safety. (2) A selected transient analysis program EMTP was verified in its accuracy in terms of transient response of active power, current, voltage and frequency of the EDG during the load sequence test in a domestic representative nuclear power plant. (author)
Directory of Open Access Journals (Sweden)
Sunday J. IBRAHIM
2013-06-01
Full Text Available Safety and transient analyses of a pressurised water reactor (PWR using the Personal Computer Transient Analyzer (PCTRAN simulator was carried out. The analyses presented a synergistic integration of a numerical model; a full scope high fidelity simulation system which adopted point reactor neutron kinetics model and movable boundary two phase fluid models to simplify the calculation of the program, so it could achieve real-time simulation on a personal computer. Various scenarios of transients and accidents likely to occur at any nuclear power plant were simulated. The simulations investigated the change of signals and parameters vis a vis loss of coolant accident, scram, turbine trip, inadvertent control rod insertion and withdrawal, containment failure, fuel handling accident in auxiliary building and containment, moderator dilution as well as a combination of these parameters. Furthermore, statistical analyses of the PCTRAN results were carried out. PCTRAN results for the loss of coolant accident (LOCA caused a rapid drop in coolant pressure at the rate of 21.8KN/m2/sec triggering a shutdown of the reactor protection system (RPS, while the turbine trip accident showed a rapid drop in total plant power at the rate of 14.3 MWe/sec causing a downtime in the plant. Fuel handling accidents mimic results showed release of radioactive materials in unacceptable doses. This work shows the potential classes of nuclear accidents likely to occur during operation in proposed reactor sites. The simulations are very appropriate in the light of Nigeria’s plan to generate nuclear energy in the region of 1000 MWe from reactors by 2017.
Directory of Open Access Journals (Sweden)
Jikai Chen
2016-12-01
Full Text Available In a power system, the analysis of transient signals is the theoretical basis of fault diagnosis and transient protection theory. Shannon wavelet entropy (SWE and Shannon wavelet packet entropy (SWPE are powerful mathematics tools for transient signal analysis. Combined with the recent achievements regarding SWE and SWPE, their applications are summarized in feature extraction of transient signals and transient fault recognition. For wavelet aliasing at adjacent scale of wavelet decomposition, the impact of wavelet aliasing is analyzed for feature extraction accuracy of SWE and SWPE, and their differences are compared. Meanwhile, the analyses mentioned are verified by partial discharge (PD feature extraction of power cable. Finally, some new ideas and further researches are proposed in the wavelet entropy mechanism, operation speed and how to overcome wavelet aliasing.
Sextant: an expert system for transient analysis of nuclear reactors and integral test facilities
International Nuclear Information System (INIS)
Barbet, N.; Dumas, M.; Mihelich, G.
1987-01-01
Expert systems provide a new way of dealing with the computer-aided management of nuclear plants by combining several knowledge bases and reasoning modes together with a set of numerical models for real-time analysis of transients. New development tools are required together with metaknowledge bases handling temporal hypothetical reasoning and planning. They have to be efficient and robust because during a transient, neither measurements nor models, nor scenarios are hold as absolute references. SEXTANT is a general purpose physical analyzer intended to provide a pattern and avoid duplication of general tools and knowledge bases for similar applications. It combines several knowledge bases concerning measurements, models and qualitative behavior of PWR with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. For its development, SEXTANT requires a powerful shell. SPIRAL is such a toolkit, oriented towards online analysis of complex processes and already used in several applications
Development of an advanced code system for fast-reactor transient analysis
International Nuclear Information System (INIS)
Konstantin Mikityuk; Sandro Pelloni; Paul Coddington
2005-01-01
FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)
Kuosheng BWR/6 recirculation pump trip transient analysis with the RETRAN02/MOD5 code
International Nuclear Information System (INIS)
Wang, J.R.; Shih, C.
1992-01-01
A recirculation pump trip (RPT) event results in a reduction in recirculation flow, which reduces the core coolant flow rate. A reduction in core flow results in an increase in core void fraction and hence a decrease in core power due to negative void reactivity feedback. Although this category of events is less severe than others and generally considered as nonlimiting, core instability still may occur such as that at LaSalle on March 9, 1988. This paper focuses on the RPT transient analysis of Kuosheng Nuclear Power Plant (KNPP), which has two units of General Electric-designed boiling water reactor (BWR)/6 with rated core thermal power of 2894 MW and rated core flow of 10645 kg/s (23472 lb m /s). The approach to investigating the RPT transient of KNPP consists of two steps. The first step is to develop a plant-specific model using the RETRAN02/MOD5 code. In this step, various plant-specific information, including design documentation, drawings, safety analysis reports, and other information supplied by vendors were collected for model development. The RPT startup test at 68% power was used for system model benchmarking to ensure the adequacy of this model and identify several sensitive parameters. The second step is to assess whether similar power oscillation phenomena may occur at KNPP because of an RPT with isolated feedwater heater event. Two transient analyses (with or without reactor scram) of the KNPP RPT with isolated feedwater heater were investigated
Measurement and Analysis of Multiple Output Transient Propagation in BJT Analog Circuits
Roche, Nicolas J.-H.; Khachatrian, A.; Warner, J. H.; Buchner, S. P.; McMorrow, D.; Clymer, D. A.
2016-08-01
The propagation of Analog Single Event Transients (ASETs) to multiple outputs of Bipolar Junction Transistor (BJTs) Integrated Circuits (ICs) is reported for the first time. The results demonstrate that ASETs can appear at several outputs of a BJT amplifier or comparator as a result of a single ion or single laser pulse strike at a single physical location on the chip of a large-scale integrated BJT analog circuit. This is independent of interconnect cross-talk or charge-sharing effects. Laser experiments, together with SPICE simulations and analysis of the ASET's propagation in the s-domain are used to explain how multiple-output transients (MOTs) are generated and propagate in the device. This study demonstrates that both the charge collection associated with an ASET and the ASET's shape, commonly used to characterize the propagation of SETs in devices and systems, are unable to explain quantitatively how MOTs propagate through an integrated analog circuit. The analysis methodology adopted here involves combining the Fourier transform of the propagating signal and the current-source transfer function in the s-domain. This approach reveals the mechanisms involved in the transient signal propagation from its point of generation to one or more outputs without the signal following a continuous interconnect path.
Application of ADINA fluid element for transient response analysis of fluid-structure system
International Nuclear Information System (INIS)
Sakurai, Y.; Kodama, T.; Shiraishi, T.
1985-01-01
Pressure propagation and Fluid-Structure Interaction (FSI) in 3D space were simulated by general purpose finite element program ADINA using the displacement-based fluid element which presumes inviscid and compressible fluid with no net flow. Numerical transient solution was compared with the measured data of an FSI experiment and was found to fairly agree with the measured. In the next step, post analysis was conducted for a blowdown experiment performed with a 1/7 scaled reactor pressure vessel and a flexible core barrel and the code performance was found to be satisfactory. It is concluded that the transient response of the core internal structure of a PWR during the initial stage of LOCA can be analyzed by the displacement-based finite fluid element and the structural element. (orig.)
Analysis of Transient Phenomena Due to a Direct Lightning Strike on a Wind Energy System
Directory of Open Access Journals (Sweden)
João P. S. Catalão
2012-07-01
Full Text Available This paper is concerned with the protection of wind energy systems against the direct effects of lightning. As wind power generation undergoes rapid growth, lightning damages involving wind turbines have come to be regarded as a serious problem. Nevertheless, very few studies exist yet in Portugal regarding lightning protection of wind energy systems using numerical codes. A new case study is presented in this paper, based on a wind turbine with an interconnecting transformer, for the analysis of transient phenomena due to a direct lightning strike to the blade. Comprehensive simulation results are provided by using models of the Restructured Version of the Electro-Magnetic Transients Program (EMTP, and conclusions are duly drawn.
Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches
International Nuclear Information System (INIS)
1996-10-01
This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs
SACI - O: A code for the analysis of transients in a pressurized water reactor core
International Nuclear Information System (INIS)
Resende Lobo, A.A. de; Soares, P.A.
1979-03-01
The SACI-O digital computer code consists basically of a pressurized water reactor core model. It is useful in the analysis of fast reactivity transients shorter than the loop transit time. The program can also be used for evaluating the core behaviour, during other transients, when the inlet coolant conditions are known. SACI-O uses point model neutron kinetics taking into account moderator and fuel reactivity effects, and fission products decay. The neutronic and thermal-hydraulic equations are solved for an average fuel pin described by a single axial node. To perform a more detailed calculation, the modeling of another cooling channel, which can be divided into axial segments, is included in the program. The reactor trip system is also partially simulated. (Author) [pt
Energy Technology Data Exchange (ETDEWEB)
Li, Y.; Wong, R. K. C. [Calgary Univ., AB (Canada); Yeung, K. C. [Suncor Energy Inc., Calgary, AB (Canada)
1998-12-31
Results of an analysis of transient pressure near a horizontal well using a coupled diffusion-deformation method are discussed. The results are compared with those obtained from the single diffusivity equation. Implications for practical applications such as well testing are addressed. Results indicate that the diffusion-deformation behaviour of porous material affects the transient pressure response near a horizontal well. Evaluation by conventional well testing, based as it is on the single diffusion equation, would likely result in an overestimate of the permeability value. Comparison of results between the coupled diffusion-deformation approach and the single diffusion equation suggests that a better prediction of pressure response could be derived from total compressibility than by using only fluid compressibility. 6 refs., 9 figs.
Data Analysis of Transient Energy Releases in the LHC Superconducting Dipole Magnets
Calvi, M; Bottura, L; Di Castro, M; Masi, A; Siemko, A
2007-01-01
Premature training quenches are caused by transient energy released within the LHC dipole magnet coils while it is energized. Voltage signals recorded across the magnet coils and on the so-called quench antenna carry information about these disturbances. The transitory events correlated to transient energy released are extracted making use of continuous wavelet transform. Several analyses are performed to understand their relevance to the so called training phenomenon. The statistical distribution of the signals amplitude, the number of events occurring at a given current level, the average frequency content of the events are the main parameters on which the analysis have been focalized. Comparisons among different regions of the magnet, among different quenches in the same magnet and among magnets made by different builders are reported. Conclusions about the efficiency of the raw data treatment and the relevance of the parameters developed with respect to the magnet global behavior are finally given.
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
Greensmith, David J
2014-01-01
Here I present an Excel based program for the analysis of intracellular Ca transients recorded using fluorescent indicators. The program can perform all the necessary steps which convert recorded raw voltage changes into meaningful physiological information. The program performs two fundamental processes. (1) It can prepare the raw signal by several methods. (2) It can then be used to analyze the prepared data to provide information such as absolute intracellular Ca levels. Also, the rates of change of Ca can be measured using multiple, simultaneous regression analysis. I demonstrate that this program performs equally well as commercially available software, but has numerous advantages, namely creating a simplified, self-contained analysis workflow. Copyright © 2013 The Author. Published by Elsevier Ireland Ltd.. All rights reserved.
Trace analysis of auxiliary feedwater capacity for Maanshan PWR loss-of-normal-feedwater transient
Energy Technology Data Exchange (ETDEWEB)
Chen, Che-Hao; Shih, Chunkuan [National Tsing Hua Univ., Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Lin, Hao-Tzu [Atomic Energy Council, Taiwan (China). Inst. of Nuclear Energy Research
2013-07-01
Maanshan nuclear power plant is a Westinghouse PWR of Taiwan Power Company (Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the power of Maanshan NPP. The assessments include NSSS (Nuclear Steam Supply System) parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of auxiliary systems, and transient analyses, etc. Since the Fukushima Daiichi accident happened, it is necessary to consider transients with multiple-failure. Base on the analysis, we further study the auxiliary feedwater capability for Loss-of-Normal-Feedwater (LONF) transient. LONF is the limiting transient of non-turbine trip initiated event for ATWS (Anticipated Transient Without Scram) which results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the turbine fails to trip immediately, the secondary water inventory will decrease significantly before the actuation of auxiliary feedwater (AFW) system. The heat removal from the primary side decreases, and this leads to increases of primary coolant temperature and pressure. The water level of pressurizer also increases subsequently. The heat removal through the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation from primary side. The pressurizer will be filled with water finally, and the RCS pressure might rise above the set point of relief valves for water discharge. RCS pressure depends on steam generator inventory, primary coolant temperature, negative reactivity feedback, and core power, etc. The RCS pressure may reach its peak after core power reduction. According to ASME Code Level C service limit criteria, the Reactor Coolant System (RCS) pressure must be under 22.06 MPa. The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development of TRACE is based on TRAC and integrating with RELAP5 and other programs. SNAP
Trace analysis of auxiliary feedwater capacity for Maanshan PWR loss-of-normal-feedwater transient
International Nuclear Information System (INIS)
Chen, Che-Hao; Shih, Chunkuan; Wang, Jong-Rong; Lin, Hao-Tzu
2013-01-01
Maanshan nuclear power plant is a Westinghouse PWR of Taiwan Power Company (Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the power of Maanshan NPP. The assessments include NSSS (Nuclear Steam Supply System) parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of auxiliary systems, and transient analyses, etc. Since the Fukushima Daiichi accident happened, it is necessary to consider transients with multiple-failure. Base on the analysis, we further study the auxiliary feedwater capability for Loss-of-Normal-Feedwater (LONF) transient. LONF is the limiting transient of non-turbine trip initiated event for ATWS (Anticipated Transient Without Scram) which results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the turbine fails to trip immediately, the secondary water inventory will decrease significantly before the actuation of auxiliary feedwater (AFW) system. The heat removal from the primary side decreases, and this leads to increases of primary coolant temperature and pressure. The water level of pressurizer also increases subsequently. The heat removal through the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation from primary side. The pressurizer will be filled with water finally, and the RCS pressure might rise above the set point of relief valves for water discharge. RCS pressure depends on steam generator inventory, primary coolant temperature, negative reactivity feedback, and core power, etc. The RCS pressure may reach its peak after core power reduction. According to ASME Code Level C service limit criteria, the Reactor Coolant System (RCS) pressure must be under 22.06 MPa. The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development of TRACE is based on TRAC and integrating with RELAP5 and other programs. SNAP
CEDNBR: a computer code for transient thermal margin analysis of a reactor core
International Nuclear Information System (INIS)
Shesler, A.T.; Lehmann, C.R.
1976-09-01
The report describes the CEDNBR computer code. This code was developed for the transient thermal analysis of a pressurized water reactor core or a critical heat flux test. Included are the code structure, conservation equations, and correlations utilized by CEDNBR. The methods of modelling a reactor core and hot channel and a CHF test are presented. Comparisons of CEDNBR calculations are made with both empirical pressure loss data and simulated loss of flow test data. The code solves the one-dimensional conservation of mass, energy, and momentum equations and the equation of state for the fluid for either steady-state or transient conditions. Tabular time dependent functions of inlet temperatures, pressure, mass velocity, axial heat flux distributions, normalized heat flux, radial peaking factors, and incremental mixing factors are required input to the code. Transient effects are included in the calculation of enthalpy rise and fluid properties. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by applying a Critical Heat Flux (CHF) correlation to the computed local fluid properties. A code user's guide is provided for preparing input to the code. In addition, descriptions of the sub-routines used by CEDNBR are given
International Nuclear Information System (INIS)
Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.
2012-01-01
The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)
LOFT transient thermal analysis for 10 inch primary coolant blowdown piping weld
International Nuclear Information System (INIS)
Howell, S.K.
1978-01-01
A flaw in a weld in the 10 inch primary coolant blowdown piping was discovered by LOFT personnel. As a result of this, a thermal analysis and fracture mechanics analysis was requested by LOFT personnel. The weld and pipe section were analyzed for a complete thermal cycle, heatup and Loss of Coolant Experiment (LOCE), using COUPLE/MOD2, a two-dimensional finite element heat conduction code. The finite element representation used in this analysis was generated by the Applied Mechanics Branch. The record of nodal temperatures for the entire transient was written on tape VSN=T9N054, and has been forwarded to the Applied Mechanics Branch for use in their mechanical analysis. Specific details and assumptions used in this analysis are found in appropriate sections of this report
Development of the containment transient analysis code for the passive reactor
Energy Technology Data Exchange (ETDEWEB)
Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-05-01
This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.
Two-dimensional transient thermal analysis of a fuel rod by finite volume method
Energy Technology Data Exchange (ETDEWEB)
Costa, Rhayanne Yalle Negreiros; Silva, Mário Augusto Bezerra da; Lira, Carlos Alberto de Oliveira, E-mail: ryncosta@gmail.com, E-mail: mabs500@gmail.com, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear
2017-07-01
One of the greatest concerns when studying a nuclear reactor is the warranty of safe temperature limits all over the system at all time. The preservation of core structure along with the constraint of radioactive material into a controlled system are the main focus during the operation of a reactor. The purpose of this paper is to present the temperature distribution for a nominal channel of the AP1000 reactor developed by Westinghouse Co. during steady-state and transient operations. In the analysis, the system was subjected to normal operation conditions and then to blockages of the coolant flow. The time necessary to achieve a new safe stationary stage (when it was possible) was presented. The methodology applied in this analysis was based on a two-dimensional survey accomplished by the application of Finite Volume Method (FVM). A steady solution is obtained and compared with an analytical analysis that disregard axial heat transport to determine its relevance. The results show the importance of axial heat transport consideration in this type of study. A transient analysis shows the behavior of the system when submitted to coolant blockage at channel's entrance. Three blockages were simulated (10%, 20% and 30%) and the results show that, for a nominal channel, the system can still be considerate safe (there's no bubble formation until that point). (author)
Directory of Open Access Journals (Sweden)
Sang Hwan Lee
Full Text Available BACKGROUND: Early discrimination between transient and persistent par-solid ground-glass nodules (PSNs at CT is essential for patient management. The objective of our study was to retrospectively investigate the value of texture analysis in differentiating pulmonary transient and persistent PSNs in addition to clinical and CT features. METHODS: This retrospective study was performed with IRB approval and a waiver of the requirement for patients' informed consent. From January 2007 to October 2009, we identified 77 individuals (39 men and 38 women; mean age, 55 years with 86 PSNs on thin-section chest CT. Thirty-nine PSNs in 31 individuals were transient and 47 PSNs in 46 patients were persistent. The clinical, CT, and texture features of PSNs were evaluated. To investigate the additional value of texture analysis in differentiating transient from persistent PSNs, logistic regression analysis and C-statistics were performed. RESULTS: Between transient and persistent PSNs, there were significant differences in age, gender, smoking history, and eosinophil count among the clinical features. As for thin-section CT features, there were significant differences in lesion size, solid portion size, and lesion multiplicity. In terms of texture features, there were significant differences in mean attenuation, skewness of whole PSN, attenuation ratio of whole PSN to inner solid portion, and 5-, 10-, 25-, 50-percentile CT numbers of whole PSN. Multivariate analysis revealed eosinophilia, lesion size, lesion multiplicity, mean attenuation of whole PSN, skewness of whole PSN, and 5-percentile CT number were significant independent predictors of transient PSNs. (P<0.05 C-statistics revealed that texture analysis incorporating clinical and CT features (AUC, 92.9% showed significantly higher differentiating performance of transient from persistent PSNs compared with the clinical and CT features alone (AUC, 79.0%. (P = 0.004. CONCLUSION: Texture analysis of
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.
Detailed Analysis of the Transient Voltage in a JT-60SA PF Coil Circuit
International Nuclear Information System (INIS)
Yamauchi, K.; Shimada, K.; Terakado, T.; Matsukawa, M.; Coletti, R.; Lampasi, A.; Gaio, E.; Coletti, A.; Novello, L.
2013-01-01
A superconducting coil system is actually complicated by the distributed parameters, e.g. the distributed mutual inductance among turns and the distributed capacitance between adjacent conductors. In this paper, such a complicated system was modeled with a reasonably simplified circuit network with lumped parameters. Then, a detailed circuit analysis was conducted to evaluate the possible voltage transient in the coil circuit. As a result, an appropriate (minimum) snubber capacitance for the Switching Network Unit, which is a fast high voltage generation circuit in JT-60SA, was obtained. (fusion engineering)
International Nuclear Information System (INIS)
Gorlandi, A.; Mazzini, M.; Oriolo, F.
1979-01-01
This works briefly describes the features of the computation codes available at the Istituto di Impianti Nucleari of the Pisa University for the analysis of the thermofluidodynamic transient in the containment system of a nuclear power plant following a LOCA (RELAP 4/MOD.S, COMPARE, FUMO and CONTEMPT-LT/026). More details are contained in the Annex. Particular attention has been devoted to the opportunity to study, through the computation codes, the effects of the sub division of a full pressure containment system
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
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
Recent developments in transient magneto-structural integrated analysis for fusion applications
International Nuclear Information System (INIS)
Crutzen, Y.; Papadopoulos, S.; Richard, N.; Siakavellas, N.; Wu, J.
1992-01-01
In this paper three different numerical approaches modelling the mutual field-structure interactions during transient electromagnetic events are presented. The application of these approaches to simple plate models, simulating flexible conducting components of fusion devices, show that a magnetic damping is encountered when coupling effects between eddy currents and plate motion are taken into account. This damping increases with the applied magnetic field, modifying the mechanical behavior. An Integrated Design/Analysis System is also proposed, in order to combine different computer codes, obtaining performing computational schemes, in the field of 3D electromagneto-mechanical analyses
Optimization of High-Resolution Continuous Flow Analysis for Transient Climate Signals in Ice Cores
DEFF Research Database (Denmark)
Bigler, Matthias; Svensson, Anders; Kettner, Ernesto
2011-01-01
Over the past two decades, continuous flow analysis (CFA) systems have been refined and widely used to measure aerosol constituents in polar and alpine ice cores in very high-depth resolution. Here we present a newly designed system consisting of sodium, ammonium, dust particles, and electrolytic...... meltwater conductivity detection modules. The system is optimized for high- resolution determination of transient signals in thin layers of deep polar ice cores. Based on standard measurements and by comparing sections of early Holocene and glacial ice from Greenland, we find that the new system features...
International Nuclear Information System (INIS)
Amico, P.J.; Hsu, C.J.; Youngblood, R.W.; Fitzpatrick, R.G.
1989-01-01
This paper reports that as part of a probabilistic assessment of the safety significance of complex transients at certain PWR power plants, it was necessary to perform a cognitive human reliability analysis. To increase the confidence in the results, it was desirable to make use of actual observations of operator response which were available for the assessment. An approach was developed which incorporated these observations into the human cognitive reliability (HCR) modeling approach. The results obtained provided additional insights over what would have been found using other approaches. These insights were supported by the observations, and it is suggested that this approach be considered for use in future probabilistic safety assessments
Safety analysis of Atucha 1 reactor pressure vessel for a typical transient
International Nuclear Information System (INIS)
Chomik, E.; Jinchuk, D.
1994-01-01
As a consequence of disturbances on the CNA I external electric grid some incidents were produced in a 6 minutes lapse, causing a sudden cooling of the primary system, while pressure was maintained nearly constant. On the basis of this event, a safety analysis based on the LInear Elastic Fracture Mechanics was carried out. This paper presents an alternative method for the calculation of transients; the Finite Element Method, particularly, the OCA-II FEM code. By using this method it was possible to demonstrate, for this event, a safe operating condition for the end of life of the RPV, with regard to brittle fracture risk. 6 refs, 11 figs, 1 tab
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
International Nuclear Information System (INIS)
Shin, Hyeong-Ki
1999-01-01
The severe accidents that occurred recently on nuclear reactors such as Chernobyl and T.M.1.2 have led many countries utilizing nuclear energy to examine their severe accident management. This thesis focuses on this problem and aims at analyzing, in terms of reactivity, degraded core behavior resulting from different accidental configurations. Two types of core degradation can be encountered: local degradation (the destruction of isolated assemblies in the core) or spreading degradation (the destruction of neighboring assemblies). The TMI accident is an example of spreading degradation in the core. The simplicity of implementing the control rod ejection accident calculation as compared to other accidental transients have motivated the choice of this accident as a determinant for local degraded core configurations. The control rod ejection accident presents important three dimensional effects and introduces neutronic/thermohydraulic coupling. The implementation and validation of already existing three dimensional coupled calculation scheme, allowed one to analyze the consequences of such an accident and to the conclusion that only unrealistic hypotheses of assembly permutation could lead to a partial core degradation. A reasonable estimate of stored energy in the assemblies with high bum up, in relation to the stored energy in the hot spot, was also obtained for the first time. The recently performed experiments (CABRI experiments) showed that in highly burned up assemblies, the capacity to store energy decreases strongly in relation to new assemblies. This first estimate of the distribution of produced energy between different assemblies, during the rod ejection accident, offers an important piece of knowledge in the study of the consequences of an eventual fuel cycle extension (presently under consideration by development companies). Finally, the analysis of degraded core reactivity itself has been performed for a vast range of the degraded core configurations
Nuclear fuel management and transients analysis in Laguna Verde nuclear power plant
International Nuclear Information System (INIS)
De Loera De Haro, M.A.; Alvarez Gasca, J.
1991-01-01
Nuclear fuel management transient analysis are the set of activities which determine the load and reload of nuclear fuel inside the reactor, with the aim of getting the maximum performance in fuel burn up and heat remotion, without have an effect in the station safety. Nuclear fuel management and transient analysis has its basis on high precision quantitative analysis methodologies by means of simulation of nuclear and physical phenomena occurring both in normal and abnormal operation of nuclear power plants. On account of complexity of simulations and the required precision, those are carry out using codes type 'best estimate'. For the use of this tools it is necessary a deep knowledge of simulated nuclear and physical phenomena, as well as the used mathematical models and the numerical methods used. If different, the simulation results will be notably different actual processes owing to the use of models out of validity range, or incorrect calculations in the input parameters. On account of complexity of simulations and the required precision, those are carry out using codes type 'best estimate'. For the use of this tools it is necessary a deep knowledge of simulated nuclear and physical phenomena, as well as the used mathematical models and the numerical methods used. If different, the simulation results will be notably different actual processes owing to the use of models out of validity range, or incorrect calculations in the input parameters
Transient dynamic and inelastic analysis of shells of revolution - a survey of programs
International Nuclear Information System (INIS)
Svalbonas, V.
1976-01-01
Advances in the limits of structural use in the aerospace and nuclear power industries over the past years have increased the requirements upon the applicable analytical computer programs to include accurate capabilities for inelastic and transient dynamic analyses. In many minds, however, this advanced capability is unequivocally linked with the large scale, general purpose, finite element programs. This idea is also combined with the view that such analyses are therefore prohibitively expensive and should be relegated to the 'last resort' classification. While this, in the general sense, may indeed be the case, if the user needs only to analyze structures falling into limited categories, however, he may find that a variety of smaller special purpose programs are available which do not put an undue strain upon his resources. One such structural category is shells of revolution. This survey of programs concentrates upon the analytical tools which have been developed predominantly for shells of revolution. The survey is subdivided into three parts: (a) consideration of programs for transient dynamic analysis; (b) consideration of programs for inelastic analysis and finally; (c) consideration of programs capable of dynamic plasticity analysis. In each part, programs based upon finite difference, finite element, and numerical integration methods are considered. The programs are compared on the basis of analytical capabilities, and ease of idealization and use. In each part of the survey sample problems are utilized to exemplify the state-of-the-art. (Auth.)
International Nuclear Information System (INIS)
Chvetsov, I.; Volkov, A.
2000-01-01
For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)
Energy Technology Data Exchange (ETDEWEB)
Navarro-Valenti, S.; Kim, S.H.; Georgevich, V. [Oak Ridge National Lab., TN (United States)] [and others
1995-09-01
The purpose of this paper is to describe the analysis performed to predict the thermal behavior of fuel miniplates under rapid transient heatup conditions. The possibility of explosive boiling was considered, and it was concluded that the heating rates are not large enough for explosive boiling to occur. However, transient boiling effects were pronounced. Because of the complexity of transient pool boiling and the unavailability of experimental data for the situations studied, an approximation was made that predicted the data very well within the uncertainties present. If pool boiling from the miniplates had been assumed to be steady during the heating pulse, the experimental data would have been greatly overestimated. This fact demonstrates the importance of considering the transient nature of heat transfer in the analysis of reactivity excursion accidents. An additional contribution of the present work is that it provided data on highly subcooled steady nulceate boiling from the cooling portion of the thermocouple traces.
International Nuclear Information System (INIS)
Hall, P.; Hutt, P.
1994-01-01
This paper describes Nuclear Electric's (NE) development of an integrated code package in support of all its reactors including Sizewell B, designed for the provision of fuel management design, core performance studies, operational support and fault transient analysis. The package uses the NE general purpose three-dimensional transient reactor physics code PANTHER with cross-sections derived in the PWR case from the LWRWIMS LWR lattice neutronics code. The package also includes ENIGMA a generic fuel performance code and for PWR application VIPRE-01 a subchannel thermal hydraulics code, RELAP5 the system thermal hydraulics transient code and SCORPIO an on-line surveillance system. The paper describes the capabilities and validation of the elements of this package for PWR, how they are coupled within the package and the way in which they are being applied for Sizewell B to on-line surveillance and fault transient analysis. (Author)
TRAC-BD1: transient reactor analysis code for boiling-water systems
International Nuclear Information System (INIS)
Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.
1981-01-01
The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented
Electromagnetic transient analysis and Novell protective relaying techniques for power transformers
Lin, X; Tian, Q; Weng, H
2015-01-01
This book addresses the technical challenges of transformer malfunction analysis as well as protection. One of the current research directions is the malfunction mechanism analysis due to nonlinearity of transformer core and comprehensive countermeasures on improving the performance of transformer differential protection. Here, the authors summarize their research outcomes and present a set of recent research advances in the electromagnetic transient analysis, the application on power transformer protections, and present a more systematic investigation and review in this field. This research area is still progressing, especially with the fast development of Smart Grid. This book is an important addition to the literature and will enhance significant advancement in research. It is a good reference book for researchers in power transformer protection research and a good text book for graduate and undergraduate students in electrical engineering.
International Nuclear Information System (INIS)
2007-12-01
Accident analysis is an important tool for ensuring the adequacy and efficiency of the provision in the defence in depth concept to cope with challenges to plant safety. Accident analysis is the milestone of the demonstration that the plant is capable of meeting any prescribed limits for radioactive releases and any other acceptable limits for the safe operation of the plant. It is used, by designers, utilities and regulators, in a number of applications such as: (a) licensing of new plants, (b) modification of existing plants, (c) analysis of operational events, (d) development, improvement or justification of the plant operational limits and conditions, and (e) safety cases. According to the defence in depth concept, the fuel rod cladding constitutes the first containment barrier of the fission products. Therefore, related safety objectives and associated criteria are defined, in order to ensure, at least for normal operation and anticipated transients, the integrity of the cladding, and for accident conditions, acceptable radiological consequences with regard to the postulated frequency of the accident, as usually identified in the safety analysis reports. Therefore, computational analysis of fuel behaviour under steady state, transient and accident conditions constitutes a major link of the safety case in order to justify the design and the safety of the fuel assemblies, as far as all relevant phenomena are correctly addressed and modelled. This publication complements the IAEA Safety Report on Accident Analysis for Nuclear Power Plants (Safety Report Series No. 23) that provides practical guidance for establishing a set of conceptual and formal methods and practices for performing accident analysis. Computational analysis of the behaviour of nuclear fuel under transient and accident conditions, including normal operation (e.g. power ramp rates) is developed in this publication. For design basis accidents, depending on the type of influence on a fuel element
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
FEA stress analysis considering cavity formation of metallic fuel pin under transient state
Energy Technology Data Exchange (ETDEWEB)
Jung, Hyun-Woo; Oh, Young-Ryun; Kim, Yun-Jae [Korea University, Seoul (Korea, Republic of)
2016-05-15
The aim of this research is to study the stress state of the fuel and the cladding under transient state using the commercial finite element analysis software, ABAQUS v6.13. It is checked out that the gap distance between the fuel and the cladding is a major factor determining FCMI stress. In this regard, initial boundary condition of the fuel pin such as the initial gap distance should be set carefully when the stress analysis of the fuel pin under transient state is conducted. In case of simulating cavity formation, it is confirmed that the new cavity simulation model that elements in cavity region lose their stiffness is valid. There is a great deal of research into SFR, which is one of GEN IV reactors. When it comes to the accidents of SFR, there are two cases of accident process. One of them is In-pin process that molten fuel is discharged into upper plenum. The other is Ex-pin process that the molten fuel is discharged into coolant because of breakage of cladding.
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)
Analysis of a Plant Transcriptional Regulatory Network Using Transient Expression Systems.
Díaz-Triviño, Sara; Long, Yuchen; Scheres, Ben; Blilou, Ikram
2017-01-01
In plant biology, transient expression systems have become valuable approaches used routinely to rapidly study protein expression, subcellular localization, protein-protein interactions, and transcriptional activity prior to in vivo studies. When studying transcriptional regulation, luciferase reporter assays offer a sensitive readout for assaying promoter behavior in response to different regulators or environmental contexts and to confirm and assess the functional relevance of predicted binding sites in target promoters. This chapter aims to provide detailed methods for using luciferase reporter system as a rapid, efficient, and versatile assay to analyze transcriptional regulation of target genes by transcriptional regulators. We describe a series of optimized transient expression systems consisting of Arabidopsis thaliana protoplasts, infiltrated Nicotiana benthamiana leaves, and human HeLa cells to study the transcriptional regulations of two well-characterized transcriptional regulators SCARECROW (SCR) and SHORT-ROOT (SHR) on one of their targets, CYCLIN D6 (CYCD6).Here, we illustrate similarities and differences in outcomes when using different systems. The plant-based systems revealed that the SCR-SHR complex enhances CYCD6 transcription, while analysis in HeLa cells showed that the complex is not sufficient to strongly induce CYCD6 transcription, suggesting that additional, plant-specific regulators are required for full activation. These results highlight the importance of the system and suggest that including heterologous systems, such as HeLa cells, can provide a more comprehensive analysis of a complex gene regulatory network.
Transient analysis of an HTS DC power cable with an HVDC system
International Nuclear Information System (INIS)
Dinh, Minh-Chau; Ju, Chang-Hyeon; Kim, Jin-Geun; Park, Minwon; Yu, In-Keun; Yang, Byeongmo
2013-01-01
Highlights: •A model of an HTS DC power cable was developed using real time digital simulator. •The simulations of the HTS DC power cable in connection with an HVDC system were performed. •The transient analysis results of the HTS DC power cable were presented. -- Abstract: The operational characteristics of a superconducting DC power cable connected to a highvoltage direct current (HVDC) system are mainly concerned with the HVDC control and protection system. To confirm how the cable operates with the HVDC system, verifications using simulation tools are needed. This paper presents a transient analysis of a high temperature superconducting (HTS) DC power cable in connection with an HVDC system. The study was conducted via the simulation of the HVDC system and a developed model of the HTS DC power cable using a real time digital simulator (RTDS). The simulation was performed with some cases of short circuits that may have caused system damage. The simulation results show that during the faults, the quench did not happen with the HTS DC power cable because the HVDC controller reduced some degree of the fault current. These results could provide useful data for the protection design of a practical HVDC and HTS DC power cable system
Transient analysis of an HTS DC power cable with an HVDC system
Energy Technology Data Exchange (ETDEWEB)
Dinh, Minh-Chau, E-mail: thanchau7787@gmail.com [Department of Electrical Engineering, Changwon National University, 9 Sarim-Dong, Changwon 641-773 (Korea, Republic of); Ju, Chang-Hyeon; Kim, Jin-Geun; Park, Minwon [Department of Electrical Engineering, Changwon National University, 9 Sarim-Dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@cwnu.ac.kr [Department of Electrical Engineering, Changwon National University, 9 Sarim-Dong, Changwon 641-773 (Korea, Republic of); Yang, Byeongmo [Korea Electric Power Research Institute, 105 Munji-Ro, Yuseong-Gu, Daejon 305-760 (Korea, Republic of)
2013-11-15
Highlights: •A model of an HTS DC power cable was developed using real time digital simulator. •The simulations of the HTS DC power cable in connection with an HVDC system were performed. •The transient analysis results of the HTS DC power cable were presented. -- Abstract: The operational characteristics of a superconducting DC power cable connected to a highvoltage direct current (HVDC) system are mainly concerned with the HVDC control and protection system. To confirm how the cable operates with the HVDC system, verifications using simulation tools are needed. This paper presents a transient analysis of a high temperature superconducting (HTS) DC power cable in connection with an HVDC system. The study was conducted via the simulation of the HVDC system and a developed model of the HTS DC power cable using a real time digital simulator (RTDS). The simulation was performed with some cases of short circuits that may have caused system damage. The simulation results show that during the faults, the quench did not happen with the HTS DC power cable because the HVDC controller reduced some degree of the fault current. These results could provide useful data for the protection design of a practical HVDC and HTS DC power cable system.
Analysis of transient state in HTS tapes under ripple DC load current
Stepien, M.; Grzesik, B.
2014-05-01
The paper concerns the analysis of transient state (quench transition) in HTS tapes loaded with the current having DC component together with a ripple component. Two shapes of the ripple were taken into account: sinusoidal and triangular. Very often HTS tape connected to a power electronic current supply (i.e. superconducting coil for SMES) that delivers DC current with ripples and it needs to be examined under such conditions. Additionally, measurements of electrical (and thermal) parameters under such ripple excitation is useful to tape characterization in broad range of load currents. The results presented in the paper were obtained using test bench which contains programmable DC supply and National Instruments data acquisition system. Voltage drops and load currents were measured vs. time. Analysis of measured parameters as a function of the current was used to tape description with quench dynamics taken into account. Results of measurements were also used to comparison with the results of numerical modelling based on FEM. Presented provisional results show possibility to use results of measurements in transient state to prepare inverse models of superconductors and their detailed numerical modelling.
System transient analysis code development for low pressure and low power
International Nuclear Information System (INIS)
Kim, Hee Cheol
1998-02-01
A real time reactor system analysis code, ARTIST, based on drift flux model has been developed to investigate the transient system behavior under low pressure, low flow and low power conditions with noncondensable gas present in the system. The governing equations of the ARTIST code consist of three mass continuity equations (steam, liquid and noncondensable), two energy equations (gas and mixture) and one momentum equation (mixture) constituted with the drift flux model. The capability of ARTIST in predicting two-phase flow void distribution in the system has been validated against experimental data. The results of the ARTIST axial void distribution at low pressure and low flow, are far better than the results of both the homogeneous model of TASS code and the two-fluid model of RELAP5/MOD3 code. Also, RELAP5/MOD3 calculation shows the large amplitude of void fraction oscillations at low pressure. These results imply that interfacial momentum transfer terms in the two-fluid model formulation should be carefully constituted, especially for the low pressure condition due to the big density differences between steam and water. Thermal-hydraulic state solution scheme is developed when noncondensable gas exists. Numerical consistency and convergence of obtaining equilibrium state is tested with the ideal problems for various situations including very low partial pressure conditions. Calculated thermal-hydraulic state for each test shows consistent and expected behaviour. A new multi-layer back propagation network algorithm for calculating the departure from nucleate boiling ratio (DNBR) is developed and adopted in ARTIST code in order to have real-time DNBR evaluation by eliminating the tandem procedure of the transient DNBR calculation. The algorithm trained by different patterns generated by latin hypercube sampling method on the performance space is tested for the randomly sampled untrained data and the transient DNBR data. The uncertainty of the algorithm is
PASP Plus Transient Pulse Monitor (TPM) - Data Analysis and Interpretation Report
National Research Council Canada - National Science Library
Adamo, Richard
1996-01-01
The Transient Pulse Monitor (TPM), part of the PASP Plus experiment aboard the APEX spacecraft, is designed to detect and characterize electromagnetic transient signals produced by electrostatic discharges on the solar array test modules...
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.
Directory of Open Access Journals (Sweden)
Mohammed Hussein
2007-01-01
Full Text Available The transient response of erodable surface thermocouples has been numerically assessed by using a two dimensional finite element analysis. Four types of base metal erodable surface thermocouples have been examined in this study, included type-K (alumel-chromel, type-E (chromel-constantan, type-T (copper-constantan, and type-J (iron-constantan with 50 mm thick- ness for each. The practical importance of these types of thermocouples is to be used in internal combustion engine studies and aerodynamics experiments. The step heat flux was applied at the surface of the thermocouple model. The heat flux from the measurements of the surface temperature can be commonly identified by assuming that the heat transfer within these devices is one-dimensional. The surface temperature histories at different positions along the thermocouple are presented. The normalized surface temperature histories at the center of the thermocouple for different types at different response time are also depicted. The thermocouple response to different heat flux variations were considered by using a square heat flux with 2 ms width, a sinusoidal surface heat flux variation width 10 ms period and repeated heat flux variation with 2 ms width. The present results demonstrate that the two dimensional transient heat conduction effects have a significant influence on the surface temperature history measurements made with these devices. It was observed that the surface temperature history and the transient response for thermocouple type-E are higher than that for other types due to the thermal properties of this thermocouple. It was concluded that the thermal properties of the surrounding material do have an impact, but the properties of the thermocouple and the insulation materials also make an important contribution to the net response.
Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS
Energy Technology Data Exchange (ETDEWEB)
Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)
2015-11-30
The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.
Transient and Steady-State Analysis of Nonlinear RF and Microwave Circuits
Directory of Open Access Journals (Sweden)
Zhu Lei(Lana
2006-01-01
Full Text Available This paper offers a review of simulation methods currently available for the transient and steady-state analysis of nonlinear RF and microwave circuits. The most general method continues to be the time-marching approach used in Spice, but more recent methods based on multiple time dimensions are particularly effective for RF and microwave circuits. We derive nodal formulations for the most widely used multiple time dimension methods. We put special emphasis on methods for the analysis of oscillators based in the warped multitime partial differential equations (WaMPDE approach. Case studies of a Colpitts oscillator and a voltage controlled Clapp-Gouriet oscillator are presented and discussed. The accuracy of the amplitude and phase of these methods is investigated. It is shown that the exploitation of frequency-domain latency reduces the computational effort.
Nhu Y, Do
2018-03-01
Vietnam has many advantages of wind power resources. Time by time there are more and more capacity as well as number of wind power project in Vietnam. Corresponding to the increase of wind power emitted into national grid, It is necessary to research and analyze in order to ensure the safety and reliability of win power connection. In national distribution grid, voltage sag occurs regularly, it can strongly influence on the operation of wind power. The most serious consequence is the disconnection. The paper presents the analysis of distribution grid's transient process when voltage is sagged. Base on the analysis, the solutions will be recommended to improve the reliability and effective operation of wind power resources.
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.
An analysis of transient thermal properties for high power GaN-based laser diodes
Energy Technology Data Exchange (ETDEWEB)
Kim, Jae Min; Kim, Seungtaek; Kang, Sung Bok; Kim, Young Jin; Jeong, Hoon; Lee, Kyeongkyun; Kim, Jongseok [Korea Institute of Industrial Technology, 35-3 Hongcheon-Ri, Ipjang-Myeon, Cheonan, Chungnam 331-825 (Korea); Lee, Sangdon; Suh, Dongsik [QSI Co., Ltd., 315-9 Cheonheung-Ri, Sungger-Eup, Cheonan, Chungnam 330-836 (Korea); Yi, Jeong Hoon; Choi, Yoonho; Jung, Seok Gu; Noh, Minsoo [LG Electronics Advanced Research Institute, 16 Woomyeon-Dong, Seocho-Gu, Seoul 137-724 (Korea)
2010-07-15
Thermal properties of 405 nm GaN-based laser diodes were investigated by employing a transient heating response method based on the temperature dependence of diode forward voltage. Thermal resistances of materials consisting of packaged laser diodes were differentiated in transient thermal response curves at a current below threshold current. With a current above threshold current, no significant change in thermal resistances and difference between junction-up and junction-down laser diodes was observed at pulses shorter than 3 sec. From an analysis with long current injections, thermal resistance of a packaged laser diode with a junction-up bonding was {proportional_to}45 C/W which was higher than that of a junction-down bonded laser diode by {proportional_to}10 C/W. Further analyses based on parameters obtained from voltage recovery curves indicated that the time constant for cooling is directly related to the thermal resistance and thermal capacitance of a laser diode package. (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)
Rayleigh-Taylor instability under curved substrates: An optimal transient growth analysis
Balestra, Gioele; Brun, P.-T.; Gallaire, François
2016-12-01
We investigate the stability of thin viscous films coated on the inside of a horizontal cylindrical substrate. In such a case, gravity acts both as a stabilizing force through the progressive drainage of the film and as a destabilizing force prone to form droplets via the Rayleigh-Taylor instability. The drainage solution, derived from lubrication equations, is found asymptotically stable with respect to infinitesimally small perturbations, although in reality, droplets often form. To resolve this paradox, we perform an optimal transient growth analysis for the first-order perturbations of the liquid's interface, generalizing the results of Trinh et al. [Phys. Fluids 26, 051704 (2014), 10.1063/1.4876476]. We find that the system displays a linear transient growth potential that gives rise to two different scenarios depending on the value of the Bond number (prescribing the relative importance of gravity and surface tension forces). At low Bond numbers, the optimal perturbation of the interface does not generate droplets. In contrast, for higher Bond numbers, perturbations on the upper hemicircle yield gains large enough to potentially form droplets. The gain increases exponentially with the Bond number. In particular, depending on the amplitude of the initial perturbation, we find a critical Bond number above which the short-time linear growth is sufficient to trigger the nonlinear effects required to form dripping droplets. We conclude that the transition to droplets detaching from the substrate is noise and perturbation dependent.
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
Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition
International Nuclear Information System (INIS)
Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.
1996-12-01
Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76 o C which is much less than the limiting maximum value of fuel temperature of 1150 0 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000 o C, the calculated outer cladding wall temperature was observed to be 702.39 0 C compared to a value of 760 o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)
Transient analysis and leakage detection algorithm using GA and HS algorithm for a pipeline system
Energy Technology Data Exchange (ETDEWEB)
Kim, Sang Hyun; Yoo, Wan Suk; Oh, Kwang Jung; Hwang, In Sung; Oh, Jeong Eun [Pusan National University, Pusan (Korea, Republic of)
2006-03-15
The impact of leakage was incorporated into the transfer functions of the complex head and discharge. The impedance transfer functions for the various leaking pipeline systems were also derived. Hydraulic transients could be efficiently analyzed by the developed method. The simulation of normalized pressure variation using the method of characteristics and the impulse response method shows good agreement to the condition of turbulent flow. The leak calibration could be performed by incorporation of the impulse response method with Genetic Algorithm (GA) and Harmony Search (HS). The objective functions for the leakage detection can be made using the pressure-head response at the valve, or the pressure-head or the flow response at a certain point of the pipeline located upstream from the valve. The proposed method is not constrained by the Courant number to control the numerical dissipation of the method of characteristics. The limitations associated with the discreteness of the pipeline system in the inverse transient analysis can be neglected in the proposed method.
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
Visual scan-path analysis with feature space transient fixation moments
Dempere-Marco, Laura; Hu, Xiao-Peng; Yang, Guang-Zhong
2003-05-01
The study of eye movements provides useful insight into the cognitive processes underlying visual search tasks. The analysis of the dynamics of eye movements has often been approached from a purely spatial perspective. In many cases, however, it may not be possible to define meaningful or consistent dynamics without considering the features underlying the scan paths. In this paper, the definition of the feature space has been attempted through the concept of visual similarity and non-linear low dimensional embedding, which defines a mapping from the image space into a low dimensional feature manifold that preserves the intrinsic similarity of image patterns. This has enabled the definition of perceptually meaningful features without the use of domain specific knowledge. Based on this, this paper introduces a new concept called Feature Space Transient Fixation Moments (TFM). The approach presented tackles the problem of feature space representation of visual search through the use of TFM. We demonstrate the practical values of this concept for characterizing the dynamics of eye movements in goal directed visual search tasks. We also illustrate how this model can be used to elucidate the fundamental steps involved in skilled search tasks through the evolution of transient fixation moments.
Numerical analysis of steady and transient natural convection in an enclosed cavity
Mehedi, Tanveer Hassan; Tahzeeb, Rahat Bin; Islam, A. K. M. Sadrul
2017-06-01
The paper presents the numerical simulation of natural convection heat transfer of air inside an enclosed cavity which can be helpful to find out the critical width of insulation in air insulated walls seen in residential buildings and industrial furnaces. Natural convection between two walls having different temperatures have been simulated using ANSYS FLUENT 12.0 in both steady and transient conditions. To simulate different heat transfer and fluid flow conditions, Rayleigh number ranging from 103 to 105 has been maintained (i.e. Laminar flow.) In case of steady state analysis, the CFD predictions were in very good agreement with the reviewed literature. Transient simulation process has been performed by using User Defined Functions, where the temperature of the hot wall varies with time linearly. To obtain and compare the heat transfer properties, Nusselt number has been calculated at the hot wall at different conditions. The buoyancy driven flow characteristics have been investigated by observing the flow pattern in a graphical manner. The characteristics of the system at different temperature differences between the wall has been observed and documented.
Transient analysis and leakage detection algorithm using GA and HS algorithm for a pipeline system
International Nuclear Information System (INIS)
Kim, Sang Hyun; Yoo, Wan Suk; Oh, Kwang Jung; Hwang, In Sung; Oh, Jeong Eun
2006-01-01
The impact of leakage was incorporated into the transfer functions of the complex head and discharge. The impedance transfer functions for the various leaking pipeline systems were also derived. Hydraulic transients could be efficiently analyzed by the developed method. The simulation of normalized pressure variation using the method of characteristics and the impulse response method shows good agreement to the condition of turbulent flow. The leak calibration could be performed by incorporation of the impulse response method with Genetic Algorithm (GA) and Harmony Search (HS). The objective functions for the leakage detection can be made using the pressure-head response at the valve, or the pressure-head or the flow response at a certain point of the pipeline located upstream from the valve. The proposed method is not constrained by the Courant number to control the numerical dissipation of the method of characteristics. The limitations associated with the discreteness of the pipeline system in the inverse transient analysis can be neglected in the proposed method
Comparison and analysis on transient characteristics of integral pressurized water reactors
International Nuclear Information System (INIS)
Zhang, Guoxu; Xie, Heng
2017-01-01
Highlights: • Two IPWR Relap5 models with different PSS design were developed. • Postulated SBO and SBLOCA were analyzed. • PRHRS in primary PSS design showed stable performance under different scenarios. • Secondary PRHRS design faced flow instability. - Abstract: In the present work, the similarities and differences of representative IPWRs (integral pressurized water reactor) are studied, and two typical reactor design schemes are summarized. To get a comprehensive understanding of their transient characteristics, SBO (station blackout) and SBLOCA (small break LOCA) are simulated and analyzed respectively by using Relap5/Mod3.2. The calculation results show that, both designs are effective in keeping reactor safe. However, the transient features of the two designs show significant differences. In the primary side passive safety system (PSS) connection design, PRHRS (passive residual heat removal system) shows a roughly congruent performance in removing residual heat under various accidents. While in secondary side PSS connection design, the capability of PRHRS is closely related to primary coolant circulation condition. In SBLOCA analysis, different design approach shows different primary coolant water inventory change trend. And primary PSS connection design could potentially keep reactor core well covered for a longer time.
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.
Analysis on the influence of the pump start transient performance with different inertia impeller
International Nuclear Information System (INIS)
Tang, Y; Cheng, J; Liu, E H; Tang, L D
2012-01-01
Centrifugal pump start-up time is very short, in the boot process, the instantaneous head and flow will have an impact role to the pipeline, and however the moment of inertia is one of the main factors affecting centrifugal pump boot acceleration. We analyzed the pump start-up transient characteristics with the different moment of inertia of the impeller corresponding to the different materials, there are three different moment of inertia of the impeller have been selected. At first, we use the 'Flowmaster' fluid system simulation software do the outer characteristics simulation to the selected-model, get the time - flow and the time - speed curve. Then, do the experiments research in the process when pump start-up, and compare with the simulation result. At last use the outer characteristics simulation result as the boundary, using the ANASYS CFX software do the transient simulation to the three groups with different inertia pump impeller, and draw the pressure distribution picture. In according to the analysis, we can confirm that the impact of inertia is one of the factors in the stability during the pump star, and we can get that the greater moment of inertia, the longer the boot stable. We also can get that combined Flowmaster with ANSYS can solved engineering practice problem in fluid system conveniently, and take it easy to solve the similar problem.
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.
International Nuclear Information System (INIS)
Barbet, N.; Dumas, M.; Mihelich, G.; Souchet, Y.; Thomas, J.B.
1987-04-01
Two developments of expert systems intended to work on line to the analysis of nuclear reactor transients are reported. During an hypothetical crisis occurring in a nuclear facility, a staff of the Institute for Protection and Nuclear Safety (IPSN) has to assess the risk to local population. The expert system is intended to work as an assistant to the staff. At the present time, it deals with the availability of the safety systems of the plant (e.g. ECCS), depending on the functional state of the support systems. A next step is to take into account the physical transient of the reactor (mass and energy balance, pressure, flows). In order to reach this goal as in the development of other similar expert systems, a physical analyser is required. This is the aim of SEXTANT, which combines several knowledge bases concerning measurements, models and qualitative behaviour of the plant with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. Both expert systems require powerful shells for their development. SPIRAL is such a toolkit for the development of expert systems devoted to the computer aided management of complex processes
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
SPM analysis and cognitive dysfunctions in patients with transient global amnesia
International Nuclear Information System (INIS)
Jeong, Young Jin; Kang, Do Young; Yun, Go Un; Park, Kyung Won; Kim, Jae Woo
2004-01-01
Transient global amnesia (TGA) is known as a disease of benign nature characterized with clinically transient global antegrade amnesia and a variable degree of global retrograde memory impairment, but it usually resolved within 24 hours. The aims of this study are to assess the alterations in regional cerebral blood flow (rCBF) by Tc-99m HMPAO SPECT imaging with statistical parametric mapping (SPM) analysis and to verify the cognitive deficits by neuropsychological test in TGA patients. Twelve patients with TGA and age-matched normal control subjects participated in this study. Tc-99m HMPAO SPECT was performed within 1 to 19 days (mean duration: 7.3:±5.2 days) after the events to measure the rCBF. SPECT images were analyzed using SPM (SPM99) with Matlab 5.3. Seoul Neuropsychological Screening Battery test was also done within 2 to 8 days (mean duration 3.8±2.2 days) for cognitive functions in 8 of 12 patients with TGA. The SPM analysis of SPECT images showed significantly decreased rCBF in the left inferior frontal gyrus (Brodmann area 9), the left supramarginal gyrus (Brodmann area 40), the left postcentral gyrus (Brodmann area 40) and the left precentral gyrus (Brodmann area 4) in patients with TGA (uncorrected p<0.01). Neuropsychological test findings represented that several cognitive functions. such as, verbal memory, visual memory, phonemic fluency and confrontational naming, were impaired in patients with TGA compared with normal control. Additionally, on SPM analysis, we found lesions of hyperperfusion in contralateral cerebral hemisphere. Our study shows perfusion deficits in the left cerebral hemisphere in patients with TGA and several cognitive dysfunctions. And we found after clinical symptoms were completely resolved, the lesions of hypoperfusion were still remained. We found that functional quantitative neuroimaging study and neuropsychological test are useful to understand underlying pathomachanism of TGA
Issues regarding transient analysis examined by the Sizewell B Public Inquiry
International Nuclear Information System (INIS)
Farmer, P.R.; Dunnicliffe, C.J.
1988-01-01
Issues on PWR safety transient analysis that were discussed at the Sizewell B Public Inquiry are presented. The Public Inquiry was set up by the UK Government under an Inspector, Sir Frank Layfield, to examine all aspects of the construction, safety and operation of a 1200 MW(e) PWR on the Sizewell site. The terms of reference were broad ranging, and the constitution of the Inquiry was to make a recommendation under three Acts of Parliament which apply to the construction and operation of nuclear electrical plant. The Inquiry also covered local planning aspects, which are the responsibility of the Local Authority - in this case the Suffolk County Council. The Inspector examined and made recommendations on the safety of the Station, but consideration by Public Inquiry is outside the formal safety and licensing process, which is the business of the Utility (the CEGB) and the Nuclear Installations Inspectorate (the NII). The paper therefore takes a broader look at the question of safety, dealing with the licensing process, the requirements of the safety case and the forward strategies adopted by the CEGB in terms of research and development. This is considered for transient analysis, and the aim is to set the discussions and conclusions of the Public Inquiry into their proper context with regard to nuclear safety in the UK. The Inquiry went into some depth on the topic of LOCA, as an example of safety analysis. In the summary of the evidence and cross-examination the Inspector accepted the adequacy of the LOCA safety case without major reservations, and was satisfied further work in progress would resolve any residual criticisms. In particular support was given for the CEGB commitment to the development and use of more physically realistic calculational methods
The transient analysis of single turbine control valve closure for Lungmen ABWR
International Nuclear Information System (INIS)
Ma Shaoshih; Yuann Yngruey; Shih Chunkuan
2012-01-01
Highlights: ► The LRM was used to evaluate the single control valve closure event. ► The purpose is to offer an updated analysis about the MCFL under the partial arc mode instead of FSAR’s result. ► It is concluded that the 112% MCFL setting is the most limiting case. ► The MCFL setting actually used in SBPCS must be kept between 112% to 114% to gain the operational margin. ► The HFF index defined by the normalized heat flux can be used to predict the CPR change. - Abstract: The single control valve closure in fast (SCVCF) event is the most limiting transient in terms of delta critical power ratio (ΔCPR) for the Lungmen Plant, which is a basis to determine the operating limit minimum critical power ratio value. The partial arc mode is adopted in Lungmen Plant to control the position of the turbine control valve. However, the transient analyses presented in the Lungmen Final Safety Analysis Report (FSAR) assume that the TCVs are in the full arc mode. In this study, the Lungmen RETRAM model with partial arc mode is used to analyze the SCVCF event to offer more realistic results than the FSAR. It is concluded that the most limiting maximum combined flow limiter (MCFL) setting in RETRAN analysis is different from that of FSAR. An optimum operating range for the MCFL is suggested to gain the margin against the operating drift. Additionally, a Heat Flux Factor index is defined to appropriately determine the ranking of these cases in terms of ΔCPR.
SPM analysis and cognitive dysfunctions in patients with transient global amnesia
Energy Technology Data Exchange (ETDEWEB)
Jeong, Young Jin; Kang, Do Young; Yun, Go Un; Park, Kyung Won; Kim, Jae Woo [School of Medicine, Donga University, Busan (Korea, Republic of)
2004-07-01
Transient global amnesia (TGA) is known as a disease of benign nature characterized with clinically transient global antegrade amnesia and a variable degree of global retrograde memory impairment, but it usually resolved within 24 hours. The aims of this study are to assess the alterations in regional cerebral blood flow (rCBF) by Tc-99m HMPAO SPECT imaging with statistical parametric mapping (SPM) analysis and to verify the cognitive deficits by neuropsychological test in TGA patients. Twelve patients with TGA and age-matched normal control subjects participated in this study. Tc-99m HMPAO SPECT was performed within 1 to 19 days (mean duration: 7.3:{+-}5.2 days) after the events to measure the rCBF. SPECT images were analyzed using SPM (SPM99) with Matlab 5.3. Seoul Neuropsychological Screening Battery test was also done within 2 to 8 days (mean duration 3.8{+-}2.2 days) for cognitive functions in 8 of 12 patients with TGA. The SPM analysis of SPECT images showed significantly decreased rCBF in the left inferior frontal gyrus (Brodmann area 9), the left supramarginal gyrus (Brodmann area 40), the left postcentral gyrus (Brodmann area 40) and the left precentral gyrus (Brodmann area 4) in patients with TGA (uncorrected p<0.01). Neuropsychological test findings represented that several cognitive functions. such as, verbal memory, visual memory, phonemic fluency and confrontational naming, were impaired in patients with TGA compared with normal control. Additionally, on SPM analysis, we found lesions of hyperperfusion in contralateral cerebral hemisphere. Our study shows perfusion deficits in the left cerebral hemisphere in patients with TGA and several cognitive dysfunctions. And we found after clinical symptoms were completely resolved, the lesions of hypoperfusion were still remained. We found that functional quantitative neuroimaging study and neuropsychological test are useful to understand underlying pathomachanism of TGA.
International Nuclear Information System (INIS)
Peterson, C.E.; Gose, G.C.; McFadden, J.H.
1983-01-01
RETRAN-02 represents a significant achievement in the development of a versatile and reliable computer program for use in best estimate transient thermal-hydraulic analysis of light water reactor systems. The RETRAN-02 computer program is an extension of the RETRAN-01 program designed to provide analysis capabilities for 1) BWR and PWR transients, 2) small break loss of coolant accidents, 3) balance of plant modeling, and 4) anticipated transients without scram, while maintaining the analysis capabilities of the predecessor code. The RETRAN-02 computer code is constructed in a semimodular and dynamic dimensioned form where additions to the code can be easily carried out as new and improved models are developed. This report (the fourth of a five volume computer code manual) describes the verification and validation of RETRAN-02
Rate transient analysis for homogeneous and heterogeneous gas reservoirs using the TDS technique
International Nuclear Information System (INIS)
Escobar, Freddy Humberto; Sanchez, Jairo Andres; Cantillo, Jose Humberto
2008-01-01
In this study pressure test analysis in wells flowing under constant wellbore flowing pressure for homogeneous and naturally fractured gas reservoir using the TDS technique is introduced. Although, constant rate production is assumed in the development of the conventional well test analysis methods, constant pressure production conditions are sometimes used in the oil and gas industry. The constant pressure technique or rate transient analysis is more popular reckoned as decline curve analysis under which rate is allows to decline instead of wellbore pressure. The TDS technique, everyday more used even in the most recognized software packages although without using its trade brand name, uses the log-log plot to analyze pressure and pressure derivative test data to identify unique features from which exact analytical expression are derived to easily estimate reservoir and well parameters. For this case, the fingerprint characteristics from the log-log plot of the reciprocal rate and reciprocal rate derivative were employed to obtain the analytical expressions used for the interpretation analysis. Many simulation experiments demonstrate the accuracy of the new method. Synthetic examples are shown to verify the effectiveness of the proposed methodology
Evaluation of time integration methods for transient response analysis of nonlinear structures
International Nuclear Information System (INIS)
Park, K.C.
1975-01-01
Recent developments in the evaluation of direct time integration methods for the transient response analysis of nonlinear structures are presented. These developments, which are based on local stability considerations of an integrator, show that the interaction between temporal step size and nonlinearities of structural systems has a pronounced effect on both accuracy and stability of a given time integration method. The resulting evaluation technique is applied to a model nonlinear problem, in order to: 1) demonstrate that it eliminates the present costly process of evaluating time integrator for nonlinear structural systems via extensive numerical experiments; 2) identify the desirable characteristics of time integration methods for nonlinear structural problems; 3) develop improved stiffly-stable methods for application to nonlinear structures. Extension of the methodology for examination of the interaction between a time integrator and the approximate treatment of nonlinearities (such as due to pseudo-force or incremental solution procedures) is also discussed. (Auth.)
International Nuclear Information System (INIS)
Yulianti, Yanti; Su'ud, Zaki; Waris, Abdul; Khotimah, S. N.; Shafii, M. Ali
2010-01-01
The research about fast transient and spatially non-homogenous nuclear reactor accident analysis of two-dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space-time diffusion equation is solved by using direct methods which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference discretization method is solved by using iterative methods ADI (Alternating Direct Implicit). The indication of accident is decreasing macroscopic absorption cross-section that results large external reactivity. The power reactor has a peak value before reactor has new balance condition. Changing of temperature reactor produce a negative Doppler feedback reactivity. The reactivity will reduce excess positive reactivity. Temperature reactor during accident is still in below fuel melting point which is in secure condition.
Development of Transient-Reactor Analysis Code (TRAC) for real-time applications
International Nuclear Information System (INIS)
Niederauer, G.F.; Giguere, P.T.; Lime, J.F.; Knight, T.D.; Ashy, O.; Fakory, R.
1997-01-01
This is the final report of a six-month, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). Nuclear-plant training simulators employ simplified one-dimensional thermal-hydraulics codes because of the demands to run in real time and with limited computing power. The objective of this project was to investigate the feasibility of using the advanced Transient-Reactor Analysis Code (TRAC) in a simulator to increase the fidelity of a simulator. Many issues need to be addressed to take such a complex code from a batch engineering environment to a real-time environment. Working with simulator vendor, GSE, the authors investigated the technical issues relating to integrating TRAC into a real-time environment. They also modified a nuclear power plant model for simulator purposes and investigated its performance in real time
International Nuclear Information System (INIS)
Zeuch, W.R.; Wang, C.Y.
1985-01-01
This paper presents some of the current capabilities of the three-dimensional piping code SHAPS and demonstrates their usefulness in handling analyses encountered in typical LMFBR studies. Several examples demonstrate the utility of the SHAPS code for problems involving fluid-structure interactions and seismic-related events occurring in three-dimensional piping networks. Results of two studies of pressure wave propagation demonstrate the dynamic coupling of pipes and elbows producing global motion and rigorous treatment of physical quantities such as changes in density, pressure, and strain energy. Results of the seismic analysis demonstrate the capability of SHAPS to handle dynamic structural response within a piping network over an extended transient period of several seconds. Variation in dominant stress frequencies and global translational frequencies were easily handled with the code. 4 refs., 10 figs
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.)
Transient analysis of printed lines using finite-difference time-domain method
Energy Technology Data Exchange (ETDEWEB)
Ahmed, Shahid [Thomas Jefferson National Accelerator Facility, 12050 Jefferson Avenue, Suite 704, Newport News, VA, 23606, USA
2012-03-29
Comprehensive studies of ultra-wideband pulses and electromagnetic coupling on printed coupled lines have been performed using full-wave 3D finite-difference time-domain analysis. Effects of unequal phase velocities of coupled modes, coupling between line traces, and the frequency dispersion on the waveform fidelity and crosstalk have been investigated in detail. To discriminate the contributions of different mechanisms into pulse evolution, single and coupled microstrip lines without (ϵ_{r} = 1) and with (ϵ_{r} > 1) dielectric substrates have been examined. To consistently compare the performance of the coupled lines with substrates of different permittivities and transients of different characteristic times, a generic metric similar to the electrical wavelength has been introduced. The features of pulse propagation on coupled lines with layered and pedestal substrates and on the irregular traces have been explored. Finally, physical interpretations of the simulation results are discussed in the paper.
Ren, Zhengyi; Huang, Tong; Feng, Jiajia; Zhou, Yuanwei
2018-05-01
In this paper, a 600Wh vertical maglev energy storage flywheel rotor system is taken as a model. The motion equation of a rigid rotor considering the gyroscopic effect and the center of mass offset is obtained by the centroid theorem, and the experimental verification is carried out. Using the state variable method, the Matlab software was used to program and simulate the radial displacement and radial electromagnetic force of the rotor system at each speed. The results show that the established system model is in accordance with the designed 600Wh vertical maglev energy storage flywheel model. The results of the simulation analysis are helpful to further understand the dynamic nature of the flywheel rotor at different transient speeds.
Computational analysis of transient gas release from a high pressure vessel
Energy Technology Data Exchange (ETDEWEB)
Pedro, G.; Oshkai, P.; Djilali, N. [Victoria Univ., BC (Canada). Inst. for Integrated Energy Systems; Penau, F. [CERAM Euro-American Inst. of Technology, Sophia Antipolis (France)
2006-07-01
Gas jets exiting from compressed vessels can undergo several regimes as the pressure in the vessel decreases, and a greater understanding of the characteristics of gas jets is needed to determine safety requirements in the transport, distribution, and use of hydrogen. This paper provided a study of the bow shock waves that typically occur during the initial stage of a gas jet incident. The transient behaviour of an initiated jet was investigated using unsteady, compressible flow simulations. The gas was considered to be ideal, and the domain was considered to be axisymmetric. Tank pressure for the analysis was set at a value of 100 atm. Jet structure was examined, as well as the shock structures and separation due to adverse pressure gradients at the nozzle. Shock structure displacement was also characterized.
Three-dimensional inverse transient heat transfer analysis of thick functionally graded plates
Energy Technology Data Exchange (ETDEWEB)
Haghighi, M.R. Golbahar; Malekzadeh, P. [Department of Mechanical Engineering, School of Engineering, Persian Gulf University, Bushehr 75168 (Iran); Eghtesad, M. [Department of Mechanical Engineering, School of Engineering, Shiraz University, Shiraz 71348-51154 (Iran); Necsulescu, D.S. [Department of Mechanical Engineering, Faculty of Engineering, University of Ottawa, Ottawa, Ontario (Canada)
2009-03-15
In this paper, a three-dimensional transient inverse heat conduction (IHC) procedure is presented to estimate the unknown boundary heat flux of thick functionally graded (FG) plates. For this purpose, the conjugate gradient method (CGM) in conjunction with adjoint problem is used. A recently developed three-dimensional efficient hybrid method is employed to solve variable-coefficient initial-boundary-value differential equations of direct problem as a part of the inverse solution. The accuracy of the inverse analysis is examined by simulating the exact and noisy data for problems with different types of boundary conditions and material properties. In addition to rectangular domain, skew plates are considered. The results obtained show good accuracy for the estimation of boundary heat fluxes. (author)
Energy Technology Data Exchange (ETDEWEB)
Golbahar Haghighi, M.R.; Eghtesad, M. [Department of Mechanical Engineering, School of Engineering, Shiraz University, Shiraz 71348-51154 (Iran, Islamic Republic of); Malekzadeh, P. [Department of Mechanical Engineering, School of Engineering, Persian Gulf University, Boushehr 75169-13798 (Iran, Islamic Republic of)], E-mail: malekzadeh@pgu.ac.ir
2008-05-15
In this paper, a mixed finite element (FE) and differential quadrature (DQ) method as a simple, accurate and computationally efficient numerical tool for two dimensional transient heat transfer analysis of functionally graded materials (FGMs) is developed. The method benefits from the high accuracy, fast convergence behavior and low computational efforts of the DQ in conjunction with the advantages of the FE method in general geometry, loading and systematic boundary treatment. Also, the boundary conditions at the top and bottom surfaces of the domain can be implemented more precisely and in strong form. The temporal derivatives are discretized using an incremental DQ method (IDQM), whose numerical stability is not sensitive to time step size. The effects of non-uniform convective-radiative conditions on the boundaries are investigated. The accuracy of the proposed method is demonstrated by comparing its results with those available in the literature. It is shown that using few grid points, highly accurate results can be obtained.
Analysis of a station blackout transient at the Kori units 3/4
International Nuclear Information System (INIS)
Bang, Young Seok; Kim, Hho Jung
1992-01-01
A transient analysis of station blackout accident is performed to evaluate the plant specific capability to cope with the accident at the Kori Units 3/4. The RELAP5/MOD3/5m5 code and full three loop modelling scheme are used in the calculation. The leak flow from reactor coolant system due to a failure of reactor coolant pump seal following the accident is assumed to be 25 gpm and the turbine driven aux feedwater unavailable. As a result, it is found that no core uncovery occurs in the plant until 7100 sec following a station blackout, the steam generator has a decay heat removal capability until 3100 sec, and the natural circulation over the reactor coolant loop until the complete loop seal voiding are observed. And the Nuclear Plant Analyzer is used and found to be effective in improving the phenomenological understanding on the accident
Transient analysis of an HTS DC power cable with an HVDC system
Dinh, Minh-Chau; Ju, Chang-Hyeon; Kim, Jin-Geun; Park, Minwon; Yu, In-Keun; Yang, Byeongmo
2013-11-01
The operational characteristics of a superconducting DC power cable connected to a highvoltage direct current (HVDC) system are mainly concerned with the HVDC control and protection system. To confirm how the cable operates with the HVDC system, verifications using simulation tools are needed. This paper presents a transient analysis of a high temperature superconducting (HTS) DC power cable in connection with an HVDC system. The study was conducted via the simulation of the HVDC system and a developed model of the HTS DC power cable using a real time digital simulator (RTDS). The simulation was performed with some cases of short circuits that may have caused system damage. The simulation results show that during the faults, the quench did not happen with the HTS DC power cable because the HVDC controller reduced some degree of the fault current. These results could provide useful data for the protection design of a practical HVDC and HTS DC power cable system.
Energy Technology Data Exchange (ETDEWEB)
Vinai, P
2007-10-15
For the development, design and licensing of a nuclear power plant (NPP), a sound safety analysis is necessary to study the diverse physical phenomena involved in the system behaviour under operational and transient conditions. Such studies are based on detailed computer simulations. With the progresses achieved in computer technology and the greater availability of experimental and plant data, the use of best estimate codes for safety evaluations has gained increasing acceptance. The application of best estimate safety analysis has raised new problems that need to be addressed: it has become more crucial to assess as to how reliable code predictions are, especially when they need to be compared against safety limits that must not be crossed. It becomes necessary to identify and quantify the various possible sources of uncertainty that affect the reliability of the results. Currently, such uncertainty evaluations are generally based on experts' opinion. In the present research, a novel methodology based on a non-parametric statistical approach has been developed for objective quantification of best-estimate code uncertainties related to the physical models used in the code. The basis is an evaluation of the accuracy of a given physical model achieved by comparing its predictions with experimental data from an appropriate set of separate-effect tests. The differences between measurements and predictions can be considered stochastically distributed, and thus a statistical approach can be employed. The first step was the development of a procedure for investigating the dependence of a given physical model's accuracy on the experimental conditions. Each separate-effect test effectively provides a random sample of discrepancies between measurements and predictions, corresponding to a location in the state space defined by a certain number of independent system variables. As a consequence, the samples of 'errors', achieved from analysis of the entire
International Nuclear Information System (INIS)
Vinai, P.
2007-10-01
For the development, design and licensing of a nuclear power plant (NPP), a sound safety analysis is necessary to study the diverse physical phenomena involved in the system behaviour under operational and transient conditions. Such studies are based on detailed computer simulations. With the progresses achieved in computer technology and the greater availability of experimental and plant data, the use of best estimate codes for safety evaluations has gained increasing acceptance. The application of best estimate safety analysis has raised new problems that need to be addressed: it has become more crucial to assess as to how reliable code predictions are, especially when they need to be compared against safety limits that must not be crossed. It becomes necessary to identify and quantify the various possible sources of uncertainty that affect the reliability of the results. Currently, such uncertainty evaluations are generally based on experts' opinion. In the present research, a novel methodology based on a non-parametric statistical approach has been developed for objective quantification of best-estimate code uncertainties related to the physical models used in the code. The basis is an evaluation of the accuracy of a given physical model achieved by comparing its predictions with experimental data from an appropriate set of separate-effect tests. The differences between measurements and predictions can be considered stochastically distributed, and thus a statistical approach can be employed. The first step was the development of a procedure for investigating the dependence of a given physical model's accuracy on the experimental conditions. Each separate-effect test effectively provides a random sample of discrepancies between measurements and predictions, corresponding to a location in the state space defined by a certain number of independent system variables. As a consequence, the samples of 'errors', achieved from analysis of the entire database, are
DEFF Research Database (Denmark)
Dorrell, David G.; Hermann, Alexander Niels August; Jensen, Bogi Bech
2013-01-01
eccentricity. The operating conditions are varied so that transient, motoring and doubly-fed induction generator modes are studied. This allows greater understanding of the radial forces involved. Wound rotor induction machines exhibit higher unbalanced magnetic pull than cage induction machines so......There has been much literature on unbalanced magnetic pull in various types of electrical machine. This can lead to bearing wear and additional vibrations in the machine. In this paper a wound rotor induction is studied. Finite element analysis studies are conducted when the rotor has 10 % rotor...
Benchmarking of Modern Data Analysis Tools for a 2nd generation Transient Data Analysis Framework
Goncalves, Nuno
2016-01-01
During the past year of operating the Large Hadron Collider (LHC), the amount of transient accelerator data to be persisted and analysed has been steadily growing. Since the startup of the LHC in 2006, the amount of weekly data storage requirements exceeded what the systems was initially designed to accommodate in a full year of operation. Moreover, it is predicted that the data acquisition rates will continue to increase in the future, due to foreseen improvements in the infrastructure within the scope of the High Luminosity LHC project. Despite the efforts for improving and optimizing the current data storage infrastructures (CERN Accelerator Logging Service and Post Mortem database), some limitations still persist and require a different approach to scale up efficiently to provide efficient services for future machine upgrades. This project aims to explore one of the possibilities among novel solutions proposed to solve the problem of working with large datasets. The configuration is composed of Spark for ...
TRAC-PF1 analysis of LOFT steam-generator feedwater transient test L9-1
International Nuclear Information System (INIS)
Meier, J.K.
1983-01-01
The Transient Reactor Analysis Code (TRAC-PF1) calculations were compared to test data from Loss-of-Fluid Test (LOFT) L9-1, which was a loss-of-feedwater transient. This paper includes descriptions of the test and the TRAC input and compares the TRAC-calculated results with the test data. We conclude that the code predicted the experiment well, given the uncertainties in the boundary conditions. The analysis indicates the need to model all the flow paths and heat structures, and to improve the TRAC wall condensation heat-transfer model
International Nuclear Information System (INIS)
1976-05-01
Results are presented of analyses of the transient thermal-hydraulic conditions and radiological release consequences which would occur in power plants which employ a Combustion Engineering Nuclear Steam Supply System during Anticipated Transients Without Scram due to a lack of insertion of the Control Element Assemblies upon signals for automatic or manual reactor shutdown. The transients analyzed include all events which meet the criterion to be considered as anticipated at least once in the plant lifetime with automatic reactor shutdown
International Nuclear Information System (INIS)
Gomes, Daniel S.; Teixeira, Antonio S.
2017-01-01
Although regulatory agencies have shown a special interest in incorporating best estimate approaches in the fuel licensing process, fuel codes are currently licensed based on only the deterministic limits such as those seen in 10CRF50, and therefore, may yield unrealistic safety margins. The concept of uncertainty analysis is employed to more realistically manage this risk. In this study, uncertainties were classified into two categories: probabilistic and epistemic (owing to a lack of pre-existing knowledge in this area). Fuel rods have three sources of uncertainty: manufacturing tolerance, boundary conditions, and physical models. The first step in successfully analyzing the uncertainties involves performing a statistical analysis on the input parameters used throughout the fuel code. The response obtained from this analysis must show proportional index correlations because the uncertainties are globally propagated. The Dakota toolkit was used to analyze the FRAPTRAN transient fuel code. The subsequent sensitivity analyses helped in identifying the key parameters with the highest correlation indices including the peak cladding temperature and the time required for cladding failures. The uncertainty analysis was performed using an IFA-650-5 fuel rod, which was in line with the tests performed in the Halden Project in Norway. The main objectives of the Halden project included studying the ballooning and rupture processes. The results of this experiment demonstrate the accuracy and applicability of the physical models in evaluating the thermal conductivity, mechanical model, and fuel swelling formulations. (author)
Transient analysis of cutoff waveguide antenna in three-dimensional space
International Nuclear Information System (INIS)
Kashiwa, Tatsuya; Yoshida, Norinobu; Fukai, Ichiro
1986-01-01
Recently, the exciting system for electric power heating as seen in nuclear fusion plasma heating and medical purpose has been actively studied and developed. Since such system treats basically a neighborhood field, various problems unlike conventional exciting system for communication arise. In such situation, the structure having the waveguides of simple and robust construction as the main body has been proposed. In this exciting system including the condition of media, the complex distribution of a neighborhood field based on a three-dimensional structure exerts an important effect on the characteristics. Especially in large power excitation, the higher mode of relatively small power distribution cannot be neglected. Besides, also a transient field distribution exerts an important effect on the characteristics, and the time response analysis is required. In this analysis, by the three-dimensional time response analysis method using Bergeron method, the unified analysis of the total system comprising a cutoff waveguide, a coaxial exciting part and a heating region was carried out for determining a radiation neighborhood electromagnetic field by a cutoff waveguide antenna. (Kako, I.)
Energy Technology Data Exchange (ETDEWEB)
Gomes, Daniel S.; Teixeira, Antonio S., E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)
2017-07-01
Although regulatory agencies have shown a special interest in incorporating best estimate approaches in the fuel licensing process, fuel codes are currently licensed based on only the deterministic limits such as those seen in 10CRF50, and therefore, may yield unrealistic safety margins. The concept of uncertainty analysis is employed to more realistically manage this risk. In this study, uncertainties were classified into two categories: probabilistic and epistemic (owing to a lack of pre-existing knowledge in this area). Fuel rods have three sources of uncertainty: manufacturing tolerance, boundary conditions, and physical models. The first step in successfully analyzing the uncertainties involves performing a statistical analysis on the input parameters used throughout the fuel code. The response obtained from this analysis must show proportional index correlations because the uncertainties are globally propagated. The Dakota toolkit was used to analyze the FRAPTRAN transient fuel code. The subsequent sensitivity analyses helped in identifying the key parameters with the highest correlation indices including the peak cladding temperature and the time required for cladding failures. The uncertainty analysis was performed using an IFA-650-5 fuel rod, which was in line with the tests performed in the Halden Project in Norway. The main objectives of the Halden project included studying the ballooning and rupture processes. The results of this experiment demonstrate the accuracy and applicability of the physical models in evaluating the thermal conductivity, mechanical model, and fuel swelling formulations. (author)
Oxide fuel pin transient performance analysis and design with the TEMECH code
International Nuclear Information System (INIS)
Bard, F.E.; Dutt, S.P.; Hinman, C.A.; Hunter, C.W.; Pitner, A.L.
1986-01-01
The TEMECH code is a fast-running, thermal-mechanical-hydraulic, analytical program used to evaluate the transient performance of LMR oxide fuel pins. The code calculates pin deformation and failure probability due to fuel-cladding differential thermal expansion, expansion of fuel upon melting, and fission gas pressurization. The mechanistic fuel model in the code accounts for fuel cracking, crack closure, porosity decrease, and the temperature dependence of fuel creep through the course of the transient. Modeling emphasis has been placed on results obtained from Fuel Cladding Transient Test (FCTT) testing, Transient Fuel Deformation (TFD) tests and TREAT integral fuel pin experiments
Development of Input/Output System for the Reactor Transient Analysis System (RETAS)
International Nuclear Information System (INIS)
Suh, Jae Seung; Kang, Doo Hyuk; Cho, Yeon Sik; Ahn, Seung Hoon; Cho, Yong Jin
2009-01-01
A Korea Institute of Nuclear Safety Reactor Transient Analysis System (KINS-RETAS) aims at providing a realistic prediction of core and RCS response to the potential or actual event scenarios in Korean nuclear power plants (NPPs). A thermal hydraulic system code MARS is a pivot code of the RETAS, and used to predict thermal hydraulic (TH) behaviors in the core and associated systems. MARS alone can be applied to many types of transients, but is sometimes coupled with the other codes developed for different objectives. Many tools have been developed to aid users in preparing input and displaying the transient information and output data. Output file and Graphical User Interfaces (GUI) that help prepare input decks, as seen in SNAP (Gitnick, 1998), VISA (K.D. Kim, 2007) and display aids include the eFAST (KINS, 2007). The tools listed above are graphical interfaces. The input deck builders allow the user to create a functional diagram of the plant, pictorially on the screen. The functional diagram, when annotated with control volume and junction numbers, is a nodalization diagram. Data required for an input deck is entered for volumes and junctions through a mouse-driven menu and pop-up dialog; after the information is complete, an input deck is generated. Display GUIs show data from MARS calculations, either during or after the transient. The RETAS requires the user to first generate a set of 'input', two dimensional pictures of the plant on which some of the data is displayed either numerically or with a color map. The RETAS can generate XY-plots of the data. Time histories of plant conditions can be seen via the plots or through the RETAS's replay mode. The user input was combined with design input from MARS developers and experts from both the GUI and ergonomics fields. A partial list of capabilities follows. - 3D display for neutronics. - Easier method (less user time and effort) to generate 'input' for the 3D displays. - Detailed view of data at volume or
Development of Input/Output System for the Reactor Transient Analysis System (RETAS)
Energy Technology Data Exchange (ETDEWEB)
Suh, Jae Seung; Kang, Doo Hyuk; Cho, Yeon Sik [ENESYS, Daejeon (Korea, Republic of); Ahn, Seung Hoon; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2009-05-15
A Korea Institute of Nuclear Safety Reactor Transient Analysis System (KINS-RETAS) aims at providing a realistic prediction of core and RCS response to the potential or actual event scenarios in Korean nuclear power plants (NPPs). A thermal hydraulic system code MARS is a pivot code of the RETAS, and used to predict thermal hydraulic (TH) behaviors in the core and associated systems. MARS alone can be applied to many types of transients, but is sometimes coupled with the other codes developed for different objectives. Many tools have been developed to aid users in preparing input and displaying the transient information and output data. Output file and Graphical User Interfaces (GUI) that help prepare input decks, as seen in SNAP (Gitnick, 1998), VISA (K.D. Kim, 2007) and display aids include the eFAST (KINS, 2007). The tools listed above are graphical interfaces. The input deck builders allow the user to create a functional diagram of the plant, pictorially on the screen. The functional diagram, when annotated with control volume and junction numbers, is a nodalization diagram. Data required for an input deck is entered for volumes and junctions through a mouse-driven menu and pop-up dialog; after the information is complete, an input deck is generated. Display GUIs show data from MARS calculations, either during or after the transient. The RETAS requires the user to first generate a set of 'input', two dimensional pictures of the plant on which some of the data is displayed either numerically or with a color map. The RETAS can generate XY-plots of the data. Time histories of plant conditions can be seen via the plots or through the RETAS's replay mode. The user input was combined with design input from MARS developers and experts from both the GUI and ergonomics fields. A partial list of capabilities follows. - 3D display for neutronics. - Easier method (less user time and effort) to generate 'input' for the 3D displays. - Detailed view
MINET, Transient Fluid Flow and Heat Transfer Power Plant Network Analysis
International Nuclear Information System (INIS)
Van Tuyle, G.J.
2002-01-01
1 - Description of program or function: MINET (Momentum Integral Network) was developed for the transient analysis of intricate fluid flow and heat transfer networks, such as those found in the balance of plant in power generating facilities. It can be utilized as a stand-alone program or interfaced to another computer program for concurrent analysis. Through such coupling, a computer code limited by either the lack of required component models or large computational needs can be extended to more fully represent the thermal hydraulic system thereby reducing the need for estimating essential transient boundary conditions. The MINET representation of a system is one or more networks of volumes, segments, and boundaries linked together via heat exchangers only, i.e., heat can transfer between networks, but fluids cannot. Volumes are used to represent tanks or other volume components, as well as locations in the system where significant flow divisions or combinations occur. Segments are composed of one or more pipes, pumps, heat exchangers, turbines, and/or valves each represented by one or more nodes. Boundaries are simply points where the network interfaces with the user or another computer code. Several fluids can be simulated, including water, sodium, NaK, and air. 2 - Method of solution: MINET is based on a momentum integral network method. Calculations are performed at two levels, the network level (volumes) and the segment level. Equations conserving mass and energy are used to calculate pressure and enthalpy within volumes. An integral momentum equation is used to calculate the segment average flow rate. In-segment distributions of mass flow rate and enthalpy are calculated using local equations of mass and energy. The segment pressure is taken to be the linear average of the pressure at both ends. This method uses a two-plus equation representation of the thermal hydraulic behavior of a system of heat exchangers, pumps, pipes, valves, tanks, etc. With the
Heide, P A W
2002-01-01
Secondary ion mass spectrometry (SIMS) depth profile analysis of Si wafers using 1 keV Cs sup + primary ions at large incidence angles (80 deg. ) is plagued by unusually strong transient effects (variations in both sputter and ion yields). Analysis of a native oxide terminated Si wafer with and without the aid of an O sub 2 leak, and an Ar sup + pre-sputtered wafer revealed correlations between the implanted Cs content and various secondary ion intensities consistent with that expected from a resonance charge transfer process (that assumed by the electron tunneling model). Cs concentrations were defined through X-ray photoelectron spectroscopy of the sputtered surface from SIMS profiles terminated within the transient region. These scaled with the surface roughening occurring under these conditions and can be explained as resulting from the associated drop in sputter rates. An O induced transient effect from the native oxide was also identified. Characterization of these effects allowed the reconstruction of ...
Greensmith, David J.
2014-01-01
Here I present an Excel based program for the analysis of intracellular Ca transients recorded using fluorescent indicators. The program can perform all the necessary steps which convert recorded raw voltage changes into meaningful physiological information. The program performs two fundamental processes. (1) It can prepare the raw signal by several methods. (2) It can then be used to analyze the prepared data to provide information such as absolute intracellular Ca levels. Also, the rates of change of Ca can be measured using multiple, simultaneous regression analysis. I demonstrate that this program performs equally well as commercially available software, but has numerous advantages, namely creating a simplified, self-contained analysis workflow. PMID:24125908
Transient flow analysis of the single cylinder for the control rod hydraulic driving system
International Nuclear Information System (INIS)
Sun, Xinming; Qin, Benke; Bo, Hanliang
2017-01-01
Highlights: • The control rod hydraulic driving system(CRHDS) is a new type of built-in control rod drive technology. The hydraulic cylinder is the main component of the CRHDS. • Transient flow phenomenon in the CRHDS is studied by experiments under different working conditions. • The working mechanism of the hydraulic cylinder step motion and the key characteristic parameters are analyzed based on the experimental results. - Abstract: The control rod hydraulic driving system (CRHDS) is a new type of built-in control rod drive technology. In the CRHDS the pulse flow from the pump into the hydraulic cylinder of the control rod hydraulic drive mechanism (CRHDM) is regulated by the integrated valve to perform the step motion of the reactor control rod. Transient flow occurs in the CRHDS during control rod step motion process which is studied by experiments. The time-history curves of flow rate, pressure and inner cylinder displacement were analyzed, and the results show that the water hammer pressure peak during the step-up motion is high, while there are no obvious pressure fluctuations in the corresponding step-down motion. In the step-up process, the pressure fluctuation amplitude increases with the increase of CRHDS driving pressure. The step-up time and the pressure increasing time before step-up decreases with the driving pressure. The step-up pressure increases with the driving pressure. In the step-down process, the step-down time, the step-down pressure and the pressure decreasing time before step-down do not change with the increase of the driving pressure. The experimental results lay the base for the working principle and vibration reduction analysis of the CRHDS and it’s also helpful for improvement of the working performance of the key facilities and instruments of the CRHDS loop.
Transient Analysis and Design Improvement of a Gas Turbine Rotor Based on Thermal-Mechanical Method
Directory of Open Access Journals (Sweden)
Yang Liu
2018-01-01
Full Text Available The rotor is the core component of a gas turbine, and more than 80% of the failures in gas turbines occur in the rotor system, especially during the start-up period. Therefore, the safety assessment of the rotor during the start-up period is essential for the design of the gas turbine. In this paper, the transient equivalent stress of a gas turbine rotor under the cold start-up condition is investigated and the novel tie rod structure is introduced to reduce the equivalent stress. Firstly, a three-dimensional finite element model of the gas turbine rotor is built, and nonlinear contact behaviors such as friction are taken into account. Secondly, the convective heat transfer coefficients of the gas turbine rotor under the cold start-up condition are calculated using thermal dynamic theory. The transient analysis of the gas turbine rotor is conducted considering the thermal load, the centrifugal load, and the pretightening force. The temperature and stress distributions of the rotor under the cold start-up condition are shown in detail. In particular, the generation mechanism of maximum equivalent stress for tie rods and the change tendency of the pretightening force are illustrated in detail. The tie rod holes of the rear shaft and the turbine tie rod are the dangerous locations during the start-up period. Finally, a novel tie rod is proposed to reduce the maximum equivalent stress at the dangerous location. The maximum equivalent stress at this location is decreased by 15%. This paper provides some reference for the design of the gas turbine rotor.
Experimental analysis on the dynamic wake of an actuator disc undergoing transient loads
Yu, W.; Hong, V. W.; Ferreira, C.; van Kuik, G. A. M.
2017-10-01
The Blade Element Momentum model, which is based on the actuator disc theory, is still the model most used for the design of open rotors. Although derived from steady cases with a fully developed wake, this approach is also applied to unsteady cases, with additional engineering corrections. This work aims to study the impact of an unsteady loading on the wake of an actuator disc. The load and flow of an actuator disc are measured in the Open Jet Facility wind tunnel of Delft University of Technology, for steady and unsteady cases. The velocity and turbulence profiles are characterized in three regions: the inner wake region, the shear layer region and the region outside the wake. For unsteady load cases, the measured velocity field shows a hysteresis effect in relation to the loading, showing differences between the cases when loading is increased and loading is decreased. The flow field also shows a transient response to the step change in loading, with either an overshoot or undershoot of the velocity in relation to the steady-state velocity. In general, a smaller reduced ramp time results in a faster velocity transient, and in turn a larger amplitude of overshoot or undershoot. Time constants analysis shows that the flow reaches the new steady-state slower for load increase than for load decrease; the time constants outside the wake are generally larger than at other radial locations for a given downstream plane; the time constants of measured velocity in the wake show radial dependence.The data are relevant for the validation of numerical models for unsteady actuator discs and wind turbines, and are made available in an open source database (see Appendix).
Analysis of Spontaneous and Nerve-Evoked Calcium Transients in Intact Extraocular Muscles in Vitro
Feng, Cheng-Yuan; Hennig, Grant W.; Corrigan, Robert D.; Smith, Terence K.; von Bartheld, Christopher S.
2012-01-01
Extraocular muscles (EOMs) have unique calcium handling properties, yet little is known about the dynamics of calcium events underlying ultrafast and tonic contractions in myofibers of intact EOMs. Superior oblique EOMs of juvenile chickens were dissected with their nerve attached, maintained in oxygenated Krebs buffer, and loaded with fluo-4. Spontaneous and nerve stimulation-evoked calcium transients were recorded and, following calcium imaging, some EOMs were double-labeled with rhodamine-conjugated alpha-bungarotoxin (rhBTX) to identify EOM myofiber types. EOMs showed two main types of spontaneous calcium transients, one slow type (calcium waves with 1/2max duration of 2–12 s, velocity of 25–50 μm/s) and two fast “flash-like” types (Type 1, 30–90 ms; Type 2, 90–150 ms 1/2max duration). Single pulse nerve stimulation evoked fast calcium transients identical to the fast (Type 1) calcium transients. Calcium waves were accompanied by a local myofiber contraction that followed the calcium transient wavefront. The magnitude of calcium-wave induced myofiber contraction far exceeded those of movement induced by nerve stimulation and associated fast calcium transients. Tetrodotoxin eliminated nerve-evoked transients, but not spontaneous transients. Alpha-bungarotoxin eliminated both spontaneous and nerve-evoked fast calcium transients, but not calcium waves, and caffeine increased wave activity. Calcium waves were observed in myofibers lacking spontaneous or evoked fast transients, suggestive of multiply-innervated myofibers, and this was confirmed by double-labeling with rhBTX. We propose that the abundant spontaneous calcium transients and calcium waves with localized contractions that do not depend on innervation may contribute to intrinsic generation of tonic functions of EOMs. PMID:22579493
Identification of a cis-regulatory element by transient analysis of co-ordinately regulated genes
Directory of Open Access Journals (Sweden)
Allan Andrew C
2008-07-01
Full Text Available Abstract Background Transcription factors (TFs co-ordinately regulate target genes that are dispersed throughout the genome. This co-ordinate regulation is achieved, in part, through the interaction of transcription factors with conserved cis-regulatory motifs that are in close proximity to the target genes. While much is known about the families of transcription factors that regulate gene expression in plants, there are few well characterised cis-regulatory motifs. In Arabidopsis, over-expression of the MYB transcription factor PAP1 (PRODUCTION OF ANTHOCYANIN PIGMENT 1 leads to transgenic plants with elevated anthocyanin levels due to the co-ordinated up-regulation of genes in the anthocyanin biosynthetic pathway. In addition to the anthocyanin biosynthetic genes, there are a number of un-associated genes that also change in expression level. This may be a direct or indirect consequence of the over-expression of PAP1. Results Oligo array analysis of PAP1 over-expression Arabidopsis plants identified genes co-ordinately up-regulated in response to the elevated expression of this transcription factor. Transient assays on the promoter regions of 33 of these up-regulated genes identified eight promoter fragments that were transactivated by PAP1. Bioinformatic analysis on these promoters revealed a common cis-regulatory motif that we showed is required for PAP1 dependent transactivation. Conclusion Co-ordinated gene regulation by individual transcription factors is a complex collection of both direct and indirect effects. Transient transactivation assays provide a rapid method to identify direct target genes from indirect target genes. Bioinformatic analysis of the promoters of these direct target genes is able to locate motifs that are common to this sub-set of promoters, which is impossible to identify with the larger set of direct and indirect target genes. While this type of analysis does not prove a direct interaction between protein and DNA
AITRAC: Augmented Interactive Transient Radiation Analysis by Computer. User's information manual
International Nuclear Information System (INIS)
1977-10-01
AITRAC is a program designed for on-line, interactive, DC, and transient analysis of electronic circuits. The program solves linear and nonlinear simultaneous equations which characterize the mathematical models used to predict circuit response. The program features 100 external node--200 branch capability; conversional, free-format input language; built-in junction, FET, MOS, and switch models; sparse matrix algorithm with extended-precision H matrix and T vector calculations, for fast and accurate execution; linear transconductances: beta, GM, MU, ZM; accurate and fast radiation effects analysis; special interface for user-defined equations; selective control of multiple outputs; graphical outputs in wide and narrow formats; and on-line parameter modification capability. The user describes the problem by entering the circuit topology and part parameters. The program then automatically generates and solves the circuit equations, providing the user with printed or plotted output. The circuit topology and/or part values may then be changed by the user, and a new analysis, requested. Circuit descriptions may be saved on disk files for storage and later use. The program contains built-in standard models for resistors, voltage and current sources, capacitors, inductors including mutual couplings, switches, junction diodes and transistors, FETS, and MOS devices. Nonstandard models may be constructed from standard models or by using the special equations interface. Time functions may be described by straight-line segments or by sine, damped sine, and exponential functions. 42 figures, 1 table
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)
Fluid transient analysis and design considerations in TVA PWR feedwater systems and steam generators
International Nuclear Information System (INIS)
Kelley, B.T.
1979-01-01
TVA has evaluated a number of fluid transients in an effort to discover areas of potential problems and to improve overall unit operation. The transients recently or currently being evaluated fall into four major areas - accident analyses, fast valving, heater drain systems, and steam generators. A discussion of each area follows
International Nuclear Information System (INIS)
Singh, R.K.; Redlinger, R.; Breitung, W.
2005-09-01
Design and analysis of blast resistant structures is an important area of safety research in nuclear, aerospace, chemical process and vehicle industries. Institute for Nuclear and Energy Technologies (IKET) of Research Centre- Karlsruhe (Forschungszentrum Karlsruhe or FZK) in Germany is pursuing active research on the entire spectrum of safety evaluation for efficient hydrogen management in case of the postulated design basis and beyond the design basis severe accidents for nuclear and non-nuclear applications. This report concentrates on the consequence analysis of hydrogen combustion accidents with emphasis on the structural safety assessment. The transient finite element simulation results obtained for 2gm, 4gm, 8gm and 16gm hydrogen combustion experiments concluded recently on the test-cell structure are described. The frequencies and damping of the test-cell observed during the hammer tests and the combustion experiments are used for the present three dimensional finite element model qualification. For the numerical transient dynamic evaluation of the test-cell structure, the pressure time history data computed with CFD code COM-3D is used for the four combustion experiments. Detail comparisons of the present numerical results for the four combustion experiments with the observed time signals are carried out to evaluate the structural connection behavior. For all the combustion experiments excellent agreement is noted for the computed accelerations and displacements at the standard transducer locations, where the measurements were made during the different combustion tests. In addition inelastic analysis is also presented for the test-cell structure to evaluate the limiting impulsive and quasi-static pressure loads. These results are used to evaluate the response of the test cell structure for the postulated over pressurization of the test-cell due to the blast load generated in case of 64 gm hydrogen ignition for which additional sets of computations were
International Nuclear Information System (INIS)
Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.; Hughes, E.D.; Solbrig, C.W.
1975-11-01
A mathematical model and a numerical solution scheme for thermal-hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media
Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions
International Nuclear Information System (INIS)
Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong
2014-01-01
The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea
Energy Technology Data Exchange (ETDEWEB)
Stevanovic, Vladimir; Studovic, Milovan [Faculty of Mechanical Engineering, University of Belgrade, Belgrade (Yugoslavia); Bratic, Aleksandar [Thermal Power Plant Nikola Tesla (Yugoslavia)
1993-11-01
Simulation and analysis of a real main steam line break transient at the coal fired 300 MW Drmno Thermal Power Plant have been performed by the computer code TEA-01. The methods and procedures used could be applied to a nuclear power plant. 9 refs., 6 figs.
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)
Transient analysis of a queue with queue-length dependent MAP and its application to SS7 network
Directory of Open Access Journals (Sweden)
Bong Dae Choi
1999-01-01
Full Text Available We analyze the transient behavior of a Markovian arrival queue with congestion control based on a double of thresholds, where the arrival process is a queue-length dependent Markovian arrival process. We consider Markov chain embedded at arrival epochs and derive the one-step transition probabilities. From these results, we obtain the mean delay and the loss probability of the nth arrival packet. Before we study this complex model, first we give a transient analysis of an MAP/M/1 queueing system without congestion control at arrival epochs. We apply our result to a signaling system No. 7 network with a congestion control based on thresholds.
Coupled transient thermo-fluid/thermal-stress analysis approach in a VTBM setting
International Nuclear Information System (INIS)
Ying, A.; Narula, M.; Zhang, H.; Abdou, M.
2008-01-01
A virtual test blanket module (VTBM) has been envisioned as a utility to aid in streamlining and optimizing the US ITER TBM design effort by providing an integrated multi-code, multi-physics modeling environment. Within this effort, an integrated simulation approach is being developed for TBM design calculations and performance evaluation. Particularly, integrated thermo-fluid/thermal-stress analysis is important for enabling TBM design and performance calculations. In this paper, procedures involved in transient coupled thermo-fluid/thermal-stress analysis are investigated. The established procedure is applied to study the impact of pulsed operational phenomenon on the thermal-stress response of the TBM first wall. A two-way coupling between the thermal strain and temperature field is also studied, in the context of a change in thermal conductivity of the beryllium pebble bed in a solid breeder blanket TBM due to thermal strain. The temperature field determines the thermal strain in beryllium, which in turn changes the temperature field. Iterative thermo-fluid/thermal strain calculations have been applied to both steady-state and pulsed operation conditions. All calculations have been carried out in three dimensions with representative MCAD models, including all the TBM components in their entirety
CENTAR code for extended nonlinear transient analysis of extraterrestrial reactor systems
International Nuclear Information System (INIS)
Nassersharif, B.; Peer, J.S.; DeHart, M.D.
1987-01-01
Current interest in the application of nuclear reactor-driven power systems to space missions has generated a need for a systems simulation code to model and analyze space reactor systems; such a code has been initiated at Texas A and M, and the first version is nearing completion; release was anticipated in the fall of 1987. This code, named CENTAR (Code for Extended Nonlinear Transient Analysis of Extraterrestrial Reactor Systems), is designed specifically for space systems and is highly vectorizable. CENTAR is composed of several specialized modules. A fluids module is used to model fluid behavior throughout the system. A wall heat transfer module models the heat transfer characteristics of all walls, insulation, and structure around the system. A fuel element thermal analysis module is used to predict the temperature behavior and heat transfer characteristics of the reactor fuel rods. A kinetics module uses a six-group point kinetics formulation to model reactivity feedback and control and the ANS 5.1 decay-heat curve to model shutdown decay-heat production. A pump module models the behavior of thermoelectric-electromagnetic pumps, and a heat exchanger module models not only thermal effects in thermoelectric heat exchangers, but also predicts electrical power production for a given configuration. Finally, an accumulator module models coolant expansion/contraction accumulators
Analysis methods of stochastic transient electro–magnetic processes in electric traction system
Directory of Open Access Journals (Sweden)
T. M. Mishchenko
2013-04-01
Full Text Available Purpose. The essence and basic characteristics of calculation methods of transient electromagnetic processes in the elements and devices of non–linear dynamic electric traction systems taking into account the stochastic changes of voltages and currents in traction networks of power supply subsystem and power circuits of electric rolling stock are developed. Methodology. Classical methods and the methods of non–linear electric engineering, as well as probability theory method, especially the methods of stationary ergodic and non–stationary stochastic processes application are used in the research. Findings. Using the above-mentioned methods an equivalent circuit and the system of nonlinear integra–differential equations for electromagnetic condition of the double–track inter-substation zone of alternating current electric traction system are drawn up. Calculations allow obtaining electric traction current distribution in the areas of feeder zones. Originality. First of all the paper is interesting and important from scientific point of view due to the methods, which allow taking into account probabilistic character of change for traction voltages and electric traction system currents. On the second hand the researches develop the most efficient methods of nonlinear circuits’ analysis. Practical value. The practical value of the research is presented in application of the methods to the analysis of electromagnetic and electric energy processes in the traction power supply system in the case of high-speed train traffic.
Study on time-frequency analysis method of very fast transient overvoltage
Li, Shuai; Liu, Shiming; Huang, Qiyan; Fu, Chuanshun
2018-04-01
The operation of the disconnector in the gas insulated substation (GIS) may produce very fast transient overvoltage (VFTO), which has the characteristics of short rise time, short duration, high amplitude and rich frequency components. VFTO can cause damage to GIS and secondary equipment, and the frequency components contained in the VFTO can cause resonance overvoltage inside the transformer, so it is necessary to study the spectral characteristics of the VFTO. From the perspective of signal processing, VFTO is a kind of non-stationary signal, the traditional Fourier transform is difficult to describe its frequency which changes with time, so it is necessary to use time-frequency analysis to analyze VFTO spectral characteristics. In this paper, we analyze the performance of short time Fourier transform (STFT), Wigner-Ville distribution (WVD), pseudo Wigner-Ville distribution (PWVD) and smooth pseudo Wigner-Ville distribution (SPWVD). The results show that SPWVD transform is the best. The time-frequency aggregation of SPWVD is higher than STFT, and it does not have cross-interference terms, which can meet the requirements of VFTO spectrum analysis.
On uncertainty and local sensitivity analysis for transient conjugate heat transfer problems
International Nuclear Information System (INIS)
Rauch, Christian
2012-01-01
The need for simulating real-world behavior of automobiles has led to more and more sophisticated models being added of various physical phenomena for being coupled together. This increases the number of parameters to be set and, consequently, the required knowledge of their relative importance for the solution and the theory behind them. Sensitivity and uncertainty analysis provides the knowledge of parameter importance. In this paper a thermal radiation solver is considered that performs conduction calculations and receives heat transfer coefficient and fluid temperate at a thermal node. The equations of local, discrete, and transient sensitivities for the conjugate heat transfer model solved by the finite difference method are being derived for some parameters. In the past, formulations for the finite element method have been published. This paper builds on the steady-state formulation published previously by the author. A numerical analysis on the stability of the solution matrix is being conducted. From those normalized sensitivity coefficients are calculated dimensionless uncertainty factors. On a simplified example the relative importance of the heat transfer modes at various locations is then investigated by those uncertainty factors and their changes over time
Directory of Open Access Journals (Sweden)
Yan Wang
2017-01-01
Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.
International Nuclear Information System (INIS)
Berta, V.T.
1977-05-01
Fourteen experiments on the Loss-of-Fluid Test (LOFT) facility pressure suppression system (PSS) are analyzed in relation to the vertical load generated on the suppression tank in the first 0.5 sec of the transient. Variations in principle parameters affecting the generation of vertical loads were included in the experiments. The internal and external vent submergences are identified from the analysis as being parameters which are first order in influencing the magnitude of the vertical load. These parameters are geometric in nature and depend only on PSS design. Physical parameters of total energy input and rate of energy input to the dry well, which influence the dry well pressurization, also are identified as being first order in influencing the magnitude of the vertical loads. The vertical load magnitude is a direct function of these geometric and physical parameters. The analysis indicates that a small value in any one of the parameters will cause the vertical load to be small and to have little dependence on the magnitude of the other parameters. In addition, the phenomena of nonuniform nonsynchronized vent inlet pressures, which have origins that are either geometric, physical, or a combination of both, act as a significant vertical load reduction mechanism
Energy Technology Data Exchange (ETDEWEB)
Park, Jeong Soon; Choi, Young Hwan; Jhung, Myung Jo [Safety Research Division, Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2016-04-15
The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.
International Nuclear Information System (INIS)
Ayazuddin, S.K.; Qureshi, A.A.; Hayat, T.
1997-11-01
The Primary Water Inlet Pipeline (PW-IPL) is of stainless steel conveying demineralized water from hold-up tank to the reactor pool of Pakistan Research Reactor-1 (PARR-1). The section of the pipeline from heat exchangers to the valve pit is hanger supported in the pump room and the rest of the section from valve pit to the reactor pool is embedded. The PW-IPL is subjected to steady state and transient vibrations. The reactor pumps, which drive the coolant through various circuits mainly contribute the steady state vibrations, while transient vibrations arise due to instant closure of the check valve (water hammer). The ASME Boiler and Pressure Vessel code provides data about the acceptable limits of stresses related to the primary static stress due to steady state vibrations. However, due to complexity in the pipe structure, stresses related to the transient vibrations are neglected in the code. In this report attempt has been made to analyzed both steady state and transient vibrations of PW-IPL of PARR-1. Since, both the steady state and transient vibrations affect the hanger-supported section of the PW-IPL, therefore, it was selected for vibration test measurements. In the analysis vibration data was compared with the allowable limits and estimations of maximum pressure build-up, eflection, natural frequency, tensile and shear load on hanger support, and the ratio of maximum combine stress to the allowable load were made. (author)
International Nuclear Information System (INIS)
Yamawaki, Michio; Tanaka, Satoru; Kiyoshi, Tsukasa
1989-01-01
To obtain further information on the transient permeation behavior of hydrogen isotopes as caused by an abrupt temperature change, numerical calculations were carried out for two typical metals, nickel and vanadium. Deuterium permeation through nickel is analyzed as a typical case of bulk-diffusion-limited permeation. Its transient behavior changed dramatically according to the specimen thickness. The transient behavior, in general, is separated into two parts, initial and latter period behaviors. Conditions which cause such a separation were evaluated. Evaluation of the hydrogen diffusivity and solubility by an analysis of transient curves of hydrogen permeation was carried out. The transient behavior of simultaneous gas- and ion-driven hydrogen permeation through vanadium was also analyzed. Overshooting of the hydrogen permeation rate appears with an abrupt temperature increase. Increasing the impinging ion flux causes the overshooting peak to become sharper, and also reduces the change of the steady-state permeation rate to be attained after the temperature change compared with the initial value. (orig.)
Li, Chenlin; Guo, Huili; Tian, Xiaogeng
2018-04-01
This paper is devoted to the thermal shock analysis for viscoelastic materials under transient heating loads. The governing coupled equations with time-delay parameter and nonlocal scale parameter are derived based on the generalized thermo-viscoelasticity theory. The problem of a thin plate composed of viscoelastic material, subjected to a sudden temperature rise at the boundary plane, is solved by employing Laplace transformation techniques. The transient responses, i.e. temperature, displacement, stresses, heat flux as well as strain, are obtained and discussed. The effects of time-delay and nonlocal scale parameter on the transient responses are analyzed and discussed. It can be observed that: the propagation of thermal wave is dynamically smoothed and changed with the variation of time-delay; while the displacement, strain, and stress can be rapidly reduced by nonlocal scale parameter, which can be viewed as an important indicator for predicting the stiffness softening behavior for viscoelastic materials.
Energy Technology Data Exchange (ETDEWEB)
Djukanovic, M [Inst. ' Nikola Tesla' , Belgrade (Yugoslavia); Sobajic, D J; Pao, Yohhan [Case Western Reserve Univ., Cleveland, OH (United States). Dept. of Electrical Engineering and Applied Physics Case Western Reserve Univ., Cleveland, OH (United States). Dept. of Computer Engineering and Science AI WARE inc., Cleveland, OH (United States)
1992-10-01
In heavily stressed power systems, post-fault transient voltage dips can lead to undesired tripping of industrial drives and large induction motors. The lowest transient voltage dips occur when fault clearing times are less than critical ones. In this paper, we propose a new iterative analytical methodology to obtain more accurate estimates of voltage dips at maximum angular swing in direct transient stability analysis. We also propose and demonstrate the possibility of storing the results of these computations in the associative memory (AM) system, which exhibits remarkable generalization capabilities. Feature-based models stored in the AM can be utilized for fast and accurate prediction of the location, duration and the amount of the worst voltage dips, thereby avoiding the need and cost for lengthy time-domain simulations. Numerical results obtained using the example of the New England power system are presented to illustrate our approach. (Author)
Analysis of loss of normal feedwater transient using RELAP5/MOD1/NSC; KNU1 plant simulation
International Nuclear Information System (INIS)
Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo
1986-01-01
Simulation of the system thermal-hydraulic parameters was carried out following the KNU1(Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on november 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS(Reactor Coolant System) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018. (Author)
NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK
Directory of Open Access Journals (Sweden)
Filip Osuský
2016-12-01
Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.
RETRAN operational transient analysis of the Big Rock Point plant boiling water reactor
International Nuclear Information System (INIS)
Sawtelle, G.R.; Atchison, J.D.; Farman, R.F.; VandeWalle, D.J.; Bazydlo, H.G.
1983-01-01
Energy Incorporated used the RETRAN computer code to model and calculate nine Consumers Power Company Big Rock Point Nuclear Power Plant transients. RETRAN, a best-estimate, one-dimensional, homogeneous-flow thermal-equilibrium code, is applicable to FSAR Chapter 15 transients for Conditions 1 through IV. The BWR analyses were performed in accordance with USNRC Standard Review Plan criteria and in response to the USNRC Systematic Evaluation Program. The RETRAN Big Rock Point model was verified by comparison to plant startup test data. This paper discusses the unique modeling techniques used in RETRAN to model this steam-drum-type BWR. Transient analyses results are also presented
Analysis of incoloy 800ht alloy tested in thermal transient conditions
International Nuclear Information System (INIS)
Velciu, L.; Meleg, T.; Nitu, A.; Popa, L.
2015-01-01
This paper investigated Incoloy 800 HT alloy after following thermal transient tests: fast heating rates (50° and 90°C/minute) up to 1,000°C, maintaining this temperature level (0 and 60 minutes), furnace-cooling until 220°C, and then air-cooling. This alloy is one of the candidate materials for construction of the steam generators of the future NPP reactors. The analysis consisted in metallographic examination and traction tests. The samples were investigated using the Olympus GX 71 optical microscope, the OPL microdurometer with automatic cycle and WALTER BAI traction device. The average grain size was determined by linear interception method. The micro hardness was calculated by the relationship from the device technical book. On the traction diagrams were obtained: strength resistance (Rm), elongation at rupture (A) and elastic modulus (E). The tested alloy was compared with the ''as received'' material, and the results showed a good behavior of this alloy in the presented conditions. (authors)
Energy Technology Data Exchange (ETDEWEB)
Sjenitzer, Bart L.; Hoogenboom, J. Eduard, E-mail: B.L.Sjenitzer@TUDelft.nl, E-mail: J.E.Hoogenboom@TUDelft.nl [Delft University of Technology (Netherlands)
2011-07-01
A new Dynamic Monte Carlo method is implemented in the general purpose Monte Carlo code Tripoli 4.6.1. With this new method incorporated, a general purpose code can be used for safety transient analysis, such as the movement of a control rod or in an accident scenario. To make the Tripoli code ready for calculating on dynamic systems, the Tripoli scheme had to be altered to incorporate time steps, to include the simulation of delayed neutron precursors and to simulate prompt neutron chains. The modified Tripoli code is tested on two sample cases, a steady-state system and a subcritical system and the resulting neutron fluxes behave just as expected. The steady-state calculation has a constant neutron flux over time and this result shows the stability of the calculation. The neutron flux stays constant with acceptable variance. This also shows that the starting conditions are determined correctly. The sub-critical case shows that the code can also handle dynamic systems with a varying neutron flux. (author)
International Nuclear Information System (INIS)
Sjenitzer, Bart L.; Hoogenboom, J. Eduard
2011-01-01
A new Dynamic Monte Carlo method is implemented in the general purpose Monte Carlo code Tripoli 4.6.1. With this new method incorporated, a general purpose code can be used for safety transient analysis, such as the movement of a control rod or in an accident scenario. To make the Tripoli code ready for calculating on dynamic systems, the Tripoli scheme had to be altered to incorporate time steps, to include the simulation of delayed neutron precursors and to simulate prompt neutron chains. The modified Tripoli code is tested on two sample cases, a steady-state system and a subcritical system and the resulting neutron fluxes behave just as expected. The steady-state calculation has a constant neutron flux over time and this result shows the stability of the calculation. The neutron flux stays constant with acceptable variance. This also shows that the starting conditions are determined correctly. The sub-critical case shows that the code can also handle dynamic systems with a varying neutron flux. (author)
You, Myung-Won; Kim, Kyung Won; Pyo, Junhee; Huh, Jimi; Kim, Hyoung Jung; Lee, So Jung; Park, Seong Ho
2017-01-01
We aimed to evaluate the correlation between liver stiffness measurement using transient elastography (TE-LSM) and hepatic venous pressure gradient and the diagnostic performance of TE-LSM in assessing clinically significant portal hypertension through meta-analysis. Eleven studies were included from thorough literature research and selection processes. The summary correlation coefficient was 0.783 (95% confidence interval [CI], 0.737-0.823). Summary sensitivity, specificity and area under the hierarchical summary receiver operating characteristic curve (AUC) were 87.5% (95% CI, 75.8-93.9%), 85.3 % (95% CI, 76.9-90.9%) and 0.9, respectively. The subgroup with low cut-off values of 13.6-18 kPa had better summary estimates (sensitivity 91.2%, specificity 81.3% and partial AUC 0.921) than the subgroup with high cut-off values of 21-25 kPa (sensitivity 71.2%, specificity 90.9% and partial AUC 0.769). In summary, TE-LSM correlated well with hepatic venous pressure gradient and represented good diagnostic performance in diagnosing clinically significant portal hypertension. For use as a sensitive screening tool, we propose using low cut-off values of 13.6-18 kPa in TE-LSM. Copyright Â© 2016 World Federation for Ultrasound in Medicine & Biology. Published by Elsevier Inc. All rights reserved.
Brain SPECT analysis using statistical parametric mapping in patients with transient global amnesia
Energy Technology Data Exchange (ETDEWEB)
Kim, E. N.; Sohn, H. S.; Kim, S. H; Chung, S. K.; Yang, D. W. [College of Medicine, The Catholic Univ. of Korea, Seoul (Korea, Republic of)
2001-07-01
This study investigated alterations in regional cerebral blood flow (rCBF) in patients with transient global amnesia (TGA) using statistical parametric mapping 99 (SPM99). Noninvasive rCBF measurements using 99mTc-ethyl cysteinate dimer (ECD) SPECT were performed on 8 patients with TGA and 17 age matched controls. The relative rCBF maps in patients with TGA and controls were compared. In patients with TGA, significantly decreased rCBF was found along the left superior temporal extending to left parietal region of the brain and left thalamus. There were areas of increased rCBF in the right temporal, right frontal region and right thalamus. We could demonstrate decreased perfusion in left cerebral hemisphere and increased perfusion in right cerebral hemisphere in patients with TGA using SPM99. The reciprocal change of rCBF between right and left cerebral hemisphere in patients with TGA might suggest that imbalanced neuronal activity between the bilateral hemispheres may be important role in the pathogenesis of the TGA. For quantitative SPECT analysis in TGA patients, we recommend SPM99 rather than the ROI method because of its definitive advantages.
Transient Thermal Analysis of 3-D Integrated Circuits Packages by the DGTD Method
Li, Ping
2017-03-11
Since accurate thermal analysis plays a critical role in the thermal design and management of the 3-D system-level integration, in this paper, a discontinuous Galerkin time-domain (DGTD) algorithm is proposed to achieve this purpose. Such as the parabolic partial differential equation (PDE), the transient thermal equation cannot be directly solved by the DGTD method. To address this issue, the heat flux, as an auxiliary variable, is introduced to reduce the Laplace operator to a divergence operator. The resulting PDE is hyperbolic, which can be further written into a conservative form. By properly choosing the definition of the numerical flux used for the information exchange between neighboring elements, the hyperbolic thermal PDE can be solved by the DGTD together with the auxiliary differential equation. The proposed algorithm is a kind of element-level domain decomposition method, which is suitable to deal with multiscale geometries in 3-D integrated systems. To verify the accuracy and robustness of the developed DGTD algorithm, several representative examples are benchmarked.
Brain SPECT analysis using statistical parametric mapping in patients with transient global amnesia
International Nuclear Information System (INIS)
Kim, E. N.; Sohn, H. S.; Kim, S. H; Chung, S. K.; Yang, D. W.
2001-01-01
This study investigated alterations in regional cerebral blood flow (rCBF) in patients with transient global amnesia (TGA) using statistical parametric mapping 99 (SPM99). Noninvasive rCBF measurements using 99mTc-ethyl cysteinate dimer (ECD) SPECT were performed on 8 patients with TGA and 17 age matched controls. The relative rCBF maps in patients with TGA and controls were compared. In patients with TGA, significantly decreased rCBF was found along the left superior temporal extending to left parietal region of the brain and left thalamus. There were areas of increased rCBF in the right temporal, right frontal region and right thalamus. We could demonstrate decreased perfusion in left cerebral hemisphere and increased perfusion in right cerebral hemisphere in patients with TGA using SPM99. The reciprocal change of rCBF between right and left cerebral hemisphere in patients with TGA might suggest that imbalanced neuronal activity between the bilateral hemispheres may be important role in the pathogenesis of the TGA. For quantitative SPECT analysis in TGA patients, we recommend SPM99 rather than the ROI method because of its definitive advantages
Two-dimensional nonlinear transient heat transfer analysis of variable section pin fins
Energy Technology Data Exchange (ETDEWEB)
Malekzadeh, P. [Department of Mechanical Engineering, School of Engineering, Persian Gulf University, Boushehr 75168 (Iran); Rahideh, H. [Department of Chemical Engineering, School of Engineering, Persian Gulf University, Boushehr 75168 (Iran)
2009-04-15
The two-dimensional nonlinear transient heat transfer analysis of variable cross section pin-fins is studied using the incremental differential quadrature method (IDQM) as a simple, accurate, and computationally efficient numerical tool. The formulations are general so that it can easily be used for arbitrary continuously varying cross section pin fins with the spatial-temperature dependent thermal parameters. On all external surfaces of the pin fin, the convective-radiative condition is considered. The effects of two different types of boundary conditions at the base of pin fin are investigated: time and spatial dependent temperature, and the convection heat transfer. The thermal conductivity of the pin fin is assumed to vary as a linear function of the temperature. The accuracy of the method is demonstrated by comparing its results with those generated by finite difference method. It is shown that using few grid points, results in excellent agreements with those of FDM are obtained. Less computational efforts of the method with respect to finite difference method is shown. (author)
International Nuclear Information System (INIS)
Mahaffy, J.; Boyack, B.E.; Steinke, R.G.
1998-05-01
The objective of this document is to ease the task of adding new system components to the Transient Reactor Analysis Code (TRAC) or altering old ones. Sufficient information is provided to permit replacement or modification of physical models and correlations. Within TRAC, information is passed at two levels. At the upper level, information is passed by system-wide and component-specific data modules at and above the level of component subroutines. At the lower level, information is passed through a combination of module-based data structures and argument lists. This document describes the basic mechanics involved in the flow of information within the code. The discussion of interfaces in the body of this document has been kept to a general level to highlight key considerations. The appendices cover instructions for obtaining a detailed list of variables used to communicate in each subprogram, definitions and locations of key variables, and proposed improvements to intercomponent interfaces that are not available in the first level of code modernization
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
Numerical analysis of transient heat conduction in downward-facing curved sections during quenching
International Nuclear Information System (INIS)
Gao, C.; El-Genk, M.S.
1996-01-01
Pool boiling from downward-facing surfaces is of interest in many applications such as cooling of electric cables, handling of containers of hazardous liquids and external cooling of nuclear reactor vessels. Here, a two-dimensional numerical analysis was performed to determine pool boiling curves from downward-facing curved stainless-steel and copper surfaces during quenching in saturated water. To ensure stability and accuracy of the numerical solution, the alternating direction implicit (ADI) method based on finite control volume representations was employed. A time dependent boundary condition was provided by the measured temperature at nine interior locations near the boiling surface. Best results were obtained using a grid of 20x20 CVs and a non-iterative approach. Calculated temperatures near the top surface of the metal sections agreed with measured values to within 0.5 K and 2.5 K for the copper and stainless-steel sections, respectively. The running time on a Pentium 90 MHz PC for the entire boiling curve was 7% of the real transient time and 4% of that of a simplified Gaussian elimination (SGE) method for the Crank-Nicolson scheme
Wavelet transform analysis of transient signals: the seismogram and the electrocardiogram
Energy Technology Data Exchange (ETDEWEB)
Anant, K.S.
1997-06-01
In this dissertation I quantitatively demonstrate how the wavelet transform can be an effective mathematical tool for the analysis of transient signals. The two key signal processing applications of the wavelet transform, namely feature identification and representation (i.e., compression), are shown by solving important problems involving the seismogram and the electrocardiogram. The seismic feature identification problem involved locating in time the P and S phase arrivals. Locating these arrivals accurately (particularly the S phase) has been a constant issue in seismic signal processing. In Chapter 3, I show that the wavelet transform can be used to locate both the P as well as the S phase using only information from single station three-component seismograms. This is accomplished by using the basis function (wave-let) of the wavelet transform as a matching filter and by processing information across scales of the wavelet domain decomposition. The `pick` time results are quite promising as compared to analyst picks. The representation application involved the compression of the electrocardiogram which is a recording of the electrical activity of the heart. Compression of the electrocardiogram is an important problem in biomedical signal processing due to transmission and storage limitations. In Chapter 4, I develop an electrocardiogram compression method that applies vector quantization to the wavelet transform coefficients. The best compression results were obtained by using orthogonal wavelets, due to their ability to represent a signal efficiently. Throughout this thesis the importance of choosing wavelets based on the problem at hand is stressed. In Chapter 5, I introduce a wavelet design method that uses linear prediction in order to design wavelets that are geared to the signal or feature being analyzed. The use of these designed wavelets in a test feature identification application led to positive results. The methods developed in this thesis; the
Transient stability analysis with equal-area criterion for out of step ...
African Journals Online (AJOL)
user
*Corresponding Author: e-mail: almoatazabdelaziz@hotmail.com, Tel. ...... Transient stability of power systems: theory and Practice, John Wiley and Sons. ... and protection and new optimization techniques in power systems operation and.
An analysis of the transient's social behavior in the radiological emergency planning zone
International Nuclear Information System (INIS)
Bang, Sun Young; Lee, Gab Bock; Chung, Yang Geun; Lee, Jae Eun
2007-01-01
The purpose of this study is to analyze the social behavior, especially, the evacuation-related social behavior, of the transients in the radiological Emergency Planning Zone (EPZ) of nuclear power plants. So, the meaning and kinds of the evacuation and the significance of the Trip Generation Time (TGT) have been reviewed. The characteristics of the social behavior of the transient around Ulchin, Wolsong and Kori sites was analyzed through field surveys by using the questionnaire. The major findings of this research implications are as follows. First, for securing the safe evacuation, the alternatives to effectively provide the information on the evacuation warning may be prepared. Second, it is necessary to establish the education and training of transient's evacuation. Third, it is needed that the cause and background of the evacuation refusal are identified and the new response plan to secure transient's safety is prepared
International Nuclear Information System (INIS)
Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa
1986-01-01
Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)
Transient photoresponse in amorphous In-Ga-Zn-O thin films under stretched exponential analysis
Luo, Jiajun; Adler, Alexander U.; Mason, Thomas O.; Bruce Buchholz, D.; Chang, R. P. H.; Grayson, M.
2013-04-01
We investigated transient photoresponse and Hall effect in amorphous In-Ga-Zn-O thin films and observed a stretched exponential response which allows characterization of the activation energy spectrum with only three fit parameters. Measurements of as-grown films and 350 K annealed films were conducted at room temperature by recording conductivity, carrier density, and mobility over day-long time scales, both under illumination and in the dark. Hall measurements verify approximately constant mobility, even as the photoinduced carrier density changes by orders of magnitude. The transient photoconductivity data fit well to a stretched exponential during both illumination and dark relaxation, but with slower response in the dark. The inverse Laplace transforms of these stretched exponentials yield the density of activation energies responsible for transient photoconductivity. An empirical equation is introduced, which determines the linewidth of the activation energy band from the stretched exponential parameter β. Dry annealing at 350 K is observed to slow the transient photoresponse.
Analysis of metal fuel transient overpower experiments with the SAS4A accident analysis code
International Nuclear Information System (INIS)
Tentner, A.M.; Kalimullah; Miles, K.J.
1990-01-01
The results of the SAS4A analysis of the M7 TREAT Metal fuel experiment are presented. New models incorporated in the metal fuel version of SAS4A are described. The computational results are compared with the experimental observations and this comparison is used in the interpretation of physical phenomena. This analysis was performed using the integrated metal fuel SAS4A version and covers a wide range of events, providing an increased degree of confidence in the SAS4A metal fuel accident analysis capabilities
A Design Method for Graded Insulation of Transformers by Transient Electric Field Intensity Analysis
Yamashita, Hideo; Cingoski, Vlatko; Namera, Akihiro; Nakamae, Eihachiro; Kitamura, Hideo
2000-01-01
In this paper, a calculation method for transient electric field distribution inside a transformer impressed with voltage is proposed: The concentrated electric network for the transformer is constructed by dividing transformer windings into several blocks, and the transient voltage and electric field intensity distributions inside the transformer are calculated by using the axisymmetrical finite element method. Moreover, an animated display of the distributions is realized: The visualization...
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.
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.
Cerebral blood flow SPET in transient global amnesia with automated ROI analysis by 3DSRT
Energy Technology Data Exchange (ETDEWEB)
Takeuchi, Ryo [Division of Nuclear Medicine, Nishi-Kobe Medical Center, Kohjidai 5-7-1, 651-2273, Nishi-ku, Kobe-City, Hyogo (Japan); Matsuda, Hiroshi [Department of Radiology, National Center Hospital for Mental, Nervous and Muscular Disorders, National Center of Neurology and Psychiatry, Tokyo (Japan); Yoshioka, Katsunori [Daiichi Radioisotope Laboratories, Ltd., Tokyo (Japan); Yonekura, Yoshiharu [Biomedical Imaging Research Center, University of Fukui, Fukui (Japan)
2004-04-01
The aim of this study was to determine the areas involved in episodes of transient global amnesia (TGA) by calculation of cerebral blood flow (CBF) using 3DSRT, fully automated ROI analysis software which we recently developed. Technetium-99m l,l-ethyl cysteinate dimer single-photon emission tomography ({sup 99m}Tc-ECD SPET) was performed during and after TGA attacks on eight patients (four men and four women; mean study interval, 34 days). The SPET images were anatomically standardized using SPM99 followed by quantification of 318 constant ROIs, grouped into 12 segments (callosomarginal, precentral, central, parietal, angular, temporal, posterior cerebral, pericallosal, lenticular nucleus, thalamus, hippocampus and cerebellum), in each hemisphere to calculate segmental CBF (sCBF) as the area-weighted mean value for each of the respective 12 segments based on the regional CBF in each ROI. Correlation of the intra- and post-episodic sCBF of each of the 12 segments of the eight patients was estimated by scatter-plot graphical analysis and Pearson's correlation test with Fisher's Z-transformation. For the control, {sup 99m}Tc-ECD SPET was performed on eight subjects (three men and five women) and repeated within 1 month; the correlation between the first and second sCBF values of each of the 12 segments was evaluated in the same way as for patients with TGA. Excellent reproducibility between the two sCBF values was found in all 12 segments of the control subjects. However, a significant correlation between intra- and post-episodic sCBF was not shown in the thalamus or angular segments of TGA patients. The present study was preliminary, but at least suggested that thalamus and angular regions are closely involved in the symptoms of TGA. (orig.)
Cerebral blood flow SPET in transient global amnesia with automated ROI analysis by 3DSRT
International Nuclear Information System (INIS)
Takeuchi, Ryo; Matsuda, Hiroshi; Yoshioka, Katsunori; Yonekura, Yoshiharu
2004-01-01
The aim of this study was to determine the areas involved in episodes of transient global amnesia (TGA) by calculation of cerebral blood flow (CBF) using 3DSRT, fully automated ROI analysis software which we recently developed. Technetium-99m l,l-ethyl cysteinate dimer single-photon emission tomography ( 99m Tc-ECD SPET) was performed during and after TGA attacks on eight patients (four men and four women; mean study interval, 34 days). The SPET images were anatomically standardized using SPM99 followed by quantification of 318 constant ROIs, grouped into 12 segments (callosomarginal, precentral, central, parietal, angular, temporal, posterior cerebral, pericallosal, lenticular nucleus, thalamus, hippocampus and cerebellum), in each hemisphere to calculate segmental CBF (sCBF) as the area-weighted mean value for each of the respective 12 segments based on the regional CBF in each ROI. Correlation of the intra- and post-episodic sCBF of each of the 12 segments of the eight patients was estimated by scatter-plot graphical analysis and Pearson's correlation test with Fisher's Z-transformation. For the control, 99m Tc-ECD SPET was performed on eight subjects (three men and five women) and repeated within 1 month; the correlation between the first and second sCBF values of each of the 12 segments was evaluated in the same way as for patients with TGA. Excellent reproducibility between the two sCBF values was found in all 12 segments of the control subjects. However, a significant correlation between intra- and post-episodic sCBF was not shown in the thalamus or angular segments of TGA patients. The present study was preliminary, but at least suggested that thalamus and angular regions are closely involved in the symptoms of TGA. (orig.)
The effect of the virtual mass force term on the stability of transient two-phase flow analysis
International Nuclear Information System (INIS)
Watanabe, Tadashi; Hirano, Masashi; Tanabe, Fumiya
1989-08-01
The effect of the virtual mass force term on the stability of transient two-phase flow analysis is studied. The objective form of the virtual mass acceleration is used. The virtual mass coefficient is determined from the stability condition of basic equations against infinitesimal high wave-number perturbations. The parameter is chosen so that a reasonable agreement between the analytical and experimental sound speed in two-phase flows can be obtained. A one-dimensional sedimentation problem is simulated by the MINCS code which is a tool for transient two-phase flow analysis. The stability analysis is performed for the numerical procedure. It is shown that calculated results are stabilized so long as the virtual mass coefficient satisfies the stability condition of differential equations. (author)
Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3
International Nuclear Information System (INIS)
Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.
2003-01-01
The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)
International Nuclear Information System (INIS)
Schultz, R.R.
1982-01-01
Operating plant transients are of great interest for many reasons, not the least of which is the potential for a mild transient to degenerate to a severe transient yielding core damage. Using the Browns Ferry (BF) Unit-1 plant as a basis of study, the station blackout sequence was investigated by the Severe Accident Sequence Analysis (SASA) Program in support of the Nuclear Regulatory Commission's Unresolved Safety Issue A-44: Station Blackout. A station blackout transient occurs when the plant's AC power from a comemrcial power grid is lost and cannot be restored by the diesel generators. Under normal operating conditions, f a loss of offsite power (LOSP) occurs [i.e., a complete severance of the BF plants from the Tennessee Valley Authority (TVA) power grid], the eight diesel generators at the three BF units would quickly start and power the emergency AC buses. Of the eight diesel generators, only six are needed to safely shut down all three units. Examination of BF-specific data show that LOSP frequency is low at Unit 1. The station blackout frequency is even lower (5.7 x 10 - 4 events per year) and hinges on whether the diesel generators start. The frequency of diesel generator failure is dictated in large measure by the emergency equipment cooling water (EECW) system that cools the diesel generators
Masuzawa, Toru; Ohta, Akiko; Tanaka, Nobuatu; Qian, Yi; Tsukiya, Tomonori
2009-01-01
The effect of the hydraulic force on magnetically levitated (maglev) pumps should be studied carefully to improve the suspension performance and the reliability of the pumps. A maglev centrifugal pump, developed at Ibaraki University, was modeled with 926 376 hexahedral elements for computational fluid dynamics (CFD) analyses. The pump has a fully open six-vane impeller with a diameter of 72.5 mm. A self-bearing motor suspends the impeller in the radial direction. The maximum pressure head and flow rate were 250 mmHg and 14 l/min, respectively. First, a steady-state analysis was performed using commercial code STAR-CD to confirm the model's suitability by comparing the results with the real pump performance. Second, transient analysis was performed to estimate the hydraulic force on the levitated impeller. The impeller was rotated in steps of 1 degrees using a sliding mesh. The force around the impeller was integrated at every step. The transient analysis revealed that the direction of the radial force changed dynamically as the vane's position changed relative to the outlet port during one circulation, and the magnitude of this force was about 1 N. The current maglev pump has sufficient performance to counteract this hydraulic force. Transient CFD analysis is not only useful for observing dynamic flow conditions in a centrifugal pump but is also effective for obtaining information about the levitation dynamics of a maglev pump.
Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code
International Nuclear Information System (INIS)
Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun
2014-01-01
The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available
Transient analysis and thermal hydraulic margins of GHARR-1 using the PARET/NAL code
International Nuclear Information System (INIS)
Adoo, N.A.
2009-06-01
The PARET code has been adapted by the IAEA for testing transient behaviour in research reactors. The PARET code provides a coupled thermal hydrodynamic and point kinetics capability with a continuous reactivity feedback and an optional voiding model that estimates the voiding produced by the subcooled boiling. The present version of the PARET/ANL 73 code provides a convenient means of assessing the various models and correlations proposed for the use in the analysis of research reactor behaviour. The Monte Carlo N-Particle code (MCNP) has been used to obtain power peaking profile for a two channel PARET/ANL model. A PARET model with the corresponding neutronics and thermal hydraulic characteristics for the miniature neutron source reactor (MNSR) has been used to simulate reactivity accidents for the Ghana Research Reactor - 1(GHARR-1) under the MNSR operation conditions of natural circulation, normal operation and reactivity insertion accidents. The simulation results via the insertion of large reactivity demonstrated the high inherent safety features of the MNSR for which the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The hot channel peaking factors for both radial and axial were found to be 1.17 and 1.44 respectively. Thermal hydraulic performance characteristics were investigated and the safety margins determined. The peak clad and coolant temperatures ranged from 59.18 0 C to 106.75 0 C and 42.95 0 C to 178.44 0 C respectively at which nucleate boiling will occur within the flow channels of the core. (au)
Pressure transient analysis in single and two-phase water by finite difference methods
International Nuclear Information System (INIS)
Berry, G.F.; Daley, J.G.
1977-01-01
An important consideration in the design of LMFBR steam generators is the possibility of leakage from a steam generator water tube. The ensuing sodium/water reaction will be largely controlled by the amount of water available at the leak site, thus analysis methods treating this event must have the capability of accurately modeling pressure transients through all states of water occurring in a steam generator, whether single or two-phase. The equation systems of the present model consist of the conservation equations together with an equation of state for one-dimensional homogeneous flow. These equations are then solved using finite difference techniques with phase considerations and non-equilibrium effects being treated through the equation of state. The basis for water property computation is Keenan's 'fundamental equation of state' which is applicable to single-phase water at pressures less than 1000 bars and temperatures less than 1300 0 C. This provides formulations allowing computation of any water property to any desired precision. Two-phase properties are constructed from values on the saturation line. The use of formulations permits the direct calculation of any thermodynamic property (or property derivative) to great precision while requiring very little computer storage, but does involve considerable computation time. For this reason an optional calculation scheme based on the method of 'transfinite interpolation' is included to give rapid computation in selected regions with decreased precision. The conservation equations were solved using the second order Lax-Wendroff scheme which includes wall friction, allows the formation of shocks and locally supersonic flow. Computational boundary conditions were found from a method-of-characteristics solution at the reservoir and receiver ends. The local characteristics were used to interpolate data from inside the pipe to the boundary
Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code
Energy Technology Data Exchange (ETDEWEB)
Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun [KHNP-CRI, Daejeon (Korea, Republic of)
2014-10-15
The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available.
International Nuclear Information System (INIS)
Leavell, W.H.; Mullens, J.A.
1981-01-01
A computational algorithm has been developed to measure transient, phase-interface velocity in two-phase, steam-water systems. The algorithm will be used to measure the transient velocity of steam-water mixture during simulated PWR reflood experiments. By utilizing signals produced by two, spatially separated impedance probes immersed in a two-phase mixture, the algorithm computes the average transit time of mixture fluctuations moving between the two probes. This transit time is computed by first, measuring the phase shift between the two probe signals after transformation to the frequency domain and then computing the phase shift slope by a weighted least-squares fitting technique. Our algorithm, which has been tested with both simulated and real data, is able to accurately track velocity transients as fast as 4 m/s/s
Analysis of the one-dimensional transient compressible vapor flow in heat pipes
Jang, Jong H.; Faghri, Amir; Chang, Won S.
1991-01-01
The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual vapor flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds as well as high mass flow rates are successfully predicted.
Shivakumar, J.; Ashok, M. H.; Khadakbhavi, Vishwanath; Pujari, Sanjay; Nandurkar, Santosh
2018-02-01
The present work focuses on geometrically nonlinear transient analysis of laminated smart composite plates integrated with the patches of Active fiber composites (AFC) using Active constrained layer damping (ACLD) as the distributed actuators. The analysis has been carried out using generalised energy based finite element model. The coupled electromechanical finite element model is derived using Von Karman type nonlinear strain displacement relations and a first-order shear deformation theory (FSDT). Eight-node iso-parametric serendipity elements are used for discretization of the overall plate integrated with AFC patch material. The viscoelastic constrained layer is modelled using GHM method. The numerical results shows the improvement in the active damping characteristics of the laminated composite plates over the passive damping for suppressing the geometrically nonlinear transient vibrations of laminated composite plates with AFC as patch material.
Severe transient analysis of the Penn State University Advanced Light Water Reactor
International Nuclear Information System (INIS)
Borkowski, J.A.
1988-08-01
The Penn State University Advanced Light Water Reactor (PSU ALWR) incorporates various passive and active ultra-safe features, such as continuous online injection and letdown for pressure control, a raised-loop primary system for enhanced natural circulation, a dedicated primary reservoir for enhanced thermal hydraulic control, and a secondary shutdown turbine. Because of the conceptual design basis of the project, the dynamic system modeling was to be performed using a code with a high degree of flexibility. For this reason the modeling has been performed with the Modular Modeling System (MMS). The basic design and normal transients have been performed successfully with MMS. However, the true test of an inherently safe concept lies in its response to more brutal transients. Therefore, such a demonstrative transient is chosen for the PSU ALWR: a turbine trip and reactor scram, concurrent with total loss of offsite ac power. Diesel generators are likewise unavailable. This transient demonstrates the utility of the pressure control system, the shutdown turbine generator, and the enhanced natural circulation of the PSU ALWR. The low flow rates, low pressure drops, and large derivative states encountered in such a transient pose special problems for the modeler and for MMS. The results of the transient analyses indicate excellent performance by the PSU ALWR in terms of inherently safe operation. The primary coolant enters full natural circulation, and removes all decay heat through the steam generators. Further, the steam generators continually supply sufficient steam to the shutdown power system, despite the abrupt changeover to the auxiliary feedwater system. Finally, even with coincident failures in the pressurization system, the primary repressurizes to near-normal values, without overpressurization. No core boiling or uncovery is predicted, and consequently fuel damage is avoided. 17 refs., 19 figs., 4 tabs
Whole-core thermal-hydraulic transient code development and verification for LMFBR analysis
International Nuclear Information System (INIS)
Spencer, D.R.
1979-04-01
Predicted performance during both steady state and transient reactor operation determines the steady state operating limits on LMFBRs. Unnecessary conservatism in performance predictions will not contribute to safety, but will restrict the reactor to more conservative, less economical steady state operation. The most general method for reducing analytical conservatism in LMFBR's without compromising safety is to develop, validate and apply more sophisticated computer models to the limiting performance analyses. The purpose of the on-going Natural Circulation Verification Program (NCVP) is to develop and validate computer codes to analyze natural circulation transients in LMFBRs, and thus, replace unnecessary analytical conservatism with demonstrated calculational capability
Temporal frequency probing for 5D transient analysis of global light transport
O'Toole, Matthew; Heide, Felix; Xiao, Lei; Hullin, Matthias B.; Heidrich, Wolfgang; Kutulakos, Kiriakos N.
2014-01-01
To overcome this complexity, we observe that transient light transport is always separable in the temporal frequency domain. This makes it possible to analyze transient transport one temporal frequency at a time by trivially adapting techniques from conventional projector-to-camera transport. We use this idea in a prototype that offers three never-seen-before abilities: (1) acquiring time-of-flight depth images that are robust to general indirect transport, such as interreflections and caustics; (2) distinguishing between direct views of objects and their mirror reflection; and (3) using a photonic mixer device to capture sharp, evolving wavefronts of "light-in-flight".
Stress analysis in pipelines submitted to internal pressure - and temperature transients
International Nuclear Information System (INIS)
Mansur, T.R.
1981-08-01
Experimental determination of the structural behaviour of a thermal-hydraulic loop, when submitted to simultaneous fast change of pressure and temperature, was performed. For this, electrical strain-gages were positioned at some critical points in order to measure the deformation conditions of the structure. The study of the kinetics of the deformation revealed the presence of important transient stresses, mainly from thermal origin. After this transient behaviour, the structure is submitted to a thermal stress, which is shown to be strongly dependent on the degree of restraint of the structure. (Author) [pt
Dilution and Mixing in transient velocity fields: a first-order analysis
Di Dato, Mariaines; de Barros, Felipe, P. J.; Fiori, Aldo; Bellin, Alberto
2017-04-01
An appealing remediation technique is in situ oxidation, which effectiveness is hampered by difficulties in obtaining good mixing of the injected oxidant with the contaminant, particularly when the contaminant plume is contained and therefore its deformation is physically constrained. Under such conditions (i.e. containment), mixing may be augmented by inducing temporal fluctuations of the velocity field. The temporal variability of the flow field may increase the deformation of the plume such that diffusive mass flux becomes more effective. A transient periodic velocity field can be obtained by an engineered sequence of injections and extractions from wells, which may serve also as a hydraulic barrier to confine the plume. Assessing the effectiveness of periodic flows to maximize solute mixing is a difficult task given the need to use a 3D setup and the large number of possible flow configurations that should be analyzed in order to identify the optimal one. This is the typical situation in which analytical solutions, though approximated, may assist modelers in screening possible alternative flow configurations such that solute dilution is maximized. To quantify dilution (i.e. a precondition that enables reactive mixing) we utilize the concept of the dilution index [1]. In this presentation, the periodic flow takes place in an aquifer with spatially variable hydraulic conductivity field which is modeled as a Stationary Spatial Random Function. We developed a novel first-order analytical solution of the dilution index under the hypothesis that the flow can be approximated as a sequence of steady state configurations with the mean velocity changing with time in intensity and direction. This is equivalent to assume that the characteristic time of the transient behavior is small compared to the period characterizing the change in time of the mean velocity. A few closed paths have been analyzed quantifying their effectiveness in enhancing dilution and thereby mixing
An analysis of transient flow in upland watersheds: interactions between structure and process
David Lawrence Brown
1995-01-01
The physical structure and hydrological processes of upland watersheds interact in response to forcing functions such as rainfall, leading to storm runoff generation and pore pressure evolution. Transient fluid flow through distinct flow paths such as the soil matrix, macropores, saprolite, and bedrock may be viewed as a consequence of such interactions. Field...
Stochastic resonance of ensemble neurons for transient spike trains: Wavelet analysis
International Nuclear Information System (INIS)
Hasegawa, Hideo
2002-01-01
By using the wavelet transformation (WT), I have analyzed the response of an ensemble of N (=1, 10, 100, and 500) Hodgkin-Huxley neurons to transient M-pulse spike trains (M=1 to 3) with independent Gaussian noises. The cross correlation between the input and output signals is expressed in terms of the WT expansion coefficients. The signal-to-noise ratio (SNR) is evaluated by using the denoising method within the WT, by which the noise contribution is extracted from the output signals. Although the response of a single (N=1) neuron to subthreshold transient signals with noises is quite unreliable, the transmission fidelity assessed by the cross correlation and SNR is shown to be much improved by increasing the value of N: a population of neurons plays an indispensable role in the stochastic resonance (SR) for transient spike inputs. It is also shown that in a large-scale ensemble, the transmission fidelity for suprathreshold transient spikes is not significantly degraded by a weak noise which is responsible to SR for subthreshold inputs
Simulation and analysis of transient over voltages due to capacitor banks switching
International Nuclear Information System (INIS)
Jadid, Sh.; Yazdanpanah, D.
2002-01-01
The switching of any capacitor bank produces over voltages. Transient overvoltage will always occur in the switching device, the switching of shunt capacitor bank has become the most common source of transient voltage on power systems. Transient over voltages due to switching the capacitor bands hurt not only to the capacitor banks, but also to other equipment, such as circuit breakers and transformers. Several methods are available for reducing energising transients. These devices include pre-insertion resistors, pre-insertion inductors,synchronous closing, and MOV arresters. However, not all are practical or economical. The other important problem is existence of capacitor banks in presence of harmonics.Capacitors do not produce harmonics;however,the addition of capacitors to the electrical system will change the frequency response characteristics of the system will change the frequency response characteristics of the system, and in some cases can result in magnification of the voltage and current distortion in the system. In other word in presence of harmonic-producing loads,the capacitors used for power factor correction,may cause parallel resonance with the system inductance, so they increase the total harmonic distortion of voltage and current waveforms
DEFF Research Database (Denmark)
Li, Zhongyu; Zhao, Rende; Xin, Zhen
2016-01-01
The Inrush Transient Current (ITC) in the output of the photovoltaic grid-connected inverters is usually generated when grid voltage sag occurs, which can trigger the protection of the grid-connected inverters, and even destroy the semiconductor switches. Then, the grid-connected inverters...
Transient isotachophoresis in carrier ampholyte-based capillary electrophoresis for protein analysis
Czech Academy of Sciences Publication Activity Database
Busnel, J. M.; Descroix, S.; Godfrin, D.; Hennion, M. C.; Kašička, Václav; Peltre, G.
2006-01-01
Roč. 27, č. 18 (2006), s. 3591-3598 ISSN 0173-0835 Institutional research plan: CEZ:AV0Z40550506 Keywords : carrier ampholyte-based capillary electrophoresis * transient isotachophoresis * proteins Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 4.101, year: 2006
Analysis of very fast transients in layer-type transformer windings
Popov, M.; Sluis, van der L.; Smeets, R.; Lopez Roldan, J.
2007-01-01
This paper deals with the measurement, modeling, and simulation of very fast transient overvoltages in layer-type distribution transformer windings. Measurements were performed by applying a step impulse with 50-ns rise time on a single-phase test transformer equipped with measuring points along the
DEFF Research Database (Denmark)
Wagner, Manfred H.; Rolon-Garrido, Victor H.; Nielsen, Jens Kromann
2008-01-01
The transient and steady-state elongational viscosity data of three bidisperse polystyrene blends were investigated recently by Nielsen et al. [J. Rheol. 50, 453-476 (2006)]. The blends contain a monodisperse high molar mass component (M-L= 390 kg/ mol) in a matrix of a monodisperse small molar m...
Transient analysis of one-sided Lévy-driven queues
Starreveld, N.J.; Bekker, R.; Mandjes, M.
2016-01-01
In this article, we analyze the transient behavior of the workload process in a Lévy-driven queue. We are interested in the value of the workload process at a random epoch; this epoch is distributed as the sum of independent exponential random variables. We consider both cases of spectrally
Transient analysis of one-sided Lévy-driven queues
Starreveld, N.J.; R. Bekker (Rene); M.R.H. Mandjes (Michel)
2016-01-01
textabstractIn this article, we analyze the transient behavior of the workload process in a Lévy-driven queue. We are interested in the value of the workload process at a random epoch; this epoch is distributed as the sum of independent exponential random variables. We consider both cases of
Transient analysis of one-sided Lévy-driven queues
Starreveld, N.; Bekker, R.; Mandjes, M.R.H.
2015-01-01
In this paper we analyze the transient behavior of the workload process in a Lévy input queue. We are interested in the value of the workload process at a random epoch; this epoch is distributed as the sum of independent exponential random variables. We consider both cases of spectrally one-sided
TRANSIENT ANALYSIS OF WIND DIESEL POWER SYSTEM WITH FLYWHEEL ENERGY STORAGE
Directory of Open Access Journals (Sweden)
S. SUJITH
2017-10-01
Full Text Available Wind-Diesel Hybrid power generation is a viable alternative for generating continuous power to isolated power system areas which have inconsistent but potential wind power. The unpredictable nature of variable power from Wind generator to the system is compensated by Diesel generator, which supplies the deficit in generated power from wind to meet the instantaneous system load. However, one of the major challenges for such a system is the higher probability of transients in the form of wind and load fluctuations. This paper analyses the application of Flywheel Energy storage system (FESS to meet the transients during wind-speed and load fluctuations around high wind operation. The power system architecture, the distributed control mechanism governing the flow of power transfer and the modelling of major system components has been discussed and the system performances have been validated using MATLAB /Simulink software. Two cases of transient stages around the high wind system operation are discussed. The simulation results highlight the effective usage of FESS in reducing the peak overshoot of active power transients, smoothes the active power curves and helps in reducing the diesel consumption during the flywheel discharge period, without affecting the continuous power supply for meeting the instantaneous load demand.
DEFF Research Database (Denmark)
Li, H.; Zhao, B.; Han, L.
2010-01-01
In order to analyze correctly the effect of different models for induction generators on the transient performances of large wind power generation, Wind turbine driven squirrel cage induction generator (SCIG) models taking into account both main and leakage flux saturation and skin effect were...
International Nuclear Information System (INIS)
Nagarajan, Vijaisri; Chen, Yitung; Wang, Qiuwang; Ma, Ting
2014-01-01
Highlights: • Rip saw fin design is considered to be the best because it has thin fins and has higher heat transfer coefficient. • Minimum principal stress and maximum safety factor are obtained for the inverted bolt fin design. • Maximum principal stress and minimum safety factor are obtained for triangular fin design. • Thermal stress has significant impact than mechanical stress. • High principal stress is found at the startup and shutdown stage. - Abstract: In this study three-dimensional model of ceramic plate-fin high temperature heat exchanger with different fin designs and arrangements is analyzed numerically using ANSYS FLUENT and ANSYS structural module. The ability of ceramics to withstand high temperature and corrosion makes silicon carbide (SiC) suitable candidate material to be used in high temperature heat exchanger. The operating temperature of heat exchanger is 950 °C and the operating pressure is 1.5 MPa. The working fluids are helium, sulfur trioxide, sulfur dioxide, oxygen and the water vapor. Fluid flow and heat transfer analysis are carried out for steady and transient state in FLUENT. The obtained thermal and pressure load for the steady and transient state from ANSYS FLUENT are imported to ANSYS structural module to obtain the principal stress and the factor of safety. Different arrangements of rectangular fins, triangular fins, inverted bolt fins and ripsaw fins are studied. From the results it is found that the minimum stress and the maximum safety factor are obtained for inverted bolt fins. The triangular fins have the maximum principal stress and minimum factor of safety. However, the fluid flow and heat transfer analysis show inverted bolt fins and triangular fins produce higher pressure drop and friction factor. The steady state maximum principal stress is 10.08 MPa, 9.90 MPa and 11.43 MPa for straight, staggered and top and bottom ripsaw fin arrangement. The corresponding safety factors are 21.80, 21.95 and 19
Energy Technology Data Exchange (ETDEWEB)
Roberto, Thiago D., E-mail: thiagodbtr@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: mabs500@gmail.com [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)
2016-01-15
The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.
International Nuclear Information System (INIS)
Roberto, Thiago D.; Silva, Mário A. B. da; Lapa, Celso M.F.
2016-01-01
The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.
Energy Technology Data Exchange (ETDEWEB)
Luna, N. [Secretaria de Energia, Direccion de Operacion Petrolera, Mexico DF (Mexico); Mendez, F. [UNAM, Facultad de Ingenieria, Mexico DF (Mexico); Bautista, O. [ITESM, Division de Ingenieria y Arquitectura, Mexico DF (Mexico)
2005-05-01
We treat numerically in this paper, the transient analysis of a conjugated heat transfer process in the thermal entrance region of a circular tube with a fully developed laminar power-law fluid flow. We apply the quasi-steady approximation for the power-law fluid, identifying the suitable time scales of the process. Thus, the energy equation in the fluids is solved analytically using the well-known integral boundary layer technique. This solution is coupled to the transient energy equation for the solid where the transverse and longitudinal heat conduction effects are taken into account. The numerical results for the temporal evolution of the average temperature of the tube wall, {theta}{sub av,} is plotted for different nondimensional parameters such as conduction parameter, {alpha}, the aspect ratios of the tube, {epsilon} and {epsilon}{sub 0} and the index of power-law fluid, n. (orig.)
International Nuclear Information System (INIS)
Miles, K.J.; Hill, D.J.
1986-01-01
The DEFORM-4 module is the segment of the SAS4A Accident Analysis Code System that calculates the fuel pin characterization in response to a steady state irradiation history, thereby providing the initial conditions for the transient calculation. The various phenomena considered include fuel porosity migration, fission gas bubble induced swelling, fuel cracking and healing, fission gas release, cladding swelling, and the thermal-mechanical state of the fuel and cladding. In the transient state, the module continues the thermal-mechanical response calculation, including fuel melting and central cavity pressurization, until cladding failure is predicted and one of the failed fuel modules is initiated. Comparisons with experimental data have demonstrated the validity of the modeling approach
International Nuclear Information System (INIS)
Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman
1988-01-01
TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory
Energy Technology Data Exchange (ETDEWEB)
Muzhang, Peng; Quan, Zhang; Guoli, Wang; Yuman, Zhang
1988-03-01
TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory.
The analysis of transient noise of PCB P/G network based on PI/SI co-simulation
Haohang, Su
2018-02-01
With the frequency of the space camera become higher than before, the power noise of the imaging electronic system become the important factor. Much more power noise would disturb the transmissions signal, and even influence the image sharpness and system noise. "Target impedance method" is one of the traditional design method of P/G network (power and ground network), which is shorted of transient power noise analysis and often made "over design". In this paper, a new design method of P/G network is provided which simulated by PI/SI co-simulation. The transient power noise can be simulated and then applied in the design of noise reduction, thus effectively controlling the change of the noise in the P/G network. The method can efficiently control the number of adding decoupling capacitor, and is very efficient and feasible to keep the power integrity.
International Nuclear Information System (INIS)
Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki
2012-01-01
FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)
Fractal analysis: A new tool in transient volcanic ash plume characterization.
Tournigand, Pierre-Yves; Peña Fernandez, Juan Jose; Taddeucci, Jacopo; Perugini, Diego; Sesterhenn, Jörn
2017-04-01
Transient volcanic plumes are time-dependent features generated by unstable eruptive sources. They represent a threat to human health and infrastructures, and a challenge to characterize due to their intrinsic instability. Plumes have been investigated through physical (e.g. visible, thermal, UV, radar imagery), experimental and numerical studies in order to provide new insights about their dynamics and better anticipate their behavior. It has been shown experimentally that plume dynamics is strongly dependent to source conditions and that plume shape evolution holds key to retrieve these conditions. In this study, a shape evolution analysis is performed on thermal high-speed videos of volcanic plumes from three different volcanoes Sakurajima (Japan), Stromboli (Italy) and Fuego (Guatemala), recorded with a FLIR SC655 thermal camera during several field campaigns between 2012 and 2016. To complete this dataset, three numerical gas-jet simulations at different Reynolds number (2000, 5000 and 10000) have been used in order to set reference values to the natural cases. Turbulent flow shapes are well known to feature scale-invariant structures and a high degree of complexity. For this reason we characterized the bi-dimensional shape of natural and synthetic plumes by using a fractal descriptor. Such method has been applied in other studies on experimental turbulent jets as well as on atmospheric clouds and have shown promising results. At each time-step plume contour has been manually outlined and measured using the box-counting method. This method consists in covering the image with squares of variable sizes and counting the number of squares containing the plume outline. The negative slope of the number of squares in function of their size in a log-log plot gives the fractal dimension of the plume at a given time. Preliminary results show an increase over time of the fractal dimension for natural volcanic plume as well as for the numerically simulated ones, but at
International Nuclear Information System (INIS)
Dallman, R.J.; Gottula, R.C.; Holcomb, E.E.; Jouse, W.C.; Wagoner, S.R.; Wheatley, P.D.
1987-05-01
An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented
International Nuclear Information System (INIS)
Ellis, R.J.
1990-10-01
This report discusses the analysis of a postulated loss-of-regulation (LOR) accident in a metal-fuelled MAPLE Research Reactor. The selected transient scenario involves a slow LOR from low reactor power; the control rods are assumed to withdraw slowly until a trip at 12 MW halts the withdrawal. The simulation was performed using the space-time reactor kinetics computer code TANK, and modelling the reactor in detail in two dimensions and in two neutron-energy groups. Emphasis in this report is placed on the modelling techniques used in TANK and the physics considerations of the analysis
An improved local radial point interpolation method for transient heat conduction analysis
Wang, Feng; Lin, Gao; Zheng, Bao-Jing; Hu, Zhi-Qiang
2013-06-01
The smoothing thin plate spline (STPS) interpolation using the penalty function method according to the optimization theory is presented to deal with transient heat conduction problems. The smooth conditions of the shape functions and derivatives can be satisfied so that the distortions hardly occur. Local weak forms are developed using the weighted residual method locally from the partial differential equations of the transient heat conduction. Here the Heaviside step function is used as the test function in each sub-domain to avoid the need for a domain integral. Essential boundary conditions can be implemented like the finite element method (FEM) as the shape functions possess the Kronecker delta property. The traditional two-point difference method is selected for the time discretization scheme. Three selected numerical examples are presented in this paper to demonstrate the availability and accuracy of the present approach comparing with the traditional thin plate spline (TPS) radial basis functions.
An improved local radial point interpolation method for transient heat conduction analysis
International Nuclear Information System (INIS)
Wang Feng; Lin Gao; Hu Zhi-Qiang; Zheng Bao-Jing
2013-01-01
The smoothing thin plate spline (STPS) interpolation using the penalty function method according to the optimization theory is presented to deal with transient heat conduction problems. The smooth conditions of the shape functions and derivatives can be satisfied so that the distortions hardly occur. Local weak forms are developed using the weighted residual method locally from the partial differential equations of the transient heat conduction. Here the Heaviside step function is used as the test function in each sub-domain to avoid the need for a domain integral. Essential boundary conditions can be implemented like the finite element method (FEM) as the shape functions possess the Kronecker delta property. The traditional two-point difference method is selected for the time discretization scheme. Three selected numerical examples are presented in this paper to demonstrate the availability and accuracy of the present approach comparing with the traditional thin plate spline (TPS) radial basis functions
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)
A modal method for transient thermal analysis of CANDU fuel channel
International Nuclear Information System (INIS)
Park, J-W.; Muzumdar, A.J.
1996-01-01
The classical modal expansion technique has been applied to predict transient fuel and coolant temperatures under on-power conditions in a CANDU fuel channel. The temperature profile across the fuel pellet is assumed to be parabolic and fuel and coolant temperatures are expanded with Fourier series. The coefficient derivatives are written in state space form and solved by the Runge-Kutta method of fifth order. To validate the present model, the calculated fuel temperatures for several sample cases were compared with HOTSPOT-II, which employs a more rigorous finite-difference model. The agreement was found to be reasonable for the operational transients simulated. The advantage of the modal method is the fast computation speed for application to the real-time system such as the CANDU simulator which is being currently developed at the Institute for Advanced Engineering (IAE). (author)
Energy Technology Data Exchange (ETDEWEB)
Kot, C A; Youngdahl, C K
1978-09-01
PTAC was developed to predict pressure transients in nuclear-power-plant piping systems in which the possibility of cavitation must be considered. The program performs linear or nonlinear fluid-hammer calculations, using a fixed-grid method-of-characteristics solution procedure. In addition to pipe friction and elasticity, the program can treat a variety of flow components, pipe junctions, and boundary conditions, including arbitrary pressure sources and a sodium/water reaction. Essential features of transient cavitation are modeled by a modified column-separation technique. Comparisons of calculated results with available experimental data, for a simple piping arrangement, show good agreement and provide validation of the computational cavitation model. Calculations for a variety of piping networks, containing either liquid sodium or water, demonstrate the versatility of PTAC and clearly show that neglecting cavitation leads to erroneous predictions of pressure-time histories.
FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code
International Nuclear Information System (INIS)
Valtonen, Keijo; Hamalainen, Anitta; Cunningham, Mitchel E.
2002-01-01
The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.
FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code
Energy Technology Data Exchange (ETDEWEB)
Valtonen, Keijo (Radiation and Nuclear Safety Authority, Finland); Hamalainen, Anitta (VTT Energy, Finland); Cunningham, Mitchel E.(BATTELLE (PACIFIC NW LAB))
2002-05-01
The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.
International Nuclear Information System (INIS)
Hanson, J.M.
1984-12-01
The report evaluates previous investigations of the gas permeability of the rock surrounding emplacement holes at the Nevada Test Site. The discussion sets the framework from which the present uncertainty in gas permeability can be overcome. The usefulness of the barometric pressure testing method has been established. Flow models were used to evaluate barometric pressure transients taken at NTS holes U2fe, U19ac and U20ai. 31 refs., 103 figs., 18 tabs
Development of the computer code for transient analysis in experimental fast reactor
International Nuclear Information System (INIS)
Moreira, M.L.; Sato, E.F.
1989-01-01
A calculational model of heat transfer and fluid coolant dynamics, for thermal-hydraulic simulation of the primary system components of a pool type experimental fast breeder reactor, has developed. Programmed in FORTRAN, the SORES code was used to simulate transients as loss of pumping and loss of secondary sodium flow in the EBRII. The SORES results compared with measured data and NATDEMO code results was found to be good. (author) [pt
Computational Analysis of Effect of Transient Fluid Force on Composite Structures
2013-12-01
as they well represent an E-glass fiber reinforced composite frequently used in research and industrial applications. The fluid domain was sized...provide unique perspectives on peak stress ratios . The two models both share increased structural rigidity. The cylinder is reinforced by... Poisson ratio of 0.3 and Young’s modulus of 20 GPa were added to the transient structural engineering data cell (Figure 69). 78 Figure 69. E-Glass
Analysis of chemical warfare using a transient semi-Markov formulation.
Kierzewski, Michael O.
1988-01-01
Approved for public release; distribution is unlimited This thesis proposes an analytical model to test various assumptions about conventional/chemical warfare. A unit's status in conventional/chemical combat is modeled as states in a semi-Markov chain with transient and absorbing states. The effects of differing chemical threat levels, availability of decontamination assets and assumed personnel degradation rates on expected unit life and capabilities are tested. The ...
Smits, F.T.; Brouwer, H.J.; Schene, A.H.; van Weert, H.C.; ter Riet, G.
2016-01-01
OBJECTIVES: Frequent attenders (FAs) suffer more and consult general practitioners (GPs) more often for chronic physical and psychiatric illnesses, social difficulties and distress than non-FAs. However, it is unclear to what extent FAs present transient episodes of care (TECs) compared with
Temporal frequency probing for 5D transient analysis of global light transport
O'Toole, Matthew
2014-07-27
We analyze light propagation in an unknown scene using projectors and cameras that operate at transient timescales. In this new photography regime, the projector emits a spatio-temporal 3D signal and the camera receives a transformed version of it, determined by the set of all light transport paths through the scene and the time delays they induce. The underlying 3D-to-3D transformation encodes scene geometry and global transport in great detail, but individual transport components (e.g., direct reflections, inter-reflections, caustics, etc.) are coupled nontrivially in both space and time. To overcome this complexity, we observe that transient light transport is always separable in the temporal frequency domain. This makes it possible to analyze transient transport one temporal frequency at a time by trivially adapting techniques from conventional projector-to-camera transport. We use this idea in a prototype that offers three never-seen-before abilities: (1) acquiring time-of-flight depth images that are robust to general indirect transport, such as interreflections and caustics; (2) distinguishing between direct views of objects and their mirror reflection; and (3) using a photonic mixer device to capture sharp, evolving wavefronts of "light-in-flight".
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.
Numerical analysis of transient pressure variation in the condenser of a nuclear power station
Energy Technology Data Exchange (ETDEWEB)
Wang, Xinjun; Zhou, Zijie; Song, Zhao [Xi' an Jiaotong University, Xi' an (China); Lu, Qiankui; Li, Jiafu [Dong Fang Turbine Co., Ltd, Deyang (China)
2016-02-15
To research the characteristics of the transient variation of pressure in a nuclear power station condenser under accident condition, a mathematical model was established which simulated the cycling cooling water, heat transfer and pressure in the condenser. The calculation program of transient variation characteristics was established in Fortran language. The pump's parameter, cooling line's organization, check valve's feature and the parameter of siphonic water-collecting well are involved in the cooling water flow's mathematical model. The initial conditions of control volume are determined by the steady state of the condenser. The transient characteristics of a 1000 MW nuclear power station's condenser and cooling water system were examined. The results show that at the condition of plant-power suspension of pump, the cooling water flow rate decreases rapidly and refluxes, then fluctuates to 0. The variation of heat transfer coefficient in the condenser has three stages: at start it decreases sharply, then increases and decreases, and keeps constant in the end. Under three conditions (design, water and summer), the condenser pressure goes up in fluctuation. The time intervals between condenser's pressure signals under three conditions are about 26.4 s, which can fulfill the requirement for safe operation of nuclear power station.
Transient Three-Dimensional Side Load Analysis of a Film Cooled Nozzle
Wang, Ten-See; Guidos, Mike
2008-01-01
Transient three-dimensional numerical investigations on the side load physics for an engine encompassing a film cooled nozzle extension and a regeneratively cooled thrust chamber, were performed. The objectives of this study are to identify the three-dimensional side load physics and to compute the associated aerodynamic side load using an anchored computational methodology. The computational methodology is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, and a transient inlet history based on an engine system simulation. Ultimately, the computational results will be provided to the nozzle designers for estimating of effect of the peak side load on the nozzle structure. Computations simulating engine startup at ambient pressures corresponding to sea level and three high altitudes were performed. In addition, computations for both engine startup and shutdown transients were also performed for a stub nozzle, operating at sea level. For engine with the full nozzle extension, computational result shows starting up at sea level, the peak side load occurs when the lambda shock steps into the turbine exhaust flow, while the side load caused by the transition from free-shock separation to restricted-shock separation comes at second; and the side loads decreasing rapidly and progressively as the ambient pressure decreases. For the stub nozzle operating at sea level, the computed side loads during both startup and shutdown becomes very small due to the much reduced flow area.
Analysis of loss of offsite power transient using RELAP5/MOD1/NSC
International Nuclear Information System (INIS)
Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo
1986-01-01
System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAP5/MOD1/NSC), based upon the sequence of events for the KNU1( Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximate the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV setting etc. are recognized to be important in the transient analyses on a bestestimate basis. (Author)
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.
Directory of Open Access Journals (Sweden)
Miroslav M Živković
2010-01-01
Full Text Available This paper deals with transient nonlinear heat conduction through the insulation wall of the tank for transportation of liquid aluminum. Tanks designed for this purpose must satisfy certain requirements regarding temperature of loading and unloading, during transport. Basic theoretical equations are presented, which describe the problem of heat conduction finite element (FE analysis, starting from the differential equation of energy balance, taking into account the initial and boundary conditions of the problem. General 3D problem for heat conduction is considered, from which solutions for two- and one-dimensional heat conduction can be obtained, as special cases. Forming of the finite element matrices using Galerkin method is briefly described. The procedure for solving equations of energy balance is discussed, by methods of resolving iterative processes of nonlinear transient heat conduction. Solution of this problem illustrates possibilities of PAK-T software package, such as materials properties, given as tabular data, or analytical functions. Software also offers the possibility to solve nonlinear and transient problems with incremental methods. Obtained results for different thicknesses of the tank wall insulation materials enable its comparison in regards to given conditions
Directory of Open Access Journals (Sweden)
Paulo Antonio Delgado-Arredondo
2015-01-01
Full Text Available Induction motors are critical components for most industries and the condition monitoring has become necessary to detect faults. There are several techniques for fault diagnosis of induction motors and analyzing the startup transient vibration signals is not as widely used as other techniques like motor current signature analysis. Vibration analysis gives a fault diagnosis focused on the location of spectral components associated with faults. Therefore, this paper presents a comparative study of different time-frequency analysis methodologies that can be used for detecting faults in induction motors analyzing vibration signals during the startup transient. The studied methodologies are the time-frequency distribution of Gabor (TFDG, the time-frequency Morlet scalogram (TFMS, multiple signal classification (MUSIC, and fast Fourier transform (FFT. The analyzed vibration signals are one broken rotor bar, two broken bars, unbalance, and bearing defects. The obtained results have shown the feasibility of detecting faults in induction motors using the time-frequency spectral analysis applied to vibration signals, and the proposed methodology is applicable when it does not have current signals and only has vibration signals. Also, the methodology has applications in motors that are not fed directly to the supply line, in such cases the analysis of current signals is not recommended due to poor current signal quality.
International Nuclear Information System (INIS)
Nie, Ren-Shi; Guo, Jian-Chun; Jia, Yong-Lu; Zhu, Shui-Qiao; Rao, Zheng; Zhang, Chun-Guang
2011-01-01
The no-type curve with negative skin of a horizontal well has been found in the current research. Negative skin is very significant to transient well test and rate decline analysis. This paper first presents the negative skin problem where the type curves with negative skin of a horizontal well are oscillatory. In order to solve the problem, we propose a new model of transient well test and rate decline analysis for a horizontal well in a multiple-zone composite reservoir. A new dimensionless definition of r D is introduced in the dimensionless mathematical modelling under different boundaries. The model is solved using the Laplace transform and separation of variables techniques. In Laplace space, the solutions for both constant rate production and constant wellbore pressure production are expressed in a unified formula. We provide graphs and thorough analysis of the new standard type curves for both well test and rate decline analysis; the characteristics of type curves are the reflections of horizontal well production in a multiple-zone reservoir. An important contribution of our paper is that our model removed the oscillation in type curves and thus solved the negative skin problem. We also show that the characteristics of type curves depend heavily on the properties of different zones, skin factor, well length, formation thickness, etc. Our research can be applied to a real case study
International Nuclear Information System (INIS)
Vittal, V.
2000-01-01
The electric utility industry is undergoing unprecedented changes in its structure worldwide. With the advent of an open market environment and competition in the industry, and restructuring of the industry into separate generation, transmission, and distribution entities, new issues in power system operation and planning are inevitable. One of the major consequences of this new electric utility environment is the greater emphasis on reliability and secure operation of the power system. This paper examines the impact of restructuring on power system dynamic analysis. It specifically addresses issues related to transient stability analysis and small-signal stability analysis. Four major topics to examine the effect on the nature of studies conducted are considered. These topics are (1) system adequacy and security, (2) system modeling data requirements, (3) system protection and control, and (4) system restoration. The consequences and impact of each of these topics on the nature of the studies conducted are examined and discussed. The emphasis on greater reliability has led to a clearer enunciation of standards, measurements, and guides in some countries. These requirements will result in: (1) more measurements on existing systems, (2) rigorous analysis of transient stability and small-signal stability to determine operating limits and plan systems, (3) greater emphasis on studies to verify coordination and proper performance of protection and controls, and (4) development of a detailed plan for system restoration in the case of wide-spread outages
Liu, Fang; Shen, Changqing; He, Qingbo; Zhang, Ao; Liu, Yongbin; Kong, Fanrang
2014-01-01
A fault diagnosis strategy based on the wayside acoustic monitoring technique is investigated for locomotive bearing fault diagnosis. Inspired by the transient modeling analysis method based on correlation filtering analysis, a so-called Parametric-Mother-Doppler-Wavelet (PMDW) is constructed with six parameters, including a center characteristic frequency and five kinematic model parameters. A Doppler effect eliminator containing a PMDW generator, a correlation filtering analysis module, and a signal resampler is invented to eliminate the Doppler effect embedded in the acoustic signal of the recorded bearing. Through the Doppler effect eliminator, the five kinematic model parameters can be identified based on the signal itself. Then, the signal resampler is applied to eliminate the Doppler effect using the identified parameters. With the ability to detect early bearing faults, the transient model analysis method is employed to detect localized bearing faults after the embedded Doppler effect is eliminated. The effectiveness of the proposed fault diagnosis strategy is verified via simulation studies and applications to diagnose locomotive roller bearing defects. PMID:24803197
International Nuclear Information System (INIS)
Muir, M.D.
1975-01-01
The design and design philosophy of a high performance, extremely versatile transient analyzer is described. This sub-system was designed to be controlled through the data acquisition computer system which allows hands off operation. Thus it may be placed on the experiment side of the high voltage safety break between the experimental device and the control room. This analyzer provides control features which are extremely useful for data acquisition from PPPL diagnostics. These include dynamic sample rate changing, which may be intermixed with multiple post trigger operations with variable length blocks using normal, peak to peak or integrate modes. Included in the discussion are general remarks on the advantages of adding intelligence to transient analyzers, a detailed description of the characteristics of the PPPL transient analyzer, a description of the hardware, firmware, control language and operation of the PPPL transient analyzer, and general remarks on future trends in this type of instrumentation both at PPPL and in general
International Nuclear Information System (INIS)
Bae, B. U.; Park, Y. S.; Kim, J. R.; Kang, K. H.; Choi, K. Y.; Sung, H. J.; Hwang, M. J.; Kang, D. H.; Lim, S. G.; Jun, S. S.
2015-01-01
Participants of DSP-03 were divided in three groups and each group has focused on the specific subject related to the enhancement of the code analysis. The group A tried to investigate scaling capability of ATLAS test data by comparing to the code analysis for a prototype, and the group C studied to investigate effect of various models in the one-dimensional codes. This paper briefly summarizes the code analysis result from the group B participants in the DSP-03 of the ATLAS test facility. The code analysis by Group B focuses highly on investigating the multi-dimensional thermal hydraulic phenomena in the ATLAS facility during the SLB transient. Even though the one-dimensional system analysis code cannot simulate the whole system of the ATLAS facility with a nodalization of the CFD (Computational Fluid Dynamics) scale, a reactor pressure vessel can be considered with multi-dimensional components to reflect the thermal mixing phenomena inside a downcomer and a core. Also, the CFD could give useful information for understanding complex phenomena in specific components such as the reactor pressure vessel. From the analysis activity of Group B in ATLAS DSP-03, participants adopted a multi-dimensional approach to the code analysis for the SLB transient in the ATLAS test facility. The main purpose of the analysis was to investigate prediction capability of multi-dimensional analysis tools for the SLB experiment result. In particular, the asymmetric cooling and thermal mixing phenomena in the reactor pressure vessel could be significantly focused for modeling the multi-dimensional components
Energy Technology Data Exchange (ETDEWEB)
Zheng, Xiaoqing; Cheng, Zeng [Department of Electrical and Computer Engineering, McMaster University (Canada); Deen, M. Jamal, E-mail: jamal@mcmaster.ca [Department of Electrical and Computer Engineering, McMaster University (Canada); School of Biomedical Engineering, McMaster University (Canada); Peng, Hao, E-mail: penghao@mcmaster.ca [Department of Electrical and Computer Engineering, McMaster University (Canada); School of Biomedical Engineering, McMaster University (Canada); Department of Medical Physics, McMaster University, Ontario L8S 4K1, Hamilton (Canada)
2016-02-01
Cadmium Zinc Telluride (CZT) semiconductor detectors are capable of providing superior energy resolution and three-dimensional position information of gamma ray interactions in a large variety of fields, including nuclear physics, gamma-ray imaging and nuclear medicine. Some dedicated Positron Emission Tomography (PET) systems, for example, for breast cancer detection, require higher contrast recovery and more accurate event location compared with a whole-body PET system. The spatial resolution is currently limited by electrode pitch in CZT detectors. A straightforward approach to increase the spatial resolution is by decreasing the detector electrode pitch, but this leads to higher fabrication cost and a larger number of readout channels. In addition, inter-electrode charge spreading can negate any improvement in spatial resolution. In this work, we studied the feasibility of achieving sub-pitch spatial resolution in CZT detectors using two methods: charge sharing effect and transient signal analysis. We noted that their valid ranges of usage were complementary. The dependences of their corresponding valid ranges on electrode design, depth-of-interaction (DOI), voltage bias and signal triggering threshold were investigated. The implementation of these two methods in both pixelated and cross-strip configuration of CZT detectors were discussed. Our results show that the valid range of charge sharing effect increases as a function of DOI, but decreases with increasing gap width and bias voltage. For a CZT detector of 5 mm thickness, 100 µm gap and biased at 400 V, the valid range of charge sharing effect was found to be about 112.3 µm around the gap center. This result complements the valid range of the transient signal analysis within one electrode pitch. For a signal-to-noise ratio (SNR) of ~17 and preliminary measurements, the sub-pitch spatial resolution is expected to be ~30 µm and ~250 µm for the charge sharing and transient signal analysis methods
International Nuclear Information System (INIS)
Billon, F.; David, J.; Procaccia, H.
1983-01-01
The operating efficiency of steam generators (S.G.s) and their structural integrity depend on the design configurations of the feedwater spray within the S.G., and on the operating procedure. To check the merit of some design modifications, and to verify the fluid-structure interaction with a view to preserve the S.G.s integrity during severe operating transients, a special instrumentation that admits the determination of the instantaneous thermal hydraulic characteristics of the flow in the secondary water and the S.G. tube sheet, has been installed by EDF on one steam generator of Tricastin unit 1 power plant. In parallel, FRAMATOME has developped a computer code, TEMPTRON, that allows the calculations of the thermal loads and the consequent stresses in the most sollicited zones of the steam generator during transient operation of the plant. This code divides the S.G. into three parts: - the first concerns the S.G.s region above the downcomer, zone where the mixing between hot water and cold feedwater occurs, - the second is the downcomer itself which is divided into n segments, - the third concerns the tube sheet zone which is also divided into n segments. The most severe transient test performed is the auxiliary cold feedwater injection into the steam generator during a hot standby of the plant: two levels of flow rate have been realised: 55 and 110 m 3 /h of 42 0 C feedwater. The tests have shown that if the cold feedwater injection occurs when the steam generator water level is below feedwater ring, the lowest fluid temperature reached at tube sheet inlet is about 230 0 C. (orig.)
Transient Side Load Analysis of Out-of-Round Film-Cooled Nozzle Extensions
Wang, Ten-See; Lin, Jeff; Ruf, Joe; Guidos, Mike
2012-01-01
There was interest in understanding the impact of out-of-round nozzle extension on the nozzle side load during transient startup operations. The out-of-round nozzle extension could be the result of asymmetric internal stresses, deformation induced by previous tests, and asymmetric loads induced by hardware attached to the nozzle. The objective of this study was therefore to computationally investigate the effect of out-of-round nozzle extension on the nozzle side loads during an engine startup transient. The rocket engine studied encompasses a regeneratively cooled chamber and nozzle, along with a film cooled nozzle extension. The computational methodology is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, and transient inlet boundary flow properties derived from an engine system simulation. Six three-dimensional cases were performed with the out-of-roundness achieved by three different degrees of ovalization, elongated on lateral y and z axes: one slightly out-of-round, one more out-of-round, and one significantly out-of-round. The results show that the separation line jump was the primary source of the peak side loads. Comparing to the peak side load of the perfectly round nozzle, the peak side loads increased for the slightly and more ovalized nozzle extensions, and either increased or decreased for the two significantly ovalized nozzle extensions. A theory based on the counteraction of the flow destabilizing effect of an exacerbated asymmetrical flow caused by a lower degree of ovalization, and the flow stabilizing effect of a more symmetrical flow, created also by ovalization, is presented to explain the observations obtained in this effort.
Transient Three-Dimensional Side Load Analysis of Out-of-Round Film Cooled Nozzles
Wang, Ten-See; Lin, Jeff; Ruf, Joe; Guidos, Mike
2010-01-01
The objective of this study is to investigate the effect of nozzle out-of-roundness on the transient startup side loads at a high altitude, with an anchored computational methodology. The out-of-roundness could be the result of asymmetric loads induced by hardware attached to the nozzle, asymmetric internal stresses induced by previous tests, and deformation, such as creep, from previous tests. The rocket engine studied encompasses a regeneratively cooled thrust chamber and a film cooled nozzle extension with film coolant distributed from a turbine exhaust manifold. The computational methodology is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, and a transient inlet history based on an engine system simulation. Transient startup computations were performed with the out-of-roundness achieved by four different degrees of ovalization: one perfectly round, one slightly out-of-round, one more out-of-round, and one significantly out-of-round. The results show that the separation-line-jump is the peak side load physics for the round, slightly our-of-round, and more out-of-round cases, and the peak side load increases as the degree of out-of-roundness increases. For the significantly out-of-round nozzle, however, the peak side load reduces to comparable to that of the round nozzle and the separation line jump is not the peak side load physics. The counter-intuitive result of the significantly out-of-round case is found to be related to a side force reduction mechanism that splits the effect of the separation-line-jump into two parts, not only in the circumferential direction and most importantly in time.
Neural control and transient analysis of the LCL-type resonant converter
Zouggar, S.; Nait Charif, H.; Azizi, M.
2000-07-01
This paper proposes a generalised inverse learning structure to control the LCL converter. A feedforward neural network is trained to act as an inverse model of the LCL converter then both are cascaded such that the composed system results in an identity mapping between desired response and the LCL output voltage. Using the large signal model, we analyse the transient output response of the controlled LCL converter in the case of large variation of the load. The simulation results show the efficiency of using neural networks to regulate the LCL converter.
D'Isanto, A.; Cavuoti, S.; Brescia, M.; Donalek, C.; Longo, G.; Riccio, G.; Djorgovski, S. G.
2016-04-01
The exploitation of present and future synoptic (multiband and multi-epoch) surveys requires an extensive use of automatic methods for data processing and data interpretation. In this work, using data extracted from the Catalina Real Time Transient Survey (CRTS), we investigate the classification performance of some well tested methods: Random Forest, MultiLayer Perceptron with Quasi Newton Algorithm and K-Nearest Neighbours, paying special attention to the feature selection phase. In order to do so, several classification experiments were performed. Namely: identification of cataclysmic variables, separation between galactic and extragalactic objects and identification of supernovae.
Dynamic transient analysis of rupture disks by the finite-element method
International Nuclear Information System (INIS)
Hsieh, B.J.
1975-02-01
A finite element method utilizing the principle of virtual work in convected coordinates is used to analyze the axisymmetric dynamic transient response of rupture disks. This method can treat non-linearities arising both from inelastic material properties and large displacements/rotations provided that the convected strains are small. This report contains extensive calculations using a variety of rupture disk geometries and attempts to relate the static buckling of such disks to their dynamic response characteristics. A majority of the calculations treat the response of 18 inch disks typical of those currently considered for use in the Clinch River Breeder Reactor intermediate heat transport system
Verification of a neutronic code for transient analysis in reactors with Hex-z geometry
Energy Technology Data Exchange (ETDEWEB)
Gonzalez-Pintor, S.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain); Ginestar, D. [Departamento de Matematica Aplicada, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain)
2012-07-01
Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)
Sayed, Sadeed Bin
2016-11-02
An explicit marching on-in-time scheme for analyzing transient electromagnetic wave interactions on ferromagnetic scatterers is described. The proposed method solves a coupled system of time domain magnetic field volume integral and Landau-Lifshitz-Gilbert (LLG) equations. The unknown fluxes and fields are discretized using full and half Schaubert-Wilton-Glisson functions in space and bandlimited temporal interpolation functions in time. The coupled system is cast in the form of an ordinary differential equation and integrated in time using a PE(CE)m type linear multistep method to obtain the unknown expansion coefficients. Numerical results demonstrating the stability and accuracy of the proposed scheme are presented.
Two-dimensional transient far-field analysis for the excess temperature from an arbitrary source
Energy Technology Data Exchange (ETDEWEB)
Witten, A.J.; Long, E.C.
1978-07-01
An analytic solution is presented for the two-dimensional time-dependent advective diffusion equation governing the distribution of excess temperature in a river of uniform width, depth, and downstream flow. The solution is also applicable to a straight coastline with uniform longshore flow. Exact solutions are obtained for a point heat source and a particular line heat source, while an approximate representation is given for an arbitrary time-varying heat source. These solutions are incorporated into a computer program which calculates excess temperature and time rate-of-change of excess temperature in a river or coast as a result of waste heat discharged from various transient sources.
Sayed, Sadeed Bin; Ulku, Huseyin Arda; Bagci, Hakan
2016-01-01
An explicit marching on-in-time scheme for analyzing transient electromagnetic wave interactions on ferromagnetic scatterers is described. The proposed method solves a coupled system of time domain magnetic field volume integral and Landau-Lifshitz-Gilbert (LLG) equations. The unknown fluxes and fields are discretized using full and half Schaubert-Wilton-Glisson functions in space and bandlimited temporal interpolation functions in time. The coupled system is cast in the form of an ordinary differential equation and integrated in time using a PE(CE)m type linear multistep method to obtain the unknown expansion coefficients. Numerical results demonstrating the stability and accuracy of the proposed scheme are presented.
Fracture mechanical analysis of relevant transients in the pressure vessel of Atucha I reactor
International Nuclear Information System (INIS)
Saavedra, Fernando M.
2001-01-01
The evolution of the applied stress intensity factor K I for 10 relevant transients of the nuclear power station Atucha I obtained from thermohydraulic data is analyzed according to the methodology proposed in Section XI of ASME Boiler and Pressure Vessel Code. Vast knowledge was thus obtained about basic concepts of fracture mechanics and its application to remanent life of nuclear components. Basic knowledge which commands the performance of nuclear power stations was also obtained, especially that related to the Atucha I utility [es
Analysis of a control rod ejection transient in a mox-fuelled PWR
International Nuclear Information System (INIS)
Lenain, R.; Mathonniere, G.; Perrutel, J.P.; Schaeffer, H.; Stelletta, S.; Lam Hime, M.
1988-09-01
The decision to use mixed-oxide (MOX) fuel in PWR's involved re-investigation of a certain number of accidents and notably control rod ejection transients. It has thus been shown that this accident would be no more severe than in the case of all-uranium cores, since the positive effects on the ejected rod worth would counterbalance the negative effects on the delayed neutron fraction. A new approach to the kinetics aspect of the calculation method for this accident is also presented, involving a 3-D kinetic calculation with only a few axial meshes
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.
FRAP-T6: a computer code for the transient analysis of oxide fuel rods
International Nuclear Information System (INIS)
Siefken, L.J.; Shah, V.N.; Berna, G.A.; Hohorst, J.K.
1983-06-01
FRAP-T6 is a computer code which is being developed to calculate the transient behavior of a light water reactor fuel rod. This report is an addendum to the FRAP-T6/MODO user's manual which provides the additional user information needed to use FRAP-T6/MOD1. This includes model changes, improvements, and additions, coding changes and improvements, change in input and control language, and example problem solutions to aid the user. This information is designed to supplement the FRAP-T6/MODO user's manual
International Nuclear Information System (INIS)
Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.
1996-01-01
The simplified boiling water reactor (SBWR) operating in natural circulation is designed with many passive safety features. An anticipated transient without scram (ATWS) initiated by inadvertent closure of the main steam isolation valve (MSIV) in an SBWR has been analyzed using the RAMONA-4B code of Brookhaven National Laboratory. This analysis demonstrates the predicted performance of the SBWR during an MSIV closure ATWS, followed by shutdown of the reactor through injection of boron into the reactor core from the standby liquid control system
International Nuclear Information System (INIS)
Soares, P.A.; Sirimarco, L.F.; Veloso, M.A.F.
1979-03-01
SACI-O is a computer code which deals with the dynamics of the core of pressurized light water reactors (PWR). Its applicability is determined by the evaluation of the models used in the simulation of the several phenomena and processes which occur in the core during transients. This report presents a comparison between the results obtained with SACI-O and those presented in the Final Safety Analysis Report (FSAR) of Angra dos Reis Nuclear Station, Unit 1. Although some data used in the calculations done by Westinghouse are not known, there was a good agreement between the mentioned results. (Author) [pt
Transient analysis of unbalanced short circuits of the ERDA-NASA 100 kW wind turbine alternator
Hwang, H. H.; Gilbert, L. J.
1976-01-01
Unbalanced short-circuit faults on the alternator of the ERDA-NASA Mod-O100-kW experimental wind turbine are studied. For each case, complete solutions for armature, field, and damper-circuit currents; short-circuit torque; and open-phase voltage are derived directly by a mathematical analysis. Formulated results are tabulated. For the Mod-O wind turbine alternator, numerical calculations are given, and results are presented by graphs. Comparisons for significant points among the more important cases are summarized. For these cases the transients are found to be potentially severe. The effect of the alternator neutral-to-ground impedance is evaluated.
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)
Transient analysis of air-water two-phase flow in channels and bends
International Nuclear Information System (INIS)
Khan, H.J.; Ye, W.; Pertmer, G.A.
1992-01-01
The algorithm used in this paper is the Newton Block Gauss Seidel method, which has been applied to both simple and complex flow conditions in two-phase flow. This paper contains a description of difference techniques and an iterative solution algorithm that is used to solve the field and constitutive equations of the two-fluid model. In practice, this solution procedure has been proven to be stable and capable of generating solutions in problems where other schemes have failed. The method converges rapidly for reasonable error tolerances and is easily extended to three-dimensional geometries. Using air-water as the two-phase medium, transient flow behavior in several geometries of interest are shown. Flow through a vertical channel with flow obstruction, large U bends, and 90-deg bends are being demonstrated with variation of inlet void fraction and slip ratio. Significant changes in the velocity and void distribution profiles have been observed. Various regions of flow recirculation are obtained in the flow domain for each phase. The phasic velocity and void distributions are dominated by gravity-induced phase separation causing air to accumulate in the upper region. The influence of inlet slip ratio and interfacial momentum transfer on the transient flow profile has been demonstrated in detail
Energy Technology Data Exchange (ETDEWEB)
Mochida, A.; Kudo, K.; Higasa, H.
1999-07-01
In the present study, transient temperature rise is analyzed in a flywheel type power storage system operated in vacuum environment. The flywheel rotor is levitated by high-temperature-superconducting magnetic bearing to reduce the bearing loss. Though the superconductor is cooled by liquid nitrogen, the temperature of the whole system rises due to Joule heating in the coils of the bearings and the motor during the operation. If the temperature should reach the critical temperature of the permanent magnet used for the magnetic bearings after long time operation, the magnetic bearings lose their effect. The heat generated in the levitated rotor diffuses within it by heat conduction and finally emitted to its surrounding solid materials by thermal radiation from the rotor surfaces across vacuum layer. Numerical simulation is carried out calculating the transient radiative-conductive heat transfer and time-dependent profiles of temperature within the rotor are obtained. The results are compared with the experimentally obtained temperatures by measured a test model of 1kWh power storage and the measured profiles of the temperature rise of the rotor fit very well with the calculated ones. Using this simulation tool, the effects of the surface emissivity of the materials of the rotor and the stator, the temperature of the surrounding casings and the thermal conductivity of the materials on the temperature profiles in the system are estimated.
Development of a reactivity worth correction scheme for the one-dimensional transient analysis
International Nuclear Information System (INIS)
Cho, J. Y.; Song, J. S.; Joo, H. G.; Kim, H. Y.; Kim, K. S.; Lee, C. C.; Zee, S. Q.
2003-11-01
This work is to develop a reactivity worth correction scheme for the MASTER one-dimensional (1-D) calculation model. The 1-D cross section variations according to the core state in the MASTER input file, which are produced for 1-D calculation performed by the MASTER code, are incorrect in most of all the core states except for exactly the same core state where the variations are produced. Therefore this scheme performs the reactivity worth correction factor calculations before the main 1-D transient calculation, and generates correction factors for boron worth, Doppler and moderator temperature coefficients, and control rod worth, respectively. These correction factors force the one dimensional calculation to generate the same reactivity worths with the 3-dimensional calculation. This scheme is applied to the control bank withdrawal accident of Yonggwang unit 1 cycle 14, and the performance is examined by comparing the 1-D results with the 3-D results. This problem is analyzed by the RETRAN-MASTER consolidated code system. Most of all results of 1-D calculation including the transient power behavior, the peak power and time are very similar with the 3-D results. In the MASTER neutronics computing time, the 1-D calculation including the correction factor calculation requires the negligible time comparing with the 3-D case. Therefore, the reactivity worth correction scheme is concluded to be very good in that it enables the 1-D calculation to produce the very accurate results in a few computing time
Studies of implicit and explicit solution techniques in transient thermal analysis of structures
International Nuclear Information System (INIS)
Adelman, H.M.; Haftka, R.T.; Robinson, J.C.
1982-08-01
Studies aimed at an increase in the efficiency of calculating transient temperature fields in complex aerospace vehicle structures are reported. The advantages and disadvantages of explicit and implicit algorithms are discussed and a promising set of implicit algorithms with variable time steps, known as GEARIB, is described. Test problems, used for evaluating and comparing various algorithms, are discussed and finite element models of the configurations are described. These problems include a coarse model of the Space Shuttle wing, an insulated frame test article, a metallic panel for a thermal protection system, and detailed models of sections of the Space Shuttle wing. Results generally indicate a preference for implicit over explicit algorithms for transient structural heat transfer problems when the governing equations are stiff (typical of many practical problems such as insulated metal structures). The effects on algorithm performance of different models of an insulated cylinder are demonstrated. The stiffness of the problem is highly sensitive to modeling details and careful modeling can reduce the stiffness of the equations to the extent that explicit methods may become the best choice. Preliminary applications of a mixed implicit-explicit algorithm and operator splitting techniques for speeding up the solution of the algebraic equations are also described
Studies of implicit and explicit solution techniques in transient thermal analysis of structures
Adelman, H. M.; Haftka, R. T.; Robinson, J. C.
1982-01-01
Studies aimed at an increase in the efficiency of calculating transient temperature fields in complex aerospace vehicle structures are reported. The advantages and disadvantages of explicit and implicit algorithms are discussed and a promising set of implicit algorithms with variable time steps, known as GEARIB, is described. Test problems, used for evaluating and comparing various algorithms, are discussed and finite element models of the configurations are described. These problems include a coarse model of the Space Shuttle wing, an insulated frame tst article, a metallic panel for a thermal protection system, and detailed models of sections of the Space Shuttle wing. Results generally indicate a preference for implicit over explicit algorithms for transient structural heat transfer problems when the governing equations are stiff (typical of many practical problems such as insulated metal structures). The effects on algorithm performance of different models of an insulated cylinder are demonstrated. The stiffness of the problem is highly sensitive to modeling details and careful modeling can reduce the stiffness of the equations to the extent that explicit methods may become the best choice. Preliminary applications of a mixed implicit-explicit algorithm and operator splitting techniques for speeding up the solution of the algebraic equations are also described.
FRAP-T1: a computer code for the transient analysis of oxide fuel rods
International Nuclear Information System (INIS)
Dearien, J.A.; Miller, R.L.; Hobbins, R.R.; Siefken, L.J.; Baston, V.F.; Coleman, D.R.
1977-02-01
FRAP-T is a FORTRAN IV computer code which can be used to solve for the transient response of a light water reactor (LWR) fuel rod during accident transients such as loss-of-coolant accident (LOCA) or a power-cooling-mismatch (PCM). The coupled effects of mechanical, thermal, internal gas, and material property response on the behavior of the fuel rod are considered. FRAP-T is a modular code with each major computational model isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAP-T is coupled to a materials properties subcode (MATPRO) which is used to provide gas, fuel, and cladding properties to the FRAP-T computational subcodes. No material properties need be supplied by the code user. The needed water properties are stored in tables built into the code. Critical heat flux (CHF) and heat transfer correlations for a wide range of coolant conditions are contained in modular subroutines. FRAP-T has been evaluated by making extensive comparisons between predictions of the code and experimental data. Comparison of predicted and experimental results are presented for a range of FRAP-T calculated parameters. The code is presently programmed and running on an IBM-360/75 and a CDC 7600 computer
Erdel, Fabian; Rippe, Karsten
2012-11-20
Interactions between nuclear proteins and chromatin frequently occur on the time scale of seconds and below. These transient binding events are important for the fast identification of target sites as concluded from our previous analysis of the human chromatin remodelers Snf2H and Snf2L from the imitation switch (ISWI) family. Both ATP-driven molecular motor proteins are able to translocate nucleosomes along the DNA and appear to exert this activity only on a small number of nucleosomes to which they bind more tightly. For mechanistic studies, one needs to distinguish such translocation reactions or other long-lived interactions associated with conformational changes and/or ATP hydrolysis from nonproductive chromatin sampling during target search. These processes can be separated by measuring the duration of nucleosome binding with subsecond time resolution. To reach this goal, we have developed a fluorescence bleaching technique termed pixel-wise photobleaching profile evolution analysis (3PEA). It exploits the inherent time structure of confocal microscopy images and yields millisecond resolution. 3PEA represents a generally applicable approach to quantitate transient chromatin interactions in the 2- to 500-ms time regime within only ∼1 s needed for a measurement. The green autofluorescent protein (GFP)-tagged Snf2H and Snf2L and the inactive Snf2L+13 splice variant were studied by 3PEA in comparison to the isolated GFP or red autofluorescent protein and a GFP pentamer. Our results reveal that the residence time for transient chromatin binding of Snf2H and Snf2L is <2 ms, and strongly support the view that ISWI-type remodelers are only rarely active in unperturbed cells during G1 phase.
Energy Technology Data Exchange (ETDEWEB)
Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear
2017-11-01
A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)
Jamison, J. W.
1994-01-01
CFORM was developed by the Kennedy Space Center Robotics Lab to assist in linear control system design and analysis using closed form and transient response mechanisms. The program computes the closed form solution and transient response of a linear (constant coefficient) differential equation. CFORM allows a choice of three input functions: the Unit Step (a unit change in displacement); the Ramp function (step velocity); and the Parabolic function (step acceleration). It is only accurate in cases where the differential equation has distinct roots, and does not handle the case for roots at the origin (s=0). Initial conditions must be zero. Differential equations may be input to CFORM in two forms - polynomial and product of factors. In some linear control analyses, it may be more appropriate to use a related program, Linear Control System Design and Analysis (KSC-11376), which uses root locus and frequency response methods. CFORM was written in VAX FORTRAN for a VAX 11/780 under VAX VMS 4.7. It has a central memory requirement of 30K. CFORM was developed in 1987.
SAFE: A computer code for the steady-state and transient thermal analysis of LMR fuel elements
International Nuclear Information System (INIS)
Hayes, S.L.
1993-12-01
SAFE is a computer code developed for both the steady-state and transient thermal analysis of single LMR fuel elements. The code employs a two-dimensional control-volume based finite difference methodology with fully implicit time marching to calculate the temperatures throughout a fuel element and its associated coolant channel for both the steady-state and transient events. The code makes no structural calculations or predictions whatsoever. It does, however, accept as input structural parameters within the fuel such as the distributions of porosity and fuel composition, as well as heat generation, to allow a thermal analysis to be performed on a user-specified fuel structure. The code was developed with ease of use in mind. An interactive input file generator and material property correlations internal to the code are available to expedite analyses using SAFE. This report serves as a complete design description of the code as well as a user's manual. A sample calculation made with SAFE is included to highlight some of the code's features. Complete input and output files for the sample problem are provided
International Nuclear Information System (INIS)
Perkó, Zoltán; Lathouwers, Danny; Kloosterman, Jan Leen; Hagen, Tim van der
2014-01-01
Highlights: • Grid and basis adaptive Polynomial Chaos techniques are presented for S and U analysis. • Dimensionality reduction and incremental polynomial order reduce computational costs. • An unprotected loss of flow transient is investigated in a Gas Cooled Fast Reactor. • S and U analysis is performed with MC and adaptive PC methods, for 42 input parameters. • PC accurately estimates means, variances, PDFs, sensitivities and uncertainties. - Abstract: Since the early years of reactor physics the most prominent sensitivity and uncertainty (S and U) analysis methods in the nuclear community have been adjoint based techniques. While these are very effective for pure neutronics problems due to the linearity of the transport equation, they become complicated when coupled non-linear systems are involved. With the continuous increase in computational power such complicated multi-physics problems are becoming progressively tractable, hence affordable and easily applicable S and U analysis tools also have to be developed in parallel. For reactor physics problems for which adjoint methods are prohibitive Polynomial Chaos (PC) techniques offer an attractive alternative to traditional random sampling based approaches. At TU Delft such PC methods have been studied for a number of years and this paper presents a large scale application of our Fully Adaptive Non-Intrusive Spectral Projection (FANISP) algorithm for performing the sensitivity and uncertainty analysis of a Gas Cooled Fast Reactor (GFR) Unprotected Loss Of Flow (ULOF) transient. The transient was simulated using the Cathare 2 code system and a fully detailed model of the GFR2400 reactor design that was investigated in the European FP7 GoFastR project. Several sources of uncertainty were taken into account amounting to an unusually high number of stochastic input parameters (42) and numerous output quantities were investigated. The results show consistently good performance of the applied adaptive PC
Transient Torsional Analysis of a Belt Conveyor Drive with Pneumatic Flexible Shaft Coupling
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Kaššay Peter
2017-03-01
Full Text Available Development and application of pneumatic flexible shaft couplings have been in the center of our department research activities for a long time. These couplings are able to change torsional stiffness by changing pressure in their flexible elements – air bellows. Until now we have dealt with the use of pneumatic flexible shaft couplings for tuning mechanical systems working with periodically alternating load torque at steady state. Some mechanical systems, however, operate with a static load torque at constant speed (e.g. hoists, elevators, etc., where it is necessary to consider the suitability of shaft coupling in terms of load torque at transient conditions (run-up and braking. Therefore we decided to analyze the use of pneumatic flexible shaft couplings also in this type of mechanical systems on an example of conveyor belt drive.
Transient analysis of a thermal storage unit involving a phase change material
Griggs, E. I.; Pitts, D. R.; Humphries, W. R.
1974-01-01
The transient response of a single cell of a typical phase change material type thermal capacitor has been modeled using numerical conductive heat transfer techniques. The cell consists of a base plate, an insulated top, and two vertical walls (fins) forming a two-dimensional cavity filled with a phase change material. Both explicit and implicit numerical formulations are outlined. A mixed explicit-implicit scheme which treats the fin implicity while treating the phase change material explicitly is discussed. A band algorithmic scheme is used to reduce computer storage requirements for the implicit approach while retaining a relatively fine grid. All formulations are presented in dimensionless form thereby enabling application to geometrically similar problems. Typical parametric results are graphically presented for the case of melting with constant heat input to the base of the cell.
Edwards, C. L. W.; Meissner, F. T.; Hall, J. B.
1979-01-01
Color computer graphics techniques were investigated as a means of rapidly scanning and interpreting large sets of transient heating data. The data presented were generated to support the conceptual design of a heat-sink thermal protection system (TPS) for a hypersonic research airplane. Color-coded vector and raster displays of the numerical geometry used in the heating calculations were employed to analyze skin thicknesses and surface temperatures of the heat-sink TPS under a variety of trajectory flight profiles. Both vector and raster displays proved to be effective means for rapidly identifying heat-sink mass concentrations, regions of high heating, and potentially adverse thermal gradients. The color-coded (raster) surface displays are a very efficient means for displaying surface-temperature and heating histories, and thereby the more stringent design requirements can quickly be identified. The related hardware and software developments required to implement both the vector and the raster displays for this application are also discussed.
Transient thermal analysis for radioactive liquid mixing operations in a large-scaled tank
International Nuclear Information System (INIS)
Lee, S. Y.; Smith, F. G. III
2014-01-01
A transient heat balance model was developed to assess the impact of a Submersible Mixer Pump (SMP) on radioactive liquid temperature during the process of waste mixing and removal for the high-level radioactive materials stored in Savannah River Site (SRS) tanks. The model results will be mainly used to determine the SMP design impacts on the waste tank temperature during operations and to develop a specification for a new SMP design to replace existing longshaft mixer pumps used during waste removal. The present model was benchmarked against the test data obtained by the tank measurement to examine the quantitative thermal response of the tank and to establish the reference conditions of the operating variables under no SMP operation. The results showed that the model predictions agreed with the test data of the waste temperatures within about 10%
TRAC analysis of steam-generator overfill transients for TMI-1
International Nuclear Information System (INIS)
Bassett, B.
1983-01-01
A reactor safety issue concerning the overfilling of once-through steam generators leading to combined primary/secondary blowdown has been raised recently. A series of six calculations, performed with the LWR best-estimate code, TRAC-PD2, on a Babcock and Wilcox Plant (TMI-1), was performed to investigate this safety issue. The base calculation assumed runaway main feedwater to one steam generator causing it to overfill and to break the main steam line. Four additional calculations build onto the base case with combinations of a pump-seal failure, a steam-generator tube rupture, and the pilot-operated relief valve not reseating. A sixth calculation involved only the rupture of a single steam-generator tube. The results of these analyses indicate that for the transients investigated, the emergency cooling system provided an adequate make-up coolant flow to mitigate the accidents
Transient core-debris bed heat-removal experiments and analysis
International Nuclear Information System (INIS)
Ginsberg, T.; Klein, J.; Klages, J.; Schwarz, C.E.; Chen, J.C.
1982-08-01
An experimental investigation is reported of the thermal interaction between superheated core debris and water during postulated light-water reactor degraded core accidents. Data are presented for the heat transfer characteristics of packed beds of 3 mm spheres which are cooled by overlying pools of water. Results of transient bed temperature and steam flow rate measurements are presented for bed heights in the range 218 mm-433 mm and initial particle bed temperatures between 530K and 972K. Results display a two-part sequential quench process. Initial frontal cooling leaves pockets or channels of unquenched spheres. Data suggest that heat transfer process is limited by a mechanism of countercurrent two-phase flow. An analytical model which combines a bed energy equation with either a quasisteady version of the Lipinski debris bed model or a critical heat flux model reasonably well predicts the characteristic features of the bed quench process. Implications with respect to reactor safety are discussed
Development and Application of a Plant Code to the Analysis of Transients in Integrated Reactors
International Nuclear Information System (INIS)
Rabiti, A.; Gimenez, M.; Delmastro, D.; Zanocco, P.
2003-01-01
In this work, a secondary system model for a CAREM-25 type nuclear power plant was developed.A two-phase flow homogenous model was used and found adequate for the scope of the present work.A finite difference scheme was used for the numerical implementation of the model.This model was coupled to the HUARPE code, a primary circuit code, in order to obtain a plant code.This plant code was used to analyze the inherent response of the system, without control feedback loops, for a transient of steam generator feed-water mass flow reduction.The results obtained are satisfactory, but a validation against other plant codes is still necessary
International Nuclear Information System (INIS)
Johnson, H.G.
1982-01-01
The Fast Flux Test Facility (FFTF) is arranged for natural circulation emergency core cooling in the event of loss of all plant electrical power. This design feature was conclusively demonstrated in a series of four natural circulation transient tests during the plant startup testing program in 1980 and 1981. Predictions, of core performance during these tests were made using the Westinghouse Hanford Company CORA computer program. The predictions, which compared well with measured plant data, were used in the extrapolation process to demonstrate the validity of the FFTF plant safety models and codes. This paper provides a brief description of the CORA code and includes typical comparisons of predictions to measured plant test data
TRAC analysis of the Crystal River Unit-3 Plant transient of February 26, 1980
International Nuclear Information System (INIS)
Coddington, P.; Willcutt, G.J.E. Jr.
1983-01-01
This paper describes the application of the TRAC-PD2 and TRAC-PF1 codes to analyze the Crystal River transient. The PD2 and PF1 analyses used the three-dimensional and one-dimensional vessel models, respectively. Both calculations predicted the plant depressurization caused by the open PORV and the subsequent repressurization caused by closing the PORV and continuing high-pressure injection flow. Also, natural circulation was calculated in loop B following reestablishment of feedwater to the loop-B steam generator. After system repressurization, the codes calculated that pressure was relieved through the safety valves, and an intermittent flow occurred in loop A because of high-pressure-injection-driven density variations
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.
COBRA-WC: a version of COBRA for single-phase multiassembly thermal hydraulic transient analysis
International Nuclear Information System (INIS)
George, T.L.; Basehore, K.L.; Wheeler, C.L.; Prather, W.A.; Masterson, R.E.
1980-07-01
The objective of this report is to provide the user of the COBRA-WC (Whole Core) code a basic understanding of the code operation and capabilities. Included in this manual are the equations solved and the assumptions made in their derivations, a general description of the code capabilities, an explanation of the numerical algorithms used to solve the equations, and input instructions for using the code. Also, the auxiliary programs GEOM and SPECSET are described and input instructions for each are given. Input for COBRA-WC sample problems and the corresponding output are given in the appendices. The COBRA-WC code has been developed from the COBRA-IV-I code to analyze liquid Metal Fast Breeder Reactor (LMFBR) assembly transients. It was specifically developed to analyze a core flow coastdown to natural circulation cooling
International Nuclear Information System (INIS)
Itoh, Hiroyasu; Tomioka, Satoshi; Enoto, Takeaki
1994-01-01
Wave information about micro-pulses of electron beams is required in order to control an electron beam precisely, and it may be possible to obtain such information by analyzing the spectrum of the electromagnetic field created by a linac electron beam. In order to derive the spectrum, we measured the spatial distribution generated by a standing wave. Furthermore we calculated the transient electromagnetic field excited by a bunched electron beam, using the finite-difference time-domain (FD-TD) method, and compared two spectra in the frequency domain, into which the calculated value in time variation and the measured values in spatial variation are transformed by using the fast Fourier transform (FFT) respectively. (author)
International Nuclear Information System (INIS)
Lee, Jun-Ha; Lee, Hoong-Joo
2005-01-01
We developed a new systematic calibration procedure which was applied to the prediction of the diffusivity, the segregation, and transient enhanced diffusion (TED) of an indium impurity. The TED of the indium impurity was studied using four different experimental conditions. Although indium is susceptible to TED, rapid thermal annealing (RTA) is effective in suppressing the TED effect and maintaining a steep retrograde profile. Like boron impurities, the indium shows significant oxidation-enhanced diffusion in silicon and has segregation coefficients much less than 1 at the Si/SiO 2 interface. In contrast to boron, the segregation coefficient of indium decreases as the temperature increases. The accuracy of the proposed procedure was validated by using secondary ion mass spectrometry (SIMS) data and by using the 0.13-μm device characteristics, such as V th and I dsat , for which the differences between simulation and experiment less than 5 %.
Gao, B. C.; Meng, X. K.; Shen, M. X.; Peng, X. D.
2016-05-01
A transient thermal-mechanical coupling model for a contacting mechanical seal during start-up has been developed. It takes into consideration the coupling relationship among thermal-mechanical deformation, film thickness, temperature and heat generation. The finite element method and multi-iteration technology are applied to solve the temperature distribution and thermal-mechanical deformation as well as their evolution behavior. Results show that the seal gap transforms from negative coning to positive coning and the contact area of the mechanical seal gradually decreases during start-up. The location of the maximum temperature and maximum contact pressure move from the outer diameter to inside diameter. The heat generation and the friction torque increase sharply at first and then decrease. Meanwhile, the contact force decreases and the fluid film force and leakage rate increase.
Sayed, Sadeed Bin; Uysal, Ismail Enes; Bagci, Hakan; Ulku, H. Arda
2018-01-01
Quantum tunneling is observed between two nanostructures that are separated by a sub-nanometer gap. Electrons “jumping” from one structure to another create an additional current path. An auxiliary tunnel is introduced between the two structures as a support for this so that a classical electromagnetic solver can account for the effects of quantum tunneling. The dispersive permittivity of the tunnel is represented by a Drude model, whose parameters are obtained from the electron tunneling probability. The transient scattering from the connected nanostructures (i.e., nanostructures plus auxiliary tunnel) is analyzed using a time domain volume integral equation solver. Numerical results demonstrating the effect of quantum tunneling on the scattered fields are provided.
Vectorization of LWR transient analysis code RELAP5/MOD1 and its effect
International Nuclear Information System (INIS)
Ishiguro, Misako; Harada, Hiroo; Shinozawa, Naohisa; Naraoka, Ken-itsu
1985-03-01
The RELAP5/MOD1 is a large thermal-hydraulic code to analyze LWR LOCA and non-LOCA transients. The code originally was designed for use on a CDC Cyber-176. This report documents vectorization of the RELAP5/MOD1 code conducted for the purpose of efficient use of VP-100 (peak speed 250 MFLOPS, clock period 7.5 ns) at the JAERI. The code was vectorized using the junction and volume level parallelisms in the hydrodynamic calculations, and the heat-structure and heat-mesh level in the heat conduction calculations. The vectorized version runs as much as 2.4 to 2.8 times faster than the original scalar version, while the speedup ratio is dependent on the number of spactial cells included in the problem. (author)
Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility
Narcisi, V.; Giannetti, F.; Del Nevo, A.; Tarantino, M.; Caruso, G.
2017-11-01
In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.
Transient Analysis of a Gas-cooled Fast Reactor for Single Control Assembly Withdrawal
International Nuclear Information System (INIS)
Choi, Hangbok
2014-01-01
The Energy Multiplier Module (EMZ) system response has been evaluated for control assembly withdrawal transients. Currently the EM2 core is equipped with six cylindrical drum-type control assemblies in the reflector zone for excess reactivity control and power maneuvering during the operating core life. This study investigates the system response to the control assembly withdrawal accident with various rotational speeds and reactivity worth to determine feasible control assembly design requirements from the physics viewpoint. The simulations have been conducted for single control assembly withdrawal transients without scram by a gas-cooled reactor plant simulator, which is based on a simplified plant nodal model, including the point reactor kinetics, single channel core thermal-fluid model, and a turbo-machinery performance model. Simulations were conducted for the middle-of- cycle core, when the excess reactivity of the core is the highest. Control assembly withdrawal times were varied from 1 (runaway) to 180 sec and reactivity worth was varied from 100 to 400 pcm. For a single control assembly withdrawal, the simulation has shown that the peak fuel temperature is expected to be ~1820°C when the assembly worth is 200 pcm and the runaway time is 1 sec per 180 degree rotation. The peak temperature could be reduced to ~1780°C if the assembly is rotated out in a moderate speed such as 1 degree/sec. These peak temperatures give a thermal margin of 22 to 24% to the melting point of uranium carbide fuel. The results also indicate that the current design with a single control assembly worth of 314 pcm may need adjustments in the future design. (author)
International Nuclear Information System (INIS)
Saito, Shinzo
1977-09-01
In the code EXCURS developed for core transient behaviour calculation of a sodium-cooled fast reactor, a one-channel model is used to represent thermal behaviour of the reactor core. Calculations are made for three different channels; i.e. average, hot and hottest. In the average channel the power density and coolant velocity are equal to the mean values of the whole core. In the hot channel, a maximum power density of the core and a specific coolant velocity are introduced. In the hottest channel, engineering hot channel factors are considered to the hot channel. A one-point neutron kinetics equation with six delayed neutron groups is used to calculate the time-dependent power behaviour. Externally introduced reactivity effect and control rod movement in the case of a scram are taken into account. In the feedback effects evaluated on the basis of the average channel temperatures are considered Doppler effect, fuel axial expansion, cladding expansion, coolant expansion and structure expansion. The decay heat after reactor scram is also considered. Heat balance is taken in each cross section, neglecting the axial heat transfer except for the coolant region. Temperature dependence of the physical properties of materials is considered by second-order polynomials approximation, and also the fuel melting process. Each channel can be divided into a maximum of 20 regions in both radially and axially. The reactor core transient behaviour due to reactivity insertion or loss-of-coolant flow can be studied by EXCURS. The calculated results are plotted optionally by connected code EXPLOT. (auth.)
Transient pseudohypoaldosteronism
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Stajić Nataša
2011-01-01
Full Text Available Introduction. Infants with urinary tract malformations (UTM presenting with urinary tract infection (UTI are prone to develop transient type 1 pseudohypoaldosteronism (THPA1. Objective. Report on patient series with characteristics of THPA1, UTM and/or UTI and suggestions for the diagnosis and therapy. Methods. Patients underwent blood and urine electrolyte and acid-base analysis, serum aldosterosterone levels and plasma rennin activity measuring; urinalysis, urinoculture and renal ultrasound were done and medical and/or surgical therapy was instituted. Results. Hyponatraemia (120.9±5.8 mmol/L, hyperkalaemia (6.9±0.9 mmol/L, metabolic acidosis (plasma bicarbonate, 11±1.4 mmol/L, and a rise in serum creatinine levels (145±101 μmol/L were associated with inappropriately high urinary sodium (51.3±17.5 mmol/L and low potassium (14.1±5.9 mmol/L excretion. Elevated plasma aldosterone concentrations (170.4±100.5 ng/dL and the very high levels of the plasma aldosterone to potassium ratio (25.2±15.6 together with diminished urinary K/Na values (0.31±0.19 indicated tubular resistance to aldosterone. After institution of appropriate medical and/or surgical therapy, serum electrolytes, creatinine, and acid-base balance were normalized. Imaging studies showed ureteropyelic or ureterovesical junction obstruction in 3 and 2 patients, respectively, posterior urethral valves in 3, and normal UT in 1 patient. According to our knowledge, this is the first report on THPA1 in the Serbian literature. Conclusion. Male infants with hyponatraemia, hyperkalaemia and metabolic acidosis have to have their urine examined and the renal ultrasound has to be done in order to avoid both, the underdiagnosis of THPA1 and the inappropriate medication.
Transients: The regulator's view
International Nuclear Information System (INIS)
Sheron, B.W.; Speis, T.P.
1984-01-01
This chapter attempts to clarify the basis for the regulator's concerns for transient events. Transients are defined as both anticipated operational occurrences and postulated accidents. Recent operational experience, supplemented by improved probabilistic risk analysis methods, has demonstrated that non-LOCA transient events can be significant contributors to overall risk. Topics considered include lessons learned from events and issues, the regulations governing plant transients, multiple failures, different failure frequencies, operator errors, and public pressure. It is concluded that the formation of Owners Groups and Regulatory Response Groups within the owners groups are positive signs of the industry's concern for safety and responsible dealing with the issues affecting both the US NRC and the industry
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
International Nuclear Information System (INIS)
Droppo, James G.
2004-01-01
An analysis is conducted of the 1996-1998 Hanford tank ventilation studies of average ventilation rates to help define characteristics of shorter term releases. This effort is being conducted as part of the design of tests of Industrial Hygiene's (IH) instrumentation ability to detect transient airborne plumes from tanks using current deployment strategies for tank operations. This analysis has improved our understanding of the variability of hourly average tank ventilation processes. However, the analysis was unable to discern the relative importance of emissions due to continuous releases and short-duration bursts of material. The key findings are as follows: (1) The ventilation of relatively well-sealed, passively ventilated tanks appears to be driven by a combination of pressure, buoyancy, and wind influences. The results of a best-fit analysis conducted with a single data set provide information on the hourly emission variability that IH instrumentation will need to detect. (2) Tank ventilation rates and tank emission rates are not the same. The studies found that the measured infiltration rates for a single tank are often a complex function of air exchanges between tanks and air exchanges with outdoor air. This situation greatly limits the usefulness of the ventilation data in defining vapor emission rates. (3) There is no evidence in the data to discern if the routine tank vapor releases occur over a short time (i.e., a puff) or over an extended time (i.e., continuous releases). Based on this analysis of the tank ventilation studies, it is also noted that (1) the hourly averaged emission peaks from the relatively well-sealed passively-vented tanks (such as U-103) are not a simple function of one meteorological parameter--but the peaks often are the result of the coincidence of temporal maximums in pressure, temperature, and wind influences and (2) a mechanistic combination modeling approach and/or field studies may be necessary to understand the short
Time-step selection considerations in the analysis of reactor transients with DIF3D-K
International Nuclear Information System (INIS)
Taiwo, T.A.; Khalil, H.S.; Cahalan, J.E.; Morris, E.E.
1993-01-01
The DIF3D-K code solves the three-dimensional, time-dependent multigroup neutron diffusion equations by using a nodal approach for spatial discretization and either the theta method or one of three space-time factorization approaches for temporal integration of the nodal equations. The three space-time factorization options (namely, improved quasistatic, adiabatic, and conventional point kinetics) were implemented because of their potential efficiency advantage for the analysis of transients in which the flux shape changes more slowly than its amplitude. In this paper, we describe the implementation of DIF3D-K as the neutronics module within the SAS-HWR accident analysis code. We also describe the neuronic-related time-step selection algorithms and their influence on the accuracy and efficiency of the various solution options
International Nuclear Information System (INIS)
Gilli, L.; Lathouwers, D.; Kloosterman, J.L.; Van der Hagen, T.H.J.J.
2011-01-01
In this paper a method to perform sensitivity analysis for a simplified multi-physics problem is presented. The method is based on the Adjoint Sensitivity Analysis Procedure which is used to apply first order perturbation theory to linear and nonlinear problems using adjoint techniques. The multi-physics problem considered includes a neutronic, a thermo-kinetics, and a thermal-hydraulics part and it is used to model the time dependent behavior of a sodium cooled fast reactor. The adjoint procedure is applied to calculate the sensitivity coefficients with respect to the kinetic parameters of the problem for two reference transients using two different model responses, the results obtained are then compared with the values given by a direct sampling of the forward nonlinear problem. Our first results show that, thanks to modern numerical techniques, the procedure is relatively easy to implement and provides good estimation for most perturbations, making the method appealing for more detailed problems. (author)
Time-step selection considerations in the analysis of reactor transients with DIF3D-K
International Nuclear Information System (INIS)
Taiwo, T.A.; Khalil, H.S.; Cahalan, J.E.; Morris, E.E.
1993-01-01
The DIF3D-K code solves the three-dimensional, time-dependent multigroup neutron diffusion equations by using a nodal approach for spatial discretization and either the theta method or one of three space-time factorization approaches for temporal integration of the nodal equations. The three space-time factorization options (namely, improved quasistatic, adiabatic and conventional point kinetics) were implemented because of their potential efficiency advantage for the analysis of transients in which the flux shape changes more slowly than its amplitude. Here we describe the implementation of DIF3D-K as the neutronics module within the SAS-HWR accident analysis code. We also describe the neutronics-related time step selection algorithms and their influence on the accuracy and efficiency of the various solution options
Chin, Jeffrey C.; Csank, Jeffrey T.
2016-01-01
The Tool for Turbine Engine Closed-Loop Transient Analysis (TTECTrA ver2) is a control design tool thatenables preliminary estimation of transient performance for models without requiring a full nonlinear controller to bedesigned. The program is compatible with subsonic engine models implemented in the MATLAB/Simulink (TheMathworks, Inc.) environment and Numerical Propulsion System Simulation (NPSS) framework. At a specified flightcondition, TTECTrA will design a closed-loop controller meeting user-defined requirements in a semi or fully automatedfashion. Multiple specifications may be provided, in which case TTECTrA will design one controller for each, producing acollection of controllers in a single run. Each resulting controller contains a setpoint map, a schedule of setpointcontroller gains, and limiters; all contributing to transient characteristics. The goal of the program is to providesteady-state engine designers with more immediate feedback on the transient engine performance earlier in the design cycle.
Directory of Open Access Journals (Sweden)
Youwei He
2018-02-01
Full Text Available Although technical advances in hydraulically fracturing and drilling enable commercial production from tight reservoirs, oil/gas recovery remains at a low level. Due to the technical and economic limitations of well-testing operations in tight reservoirs, rate-transient analysis (RTA has become a more attractive option. However, current RTA models hardly consider the effect of the non-uniform production on rate decline behaviors. In fact, PLT results demonstrate that production profile is non-uniform. To fill this gap, this paper presents an improved RTA model of multi-fractured horizontal wells (MFHWs to investigate the effects of non-uniform properties of hydraulic fractures (production of fractures, fracture half-length, number of fractures, fracture conductivity, and vertical permeability on rate transient behaviors through the diagnostic type curves. Results indicate obvious differences on the rate decline curves among the type curves of uniform properties of fractures (UPF and non-uniform properties of fractures (NPF. The use of dimensionless production integral derivative curve magnifies the differences so that we can diagnose the phenomenon of non-uniform production. Therefore, it’s significant to incorporate the effects of NPF into the RDA models of MFHWs, and the model proposed in this paper enables us to better evaluate well performance based on long-term production data.
Tournier, Jean-Michel; El-Genk, Mohamed S.
1995-01-01
A two-dimensional Heat Pipe Transient Analysis Model, 'HPTAM,' was developed to simulate the transient operation of fully-thawed heat pipes and the startup of heat pipes from a frozen state. The model incorporates: (a) sublimation and resolidification of working fluid; (b) melting and freezing of the working fluid in the porous wick; (c) evaporation of thawed working fluid and condensation as a thin liquid film on a frozen substrate; (d) free-molecule, transition, and continuum vapor flow regimes, using the Dusty Gas Model; (e) liquid flow and heat transfer in the porous wick; and (f) thermal and hydrodynamic couplings of phases at their respective interfaces. HPTAM predicts the radius of curvature of the liquid meniscus at the liquid-vapor interface and the radial location of the working fluid level (liquid or solid) in the wick. It also includes the transverse momentum jump condition (capillary relationship of Pascal) at the liquid-vapor interface and geometrically relates the radius of curvature of the liquid meniscus to the volume fraction of vapor in the wick. The present model predicts the capillary limit and partial liquid recess (dryout) in the evaporator wick, and incorporates a liquid pooling submodel, which simulates accumulation of the excess liquid in the vapor core at the condenser end.
Transient analysis for lead-bismuth-cooled accelerator-driven system proposed by JAEA
International Nuclear Information System (INIS)
Sugawara, T.; Nishihara, K.; Tsujimoto, K.
2015-01-01
It is supposed that an Accelerator-driven System (ADS) is safer than conventional critical reactors since an ADS is driven by the external neutron source in the subcritical state. In this study, the transient analyses for the lead-bismuth cooled ADS proposed by JAEA were performed using the SIMMER-III and RELAP5/mod3.2 codes to investigate the possibility of core damage. In this research, 3 accidents: the protected loss of heat sink, the protected overcooling and the unprotected blockage accident were considered as typical ADS accidents. Through these calculations, it was confirmed that all calculation results, except for the protected loss of heat sink, fulfilled the no-damage criteria. In the protected loss of heat sink, the cladding tube temperature reached its melting temperature after 18-21 hours, although the calculation condition was very conservative. These results have led to requirements to design a safety system of the ADS to decrease the frequencies of accidents. (authors)
Non-modal stability analysis and transient growth in a magnetized Vlasov plasma
Ratushnaya, V.
2014-12-01
Collisionless plasmas, such as those encountered in tokamaks, exhibit a rich variety of instabilities. The physical origin, triggering mechanisms and fundamental understanding of many plasma instabilities, however, are still open problems. We investigate the stability properties of a 3-dimensional collisionless Vlasov plasma in a stationary homogeneous magnetic field. We narrow the scope of our investigation to the case of Maxwellian plasma and examine its evolution with an electrostatic approximation. For the first time using a fully kinetic approach we show the emergence of the local instability, a transient growth, followed by classical Landau damping in a stable magnetized plasma. We show that the linearized Vlasov operator is non-normal leading to the algebraic growth of the perturbations using non-modal stability theory. The typical time scales of the obtained instabilities are of the order of several plasma periods. The first-order distribution function and the corresponding electric field are calculated and the dependence on the magnetic field and perturbation parameters is studied. Our results offer a new scenario of the emergence and development of plasma instabilities on the kinetic scale.
Use of one delayed-neutron precursor group in transient analysis
International Nuclear Information System (INIS)
Diamond, D.J.
1983-01-01
In most reactor dynamics calculations six groups of delayed-neutron precursors are usually accounted for. However, under certain circumstances it may be advantageous to simplify the calculation and utilize a single delayed-neutron group. The motivation for going to one precursor group is economy. For LWR transient codes that use point kinetics the equations are solved very rapidly and six precursor groups should always be used. However, codes with spatially dependent neutron kinetics are very long running and the use of one precursor group may save computer costs and not impair the accuracy of the results significantly. Furthermore, in some codes, the elimation of five presursor groups makes additional memory available which may be used to give a net increase in the accuracy of the calculations, e.g., by allowing for an increase in mesh density. In order to use one delayed neutron precursor group it is necessary to derive a single decay constant, 6 lambda-, which, along with the total (or one group) delayed neutron fraction β = Σ/sub i = 1/β/sub i/, will adequately describe the transeint precursor behavior. The present summary explains how a recommendation for lambda- was derived
Transient Mass-loss Analysis of Solar Observations Using Stellar Methods
Energy Technology Data Exchange (ETDEWEB)
Crosley, M. K.; Norman, C. [Johns Hopkins University, Department of Physics and Astronomy, 3400 N. Charles Street, Baltimore, MD 21218 (United States); Osten, R. A. [Space Telescope Science Institute, 3700 San Martin Drive, Baltimore, MD 21218 (United States)
2017-08-10
Low-frequency dynamic spectra of radio bursts from nearby stars offer the best chance to directly detect the stellar signature of transient mass loss on low-mass stars. Crosley et al. (2016) proposes a multi-wavelength methodology to determine coronal mass ejection (CME) parameters, such as speed, mass, and kinetic energy. We test the validity and accuracy of the results derived from the methodology by using Geostationary Operational Environmental Satellite X-ray observations and Bruny Island Radio Spectrometer radio observations. These are analogous observations to those that would be found in the stellar studies. Derived results from these observations are compared to direct white light measurements of the Large Angle and Spectrometric Coronagraph. We find that, when a pre-event temperature can be determined, the accuracy of CME speeds are within a few hundred km s{sup −1}, and are reliable when specific criteria has been met. CME mass and kinetic energies are only useful in determining the approximate order of magnitude measurements when considering the large errors associated to them. These results will be directly applicable to the interpretation of any detected stellar events and the derivation of stellar CME properties.
Techno-economic analysis of a transient plant-based platform for monoclonal antibody production
Nandi, Somen; Kwong, Aaron T.; Holtz, Barry R.; Erwin, Robert L.; Marcel, Sylvain; McDonald, Karen A.
2016-01-01
ABSTRACT Plant-based biomanufacturing of therapeutic proteins is a relatively new platform with a small number of commercial-scale facilities, but offers advantages of linear scalability, reduced upstream complexity, reduced time to market, and potentially lower capital and operating costs. In this study we present a detailed process simulation model for a large-scale new “greenfield” biomanufacturing facility that uses transient agroinfiltration of Nicotiana benthamiana plants grown hydroponically indoors under light-emitting diode lighting for the production of a monoclonal antibody. The model was used to evaluate the total capital investment, annual operating cost, and cost of goods sold as a function of mAb expression level in the plant (g mAb/kg fresh weight of the plant) and production capacity (kg mAb/year). For the Base Case design scenario (300 kg mAb/year, 1 g mAb/kg fresh weight, and 65% recovery in downstream processing), the model predicts a total capital investment of $122 million dollars and cost of goods sold of $121/g including depreciation. Compared with traditional biomanufacturing platforms that use mammalian cells grown in bioreactors, the model predicts significant reductions in capital investment and >50% reduction in cost of goods compared with published values at similar production scales. The simulation model can be modified or adapted by others to assess the profitability of alternative designs, implement different process assumptions, and help guide process development and optimization. PMID:27559626
A thermal-hydraulic code for transient analysis in a channel with a rod bundle
International Nuclear Information System (INIS)
Khodjaev, I.D.
1995-01-01
The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident
Uysal, Ismail Enes
2016-08-09
Transient electromagnetic interactions on plasmonic nanostructures are analyzed by solving the Poggio-Miller-Chan-Harrington-Wu-Tsai (PMCHWT) surface integral equation (SIE). Equivalent (unknown) electric and magnetic current densities, which are introduced on the surfaces of the nanostructures, are expanded using Rao-Wilton-Glisson and polynomial basis functions in space and time, respectively. Inserting this expansion into the PMCHWT-SIE and Galerkin testing the resulting equation at discrete times yield a system of equations that is solved for the current expansion coefficients by a marching on-in-time (MOT) scheme. The resulting MOT-PMCHWT-SIE solver calls for computation of additional convolutions between the temporal basis function and the plasmonic medium\\'s permittivity and Green function. This computation is carried out with almost no additional cost and without changing the computational complexity of the solver. Time-domain samples of the permittivity and the Green function required by these convolutions are obtained from their frequency-domain samples using a fast relaxed vector fitting algorithm. Numerical results demonstrate the accuracy and applicability of the proposed MOT-PMCHWT solver. © 2016 Optical Society of America.
Combined analysis of steady state and transient transport by the maximum entropy method
Energy Technology Data Exchange (ETDEWEB)
Giannone, L.; Stroth, U; Koellermeyer, J [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); and others
1996-04-01
A new maximum entropy approach has been applied to analyse three types of transient transport experiments. For sawtooth propagation experiments in the ASDEX Upgrade and ECRH power modulation and power-switching experiments in the Wendelstein 7-AS Stellarator, either the time evolution of the temperature perturbation or the phase and amplitude of the modulated temperature perturbation are used as non-linear constraints to the {chi}{sub e} profile to be fitted. Simultaneously, the constraints given by the equilibrium temperature profile for steady-state power balance are fitted. In the maximum entropy formulation, the flattest {chi}{sub e} profile consistent with the constraints is found. It was found that {chi}{sub e} determined from sawtooth propagation was greater than the power balance value by a factor of five in the ASDEX Upgrade. From power modulation experiments, employing the measurements of four modulation frequencies simultaneously, the power deposition profile as well as the {chi}{sub e} profile could be determined. A comparison of the predictions of a time-independent {chi}{sub e} model and a power-dependent {chi}{sub e} model is made. The power-switching experiments show that the {chi}{sub e} profile must change within a millisecond to a new value consistent with the power balance value at the new input power. Neither power deposition broadening due to suprathermal electrons nor temperature or gradient dependences of {chi}{sub e} can explain this observation. (author).
A thermal-hydraulic code for transient analysis in a channel with a rod bundle
Energy Technology Data Exchange (ETDEWEB)
Khodjaev, I.D. [Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)
1995-09-01
The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.
PRONTO3D users` instructions: A transient dynamic code for nonlinear structural analysis
Energy Technology Data Exchange (ETDEWEB)
Attaway, S.W.; Mello, F.J.; Heinstein, M.W.; Swegle, J.W.; Ratner, J.A. [Sandia National Labs., Albuquerque, NM (United States); Zadoks, R.I. [Univ. of Texas, El Paso, TX (United States)
1998-06-01
This report provides an updated set of users` instructions for PRONTO3D. PRONTO3D is a three-dimensional, transient, solid dynamics code for analyzing large deformations of highly nonlinear materials subjected to extremely high strain rates. This Lagrangian finite element program uses an explicit time integration operator to integrate the equations of motion. Eight-node, uniform strain, hexahedral elements and four-node, quadrilateral, uniform strain shells are used in the finite element formulation. An adaptive time step control algorithm is used to improve stability and performance in plasticity problems. Hourglass distortions can be eliminated without disturbing the finite element solution using either the Flanagan-Belytschko hourglass control scheme or an assumed strain hourglass control scheme. All constitutive models in PRONTO3D are cast in an unrotated configuration defined using the rotation determined from the polar decomposition of the deformation gradient. A robust contact algorithm allows for the impact and interaction of deforming contact surfaces of quite general geometry. The Smooth Particle Hydrodynamics method has been embedded into PRONTO3D using the contact algorithm to couple it with the finite element method.
Non-modal stability analysis and transient growth in a magnetized Vlasov plasma
Ratushnaya, V.; Samtaney, Ravi
2014-01-01
Collisionless plasmas, such as those encountered in tokamaks, exhibit a rich variety of instabilities. The physical origin, triggering mechanisms and fundamental understanding of many plasma instabilities, however, are still open problems. We investigate the stability properties of a 3-dimensional collisionless Vlasov plasma in a stationary homogeneous magnetic field. We narrow the scope of our investigation to the case of Maxwellian plasma and examine its evolution with an electrostatic approximation. For the first time using a fully kinetic approach we show the emergence of the local instability, a transient growth, followed by classical Landau damping in a stable magnetized plasma. We show that the linearized Vlasov operator is non-normal leading to the algebraic growth of the perturbations using non-modal stability theory. The typical time scales of the obtained instabilities are of the order of several plasma periods. The first-order distribution function and the corresponding electric field are calculated and the dependence on the magnetic field and perturbation parameters is studied. Our results offer a new scenario of the emergence and development of plasma instabilities on the kinetic scale.
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.