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)
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)
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
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.
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
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
Pressure transients in pipeline systems
DEFF Research Database (Denmark)
Voigt, Kristian
1998-01-01
This text is to give an overview of the necessary background to do investigation of pressure transients via simulations. It will describe briefly the Method of Characteristics which is the defacto standard for simulating pressure transients. Much of the text has been adopted from the book Pressur...
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
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.
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
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
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
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
Pressure transients across HEPA filters
International Nuclear Information System (INIS)
Gregory, W.; Reynolds, G.; Ricketts, C.; Smith, P.R.
1977-01-01
Nuclear fuel cycle facilities require ventilation for health and safety reasons. High efficiency particulate air (HEPA) filters are located within ventilation systems to trap radioactive dust released in reprocessing and fabrication operations. Pressure transients within the air cleaning systems may be such that the effectiveness of the filtration system is questioned under certain accident conditions. These pressure transients can result from both natural and man-caused phenomena: atmospheric pressure drop caused by a tornado or explosions and nuclear excursions initiate pressure pulses that could create undesirable conditions across HEPA filters. Tornado depressurization is a relatively slow transient as compared to pressure pulses that result from combustible hydrogen-air mixtures. Experimental investigation of these pressure transients across air cleaning equipment has been undertaken by Los Alamos Scientific Laboratory and New Mexico State University. An experimental apparatus has been constructed to impose pressure pulses across HEPA filters. The experimental equipment is described as well as preliminary results using variable pressurization rates. Two modes of filtration of an aerosol injected upstream of the filter is examined. A laser instrumentation for measuring the aerosol release, during the transient, is described
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.
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.
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
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.
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
International Nuclear Information System (INIS)
Kitoh, Kazuaki; Koshizuka, Seiichi; Oka, Yoshiaki
1996-01-01
The features of the direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor (SCFBR) are high thermal efficiency and simple reactor system. The safety principle is basically the same as that of an LWR since it is a water-cooled reactor. Maintaining the core flow is the basic safety requirement of the reactor, since its coolant system is the one through type. The transient behaviors at control rod, pressure and flow-induced abnormalities are analyzed and presented in this paper. The results of flow-induced transients of SCFBR were reported at ICONE-3, though pressure change was neglected. The change of fuel temperature distribution is also considered for the analysis of the rapid reactivity-induced transients such as control rod withdrawal. Total loss of flow and pump seizure are analyzed as the accidents. Loss of load, control rod withdrawal from the normal operation, loss of feedwater heating, inadvertent start of an auxiliary feedwater pump, partial loss of coolant flow and loss of external power are analyzed as the transients. The behavior of the flow-induced transients is not so much different from the analyses assuming constant pressure. Fly wheels should be equipped with the feedwater pumps to prolong the coast-down time more than 10s and to cope with the total loss of flow accident. The coolant density coefficient of the SCFBR is less than one tenth of a BWR in which the recirculation flow is used for the power control. The over pressurization transients at the loss of load is not so severe as that of a BWR. The power reaches 120%. The minimum deterioration heat flux ratio (MDHFR) and the maximum pressure are sufficiently lower than the criteria; MDHFR above 1.0 and pressure ratio below 1.10 of 27.5 MPa, maximum pressure for operation. Among the reactivity abnormalities, the control rod withdrawal transient from the normal operation is analyzed
Pressure transient in liquid lines
International Nuclear Information System (INIS)
Sun, J.G.; Wang, X.Q.
1995-01-01
The pressure surge that results from a step change of flow in liquid pipelines, commonly known as water hammer, was analyzed by an eigenfunction method. A differential-integral Pressure wave equation and a linearized velocity equation were derived from the equations of mass and momentum conservation. Waveform distortion due to viscous dissipation and pipe-wall elastic expansion is characterized by a dimensionless transmission number K. The pressure surge condition, which is mathematically singular, was used in the solution procedure. The exact solutions from numerical calculation of the differential-integral equation provide a complete Pressure transient in the pipe. The problems are also calculated With the general-purpose computer code COMMIX, which solves the exact mass conservation equation and Navier-Stokes equations. These solutions were compared with published experimental results, and agreement was good. The effect of turbulence on the pressure transient is discussed in the light of COMMIX calculational results
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.
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.
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
Intermediate Leg SBLOCA - Long Lasting Pressure Transient
International Nuclear Information System (INIS)
Konjarek, D.; Bajs, T.; Vukovic, J.
2010-01-01
The basic phenomenology of Small Break Loss of Coolant Accident (SBLOCA) for PWR plant is described with focus on analysis of scenario in which reactor coolant pressure decreases below secondary system pressure. Best estimate light water reactor transient analysis code RELAP5/mod3.3 was used in calculation. Rather detailed model of the plant was used. The break occurs in intermediate leg on lowest elevation near pump suction. The size of the break is chosen to be small enough to cause cycling of safety valves (SVs) on steam generators (SGs) for some time, but, afterwards, it is large enough to remove decay heat through the break, causing cooling the secondary side. In this case of SBLOCA, when primary pressure decreases below secondary pressure, long lasting pressure transients with significant amplitude occur. Reasons for such behavior are explained.(author).
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
Measurement of fast transient pressures
International Nuclear Information System (INIS)
Procaccia, Henri
1978-01-01
The accuracy, reliability and sensitivity of a pressure transducers define its principal static characteristics. When the quantity measured varies with time, the measurement carries a dynamic error and a delay depending on the frequency of this variation. Hence, when fast pressure changes in a fluid have to be determined, different kinds of pressure transducers can be used depending on their inherent dynamic characteristics which must be compared with those of the transient phenomenon to be analysed. The text describes the pressure transducers generally employed in industry for analysing such phenomenon and gives two practical applications developed in the EDF: the first submits the measurements and results of pump cavitation tests carried out at the Vitry II EDF power station; the second deals with hammer blows particularly noticed in nuclear power stations and required the use of transducers of exceptionally high performance such as strain gauge transducers and piezoelectric transducers (response time within 1m sec.) [fr
Transient study of a PWR pressurizer
International Nuclear Information System (INIS)
Sotoma, H.
1973-01-01
An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)
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
Directory of Open Access Journals (Sweden)
Yan Zeng
2018-01-01
Full Text Available Multistage fractured horizontal wells (MFHWs have become the main technology for shale gas exploration. However, the existing models have neglected the percolation mechanism in nanopores of organic matter and failed to consider the differences among the reservoir properties in different areas. On that account, in this study, a modified apparent permeability model was proposed describing gas flow in shale gas reservoirs by integrating bulk gas flow in nanopores and gas desorption from nanopores. The apparent permeability was introduced into the macroseepage model to establish a dynamic pressure analysis model for MFHWs dual-porosity formations. The Laplace transformation and the regular perturbation method were used to obtain an analytical solution. The influences of fracture half-length, fracture permeability, Langmuir volume, matrix radius, matrix permeability, and induced fracture permeability on pressure and production were discussed. Results show that fracture half-length, fracture permeability, and induced fracture permeability exert a significant influence on production. A larger Langmuir volume results in a smaller pressure and pressure derivative. An increase in matrix permeability increases the production rate. Besides, this model fits the actual field data relatively well. It has a reliable theoretical foundation and can preferably describe the dynamic changes of pressure in the exploration process.
Energy Technology Data Exchange (ETDEWEB)
Dang, M.; Dupont, J. F.; Jacquemoud, P.; Mylonas, R. [Eidgenoessisches Inst. fuer Reaktorforschung, Wuerenlingen (Switzerland)
1981-01-15
The direct coupling of a gas cooled reactor with a closed gas turbine cycle leads to a specific dynamic plant behaviour, which may be summarized as follows: a) any operational transient involving a variation of the core mass flow rate causes a variation of the pressure ratio of the turbomachines and leads unavoidably to pressure and temperature transients in the gas turbine cycle; and b) very severe pressure equalization transients initiated by unlikely events such as the deblading of one or more turbomachines must be taken into account. This behaviour is described and illustrated through results gained from computer analyses performed at the Swiss Federal Institute for Reactor Research (EIR) in Wurenlingen within the scope of the Swiss-German HHT project.
International Nuclear Information System (INIS)
Lin, E.I.H.
1977-01-01
A large-strain time-dependent thermoplastic analysis has been developed for the ballooning deformation of a thin-wall tube subjected to internal pressure, axial loading, and fast thermal transients. This deformation initiates with the onset of plastic instability in the material, the onset being determined by a plastic-instability criterion for strain-rate sensitive materials. The interaction among the local ballooning geometry, the state of stress, and the plastic flow process was considered, and integration of the flow equations yields the local curvature and the states of stress and strain in the vicinity of the maximum ballooning site. The effects of axial constraint and heating rate were also discussed. The analysis was applied to a LWR Zircaloy cladding subjected to a constant heating rate and a range of internal pressures. The results agree very well with experimental strain-time data obtained from tube-burst tests. In most cases, the time of rupture was accurately predicted despite the lack of complete material-property data
Lei Wang; Hongjun Yin; Xiaoshuang Yang; Chuncheng Yang; Jing Fu
2015-01-01
Based on fractal geometry, fractal medium of coalbed methane mathematical model is established by Langmuir isotherm adsorption formula, Fick's diffusion law, Laplace transform formula, considering the well bore storage effect and skin effect. The Laplace transform finite difference method is used to solve the mathematical model. With Stehfest numerical inversion, the distribution of dimensionless well bore flowing pressure and its derivative was obtained in real space. According to compare wi...
Directory of Open Access Journals (Sweden)
Wang Dongying
2017-01-01
Full Text Available In this paper, a triple-medium flow model for carbonate geothermal reservoirs with an exponential external boundary ﬂux is established. The pressure solution under constant production conditions in Laplace space is solved. The geothermal wellbore pressure change considering wellbore storage and skin factor is obtained by Stehfest numerical inversion. The well test interpretation charts and Fetkovich production decline chart for carbonate geothermal reservoirs are proposed for the first time. The proposed Fetkovich production decline curves are applied to analyze the production decline behavior. The results indicate that in carbonate geothermal reservoirs with exponential external boundary ﬂux, the pressure derivative curve contains a triple dip, which represents the interporosity flow between the vugs or matrix and fracture system and the invading flow of the external boundary ﬂux. The interporosity flow of carbonate geothermal reservoirs and changing external boundary flux can both slow down the extent of production decline and the same variation tendency is observed in the Fetkovich production decline curve.
Transient performance analysis of pressurized safety injection tank with a partition
International Nuclear Information System (INIS)
Bae, Youngmin; Kim, Young In; Kim, Keung Koo
2015-01-01
Highlights: • Functional performance of safety injection tanks with a partition is evaluated. • Effects of key design parameters are scrutinized. • Distinctive features of the flow in multi-unit safety injection tanks are explored. - Abstract: A parametric study has been performed to evaluate the functional performance of a pressurized multi-unit safety injection tank, which would be considered as one of the candidates for a passive safety injection system in a nuclear power plant. The influences of key design parameters including the orifice size, initial gas fraction, and resistance coefficients and operating condition on the injection flow rate are scrutinized with a discussion of the relevant flow features such as the choked flow of gas through an orifice and two interconnected regions of differing gaseous pressure. The obtained results indicate that a multi-unit safety injection tank can passively control the injection flow rate and provide a stable safety injection over a relatively long period even in the case of drastic depressurization of a reactor coolant system
Directory of Open Access Journals (Sweden)
Lei Wang
2015-09-01
Full Text Available Based on fractal geometry, fractal medium of coalbed methane mathematical model is established by Langmuir isotherm adsorption formula, Fick's diffusion law, Laplace transform formula, considering the well bore storage effect and skin effect. The Laplace transform finite difference method is used to solve the mathematical model. With Stehfest numerical inversion, the distribution of dimensionless well bore flowing pressure and its derivative was obtained in real space. According to compare with the results from the analytical method, the result from Laplace transform finite difference method turns out to be accurate. The influence factors are analyzed, including fractal dimension, fractal index, skin factor, well bore storage coefficient, energy storage ratio, interporosity flow coefficient and the adsorption factor. The calculating error of Laplace transform difference method is small. Laplace transform difference method has advantages in well-test application since any moment simulation does not rely on other moment results and space grid.
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
Analysis of a high pressure ATWS [anticipated transient without scram] with very low make-up flow
International Nuclear Information System (INIS)
Wagner, K.C.
1988-10-01
A series of calculations were performed to analyze the response of General Electric Company's (GE) advanced boiling water reactor (ABWR) during an anticipated transient without scram (ATWS). This work investigated the early plant response with an assumed failure or manual inhibit of the high pressure core flooder (HPCF). Consequently, the reactor core isolation cooling (RCIC) and control rod drive (CRD) systems are the only sources of high pressure injection available to maintain core cooling. Steam leaving the reactor pressure vessel was diverted to the pressure suppression pool (PSP) via the steam line and the safety relief valves. The combination of an unscrammed core and the CRD and RCIC injection sources make this a particularly challenging transient. System energy balance calculations were performed to predict the core power and PSP heat-up rate. The amount of vessel vapor superheat and the PSP temperature were found to significantly affect the resultant core power. Consequently, detailed thermal-hydraulic calculations were performed to simulate the system response during the postulated transient. 15 refs., 15 figs., 4 tabs
International Nuclear Information System (INIS)
Lin, E.I.H.
1977-01-01
A large-strain, time-dependent thermoplastic analysis of ballooning deformation was developed. The true (or lagorithmic) strains, the Von Mises yield criterion and Prandtl-Reuss flow rules were used. The constitutive equation was expressed in terms of the temperature, effective stress, strain and strain rate. Material isotropy was assumed as a first approximation; note that at high temperatures even zircaloy tends to lose a substantial amount of its low-temperature anisotropy. The axisymmetry of ballooning was also assumed, which has actually been verified by numerous experiments to be accurate throughout the course of ballooning, except in the final stage when rupture is imminent. The thin-shell approximation was made, which proved to be adequate for the standard fuel claddings and which was advantageous in that the averaged state of stress was rendered determinate. The analysis led to a set of non-linear ordinary differential equations, which was then integrated by a fifth-order Runge-Kutta routine. The general formulation allows for a direct interpretation of the experimentally-observed effects of the heating rate and cladding axial constraints on the ballooning behavior. Its implications on the flow-blockage and cladding-rupture evaluations were discussed. The analysis was applied to zircaloy claddings subjected to simulated thermal transient conditions. Most of the required material properties were taken from the existing uniaxial tensile test data. Analyses were performed at a uniform heating rate of 115 0 C/sec with internal pressures ranging from 100 to 1200 psi. Satisfactory agreement was obtained between the predictions and the diametral strain-time data available from tube-burst tests
PWR [pressurized water reactor] pressurizer transient response: Final report
International Nuclear Information System (INIS)
Murphy, S.I.
1987-08-01
To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model
Reactor vessel pressure transient protection for pressurized water reactors
International Nuclear Information System (INIS)
Zech, G.
1978-09-01
During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity
Pressurizer and steam-generator behavior under PWR transient conditions
International Nuclear Information System (INIS)
Wahba, A.B.; Berta, V.T.; Pointner, W.
1983-01-01
Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled
Interface Evolution During Transient Pressure Solution Creep
Dysthe, D. K.; Podladchikov, Y. Y.; Renard, F.; Jamtveit, B.; Feder, J.
When aggregates of small grains are pressed together in the presence of small amounts of solvent the aggregate compacts and the grains tend to stick together. This hap- pens to salt and sugar in humid air, and to sediments when buried in the Earths crust. Stress concentration at the grain contacts cause local dissolution, diffusion of the dissolved material out of the interface and deposition on the less stressed faces of the grains{1}. This process, in geology known as pressure solution, plays a cen- tral role during compaction of sedimentary basins{1,2}, during tectonic deformation of the Earth's crust{3}, and in strengthening of active fault gouges following earth- quakes{4,5}. Experimental data on pressure solution has so far not been sufficiently accurate to understand the transient processes at the grain scale. Here we present ex- perimental evidence that pressure solution creep does not establish a steady state inter- face microstructure as previously thought. Conversely, cumulative creep strain and the characteristic size of interface microstructures grow as the cubic root of time. A sim- ilar transient phenomenon is known in metallurgy (Andrade creep) and is explained here using an analogy with spinodal dewetting. 1 Weyl, P. K., Pressure solution and the force of crystallization - a phenomenological theory. J. Geophys. Res., 64, 2001-2025 (1959). 2 Heald, M. T., Cementation of Simpson and St. Peter Sandstones in parts of Okla- homa, Arkansas and Missouri, J. Geol. Chicago, 14, 16-30 (1956). 3 Schwartz, S., Stöckert, B., Pressure solution in siliciclastic HP-LT metamorphic rocks constraints on the state of stress in deep levels of accretionary complexes. Tectonophysics, 255, 203-209 (1996). 4 Renard, F., Gratier, J.P., Jamtveit, B., Kinetics of crack-sealing, intergranular pres- sure solution, and compaction around active faults. J. Struct. Geol., 22, 1395-1407, (2000). 5 Miller, S. A., BenZion, Y., Burg, J. P.,A three-dimensional fluid-controlled earth
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)
Computer program TMOC for calculating of pressure transients in fluid filled piping networks
International Nuclear Information System (INIS)
Siikonen, T.
1978-01-01
The propagation of a pressure wave in fluid filles tubes is significantly affected by the pipe wall motion and vice versa. A computer code TMOC (Transients by the Method of Characteristics) is being developed for the analysis of the coupled fluid and pipe wall transients. Because of the structural feedback, the pressure can be calculated more accurately than in the programs commonly used. (author)
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
Response of steam-water mixtures to pressure transients
International Nuclear Information System (INIS)
Hull, L.M.
1985-01-01
During the transition phase of a hypothetical core-disruptive accident in a liquid-metal fast breeder reactor, melting fuel-steel mixtures may begin to boil, resulting in a two-phase mixture of molten reactor fuel and steel vapor. Dispersal of this mixture by pressure transients may prevent recriticality of the fuel material. This paper describes the results of a series of experiments that investigated the response of two-phase mixtures to pressure transients. Simulant fluids (steam/water) were used in a transparent 10.2-cm-dia, 63.5-cm-long acrylic tube. The pressure transient was provided by releasing pressurized nitrogen from a supply tank. The data obtained are in the form of pressure-time records and high-speed movies. The varied parameters are initial void fraction (10% and 40%) and transient pressure magnitude (3.45 and 310 kPa)
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.)
International Nuclear Information System (INIS)
Carvalho, F. de A.T. de.
1985-01-01
Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt
Ballooning of CANDU pressure tube in local thermal transients
International Nuclear Information System (INIS)
Mihalache, Maria; Ionescu, Viorel
2008-01-01
In certain LOCA scenarios for the CANDU fuel channel, the ballooning of the pressure tube and contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator takes place through the contact area. If the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. In INR-Pitesti the DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after occurrence of the contact between the two tubes. The code contains few models: thermal creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This paper gives a DELOCA code description and the fuel channel behaviour analysis, in transient temperature conditions of the pressure tube, using the materials properties, time and temperature dependencies of these properties as obtained in the different laboratories of the world and in the INR - Pitesti in the last years. DELOCA computer code simulated the fuel channel response to the constant heating rates of inside pressure tube surface. The paper presents contact temperature and time dependencies on the heating rate, and the appropriate fitting functions. (authors)
Computational model for transient studies of IRIS pressurizer behavior
International Nuclear Information System (INIS)
Rives Sanz, R.; Montesino Otero, M.E.; Gonzalez Mantecon, J.; Rojas Mazaira, L.
2014-01-01
International Reactor Innovative and Secure (IRIS) excels other Small Modular Reactor (SMR) designs due to its innovative characteristics regarding safety. IRIS integral pressurizer makes the design of larger pressurizer system than the conventional PWR, without any additional cost. The IRIS pressurizer volume of steam can provide enough margins to avoid spray requirement to mitigate in-surge transient. The aim of the present research is to model the IRIS pressurizer's dynamic using the commercial finite volume Computational Fluid Dynamic code CFX 14. A symmetric tridimensional model equivalent to 1/8 of the total geometry was adopted to reduce mesh size and minimize processing time. The model considers the coexistence of three phases: liquid, steam, and vapor bubbles in liquid volume. Additionally, it takes into account the heat losses between the pressurizer and primary circuit. The relationships for interfacial mass, energy, and momentum transport are programmed and incorporated into CFX by using expressions in CFX Command Language (CCL) format. Moreover, several additional variables are defined for improving the convergence and allow monitoring of boron dilution sequences and condensation-evaporation rate in different control volumes. For transient states a non - equilibrium stratification in the pressurizer is considered. This paper discusses the model developed and the behavior of the system for representative transients sequences such as the in/out-surge transients and boron dilution sequences. The results of analyzed transients of IRIS can be applied to the design of pressurizer internal structures and components. (author)
Wang, Cheng; Redgrave, Jessica; Shafizadeh, Mohsen; Majid, Arshad; Kilner, Karen; Ali, Ali N
2018-05-09
Secondary vascular risk reduction is critical to preventing recurrent stroke. We aimed to evaluate the effect of exercise interventions on vascular risk factors and recurrent ischaemic events after stroke or transient ischaemic attack (TIA). Intervention systematic review and meta-analysis. OVID MEDLINE, PubMed, The Cochrane Library, Web of Science, The National Institute for Health and Care Excellence, TRIP Database, CINAHL, PsycINFO, SCOPUS, UK Clinical Trials Gateway and the China National Knowledge Infrastructure were searched from 1966 to October 2017. Randomised controlled trials evaluating aerobic or resistance exercise interventions on vascular risk factors and recurrent ischaemic events among patients with stroke or TIA, compared with control. Twenty studies (n=1031) were included. Exercise interventions resulted in significant reductions in systolic blood pressure (SBP) -4.30 mm Hg (95% CI -6.77 to -1.83) and diastolic blood pressure -2.58 mm Hg (95% CI -4.7 to -0.46) compared with control. Reduction in SBP was most pronounced among studies initiating exercise within 6 months of stroke or TIA (-8.46 mm Hg, 95% CI -12.18 to -4.75 vs -2.33 mm Hg, 95% CI -3.94 to -0.72), and in those incorporating an educational component (-7.81 mm Hg, 95% CI -14.34 to -1.28 vs -2.78 mm Hg, 95% CI -4.33 to -1.23). Exercise was also associated with reductions in total cholesterol (-0.27 mmol/L, 95% CI -0.54 to 0.00), but not fasting glucose or body mass index. One trial reported reductions in secondary vascular events with exercise, but was insufficiently powered. Exercise interventions can result in clinically meaningful blood pressure reductions, particularly if initiated early and alongside education. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2018. All rights reserved. No commercial use is permitted unless otherwise expressly granted.
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.
The calculation of dryout during flow and pressure transients
International Nuclear Information System (INIS)
James, P.W.; Whalley, P.B.
1981-01-01
The method for predicting dryout in a round tube by means of an annular flow model (Whalley et al 1974) is extended to cover the case where both inlet mass flux and pressure are time-dependent. The qualitative effects of an inlet pressure transient are assessed by performing a 'numerical experiment' and it is found that the predictions of the model represent reasonable physical trends. The relative merits of wo numerical solution schemes are also discussed
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
International Nuclear Information System (INIS)
Porter, W.H.L.
1982-11-01
To check containment performance of the CVTR, steam was injected above the operating floor through a 10 foot pipe cap containing the 1 inch diameter holes, at a steady rate of 102.8 lb/sec for a period of 166 seconds. This steam had an enthalpy of 1195 Btu/lb and was therefore not entirely typical of the much wetter material which would be rejected for the greater part of a true breached circuit accident. Pressure transients measured experimentally within the containment were compared with results calculated by the American code CONTEMPT and these results in turn have allowed the Winfrith code CLAPTRAP to be tested for consistency and to establish that the use of this code would have led to similar conclusions about the heat transfer coefficients at the heat absorbent surfaces. (U.K.)
Transient thermal creep of nuclear reactor pressure vessel type concretes
International Nuclear Information System (INIS)
Khoury, G.A.
1983-01-01
The immediate aim of the research was to study the transient thermal strain behaviour of four AGR type nuclear reactor concretes during first time heating in an unsealed condition to 600 deg. C. The work being also relevant to applications of fire exposed concrete structures. The programme was, however, expanded to serve a second more theoretical purpose, namely the further investigation of the strain development of unsealed concrete under constant, transient and cyclic thermal states in particular and the effect of elevated temperatures on concrete in general. The range of materials investigated included seven different concretes and three types of cement paste. Limestone, basalt, gravel and lightweight aggregates were employed as well as OPC and SRC cements. Cement replacements included pfa and slag. Test variables comprised two rates of heating (0.2 and 1 deg. C/minute), three initial moisture contents (moist as cast, air-dry and oven dry at 105 deg. C), two curing regimes (bulk of tests represented mass cured concrete), five stress levels (0, 10, 20, 30 and a few tests at 60% of the cold strength), two thermal cycles and levels of test temperature up to 720 deg. C. Supplementary, dilatometry, TGA and DTA tests were performed at CERL on individual samples of aggregate and cement paste which helped towards explaining the observed trends in the concretes. A simple formula was developed which relates the elastic thermal stresses generated from radial temperature gradients to the solution obtained from the transient heat conduction equation. Thermal stresses can, therefore, be minimized by reductions in the radius of the specimen and the rate of heating The results were confirmed by finite element analysis which indicate( tensile stresses in the central region and compressive stresses near the surf ace during heating which are reversed during cooling. It is shown that the temperature gradients, pore pressures and tensile thermal stresses during both heating and
CHF during flow rate, pressure and power transients in heated channels
International Nuclear Information System (INIS)
Celata, G.P.; Cumo, M.
1987-01-01
The behaviour of forced two-phase flows following inlet flow rate, pressure and power transients is presented here with reference to experiments performed with a R-12 loop. A circular duct, vertical test section (L = 2300 mm; D = 7.5 mm) instrumented with fluid (six) and wall (twelve) thermocouples has been employed. Transients have been carried out performing several values of flow decays (exponential decrease), depressurization rates (exponential decrease) and power inputs (step-wise increase). Experimental data have shown the complete inadequacy of steady-state critical heat flux correlations in predicting the onset of boiling crisis during fast transients. Data analysis for a better theoretical prediction of CHF occurrence during transient conditions has been accomplished, and design correlations for critical heat flux and time-to-crisis predictions have been proposed for the different types of transients
Some local dilution transient in a pressurized water reactor
International Nuclear Information System (INIS)
Jacobson, S.
1989-01-01
Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)
Pressure Transient Model of Water-Hydraulic Pipelines with Cavitation
Directory of Open Access Journals (Sweden)
Dan Jiang
2018-03-01
Full Text Available Transient pressure investigation of water-hydraulic pipelines is a challenge in the fluid transmission field, since the flow continuity equation and momentum equation are partial differential, and the vaporous cavitation has high dynamics; the frictional force caused by fluid viscosity is especially uncertain. In this study, due to the different transient pressure dynamics in upstream and downstream pipelines, the finite difference method (FDM is adopted to handle pressure transients with and without cavitation, as well as steady friction and frequency-dependent unsteady friction. Different from the traditional method of characteristics (MOC, the FDM is advantageous in terms of the simple and convenient computation. Furthermore, the mechanism of cavitation growth and collapse are captured both upstream and downstream of the water-hydraulic pipeline, i.e., the cavitation start time, the end time, the duration, the maximum volume, and the corresponding time points. By referring to the experimental results of two previous works, the comparative simulation results of two computation methods are verified in experimental water-hydraulic pipelines, which indicates that the finite difference method shows better data consistency than the MOC.
Pressurized transient otoacoustic emissions measured using click and chirp stimuli.
Keefe, Douglas H; Patrick Feeney, M; Hunter, Lisa L; Fitzpatrick, Denis F; Sanford, Chris A
2018-01-01
Transient-evoked otoacoustic emission (TEOAE) responses were measured in normal-hearing adult ears over frequencies from 0.7 to 8 kHz, and analyzed with reflectance/admittance data to measure absorbed sound power and the tympanometric peak pressure (TPP). The mean TPP was close to ambient. TEOAEs were measured in the ear canal at ambient pressure, TPP, and fixed air pressures from 150 to -200 daPa. Both click and chirp stimuli were used to elicit TEOAEs, in which the incident sound pressure level was constant across frequency. TEOAE levels were similar at ambient and TPP, and for frequencies from 0.7 to 2.8 kHz decreased with increasing positive and negative pressures. At 4-8 kHz, TEOAE levels were larger at positive pressures. This asymmetry is possibly related to changes in mechanical transmission through the ossicular chain. The mean TEOAE group delay did not change with pressure, although small changes were observed in the mean instantaneous frequency and group spread. Chirp TEOAEs measured in an adult ear with Eustachian tube dysfunction and TPP of -165 daPa were more robust at TPP than at ambient. Overall, results demonstrate the feasibility and clinical potential of measuring TEOAEs at fixed pressures in the ear canal, which provide additional information relative to TEOAEs measured at ambient pressure.
Energy Technology Data Exchange (ETDEWEB)
Massoud, M
1987-01-01
Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.
International Nuclear Information System (INIS)
Massoud, M.
1987-01-01
Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients
R.B. pressure and temperature transient following main steam line break
International Nuclear Information System (INIS)
Das, M.; Bhawal, R.N.; Prakash, P.
1989-01-01
The R.B. containment plays an important role in mitigating the consequences of any accident core. The analysis of Main Steam Line Break (MSLB), though not of relevance from activity release considerations, is essentially from structural integrity point of view. In this paper the outline of the likely scenario is drawn and the approach for thermal hydraulic simulation of the system for carrying out transient blowdown analysis is discussed. The results of the containment pressure and temperature transient analysis are also presented. (author). 4 refs., 7 figs
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 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.
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
Solar wind dynamic pressure variations and transient magnetospheric signatures
International Nuclear Information System (INIS)
Sibeck, D.G.; Baumjohann, W.
1989-01-01
Contrary to the prevailing popular view, we find some transient ground events with bipolar north-south signatures are related to variations in solar wind dynamic pressure and not necessarily to magnetic merging. We present simultaneous solar wind plasma observations for two previously reported transient ground events observed at dayside auroral latitudes. During the first event, originally reported by Lanzerotti et al. [1987], conjugate ground magnetometers recorded north-south magetic field deflections in the east-west and vertical directions. The second event was reported by Todd et al. [1986], we noted ground rader observations indicating strong northward then southward ionospheric flows. The events were associated with the postulated signatures of patchy, sporadic, merging of magnetosheath and magnetospheric magnetic field lines at the dayside magnetospause, known as flux transfer events. Conversely, we demonstrate that the event reported by Lanzerotti et al. was accompanied by a sharp increase in solar wind dynamic pressure, a magnetospheric compression, and a consequent ringing of the magnetospheric magnetic field. The event reported by Todd et al. was associated with a brief but sharp increase in the solar wind dynamic pressure. copyright American Geophysical Union 1989
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 modelling of a natural circulation loop under variable pressure
International Nuclear Information System (INIS)
Vianna, Andre L.B.; Faccini, Jose L.H.; Su, Jian; Instituto de Engenharia Nuclear
2017-01-01
The objective of the present work is to model the transient operation of a natural circulation loop, which is one-tenth scale in height to a typical Passive Residual Heat Removal system (PRHR) of an Advanced Pressurized Water Nuclear Reactor and was designed to meet the single and two-phase flow similarity criteria to it. The loop consists of a core barrel with electrically heated rods, upper and lower plena interconnected by hot and cold pipe legs to a seven-tube shell heat exchanger of countercurrent design, and an expansion tank with a descending tube. A long transient characterized the loop operation, during which a phenomenon of self-pressurization, without self-regulation of the pressure, was experimentally observed. This represented a unique situation, named natural circulation under variable pressure (NCVP). The self-pressurization was originated in the air trapped in the expansion tank and compressed by the loop water dilatation, as it heated up during each experiment. The mathematical model, initially oriented to the single-phase flow, included the heat capacity of the structure and employed a cubic polynomial approximation for the density, in the buoyancy term calculation. The heater was modelled taking into account the different heat capacities of the heating elements and the heater walls. The heat exchanger was modelled considering the coolant heating, during the heat exchanging process. The self-pressurization was modelled as an isentropic compression of a perfect gas. The whole model was computationally implemented via a set of finite difference equations. The corresponding computational algorithm of solution was of the explicit, marching type, as for the time discretization, in an upwind scheme, regarding the space discretization. The computational program was implemented in MATLAB. Several experiments were carried out in the natural circulation loop, having the coolant flow rate and the heating power as control parameters. The variables used in the
Transient modelling of a natural circulation loop under variable pressure
Energy Technology Data Exchange (ETDEWEB)
Vianna, Andre L.B.; Faccini, Jose L.H.; Su, Jian, E-mail: avianna@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br, E-mail: faccini@ien.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Termo-Hidraulica Experimental
2017-07-01
The objective of the present work is to model the transient operation of a natural circulation loop, which is one-tenth scale in height to a typical Passive Residual Heat Removal system (PRHR) of an Advanced Pressurized Water Nuclear Reactor and was designed to meet the single and two-phase flow similarity criteria to it. The loop consists of a core barrel with electrically heated rods, upper and lower plena interconnected by hot and cold pipe legs to a seven-tube shell heat exchanger of countercurrent design, and an expansion tank with a descending tube. A long transient characterized the loop operation, during which a phenomenon of self-pressurization, without self-regulation of the pressure, was experimentally observed. This represented a unique situation, named natural circulation under variable pressure (NCVP). The self-pressurization was originated in the air trapped in the expansion tank and compressed by the loop water dilatation, as it heated up during each experiment. The mathematical model, initially oriented to the single-phase flow, included the heat capacity of the structure and employed a cubic polynomial approximation for the density, in the buoyancy term calculation. The heater was modelled taking into account the different heat capacities of the heating elements and the heater walls. The heat exchanger was modelled considering the coolant heating, during the heat exchanging process. The self-pressurization was modelled as an isentropic compression of a perfect gas. The whole model was computationally implemented via a set of finite difference equations. The corresponding computational algorithm of solution was of the explicit, marching type, as for the time discretization, in an upwind scheme, regarding the space discretization. The computational program was implemented in MATLAB. Several experiments were carried out in the natural circulation loop, having the coolant flow rate and the heating power as control parameters. The variables used in the
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)
Energy Technology Data Exchange (ETDEWEB)
Montazeri, G.H. [Islamic Azad University, Mahshahr (Iran, Islamic Republic of). Dept. of Chemical and Petroleum Engineering], E-mail: montazeri_gh@yahoo.com; Tahami, S.A. [Mad Daneshgostar Tabnak Co. (MDT),Tehran (Iran, Islamic Republic of); Moradi, B.; Safari, E. [Iranian Central Oil Fields Co, Tehran (Iran, Islamic Republic of)], E-mail: morady.babak@gmail.com
2011-07-15
This paper presents a model for pressure transient and derivative analysis for naturally fractured reservoirs by a formulation of inter porosity flow incorporating variations in matrix block size, which is inversely related to fracture intensity. Geologically realistic Probability Density Functions (PDFs) of matrix block size, such as uniform, bimodal, linear and exponential distributions, are examined and pseudo-steady-state and transient models for inter porosity flow are assumed. The results have been physically interpreted, and, despite results obtained by other authors, it was found that the shape of pressure derivative curves for different PDFs are basically identical within some ranges of block size variability, inter porosity skin, PDFs parameters and matrix storage capacity. This tool can give an insight on the distribution of block sizes and shapes, together with other sources of information such as Logs and geological observations. (author)
Computing the effect of plastic deformation of piping on pressure transient propagation
International Nuclear Information System (INIS)
Youngdahl, C.K.; Kot, C.A.
1977-01-01
The computer program PTA-1 performs pressure-transient analysis of large piping networks using the one-dimensional method of characteristics applied to a fluid-hammer formulation. The effect of elastic-plastic deformation of piping on pulse propagation is included in the computation. Each pipe is modeled as a series of rings, neglecting axial effects, bending moments, and inertia. The fluid wave speed is a function of pipe deformation and, consequently, of position and time. Comparison with existing experimental data indicate that this simple fluid-structure interaction model gives suprisingly accurate results for both pressure histories in the fluid and strain histories in the piping
A Study of System Pressure Transients Generated by Isolation Valve Open/Closure in Orifice Manifold
Energy Technology Data Exchange (ETDEWEB)
Kim, M. [KEPCO, Daejeon (Korea, Republic of); Bae, S. W.; Kim, J. I.; Park, S. J. [KHNP, Abu Dhabi (United Arab Emirates)
2016-05-15
In this study, we explore the effects of pressure transients on peak and minimal pressures caused by the actuation of isolation valve and control valve reacting to the combined orifice operation of orifice manifold with motor-operated valve installed in the rear of the orifice. We then use the collected data to direct our effort towards cause analysis and propose improvements to efficiency and safety of operation. This formation is used to by domestic and foreign nuclear power plants as a mean to control flow rate, producing required flow rate jointly together by combination of the orifices. No significant impacts on the internals of manifold orifice due to peak pressure has been observed, although chance of cavitation at the outlet of control valve is significant. Considering the peak pressure, as well as minimum pressure occurs in low flow rate conditions, the pressure transient is more so affected by the characteristics (modified equal percentage) of control valve. Isolation valve of the orifice and control valve operate organically, therefore stroke time for valves need to be applied in order for both valves to cooperatively formulate an optimized operation.
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.
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
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
Transient response of a five-region nonequilibrium real-time pressurizer model
International Nuclear Information System (INIS)
Fakory, M.R.; Seifaee, F.
1987-01-01
Recent accidents at nuclear power plants in the US and abroad have prompted accurate analysis and simulation of the plant systems and the training of reactor operators on plant-specific simulators that are equipped with the simulation models. Consequently, several models for real-time and off-time simulation of nuclear reactor systems, with various levels of accuracy and simulation fidelity, have been introduced. Experience with power plant simulation demonstrates that in order to realistically predict and simulate reactor responses during unanticipated transients, it is necessary to equip the simulation model with a multielement pressurizer model. The objective of this paper is to present the results of a five-region drift-flux-based pressurizer model, which has been developed for integration with real-time training simulators. A comparison between the plant data and the results of the nonequilibrium pressurizer model demonstrates that the model is well capable of close simulation of dynamic behavior of the pressurizer system
International Nuclear Information System (INIS)
Xu Mingyu; Lin Tengjiao; Li Runfang; Du Xuesong; Li Shuian; Yang Yu
2005-01-01
There are some complex operating cases such as high temperature and high pressure during the operating process of nuclear reactor pressure vessel. It is necessary to carry out mechanical analysis and experimental investigation for its sealing ability. On the basis of the self-developed program for 3-D transient sealing analysis for nuclear reactor pressure vessel, some specific measures are presented to enhance the calculation efficiency in several aspects such as the non-linear solution of elasto-plastic problem, the mixed solution algorithm for contact problem as well as contract heat transfer problem and linear equation set solver. The 3-D transient sealing analysis program is amended and complemented, with which the sealing analysis result of the pressure vessel model can be obtained. The calculation results have good regularity and the calculation efficiency is twice more than before. (authors)
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
Energy Technology Data Exchange (ETDEWEB)
Kliem, Soeren
2010-08-15
A model for the realistic description of the coolant mixing inside the pressure vessel of a pressurized water reactors has been developed and validated. This fast running model is based on the linear superposition of response functions on Dirac-pulse-like perturbations of the coolant parameters. It has been implemented into the coupled code system DYN3D/ATHLET nad serves as the interface between the one-dimensional thermal hydraulic system code ATHLETE and the 3D neutron kinetic core model DYN3D. By help of this model, the coolant mixing inside the reactor pressure vessel can e simulated in an efficient manner. A methodology for this analysis of hypothetical boron dilution accidents has been developed, which is based on the newly developed model for the coolant mixing. This methodology consists of a combination of stationary and transient calculations including a realistic treatment of the mixing of deborated slugs on the way towards the reactor core. The degree of conservatism can be adjusted by the variation of the initial size of the deborated slug. This new method was applied to two different boron dilution accidents. Besides the start of the first main coolant pump with a deborated slug of coolant in the cold leg of the primary circuit, a deboration event during the operation of the residual heat removal system was investigated. The results of the parameter study for a reactor core with a generic loading pattern demonstrated in both cases, that although the shut-down reactor becomes re-critical safety relevant cladding temperature limits are not reached, even if maximum possible volumes of the deborated slug are considered. The main reason for these results is the use of realistic time-dependent distributions of the boron concentration at the inlet of each fuel assembly.
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)
International Nuclear Information System (INIS)
Porter, W.H.L.
1976-05-01
This paper gives an appreciation and commentary of the basic calculation methods under development at AEE Winfrith for the analysis of multicompartment total containments. The assumptions introduced and the effects of their variation are important in establishing a parametric survey of the range of possible conditions which the containment may be required to meet. These aspects of the performance will be discussed as each individual factor in the train of events is examined in turn. (U.K.)
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)
Consideration of loading conditions initiated by thermal transients in PWR pressure vessels
International Nuclear Information System (INIS)
Azodi; Glahn; Kersting; Schulz; Jansky.
1983-01-01
This report describes the present state of PWR-plants in the Federal Republic of Germany with respect to - the design of the primary pressure boundary - the analysis of thermal transients and resulting loads - the material conditions and neutron fluence - the requirements for protection against fast fracture. The experimental and analytical research and development programs are delineated together with some foreign R and D programs. It is shown that the parameters investigated (loading condition, crack shape and orientation etc.) cover a broad range. Extensive analytical investigations are emphasized. (orig./RW) [de
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
Study of Fast Transient Pressure Drop in VVER-1000 Nuclear Reactor Using Acoustic Phenomenon
Directory of Open Access Journals (Sweden)
Soroush Heidari Sangestani
2018-01-01
Full Text Available This article aims to simulate the sudden and fast pressure drop of VVER-1000 reactor core coolant, regarding acoustic phenomenon. It is used to acquire a more accurate method in order to simulate the various accidents of reactor core. Neutronic equations should be solved concurrently by means of DRAGON 4 and DONJON 4 coupling codes. The results of the developed package are compared with WIMS/CITATION and final safety analysis report of Bushehr VVER-1000 reactor (FSAR. Afterwards, time dependent thermal-hydraulic equations are answered by employing Single Heated Channel by Sectionalized Compressible Fluid method. Then, the obtained results were validated by the same transient simulation in a pressurized water reactor core. Then, thermal-hydraulic and neutronic modules are coupled concurrently by use of producing group constants regarding the thermal feedback effect. Results were compared to the mentioned transient simulation in RELAP5 computer code, which show that mass flux drop is sensed at the end of channel in several milliseconds which causes heat flux drop too. The thermal feedback resulted in production of some perturbations in the changes of these parameters. The achieved results for this very fast pressure drop represent accurate calculations of thermoneutronic parameters fast changes.
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.)
PRESSURE PULSES AT VOYAGER 2 : DRIVERS OF INTERSTELLAR TRANSIENTS?
Energy Technology Data Exchange (ETDEWEB)
Richardson, J. D. [Kavli Center for Astrophysics and Space Science, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Wang, C.; Liu, Y. D. [State Key Laboratory for Space Weather, Chinese Academy of Sciences, Beijing (China); Šafránková, J.; Němeček, Z. [Charles University, Faculty of Mathematics and Physics, V Holešovičkách 2, 180 00 Prague 8 (Czech Republic); Kurth, W. S., E-mail: jdr@space.mit.edu, E-mail: cw@spaceweather.ac.cn, E-mail: liuxying@spaceweather.ac.cn, E-mail: jana.safrankova@mff.cuni.cz, E-mail: william-kurth@uiowa.edu [University of Iowa, Iowa City, IA 52242 (United States)
2017-01-10
Voyager 1 ( V1 ) crossed the heliopause into the local interstellar medium (LISM) in 2012. The LISM is a dynamic region periodically disturbed by solar transients with outward-propagating shocks, cosmic-ray intensity changes and anisotropies, and plasma wave oscillations. Voyager 2 ( V2 ) trails V1 and thus may observe the solar transients that are later observed at V1. V2 crossed the termination shock in 2007 and is now in the heliosheath. Starting in 2012, when solar maximum conditions reached V2 , five possible merged interaction regions (MIRs) have been observed by V2 in the heliosheath. The timing is consistent with these MIRs driving the transients observed by V1 in the LISM. The largest heliosheath MIR was observed by V2 in late 2015 and should reach V1 in 2018.
TRAC-PF1 analyses of potential pressurized-thermal-shock transients at a Combustion-Engineering PWR
International Nuclear Information System (INIS)
Koenig, J.E.; Spriggs, G.D.; Smith, R.C.
1984-01-01
Los Alamos is participating in a program to assess the risk of pressurized thermal shock (PTS) to a reactor vessel. Our role is to provide best-estimate thermal-hydraulic analyses of 12 postulated overcooling transients using TRAC-PF1. These transients are hypothetical and include multiple operator/equipment failures. Calvert Cliffs/Unit-1, a Combustion-Engineering plant, is the pressurized water reactor modeled for this study. The utility and the vendor supplied information for the comprehensive TRAC-PF1 model. Secondary and primary breaks from both hot-zero-power and full-power conditions were simulated for 7200 s (2 h). Low bulk temperatures and loop-flow stagnation while the system was at a high pressure were of particular interest for PTS analysis. Three transients produced primary temperatures below 405 K (270 0 F - the NRC screening criterion) with system repressurization. Six transients indicated flow stagnation would occur in one loop but not both. One transient showed flow stagnation might occur in both loops. Oak Ridge National Laboratory will do fracture-mechanics analysis using these TRAC-PF1 results and make the final determination of the risk of PTS
Transient 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.
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)
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
International Nuclear Information System (INIS)
Rajput, A.K.
1984-01-01
The study of sodium water reaction, following a large leak, concerns primarily with the estimation of pressure/flow transients that are developed in the steam generator and the associated secondary circuit. This paper describes the mathematical formulations used in SWRT (Sodium Water Reaction Transients) code developed to estimate such pressure transients for FBTR plant. The results, obtained using SWRT have been presented for a leak in economiser (20m from bottom water header) and for a leak in super heater portions. A time lag of 50 m sec was considered for rupture disc takes to burst once the pressure experienced by it exceeds the set value. Also described in annexure to this paper is the mathematical formulation for two phase transient flow for the better estimation of leak rate from the ruptured end of the damaged heat transfer tube. This leak model considers slip but assumes thermal equilibrium between the liquid and vapour phases
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)
International Nuclear Information System (INIS)
Youngdahl, C.K.; Kot, C.A.
1977-01-01
Pressure pulses in the intermediate sodium system of a liquid-metal-cooled fast breeder reactor, such as may originate from a sodium/water reaction in a steam generator, are propagated through the complex sodium piping network to system components such as the pump and intermediate heat exchanger. To assess the effects of such pulses on continued reliable operation of these components and to contribute to system designs which result in the mitigation of these effects, Pressure Transient Analysis (PTA) computer codes are being developed for accurately computing the transmission of pressure pulses through a complicated fluid transport system, consisting of piping, fittings and junctions, and components. PTA-1 provides an extension of the well-accepted and verified fluid hammer formulation for computing hydraulic transients in elastic or rigid piping systems to include plastic deformation effects. The accuracy of the modeling of pipe plasticity effects on transient propagation has been validated using results from two sets of Stanford Research Institute experiments. Validation of PTA-1 using the latter set of experiments is described briefly. The comparisons of PTA-1 computations with experiments show that (1) elastic-plastic deformation of LMFBR-type piping can have a significant qualitative and quantitative effect on pressure pulse propagation, even in simple systems; (2) classical fluid-hammer theory gives erroneous results when applied to situations where piping deforms plastically; and (3) the computational model incorporated in PTA-1 for predicting plastic deformation and its effect on transient propagation is accurate
High-temperature transient creep properties of CANDU pressure tubes
International Nuclear Information System (INIS)
Fong, R.W.L.; Chow, C.K.
2002-06-01
During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)
Transient heating effects in high pressure Diesel injector nozzles
International Nuclear Information System (INIS)
Strotos, George; Koukouvinis, Phoevos; Theodorakakos, Andreas; Gavaises, Manolis; Bergeles, George
2015-01-01
Highlights: • Simulation of friction-induced heating in high pressure Diesel fuel injectors. • Injection pressures up to 3000 bar. • Simulations with variable fuel properties significantly affect predictions. • Needle motion affects flow and temperature fields. • Possible heterogeneous boiling as injection pressures increase above 2000 bar. - Abstract: The tendency of today’s fuel injection systems to reach injection pressures up to 3000 bar in order to meet forthcoming emission regulations may significantly increase liquid temperatures due to friction heating; this paper identifies numerically the importance of fuel pressurization, phase-change due to cavitation, wall heat transfer and needle valve motion on the fluid heating induced in high pressure Diesel fuel injectors. These parameters affect the nozzle discharge coefficient (C d ), fuel exit temperature, cavitation volume fraction and temperature distribution within the nozzle. Variable fuel properties, being a function of the local pressure and temperature are found necessary in order to simulate accurately the effects of depressurization and heating induced by friction forces. Comparison of CFD predictions against a 0-D thermodynamic model, indicates that although the mean exit temperature increase relative to the initial fuel temperature is proportional to (1 − C d 2 ) at fixed needle positions, it can significantly deviate from this value when the motion of the needle valve, controlling the opening and closing of the injection process, is taken into consideration. Increasing the inlet pressure from 2000 bar, which is the pressure utilized in today’s fuel systems to 3000 bar, results to significantly increased fluid temperatures above the boiling point of the Diesel fuel components and therefore regions of potential heterogeneous fuel boiling are identified
Linear pressure profile estimation along a penstock associated with transients due to severe defects
Kueny, J. L.; Combes, G.; Lourenço, M.; Clary, V.; Ballester, J. L.
2014-03-01
The purpose of this article is to show how the pressure load profile along a penstock of an hydroplant and the corresponding flow rate is obtained from the pressure signal using a code called ACHYL CF. In particular the paper will present how it is possible to reconstruct the history of the incident after a strong transient state, in the case of two plants with Pelton turbines and one DSPCF device on a branch of the circuit. For plant1 the DSPCF device observes an overrun of the maximal allowed pressure after the filling of the injector branch and for plant_2, a strong transient leads to the rupture of the penstock.
Linear pressure profile estimation along a penstock associated with transients due to severe defects
International Nuclear Information System (INIS)
Kueny, J L; Clary, V; Combes, G; Lourenço, M; Ballester, J L
2014-01-01
The purpose of this article is to show how the pressure load profile along a penstock of an hydroplant and the corresponding flow rate is obtained from the pressure signal using a code called ACHYL CF. In particular the paper will present how it is possible to reconstruct the history of the incident after a strong transient state, in the case of two plants with Pelton turbines and one DSPCF device on a branch of the circuit. For plant 1 the DSPCF device observes an overrun of the maximal allowed pressure after the filling of the injector branch and for plant 2 , a strong transient leads to the rupture of the penstock
Comparison of pressure transient response in intensely and sparsely fractured reservoirs
Energy Technology Data Exchange (ETDEWEB)
Johns, R.T.
1989-04-01
A comprehensive analytical model is presented to study the pressure transient behavior of a naturally fractured reservoir with a continuous matrix block size distribution. Geologically realistic probability density functions of matrix block size are used to represent reservoirs of varying fracture intensity and uniformity. Transient interporosity flow is assumed and interporosity skin is incorporated. Drawdown and interference pressure transient tests are investigated. The results show distinctions in the pressure response from intensely and sparsely fractured reservoirs in the absence of interporosity skin. Also, uniformly and nonuniformly fractured reservoirs exhibit distinct responses, irrespective of the degree of fracture intensity. The pressure response in a nonuniformly fractured reservoir with large block size variability, approaches a nonfractured (homogeneous) reservoir response. Type curves are developed to estimate matrix block size variability and the degree of fracture intensity from drawdown and interference well tests.
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)
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
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
Characterizing hydraulic fractures in shale gas reservoirs using transient pressure tests
Directory of Open Access Journals (Sweden)
Cong Wang
2015-06-01
This work presents an unconventional gas reservoir simulator and its application to quantify hydraulic fractures in shale gas reservoirs using transient pressure data. The numerical model incorporates most known physical processes for gas production from unconventional reservoirs, including two-phase flow of liquid and gas, Klinkenberg effect, non-Darcy flow, and nonlinear adsorption. In addition, the model is able to handle various types and scales of fractures or heterogeneity using continuum, discrete or hybrid modeling approaches under different well production conditions of varying rate or pressure. Our modeling studies indicate that the most sensitive parameter of hydraulic fractures to early transient gas flow through extremely low permeability rock is actually the fracture-matrix contacting area, generated by fracturing stimulation. Based on this observation, it is possible to use transient pressure testing data to estimate the area of fractures generated from fracturing operations. We will conduct a series of modeling studies and present a methodology using typical transient pressure responses, simulated by the numerical model, to estimate fracture areas created or to quantity hydraulic fractures with traditional well testing technology. The type curves of pressure transients from this study can be used to quantify hydraulic fractures in field application.
Cavitation nuclei in water exposed to transient pressures
DEFF Research Database (Denmark)
Andersen, Anders Peter; Mørch, Knud Aage
2015-01-01
A model of skin-stabilized interfacial cavitation nuclei and their response to tensile and compressive stressing is presented. The model is evaluated in relation to experimental tensile strength results for water at rest at the bottom of an open water-filled container at atmospheric pressure...... and room temperature. These results are obtained by recording the initial growth of cavities generated by a short tensile pulse applied to the bottom of the container. It is found that the cavitation nuclei shift their tensile strength depending on their pressure history. Static pressurization...... for an extended period of time prior to testing is known to increase the tensile strength of water, but little information is available on how it is affected by compression pulses of short duration. This is addressed by imposing compression pulses of approximately 1 ms duration and a peak intensity of a few bar...
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
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
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.
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)
The assessment of RELAP5/MOD2 based on pressurizer transient experiments
International Nuclear Information System (INIS)
Xue Hanjun; Tanrikut, A.; Menzel, R.
1992-03-01
Two typical experiments have been performed in Chinese test facility under full pressure load corresponding to typical PWRs, 1) dynamic behavior of pressurizer due to relief valve operations (Case-I) is extremely important in transients and accident conditions regarding depressurization of PWR primary system; 2) Outsurge/Insurge operation is one of the transient which is often encountered and experienced in pressurizer systems due to pressure transients in primary system of PWRs. The simulation capability of RELAP5/MOD2 is good in comparison to experimental results. The physical models (such as interface model, stratification model), playing a major role in such simulation, seems to be realistic. The effect of realistic valve modeling in depressurization simulation is extremely important. Sufficient data for relief valve (the dynamic characteristics of valve) play a major role. The time dependent junction model and the trip valve model with a reduced discharge coefficient of 0.2 give better predictions in agreement with the experiment data while the trip valve models with discharge coefficient 1.0 yield overdepressurization. The simulation of outsurge/insurge transient yields satisfactory results. The thermal non-equilibrium model is important with respect to simulation of complicated physical phenomena in outsurge/insurge transient but has a negligible effect upon the depressurization simulation. (orig./HP)
Leak detection in medium density polyethylene (MDPE) pipe using pressure transient method
Amin, M. M.; Ghazali, M. F.; PiRemli, M. A.; Hamat, A. M. A.; Adnan, N. F.
2015-12-01
Water is an essential part of commodity for a daily life usage for an average person, from personal uses such as residential or commercial consumers to industries utilization. This study emphasizes on detection of leaking in medium density polyethylene (MDPE) pipe using pressure transient method. This type of pipe is used to analyze the position of the leakage in the pipeline by using Ensemble Empirical Mode Decomposition Method (EEMD) with signal masking. Water hammer would induce an impulse throughout the pipeline that caused the system turns into a surge of water wave. Thus, solenoid valve is used to create a water hammer through the pipelines. The data from the pressure sensor is collected using DASYLab software. The data analysis of the pressure signal will be decomposed into a series of wave composition using EEMD signal masking method in matrix laboratory (MATLAB) software. The series of decomposition of signals is then carefully selected which reflected intrinsic mode function (IMF). These IMFs will be displayed by using a mathematical algorithm, known as Hilbert transform (HT) spectrum. The IMF signal was analysed to capture the differences. The analyzed data is compared with the actual measurement of the leakage in term of percentage error. The error recorded is below than 1% and it is proved that this method highly reliable and accurate for leak detection.
Effect of helium pressure on the response of unirradiated UO2 subjected to thermal transients
International Nuclear Information System (INIS)
Fenske, G.R.; Chapello, P.M.; Emerson, J.E.; Poeppel, R.B.
1983-01-01
The effect of helium pressure on the transient response of unirradiated depleted UO 2 subjected to simulated hypothetical loss-of-flow accidents in a gas-cooled fast reactor was examined by use of the direct electrical heating technique. Transient tests were performed at pressures ranging from 7 to 10 X 10 5 Pa(7 to 10 atm) to 7 to 8 MPa (70 to 80 atm) on radially restrained and unrestrained fuel segments. The average heating rates ranged from about17 to 240 J/g x s. The results indicate that while the mechanical integrity of the fuel segment was independent of the test pressure, the rapid ejection of molten fuel from pellet interfaces of unrestrained fuel, observed at the lower pressures, was delayed or suppressed at the higher pressures
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
RELAP5 analyses of overcooling transients in a pressurized water reactor
International Nuclear Information System (INIS)
Bolander, M.A.; Fletcher, C.D.; Ogden, D.M.; Stitt, B.D.; Waterman, M.E.
1983-01-01
In support of the Pressurized Thermal Shock Integration Study sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.5 computer code. These analyses were performed for Oconee Plants 1 and 3, which are pressurized water reactors of Babcock and Wilcox lowered-loop design. Results of the RELAP5 analyses are presented, including a comparison with plant data. The capabilities and limitations of the RELAP5/MOD1.5 computer code in analyzing integral plant transients are examined. These analyses require detailed thermal-hydraulic and control system computer models
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.
Pressurized water reactor iodine spiking behavior under power transient conditions
International Nuclear Information System (INIS)
Ho, J.C.
1992-01-01
The most accepted theory explaining the cause of pressurized water reactor iodine spiking is steam formation and condensation in damaged fuel rods. The phase transformation of the primary coolant from water to steam and back again is believed to cause the iodine spiking phenomenon. But due to the complex nature of the phenomenon, a comprehensive model of the behavior has not yet been successfully developed. This paper presents a new model based on an empirical approach, which gives a first-order estimation of the peak iodine spiking magnitude. Based on the proposed iodine spiking model, it is apparent that it is feasible to derive a correlation using the plant operating data base to monitor and control the peak iodine spiking magnitude
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)
Fatigue status assessment for reactor pressure vessel based on actual operational transient
International Nuclear Information System (INIS)
Zhu Guangqiang; Liao Changbin; Dai Bing; Gui Chun
2013-01-01
Background: Fatigue is an important aging mechanism in RPV and it must be contained to aging management working range. Purpose: In order to ensure the safety operation of nuclear power plants, as extension of RPV service time, it is necessary to assess the fatigue damage caused by actual operation transient. Methods: Based on monitoring data of actual operation during the past eleven years, refer to design transient, the statistic analysis for types and occurrence times of actual transient is carried out, at the same time, every transients are combined as different operation cycles and the temperature field and stress field of typical components are analyzed by FEM. Results: Based on these information, fatigue analysis and assessment are finished, if later-actual transients are similar with the previous transients, the calculation result shows that the ratio between maximum of cumulative usage factors and design calculation value is 0.4967 the design transients is conservative. Conclusions: Fatigue status of RPV could be assessed and traced quickly through fatigue status assessment method in this paper based on actual operational transient and assessment result would be a good reference for RPV aging management. (authors)
A model for the calculation of vent clearing transients in pressure suppression systems
International Nuclear Information System (INIS)
Brosche, D.
1975-01-01
For the layout of a pressure suppression system of a light water cooled reactor (boiling water reactor) it is important to know the time dependent behavior of the vent clearing transient after a loss-of-coolant accident for two main reasons: time of the end of the vent clearing transient influences strongly the pressure and temperature maxima in the drywell and wetwell. Time-dependent behavior of the vent clearing transient influences pressure loads in the condensation pool of the wetwell and therefore pressure induced stresses to the structure. The time-dependent behavior of the water masses in the vent pipes and wetwell are described by the basic equations for a nonstationary incompressible friction flow: momentum equation, continuity equation and a correlation for the variation of the state of the gas volume in the wetwell above the water level. After many algebraic operations and integrations along the flow path, a single ordinary nonlinear differential equation for the variations of the water levels in the vent pipes and wetwell is obtained. Therefore the time-dependent velocities and accelerations of the water levels and the moment of the end clearing transient are known. The time-dependent pressure behavior in the drywell, geometrical conditions, initial submergence depth of the vent pipes and different friction and pressure loss factors are presented. The theoretical model has been tested at corresponding experiments performed at a full scale 1/48 segment of the Humboldt Bay pressure suppression containment in the USA and at the pressure suppression containment at the Marviken nuclear power station in Sweden. All these comparisons have shown good agreement between theory and experiment
SOIL-AIR PERMEABILITY MEASUREMENT WITH A TRANSIENT PRESSURE BUILDUP METHOD
An analytical solution for transient pressure change in a single venting well was derived from mass conservation of air, Darcy's law of flow in porous media, and the ideal gas law equation of state. Slopes of plots of Pw2 against ln (t+Δt)/Δt similar to Homer's plot were used to ...
Transient receptor potential canonical type 3 channels and blood pressure in humans
DEFF Research Database (Denmark)
Thilo, Florian; Baumunk, Daniel; Krause, Hans
2009-01-01
There is evidence that transient receptor potential canonical type 3 (TRPC3) cation channels are involved in the regulation of blood pressure, but this has not been studied using human renal tissue. We tested the hypothesis that the expression of TRPC3 in human renal tissue is associated with blood...
Scaling of two-phase flow transients using reduced pressure system and simulant fluid
International Nuclear Information System (INIS)
Kocamustafaogullari, G.; Ishii, M.
1987-01-01
Scaling criteria for a natural circulation loop under single-phase flow conditions are derived. Based on these criteria, practical applications for designing a scaled-down model are considered. Particular emphasis is placed on scaling a test model at reduced pressure levels compared to a prototype and on fluid-to-fluid scaling. The large number of similarty groups which are to be matched between modell and prototype makes the design of a scale model a challenging tasks. The present study demonstrates a new approach to this clasical problen using two-phase flow scaling parameters. It indicates that a real time scaling is not a practical solution and a scaled-down model should have an accelerated (shortened) time scale. An important result is the proposed new scaling methodology for simulating pressure transients. It is obtained by considerung the changes of the fluid property groups which appear within the two-phase similarity parameters and the single-phase to two-phase flow transition prameters. Sample calculations are performed for modeling two-phase flow transients of a high pressure water system by a low-pressure water system or a Freon system. It is shown that modeling is possible for both cases for simulation pressure transients. However, simulation of phase change transitions is not possible by a reduced pressure water system without distortion in either power or time. (orig.)
LeChevallier, Mark W; Gullick, Richard W; Karim, Mohammad R; Friedman, Melinda; Funk, James E
2003-03-01
The potential for public health risks associated with intrusion of contaminants into water supply distribution systems resulting from transient low or negative pressures is assessed. It is shown that transient pressure events occur in distribution systems; that during these negative pressure events pipeline leaks provide a potential portal for entry of groundwater into treated drinking water; and that faecal indicators and culturable human viruses are present in the soil and water exterior to the distribution system. To date, all observed negative pressure events have been related to power outages or other pump shutdowns. Although there are insufficient data to indicate whether pressure transients are a substantial source of risk to water quality in the distribution system, mitigation techniques can be implemented, principally the maintenance of an effective disinfectant residual throughout the distribution system, leak control, redesign of air relief venting, and more rigorous application of existing engineering standards. Use of high-speed pressure data loggers and surge modelling may have some merit, but more research is needed.
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
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
Haemers, Peter; Sutherland, George; Cikes, Maja; Jakus, Nina; Holemans, Patricia; Sipido, Karin R; Willems, Rik; Claus, Piet
2015-11-01
An acute increase in blood pressure is associated with the occurrence of premature ventricular complexes (PVCs). We aimed to study the timing of these PVCs with respect to afterload-induced changes in myocardial deformation in a controlled, preclinically relevant, novel closed-chest pig model. An acute left ventricular (LV) afterload challenge was induced by partial balloon inflation in the descending aorta, lasting 5-10 heartbeats (8 pigs; 396 inflations). Balloon inflation enhanced the reflected wave (augmentation index 30% ± 8% vs 59% ± 6%; P blood pressure by 35% ± 4%. This challenge resulted in a more abrupt LV pressure decline, which was delayed beyond ventricular repolarization (rate of pressure decline 0.16 ± 0.01 mm Hg/s vs 0.27 ± 0.04 mm Hg/ms; P pressure 1 ± 12 ms vs 36 ± 9 ms; P = .008), during which the velocity of myocardial shortening at the basal septum increased abruptly (ie, postsystolic shortening) (peak strain rate -0.6 ± 0.5 s(-1) vs -2.5 ± 0.8 s(-1); P pressure decline, with increased postsystolic shortening, and not at peak pressure, that PVCs occur (22% of inflations). These PVCs preferentially occurred at the basal and apical segments. In the same regions, monophasic action potentials demonstrated the appearance of delayed afterdepolarization-like transient depolarizations as origin of PVCs. An acute blood pressure increase results in a more abrupt LV pressure decline, which is delayed after ventricular repolarization. This has a profound effect on myocardial mechanics with enhanced postsystolic shortening. Coincidence with induced transient depolarizations and PVCs provides support for the mechanoelectrical origin of pressure-induced premature beats. Copyright © 2015 Heart Rhythm Society. Published by Elsevier Inc. All rights reserved.
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
International Nuclear Information System (INIS)
Shin, Y.W.; Wiedermann, A.H.
1984-02-01
A method was published, based on the integral method of characteristics, by which the junction and boundary conditions needed in computation of a flow in a piping network can be accurately formulated. The method for the junction and boundary conditions formulation together with the two-step Lax-Wendroff scheme are used in a computer program; the program in turn, is used here in calculating sample problems related to the blowdown transient of a two-phase flow in the piping network downstream of a PWR pressurizer. Independent, nearly exact analytical solutions also are obtained for the sample problems. Comparison of the results obtained by the hybrid numerical technique with the analytical solutions showed generally good agreement. The good numerical accuracy shown by the results of our scheme suggest that the hybrid numerical technique is suitable for both benchmark and design calculations of PWR pressurizer blowdown transients
Energy Technology Data Exchange (ETDEWEB)
Fukuda, K. [Kobe Univ. of Mercantile Marine (Japan); Shiotsu, M.; Sakurai, A. [Kyoto Univ. (Japan)
1995-09-01
Understanding of transient boiling phenomenon caused by increasing heat inputs in subcooled water at high pressures is necessary to predict correctly a severe accident due to a power burst in a water-cooled nuclear reactor. Transient maximum heat fluxes, q{sub max}, on a 1.2 mm diameter horizontal cylinder in a pool of saturated and subcooled water for exponential heat inputs, q{sub o}e{sup t/T}, with periods, {tau}, ranging from about 2 ms to 20 s at pressures from atmospheric up to 2063 kPa for water subcoolings from 0 to about 80 K were measured to obtain the extended data base to investigate the effect of high subcoolings on steady-state and transient maximum heat fluxes, q{sub max}. Two main mechanisms of q{sub max} exist depending on the exponential periods at low subcoolings. One is due to the time lag of the hydrodynamic instability which starts at steady-state maximum heat flux on fully developed nucleate boiling (FDNB), and the other is due to the heterogenous spontaneous nucleations (HSN) in flooded cavities which coexist with vapor bubbles growing up from active cavities. The shortest period corresponding to the maximum q{sub max} for long period range belonging to the former mechanism becomes longer and the q{sub max}mechanism for long period range shifts to that due the HSN on FDNB with the increase of subcooling and pressure. The longest period corresponding to the minimum q{sub max} for the short period range belonging to the latter mechanism becomes shorter with the increase in saturated pressure. On the contrary, the longest period becomes longer with the increase in subcooling at high pressures. Correlations for steady-state and transient maximum heat fluxes were presented for a wide range of pressure and subcooling.
International Nuclear Information System (INIS)
Fukuda, K.; Shiotsu, M.; Sakurai, A.
1995-01-01
Understanding of transient boiling phenomenon caused by increasing heat inputs in subcooled water at high pressures is necessary to predict correctly a severe accident due to a power burst in a water-cooled nuclear reactor. Transient maximum heat fluxes, q max , on a 1.2 mm diameter horizontal cylinder in a pool of saturated and subcooled water for exponential heat inputs, q o e t/T , with periods, τ, ranging from about 2 ms to 20 s at pressures from atmospheric up to 2063 kPa for water subcoolings from 0 to about 80 K were measured to obtain the extended data base to investigate the effect of high subcoolings on steady-state and transient maximum heat fluxes, q max . Two main mechanisms of q max exist depending on the exponential periods at low subcoolings. One is due to the time lag of the hydrodynamic instability which starts at steady-state maximum heat flux on fully developed nucleate boiling (FDNB), and the other is due to the heterogenous spontaneous nucleations (HSN) in flooded cavities which coexist with vapor bubbles growing up from active cavities. The shortest period corresponding to the maximum q max for long period range belonging to the former mechanism becomes longer and the q max mechanism for long period range shifts to that due the HSN on FDNB with the increase of subcooling and pressure. The longest period corresponding to the minimum q max for the short period range belonging to the latter mechanism becomes shorter with the increase in saturated pressure. On the contrary, the longest period becomes longer with the increase in subcooling at high pressures. Correlations for steady-state and transient maximum heat fluxes were presented for a wide range of pressure and subcooling
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
Comparison of BWR-6 pressurization transients with one-dimensional and point kinetics
International Nuclear Information System (INIS)
Serra, J.M.; Mata, P.; Cronin, J.T.
1992-01-01
This paper focuses on the differences between the results of core reload licensing calculations for the BWR-6 plant when performed with a one-dimensional (1-D) versus a point kinetics model. More specifically, the improvement in critical power ratio which would be expected from a change in methods from a point to a 1-D kinetics core wide transient calculation for pressurization transients is investigated. To qualitatively assess critical power ratio (CPR) improvement, core wide transient and hot channel calculations of a generator load rejection with failure of the steam by-pass system and a feedwater controller failure of maximum demand are performed with both, point and 1-D kinetics models in the core wide simulation. Additionally, a sensitivity study on the frequency of power shape function updating in the 1-D kinetics calculation is performed
Transient response of a liquid injector to a steep-fronted transverse pressure wave
Lim, D.; Heister, S.; Stechmann, D.; Kan, B.
2017-12-01
Motivated by the dynamic injection environment posed by unsteady pressure gain combustion processes, an experimental apparatus was developed to visualize the dynamic response of a transparent liquid injector subjected to a single steep-fronted transverse pressure wave. Experiments were conducted at atmospheric pressure with a variety of acrylic injector passage designs using water as the working fluid. High-speed visual observations were made of the injector exit near field, and the extent of backflow and the time to refill the orifice passage were characterized over a range of injection pressures. A companion transient one-dimensional model was developed for interpretation of the results and to elucidate the trends with regard to the strength of the transverse pressure wave. Results from the model were compared with the experimental observations.
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
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 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)
International Nuclear Information System (INIS)
Wang, Lei; Wang, Xiaodong
2014-01-01
Resulting from the nature of anisotropy of coal media, it is a meaningful work to evaluate pressure transient behavior and flow characteristics within coals. In this article, a complete analytical model called the elliptical flow model is established by combining the theory of elliptical flow in anisotropic media and Fick's laws about the diffusion of coalbed methane. To investigate pressure transient behavior, analytical solutions were first obtained through introducing a series of special functions (Mathieu functions), which are extremely complex and are hard to calculate. Thus, a computer program was developed to establish type curves, on which the effects of the parameters, including anisotropy coefficient, storage coefficient, transfer coefficient and rate constant, were analyzed in detail. Calculative results show that the existence of anisotropy would cause great pressure depletion. To validate new analytical solutions, previous results were used to compare with the new results. It is found that a better agreement between the solutions obtained in this work and the literature was achieved. Finally, a case study is used to explain the effects of the parameters, including rock total compressibility coefficient, coal medium porosity and anisotropic permeability, sorption time constant, Langmuir volume and fluid viscosity, on bottom-hole pressure behavior. It is necessary to coordinate these parameters so as to reduce the pressure depletion. (paper)
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)
The PARET code and the analysis of the SPERT I transients
Energy Technology Data Exchange (ETDEWEB)
Woodruff, William L [Argonne National Laboratory, Argonne (United States)
1983-09-01
The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients.
The PARET code and the analysis of the SPERT I transients
International Nuclear Information System (INIS)
Woodruff, William L.
1983-01-01
The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients
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.)
McManus, Richard J; Roalfe, Andrea; Fletcher, Kate; Taylor, Clare J; Martin, Una; Virdee, Satnam; Greenfield, Sheila; Hobbs, F D Richard
2016-01-01
Objective To assess whether using intensive blood pressure targets leads to lower blood pressure in a community population of people with prevalent cerebrovascular disease. Design Open label randomised controlled trial. Setting 99 general practices in England, with participants recruited in 2009-11. Participants People with a history of stroke or transient ischaemic attack whose systolic blood pressure was 125 mm Hg or above. Interventions Intensive systolic blood pressure target (different target, patients in both arms were actively managed in the same way with regular reviews by the primary care team. Main outcome measure Change in systolic blood pressure between baseline and 12 months. Results 529 patients (mean age 72) were enrolled, 266 to the intensive target arm and 263 to the standard target arm, of whom 379 were included in the primary analysis (182 (68%) intensive arm; 197 (75%) standard arm). 84 patients withdrew from the study during the follow-up period (52 intensive arm; 32 standard arm). Mean systolic blood pressure dropped by 16.1 mm Hg to 127.4 mm Hg in the intensive target arm and by 12.8 mm Hg to 129.4 mm Hg in the standard arm (difference between groups 2.9 (95% confidence interval 0.2 to 5.7) mm Hg; P=0.03). Conclusions Aiming for target below 130 mm Hg rather than 140 mm Hg for systolic blood pressure in people with cerebrovascular disease in primary care led to a small additional reduction in blood pressure. Active management of systolic blood pressure in this population using a blood pressure. Trial registration Current Controlled Trials ISRCTN29062286. PMID:26919870
International Nuclear Information System (INIS)
Knudson, D.L.; Dobbe, C.A.
1993-11-01
Containment integrity could be challenged by direct heating associated with a high pressure melt ejection (HPME) of core materials following reactor vessel breach during certain severe accidents. Intentional reactor coolant system (RCS) depressurization, where operators latch pressurizer relief valves open, has been proposed as an accident management strategy to reduce risks by mitigating the severity of HPME. However, decay heat levels, valve capacities, and other plant-specific characteristics determine whether the required operator action will be effective. Without operator action, natural circulation flows could heat ex-vessel RCS pressure boundaries (surge line and hot leg piping, steam generator tubes, etc.) to the point of failure before vessel breach, providing an alternate mechanism for RCS depressurization and HPME mitigation. This report contains an assessment of the potential for HPME during a Surry station blackout transient without operator action and without recovery. The assessment included a detailed transient analysis using the SCDAP/RELAP5/MOD3 computer code to calculate the plant response with and without hot leg countercurrent natural circulation, with and without reactor coolant pump seal leakage, and with variations on selected core damage progression parameters. RCS depressurization-related probabilities were also evaluated, primarily based on the code results
Characterization of Pressure Transients Generated by Nanosecond Electrical Pulse (nsEP) Exposure
Roth, Caleb C.; Barnes Jr., Ronald A.; Ibey, Bennett L.; Beier, Hope T.; Christopher Mimun, L.; Maswadi, Saher M.; Shadaram, Mehdi; Glickman, Randolph D.
2015-01-01
The mechanism(s) responsible for the breakdown (nanoporation) of cell plasma membranes after nanosecond pulse (nsEP) exposure remains poorly understood. Current theories focus exclusively on the electrical field, citing electrostriction, water dipole alignment and/or electrodeformation as the primary mechanisms for pore formation. However, the delivery of a high-voltage nsEP to cells by tungsten electrodes creates a multitude of biophysical phenomena, including electrohydraulic cavitation, electrochemical interactions, thermoelastic expansion, and others. To date, very limited research has investigated non-electric phenomena occurring during nsEP exposures and their potential effect on cell nanoporation. Of primary interest is the production of acoustic shock waves during nsEP exposure, as it is known that acoustic shock waves can cause membrane poration (sonoporation). Based on these observations, our group characterized the acoustic pressure transients generated by nsEP and determined if such transients played any role in nanoporation. In this paper, we show that nsEP exposures, equivalent to those used in cellular studies, are capable of generating high-frequency (2.5 MHz), high-intensity (>13 kPa) pressure transients. Using confocal microscopy to measure cell uptake of YO-PRO®-1 (indicator of nanoporation of the plasma membrane) and changing the electrode geometry, we determined that acoustic waves alone are not responsible for poration of the membrane. PMID:26450165
Characterization of Pressure Transients Generated by Nanosecond Electrical Pulse (nsEP) Exposure.
Roth, Caleb C; Barnes, Ronald A; Ibey, Bennett L; Beier, Hope T; Christopher Mimun, L; Maswadi, Saher M; Shadaram, Mehdi; Glickman, Randolph D
2015-10-09
The mechanism(s) responsible for the breakdown (nanoporation) of cell plasma membranes after nanosecond pulse (nsEP) exposure remains poorly understood. Current theories focus exclusively on the electrical field, citing electrostriction, water dipole alignment and/or electrodeformation as the primary mechanisms for pore formation. However, the delivery of a high-voltage nsEP to cells by tungsten electrodes creates a multitude of biophysical phenomena, including electrohydraulic cavitation, electrochemical interactions, thermoelastic expansion, and others. To date, very limited research has investigated non-electric phenomena occurring during nsEP exposures and their potential effect on cell nanoporation. Of primary interest is the production of acoustic shock waves during nsEP exposure, as it is known that acoustic shock waves can cause membrane poration (sonoporation). Based on these observations, our group characterized the acoustic pressure transients generated by nsEP and determined if such transients played any role in nanoporation. In this paper, we show that nsEP exposures, equivalent to those used in cellular studies, are capable of generating high-frequency (2.5 MHz), high-intensity (>13 kPa) pressure transients. Using confocal microscopy to measure cell uptake of YO-PRO®-1 (indicator of nanoporation of the plasma membrane) and changing the electrode geometry, we determined that acoustic waves alone are not responsible for poration of the membrane.
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
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
Thermal-hydraulics of the Loviisa reactor pressure vessel overcooling transients
International Nuclear Information System (INIS)
Tuomisto, Harri.
1987-06-01
In the Loviisa reactor pressure vessel safety analyses, the thermal-hydraulics of various overcooling transients has been evaluated to give pertinent initial data for fracture-mechanics calculations. The thermal-hydraulic simulations of the developed overcooling scenarios have been performed using best-estimate thermal-hydraulic computer codes. Experimental programs have been carried out to study phenomena related to natural circulation interruptions in the reactor coolant system. These experiments include buoyancy-induced phenomena such as thermal mixing and stratification of cold high-pressure safety injection water in the cold legs and the downcomer, and oscillations of the single-phase natural circulation. In the probabilistic pressurized thermal shock study, the Loviisa training simulator and the advanced system code RELAP5/MOD2 were utilized to simulate selected sequences. Flow stagnation cases were separately calculated with the REMIX computer program. The methods employed were assessed for these calculations against the plant data and own experiments
Transients in low pressure pumping circuits: a language oriented for the problem
International Nuclear Information System (INIS)
De Bernardinis, B.; Siccardi, F.
1977-01-01
Following a previous work (Vallombrosa 1974) a specialized language was developed for transients in low pressure pumping circuits, when the liquid column separation phenomenon may happen or is to be avoided. The first generation of the programming code is given. Numerical schemes go beyond the usual characteristic integration techniques now available and make it possible to atrack the solution of problems in which on the one hand, the differential equations are nonlinear on account of the variations of the celerity with pressure, and on the other, the pressure of a dispersed gaseous phase in the liquid influences the energetic dissipation mechanisms. The oriented language allows the simulation of the main constituents of the circuits, pumping stations, reservoirs, air tanks, piezometric wells, condensers, variable resistances, conduit junctions, both during normal functioning and in cavitation conditions. Special control instructions on the programming code allow such a simulation language to be easily employed even by people not specifically competent in computer progr
Steam-generator-tube-rupture transients for pressurized-water reactors
International Nuclear Information System (INIS)
Dobranich, D.; Henninger, R.J.; DeMuth, N.S.
1982-01-01
Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures
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
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.
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
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
International Nuclear Information System (INIS)
Gulshani, P.; So, C.B.
1986-10-01
In a number of postulated accident scenarios in a CANDU reactor, some of the horizontal fuel channels are predicted to experience periods of stratified channel coolant condition which can lead to a circumferential temperature gradient around the pressure tube. To study pressure tube strain and integrity under stratified flow channel conditions, it is, necessary to determine the pressure tube circumferential temperature distribution. This paper presents an algebraic model, called AMPTRACT (Algebraic Model for Pressure Tube TRAnsient Circumferential Temperature), developed to give the transient temperature distribution in a closed form. AMPTRACT models the following modes of heat transfer: radiation from the outermost elements to the pressure tube and from the pressure to calandria tube, convection between the fuel elements and the pressure tube and superheated steam, and circumferential conduction from the exposed to submerged part of the pressure tube. An iterative procedure is used to solve the mass and energy equations in closed form for axial steam and fuel-sheath transient temperature distributions. The one-dimensional conduction equation is then solved to obtain the pressure tube circumferential transient temperature distribution in a cosine series expansion. In the limit of large times and in the absence of convection and radiation to the calandria tube, the predicted pressure tube temperature distribution reduces identically to a parabolic profile. In this limit, however, radiation cannot be ignored because the temperatures are generally high. Convection and radiation tend to flatten the parabolic distribution
Modeling of boron control during power transients in a pressurized water reactor
International Nuclear Information System (INIS)
Mathieu, P.; Distexhe, E.
1986-01-01
Accurate control instructions in a reactor control aid computer are included in order to realize the boron makeup throughput, which is required to obtain the boron concentration in the primary coolant loop, predicted by a neutronic code. A modeling of the transfer function between the makeup and the primary loop is proposed. The chemical and volumetric control system, the pressurizer, and the primary loop are modeled as instantaneous diffusion cells. The pipes are modeled as time lag lines. The model provides the unstationary boron distributions in the different elements of the setup. A numerical code is developed to calculate the time evolutions of the makeup throughput during power transients
International Nuclear Information System (INIS)
Hamilton, M.L.; Johnson, G.D.; Hunter, C.W.; Duncan, D.R.
1982-11-01
Fast breeder fuel-pin cladding has been tested under experimental conditions simulating the temperature and pressure history characteristic of anticipated transient events. Irradiation induces severe reductions in both strength and ductility. Ductility losses are independent of the rate of temperature increase and saturate by a fluence of approx. 2 x 10 22 n/cm 2 (E > 0.1 MeV). Losses in strength are dependent on the rate of temperature increase but saturate at a fluence of approx.5 x 10 22 n/cm 2 . Evidence is presented to show that fission products are probably responsible for the degradation in mechanical properties
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
Effect of pressure on the transient swelling rate of oxide fuel
International Nuclear Information System (INIS)
Gruber, E.E.
1982-04-01
An analysis of the transient swelling rate of oxide fuel, based on fission-gas bubble conditions calculated with the FRAS3 code, has been developed and implemented in the code. The need for this capability arises in the coupling of the FRAS3 fission-gas analysis code to the FPIN fuel-pin mechanics code. An efficient means of closely coupling the calculations of swelling strains and stresses between the modules is required. The present analysis provides parameters that allow the FPIN calculation to proceed through a fairly large time step, using estimated swelling rates, to calculate the stresses. These stress values can then be applied in the FRAS3 detailed calculation to refine the swelling calculation, and to provide new values for the parameters to estimate the swelling in the next time step. The swelling rates were calculated for two representative transients and used to estimate swelling over a short time period for various stress levels
Reduced-order modellin for high-pressure transient flow of hydrogen-natural gas mixture
Agaie, Baba G.; Khan, Ilyas; Alshomrani, Ali Saleh; Alqahtani, Aisha M.
2017-05-01
In this paper the transient flow of hydrogen compressed-natural gas (HCNG) mixture which is also referred to as hydrogen-natural gas mixture in a pipeline is numerically computed using the reduced-order modelling technique. The study on transient conditions is important because the pipeline flows are normally in the unsteady state due to the sudden opening and closure of control valves, but most of the existing studies only analyse the flow in the steady-state conditions. The mathematical model consists in a set of non-linear conservation forms of partial differential equations. The objective of this paper is to improve the accuracy in the prediction of the HCNG transient flow parameters using the Reduced-Order Modelling (ROM). The ROM technique has been successfully used in single-gas and aerodynamic flow problems, the gas mixture has not been done using the ROM. The study is based on the velocity change created by the operation of the valves upstream and downstream the pipeline. Results on the flow characteristics, namely the pressure, density, celerity and mass flux are based on variations of the mixing ratio and valve reaction and actuation time; the ROM computational time cost advantage are also presented.
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 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.
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
Stress analysis of pressure vessels
International Nuclear Information System (INIS)
Kim, B.K.; Song, D.H.; Son, K.H.; Kim, K.S.; Park, K.B.; Song, H.K.; So, J.Y.
1979-01-01
This interim report contains the results of the effort to establish the stress report preparation capability under the research project ''Stress analysis of pressure vessels.'' 1978 was the first year in this effort to lay the foundation through the acquisition of SAP V structural analysis code and a graphic terminal system for improved efficiency of using such code. Software programming work was developed in pre- and post processing, such as graphic presentation of input FEM mesh geometry and output deformation or mode shope patterns, which was proven to be useful when using the FEM computer code. Also, a scheme to apply fracture mechanics concept was developed in fatigue analysis of pressure vessels. (author)
LOFT differential pressure uncertainty analysis
International Nuclear Information System (INIS)
Evans, R.P.; Biladeau, G.L.; Quinn, P.A.
1977-03-01
A performance analysis of the LOFT differential pressure (ΔP) measurement is presented. Along with completed descriptions of test programs and theoretical studies that have been conducted on the ΔP, specific sources of measurement uncertainty are identified, quantified, and combined to provide an assessment of the ability of this measurement to satisfy the SDD 1.4.1C (June 1975) requirement of measurement of differential pressure
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
International Nuclear Information System (INIS)
Chalhoub, E.S.
1980-09-01
A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author) [pt
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
Leak detection in pipelines through spectral analysis of pressure signals
Directory of Open Access Journals (Sweden)
Souza A.L.
2000-01-01
Full Text Available The development and test of a technique for leak detection in pipelines is presented. The technique is based on the spectral analysis of pressure signals measured in pipeline sections where the formation of stationary waves is favoured, allowing leakage detection during the start/stop of pumps. Experimental tests were performed in a 1250 m long pipeline for various operational conditions of the pipeline (liquid flow rate and leakage configuration. Pressure transients were obtained by four transducers connected to a PC computer. The obtained results show that the spectral analysis of pressure transients, together with the knowledge of reflection points provide a simple and efficient way of identifying leaks during the start/stop of pumps in pipelines.
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
Li, Weicong; Almeida, André; Smith, John; Wolfe, Joe
2016-02-01
Articulation, including initial and final note transients, is important to tasteful music performance. Clarinettists' tongue-reed contact, the time variation of the blowing pressure P¯mouth, the mouthpiece pressure, the pressure in the instrument bore, and the radiated sound were measured for normal articulation, accents, sforzando, staccato, and for minimal attack, i.e., notes started very softly. All attacks include a phase when the amplitude of the fundamental increases exponentially, with rates r ∼1000 dB s(-1) controlled by varying both the rate of increase in P¯mouth and the timing of tongue release during this increase. Accented and sforzando notes have shorter attacks (r∼1300 dB s(-1)) than normal notes. P¯mouth reaches a higher peak value for accented and sforzando notes, followed by a steady decrease for accented notes or a rapid fall to a lower, nearly steady value for sforzando notes. Staccato notes are usually terminated by tongue contact, producing an exponential decrease in sound pressure with rates similar to those calculated from the bandwidths of the bore resonances: ∼400 dB s(-1). In all other cases, notes are stopped by decreasing P¯mouth. Notes played with different dynamics are qualitatively similar, but louder notes have larger P¯mouth and larger r.
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.
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
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
International Nuclear Information System (INIS)
Balino, J.L.; Carrica, P.M.; Larreteguy, A.E.
1993-01-01
The pressure transient occurred at Atucha I Nuclear Power Plant in March 1990 is simulated. The transient was due to the fast closure of a flow control valve at the steam generators feedwater lines. The system was modelled, including the actuation of the relief valves. The minimum closure time for no actuation of the relief valves and the evolution of the velocity and piezo metric head for different cases were calculated. (author)
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)
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
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
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
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)
International Nuclear Information System (INIS)
Keller, Sandra; Rajasekaran, Priyadarshini; Bibinov, Nikita; Awakowicz, Peter
2012-01-01
The plasma parameters such as electron distribution function and electron density of three atmospheric-pressure transient discharges namely filamentary and homogeneous dielectric barrier discharges in air, and the spark discharge of an argon plasma coagulation (APC) system are determined. A combination of numerical simulation as well as diagnostic methods including current measurement and optical emission spectroscopy (OES) based on nitrogen emissions is used. The applied methods supplement each other and resolve problems, which arise when these methods are used individually. Nitrogen is used as a sensor gas and is admixed in low amount to argon for characterizing the APC discharge. Both direct and stepwise electron-impact excitation of nitrogen emissions are included in the plasma-chemical model applied for characterization of these transient discharges using OES where ambiguity arises in the determination of plasma parameters under specific discharge conditions. It is shown that the measured current solves this problem by providing additional information useful for the determination of discharge-specific plasma parameters. (paper)
Vent clearing analysis of a Mark III pressure suppression containment
International Nuclear Information System (INIS)
Quintana, R.
1979-01-01
An analysis of the vent clearing transient in a Mark III pressure suppression containment after a hypothetical LOCA is carried out. A two-dimensional numerical model solving the transient fluid dynamic equations is used. The geometry of the pressure suppression pool is represented and the pressure and velocity fields in the pool are obtained from the moment the LOCA occurs until the first vent in the drywell wall clears. The results are compared to those obtained with the one-diemensional model used for containment design, with special interest on two-dimensional effects. Some conclusions concerning the effect of the water discharged into the suppression pool through the vents on submerged structures are obtained. Future improvements to the model are suggested. (orig.)
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
International Nuclear Information System (INIS)
Joo, Jae Hwang; Kang, Ki Ju; Jhung, Myung Jo
2002-01-01
Performed here is an assessment study for deterministic fracture mechanics analysis of a pressurized thermal shock (PTS). The PTS event means an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. The problems consisting of two transients and 10 cracks are solved and maximum stress intensity factors and maximum allowable nil-ductility reference temperatures are calculated. Their results are compared each other to address the general characteristics between transients, crack types and analysis methods. The effects of elastic-plastic material behavior and clad coating on the inner surface are explored
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.
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
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
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.
Energy Technology Data Exchange (ETDEWEB)
Zhang, Fan, E-mail: zhangfan4060@gmail.com; Yuan, Shouqi; Fu, Qiang; Tao, Yi
2015-11-15
Highlights: • The transient flow characteristics of the charging pump with the first stage impeller in the HPSI process have been investigated numerically by CFD. • The hydraulic performance of the charging pump during the HPSI are discussed, andthe absolute errors between the simulated and measured results are analyzed in the paper. • Pressure fluctuation in the impeller and flow pattern in the impeller were studied in the HPSI process. It is influenced little at the beginning of the HPSI process while fluctuates strongly in the end of the HPSI process. - Abstract: In order to investigate the transient flow characteristics of the centrifugal charging pump during the transient transition process of high pressure safety injection (HPSI) from Q = 148 m{sup 3}/h to Q = 160 m{sup 3}/h, numerical simulation and experiment are implemented in this study. The transient flow rate, which is the most important factor, is obtained from the experiment and works as the boundary condition to accurately accomplish the numerical simulation in the transient process. Internal characteristics under the variable operating conditions are analyzed through the transient simulation. The results shows that the absolute error between the simulated and measured heads is less than 2.26% and the absolute error between the simulated and measured efficiency is less than 2.04%. Pressure fluctuation in the impeller is less influenced by variable flow rate in the HPSI process, while flow pattern in the impeller is getting better and better with the flow rate increasing. As flow rate increases, fluid blocks on the tongue of the volute and it strikes in this area at large flow rate. Correspondingly, the pressure fluctuation is intense and vortex occurs gradually during this period, which obviously lowers the efficiency of the pump. The contents of the current work can provide references for the design optimization and fluid control of the pump used in the transient process of variable operating
International Nuclear Information System (INIS)
Zhang, Fan; Yuan, Shouqi; Fu, Qiang; Tao, Yi
2015-01-01
Highlights: • The transient flow characteristics of the charging pump with the first stage impeller in the HPSI process have been investigated numerically by CFD. • The hydraulic performance of the charging pump during the HPSI are discussed, andthe absolute errors between the simulated and measured results are analyzed in the paper. • Pressure fluctuation in the impeller and flow pattern in the impeller were studied in the HPSI process. It is influenced little at the beginning of the HPSI process while fluctuates strongly in the end of the HPSI process. - Abstract: In order to investigate the transient flow characteristics of the centrifugal charging pump during the transient transition process of high pressure safety injection (HPSI) from Q = 148 m"3/h to Q = 160 m"3/h, numerical simulation and experiment are implemented in this study. The transient flow rate, which is the most important factor, is obtained from the experiment and works as the boundary condition to accurately accomplish the numerical simulation in the transient process. Internal characteristics under the variable operating conditions are analyzed through the transient simulation. The results shows that the absolute error between the simulated and measured heads is less than 2.26% and the absolute error between the simulated and measured efficiency is less than 2.04%. Pressure fluctuation in the impeller is less influenced by variable flow rate in the HPSI process, while flow pattern in the impeller is getting better and better with the flow rate increasing. As flow rate increases, fluid blocks on the tongue of the volute and it strikes in this area at large flow rate. Correspondingly, the pressure fluctuation is intense and vortex occurs gradually during this period, which obviously lowers the efficiency of the pump. The contents of the current work can provide references for the design optimization and fluid control of the pump used in the transient process of variable operating conditions.
Arterial Blood Pressure Induces Transient C4b-Binding Protein in Human Saphenous Vein Grafts.
Kupreishvili, Koba; Meischl, Christof; Vonk, Alexander B A; Stooker, Wim; Eijsman, Leon; Blom, Anna M; Quax, Paul H A; van Hinsbergh, Victor W M; Niessen, Hans W M; Krijnen, Paul A J
2017-05-01
Complement is an important mediator in arterial blood pressure-induced vein graft failure. Previously, we noted activation of cell protective mechanisms in human saphenous veins too. Here we have analyzed whether C4b-binding protein (C4bp), an endogenous complement inhibitor, is present in the vein wall. Human saphenous vein segments obtained from patients undergoing coronary artery bypass grafting (n = 55) were perfused in vitro at arterial blood pressure with either autologous blood for 1, 2, 4, or 6 hr or with autologous blood supplemented with reactive oxygen species scavenger N-acetylcysteine. The segments were subsequently analyzed quantitatively for presence of C4bp and complement activation product C3d using immunohistochemistry. Perfusion induced deposition of C3d and C4bp within the media of the vessel wall, which increased reproducibly and significantly over a period of 4 hr up to 3.8% for C3d and 81% for C4bp of the total vessel area. Remarkably after 6 hr of perfusion, the C3d-positive area decreased significantly to 1.3% and the C4bp-positive area to 19% of the total area of the vein. The areas positive for both C4bp and C3d were increased in the presence of N-acetylcysteine. Exposure to arterial blood pressure leads to a transient presence of C4bp in the vein wall. This may be part of a cell-protective mechanism to counteract arterial blood pressure-induced cellular stress and inflammation in grafted veins. Copyright © 2017 Elsevier Inc. All rights reserved.
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
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
SOCOOL-2, Molten Materials Na Coolant Interaction, Temperature and Pressure Transient
International Nuclear Information System (INIS)
Padilla, A. Jr.
1973-01-01
1 - Description of problem or function: SOCOOL2 calculates the transient temperatures, pressures, and mechanical work energy when a molten material is instantaneously and uniformly dispersed in liquid sodium which is initially under acoustic constraint. 2 - Method of solution: A unit cell consisting of a single spherical particle of molten material surrounded concentrically by sodium is used as the basis for the calculation. Heat transfer from the molten particle to the sodium is calculated by an implicit numerical technique assuming negligible contact resistance at the interface of the particle. The expansion of the heated sodium is calculated by the one-dimensional acoustic equation until vaporization conditions are attained. Upon vaporization, it is assumed that the particle becomes vapor-blanketed and that no further heat transfer to or from the sodium occurs. The heated sodium is then expanded to the specific final pressure in an isentropic expansion process. 3 - Restrictions on the complexity of the problem: The presence of an initial amount of sodium vapor or noncondensable gas cannot be taken into account. Time delays in the process of fragmentation and mixing of the molten material into the sodium cannot be considered. Heat transfer during the two-phase expansion of sodium is neglected
Polynomial analysis of ambulatory blood pressure measurements
Zwinderman, A. H.; Cleophas, T. A.; Cleophas, T. J.; van der Wall, E. E.
2001-01-01
In normotensive subjects blood pressures follow a circadian rhythm. A circadian rhythm in hypertensive patients is less well established, and may be clinically important, particularly with rigorous treatments of daytime blood pressures. Polynomial analysis of ambulatory blood pressure monitoring
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
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.
Energy Technology Data Exchange (ETDEWEB)
Tare, U. A.; Mody, F. K.; Mese, A. I. [Haliburton Energy Services, TX (United States)
2002-07-01
In order to develop a real-time wellbore (in)stability modelling capability, experimental work was carried out to investigate the role of the chemical potential of drilling fluids on transient pore pressure and time-dependent rock property alterations of shale formations. Time-dependent alterations in the pore pressure, acoustic and rock properties of formations subjected to compressive tri-axial test were recorded during the experiments involving the Pore Pressure Transmission (PPT) test. Based on the transient pore pressure of shale exposed to the test fluid presented here, the 20 per cent calcium chloride showed a very low membrane efficiency of 4.45 per cent. The need for a thorough understanding of the drilling fluid/shale interaction prior to applying any chemical potential wellbore (in)stability model to real-time drilling operations was emphasized. 9 refs., 5 figs.
Pressure relieving support surfaces (PRESSURE) trial: cost effectiveness analysis.
Iglesias, Cynthia; Nixon, Jane; Cranny, Gillian; Nelson, E Andrea; Hawkins, Kim; Phillips, Angela; Torgerson, David; Mason, Su; Cullum, Nicky
2006-06-17
To assess the cost effectiveness of alternating pressure mattresses compared with alternating pressure overlays for the prevention of pressure ulcers in patients admitted to hospital. Cost effectiveness analysis carried out alongside the pressure relieving support surfaces (PRESSURE) trial; a multicentre UK based pragmatic randomised controlled trial. 11 hospitals in six UK NHS trusts. Intention to treat population comprising 1971 participants. Kaplan Meier estimates of restricted mean time to development of pressure ulcers and total costs for treatment in hospital. Alternating pressure mattresses were associated with lower overall costs (283.6 pounds sterling per patient on average, 95% confidence interval--377.59 pounds sterling to 976.79 pounds sterling) mainly due to reduced length of stay in hospital, and greater benefits (a delay in time to ulceration of 10.64 days on average,--24.40 to 3.09). The differences in health benefits and total costs for hospital stay between alternating pressure mattresses and alternating pressure overlays were not statistically significant; however, a cost effectiveness acceptability curve indicated that on average alternating pressure mattresses compared with alternating pressure overlays were associated with an 80% probability of being cost saving. Alternating pressure mattresses for the prevention of pressure ulcers are more likely to be cost effective and are more acceptable to patients than alternating pressure overlays.
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.
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
International Nuclear Information System (INIS)
Lockwood, M.
1991-01-01
The suggestion is discussed that characteristic particle and field signatures at the dayside magnetopause, termed flux transfer events, are, in at least some cases, due to transient solar wind and/or magnetosheath dynamic pressure increases, rather than time-dependent magnetic reconnection. It is found that most individual cases of FTEs observed by a single spacecraft can, at least qualitatively, be explained by the pressure pulse model, provided a few rather unsatisfactory features of the predictions are explained in terms of measurement uncertainties. The most notable exceptions to this are some two-regime observations made by two satellites simultaneously, one on either side of the magnetopause. However, this configuration has not been frequently achieved for sufficient time, such observations are rare, and the relevant tests are still not conclusive. The strongest evidence that FTEs are produced by magnetic reconnection is the dependence of their occurence on the north-south component of the interplanetary magnetic field (IMF) or of the magnetosheath field. The pressure pulse model provides an explanation for this dependence in the case of magnetosheath FTEs, but does not apply to magnetosphere FTEs. The only surveys of magnetosphere FTEs have not employed the simultaneous IMF, but have shown that their occurence is strongly dependent on the north-south component of the magnetosheath field, as observed earlier/later on the same magnetopause crossing. This paper employs statistics on the variability of the IMF orientation to investigate the effects of IMF changes between the times of the magnetosheath and FTE observations. It is shown that the previously published results are consistent with magnetospheric FTEs being entirely absent when the magentosheath field is northward
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.
International Nuclear Information System (INIS)
Aronne, Ivan Dionysio
2009-01-01
The demand for energy in the modern world is growing, particularly in the developing countries. The nuclear options has been deserving prominence for their qualities of not impacting the environment through emissions of greenhouse gases and nor to demand great areas.. However society requests improvement in the safety of new reactors and the utilities request larger availability of the power plants. The IRIS project of an integral nuclear pressurized water reactor proposes to fulfill those requirements. A system for identification and classification of transients would help to improve the safety and to increase the availability of the IRIS increasing its competitiveness. In order to contribute to the development o such a system it was developed in this work a System for Identification and Classification of Transients - SDICT - capable of monitoring the operation of the reactor and of providing information on its operational state. SICT was developed using the technique of neural networks, more specifically the Self-Organizing Maps. Results of IRIS simulation with RELAP5 code were used to train the neural network of SICT. To demonstrate the correctness of the methodology of using simulations results, whose values have characteristics different from the measured ones, it was made a version of SICT for an experimental installation, the The Circuit no. 1 - CT1. Experiments were run in this test facility and simulations of its operation were done with RELAP5. This CT1 version of SICT was then checked against the simulation and experimental data validating the methodology. As a result of the activities to develop SICT, it is now available for futures studies: the developed application, SICT, a database of experiments in CT1, a validate nodalization of CT1, a database of results of CT1 simulations , a nodalization of the IRIS tested for several normal and abnormal transients and a database with the results of IRIS simulations. Attached to this thesis is a CD with the
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.
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
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.
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
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)
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.
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
Analysis of stress in reactor core vessel under effect of pressure lose shock wave
International Nuclear Information System (INIS)
Li Yong; Liu Baoting
2001-01-01
High Temperature gas cooled Reactor (HTR-10) is a modular High Temperature gas cooled Reactor of the new generation. In order to analyze the safety characteristics of its core vessel in case of large rupture accident, the transient performance of its core vessel under the effect of pressure lose shock wave is studied, and the transient pressure difference between the two sides of the core vessel and the transient stresses in the core vessel is presented in this paper, these results can be used in the safety analysis and safety design of the core vessel of HTR-10. (author)
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
PREST, Pressure Temperature Transients, I Inhalation in Containment Building from LOCA
Energy Technology Data Exchange (ETDEWEB)
Gaggero, G [CETIS, EURATOM C.C.R., 21020 - Ispra - Varese (Italy); Gerini, P M [CISE, Segrate, Milano (Italy); Leoni, G [AGIP Nucleare, San Donato Milanese - Milano (Italy); Van Erp, J B [EURATOM C.C.R., 21020 - Ispra - Varese (Italy)
1969-06-01
1 - Nature of physical problem solved: The programme is intended for the determination of pressure and temperature transient inside the containment building, following a loss-of-coolant accident due to a rupture in the primary cooling system of a nuclear power plant having water as the primary coolant. The model includes the calculation of the radiation doses incurred to the thyroid due to inhalation of radioactive iodine released outside the containment building. 2 - Method of solution: The energy equation is solved at each time step by using the Newton method. In order to determine the heat exchange with structures inside the containment building as well as with the outside atmosphere, the structures are treated in slab geometry. The resulting Fourier equations for heat conduction are solved numerically by using an implicit form to avoid stability problems. 3 - Restrictions on the complexity of the problem: max. number of internal slabs - 6; max. number of external slabs - 4; max. number of meshes in each slab - 100.
Simulation of corrosion product activity in pressurized water reactors under flow rate transients
International Nuclear Information System (INIS)
Mirza, Anwar M.; Mirza, Nasir M.; Mir, Imran
1998-01-01
Simulation of coolant activation due to corrosion products and impurities in a typical pressurized water reactor has been done under flow rate transients. Employing time dependent production and losses of corrosion products in the primary coolant path an approach has been developed to calculate the coolant specific activity. Results for 24 Na, 56 Mn, 59 Fe, 60 Co and 99Mo show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. Predominant corrosion product activity is due to Mn-56. Parametric studies at full power for various ramp decreases in flow rate show initial decline in the activity and then a gradual rise to relatively higher saturation values. The minimum value and the time taken to reach the minima are strong functions of the slope of linear decrease in flow rate. In the second part flow rate coastdown was allowed to occur at different flow half-times. The reactor scram was initiated at 90% of the normal flow rate. The results show that the specific activity decreases and the rate of decrease depends on pump half time and the reactor scram conditions
International Nuclear Information System (INIS)
Tennankore, K.N.; Kumar, R.K.; Razzaghi, M.
1987-01-01
One-dimensional flame models are often used to predict the pressure transients caused by hydrogen combustion in containments during postulated severe accidents. In the absence of data, these models account for prevailing flame acceleration mechanisms, such as initial turbulence, venting and obstacle-induced turbulence, by using arbitrarily large burning velocities that are much higher than laminar burning velocities. Using an intermediate-scale test facility at the Whiteshell Nuclear Research Establishment we have obtained necessary data on the effects of flame acceleration mechanisms, to estimate the safety margin in the buring velocities used in the models. So far, data have been analyzed, with a one-dimensional model, to determine effective burning velocities and burning-rate enhancement factors. The results of the analyses indicate that the effect of initial turbulence on the burning rate can be bounded only if the effect of flame-generated turbulence is included. The effect of venting can be accounted for by using two burning velocities, one for the pre-vent duration and a second increased value during the vented-combustion stage. The enhancement factors due to these two mechanisms, for the different conditions analyzed, varied up to 5.4, and the effective burning velocities varied up to 8.4 m/s
Stover, E. K.; York, T. M.
1971-01-01
The transient pinched plasma column generated in a linear Z-pinch was studied experimentally and analytically. The plasma column was investigated experimentally with several plasma diagnostics; they were: a rapid response pressure transducer, a magnetic field probe, a voltage probe, and discharge luminosity. Axial pressure profiles on the discharge chamber axis were used to identify three characteristic regions of plasma column behavior: (1) strong axial pressure asymmetry noted early in plasma column lifetime, (2) followed by plasma heating in which there is a rapid rise in static pressure, and (3) a slight decrease static pressure before plasma column breakup. Plasma column lifetime was approximately 5 microseconds. The axial pressure asymmetry was attributed to nonsimultaneous pinching of the imploding current sheet along the discharge chamber axis. The rapid heating could be attributed in part to viscous effects introduced by radial gradients in the axial streaming velocity.
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
International Nuclear Information System (INIS)
Reventos, F.; Baptista, J.S.; Navas, A.P.; Moreno, P.
1993-12-01
The Asociacion Nuclear Asco has prepared a model of Asco NPP using RELAP5/MOD2. This model, which include thermalhydraulics, kinetics and protection and controls, has been qualified in previous calculations of several actual plant transients. One of the transients of the qualification process is a ''Pressurizer spray valve faulty opening'' presented in this report. It consists in a primary coolant depressurization that causes the reactor trip by overtemperature and later on the actuation of the safety injection. The results are in close agreement with plant data
Comparison of transient PCRV model test results with analysis
International Nuclear Information System (INIS)
Marchertas, A.H.; Belytschko, T.B.
1979-01-01
Comparisons are made of transient data derived from simple models of a reactor containment vessel with analytical solutions. This effort is a part of the ongoing process of development and testing of the DYNAPCON computer code. The test results used in these comparisons were obtained from scaled models of the British sodium cooled fast breeder program. The test structure is a scaled model of a cylindrically shaped reactor containment vessel made of concrete. This concrete vessel is prestressed axially by holddown bolts spanning the top and bottom slabs along the cylindrical walls, and is also prestressed circumferentially by a number of cables wrapped around the vessel. For test purposes this containment vessel is partially filled with water, which comes in direct contact with the vessel walls. The explosive charge is immersed in the pool of water and is centrally suspended from the top of the vessel. The 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
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
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
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.
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)
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
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.).
International Nuclear Information System (INIS)
Kunze, J.F.; Loyalka, S.K.; Hultsch, R.A.; Oladiran, O.; McKibben, J.C.
1990-01-01
This paper reports on benchmark experiments needed to verify the accuracy of thermal hydraulic codes (such as RELAP5/MOD2) with respect to their capability to simulate transient boiling conditions both with and without a closed recirculation path in narrow channels, under essentially atmospheric pressure conditions characteristic of plate-type research reactors. An experimental apparatus with this objective has been constructed, and data for surface heat flux of 1.2 x 10 5 w/m 2 are reported
Nishino, Kazuaki; Yoshida, Fujiko; Nitta, Akari; Saito, Mieko; Saito, Kazuuchi
2013-12-01
To evaluate retrospectively seasonal fluctuations of transient intraocular pressure (IOP) elevation after automated visual field examination in patients with primary open-angle glaucoma (POAG). We reviewed 53 consecutive patients with POAG who visited Kaimeido Ophthalmic and Dental Clinic from January 2011 to March 2013, 21 men and 32 women aged 67.7 +/- 11.2 years. The patients were divided into 4 groups, spring, summer, autumn, and winter according to the month of automated visual field examination and both eyes of each patient were enrolled. IOP was measured immediately after automated visual field examination (vf IOP) and compared with the average IOP from the previous 3 months (pre IOP) and with the average IOP from the following 3 months (post IOP) in each season. IOP elevation rate was defined as (vf IOP- pre IOP)/pre IOP x 100% and calculated for each season (paired t test). Additionally, the correlation between mean deviation (MD) and IOP elevation rate was evaluated (single regression analysis). Exclusion criteria were patients who received cataract surgery during this study or had a history of any previous glaucoma surgery. The automated visual field test was performed with a Humphrey field analyzer and the Central 30-2 FASTPAC threshold program. The average vf IOP was 14.5 +/- 2.5 mmHg, higher than pre IOP 13.8 +/- 2.4 mmHg (p field examination, especially in the winter but not in the summer.
International Nuclear Information System (INIS)
Ball, D.G.; Drake, J.B.; Cheverton, R.D.; Iskander, S.K.
1984-02-01
The OCA-II computer code, like its predecessor OCA-I, performs the thermal, stress, and linear elastic fracture-mechanics analysis for long flaws on the surface of a cylinder that is subjected to thermal and pressure transients. OCA-II represents a revised and expanded version of OCA-I and includes as new features (1) cladding as a discrete region, (2) a finite-element subroutine for calculating the stresses, and (3) the ability to calculate stress intensity factors for certain three-dimensional flaws, for two-dimensional circumferential flaws on the inner surface, and for both axial and circumferential flaws on the outer surface. OCA-I considered only inner-surface flaws. An option is included in OCA-II that permits a search for critical values of fluence or nil-ductility reference temperature corresponding to a specified failure criterion. These and other features of OCA-II are described in the report, which also includes user instructions for the code
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)
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.
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...
Pressure thermal shock analysis for nuclear reactor pressure vessel
International Nuclear Information System (INIS)
Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.
2015-01-01
The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)
Pressure rise analysis in superconducting coils during dumping
International Nuclear Information System (INIS)
Tada, E.; Shimamoto, S.
1984-01-01
This chapter describes the ALPHE computer code, whose purpose is to calculate transient helium behavior in a poolboiling coil and to determine suitable characteristics of safety devices to minimize the maximum pressure and the liquid helium lost during dumping due to quench, or when discharging without normalcy. The analysis is compared with the measurements obtained in the domestic test of the Japanese LCT coil. Topics considered include basic equations (helium behavior, heat generation), manual dump without quench, and dumping due to quench. It is demonstrated that the transient behavior, calculated by ALPHE assuming quasi-static equilibrium between helium and coil, is in good agreement with the experimental measurements observed in the domestic test of the Japanese LCT coil. The engineering technique required for the design criteria of superconducting coils and safety device during dumping is established. ALPHE can be used to design an emergency safety system for a helium refrigerator during dumping
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.
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
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)
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....
Vapor Pressure Data Analysis and Statistics
2016-12-01
near 8, 2000, and 200, respectively. The A (or a) value is directly related to vapor pressure and will be greater for high vapor pressure materials...1, (10) where n is the number of data points, Yi is the natural logarithm of the i th experimental vapor pressure value, and Xi is the...VAPOR PRESSURE DATA ANALYSIS AND STATISTICS ECBC-TR-1422 Ann Brozena RESEARCH AND TECHNOLOGY DIRECTORATE
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
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...
International Nuclear Information System (INIS)
Konovalyuk, L.N.; Shevelev, D.V.; Kravchenko, V.G.
2003-01-01
PRZ model is proposed which allows taking into account in pressurizer convective heat- and mass transfer influence effects at the transients in VVER (PWR) Type Reactors case when calculations performed with using 1D thermohydraulic codes. The theoretical backgrounds are given to define the transients with the convective coolant instability in PRZ. The instability threshold is given for real PRZ geometry
Pressurized Thermal Shock Analysis for OPR1000 Pressure Vessel
Energy Technology Data Exchange (ETDEWEB)
Bhowmik, P. K.; Shamim, J. A.; Gairola, A.; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)
2014-10-15
The study provides a brief understanding of the analysis procedure and techniques using ANSYS, such as the acceptance criteria, selection and categorization of events, thermal analysis, structural analysis including fracture mechanics assessment, crack propagation and evaluation of material properties. PTS may result from instrumentation and control malfunction, inadvertent steam dump, and postulated accidents such as smallbreak (SB) LOCA, large-break (LB) LOCA, main steam line break (MSLB), feedwater line breaks and steam generator overfill. In this study our main focus is to consider only the LB LOCA due to a cold leg break of the Optimized Power Reactor 1000 MWe (OPR1000). Consideration is given as well to the emergency core cooling system (ECCS) specific sequence with the operating parameters like pressure, temperature and time sequences. The static structural and thermal analysis to investigate the effects of PTS on RPV is the main motivation of this study. Specific surface crack effects and its propagation is also considered to measure the integrity of the RPV. This study describes the procedure for pressurized thermal shock analysis due to a loss of coolant accidental condition and emergency core cooling system operation for reactor pressure vessel.. Different accidental events that cause pressurized thermal shock to nuclear RPV that can also be analyzed in the same way. Considering the limitations of low speed computer only the static analysis is conducted. The modified LBLOCA phases and simplified geometry can is utilized to analyze the effect of PTS on RPV for general understanding not for specific specialized purpose. However, by integrating the disciplines of thermal and structural analysis, and fracture mechanics analysis a clearer understanding of the total aspect of the PTS problem has resulted. By adopting the CFD, thermal hydraulics, uncertainties and risk analysis for different type of accidental conditions, events and sequences with proper
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
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
International Nuclear Information System (INIS)
Dickson, T.L.; Cheverton, R.D.; Bryson, J.W.; Bass, B.R.; Shum, D.K.M.; Keeney, J.A.
1993-08-01
The Nuclear Regulatory Commission (NRC) requested Oak Ridge National Laboratory (ORNL) to perform a pressurized-thermal-shock (PTS) probabilistic fracture mechanics (PFM) sensitivity analysis for the Yankee Rowe reactor pressure vessel, for the fluences corresponding to the end of operating cycle 22, using a specific small-break-loss- of-coolant transient as the loading condition. Regions of the vessel with distinguishing features were to be treated individually -- upper axial weld, lower axial weld, circumferential weld, upper plate spot welds, upper plate regions between the spot welds, lower plate spot welds, and the lower plate regions between the spot welds. The fracture analysis methods used in the analysis of through-clad surface flaws were those contained in the established OCA-P computer code, which was developed during the Integrated Pressurized Thermal Shock (IPTS) Program. The NRC request specified that the OCA-P code be enhanced for this study to also calculate the conditional probabilities of failure for subclad flaws and embedded flaws. The results of this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative influence of a number of key input parameters in the Yankee Rowe PTS analysis and (2) data that can be used for readily determining the probability of vessel failure once a more accurate indication of vessel embrittlement becomes available. This report is designated as HSST report No. 117
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)
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
International Nuclear Information System (INIS)
Zhou, J X; Hu, M; Cai, F L; Huang, X T
2014-01-01
For a hydropower station with longer water conveyance system, an optimum turbine's selection will be beneficial to its reliable and stable operation. Different optional turbines will result in possible differences of the hydraulic characteristics in the hydromechanical system, and have different effects on the hydraulic transients' analysis and control. Therefore, the premise for turbine's selection is to fully understand the properties of the optional turbines and their effects on the hydraulic transients. After a brief introduction of the simulation models for hydraulic transients' computation and stability analysis, the effects of hydraulic turbine's characteristics at different operating points on the hydro-mechanical system's free vibration analysis were theoretically investigated with the hydraulic impedance analysis of the hydraulic turbine. For a hydropower station with long water conveyance system, based on the detailed hydraulic transients' computation respectively for two different optional turbines, the effects of the turbine's selection on hydraulic transients were analyzed. Furthermore, considering different operating conditions for each turbine and the similar operating conditions for these two turbines, free vibration analysis was comprehensively carried out to reveal the effects of turbine's impedance on system's vibration characteristics. The results indicate that, respectively with two different turbines, most of the controlling parameters under the worst cases have marginal difference, and few shows obvious differences; the turbine's impedances under different operating conditions have less effect on the natural angular frequencies; different turbine's characteristics and different operating points have obvious effects on system's vibration stability; for the similar operating conditions of these two turbines, system's vibration characteristics are basically consistent with
History matching of transient pressure build-up in a simulation model using adjoint method
Energy Technology Data Exchange (ETDEWEB)
Ajala, I.; Haekal, Rachmat; Ganzer, L. [Technische Univ. Clausthal, Clausthal-Zellerfeld (Germany); Almuallim, H. [Firmsoft Technologies, Inc., Calgary, AB (Canada); Schulze-Riegert, R. [SPT Group GmbH, Hamburg (Germany)
2013-08-01
The aim of this work is the efficient and computer-assisted history-matching of pressure build-up and pressure derivatives by small modification to reservoir rock properties on a grid by grid level. (orig.)
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.
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.
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
Liu, Y.; Rice, J. R.
2005-12-01
-equilibrate with that of its surroundings). This is consistent with our previous simulations, which show that the aseismic transients migrate along the strike at a higher speed under a lower, constant in time, effective normal stress. As a combination of the two factors, we show the pore pressure evolution with drops (due to dilatancy during slip) and then rises (due to shear heating) on the fault over multiple time scales. We next plan to formulate, and merge with the slip-rupture analysis, fuller fluid release models based on phase equilibria and models of transport in which the average fault-parallel permeability is a decreasing function of the effective normal stress. The thrust fault zone, at seismogenic depths and slightly downdip, is represented in a conceptually similar manner to the well-studied major continental faults, assuming the fault core materials have a lower permeability than the neighboring damaged zone. Heat diffusion in the fault core and damaged zone will also be considered in the modeling. The simulation results may help to improve our understanding of the processes of the aseismic transients observed within a transform plate boundary along the SAF near Cholame, California [Nadeau and Dolenc, 2005].
An investigation of transient pressure and plasma properties in a pinched plasma column. M.S. Thesis
Stover, E. K.; York, T. M.
1971-01-01
The transient pinched plasma column generated in a linear Z-pinch was studied experimentally and analytically. The plasma column was investigated experimentally with the following plasma diagnostics: a special rapid response pressure transducer, a magnetic field probe, a voltage probe and discharge luminosity. Axial pressure profiles on the discharge chamber axis were used to identify three characteristic regions of plasma column behavior; they were in temporal sequence: strong axial pressure asymmetry noted early in plasma column lifetime followed by plasma heating in which there is a rapid rise in static pressure and a slight decrease static pressure before plasma column breakup. Plasma column lifetime was approximately 5 microseconds. The axial pressure asymmetry was attributed to nonsimultaneous pinching of the imploding current sheet along the discharge chamber axis. The rapid heating is attributed in part to viscous effects introduced by radial gradients in the axial streaming velocity. Turbulent heating arising from discharge current excitation of the ion acoustic wave instability is also considered a possible heating mechanism.
International Nuclear Information System (INIS)
Forster, C.B.; Gale, J.E.
1981-06-01
A field experiment to evaluate the transient pressure pulse technique as a method of determining the in-situ hydraulic conductivity of low permeability fractured rock was made. The experiment attempted to define: the radius of influence of a pressure pulse-test in fractured rock and the correlation between pressure-pulse tests and steady-state flow tests performed in five boreholes drilled in fractured granite. Twenty-five test intervals, 2 to 3 m in length, were isolated in the boreholes, using air-inflated packers. During pressure pulse and steady-state tests, pressures were monitored in both the test and observation cavities. Rock-mass conductivities were calculated from steady-state test results and were found to range from less than 10 - 11 to 10 - 7 cm/sec. However, there was no consistent correlation between the steady-state conductivity and the pressure pulse decay characteristics of individual intervals. These conflicting test results can be attributed to the following factors: differences in volumes of rock affected by the test techniques; effects of equipment configuration and compliance; and complexity of the fracture network. Although the steady-state flow tests indicate that hydraulic connections exist between most of the test cavities, no pressure responses were noted in the observation cavities (located at least 0.3 m from the test cavities) during the pulse tests. This does not mean, however, that the pressure-pulse radius of influence is <0.3 m, because the observation cavities were too large (about 7 liters). The lack of correlation between steady-state conductivities and the corresponding pressure pulse decay times does not permit use of existing single-fracture type curves to analyze pulse tests performed in multiple-fracture intervals. Subsequent work should focus on the detailed interpretation of field results with particular reference to the effects of the fracture system at the test site
Performance Analysis of Waste Heat Driven Pressurized Adsorption Chiller
LOH, Wai Soong; SAHA, Bidyut Baran; CHAKRABORTY, Anutosh; NG, Kim Choon; CHUN, Won Gee
2010-01-01
This article presents the transient modeling and performance of waste heat driven pressurized adsorption chillers for refrigeration at subzero applications. This innovative adsorption chiller employs pitch-based activated carbon of type Maxsorb III
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.
SOLA-LOOP analysis of a back pressure check valve
International Nuclear Information System (INIS)
Travis, J.R.
1984-01-01
The SOLA-LOOP computer code for transient, nonequilibrium, two-phase flows in networks has been coupled with a simple valve model to analyze a feedwater pipe breakage with a back-pressure check valve. Three tests from the Superheated Steam Reactor Safety Program Project (PHDR) at Kahl, West Germany, are analyzed, and the calculated transient back-pressure check valve behavior and fluid dynamics effects are found to be in excellent agreement with the experimentally measured data
Energy Technology Data Exchange (ETDEWEB)
Rives Sanz, R.; Montesino Otero, M.E.; Gonzalez Mantecon, J.; Rojas Mazaira, L., E-mail: mmontesi@instec.cu [Higher Institute of Technology and Applied Science, La Habana (Cuba). Department of Nuclear Engineering; Lira, C.A. Brayner de Oliveira [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)
2014-07-01
International Reactor Innovative and Secure (IRIS) excels other Small Modular Reactor (SMR) designs due to its innovative characteristics regarding safety. IRIS integral pressurizer makes the design of larger pressurizer system than the conventional PWR, without any additional cost. The IRIS pressurizer volume of steam can provide enough margins to avoid spray requirement to mitigate in-surge transient. The aim of the present research is to model the IRIS pressurizer's dynamic using the commercial finite volume Computational Fluid Dynamic code CFX 14. A symmetric tridimensional model equivalent to 1/8 of the total geometry was adopted to reduce mesh size and minimize processing time. The model considers the coexistence of three phases: liquid, steam, and vapor bubbles in liquid volume. Additionally, it takes into account the heat losses between the pressurizer and primary circuit. The relationships for interfacial mass, energy, and momentum transport are programmed and incorporated into CFX by using expressions in CFX Command Language (CCL) format. Moreover, several additional variables are defined for improving the convergence and allow monitoring of boron dilution sequences and condensation-evaporation rate in different control volumes. For transient states a non - equilibrium stratification in the pressurizer is considered. This paper discusses the model developed and the behavior of the system for representative transients sequences such as the in/out-surge transients and boron dilution sequences. The results of analyzed transients of IRIS can be applied to the design of pressurizer internal structures and components. (author)
Sun, Hao; Li, De-Pei; Chen, Shao-Rui; Hittelman, Walter N.; Pan, Hui-Lin
2009-01-01
The arterial baroreceptor is critically involved in the autonomic regulation of homoeostasis. The transient receptor potential vanilloid 1 (TRPV1) receptor is expressed on both somatic and visceral sensory neurons. Here, we examined the TRPV1 innervation of baroreceptive pathways and its functional significance in the baroreflex. Resiniferatoxin (RTX), an ultrapotent analog of capsaicin, was used to ablate TRPV1-expressing afferent neurons and fibers in adult rats. Immunofluorescence labeling...
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)
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
Rod-bundle transient-film boiling of high-pressure water in the liquid-deficient regime
International Nuclear Information System (INIS)
Morris, D.G.; Mullins, C.B.; Yoder, G.L.
1982-01-01
Results are reported from a recent experiment investigating dispersed flow film boiling of high pressure water in upflow through a rod bundle. The data, obtained under mildly transient conditions, are used to assess correlations currently used to predict heat transfer in these circumstances. In light of the scarcity of similar data, the data should prove useful in the development and assessment of new heat transfer models. The experiment was conducted at the Oak Ridge National Laboratory in the Thermal-Hydraulic Test Facility, a highly instrumented, non-nuclear, pressurized-water loop containing 64, 3.66-m (12-ft) long rods (of which 60 are electrically heated). The rods are arranged in a square array typical of 17 x 17 fuel rod assemblies in late generation PWRs. Data were collected over typical reactor blowdown parameter ranges
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.
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.
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.
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...
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
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)
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.
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
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
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.
Energy Technology Data Exchange (ETDEWEB)
Tare, U.A.; Mody, F.K.; Mese, A.I. [Halliburton Energy Services, Cairo (Egypt)
2000-11-01
Experimental studies were conducted to explain the concept of a real-time wellbore (in)stability logging methodology. The role of the chemical potential of drilling fluids on transient pore pressure and time-dependent rock property alterations of shale formations was examined by providing details about a pore pressure transmission (PPT) test. The PPT experiments exposed formation (shale) cores under simulated downhole conditions to various salt solutions and drilling fluids. The main objective was to translate the results of the PPT tests to actual drilling conditions. A 20 per cent w/w calcium chloride solution was exposed to a Pierre II shale under high pressure in the PPT apparatus. The PPT test was used to estimate the impact of a drilling fluid on shale pore pressure. The efficiency of the salt solution/shale system was also estimated. Estimates of the dynamic rock properties were made based on the obtained acoustic data. It was determined that in order to accurately model time-dependent wellbore (in)stability in the field, it is important to calibrate representative shale core response to drilling fluids under realistic in-situ conditions. The 20 per cent w/w calcium chloride solution showed very low membrane efficiency of 4.45 per cent. It was concluded that changes in the shale dynamic rock properties as a function of test fluid exposure can be obtained from the simultaneous acquisition of sonic compression and shear wave velocity data. 12 refs., 5 figs.
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
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.
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
Dynamic analysis of solid propellant grains subjected to ignition pressurization loading
Chyuan, Shiang-Woei
2003-11-01
Traditionally, the transient analysis of solid propellant grains subjected to ignition pressurization loading was not considered, and quasi-elastic-static analysis was widely adopted for structural integrity because the analytical task gets simplified. But it does not mean that the dynamic effect is not useful and could be neglected arbitrarily, and this effect usually plays a very important role for some critical design. In order to simulate the dynamic response for solid rocket motor, a transient finite element model, accompanied by concepts of time-temperature shift principle, reduced integration and thermorheologically simple material assumption, was used. For studying the dynamic response, diverse ignition pressurization loading cases were used and investigated in the present paper. Results show that the dynamic effect is important for structural integrity of solid propellant grains under ignition pressurization loading. Comparing the effective stress of transient analysis and of quasi-elastic-static analysis, one can see that there is an obvious difference between them because of the dynamic effect. From the work of quasi-elastic-static and transient analyses, the dynamic analysis highlighted several areas of interest and a more accurate and reasonable result could be obtained for the engineer.
Measurement and analysis of transient vaporization in oxide fuel materials
International Nuclear Information System (INIS)
Benson, D.A.; Bergeron, E.G.
1979-01-01
This paper describes a series of experiments in which samples are heated to produce high vapor pressure states in times of 10 -6 to 10 -3 seconds. Experimental measurements of vapor pressures over fresh UO 2 from the pulsed electron beam and pulsed reactor heating tests are presented and compared with other high temperature data. The interpretation of the vapor pressure measured in the tests is discussed in detail. Effects of original sample stoichiometry, chemical interactions with the container and non-equilibrium evaporation due to induced temperature gradients are discussed. Special attention is given to dynamic behavior in rapid heating and vaporization of the oxide due to chemical non-equilibrium. Finally, similar projected reactor experiments on irradiated fuel are described and vapor pressure predictions made using available equilibrium models. A discussion of information accessible from such future tests and its importance is presented. (orig.) [de
Measurement and analysis of transient vaporization in oxide fuel materials
International Nuclear Information System (INIS)
Gorham-Bergeron, E.; Benson, D.A.
1978-01-01
A series of experiments is described in which samples are heated to produce high vapor pressure states in times of 10 -6 to 10 -3 seconds. Experimental measurements of vapor pressures over fresh UO 2 from the pulsed electron beam and pulsed reactor heating tests are presented and compared with other high temperature data. The interpretation of the vapor pressures measured in the tests is discussed in detail. Effects of original sample stoichiometry, chemical interactions with the container and non-equilibrium evaporation due to induced temperature gradients are discussed. Special attention is given to dynamic behavior in rapid heating and vaporization of the oxide due to chemical nonequilibrium. Finally, similar projected reactor experiments on irradiated fuel are described and vapor pressure predictions made using available equilibrium models. A discussion of information accessible from such future tests and its importance is presented
Possibilities of optimizing non-nuclear simulation of pressurized water reactor transients
International Nuclear Information System (INIS)
Silva Filho, E.
1985-01-01
The GKSS-Forschungszentrum Geesthacht GmbH has instituted the concept of a scaled test facility (volume scale factor of 1/100) of a typical PWR of the 1 300 MWe class for the purpose of studying small breaks Loss-of-Coolant Accidents (LOCA) and transients. Having in mind the goal of an optimization of this concept has been choosen a station blackout with and without reactor shutdown and a small break LOCA in a primary loop piping to investigate the thermohydraulic behaviour of the test facility in comparison to the reactor plant. The computer code RELAP 5/MOD 1 has been utilized to compare the test facility behaviour with the reactor plant one. Recommendations are given for minimization of distortions between test facility and reactor plant. (orig./HP) [de
International Nuclear Information System (INIS)
Ho, J.C.
2004-01-01
Among those theories to interpret the PWR iodine spiking behaviors, the most accepted concept is based on steam formation and condensation in damaged fuel rods. Due to the complex nature of the phenomenon, a comprehensive model of the iodine behavior has not yet been successfully developed. In 1992 a new empirical model was introduced to establish a correlation with the operating parameters. The comparison results of the predicted iodine-131 equivalent activity value with the operating radiochemistry database was off by 23%. This paper presents an improved model. Although it is still an empirical model which also gives a first order estimation of the peak iodine spiking magnitude, the deviation between prediction and measurement was reduced to ∼7%. It is believed that this improved model can be used for better prediction and control of the iodine spiking magnitude resulted from failed fuel rods during power transients or plant shutdown. (author)
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.
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
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)
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
Clamens, Olivier; Lecerf, Johann; Hudelot, Jean-Pascal; Duc, Bertrand; Cadiou, Thierry; Blaise, Patrick; Biard, Bruno
2018-01-01
CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN) and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.
Directory of Open Access Journals (Sweden)
Clamens Olivier
2018-01-01
Full Text Available CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.
Nonlinear transient dynamic response of pressure relief valves for a negative containment system
International Nuclear Information System (INIS)
Aziz, T.S.; Duff, C.G.; Tang, J.H.K.
1979-01-01
The response of the piston for the postulated simultaneous effect of pressure and an earthquake is obtained for different parameters and accident conditions. Response quantities such as accelerations, displacements, rotations, diaphragm forces as well as opening time during a design basis earthquake are obtained. The results of the different analyses, as related to the functional operability of the valves, are evaluated and discussed. (orig.)
Transient effects caused by pulsed gas and liquid injections into low pressure plasmas
International Nuclear Information System (INIS)
Ogawa, D; Goeckner, M; Overzet, L; Chung, C W
2010-01-01
The fast injection of liquid droplets into a glow discharge causes significant time variations in the pressure, the chemical composition of the gas and the phases present (liquid and/or solid along with gas). While the variations can be large and important, very few studies, especially kinetic studies, have been published. In this paper we examine the changes brought about in argon plasma by injecting Ar (gas), N 2 (gas) hexane (gas) and hexane (liquid droplets). The changes in the RF capacitively coupled power (forward and reflected), electron and ion density (n e , n i ), electron temperature (T e ) and optical emissions were monitored during the injections. It was found that the Ar injection (pressure change only) caused expected variations. The electron temperature reduced, the plasma density increased and the optical emission intensity remained nearly constant. The N 2 and hexane gas injections (chemical composition and pressure changes) also followed expected trends. The plasma densities increased and electron temperature decreased while the optical emissions changed from argon to the injected gas. These all serve to highlight the fact that the injection of evaporating hexane droplets in the plasma caused very little change. This is because the number of injected droplets is too small to noticeably affect the plasma, even though the shift in the chemical composition of the gas caused by evaporation from those same droplets can be very significant. The net conclusion is that using liquid droplets to inject precursors for low pressure plasmas is both feasible and controllable.
Virtual cylinder pressure sensor for transient operation in heavy-duty engines
Kulah, S.; Donkers, M.C.F.; Willems, F.P.T.
2015-01-01
Cylinder pressure-based combustion control is widely introduced for passenger cars. Benefits include enhanced emission robustness to fuel quality variation, reduced fuel consumption due to more accurate (multi-pulse) fuel injection, and minimized after treatment size. In addition, it enables the
Virtual Cylinder Pressure Sensor for Transient Operation in Heavy-Duty Engines
Kulah, S.; Donkers, T.; Willems, F.
2015-01-01
Cylinder pressure-based combustion control is widely introduced for passenger cars. Benefits include enhanced emission robustness to fuel quality variation, reduced fuel consumption due to more accurate (multi-pulse) fuel injection, and minimized after treatment size.In addition, it enables the
Transient upregulation of protein kinase C in pressure-overloaded neonatal rat myocardium
Czech Academy of Sciences Publication Activity Database
Hamplová, B.; Novák, F.; Kolář, František; Nováková, O.
2010-01-01
Roč. 59, č. 1 (2010), s. 25-33 ISSN 0862-8408 R&D Projects: GA MŠk(CZ) 1M0510 Institutional research plan: CEZ:AV0Z50110509 Keywords : protein kinase C * cardiac development * pressure overload Subject RIV: ED - Physiology Impact factor: 1.646, year: 2010
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)
International Nuclear Information System (INIS)
Martin, R.P.; Nassersharif, B.
1988-01-01
The state of the art in artificial intelligence (AI) and expert system (ES) technology has matured to a degree that the potential development of a computer-aided/automated diagnostic and transient mitigation system in the area of nuclear reactor operation can be considered. Since traditional methods cannot handle complex systems efficiently, AI techniques provide a means to emulate an expert reactor operator rather than follow mechanistic methods. Computer-aided transient analysis coded in LISP (CATA-Lisp) is a confidence level based expert system written in Common LISP on the SYMBOLICS 3640 computer system. New versions are being developed in Common LISP for the Texas Instruments (TI) Explorer and the Sun microsystems machines, CATALisp manipulates both a knowledge base of transient identifier patterns (tree structured to allow for zooming in diagnostics) and a knowledge base containing a qualitative model of a nuclear power plant. The interference engine used by CATALisp uses the information stored in both knowledge bases to arrive at confidence level values that are used to infer particular plant states
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
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
Detailed Analysis of ECMWF Surface Pressure Data
Fagiolini, E.; Schmidt, T.; Schwarz, G.; Zenner, L.
2012-04-01
Investigations of temporal variations within the gravity field of the Earth led us to the analysis of common surface pressure data products delivered by ECMWF. We looked into the characteristics of global as well as spatially and temporally confined phenomena being visible in the data. In particular, we were interested in the overall data quality, the local and temporal signal-to-noise ratio of surface pressure data sets, and the identification of irregular data. To this end, we analyzed a time series of a full year of surface pressure operational analysis data and their nominal standard deviations. The use of pressure data on a Gaussian grid data allowed us to remain close to the internal computations at ECMWF during data assimilation. Thus, we circumvented potential interpolation effects that would otherwise occur in cylindrical projections of conventional map products. The results obtained by us demonstrate the identification of a few distinct outliers, data quality effects over land or water and along coastlines as well as neighborhood effects of samples within and outside of the tropics. Small scale neighborhood effects depend on their geographical direction, sampling distance, land or water, and local time. In addition, one notices large scale seasonal effects that are latitude and longitude dependent. As a consequence, we obtain a cause-and-effect survey of pressure data peculiarities. One can then use background corrected pressure data to analyze seasonal effects within given latitude belts. Here time series of pressure data allow the tracking of high and low pressure areas together with the identification of their actual extent, velocity and life time. This information is vital to overall mass transport calculations and the determination of temporally varying gravity fields. However, one has to note that the satellite and ground-based instruments and the assimilation software being used for the pressure calculations will not remain the same over the years
Miller, Andrew; Villegas, Arturo; Diez, F Javier
2015-03-01
The solution to the startup transient EOF in an arbitrary rectangular microchannel is derived analytically and validated experimentally. This full 2D transient solution describes the evolution of the flow through five distinct periods until reaching a final steady state. The derived analytical velocity solution is validated experimentally for different channel sizes and aspect ratios under time-varying pressure gradients. The experiments used a time resolved micro particle image velocimetry technique to calculate the startup transient velocity profiles. The measurements captured the effect of time-varying pressure gradient fields derived in the analytical solutions. This is tested by using small reservoirs at both ends of the channel which allowed a time-varying pressure gradient to develop with a time scale on the order of the transient EOF. Results showed that under these common conditions, the effect of the pressure build up in the reservoirs on the temporal development of the transient startup EOF in the channels cannot be neglected. The measurements also captured the analytical predictions for channel walls made of different materials (i.e., zeta potentials). This was tested in channels that had three PDMS and one quartz wall, resulting in a flow with an asymmetric velocity profile due to variations in the zeta potential between the walls. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.
Yuji Nishi; Tsuneo Ishido
2012-01-01
In order to appraise the utility of self-potential (SP) measurements to characterize fractured reservoirs, we carried out continuous SP monitoring using multi Ag-AgCl electrodes installed within two open holes at the Kamaishi Mine, Japan. The observed ratio of SP change to pressure change associated with fluid flow showed different behaviors between intact host rock and fractured rock regions. Characteristic behavior peculiar to fractured reservoirs, which is predicted from numerical simulati...
Fundamental Studies of Transient, Atmospheric-Pressure, Small-Scale Plasmas
2017-01-23
C. Jiang, R. Heller, J. Lane, and K. H. Schoenbach, " Ozone -free nitric oxide production using an atmospheric pressure surface discharge – a way to...Electrostatic modeling and energy-dependent studies showed that the direct and indirect electron-induced processes in the pulsed plasma jet are responsible for...Coupled sliding discharges : a scalable nonthermal plasma system utilizing positive and negative streamers on DISTRIBUTION A: Distribution
International Nuclear Information System (INIS)
Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.
1991-09-01
The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)
Directory of Open Access Journals (Sweden)
Yuji Nishi
2012-01-01
Full Text Available In order to appraise the utility of self-potential (SP measurements to characterize fractured reservoirs, we carried out continuous SP monitoring using multi Ag-AgCl electrodes installed within two open holes at the Kamaishi Mine, Japan. The observed ratio of SP change to pressure change associated with fluid flow showed different behaviors between intact host rock and fractured rock regions. Characteristic behavior peculiar to fractured reservoirs, which is predicted from numerical simulations of electrokinetic phenomena in MINC (multiple interacting continua double-porosity media, was observed near the fractures. Semilog plots of the ratio of SP change to pressure change observed in one of the two wells show obvious transition from intermediate time increasing to late time stable trends, which indicate that the time required for pressure equilibration between the fracture and matrix regions is about 800 seconds. Fracture spacing was estimated to be a few meters assuming several micro-darcies (10-18 m2 of the matrix region permeability, which is consistent with geological and hydrological observations.
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.
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.
Integrated Software Environment for Pressurized Thermal Shock Analysis
Directory of Open Access Journals (Sweden)
Dino Araneo
2011-01-01
Full Text Available The present paper describes the main features and an application to a real Nuclear Power Plant (NPP of an Integrated Software Environment (in the following referred to as “platform” developed at University of Pisa (UNIPI to perform Pressurized Thermal Shock (PTS analysis. The platform is written in Java for the portability and it implements all the steps foreseen in the methodology developed at UNIPI for the deterministic analysis of PTS scenarios. The methodology starts with the thermal hydraulic analysis of the NPP with a system code (such as Relap5-3D and Cathare2, during a selected transient scenario. The results so obtained are then processed to provide boundary conditions for the next step, that is, a CFD calculation. Once the system pressure and the RPV wall temperature are known, the stresses inside the RPV wall can be calculated by mean a Finite Element (FE code. The last step of the methodology is the Fracture Mechanics (FM analysis, using weight functions, aimed at evaluating the stress intensity factor (KI at crack tip to be compared with the critical stress intensity factor KIc. The platform automates all these steps foreseen in the methodology once the user specifies a number of boundary conditions at the beginning of the simulation.
International Nuclear Information System (INIS)
Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ernestova, M.; Zamboch, M.; Seifert, H.P.; Ritter, S.; Roth, A.; Devrient, B.; Ehrnsten, U.
2004-01-01
The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the transferability of the results to components in service. The obtained crack growth rate data were reflected on literature data and on commonly applied prediction curves as presented in the appropriate Code. At relevant stress intensity factors it could be shown that immediate cessation of growing cracks occurs after changing from cyclic to static load in high purity oxygenated BWR water and oxygen-free VVER water corresponding to steady state operation conditions. Susceptibility to environmentally assisted cracking under static load was observed for a heat affected zone material in oxygenated high purity water and also in base materials during a chloride transient representing BWR water condition below Action Level 1 of the EPRI Water Chemistry Guidelines according to the lectrical conductivity of the water but in the range of Action Level 2 according to the content of chlorides. Time based crack growth was also observed in one Russian type base material in oxygenated VVER water and in one Western type base material in oxygenated high purity BWR
Sensitivity analysis of a PWR pressurizer
International Nuclear Information System (INIS)
Bruel, Renata Nunes
1997-01-01
A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)
Shukla, Akash; Meshram, Megha; Gopan, Amrit; Ganjewar, Vaibhav; Kumar, Praveen; Bhatia, Shobna J
2012-06-01
Transient lower esophageal sphincter relaxation (tLESR) and decreased basal lower esophageal sphincter (LES) pressure are postulated mechanisms of gastroesophageal reflux (GER). There is conflicting evidence on the effect of carbonated drinks on lower esophageal sphincter function. This study was conducted to assess the effect of a carbonated beverage on tLESR and LES pressure. High resolution manometry tracings (16 channel water-perfused, Trace 1.2, Hebbard, Australia) were obtained in 18 healthy volunteers (6 men) for 30 min each at baseline, and after 200 mL of chilled potable water and 200 mL of chilled carbonated cola drink (Pepsi [Pepsico India Ltd]). The sequence of administration of the drinks was determined by random number method generated by a computer. The analysis of tracings was done using TRACE 1.2 software by a physician who was unaware of the sequence of administration of fluids. The mean (SD) age of the participant was 37.3 (12.9) years. The median (range) frequency of tLESr was higher after the carbonated beverage (10.5 [0-26]) as compared to baseline (0 [0-3], p = 0.005) as well as after water (1 [0-14], p = 0.010). The LES pressure decreased after ingestion of the carbonated beverage (18.5 [11-37] mmHg) compared to baseline (40.5 [25-66] mmHg, p = 0.0001) and after water (34 [15-67] mmHg, p = 0.003). Gastric pressure was not different in the three groups. Ingestion of a carbonated beverage increases tLESr and lowers LES pressure in healthy subjects.
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
Performance Analysis of Waste Heat Driven Pressurized Adsorption Chiller
LOH, Wai Soong
2010-01-01
This article presents the transient modeling and performance of waste heat driven pressurized adsorption chillers for refrigeration at subzero applications. This innovative adsorption chiller employs pitch-based activated carbon of type Maxsorb III (adsorbent) with refrigerant R134a as the adsorbent-adsorbate pair. It consists of an evaporator, a condenser and two adsorber/desorber beds, and it utilizes a low-grade heat source to power the batch-operated cycle. The ranges of heat source temperatures are between 55 to 90°C whilst the cooling water temperature needed to reject heat is at 30°C. A parametric analysis is presented in the study where the effects of inlet temperature, adsorption/desorption cycle time and switching time on the system performance are reported in terms of cooling capacity and coefficient of performance. © 2010 by JSME.
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.)
Seal analysis technology for reactor pressure vessel
International Nuclear Information System (INIS)
Zheng Liangang; Zhang Liping; Yang Yu; Zang Fenggang
2009-01-01
There is the coolant with radiation, high temperature and high pressure in the reactor pressure vessel (RPV). It is closely correlated to RPV sealing capability whether the whole nuclear system work well or not. The aim of this paper is to study the seal analysis method and technology, such as the pre-tensioning of the bolt, elastoplastic contact and coupled technology of thermal and structure. The 3 D elastoplastic seal analysis method really and generally consider the loads and model the contact problem with friction between the contact plates. This method is easier than the specialized seal program and used widely. And it is more really than the 2 D seal analysis method. This 3 D elastoplastic seal analysis method has been successfully used in the design and analysis of RPV. (authors)
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
associated to various individual points over the state space. By applying a novel multi-dimensional clustering technique, based on the non-parametric statistical Kruskal-Wallis test, it has been possible to achieve a partitioning of the state space into regions differing in terms of the quality of the physical model's predictions. The second step has been the quantification of the model's uncertainty, for each of the identified state space regions, by applying a probability density function (pdf) estimator. This is a kernel-type estimator, modelled on a universal orthogonal series estimator, such that its behaviour takes advantage of the good features of both estimator types and yields reasonable pdfs, even with samples of small size and not very compact distributions. The pdfs provide a reliable basis for sampling 'error values' for use in Monte-Carlo-type uncertainty propagation studies, aimed at quantifying the impact of the physical model's uncertainty on the code's output variables of interest. The effectiveness of the developed methodology was demonstrated by applying it to the quantification of the uncertainty related to thermal-hydraulic (drift-flux) models implemented in the best-estimate safety analysis code RETRAN-3D. This has been done via the usage of a wide database of void-fraction experiments for saturated and sub-cooled conditions. Appropriate pdfs were generated for quantification of the physical model's uncertainty in a 2-dimensional (pressure/mass-flux) state space, partitioned into 3 separate regions. The impact of the RETRAN-3D drift-flux model uncertainties has been assessed at three different levels of the code's application: (a) Achilles Experiment No. 2, a separate effect experiment not included in the original assessment database; (b) Omega Rod Bundle Test No. 9, an integral experiment simulating a PWR loss-of-coolant accident (LOCA); and (c) the Peach Bottom turbine trip test, a NPP (BWR) plant transient in which the void feedback mechanism plays
Design by analysis of composite pressure equipment
International Nuclear Information System (INIS)
Durand, S.; Mallard, H.
2004-01-01
Design by analysis has been particularly pointed out by the european pressure equipment directive. Advanced mechanical analysis like finite element method are used instead of classical design by formulas or graphs. Structural behaviour can be understood by the designer. Design by analysis of metallic pressure equipments is widely used. Material behaviour or limits analysis is based on sophisticated approach (elasto-plastic analysis,..). Design by analysis of composite pressure equipments is not systematically used for industrial products. The difficulty comes from the number of information to handle. The laws of mechanics are the same for composite materials than for steel. The authors want to show that in design by analysis, the composite material approach is only more complete than the metallic approach. Mechanics is more general but not more complicated. A multi-material approach is a natural evolution of design by analysis of composite equipments. The presentation is illustrated by several industrial cases - composite vessel: analogy with metallic calculations; - composite pipes and fittings; - welding and bounding of thermoplastic equipments. (authors)
Assessment of the TRINO reactor pressure vessel integrity: theoretical analysis and NDE
Energy Technology Data Exchange (ETDEWEB)
Milella, P P; Pini, A [ENEA, Rome (Italy)
1988-12-31
This document presents the method used for the capability assessment of the Trino reactor pressure vessel. The vessel integrity assessment is divided into the following parts: transients evaluation and selection, fluence estimate for the projected end of life of the vessel, characterization of unirradiated and irradiated materials, thermal and stress analysis, fracture mechanics analysis and eventually fracture input to Non Destructive Examination (NDE). For each part, results are provided. (TEC).
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.
Buoyancy effects in overcooling transients calculated for the NRC pressurized thermal shock study
International Nuclear Information System (INIS)
Theofanous, T.G.; Iyer, K.; Nourbakhsh, H.P.; Gherson, P.
1986-05-01
The thermal-hydraulic responses of three PWRs (Oconee, Calvert Cliffs, and H.B. Robinson), to postulated Pressurized Thermal Shock (PTS) scenarios, which were originally determined by RELAP5 and TRAC calculations, are being further developed here with regard to buoyancy/stratification effects. These three PWRs were the subject of the NRC PTS study, and the present results helped define the thermal-hydraulic conditions utilized in the fracture mechanics calculations carried out at ORNL. The computer program REMIX, which is based on the Regional Mixing Model (RMM), was the analytical tool employed, while Purdue's 1/2-Scale HPI Thermal Mixing facility provided the basis for experimental support. Important mixing and wall heat transfer regimes are delineated on the basis of these results. We conclude that stratification is important only in cases of complete loop stagnation and that mixed-convection effects are important for downcomer flow velocities below approx.0.25 m/s. The stratification is small in magnitude, however it is important in creating a recirculating flow pattern which activates the lower plenum, pump and loop seal volumes, to participate in the mixing process. This mixing process together with the heat input from the wall metal significantly impact the cooldown rates. Heat transfer in the plume region is dominated by forced convection. On the other hand, the presence of the Reactor Pressure Vessel (RPV) wall cladding and wall conduction significantly dampen the free convection effects in the low velocity, mixed-convection, regime. For the stagnant loop cases, all locations outside the plume region are included in this regime. In the presence of natural loop circulation and a uniformly distributed downcomer flow, the mixed convection regime is also expected, however, the forced convection regime can also be observed in highly asymmetric flow behavior
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
Numerical issues for liquid-metal boiling transient analysis
International Nuclear Information System (INIS)
Rowe, D.S.
1986-01-01
The large liquid-to-vapor density ratio of a boiling liquid-metal leads to a very abrupt change of the two-phase mixture density at the inception of boiling. Unfortunately, the strong dependence of mixture density on pressure leads to a key numerical issue that adversely affects the behavior of numerical solutions. The difficulties can be reduced by using techniques that acknowledge this functional behavior at the start of boiling. Some of the methods used include a spatially averaged density function, mathematical smoothing, and under relaxation. Nonequilibrium two-fluid models also seem to offer aid in obtaining reliable numerical solutions. (author)
Analysis of nuclear reactor pressure vessel flanges
International Nuclear Information System (INIS)
Oliveira, C.A.N. de; Augusto, O.B.
1985-01-01
This work proposes a methodology for the structural analysis of high diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem, and the results are compared with results obtained by the finite element method. (Author) [pt
Time series analysis of barometric pressure data
International Nuclear Information System (INIS)
La Rocca, Paola; Riggi, Francesco; Riggi, Daniele
2010-01-01
Time series of atmospheric pressure data, collected over a period of several years, were analysed to provide undergraduate students with educational examples of application of simple statistical methods of analysis. In addition to basic methods for the analysis of periodicities, a comparison of two forecast models, one based on autoregression algorithms, and the other making use of an artificial neural network, was made. Results show that the application of artificial neural networks may give slightly better results compared to traditional methods.
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)
Analysis of pressurization of plutonium oxide storage vials during a postulated fire
Energy Technology Data Exchange (ETDEWEB)
Laurinat, J.; Kesterson, M.; Hensel, S.
2015-02-10
The documented safety analysis for the Savannah River Site evaluates the consequences of a postulated 1000 °C fire in a glovebox. The radiological dose consequences for a pressurized release of plutonium oxide powder during such a fire depend on the maximum pressure that is attained inside the oxide storage vial. To enable evaluation of the dose consequences, pressure transients and venting flow rates have been calculated for exposure of the storage vial to the fire. A standard B vial with a capacity of approximately 8 cc was selected for analysis. The analysis compares the pressurization rate from heating and evaporation of moisture adsorbed onto the plutonium oxide contents of the vial with the pressure loss due to venting of gas through the threaded connection between the vial cap and body. Tabulated results from the analysis include maximum pressures, maximum venting velocities, and cumulative vial volumes vented during the first 10 minutes of the fire transient. Results are obtained for various amounts of oxide in the vial, various amounts of adsorbed moisture, different vial orientations, and different surface fire exposures.
Directory of Open Access Journals (Sweden)
Möhlenkamp Stefan
2006-06-01
Full Text Available Abstract Background The pressure drop – flow relations in myocardial bridges and the assessment of vascular heart disease via fractional flow reserve (FFR have motivated many researchers the last decades. The aim of this study is to simulate several clinical conditions present in myocardial bridges to determine the flow reserve and consequently the clinical relevance of the disease. From a fluid mechanical point of view the pathophysiological situation in myocardial bridges involves fluid flow in a time dependent flow geometry, caused by contracting cardiac muscles overlying an intramural segment of the coronary artery. These flows mostly involve flow separation and secondary motions, which are difficult to calculate and analyse. Methods Because a three dimensional simulation of the haemodynamic conditions in myocardial bridges in a network of coronary arteries is time-consuming, we present a boundary layer model for the calculation of the pressure drop and flow separation. The approach is based on the assumption that the flow can be sufficiently well described by the interaction of an inviscid core and a viscous boundary layer. Under the assumption that the idealised flow through a constriction is given by near-equilibrium velocity profiles of the Falkner-Skan-Cooke (FSC family, the evolution of the boundary layer is obtained by the simultaneous solution of the Falkner-Skan equation and the transient von-Kármán integral momentum equation. Results The model was used to investigate the relative importance of several physical parameters present in myocardial bridges. Results have been obtained for steady and unsteady flow through vessels with 0 – 85% diameter stenosis. We compare two clinical relevant cases of a myocardial bridge in the middle segment of the left anterior descending coronary artery (LAD. The pressure derived FFR of fixed and dynamic lesions has shown that the flow is less affected in the dynamic case, because the distal
Rohof, W. O. A.; Boeckxstaens, G. E. E.; Hirsch, D. P.
2011-01-01
Transient lower esophageal sphincter relaxations (TLESRs) are the main mechanism underlying gastro-esophageal reflux and are detected during manometric studies using well defined criteria. Recently, high-resolution esophageal pressure topography (HREPT) has been introduced and is now considered as
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.
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)
Prasser, Horst-Michael; Grunwald, Gerhard; Hoehne, Thomas; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter
2003-01-01
The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold
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
RETRAN sensitivity studies of light water reactor transients. Final report
International Nuclear Information System (INIS)
Burrell, N.S.; Gose, G.C.; Harrison, J.F.; Sawtelle, G.R.
1977-06-01
This report presents the results of sensitivity studies performed using the RETRAN/RELAP4 transient analysis code to identify critical parameters and models which influence light water reactor transient predictions. Various plant transients for both boiling water reactors and pressurized water reactors are examined. These studies represent the first detailed evaluation of the RETRAN/RELAP4 transient code capability in predicting a variety of plant transient responses. The wide range of transients analyzed in conjunction with the parameter and modeling studies performed identify several sensitive areas as well as areas requiring future study and model development
Atwell, Matthew J.; Hurlbert, Eric A.; Melcher, J. C.; Morehead, Robert L.
2017-01-01
some tests to demonstrate the capability to quickly condition the engines for higher pulsing demand scenarios. A thermocouple at the TVS outlet allows for the calculation of energy absorbed by the vented propellant. Lastly, tests with longer pulses and multiple engines firing either in sequence or simultaneously were run in order to gather transient system response data on waterhammer. Six total high-speed pressure transducers are installed on the RCS system, one sensor at the end of each propellant manifold line on the pods, and one at the tap-off location for each commodity. This will allow for the accurate characterization of waterhammer in the system under various propellant conditions and firing sequences. Other instrumentation for this test series includes nozzle throat thermocouples, chamber pressure measurement, heat soakback measurement, and tank wall plume impingement temperature measurement. The next set of tests were performed to demonstrate simultaneous main engine and RCS operation. Data from this test will be used to examine if there is any change to nominal operation of the RCS as a result of feed system interaction or other phenomenon. Some of these tests began under high vacuum conditions (target ambient pressure less than 1x10(exp -3) torr) and others began at altitude conditions. The last set of tests were performed with the B-2 cold wall active. Under these tests, many of the same low duty cycle MIB tests were repeated in order to characterize how propellant conditions changed with the lower heat leak. In this scenario the RCS manifold experiences much less heat leak, resulting in a change to how well the engines self-condition. As a result, an increase in maximum waterhammer pressures and a change in natural frequency of the system was expected due to higher density propellants. The lower heat leak should also result in a change to the MIB pulse profile, and data will be examined to understand how MIB repeatability is affected in the different
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...
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
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
Computational methods for fracture mechanics analysis of pressurized-thermal-shock experiments
International Nuclear Information System (INIS)
Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.
1984-01-01
Extensive computational analyses are required to determine material parameters and optimum pressure-temperature transients compatible with proposed pressurized-thermal-shock (PTS) test scenarios and with the capabilities of the PTS test facility at the Oak Ridge National Laboratory (ORNL). Computational economy has led to the application of techniques suitable for parametric studies involving the analysis of a large number of transients. These techniques, which include analysis capability for two- and three-dimensional (2-D and 3-D) superposition, inelastic ligament stability, and upper-shelf arrest, have been incorporated into the OCA/USA computer program. Features of the OCA/USA program are discussed, including applications to the PTS test configuration
Computational methods for fracture mechanics analysis of pressurized-thermal-shock experiments
International Nuclear Information System (INIS)
Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.
1984-01-01
Extensive computational analyses are required to determine material parameters and optimum pressure-temperature transients compatible with proposed pressurized-thermal-shock (PTS) test scenarios and with the capabilities of the PTS test facility at the Oak Ridge National Laboratory (ORNL). Computational economy has led to the application of techniques suitable for parametric studies involving the analysis of a large number of transients. These techniques, which include analysis capability for two- and three-dimensional (2-D and 3-D) superposition, inelastic ligament stability, and upper-shelf arrest, have been incorporated into the OCA/ USA computer program. Features of the OCA/USA program are discussed, including applications to the PTS test configuration. (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.
International Nuclear Information System (INIS)
Zhang, S; Sobota, A; Van Veldhuizen, E M; Bruggeman, P J
2015-01-01
The ozone density distribution in the effluent of a time modulated RF atmospheric pressure plasma jet (APPJ) is investigated by time and spatially resolved by UV absorption spectroscopy. The plasma jet is operated with an averaged dissipated power of 6.5 W and gas flow rate 2 slm argon +2% O 2 . The modulation frequency of the RF power is 50 Hz with a duty cycle of 50%. To investigate the production and destruction mechanism of ozone in the plasma effluent, the atomic oxygen and gas temperature is also obtained by TALIF and Rayleigh scattering, respectively. A temporal increase in ozone density is observed close to the quartz tube exit when the plasma is switched off due to the decrease in O density and gas temperature. Ozone absorption at different axial positions indicates that the ozone distribution is dominated by the convection induced by the gas flow and allows estimating the on-axis local gas velocity in the jet effluent. Transient vortex structures occurring during the switch on and off of the RF power also significantly affect the ozone density in the far effluent. (paper)
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
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.
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)
Thermohydraulic analysis of pressurized water reactors
International Nuclear Information System (INIS)
Veloso, M.A.
1980-01-01
The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt
Reliability analysis of reactor pressure vessel intensity
International Nuclear Information System (INIS)
Zheng Liangang; Lu Yongbo
2012-01-01
This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)
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.
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
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
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.)
Directory of Open Access Journals (Sweden)
Thivaharan Albin
2016-07-01
Full Text Available Increasingly complex air path concepts are investigated to achieve a substantial reduction in fuel consumption while improving the vehicle dynamics. One promising technology is the two-stage turbocharging for gasoline engines, where a high pressure and a low pressure turbocharger are placed in series. For exploiting the high potential, a control concept has to be developed that allows for coordinated management of the two turbocharger stages. In this paper, the control strategy is investigated. Therefore, the effect of the actuated values on transient response and pumping losses is analyzed. Based on these findings, an optimization-based control algorithm is developed that allows taking both requirements into account. The developed new controller allows achieving a fast transient response, while at the same time reducing pumping losses in stationary operation.
Thermo-mechanical analysis of the pressure plate of clutch
Directory of Open Access Journals (Sweden)
P.V.N. Venkata Mallikarjuna
2017-09-01
Full Text Available High Temperature appears in the contact surfaces of a clutch system (friction surface and pressure plate due to the relative motion between these parts during the sliding period. These high temperatures are responsible for several failures such as pressure plate crack, pressure plate warpage etc. With the help of Finite element analysis, the sliding friction process of the pressure plate and friction during clutch engagement is simulated to get temperature field characteristics and contact pressure of pressure plate.
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
International Nuclear Information System (INIS)
Wheeler, A.J.
1978-02-01
An analytical model is described that computes the transient pressures, velocities and forces in the safety/relief valve discharge line immediately after safety/relief valve opening. Equations of motion are defined for the gas-flow and water-flow models. Results are not only verified by comparing them with an earlier version of the model, but also with Quad Cities and Monticello plant data. The model shows reasonable agreement with the earlier model and the plant data
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.
Urquiza, Eugenio
This work presents a comprehensive thermal hydraulic analysis of a compact heat exchanger using offset strip fins. The thermal hydraulics analysis in this work is followed by a finite element analysis (FEA) to predict the mechanical stresses experienced by an intermediate heat exchanger (IHX) during steady-state operation and selected flow transients. In particular, the scenario analyzed involves a gas-to-liquid IHX operating between high pressure helium and liquid or molten salt. In order to estimate the stresses in compact heat exchangers a comprehensive thermal and hydraulic analysis is needed. Compact heat exchangers require very small flow channels and fins to achieve high heat transfer rates and thermal effectiveness. However, studying such small features computationally contributes little to the understanding of component level phenomena and requires prohibitive computational effort using computational fluid dynamics (CFD). To address this issue, the analysis developed here uses an effective porous media (EPM) approach; this greatly reduces the computation time and produces results with the appropriate resolution [1]. This EPM fluid dynamics and heat transfer computational code has been named the Compact Heat Exchanger Explicit Thermal and Hydraulics (CHEETAH) code. CHEETAH solves for the two-dimensional steady-state and transient temperature and flow distributions in the IHX including the complicating effects of temperature-dependent fluid thermo-physical properties. Temperature- and pressure-dependent fluid properties are evaluated by CHEETAH and the thermal effectiveness of the IHX is also calculated. Furthermore, the temperature distribution can then be imported into a finite element analysis (FEA) code for mechanical stress analysis using the EPM methods developed earlier by the University of California, Berkeley, for global and local stress analysis [2]. These simulation tools will also allow the heat exchanger design to be improved through an
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)
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
Hanley, Janet; Fairbrother, Peter; Krishan, Ashma; McCloughan, Lucy; Padfield, Paul; Paterson, Mary; Pinnock, Hilary; Sheikh, Aziz; Sudlow, Cathie; Todd, Allison; McKinstry, Brian
2015-03-25
Good blood pressure (BP) control reduces the risk of recurrence of stroke/transient ischaemic attack (TIA). Although there is strong evidence that BP telemonitoring helps achieve good control, none of the major trials have considered the effectiveness in stroke/TIA survivors. We therefore conducted a feasibility study for a trial of BP telemonitoring for stroke/TIA survivors with uncontrolled BP in primary care. Phase 1 was a pilot trial involving 55 patients stratified by stroke/TIA randomised 3:1 to BP telemonitoring for 6 months or usual care. Phase 2 was a qualitative evaluation and comprised semi-structured interviews with 16 trial participants who received telemonitoring and 3 focus groups with 23 members of stroke support groups and 7 carers. Overall, 125 patients (60 stroke patients, 65 TIA patients) were approached and 55 (44%) patients were randomised including 27 stroke patients and 28 TIA patients. Fifty-two participants (95%) attended the 6-month follow-up appointment, but one declined the second daytime ambulatory blood pressure monitoring (ABPM) measurement resulting in a 93% completion rate for ABPM - the proposed primary outcome measure for a full trial. Adherence to telemonitoring was good; of the 40 participants who were telemonitoring, 38 continued to provide readings throughout the 6 months. There was a mean reduction of 10.1 mmHg in systolic ABPM in the telemonitoring group compared with 3.8 mmHg in the control group, which suggested the potential for a substantial effect from telemonitoring. Our qualitative analysis found that many stroke patients were concerned about their BP and telemonitoring increased their engagement, was easy, convenient and reassuring. A full-scale trial is feasible, likely to recruit well and have good rates of compliance and follow-up. ISRCTN61528726 15/12/2011.
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
Liu, Yan; Qi, Hanping; E, Mingyao; Shi, Pilong; Zhang, Qianhui; Li, Shuzhi; Wang, Ye; Cao, Yonggang; Chen, Yunping; Ba, Lina; Gao, Jingquan; Huang, Wei; Sun, Hongli
2018-02-01
Cardiac fibrosis is a common pathologic change along with pressure overload. Recent studies indicated that transient receptor potential (TRP) channels played multiple roles in heart. However, the functional role of transient receptor potential vanilloid-3 (TRPV3) in cardiac fibrosis remained unclear. The present study was designed to investigate the relationship between TRPV3 activation and pressure overload-induced cardiac fibrosis. Pressure overload rats were successfully established by abdominal aortic constriction (AAC), and cardiac fibrosis was simulated by 100 nM angiotensin II (Ang II) in neonatal cardiac fibroblasts. Echocardiographic parameters, cardiac fibroblast proliferation, cell cycle, intracellular calcium concentration ([Ca 2+ ] i ), and the protein expressions of collagen I, collagen III, transforming growth factor beta 1 (TGF-β 1 ), cyclin E, and cyclin-dependent kinase 2 (CDK2) were measured. Echocardiographic and histological measurements suggested that the activation of TRPV3 exacerbated the cardiac dysfunction and increased interstitial fibrosis in pressure overload rats. Further results showed that TRPV3 activation upregulated the expressions of collagen I, collagen III, TGF-β 1 , cyclin E, and CDK2 in vivo and in vitro. At the same time, blocking TGF-β 1 pathway could partially reverse the effect of TRPV3 activation. These results suggested that TRPV3 activation exacerbated cardiac fibrosis by promoting cardiac fibroblast proliferation through TGF-β 1 /CDK2/cyclin E pathway in the pressure-overloaded rat hearts.
Energy Technology Data Exchange (ETDEWEB)
Faydide, B. [Commissariat a l`Energie Atomique, Grenoble (France)
1997-07-01
This paper presents the current and planned numerical development for improving computing performance in case of Cathare applications needing real time, like simulator applications. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the general characteristics of the code are presented, dealing with physical models, numerical topics, and validation strategy. Then, the current and planned applications of Cathare in the field of simulators are discussed. Some of these applications were made in the past, using a simplified and fast-running version of Cathare (Cathare-Simu); the status of the numerical improvements obtained with Cathare-Simu is presented. The planned developments concern mainly the Simulator Cathare Release (SCAR) project which deals with the use of the most recent version of Cathare inside simulators. In this frame, the numerical developments are related with the speed up of the calculation process, using parallel processing and improvement of code reliability on a large set of NPP transients.
International Nuclear Information System (INIS)
Faydide, B.
1997-01-01
This paper presents the current and planned numerical development for improving computing performance in case of Cathare applications needing real time, like simulator applications. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the general characteristics of the code are presented, dealing with physical models, numerical topics, and validation strategy. Then, the current and planned applications of Cathare in the field of simulators are discussed. Some of these applications were made in the past, using a simplified and fast-running version of Cathare (Cathare-Simu); the status of the numerical improvements obtained with Cathare-Simu is presented. The planned developments concern mainly the Simulator Cathare Release (SCAR) project which deals with the use of the most recent version of Cathare inside simulators. In this frame, the numerical developments are related with the speed up of the calculation process, using parallel processing and improvement of code reliability on a large set of NPP transients
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.)
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.
Chernobyl reactor transient simulation study
International Nuclear Information System (INIS)
Gaber, F.A.; El Messiry, A.M.
1988-01-01
This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors
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,
International Nuclear Information System (INIS)
Kusunoki, Tsuyoshi; Yokomura, Takeyoshi; Nabeshima, Kunihiko; Shimazaki, Junya; Shinohara, Yoshikuni.
1988-01-01
This report describes the development of plant dynamic analysis code (ISPDYN) for integrated self-pressurized water reactor, and comparative study of pressure control methods with this code. ISPDYN is developed for integrated self-pressurized water reactor, one of the trial design by JAERI. In the transient responses, the calculated results by ISPDYN are in good agreement with the DRUCK calculations. In addition, this report presents some sensitivity studies for selected cases. Computing time of this code is very short so as about one fifth of real time. The comparative study of self-pressurized system with forced-pressurized system by this code, for rapid load decrease and increase cases, has provided useful informations. (author)
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)
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)
Webb, Alastair J S; Mazzucco, Sara; Li, Linxin; Rothwell, Peter M
2018-01-01
Visit-to-visit and day-to-day blood pressure (BP) variability (BPV) predict an increased risk of cardiovascular events but only reflect 1 form of BPV. Beat-to-beat BPV can be rapidly assessed and might also be predictive. In consecutive patients within 6 weeks of transient ischemic attack or nondisabling stroke (Oxford Vascular Study), BPV (coefficient of variation) was measured beat-to-beat for 5 minutes (Finometer), day-to-day for 1 week on home monitoring (3 readings, 3× daily), and on awake ambulatory BP monitoring. BPV after 1-month standard treatment was related (Cox proportional hazards) to recurrent stroke and cardiovascular events for 2 to 5 years, adjusted for mean systolic BP. Among 520 patients, 26 had inadequate beat-to-beat recordings, and 22 patients were in atrial fibrillation. Four hundred five patients had all forms of monitoring. Beat-to-beat BPV predicted recurrent stroke and cardiovascular events independently of mean systolic BP (hazard ratio per group SD, stroke: 1.47 [1.12-1.91]; P =0.005; cardiovascular events: 1.41 [1.08-1.83]; P =0.01), including after adjustment for age and sex (stroke: 1.47 [1.12-1.92]; P =0.005) and all risk factors (1.40 [1.00-1.94]; P =0.047). Day-to-day BPV was less strongly associated with stroke (adjusted hazard ratio, 1.29 [0.97-1.71]; P =0.08) but similarly with cardiovascular events (1.41 [1.09-1.83]; P =0.009). BPV on awake ambulatory BP monitoring was nonpredictive (stroke: 0.89 [0.59-1.35]; P =0.59; cardiovascular events: 1.08 [0.77-1.52]; P =0.65). Despite a weak correlation ( r =0.119; P =0.02), beat-to-beat BPV was associated with risk of recurrent stroke independently of day-to-day BPV (1.41 [1.05-1.90]; P =0.02). Beat-to-beat BPV predicted recurrent stroke and cardiovascular events, independently of mean systolic BP and risk factors but short-term BPV on ambulatory BP monitoring did not. Beat-to-beat BPV may be a useful additional marker of cardiovascular risk. © 2017 The Authors.
Containment pressure analysis model using CONTEMPT-LT
International Nuclear Information System (INIS)
Gupta, R.N.
1975-09-01
An analytical model for evaluating the reactor containment pressure transient following a loss-of-coolant accident (LOCA) is presented. The model uses the CONTEMPT-LT computer program developed by Aerojet Nuclear Company. The sample problem studied is the containment response following the most severe postulated LOCA at the Yankee Rowe Nuclear Power Station. The results show good agreement with the response predicted by Westinghouse Electric Corporation. (auth)
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.
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
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.
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.
International Nuclear Information System (INIS)
Benedetti, R.L.; Lords, L.V.; Kiser, D.M.
1978-02-01
The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage
Conformable Pressurized Structures : Design and Analysis
Geuskens, F.J.J.M.M.
2012-01-01
There are many applications where volume needs to be pressurised within a geometrical space for which conventional pressure vessels do not provide suitable solutions. Applications are for example found in pressure cabins for Blended Wing Body Aircraft and conformable pressure vessels for an
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.
International Nuclear Information System (INIS)
Soeren Kliem; Siegfried Mittag; Siegfried Langenbuch
2005-01-01
Full text of publication follows: The transition from the application of conservative models to the use of best-estimate models raises the question about the uncertainty of the obtained results. This question becomes especially important, if the best-estimate models should be used for safety analyses in the field of nuclear engineering. Different methodologies were developed to assess the uncertainty of the calculation results of computer simulation codes. One of them is the methodology developed by Gesellschaft fuer Anlagenund Reaktorsicherheit (GRS) which uses the statistical code package SUSA. In the past, this methodology was applied to the calculation results of the advanced thermal hydraulic system code ATHLET. In the frame of the recently finished EU FP5 funded research project VALCO, that methodology was extended and successfully applied to different coupled code systems, including the uncertainty analysis for neutronics. These code systems consist of a thermal hydraulic system code and a 3D neutron kinetic core model. One of the code systems applied was ATHLET coupled with the Rossendorf kinetics code DYN3D. Two real transients at NPPs with VVER-type reactors documented within the VALCO project were selected for analyses. One was the load drop of one of two turbines to house load level at the Loviisa-1 NPP (VVER-440), the second was a test with the switching-off of one of two main feed water pumps at the VVER-1000 Balakovo-4 NPP. The current paper is dedicated to the different steps of the use and implementation of the GRS methodology to coupled code systems and to the assessment of the results obtained by the DYN3D/ATHLET code. Based on the relevant physical processes in both transients, lists of possible sources of uncertainties were compiled. They are specific for the two transients. Besides control parameters like control rod movement and thermal hydraulic parameters like secondary side pressure, mass flow rates, pressurizer sprayer and heater
International Nuclear Information System (INIS)
Nagel, H.
1986-01-01
The flow induced valve operation is calculated for single and two-phase flow conditions by the fluid dynamic computer code DYVRO and results are compared to experimental data. The analysis show that the operational behaviour of the valves is not only dependent on the condition of the induced flow, but also the pipe flow can cause a feedback as a result of the induced pressure waves. For the calculation of pressure wave propagation in pipes of which the operation of flow induced valves has a considerable influence it is therefore necessary to have a coupled analysis of the pressure wave propagation and the operational behaviour of the valves. The analyses of the fast transient transfer from steam to two-phase flow show a good agreement with experimental data. Hence even these very high loads on pipes resulting from such fluid dynamic transients can be calculated realistically. (orig.)
International Nuclear Information System (INIS)
Ishiwatari, Y.; Oka, Y.; Koshizuka, S.
2002-01-01
A safety analysis code for a high temperature supercritical pressure light water cooled reactor (SCLWR-H) with water rods cooled by descending flow, SPRAT-DOWN, is developed. The hottest channel, a water rod, down comer, upper and lower plenums, feed pumps, etc. are modeled as junction of nodes. Partial of the feed water flows downward from the upper dome of the reactor pressure vessel to the water rods. The accidents analyzed here are total loss of feed water flow, feed water pump seizure, and control rods ejection. All the accidents satisfy the criteria. The accident event at which the maximum cladding temperature is the highest is total loss of feedwater flow. The transients analyzed here are loss of feed water heating, inadvertent start-up of an auxiliary water supply system, partial loss of feed water flow, loss of offsite power, loss of load, and abnormal withdrawal of control rods. All the transients satisfied the criteria. The transient event for which the maximum cladding temperature is the highest is control rod withdrawal at normal operation. The behavior of loss of load transient is different from that of BWR. The power does not increase because loss of flow occurs and the density change is small. The sensitivities of the system behavior to various parameters during transients and accidents are analyzed. The parameters having strong influence are the capacity of the auxiliary water supply system, the coast down time of the main feed water pumps, and the time delay of the main feed water pumps trip. The control rod reactivity also has strong influence. (authors)
International Nuclear Information System (INIS)
Fortov, V.E.
1996-01-01
The proposals on cooperation in the area of thermophysical properties of reactor materials in a broad band of pressure and temperature realized at normal transient and emergency operation activity of nuclear power plants are discussed. 1 fig
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)
Energy Technology Data Exchange (ETDEWEB)
Bae, Seong Jun; Oh, Bongseong; Ahn, Yoonhan; Baik, Seongjoon; Lee, Jekyoung; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)
2016-05-15
It was identified that controlling CO{sub 2} compressor operation near the critical point is one of the most important issues to operate a S-CO{sub 2} Brayton cycle with a high efficiency. Despite the growing interest in the S-CO{sub 2} Brayton cycle, a few previous research on the transient analysis of the S-CO{sub 2} system has been conducted previously. Moreover, previous studies have some limitation in the modelled test facility, and the experiment was not performed to observe specific scenario. The KAIST research team has conducted S-CO{sub 2} system transient experiments with the CO{sub 2} compressing test facility called SCO{sub 2}PE (Supercritical CO{sub 2} Pressurizing Experiment) at KAIST In this study, authors use the transient analysis code GAMMA (Gas Multidimensional Multicomponent mixture Analysis) code for analyzing the experiment. Two transient scenarios were selected in this study; over cooling and under cooling situations. The selected transient situation is of particular interest since the compressor inlet conditions start to drift away from the critical point of CO{sub 2}. The results represent that the GAMMA code can simulate the S-CO{sub 2} test facility, SCO{sub 2}PE. However, as shown in the cooling water flow rate increasing scenario, the GAMMA code shows calculation error when the phase change occurs. Furthermore, although the results of the cooling water flow rate decrease case shows reasonable agreement with the experimental data, there are still some unexplained differences between the experimental data and the GAMMA code prediction.
International Nuclear Information System (INIS)
Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.
2012-01-01
The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)
Transient Torsional Analysis of a Belt Conveyor Drive with Pneumatic Flexible Shaft Coupling
Directory of Open Access Journals (Sweden)
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.
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.
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
Directory of Open Access Journals (Sweden)
Wagner P. B.
2006-11-01
Full Text Available When the drainage area of a flowing well is long and narrow, the pressure transient behavior of the well often cannot be analyzed by conventional means. Well tests which appear to exhibit linear flow behavior for extended periods, beyond the length of time which could be attributed to flow to a fracture, have been observed for both geothermal wells and for oil and gas wells. In geothermal formations and in some of the deep low-permeability gas fields the elongated reservoir shape results from long parallel faults which appear to bound the formations. Typically such wells are also penetrating natural fractures. In other reservoirs the elongated drainage shape is due to the river delta or alluvial environment in which the reservoir sands were originally deposited. Using a recently developed scheme for custom type curve generation for a wide variety of flow geometries, the particular behavior of the well producing from an elongated linear flow system is presented in detail. Cases examined include fractured wells of variable fracture length in varions positions within the flow channel and also of variable length. The effects of partial penetration are also demonstrated. This paper illustrates the use of custom type curves in computer-aided type-curve matching. With the help of relatively inexpensive computer graphics, custom type curves are displayed which exactly represent whatever is known about a particular drainage geometry, including such variables as formation thickness and areal extent, fracture length, partial penetration, and well location. Familles of curves can be generated and displayed along with the data to help determine unknown features of the reservoir geometry. An important contribution included in this work was the development of a set of orthogonal dimensionless parameters which facilitate the generation of familles of type curves which vary one reservoir shape parameter with implied effects on other parameters. The dimensionless
Statistical analysis of silo wall pressures
DEFF Research Database (Denmark)
Ditlevsen, Ove Dalager; Berntsen, Kasper Nikolaj
1998-01-01
Previously published silo wall pressure measurements during plug flow of barley in alarge concrete silo are re-analysed under the hypothesis that the wall pressures are gamma-distributed.The fits of the gamma distribution type to the local pressure data from each measuring cell are satisfactory.......However, the estimated parameters of the gamma distributions turn out to be significantly inhomogeneous overthe silo wall surface. This inhomogeneity is attributed to the geometrical imperfections of the silo wall.Motivated by the engineering importance of the problem a mathematical model for constructing astochastic...... gamma-type continuous pressure field is given. The model obeys the necessary equilibrium conditionsof the wall pressure field and reflects the spatial correlation properties as estimated from simultaneouslymeasured pressures at different locations along a horizontal perimeter....
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.
International Nuclear Information System (INIS)
Osakabe, Kazuya; Onizawa, Kunio; Shibata, Katsuyuki; Kato, Daisuke
2006-09-01
As a part of the aging structural integrity research for LWR components, the probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in JAEA. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics and computer performance. PASCAL Ver.1 has functions of optimized sampling in the stratified Monte Carlo simulation, elastic-plastic fracture criterion of the R6 method, crack growth analysis models for a semi-elliptical crack, recovery of fracture toughness due to thermal annealing and so on. Since then, under the contract between the Ministry of Economy, Trading and Industry of Japan and JAEA, we have continued to develop and introduce new functions into PASCAL Ver.2 such as the evaluation method for an embedded crack, K I database for a semi-elliptical crack considering stress discontinuity at the base/cladding interface, PTS transient database, and others. A generalized analysis method is proposed on the basis of the development of PASCAL Ver.2 and results of sensitivity analyses. Graphical user interface (GUI) including a generalized method as default values has been also developed for PASCAL Ver.2. This report provides the user's manual and theoretical background of PASCAL Ver.2. (author)
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)
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
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.)
Directory of Open Access Journals (Sweden)
Zhi-Juan Pei
2017-12-01
Full Text Available AIM: To study the pathogenesis of transient intraocular pressure(IOPafter laser iridectomy with Krypton laser combined with Q-switched Nd:YAG laser. METHODS: Totally 42 healthy rabbits(84 eyesprovided by the Animal Experimental Center of our hospital were selected, including 18 female rabbits, 24 male rabbits, average weight 2.24±0.31kg, and they were randomly divided into 6 groups, 7 rats in each group(14 eyes. We observed the change of intraocular pressure after laser iridectomy surgery at 20min, 2, 6, 18, 24h and the nitric oxide(NO, malondialdehyde(MDA, superoxide dismutase(SOD, 6-keto-prostaglandin(6-keto-PGF1αand nitric oxide synthase(NOScontent in aqueous. RESULTS: There was no significant difference in intraocular pressure, NO, NOS, SOD, MAD and 6-keto- PGF1α before operation(P>0.05. The intraocular pressure increased after operation, and the difference was statistically significant(PP>0.05. The levels of NO, NOS and SOD in the aqueous humor of the two groups decreased 20min, 2 and 6h after the operation(PP>0.05. The levels of MDA and 6-keto-prostaglandin in the aqueous humor increased after the operation, and the difference was statistically significant at 20min, 2 and 6h after operation(PP>0.05.CONCLUSION: The increase of transient intraocular pressure after laser iridectomy may relate to the increase of malondialdehyde, 6-keto-prostaglandin content and the decrease of superoxide dismutase and nitric oxide in the aqueous humor after operation.
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)
Pressurizer pump reliability analysis high flux isotope reactor
International Nuclear Information System (INIS)
Merryman, L.; Christie, B.
1993-01-01
During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability
Numerical analysis of the transient flow in a scroll refrigeration compressor
Sun, Shuaihui; Wu, Kai; Guo, Pengcheng; Luo, Xingqi
2017-08-01
In the present paper, the CFD technology is adopted to simulate the working process of a scroll refrigeration compressor with R22 as working fluid. The structural grids in the scroll compressor were updated continually during the solving process to cope with the movement boundaries of the fluid domain. The radial meshing clearance was 0.008 mm which was the same with that in the real prototype. The pressure, velocity and temperature distribution in chambers of compressor were computed. Also, the transient mass flux diagrams were calculated out. The results indicated that the pressure was asymmetrical in the two symmetrical suction chambers, because the suction port and passage were not absolutely symmetrical. The gradient of temperature was great in each working chamber due to leakage flow. Velocity vector distribution was asymmetrical in each pair of working chamber owing to the movement of orbiting scroll; the flow was complicated in the central working chamber. The movement of the orbiting scroll had different influence on the vortexes formation in each pair of compression chamber. The inlet and outlet mass flux fluctuated with the crank angle obviously. Because of the ‘cut-off’ of the refrigeration fluid in the suction chamber when the crank angle was larger than 220°, the inlet mass flux decreased remarkably. Finally, some useful advices were given to improve the performance of the scroll refrigeration compressor.
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)
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)
Ceuca, S.C.; Herb, J.; Schoeffel, P.J.; Hollands, T.; Austregesilo, H.; Hristov, H.V.
2017-01-01
The realistic numerical prediction of transient fluid-dynamic scenarios including the complex, three-dimensional flow mixing phenomena occurring in the reactor pressure vessel (RPV) both in normal or abnormal operation are an important issue in today's reactor safety assessment studies. Both Computational Fluid Dynamics (CFD) tools as well as fluid-dynamic system analysis codes, each with its advantages and drawbacks, are commonly used to model such transients. Simulation results obtained with the open-source CFD tool-box OpenFOAM and the German thermal-hydraulic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients), the later developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) for the analysis of the whole spectrum of operational transients, design-basis accidents and beyond design basis accidents anticipated for nuclear energy facilities, are compared against experimental data from the ROssendorf Coolant Mixing (ROCOM) test facility. In the case of the OpenFOAM CFD simulations the influence of various turbulence models and numerical schemes has been assessed while in the case of the system analysis code ATHLET a multidimensional nodalization recommended for real power plant applications has been employed. The simulation results show a good agreement with the experimental data, indicating that both OpenFOAM and ATHLET can capture the key flow features of the mixing processes in the Reactor Pressure Vessel (RPV). (author)
Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor
International Nuclear Information System (INIS)
Krishnan, S.; Bhasin, V.; Mahajan, S.C.
1997-01-01
Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered
Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor
Energy Technology Data Exchange (ETDEWEB)
Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others
1997-04-01
Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.