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)
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).
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
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
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...
Unified fluid flow model for pressure transient analysis in naturally fractured media
International Nuclear Information System (INIS)
Babak, Petro; Azaiez, Jalel
2015-01-01
Naturally fractured reservoirs present special challenges for flow modeling with regards to their internal geometrical structure. The shape and distribution of matrix porous blocks and the geometry of fractures play key roles in the formulation of transient interporosity flow models. Although these models have been formulated for several typical geometries of the fracture networks, they appeared to be very dissimilar for different shapes of matrix blocks, and their analysis presents many technical challenges. The aim of this paper is to derive and analyze a unified approach to transient interporosity flow models for slightly compressible fluids that can be used for any matrix geometry and fracture network. A unified fractional differential transient interporosity flow model is derived using asymptotic analysis for singularly perturbed problems with small parameters arising from the assumption of a much smaller permeability of the matrix blocks compared to that of the fractures. This methodology allowed us to unify existing transient interporosity flow models formulated for different shapes of matrix blocks including bounded matrix blocks, unbounded matrix cylinders with any orthogonal crossection, and matrix slabs. The model is formulated using a fractional order diffusion equation for fluid pressure that involves Caputo derivative of order 1/2 with respect to time. Analysis of the unified fractional derivative model revealed that the surface area-to-volume ratio is the key parameter in the description of the flow through naturally fractured media. Expressions of this parameter are presented for matrix blocks of the same geometrical shape as well as combinations of different shapes with constant and random sizes. Numerical comparisons between the predictions of the unified model and those obtained from existing transient interporosity ones for matrix blocks in the form of slabs, spheres and cylinders are presented for linear, radial and spherical flow types for
Analysis of LOFT pressurizer spray and surge nozzles to include a 4500F step transient
International Nuclear Information System (INIS)
Nitzel, M.E.
1978-01-01
This report presents the analysis of the LOFT pressurizer spray and surge nozzles to include a 450 0 F step thermal transient. Previous analysis performed under subcontract by Basic Technology Incorporated was utilized where applicable. The SAASIII finite element computer program was used to determine stress distributions in the nozzles due to the step transient. Computer results were then incorporated in the necessary additional calculations to ascertain that stress limitations were not exceeded. The results of the analysis indicate that both the spray and surge nozzles will be within stress allowables prescribed by subsubarticle NB-3220 of the 1974 edition of the ASME Boiler and Pressure Vessel Code when subjected to currently known design, normal operating, upset, emergency, and faulted condition loads
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.
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)
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
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.
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
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.
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
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)
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
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
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)
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
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
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
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
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.
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.
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 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
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
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.
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
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)
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
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.
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)
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
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.
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
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
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
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.
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)
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
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)
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
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
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
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.
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.
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
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
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
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
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)
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
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.
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
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.
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)
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
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.
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
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.
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)
Transient response of a five-region nonequilibrium real-time pressurizer model
International Nuclear Information System (INIS)
Fakory, M.R.; Seifaee, F.
1987-01-01
Recent accidents at nuclear power plants in the US and abroad have prompted accurate analysis and simulation of the plant systems and the training of reactor operators on plant-specific simulators that are equipped with the simulation models. Consequently, several models for real-time and off-time simulation of nuclear reactor systems, with various levels of accuracy and simulation fidelity, have been introduced. Experience with power plant simulation demonstrates that in order to realistically predict and simulate reactor responses during unanticipated transients, it is necessary to equip the simulation model with a multielement pressurizer model. The objective of this paper is to present the results of a five-region drift-flux-based pressurizer model, which has been developed for integration with real-time training simulators. A comparison between the plant data and the results of the nonequilibrium pressurizer model demonstrates that the model is well capable of close simulation of dynamic behavior of the pressurizer system
The 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
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
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
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
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)
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
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.
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
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
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
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
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
Modelling and transient simulation of water flow in pipelines using WANDA Transient software
Directory of Open Access Journals (Sweden)
P.U. Akpan
2017-09-01
Full Text Available Pressure transients in conduits such as pipelines are unsteady flow conditions caused by a sudden change in the flow velocity. These conditions might cause damage to the pipelines and its fittings if the extreme pressure (high or low is experienced within the pipeline. In order to avoid this occurrence, engineers usually carry out pressure transient analysis in the hydraulic design phase of pipeline network systems. Modelling and simulation of transients in pipelines is an acceptable and cost effective method of assessing this problem and finding technical solutions. This research predicts the pressure surge for different flow conditions in two different pipeline systems using WANDA Transient simulation software. Computer models were set-up in WANDA Transient for two different systems namely; the Graze experiment (miniature system and a simple main water riser system based on some initial laboratory data and system parameters. The initial laboratory data and system parameters were used for all the simulations. Results obtained from the computer model simulations compared favourably with the experimental results at Polytropic index of 1.2.
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
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.
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.
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)
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.)
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
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
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 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.
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.
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
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
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)
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
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
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
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)
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
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
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
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 ...
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.)
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
Transient thermal-hydraulic characteristics analysis software for PWR nuclear power systems
International Nuclear Information System (INIS)
Wu Yingwei; Zhuang Chengjun; Su Guanghui; Qiu Suizheng
2010-01-01
A point reactor neutron kinetics model, a two-phase drift-flow U-tube steam generator model, an advanced non-equilibrium three regions pressurizer model, and a passive emergency core decay heat-removed system model are adopted in the paper to develop the computerized analysis code for PWR transient thermal-hydraulic characteristics, by Compaq Visual Fortran 6.0 language. Visual input, real-time processing and dynamic visualization output are achieved by Microsoft Visual Studio. NET language. The reliability verification of the soft has been conducted by RELAP 5, and the verification results show that the software is with high calculation precision, high calculation speed, modern interface, luxuriant functions and strong operability. The software was applied to calculate the transient accident conditions for QSNP, and the analysis results are significant to the practical engineering applications. (authors)
Transient 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
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
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
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)
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
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.)
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
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
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)
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)
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
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
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.
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.)
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.
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)
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
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)
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.
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.
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.
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
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)
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
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
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)
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.
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
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
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.
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
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...
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
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
Analysis of the Mannshan Unit 2 full load rejection transient
International Nuclear Information System (INIS)
Kang, J.C.; Pei, B.S.; Yu, G.P.; Yuann, R.Y.
1987-01-01
Mannshan Unit 2 is a Westinghouse three-loop pressurized water reactor with a rated core power of 2775 MW(thermal) and a rated core flow of 4702 kg/s. Before full power operation, a planned net load rejection was performed during the startup test by opening the main transformer highside breakers. The generator power rapidly reduced to station load. All 16 steam dump valves immediately popped open, and control bank-D rods automatically stepped in as the temperature difference T/sub avg/ - T/sub ref/ reached a programmed 2.8 0 C. Nuclear power decreased smoothly as control rods were inserted into the core. The pressurizer pressure and liquid levels also dropped. Neither safety injection nor reactor trip occurred during this transient. The test was done to verify that the whole system would function properly under a transient to keep the reactor from scramming and that the vessel integrity would also be protected. In this study, which is the preliminary stage of RELAP5/MOD2 transient simulation of the Mannshan PWR plants, system thermal-hydraulic response is tested first and isolated from the neutronic effects. The variation of core power versus time curve was extracted from the power test data to serve as a time varying boundary condition. The comparison of the analytical results of four major parameters (pressurizer pressure, average temperature of the core, steam dump flow rate, and feedwater flow rate) from RELAP5/MOD2 and the power test data is illustrated
Transient analysis 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 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
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
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)
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.
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
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
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.
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.
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
PWR plant transient analyses using TRAC-PF1
International Nuclear Information System (INIS)
Ireland, J.R.; Boyack, B.E.
1984-01-01
This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public
Thermal stratification in the pressurizer
International Nuclear Information System (INIS)
Baik, S.J.; Lee, K.W.; Ro, T.S.
2001-01-01
The thermal stratification in the pressurizer due to the insurge from the hot leg to the pressurizer has been studied. The insurge flow of the cold water into the pressurizer takes place during the heatup/cooldown and the normal or abnormal transients during power operation. The pressurizer vessel can undergo significant thermal fatigue usage caused by insurges and outsurges. Two-dimensional axisymmetric transient analysis for the thermal stratification in the pressurizer is performed using the computational fluid dynamics code, FLUENT, to get the velocity and temperature distribution. Parametric study has been carried out to investigate the effect of the inlet velocity and the temperature difference between the hot leg and the pressurizer on the thermal stratification. The results show that the insurge flow of cold water into the pressurizer does not mix well with hot water, and the cold water remains only in the lower portion of the pressurizer, which leads to the thermal stratification in the pressurizer. The thermal load on the pressurizer due to the thermal stratification or the cyclic thermal transient should be examined with respect to the mechanical integrity and this study can serve the design data for the stress analysis. (authors)
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.
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.
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
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.)
Oxide fuel pin transient performance analysis and design with the TEMECH code
International Nuclear Information System (INIS)
Bard, F.E.; Dutt, S.P.; Hinman, C.A.; Hunter, C.W.; Pitner, A.L.
1986-01-01
The TEMECH code is a fast-running, thermal-mechanical-hydraulic, analytical program used to evaluate the transient performance of LMR oxide fuel pins. The code calculates pin deformation and failure probability due to fuel-cladding differential thermal expansion, expansion of fuel upon melting, and fission gas pressurization. The mechanistic fuel model in the code accounts for fuel cracking, crack closure, porosity decrease, and the temperature dependence of fuel creep through the course of the transient. Modeling emphasis has been placed on results obtained from Fuel Cladding Transient Test (FCTT) testing, Transient Fuel Deformation (TFD) tests and TREAT integral fuel pin experiments
Development of 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
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)
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)
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
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)
Pressure-temperature response of a full-pressure PWR containment to a loss-of-coolant accident
International Nuclear Information System (INIS)
Misak, J.
1976-01-01
A mathematical model and computer code TRACO III for pressure-temperature transients in the full-pressure containment of PWR during LOCA is described. Main attention is devoted to the analysis of parametric calculations with respect to the estimation of effect of various factors on the transient process and to the comparison of the theoretical and the experimental results on CVTR. (author)
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)
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].
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
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
Status report for anticipated transients without scram for Combustion Engineering reactors
International Nuclear Information System (INIS)
1975-01-01
The NRC staff review of Combustion ATWS analyses included the anticipated transients expected to occur, the initial conditions and system parameters assumed in the analyses, the reliability of systems, the analytical techniques, the results of transient analysis of ATWS events and the design of the Reactor Protection System. Using the requirements of WASH-1270 as a guideline, the staff reviewed each relevant aspect of the Combustion model and analysis. The discussion of anticipated transients is presented, and the initial conditions, system parameters, and operating systems assumed in the analyses of these transients are discussed. The analytical techniques and computer programs are reviewed. An independent calculation conducted by the staff using the RELAP-3B code to determine the pressure within the reactor coolant pressure boundary during a complete loss of main feedwater ATWS event is described. A set of standard problems is defined for all pressurized water reactor vendors and the Regulatory staff to insure acceptability of computer codes used in all systems transient analyses. The model for calculating water discharge through primary valves is described. The comparison of the Combustion analyses to the requirements of WASH-1270 is presented. Certain outstanding issues are identified which require that Combustion or the applicant provide additional information or modify existing designs
Discussion on Design Transients of Pebble-bed High Temperature Gas-cooled Reactor
International Nuclear Information System (INIS)
Wang Yan; Li Fu; Zheng Yanhua
2014-01-01
In order to assure high quality for the components and their supports in the reactor coolant system, etc., some thermal-hydraulic transient conditions will be selected and researched for equipment design evaluation to satisfy the requirements ASME code, which are based on the conservative estimates of the magnitude and frequency of the temperature and pressure transients resulting from various operating conditions in the plant. In the mature design on pressurized water reactor, five conditions are considered. For the developing advanced pebble-bed high temperature gas-cooled reactor(HTGR), its design and operation has much difference with other reactors, so the transients of the pebble-bed high temperature gas-cooled reactor have distinctive characteristics. In this paper, the possible design transients of the pebble-bed HTGR will be discussed, and the frequency of design transients for equipment fatigue analysis and stress analysis due to cyclic stresses is also studied. The results will provide support for the design and construct of the pebble-bed HTGR. (author)
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)
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
Opportunities for practical improvements in the management of plant transients
International Nuclear Information System (INIS)
Zebroski, E.L.
1984-01-01
This chapter attempts to provide some perspectives on the steps involved in analyzing, evaluating, and implementing remedies for transients and for potentially severe events. The importance of improved response and control of plant transients is stressed. The main steps involved in the attainment of improved control of plant transients are listed. Topics considered include the acquisition of plant data, sensitivity and risk analysis, the options for improvements, the managerial role, and some priorities for data, analysis, and evaluation. The ten most frequent types of transients for pressurized water reactors (PWRs) and boiling water reactors (BWRs) are listed according to frequency of occurrence. It is concluded that the two main needs of transient management are to avoid preoccupation with end-of-spectrum accidents and to improve the rate of technology transfer from best-available analysis and implementation
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.
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)
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
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.
Directory of Open Access Journals (Sweden)
Antoine Bruneau
Full Text Available Decreased arterial oxygen pressure obtained at peak exercise is strong evidence of walking-induced hypoxemia, assuming that the lower pressure occurs just before exercise is stopped. Using empirical predefined models and transcutaneous oximetry, we have shown that some patients reporting exercise intolerance show a minimal value at the onset of walking and a post-exercise overshoot. These changes are referred to as transcutaneous "walking-induced transient hacks".In 245 patients, walking-induced transcutaneous oxygen pressure changes in the chest were analyzed using observer-independent clustering techniques. Clustering classes were compared to the profile types previously proposed with the cross-correlation technique. The classifications of patients according to both approaches were compared using kappa statistics. In 10 patients showing a hack on transcutaneous oximetry, we analyzed the results of direct iterative arterial sampling recorded during a new walking treadmill test.Clustering analysis resulted in 4 classes that closely fit the 4 most frequently proposed empirical models (cross-correlation coefficients: 0.93 to 0.97. The kappa between the two classifications was 0.865. In 10 patients showing transcutaneous hacks, the minimal direct arterial oxygen pressure value occurred at exercise onset, and these patients exhibited a recovery overshoot reaching a maximum at two minutes of recovery, confirming the walking-induced transient hypoxemia.In patients reporting exercise intolerance, transcutaneous oximetry could help to detect walking-induced transient hypoxemia, while peak-exercise arterial oximetry might be normal.
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.
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.)
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
Simulation of the Three Mile Island transient in Semiscale
International Nuclear Information System (INIS)
Larson, T.K.; Loomis, G.G.; Shumway, R.W.
1979-07-01
This report presents the results of a preliminary review and analysis of the data obtained from eight simulations of the Three Mile Island Unit 2 Nuclear Power Generating Station transient (March 28, 1979) that have been conducted in the Semiscale Mod-3 System. The Semiscale simulations of the Three Mile Island (TMI) transient were basically conducted from the same sequence of events as those recorded in the plant. System initial conditions representative of those in the TMI system were established and the transient was initiated by terminating steam generator feedwater and steam valve flow. The steam generator secondaries were drained to control primary to secondary heat transfer. The pressurizer power operated relief valve, pressurizer code safety valve, and core power trip were operated on system pressure. High pressure safety injection was activated for about one minute during the Semiscale simulations. In addition, both primary loop coolant pumps were shut off in the Semiscale simulation at the same time that the Three Mile Island loop 2A pump was shut off
PRACTICAL IMPLICATIONS OF USING INDUCED TRANSIENTS FOR LEAK DETECTION
Directory of Open Access Journals (Sweden)
Marko V. Ivetic
2007-06-01
Full Text Available This paper deals with practical problems of leak detection by methods based on hydraulic transient analysis. Controlled and safe transients can be generated and the response of the network, with the relevant information, can be monitored and analysed. Information about leaks, contained in the monitored pressure signal, cannot be easily retrieved, due to reflections, noise etc. On the basis of numerical experiments on a simple network, merits and limitations of several methods for signal analysis (time domain analysis, spectral density function and wavelet transform have been examined. Certain amount of information can be extracted from the time history of the pressure signal, assuming the first reflection of the pressure wave is captured with very high time resolution and accuracy. Only relatively large leaks can be detected using this methodology. As a way to increase the sensitivity of this method it is suggested that transforms in frequency domain and, especially, wavelet transforms, are used. The most promising method for leakage location and quantification seems to be based on wavelet analysis.
PRACTICAL IMPLICATIONS OF USING INDUCED TRANSIENTS FOR LEAK DETECTION
Directory of Open Access Journals (Sweden)
Marko V. Ivetic
2007-01-01
Full Text Available This paper deals with practical problems of leak detection by methods based on hydraulic transient analysis. Controlled and safe transients can be generated and the response of the network, with the relevant information, can be monitored and analysed. Information about leaks, contained in the monitored pressure signal, cannot be easily retrieved, due to reflections, noise etc. On the basis of numerical experiments on a simple network, merits and limitations of several methods for signal analysis (time domain analysis, spectral density function and wavelet transform have been examined. Certain amount of information can be extracted from the time history of the pressure signal, assuming the first reflection of the pressure wave is captured with very high time resolution and accuracy. Only relatively large leaks can be detected using this methodology. As a way to increase the sensitivity of this method it is suggested that transforms in frequency domain and, especially, wavelet transforms, are used. The most promising method for leakage location and quantification seems to be based on wavelet analysis.
Fundamental study on thermo-hydraulic behaviors during power transient, 2
International Nuclear Information System (INIS)
Shinano, M.; Inoue, A.
1988-01-01
Thermo-hydraulic behaviors during power transient of nuclear reactors are studied. Boiling around test rod heated transiently forces to flow out liquid in the test section and generates high pressure pulse. In this study, it is investigated experimentally and analytically that magnitude of pressure pulse and energy conversion efficiency to the mechanical works in cases of fragmentation and non-fragmentation. In analysis, effects of increasing of heat transfer and of interaction area due to fragmentation is considered. Consequently, 1) magnitude of pressure pulse on fragmentation is about 10 times greater than that on non-fragmentation. 2) analytical model can show characteristics of fragmentation processes qualitatively. (author)
International Nuclear Information System (INIS)
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)
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.
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
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
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
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
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
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
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.)
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
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
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
TRAC analyses of severe overcooling transients for the Oconee-1 PWR
Energy Technology Data Exchange (ETDEWEB)
Ireland, J R [comp.
1985-05-01
This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs.
TRAC analyses of severe overcooling transients for the Oconee-1 PWR
International Nuclear Information System (INIS)
Ireland, J.R.
1985-05-01
This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs
The economic impact of reactor transients
International Nuclear Information System (INIS)
Rossin, A.D.; Vine, G.L.
1984-01-01
This chapter discusses the cost estimation of transients and the causal relationship between transients and accidents. It is suggested that the calculation of the actual cost of a transient that has occurred is impossible without computerized records. Six months of operating experience reports, based on a survey of pressurized water reactors (PWRs) and boiling water reactors (BWRs) conducted by the Nuclear Safety Analysis Center (NSAC), are analyzed. The significant costs of a reactor transient are the repair costs resulting from severe damage to plant equipment, the cost of scrams (the actions the system is designed to take to avoid safety risks), US NRC fines, negative publicity, utility rates and revenues. It is estimated that the Three Mile Island-2 accident cost the US over $100 billion in nuclear plant delays and cancellations, more expensive fuel, oil imports, backfits, bureaucratic, administrative and legal costs, and lost productivity
The development of a transient neutron flux solution in the PANTHER code
International Nuclear Information System (INIS)
Hutt, P.K.; Knight, M.P.
1990-01-01
In the United Kingdom a new three-dimensional, two-group, homogeneous reactor diffusion code, PANTHER, has been developed for the analysis of pressurized water reactors (PWRs) and advanced gas-cooled reactors (AGRs). The code can perform a comprehensive range of calculations, steady state, depletion, and transient with either a finite difference or analytic nodal flux solution. The nodal solution allows the representation of within-node burnup variation and pin-power reconstruction in either steady-state or transient mode. Specific steady-state and transient thermal feedback modules are included for both PWRs and AGRs. The code is being developed to perform a complete range of reactor calculations from online operational support to fuel management and fault transient analysis. In the area of transient analysis, the code is currently being used for a number of PWR fault transient assessments, including rod ejection and steam-line break. In addition, work is proceeding to incorporate the PANTHER 3D nodal transient solution in the TRAC-P code. This paper outlines the development of the transient flux solutions within PANTHER
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)
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
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.
Directory of Open Access Journals (Sweden)
Fuxiang Zhang
2015-03-01
Full Text Available Addressing to the deteriorated load conditions of working string and packers caused by annular pressure drop between packers during the staged stimulation of high-pressure deep well, one 2D temperature field transient prediction model for borehole under injecting conditions which considers such influences as friction heat, convection heat exchange was set up, based on energy conservation principle and borehole heat transfer theory. By means of analyzing the influences of borehole temperature and pressure changes on the annular volume between packers, and in combination with borehole temperature transient prediction model, annular fluid PVT equations of state, radial deformation model of tubing and formation transient seepage equation, a typical high-pressure deep well inter-packer annular pressure prediction model was established. Taking a high-pressure gas well in Tarim Oilfield for example, the inter-packer annular pressure prediction was conducted, on which, the mechanical analysis on packers and working strings was carried out. The analysis results show that although the pipe string is safe in the viewpoint of conventional design methods, it is still susceptible to failure after the annular pressure drop between packers was taken into consideration. Such factor should be fully considered in the design of staged stimulation pipe strings, and this prediction model provides new thoughts for the optimal design of high-pressure deep well staged stimulation pipe strings.
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
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
Loss-of-feedwater transients in PWRs
International Nuclear Information System (INIS)
Burns, R.D. III.
1980-01-01
Recent severe accident sequence analysis (SASA) work in LASL's Multifault Accident Analysis Section has focused on loss-of-feedwater (LOFW) transients at a 4-loop Westinghouse nuclear power reactor. In all transients studied, the initiator was loss of main feedwater and reactor coolant pump (RCP) trip, caused by temporary loss of off-site power. Subsequent automatic actions included reactor scram, closure of the main steam isolation valves, and initiation of auxiliary feedwater (AFW) flow. TRAC-PD2 calculations were designed to study the consequences of AFW delivery rates below the minimum specified in the emergency operating procedures (EOPs) for the reference 4-loop plant. Six types of LOFW scenarios have been studied, including (1) zero AFW availability (nominal case), (2) initially zero AFW but full recovery after 2 h, (3) zero AFW with steam generator (SG) atmospheric relief valve (ARV) malfunction, (4) zero AFW with high pressure charging flow initiated after 2 h, and (5) zero AFW with delay in reactor scram. Additional cases were considered to study the effects of uncertainties in pressurizer heater/spray operation, operator manual initiation of high pressure charging flow, reactor initial conditions, and RCP and power coastdown characteristics. Nominal case results, rationale for selections of other cases, and lessons learned are summarized
A follow-up of transients. Stage 1
International Nuclear Information System (INIS)
Enekull, Aa.; Wallner, B.
1981-09-01
A follow-up of the transients of temperature and pressure in the primary pressurized system of a nuclear power plant has been completed for the Barsebaeck-1 reactor. The investigation consists of the following steps:- the collation of transients - drawing up load data based on transients-analyses of stress - recommendations for future programs. It was found that the lifetime of the system will exceed 40 years excluding some of the pipes for feed water. The appendices give a detailed description of the transients.(G.B.)
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.
Design criteria of integrated reactors based on transients
International Nuclear Information System (INIS)
Zanocco, P.; Gimenez, M.; Delmastro, D.
1999-01-01
A new tendency in integrated reactors conceptual design is to include safety criteria through accident analysis. In this work, the effect of design parameters in a Loss of Heat Sink transient using design maps is analyzed. Particularly, geometry related parameters and reactivity coefficients are studied. Also the effect of primary relief/safety valve during the transient is evaluated. A design map for valve area vs. coolant density reactivity coefficient is obtained. A computer code (HUARPE) is developed in order to simulate these transients. Coolant, steam dome, pressure vessel structures and core models are implemented. This code is checked against TRAC with satisfactory results. (author)
LWR fuel performance during anticipated transients with scram
International Nuclear Information System (INIS)
Martinson, Z.R.; McCardell, R.K.; MacDonanl, P.E.; Rowland, T.C.; Tokar, M.
1983-01-01
Operational transients occur occasionally in light water reactors when minor malfunctions of certain system components affect the reactor core. Potential effects of such malfunctions include a loss of the secondary heat sink, an increase in system pressure, and, in boiling water reactors, void collapse and a brief increase in reactor power. The most severe postulated Boiling Water Reactor (BWR) anticipated transient is characterized by a power peak of up to 495% rated power for about 1 second (according to a recent General Electric Co., generic analysis). The results of a series of fuel behaviour tests in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory are presented in this paper. Four progressively higher and broader power transients at a constant coolant flow rate were performed. The first transient simulated a BWR-5 turbine trip without steam bypass with fuel rods operating at BWR-6 core average rod powers. The second transient simulated a generator load rejection without steam bypass with fuel rods operating at above core average powers. The last two transients were performed at higher powers than safety analysis predicts to be possible in commercial reactors to be defined failure threshold margins. The test rods did not fail and were not damaged during any of the four transients. (author)
Condensation effects in a pressurizer scaled from a pressurized water reactor
International Nuclear Information System (INIS)
Loomis, G.G.; Shaw, R.A.
1985-01-01
This paper presents results from an experimental investigation of phenomena associated with pressurizer auxiliary spray during an abnormal plant transient in a commercial PWR. If normal pressurizer spray is unavailable (main coolant pumps are off) or the pressurizer power operated relief valve cannot be used during abnormal transients, pressurizer auxiliary spray can be used to reduce primary system pressure. Results from both transient integral experiments involving pressurizer auxiliary spray during tube rupture and separate effects spray experiments are presented. The experimental investigation was conducted in the Semiscale MOD-2B facility. Phenomenon of interest that occurred in the pressurizer during the pressurized auxiliary spray was desuperheating of the pressurizer steam space and quenching of metal walls followed by dropwise condensation of the pressurizer steam. The data from both the transient integral experiments and the separate effects experiments are compared to RELAP5 computer calculations and the capability of existing models in the code is discussed
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.
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.
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.
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.
Measurement of transient hydrodynamic characteristics of the reactor RA primary cooling system
International Nuclear Information System (INIS)
Jovic, L.; Majstorovic, D.; Zeljkovic, I.
1987-01-01
Experimental study of transient hydrodynamic characteristics of the research nuclear reactor RA by simultaneous measurements of fluid flow and pressure on several locations of the RA primary coolant system is done. Loss of electric power transient on the main circulation pumps is simulated. measurement methodology, data processing and results of measured data analysis are given. (author)
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
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.
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
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
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
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
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.
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
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)
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)
Transient Seepage for Levee Engineering Analyses
Tracy, F. T.
2017-12-01
Historically, steady-state seepage analyses have been a key tool for designing levees by practicing engineers. However, with the advances in computer modeling, transient seepage analysis has become a potentially viable tool. A complication is that the levees usually have partially saturated flow, and this is significantly more complicated in transient flow. This poster illustrates four elements of our research in partially saturated flow relating to the use of transient seepage for levee design: (1) a comparison of results from SEEP2D, SEEP/W, and SLIDE for a generic levee cross section common to the southeastern United States; (2) the results of a sensitivity study of varying saturated hydraulic conductivity, the volumetric water content function (as represented by van Genuchten), and volumetric compressibility; (3) a comparison of when soils do and do not exhibit hysteresis, and (4) a description of proper and improper use of transient seepage in levee design. The variables considered for the sensitivity and hysteresis studies are pore pressure beneath the confining layer at the toe, the flow rate through the levee system, and a levee saturation coefficient varying between 0 and 1. Getting results for SEEP2D, SEEP/W, and SLIDE to match proved more difficult than expected. After some effort, the results matched reasonably well. Differences in results were caused by various factors, including bugs, different finite element meshes, different numerical formulations of the system of nonlinear equations to be solved, and differences in convergence criteria. Varying volumetric compressibility affected the above test variables the most. The levee saturation coefficient was most affected by the use of hysteresis. The improper use of pore pressures from a transient finite element seepage solution imported into a slope stability computation was found to be the most grievous mistake in using transient seepage in the design of levees.
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
EP1000 anticipated transient without scram analyses
International Nuclear Information System (INIS)
Saiu, G.; Frogheri, M.; Schulz, T.L.
2001-01-01
The present paper summarizes the main results of the Anticipated Transient Without Scram (ATWS) analysis activity, performed for the European Passive Plant Program (EPP). The behavior of the EP1000 plant following an ATWS has been analyzed by means of the RELAP5/Mod3.2 code. An ATWS is defined as an Anticipated Transient accompanied by a common mode failure in the reactor protection system, such that the control rods do not scram as required to mitigate the consequences of the transient. According to the experience gained in PWR design, the limiting ATWS events, in a PWR, have been found to be the heatup transients caused by a reduction of heat removal capability by the secondary side of the plant. For this reason, the Loss of Normal Feedwater initiating event, to which the failure of the reactor scram is associated, has been analyzed. The purpose of the study is to verify the performance requirements set for the core feedback characteristics (that is to evaluate the effect of the low boron core neutron kinetic parameters), the overpressure protection system, and boration systems to cope with the EUR Acceptance Criteria for ATWS. Another purpose of this analysis was to support development of revised PSA success criteria that would reduce the contribution of ATWS to the large release frequency (LRF). The low boron core improved the basic EP1000 response to an ATWS event. In particular, the peak pressure was significantly lower than that which would result from a standard core configuration. The improved ATWS analysis results also permitted improved ATWS PSA success criteria. For example, the reduced peak pressure allows the use of other plant features to mitigate the event, including manual initiation of feed-bleed cooling in the event of PRHR HX failure. As a result, the core melt frequency and especially the LRF are significantly reduced. (author)
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
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
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
Directory of Open Access Journals (Sweden)
Itissam ABUIZIAH
2014-03-01
Full Text Available When transient conditions (water hammer exist, the life expectancy of the system can be adversely impacted, resulting in pump and valve failures and catastrophic pipe rupture. Hence, transient control has become an essential requirement for ensuring safe operation of water pipeline systems. To protect the pipeline systems from transient effects, an accurate analysis and suitable protection devices should be used. This paper presents the problem of modeling and simulation of transient phenomena in hydraulic systems based on the characteristics method. Also, it provides the influence of using the protection devices to control the adverse effects due to excessive and low pressure occuring in the transient. We applied this model for two main pipeline systems: Valve and pump combined with a simple surge tank connected to reservoir. The results obtained by using this model indicate that the model is an efficient tool for water hammer analysis. Moreover, using a simple surge tank reduces the unfavorable effects of transients by reducing pressure fluctuations.
Flow transients experiments with refrigerant-12
International Nuclear Information System (INIS)
Celata, G.P.; D'Annibale, F.; Farello, G.E.; Setaro, T.
1986-01-01
Flow transients have been investigated in a wide range of thermal-hydraulics situations with Refrigerannt-12. Six pressures (including the reference to PWR and BWR characteristic liquid to vapour densities ratios), several periods of the flowrate transients coastdown during the simulated flow decays, and different specific mass flowrate have been studied emploiyng a circular duct test section (Dsub(i)=7,5 mm). Two heated lengths of the test section have been considered (L = 2300 and 1180 mm). Experimental data have shown the complete inadequacy of steady-state critical heat flux correlations in predicting the onset of boiling crisis during fast flow transients (half-flow decay time, tsub(h)lt5.0-6.0 s). The flow transient does not show dependence, in terms of DNB conditions ,upon the length of the test section: the ratio between transient and steady-state critical mass flowrate is not dependent on the tested geometry. The time interval from the start of the flowrate transient to the onset of DNB (time to crisis), has been experimentally determined for all the runs. Data analysis for a better theoretical prediction of the phenomenon has been accomplished, and a design correlation for DNB conditons and time to crisis prediction has been proposed
Shi, Shanbin
The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal
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)
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
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.
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
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
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)
Energy Technology Data Exchange (ETDEWEB)
Ishii, Mamoru [Purdue Univ., West Lafayette, IN (United State
2016-11-30
The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup
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
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)
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
Transient leak detection in crude oil pipelines
Energy Technology Data Exchange (ETDEWEB)
Beushausen, R.; Tornow, S.; Borchers, H. [Nord-West Oelleitung, Wilhelmshaven (Germany); Murphy, K.; Zhang, J. [Atmos International Ltd., Manchester (United Kingdom)
2004-07-01
Nord-West Oelleitung (NWO) operates 2 crude oil pipelines from Wilhemshaven to Koln and Hamburg respectively. German regulations for transporting flammable substances stipulate that 2 independent continuously working procedures be used to detect leaks. Leak detection pigs are used routinely to complement the surveillance system. This paper described the specific issues of transient leak detection in crude oil pipelines. It was noted that traditional methods have failed to detect leaks that occur immediately after pumps are turned on or off because the pressure wave generated by the transient dominates the pressure wave that results from the leak. Frequent operational changes in a pipeline are often accompanied by an increased number of false alarms and failure to detect leaks due to unsteady operations. NWO therefore decided to have the Atmos statistical pipeline leak detection (SPLD) system installed on their pipelines. The key to the SPLD system is the sequential probability ratio test. Comprehensive data validation is performed following reception of pipeline data from the supervisory control and data acquisition (SCADA) system. The validated data is then used to calculate the corrected flow imbalance, which is fed into the SPRT to determine if there is an increase in the flow imbalance. Pattern recognition is then used to distinguish a leak from operational changes. The SPLD is unique because it uses 3 computational pipeline monitoring methods simultaneously, namely modified volume balance, statistical analysis, and pressure and flow monitoring. The successful installation and testing of the SPLD in 2 crude oil pipelines was described along with the main difficulties associated with transient leaks. Field results were presented for both steady-state and transient conditions. 5 refs., 2 tabs., 16 figs.
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.
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
ATHENA simulations of divertor loss of heat sink transient for the GSSR - Final report with updates
Energy Technology Data Exchange (ETDEWEB)
Sponton, L.L
2001-05-01
The ITER-FEAT Generic Site Safety Report includes evaluations of the consequences of various types of conceivable transients that can occur during operation. The transients that have to be considered in this respect are specified in the Accident Analysis Specifications document of the safety report. For the divertor primary heat transport system the ranges of transients include amongst others a loss of heat sink at full fusion power operation. The thermal-hydraulic consequences related to the coolability of the divertor primary heat transport system components for this transient have been evaluated and summarised in the safety report and in the current report an overview of those efforts and associated outcome is provided. The analyses have been made with the ATHENA thermal-hydraulic code using a separately developed ATHENA model of the ITER-FEAT divertor cooling system. In the current report results from calculations with an updated pressurizer model and pressurizer control system are outlined. The results show that the pressurizer safety valve does not open, that the pressurizer level increase is moderate and that no temperature increases jeopardize the structure integrity.
ATHENA simulations of divertor loss of heat sink transient for the GSSR - Final report with updates
International Nuclear Information System (INIS)
Sponton, L.L.
2001-05-01
The ITER-FEAT Generic Site Safety Report includes evaluations of the consequences of various types of conceivable transients that can occur during operation. The transients that have to be considered in this respect are specified in the Accident Analysis Specifications document of the safety report. For the divertor primary heat transport system the ranges of transients include amongst others a loss of heat sink at full fusion power operation. The thermal-hydraulic consequences related to the coolability of the divertor primary heat transport system components for this transient have been evaluated and summarised in the safety report and in the current report an overview of those efforts and associated outcome is provided. The analyses have been made with the ATHENA thermal-hydraulic code using a separately developed ATHENA model of the ITER-FEAT divertor cooling system. In the current report results from calculations with an updated pressurizer model and pressurizer control system are outlined. The results show that the pressurizer safety valve does not open, that the pressurizer level increase is moderate and that no temperature increases jeopardize the structure integrity
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
Mitigation method of thermal transient stress by thermalhydraulic-structure total analysis
International Nuclear Information System (INIS)
Kasahara, Naoto; Jinbo, Masakazu; Hosogai, Hiromi
2003-01-01
This study proposes a rational evaluation and mitigation method of thermal transient loads in fast reactor components by utilizing relationships among plant system parameters and stresses induced by thermal transients of plants. A thermalhydraulic-structure total analysis procedure helps us to grasp relationship among system parameters and thermal stresses. Furthermore, it enables mitigation of thermal transient loads by adjusting system parameters. In order to overcome huge computations, a thermalhydraulic-structure total analysis code and the Design of Experiments methodology are utilized. The efficiency of the proposed mitigation method is validated through thermal stress evaluation of an intermediate heat exchanger in Japanese demonstration fast reactor. (author)
Transient 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)
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.
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
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.)
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
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...
Computational scheme for transient temperature distribution in PWR vessel wall
International Nuclear Information System (INIS)
Dedovic, S.; Ristic, P.
1980-01-01
Computer code TEMPNES is a part of joint effort made in Gosa Industries in achieving the technique for structural analysis of heavy pressure vessels. Transient heat conduction problems analysis is based on finite element discretization of structures non-linear transient matrix formulation and time integration scheme as developed by Wilson (step-by-step procedure). Convection boundary conditions and the effect of heat generation due to radioactive radiation are both considered. The computation of transient temperature distributions in reactor vessel wall when the water temperature suddenly drops as a consequence of reactor cooling pump failure is presented. The vessel is treated as as axisymmetric body of revolution. The program has two finite time element options a) fixed predetermined increment and; b) an automatically optimized time increment for each step dependent on the rate of change of the nodal temperatures. (author)
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)
Improving MODPRESS heat loss calculations for PWR pressurizers
International Nuclear Information System (INIS)
Ramos, Natalia V.; Lira, Carlos A. Brayner O.; Castrillho, Lazara S.
2009-01-01
The improvement of heat loss calculations in MODPRESS transient code for PWR pressurizer analysis is the main focus of this investigation. Initially, a heat loss model was built based on heat transfer coefficient (HTC) correlations obtained in handbooks of thermal engineering. A hand calculation for Neptunus experimental test number U47 yielded a thermal power loss of 11.2 kW against 17.3 kW given by MODPRESS at the same conditions, while the experimental estimate is given as 17 kW. This comparison is valid only for steady state or before starting the transient experiment, because MODPRESS does not update HTC's when the transient phase begins. Furthermore, it must be noted that MODPRESS heat transfer coefficients are adjusted to reproduce the experimental value of the specific type of pressurizer. After inserting the new routine for HTC's into MODPRESS, the heat loss was calculated as 11.4 kW, a value very close to the first estimate but far below 17 kW found in the U47 experiment. In this paper, the heat loss model and results will be described. Further research is being developed to find a more general HTC that allows the analysis of the effects of heat losses on transient behavior of Neptunus and IRIS pressurizers. (author)
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
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
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.
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
Energy Technology Data Exchange (ETDEWEB)
Mohanty, Subhasish, E-mail: smohanty@anl.gov; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti
2016-12-15
Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.
International Nuclear Information System (INIS)
Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti
2016-01-01
Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.
Pressurized thermal shock program sponsored by EPRI
International Nuclear Information System (INIS)
Stahlkopf, K.E.
1983-01-01
The potential for long term neutron embrittlement of reactor vessels has been recognized for a number of years. Reactor vessel thermal shock is not a new concern, but with a growing number of plants approaching their mid-lives, it is a concern that must be understood and dealt with. Recent attention has focused on the performance of vessels during overcooling transients. This concern was designated as Unresolved Safety Issue A-49 by the Nuclear Regulatory Commission in December 1981. The USNRC staff has identified eight overcooling events of concern in U.S. PWRs. The concern is currently limited to Pressurized Water Reactors. The Electric Power Research Institute (EPRI) has supported research on reactor vessel integrity for a number of years and has supported an extensive effort on reactor vessel pressurized thermal shock (PTS) over the last three years. In addition, EPRI has developed a linked set of computer codes to simulate the pressurized thermal shock transients and assess the integrity of the nuclear reactor vessels for various overcooling transients. This paper focuses on the integrated analysis approach being used by EPRI in performing such analysis. (orig.)
Temporal neural network for the identification of nuclear power plant transients
International Nuclear Information System (INIS)
Uluyol, O.; Ragheb, M.
1993-01-01
In this paper a layered spatiotemporal neural network is proposed for the identification of nuclear power plant transients. The developed layered spatiotemporal network is inspired by the formal avalanche structure developed by S. Grossberg and offers advantages compared with the stationary pattern approach using the perceptron paradigm. Each layer in the network is trained to recognize a separate time-dependent accident scenario. Within each scenario, the temporal behavior of the relevant parameters such as pressurizer pressure, pressurizer water volume, cold and hot legs temperatures, vessel flow, and power, are considered. Numerical cases are considered where the proposed methodology is applied to two nuclear power plant anticipated transient scenarios: the Station Blackout and the Anticipated Transient without Scram transients in a pressurized water reactor . The transient signatures used were generated by modeling the accidents using RELAP5/MOD2, a best-estimate thermal-hydraulics numerical code. The ability of the proposed layered spatiotemporal network to operate at different noise levels is investigated. Its incorporation within an Insightful Algorithm and Anticipatory Systems context for identifying and in predicting the course of nuclear transients is discussed
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)
Transient-Switch-Signal Suppressor
Bozeman, Richard J., Jr.
1995-01-01
Circuit delays transmission of switch-opening or switch-closing signal until after preset suppression time. Used to prevent transmission of undesired momentary switch signal. Basic mode of operation simple. Beginning of switch signal initiates timing sequence. If switch signal persists after preset suppression time, circuit transmits switch signal to external circuitry. If switch signal no longer present after suppression time, switch signal deemed transient, and circuit does not pass signal on to external circuitry, as though no transient switch signal. Suppression time preset at value large enough to allow for damping of underlying pressure wave or other mechanical transient.
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.)
A reliability analysis of a natural-gas pressure-regulating installation
International Nuclear Information System (INIS)
Gerbec, Marko
2010-01-01
A case study involving analyses of the operability, reliability and availability was made for a selected, typical, high-pressure, natural-gas, pressure-regulating installation (PRI). The study was commissioned by the national operator of the natural-gas, transmission-pipeline network for the purpose of validating the existing operability and maintenance practices and policies. The study involved a failure-risk analysis (HAZOP) of the selected typical installation, retrieval and analysis of the available corrective maintenance data for the PRI's equipment at the network level in order to obtain the failure rates followed by an elaboration of the quantitative fault trees. Thus, both operator-specific and generic literature data on equipment failure rates were used. The results obtained show that two failure scenarios need to be considered: the first is related to the PRI's failure to provide gas to the consumer(s) due to a low-pressure state and the second is related to a failure of the gas pre-heating at the high-pressure reduction stage, leading to a low temperature (a non-critical, but unfavorable, PRI state). Related to the first scenario, the most important cause of failure was found to be a transient pressure disturbance back from the consumer side. The network's average PRI failure frequency was assessed to be about once per 32 years, and the average unavailability to be about 4 minutes per year (the confidence intervals were also assessed). Based on the results obtained, some improvements to the monitoring of the PRI are proposed.
A reliability analysis of a natural-gas pressure-regulating installation
Energy Technology Data Exchange (ETDEWEB)
Gerbec, Marko, E-mail: marko.gerbec@ijs.s [Jozef Stefan Institute, Jamova 39, 1000 Ljubljana (Slovenia)
2010-11-15
A case study involving analyses of the operability, reliability and availability was made for a selected, typical, high-pressure, natural-gas, pressure-regulating installation (PRI). The study was commissioned by the national operator of the natural-gas, transmission-pipeline network for the purpose of validating the existing operability and maintenance practices and policies. The study involved a failure-risk analysis (HAZOP) of the selected typical installation, retrieval and analysis of the available corrective maintenance data for the PRI's equipment at the network level in order to obtain the failure rates followed by an elaboration of the quantitative fault trees. Thus, both operator-specific and generic literature data on equipment failure rates were used. The results obtained show that two failure scenarios need to be considered: the first is related to the PRI's failure to provide gas to the consumer(s) due to a low-pressure state and the second is related to a failure of the gas pre-heating at the high-pressure reduction stage, leading to a low temperature (a non-critical, but unfavorable, PRI state). Related to the first scenario, the most important cause of failure was found to be a transient pressure disturbance back from the consumer side. The network's average PRI failure frequency was assessed to be about once per 32 years, and the average unavailability to be about 4 minutes per year (the confidence intervals were also assessed). Based on the results obtained, some improvements to the monitoring of the PRI are proposed.
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)
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
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
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.
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)
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...
Aeroelastic Modeling of a Nozzle Startup Transient
Wang, Ten-See; Zhao, Xiang; Zhang, Sijun; Chen, Yen-Sen
2014-01-01
Lateral nozzle forces are known to cause severe structural damage to any new rocket engine in development during test. While three-dimensional, transient, turbulent, chemically reacting computational fluid dynamics methodology has been demonstrated to capture major side load physics with rigid nozzles, hot-fire tests often show nozzle structure deformation during major side load events, leading to structural damages if structural strengthening measures were not taken. The modeling picture is incomplete without the capability to address the two-way responses between the structure and fluid. The objective of this study is to develop a tightly coupled aeroelastic modeling algorithm by implementing the necessary structural dynamics component into an anchored computational fluid dynamics methodology. The computational fluid dynamics component is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, while the computational structural dynamics component is developed under the framework of modal analysis. Transient aeroelastic nozzle startup analyses at sea level were performed, and the computed transient nozzle fluid-structure interaction physics presented,
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)
Transient Characteristics of Free Piston Vuilleurnier Cycle Heat Pumps
Matsue, Junji; Fujimoto, Norioki; Shirai, Hiroyuki
A dynamic analysis of a free piston Vuilleumier cycle heat pump was performed using a time-stepping integration method to investigate transient characteristics under power controlling. The nonlinear relationship between displacement and force for pistons was taken into account for the motion of reciprocating components. The force for pistons is mainly caused by the pressure change of working gas varying with piston displacements; moreover nonlinear viscous dissipative force due to the oscillating flow of working gas in heat exchangers and discontinuous damping force caused by solid friction at piston seals and rod seals are included. The displacements of pistons and pressure changes in the Vuilleumier cycle heat pump were integrated by an ideal isothermal thermodynamic relationship. It was assumed that the flow friction was proportional to the kinematic pressure of working gas, and that the solid friction at the seals was due to the functions of the working gas pressure and the tension of seal springs. In order to investigate the transient characteristics of a proposed free piston Vuilleumier cycle heat pump machine when hot-side working gas temperatures and alternate force were changed, some calculations were performed and discussed. These calculation results make clear transient characteristics at starting and power controlling. It was further found that only a small amount of starter power is required in particular conditions. During controlling, the machine becomes unstable when there is ar elatively large reduction in cooling or heating power. Therefore, an auxiliary device is additionally needed to obtain stable operation, such as al inear motor.
Simulation of SBWR startup transient and stability
International Nuclear Information System (INIS)
Cheng, H.S.; Khan, H.J.; Rohatgi, U.S.
1998-01-01
The Simplified Boiling Water Reactor (SBWR) designed by General Electric is a natural circulation reactor with enhanced safety features for potential accidents. It has a strong coupling between power and flow in the reactor core, hence the neutronic coupling with thermal-hydraulics is specially important. The potential geysering instability during the early part of a SBWR startup at low flow, low power and low pressure is of particular concern. The RAMONA-4B computer code developed at Brookhaven National Laboratory (BNL) for the SBWR has been used to simulate a SBWR startup transient and evaluate its stability, using a simplified four-channel representation of the reactor core for the thermal-hydraulics. This transient was run for 20,000 sec (5.56 hrs) in order to cover the essential aspect of the SBWR startup. The simulation showed that the SBWR startup was a very challenging event to analyze as it required accurate modeling of the thermal-hydraulics at low pressures. This analysis did not show any geysering instability during the startup, following the startup procedure as proposed by GE
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
Stability of infinite slopes under transient partially saturated seepage conditions
Godt, Jonathan W.; ŞEner-Kaya, BaşAk; Lu, Ning; Baum, Rex L.
2012-05-01
Prediction of the location and timing of rainfall-induced shallow landslides is desired by organizations responsible for hazard management and warnings. However, hydrologic and mechanical processes in the vadose zone complicate such predictions. Infiltrating rainfall must typically pass through an unsaturated layer before reaching the irregular and usually discontinuous shallow water table. This process is dynamic and a function of precipitation intensity and duration, the initial moisture conditions and hydrologic properties of the hillside materials, and the geometry, stratigraphy, and vegetation of the hillslope. As a result, pore water pressures, volumetric water content, effective stress, and thus the propensity for landsliding vary over seasonal and shorter time scales. We apply a general framework for assessing the stability of infinite slopes under transient variably saturated conditions. The framework includes profiles of pressure head and volumetric water content combined with a general effective stress for slope stability analysis. The general effective stress, or suction stress, provides a means for rigorous quantification of stress changes due to rainfall and infiltration and thus the analysis of slope stability over the range of volumetric water contents and pressure heads relevant to shallow landslide initiation. We present results using an analytical solution for transient infiltration for a range of soil texture and hydrological properties typical of landslide-prone hillslopes and show the effect of these properties on the timing and depth of slope failure. We follow by analyzing field-monitoring data acquired prior to shallow landslide failure of a hillside near Seattle, Washington, and show that the timing of the slide was predictable using measured pressure head and volumetric water content and show how the approach can be used in a forward manner using a numerical model for transient infiltration.
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.
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 ...
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
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.
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.
Determination of the failure probability in the weld region of ap-600 vessel for transient condition
International Nuclear Information System (INIS)
Wahyono, I.P.
1997-01-01
Failure probability in the weld region of AP-600 vessel was determined for transient condition scenario. The type of transient is increase of the heat removal from primary cooling system due to sudden opening of safety valves or steam relief valves on the secondary cooling system or the steam generator. Temperature and pressure in the vessel was considered as the base of deterministic calculation of the stress intensity factor. Calculation of film coefficient of the convective heat transfers is a function of the transient time and water parameter. Pressure, material temperature, flaw depth and transient time are variables for the stress intensity factor. Failure probability consideration was done by using the above information in regard with the flaw and probability distributions of Octavia II and Marshall. Calculation of the failure probability by probability fracture mechanic simulation is applied on the weld region. Failure of the vessel is assumed as a failure of the weld material with one crack which stress intensity factor applied is higher than the critical stress intensity factor. VISA II code (Vessel Integrity Simulation Analysis II) was used for deterministic calculation and simulation. Failure probability of the material is 1.E-5 for Octavia II distribution and 4E-6 for marshall distribution for each transient event postulated. The failure occurred at the 1.7th menit of the initial transient under 12.53 ksi of the pressure
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
Development of transient initiating event frequencies for use in probabilistic risk assessments
International Nuclear Information System (INIS)
Mackowiak, D.P.; Gentillon, C.D.; Smith, K.L.
1985-05-01
Transient initiating event frequencies are an essential input to the analysis process of a nuclear power plant probabilistic risk assessment. These frequencies describe events causing or requiring scrams. This report documents an effort to validate and update from other sources a computer-based data file developed by the Electric Power Research Institute (EPRI) describing such events at 52 United States commercial nuclear power plants. Operating information from the United States Nuclear Regulatory Commission on 24 additional plants from their date of commercial operation has been combined with the EPRI data, and the entire data base has been updated to add 1980 through 1983 events for all 76 plants. The validity of the EPRI data and data analysis methodology and the adequacy of the EPRI transient categories are examined. New transient initiating event frequencies are derived from the expanded data base using the EPRI transient categories and data display methods. Upper bounds for these frequencies are also provided. Additional analyses explore changes in the dominant transients, changes in transient outage times and their impact on plant operation, and the effects of power level and scheduled scrams on transient event frequencies. A more rigorous data analysis methodology is developed to encourage further refinement of the transient initiating event frequencies derived herein. Updating the transient event data base resulted in approx.2400 events being added to EPRI's approx.3000-event data file. The resulting frequency estimates were in most cases lower than those reported by EPRI, but no significant order-of-magnitude changes were noted. The average number of transients per year for the combined data base is 8.5 for pressurized water reactors and 7.4 for boiling water reactors
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
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)
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.
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.
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)
ATHENA simulations of divertor pump trip and loss of heat sink transients for the GSSR
Energy Technology Data Exchange (ETDEWEB)
Sjoeberg, A
2001-04-01
The ITER-FEAT Generic Site Safety Report includes evaluations of the consequences of various types of conceivable transients that may occur during operation. The transients that have to be considered in this respect are specified in the Accident Analysis Specifications document of the safety report. For the divertor primary heat transport system the ranges of transients include amongst others a trip of the main circulation pump in the divertor cooling loop as well as a loss of heat sink, both initiated at full fusion power operation. The thermal-hydraulic consequences related to the coolability of the divertor primary heat transport system components for these two transients have been evaluated and summarized in the safety report and in the current report an overview of those efforts and associated outcome is provided. The analyses have been made with the ATHENA thermal-hydraulic code using a separately developed ATHENA model of the ITER-FEAT divertor cooling system. The results from the analyses indicate that for the pump trip transient the margin against overheating of critical highly loaded parts of the divertor cassette is small but seems sufficient. In case of the loss of heat sink transient the conservative analysis reveals that the pressurizer safety valve will be opened for an extended period of time and the long term transient development indicates a risk of completely filling up the pressurizer vessel. Thus the margins against jeopardizing the integrity of the divertor cooling system with the current design are for this case small but can for a long term operation at associate conditions pose a problem.
Yao, Z.; Bi, H. L.; Huang, Q. S.; Li, Z. J.; Wang, Z. W.
2013-12-01
In load rejection transient process, the sudden shut down of guide vanes may cause units speed rise and a sharp increase in water hammer pressure of diversion system, which endangers the safety operation of the power plant. Adopting reasonable guide vane closure law is a kind of economic and effective measurement to reduce the water hammer pressure and limit rotational speed increases. In this paper, combined with Guangzhou Pumped Storage Power Station plant A, the load rejection condition under different guide vanes closure laws is calculated and the key factor of guide vanes closure laws on the impact of the load rejection transition process is analyzed. The different inflection points, which are the closure modes, on the impact of unit speed change, water level fluctuation of surge tank, and the pressure fluctuation of volute inlet and draft tube inlet are further discussed. By compared with the calculation results, a reasonable guide vanes inflection point position can be determined according to security requirements and a reasonable guide vanes closure law can be attained to effectively coordinate the unit speed rise and the rapid pressure change in the load rejection transient process.
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
PRETTA：A COMPUTER PROGRAM FOR PWR PRESSURIZER’S TRANSIENT THERMODYNAMICS
Institute of Scientific and Technical Information of China (English)
阿谢德; 徐济鋆
2001-01-01
A computer program PRETTA “Pressurizer Transient Thermodynamics Analysis” was developed for the prediction of pressurizer under transient conditions. It is based on the solution of the conservation laws of heat and mass applied to the three separate and non equilibrium thermodynamic regions. In the program all of the important thermal-hydraulics phenomena occurring in the pressurizer: stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer have been considered. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented computer program will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant's pressurizer performance.
RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors
Energy Technology Data Exchange (ETDEWEB)
Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael
2003-04-01
The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.
Evaluation to Mitigate Secondary System Peak Pressure for Loss of Condenser Vacuum Event
Energy Technology Data Exchange (ETDEWEB)
Choi, Bong Oh; Park, Jong Cheol; Park, Min Soo; Lee, Gyu Cheon; Kim, Shin Whan [KEPCO E and C, Inc., Daejeon (Korea, Republic of)
2015-10-15
In this paper, countermeasures to compensate the increased secondary pressure are introduced and evaluated. From the standpoint of the secondary system pressurization, consideration of the PPCS may result in a conservative secondary system peak pressure. The control systems are generally credited for the safety analysis if the analysis produces conservative results. However, in most of all non-loss of coolant accident (non-LOCA) events, the control system helps to mitigate a transient state. Accordingly, the safety analysis of non-LOCA assumes the control systems are in the manual mode of operation. The loss of condenser vacuum event (LOCV) is a typical anticipated operational occurrence (AOO) which results in an increase in primary and secondary system pressure. The pressurizer (PZR) pressure control system (PPCS) will function to reduce the primary system pressure increase during the transient. Therefore, it is assumed to be in manual mode and credit is not taken for its functioning. However, crediting the function of PPCS has been found to be more conservative with regard to the secondary system pressure. This is due to the delay of the reactor trip on high pressurizer pressure (HPP) and results in an increase in secondary pressure.
Improvements to the transient solution in the PANTHER space-time code
International Nuclear Information System (INIS)
Kutt, P.K.; Knight, M.P.
1993-01-01
The three dimensional, two-group, nodal diffusion code PANTHER has been developed for the analysis of almost all thermal reactor types [pressurized water reactor (PWR), boiling water reactor, VVER, RBMK, advanced gas-cooled reactor, MAGNOX]. It can perform a comprehensive range of calculations for fuel management, operational support including on-line application, and transient analysis. Transient results for a number of light water reactor (LWR) benchmark problems have been reported previously. This paper outlines some recent developments of the transient solution in PANTHER, showing results for two LWR benchmark problems. Recently, PANTHER results have been accepted as the reference solutions for a Nuclear Energy Agency Committee on Reactor Physics (NEACRP) rod ejection benchmark Unlike previous simplified rod ejection benchmarks, it represents a real PWR with a detailed thermal model and cross sections dependent on boron, fuel temperature, and water density and temperature. This reference solution was computed with fine time steps
Preliminary Evaluations of CSPACE for a Station Blackout Transient in APR1400
Energy Technology Data Exchange (ETDEWEB)
Lee, T. B.; Lee, D. K.; Lee, H. S.; Lee, G. W.; Choi, T. S. [KEPCO, Daejeon (Korea, Republic of); Park, R. J.; Kim, D. H. [KAERI, Daejeon (Korea, Republic of)
2016-05-15
This paper discusses the preliminary results of the simulated station blackout (SBO) transients using the CSPACE code and presents the information pertinent to the related safety issues. CSPACE is a merged program of a master processer of Safety and Performance Analysis Code (SPACE) for nuclear power plants and a child processer of Core Meltdown Progression Accident Simulation Software (COMPASS) generated as a dynamic-link library (DLL) codes. It has been developed to predict the best-estimate transient in the pressurized water reactor (PWR) for severe accidents. SPACE and COMPASS codes take charge of the thermal-hydraulic response of PWRs and the analysis of the severe accident progression in a vessel, respectively. The initial phase is estimated starting from time zero when the loss of off-site and on-site powers occurs simultaneously. Shortly after the RCS pressure initially falls and rises slightly due to the effects of the reactor and turbine trips, the RCS pressure declines in response to the cooling provided by heat removed to the SGs. During the period of the primary heat-up and boil-off, the RCS pressure increase is limited by two cycles of the POSRV. The RCS fluid mass is lost through the pressurizer POSRV and then the core uncovers and superheated steam flows out from the RV into the coolant loops starting at 5513.0 seconds.
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
Tacina, R. R.
1984-01-01
Non-steady combustion problems can result from engine sources such as accelerations, decelerations, nozzle adjustments, augmentor ignition, and air perturbations into and out of the compressor. Also non-steady combustion can be generated internally from combustion instability or self-induced oscillations. A premixed-prevaporized combustor would be particularly sensitive to flow transients because of its susceptability to flashback-autoignition and blowout. An experimental program, the Transient Flow Combustion Study is in progress to study the effects of air and fuel flow transients on a premixed-prevaporized combustor. Preliminary tests performed at an inlet air temperature of 600 K, a reference velocity of 30 m/s, and a pressure of 700 kPa. The airflow was reduced to 1/3 of its original value in a 40 ms ramp before flashback occurred. Ramping the airflow up has shown that blowout is more sensitive than flashback to flow transients. Blowout occurred with a 25 percent increase in airflow (at a constant fuel-air ratio) in a 20 ms ramp. Combustion resonance was found at some conditions and may be important in determining the effects of flow transients.
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
Response of air cleaning system dampers and blowers to simulated tornado transients
International Nuclear Information System (INIS)
Gregory, W.; Idar, E.; Smith, P.; Hensel, E.; Smith, E.
1985-01-01
The effects of tornado-like pressure transients upon dampers and blowers in nuclear air cleaning systems were studied. For the dampers pressure drop as a function of flow rate was obtained and an empirical relationship developed. Transient response was examined for several types of dampers, as was structural integrity. Both centrifugal and axi-vane blowers were tested and transient characteristic curves were generated in outrunning and backflow situations. The transient characteristic curves do not necessarily match the quasi-steady characteristic curves
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.
Fluid-structure interaction analysis for pressurizer surge line subjected to thermal stratification
International Nuclear Information System (INIS)
Kang, Dong Gu; Jhung, Myung Jo; Chang, Soon Heung
2011-01-01
Research highlights: → Temperature of surge line due to stratified flow is defined using CFD analysis. → Fluid-structure interaction analysis is performed to investigate the response characteristics due to thermal stress. → Fatigue usage factors due to thermal stratification are relatively low. → Simplifying temperature distribution in surge line is not always conservative. - Abstract: Serious mechanical damages such as cracks and plastic deformations due to excessive thermal stress caused by thermal stratification have been experienced in several nuclear power plants. In particular, the thermal stratification in the pressurizer surge line has been addressed as one of the significant safety and technical issues. In this study, a detailed unsteady computational fluid dynamics (CFD) analysis involving conjugate heat transfer analysis is performed to obtain the transient temperature distributions in the wall of the pressurizer surge line subjected to stratified internal flows either during out-surge or in-surge operation. The thermal loads from CFD calculations are transferred to the structural analysis code which is employed for the thermal stress analysis to investigate the response characteristics, and the fatigue analysis is ultimately performed. In addition, the thermal stress and fatigue analysis results obtained by applying the realistic temperature distributions from CFD calculations are compared with those by assuming the simplified temperature distributions to identify some requirements for a realistic and conservative thermal stress analysis from a safety point of view.
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.
Development of transient initiating event frequencies for use in probabilistic risk assessments
Energy Technology Data Exchange (ETDEWEB)
Mackowiak, D.P.; Gentillon, C.D.; Smith, K.L.
1985-05-01
Transient initiating event frequencies are an essential input to the analysis process of a nuclear power plant probabilistic risk assessment. These frequencies describe events causing or requiring scrams. This report documents an effort to validate and update from other sources a computer-based data file developed by the Electric Power Research Institute (EPRI) describing such events at 52 United States commercial nuclear power plants. Operating information from the United States Nuclear Regulatory Commission on 24 additional plants from their date of commercial operation has been combined with the EPRI data, and the entire data base has been updated to add 1980 through 1983 events for all 76 plants. The validity of the EPRI data and data analysis methodology and the adequacy of the EPRI transient categories are examined. New transient initiating event frequencies are derived from the expanded data base using the EPRI transient categories and data display methods. Upper bounds for these frequencies are also provided. Additional analyses explore changes in the dominant transients, changes in transient outage times and their impact on plant operation, and the effects of power level and scheduled scrams on transient event frequencies. A more rigorous data analysis methodology is developed to encourage further refinement of the transient initiating event frequencies derived herein. Updating the transient event data base resulted in approx.2400 events being added to EPRI's approx.3000-event data file. The resulting frequency estimates were in most cases lower than those reported by EPRI, but no significant order-of-magnitude changes were noted. The average number of transients per year for the combined data base is 8.5 for pressurized water reactors and 7.4 for boiling water reactors.
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.
Transient burnout in flow reduction condition
International Nuclear Information System (INIS)
Iwamura, Takamichi; Kuroyanagi, Toshiyuki
1981-01-01
A transient flow reduction burnout experiment was conducted with water in a uniformly heated, vertically oriented tube. Test pressures ranged from 0.5 to 3.9 MPa. An analytical method was developed to obtain transient burnout conditions at the exit. A simple correlation to predict the deviation of the transient burnout mass velocity at the tube exit from the steady state mass velocity obtained as a function of steam-water density ratio and flow reduction rate. The correlation was also compared with the other data. (author)
SENARIET, A Programme To Solve Transient Flows Of Liquids In Complex Circuits
Vargas-Munoz, M.; Rodriguez-Fernandez, M.; Perena-Tapiador, A.
2011-05-01
SENARIET is a programme to study fluid transients in pipeline systems in order to obtain pressure and velocity distributions along a circuit. When a transient process occurs in periods of the same order of the pressure waves’ travelling time along a circuit (the order of the circuit length divided by the effective propagation speed), the compressibility effects in liquids have to be considered. Taking this effect into account, the appropriate equations of continuity and momentum are solved by the method of characteristics, to obtain pressure and velocity along pipes as a function of time. The simulated results have been compared to theoretical and experimental ones to validate and evaluate the precision of the software. The results help to perform efficient and accurate predictions in order to define the propulsion sub-system. This type of analysis is very important in order to evaluate the water hammer effects in propulsion systems used on spacecrafts and launchers.
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
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
International Nuclear Information System (INIS)
Liles, D.R.; Mahaffy, J.H.
1986-07-01
The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients
Energy Technology Data Exchange (ETDEWEB)
Liles, D.R.; Mahaffy, J.H.
1986-07-01
The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.
Transient risk factors of acute occupational injuries
DEFF Research Database (Denmark)
Østerlund, Anna H; Lander, Flemming; Nielsen, Kent
2017-01-01
Objectives The objectives of this study were to (i) identify transient risk factors of occupational injuries and (ii) determine if the risk varies with age, injury severity, job task, and industry risk level. Method A case-crossover design was used to examine the effect of seven specific transient...... risk factors (time pressure, disagreement with someone, feeling sick, being distracted by someone, non-routine task, altered surroundings, and broken machinery and materials) for occupational injuries. In the study, 1693 patients with occupational injuries were recruited from a total of 4002...... in relation to sex, age, job task, industry risk level, or injury severity. Conclusion Use of a case-crossover design identified several worker-related transient risk factors (time pressure, feeling sick, being distracted by someone) that led to significantly increased risks for occupational injuries...
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
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
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
Determination of optimum pressurizer level for kori unit 1
Energy Technology Data Exchange (ETDEWEB)
Song, Dong Soo; Lee, Chang Sup; Yong, Lee Jae; Kim, Yo Han; Lee, Dong Hyuk [Korea Electric Power Research Institute, Taejon (Korea, Republic of)
1998-12-31
To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are performed. The methodology is developed by evaluating {sup d}ecrease in secondary heat removal{sup e}vents such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling. The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. 6 refs., 5 figs. (Author)
Determination of optimum pressurizer level for kori unit 1
Energy Technology Data Exchange (ETDEWEB)
Song, Dong Soo; Lee, Chang Sup; Lee Jae Yong; Kim, Yo Han; Lee, Dong Hyuk [Korea Electric Power Research Institute, Taejon (Korea, Republic of)
1997-12-31
To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are performed. The methodology is developed by evaluating {sup d}ecrease in secondary heat removal{sup e}vents such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling. The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. 6 refs., 5 figs. (Author)
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
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
Simulation Analysis of Computer-Controlled pressurization for Mixture Ratio Control
Alexander, Leslie A.; Bishop-Behel, Karen; Benfield, Michael P. J.; Kelley, Anthony; Woodcock, Gordon R.
2005-01-01
A procedural code (C++) simulation was developed to investigate potentials for mixture ratio control of pressure-fed spacecraft rocket propulsion systems by measuring propellant flows, tank liquid quantities, or both, and using feedback from these measurements to adjust propellant tank pressures to set the correct operating mixture ratio for minimum propellant residuals. The pressurization system eliminated mechanical regulators in favor of a computer-controlled, servo- driven throttling valve. We found that a quasi-steady state simulation (pressure and flow transients in the pressurization systems resulting from changes in flow control valve position are ignored) is adequate for this purpose. Monte-Carlo methods are used to obtain simulated statistics on propellant depletion. Mixture ratio control algorithms based on proportional-integral-differential (PID) controller methods were developed. These algorithms actually set target tank pressures; the tank pressures are controlled by another PID controller. Simulation indicates this approach can provide reductions in residual propellants.
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
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
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
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.
Turbofan compressor dynamics during afterburner transients
Kurkov, A. P.
1976-01-01
The effects of afterburner light-off and shut-down transients on the compressor stability are investigated. The reported experimental results are based on detailed high response pressure and temperature measurements on the TF30-P-3 turbofan engine. The tests were performed in an altitude test chamber simulating high altitude engine operation. It is shown that during both types of transients, flow breaks down in the forward part of the fan bypass duct. At a sufficiently low engine inlet pressure this resulted in a compressor stall. Complete flow breakdown within the compressor was preceded by a rotating stall. At some locations in the compressor, rotating stall cells initially extended only through part of the blade span. For the shutdown transient the time between first and last detected occurrence of rotating stall is related to the flow Reynolds number. An attempt was made to deduce the number and speed of propagation of rotating stall cells.
Transient Diagnosis and Prognosis for Secondary System in Nuclear Power Plants
Directory of Open Access Journals (Sweden)
Sangjun Park
2016-10-01
Full Text Available This paper introduces the development of a transient monitoring system to detect the early stage of a transient, to identify the type of the transient scenario, and to inform an operator with the remaining time to turbine trip when there is no operator's relevant control. This study focused on the transients originating from a secondary system in nuclear power plants (NPPs, because the secondary system was recognized to be a more dominant factor to make unplanned turbine-generator trips which can ultimately result in reactor trips. In order to make the proposed methodology practical forward, all the transient scenarios registered in a simulator of a 1,000 MWe pressurized water reactor were archived in the transient pattern database. The transient patterns show plant behavior until turbine-generator trip when there is no operator's intervention. Meanwhile, the operating data periodically captured from a plant computer is compared with an individual transient pattern in the database and a highly matched section among the transient patterns enables isolation of the type of transient and prediction of the expected remaining time to trip. The transient pattern database consists of hundreds of variables, so it is difficult to speedily compare patterns and to draw a conclusion in a timely manner. The transient pattern database and the operating data are, therefore, converted into a smaller dimension using the principal component analysis (PCA. This paper describes the process of constructing the transient pattern database, dealing with principal components, and optimizing similarity measures.
New transient-flow modelling of a multiple-fractured horizontal well
International Nuclear Information System (INIS)
Jia, Yong-Lu; Wang, Ben-Cheng; Nie, Ren-Shi; Wang, Dan-Ling
2014-01-01
A new transient-flow modelling of a multiple-fractured horizontal well is presented. Compared to conventional modelling, the new modelling considered more practical physical conditions, such as various inclined angles for different fractures, different fracture intervals, different fracture lengths and partially penetrating fractures to formation. A kind of new mathematical method, including a three-dimensional eigenvalue and orthogonal transform, was created to deduce the exact analytical solutions of pressure transients for constant-rate production in real space. In order to consider a wellbore storage coefficient and skin factor, we used a Laplace-transform approach to convert the exact analytical solutions to the solutions in Laplace space. Then the numerical solutions of pressure transients in real space were gained using a Stehfest numerical inversion. Standard type curves were plotted to describe the transient-flow characteristics. Flow regimes were clearly identified from type curves. Furthermore, the differences between the new modelling and the conventional modelling in pressure transients were especially compared and discussed. Finally, an example application to show the accordance of the new modelling with real conditions was implemented. Our new modelling is different from, but more practical than, conventional modelling. (paper)
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.
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)
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)
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
International Nuclear Information System (INIS)
Carmo, E.G.D. do; Oliveira, L.F.S. de; Roberty, N.C.
1984-01-01
A method for the determination of operational limit curves (primary pressure versus temperature) for PWR is presented. Such curves give the operators indications related to the safety status of the plant concerning the possibility of a pressurized thermal shock. The method begins by a thermal analysis for several postulated transients, followed by the determination of the thermomechanical stresses in the vessel and finally it makes use of the linear elasticity fracture mechanics. Curves are shown for a typical PWR. (Author) [pt
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)
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)
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
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
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
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
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.
PSH Transient Simulation Modeling
Energy Technology Data Exchange (ETDEWEB)
Muljadi, Eduard [National Renewable Energy Laboratory (NREL), Golden, CO (United States)
2017-12-21
PSH Transient Simulation Modeling presentation from the WPTO FY14 - FY16 Peer Review. Transient effects are an important consideration when designing a PSH system, yet numerical techniques for hydraulic transient analysis still need improvements for adjustable-speed (AS) reversible pump-turbine applications.
International Nuclear Information System (INIS)
Wang Xiuli; Yuan Shouqi; Zhu Rongsheng; Yu Zhijun
2013-01-01
The numerical simulation calculation of the transient flow characteristics of nuclear reactor coolant pump in the recessive cavitation transition process in the nuclear reactor coolant pump impeller passage is conducted by CFX, and the transient flow characteristics of nuclear reactor coolant pump in the transition process from reducing the inlet pressure at cavitation-born conditions to NPSHc condition is studied and analyzed. The flow field analysis shows that, in the recessive cavitation transition process, the speed diversification at the inlet is relative to the bubble increasing, and makes the speed near the blade entrance increase when the bubble phase region becomes larger. The bubble generation and collapse will affect the the speed fluctuation near the entrance. The vorticity close to the blade entrance gradually increasing is influenced by the bubble phase, and the collapse of bubble generated by cavitation will reduce the vorticity from the collapse to impeller outlet. Pump asymmetric structure causes the asymmetry of the flow, velocity and outlet pressure distribution within every impeller flow passage, which cause the asymmetry of the transient radial force. From the dimensionless t/T = 0.6, the bubble phase starts to have impact on the impeller transient radial force, and results in the irregular fluctuations. (authors)
Structural Integrity Assessment of VVER-1000 RPV under Accidental Cool down Transients
International Nuclear Information System (INIS)
Shrivastav, V.; Sen, R.N.; Yadav, R.S.
2012-01-01
Corrosion, Fatigue and Irradiation embrittlement are the major degradation mechanisms responsible for ageing of RPV (and its internals) of a Pressurized Water Reactor. While corrosion and fatigue can generate cracks, irradiation damage can lead to brittle fracture initiating from these cracks. Ageing in nuclear power plants needs to be managed so as to ensure that design functions remain available throughout the life of the plant. From safety perspective, this implies that ageing degradation of systems, structures and components important to safety remain within acceptable limits. Reactor Pressure Vessel has been identified as the highest priority key component in plant life management for Pressurized Water Reactors. Therefore special attention is required to ensure its structural integrity during its lifetime. In this paper, structural integrity assessment for typical VVER-1000 RPV is carried out under severe accidental cool down transients using the Finite Element Method. Three different accidental scenarios are postulated and safety of the vessel is conservatively assessed under these transients using the Linear Elastic Fracture Mechanics approach. Transient thermo mechanical stress analysis of the core belt region of the RPV is carried out in presence of postulated cracks and stress intensity factors are calculated and compared with the material fracture toughness to assess the structural integrity of the vessel. The paper also include some parametric analyses to justify the methodology. (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.
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.
Transient evaluation for LPG and oil pipelines
Energy Technology Data Exchange (ETDEWEB)
Meliande, Patricia; Do Nascimento, Elson Antonio; Fernandes Lacerda, Rogerio [Federal Fluminense University, Niteroi, RJ (Brazil)
2010-07-01
In the last decades, the offshore industry has expanded thanks to the development of equipment for subsea operation. In such installations, protection equipment can be adapted to the installation's design in order to avoid surge pressure effects and ensure safety and integrity of the system, predicting transient effect is therefore very important. The purpose of this study is to predict the surge pressure on refrigerated LPG and gasoline pipelines during unpredicted closure of valves. Flowmaster software was used to carry out the simulations and its results were then validated with a methodology that applied the characteristics method based on Wylie and Streeter assumptions. Results showed that surge pressures during transient effect for both refrigerated LPG and Gasoline pipeline systems were less than the maximum allowable operating pressure and therefore no control dispositive installations are geeded. This study showed that the use of these computer models can ensure the optimization of the system and provide cost effective solutions.
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,
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
Line pressure effects on differential pressure measurements
International Nuclear Information System (INIS)
Neff, G.G.; Evans, R.P.
1982-01-01
The performance of differential pressure transducers in experimental pressurized water reactor (PWR) systems was evaluated. Transient differential pressure measurements made using a simple calibration proportionality relating differential pressure to output voltage could have large measurement uncertainties. A more sophisticated calibration equation was derived to incorporate the effects of zero shifts and sensitivity shifts as pressure in the pressure sensing line changes with time. A comparison made between the original calibration proportionality equation and the derived compensation equation indicates that potential measurement uncertainties can be reduced
Severe accident analysis to prevent high pressure scenarios in the EPR TM
International Nuclear Information System (INIS)
Azarian, G.; Gandrille, P.; Gasperini, M.; Klein, R.
2010-01-01
The EPR TM has incorporated several design features in order to specifically address major severe accident safety issues. In particular, it was designed with the objective to transfer high pressure core melt scenarios into a low pressure scenario with high reliability so that a high pressure vessel failure can be practically eliminated. It is the key issue in the defense-in-depth approach, for a postulated severe accident with core melting, to prevent any risk of containment failure due to possible Direct Containment Heating or due to reactor vessel rocketing which results from vessel failure at high pressure. Temperature-induced steam generator tube rupture, which could lead to a radiological containment bypass, has also to be prevented. On the basis of the analysis of the main high pressure core melt scenarios which are calculated with the MAAP4.07 code which was developed to support the EPR TM, this paper explores the benefits of primary depressurization by dedicated valves on transient evolutions. It specifically addresses the thermal response of the structures by sensitivity studies involving the timing of valve actuation. It outlines that a grace period of at least one hour is available for a delayed valve actuation without inducing excessive loads and without increasing the risk of a temperature-induced steam generator tube rupture. (authors)
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.
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
International Nuclear Information System (INIS)
Wang Yan
2014-01-01
A certain amount of hydrogen will be generated due to zirconium-steam reaction or molten corium concrete interaction during severe accidents in the pressurized water reactor (PWR). The generated hydrogen releases into the containment, and the formed flammable mixture might cause deflagration or detonation to produce high thermal and pressure loads on the containment, which may threaten the integrity of the containment. The non-condensable hydrogen in containment may also reduce the steam condensation on the containment surface to affect the performance of the passive containment cooling system (PCCS). To study the transient hydrogen behavior in containment with the PCCS performance during the accidents is significant for the further study on the PCCS design and the hydrogen risk mitigation. In this paper, a new developed PCCS analysis code with self-reliance intellectual property rights, which had been validated by comparison on the transients in the containment during the design basis accidents with other developed PCCS analysis code, is brief introduced and used for the transient simulation in the containment under a postulated small break LOCA of cold-leg. The results show that the hydrogen will flow upwards with the coolant released from the break and spread in the containment by convection and diffusion, and it results in the increase of the pressure in the containment due to reducing the heat removal capacity of the PCCS. (author)
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)
Study of transient burnout characteristics under flow reduction condition
International Nuclear Information System (INIS)
Iwamura, Takamichi; Kuroyanagi, Toshiyuki
1984-03-01
As part of a study of the thermal behavior of fuel rods during Power-Cooling-Mismatch (PCM) accidents in light water reactors, burnout characteristics in a uniformly heated, vertically oriented tube or annulus, under flow reduction condition, were experimentally studied. Test pressures ranged 0.1--3.9 MPa and flow reduction rates 0.44--1100%/s. An analytical method is developed to obtain the local mass velocity during a transient condition. The major results are as follows: With increasing flow reduction rate beyond a threshold, transient burnout mass velocity at the inlet was lower than that in steady state tests under the experimental pressures. The higher system pressure resulted in the less transient effects. At pressures higher than 2.0 MPa and flow reduction rates lower than 20%/s, the local burnout mass velocity agreed with the steady state burnout mass velocity, whereas the local burnout mass velocity became higher than the steady state burnout mass velocity at flow reduction rates higher than 20%/s. At pressures lower than 1 MPa, with increasing flow reduction rate beyond the threshold value of 2%/s, the local burnout mass velocity was lower than the steady state burnout mass velocity. An empirical correlation is presented to give the ratio of the transient to the steady state burnout mass velocities at the burnout location as a function of the steam-water density ratio and the flow reduction rate. The experimental results by Cumo et al. agree with the correlation. The correlation, however, cannot predict the experimental results at higher flow reduction rates beyond 40%/s. (author)
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
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 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
Integrity of PWR pressure vessels during overcooling accidents
International Nuclear Information System (INIS)
Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.
1982-01-01
The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation
Integrity of PWR pressure vessels during overcooling accidents
International Nuclear Information System (INIS)
Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.
1982-01-01
The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation
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
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)
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
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...
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)
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.
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...
Thermohydraulics in rod bundles and critical heat flux in transient conditions in a tube
International Nuclear Information System (INIS)
Courtaud, M.; Roumy, R.
1975-01-01
After the determination of the scaling factor of Stevens's similitude for the pressure range of pressurized water vectors by comparison of critical heat flux data in from and in water, some examples of studies performed with freon are shown. The efficiency of the mixing vanes of spacer grids has been determined on the mixing phenomenon in single phase on critical heat flux. A calculation performed with the code FLICA using subchannel analysis on freon data transposed in water is in good agreement with the experiment. The influence of the number of spacer grids has been also shown. Critical heat fluxes have been determined in water at 140 bar in steady state and transient conditions on two tubular test sections. During the transient tests the flow rate was reduced by half in 0.5 seconds and the reincreased heat flux and inlet temperature remaining constant. These tests have shown the validity of the method which consists in using a critical heat flux correlation determined in steady state conditions applied with local transient conditions of enthalpy and mass velocity computed with the FLICA code [fr
TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis
International Nuclear Information System (INIS)
Liles, D.R.; Mahaffy, J.H.
1984-02-01
The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided
Sample problem calculations related to two-phase flow transients in a PWR relief-piping network
International Nuclear Information System (INIS)
Shin, Y.W.; Wiedermann, A.H.
1981-03-01
Two sample problems related with the fast transients of water/steam flow in the relief line of a PWR pressurizer were calculated with a network-flow analysis computer code STAC (System Transient-Flow Analysis Code). The sample problems were supplied by EPRI and are designed to test computer codes or computational methods to determine whether they have the basic capability to handle the important flow features present in a typical relief line of a PWR pressurizer. It was found necessary to implement into the STAC code a number of additional boundary conditions in order to calculate the sample problems. This includes the dynamics of the fluid interface that is treated as a moving boundary. This report describes the methodologies adopted for handling the newly implemented boundary conditions and the computational results of the two sample problems. In order to demonstrate the accuracies achieved in the STAC code results, analytical solutions are also obtained and used as a basis for comparison
Development and assessment of a sub-channel code applicable for trans-critical transient of SCWR
International Nuclear Information System (INIS)
Liu, X.J.; Yang, T.; Cheng, X.
2013-01-01
Highlights: • A new sub-channel code COBRA-SC for SCWR is developed. • Pseudo two-phase method is employed to realize trans-critical transient calculation. • Good suitability of COBRA-SC is demonstrated by preliminary assessment. • The calculation results of COBRA-SC agree well with ATHLET code. -- Abstract: In the last few years, extensive R and D activities have been launched covering various aspects of supercritical water-cooled reactor (SCWR), especially the thermal-hydraulic analysis. Sub-channel code plays an indispensable role to predict the detail thermal-hydraulic behavior of the SCWR fuel assembly. This paper develops a new version of sub-channel code COBRA-SC based on the previous COBRA-IV code. The supercritical water property and heat transfer/pressure drop correlations under supercritical pressure are implemented to this code. Moreover, in order to simulate the trans-critical transient (the pressure undergo a decrease from the supercritical pressure to the subcritical pressure), pseudo two-phase method is employed in COBRA-SC code. This work is completed by introduction of a virtual two-phase region near the pseudo-critical line. A smooth transition of void fraction can be realized. In addition, several heat transfer correlations right underneath the critical point are introduced into this code to capture the heat transfer behavior during the trans-critical transient. Some experimental data from simple geometry, e.g. the single tube, small rod bundle, is used to validate and evaluate this new developed COBRA-SC code. The predicted results show a good agreement with the experimental data, demonstrating good feasibility of this code for SCWR condition. A code to code comparison between COBRA-SC and ATHLET for a blowdown transient of a small fuel assembly is also presented and discussed in this paper
International Nuclear Information System (INIS)
Medeiros, Eduarda da C.A.; Castrillo, Lazara S.
2015-01-01
Insurge and outsurge phenomena are transient states and could be analyzed by thermodynamics principles, the pressurizer behavior will vary in response to mass flow changes. These surges can occur in the presence of noncondensable gases. On this paper, with the code RELAP5, the IRIS reactor pressurizer is described to analyze surges phenomena in their control volumes with non-condensable gases since they modify the pressure response. A set of three pipes components represents the pressurizer regions, connected with each other by single junctions components, the bottom volume control is connected to the primary circuit, represented by a time dependent volume component, through a time dependent junction component, which describes the mass flow behavior during surges through surge orifices. The hydrodynamic components representing the pressurizer are surrounded by heat structures, in addition there are heat structures inside the bottom volume control describing the behavior of electrical heaters, that operate in cases of outsurges. The analysis are intended to detail the behavior variables as pressure, temperature and volume of liquid inside the pressurizer during a water surge coming from the primary circuit or a water surge coming from the pressurizer to the primary circuit. (author)
Energy Technology Data Exchange (ETDEWEB)
Medeiros, Eduarda da C.A.; Castrillo, Lazara S., E-mail: e.camedeiros@gmail.com, E-mail: lazara@poli.br [Universidade de Pernambuco, Recife, PE (Brazil). Escola Politecnica. Departamento de Engenharia Mecanica
2015-07-01
Insurge and outsurge phenomena are transient states and could be analyzed by thermodynamics principles, the pressurizer behavior will vary in response to mass flow changes. These surges can occur in the presence of noncondensable gases. On this paper, with the code RELAP5, the IRIS reactor pressurizer is described to analyze surges phenomena in their control volumes with non-condensable gases since they modify the pressure response. A set of three pipes components represents the pressurizer regions, connected with each other by single junctions components, the bottom volume control is connected to the primary circuit, represented by a time dependent volume component, through a time dependent junction component, which describes the mass flow behavior during surges through surge orifices. The hydrodynamic components representing the pressurizer are surrounded by heat structures, in addition there are heat structures inside the bottom volume control describing the behavior of electrical heaters, that operate in cases of outsurges. The analysis are intended to detail the behavior variables as pressure, temperature and volume of liquid inside the pressurizer during a water surge coming from the primary circuit or a water surge coming from the pressurizer to the primary circuit. (author)
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.
The Transient Elliptic Flow of Power-Law Fluid in Fractal Porous Media
Institute of Scientific and Technical Information of China (English)
宋付权; 刘慈群
2002-01-01
The steady oil production and pressure distribution formulae of vertically fractured well for power-law non-Newtonian fluid were derived on the basis of the elliptic flow model in fractal reservoirs. The corresponding transient flow in fractal reservoirs was studied by numerical differentiation method: the influence of fractal index to transient pressure of vertically fractured well was analyzed. Finally the approximate analytical solution of transient flow was given by average mass conservation law. The study shows that using elliptic flow method to analyze the flow of vertically fractured well is a simple method.
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
CANDU 6 liquid injection shutdown system waterhammer analysis using PTRAN
International Nuclear Information System (INIS)
Ko, Deuk Yoon; Kim, Eun Ki; Ko, Yong Sang; Park, Byung Ho; Kim, Seok Bum
1996-06-01
An in-core LOCA could result in flooding of the helium header in the liquid injection shutdown system. Flooding of the helium header will result in severe pressure transients (waterhammer) in the liquid injection shutdown system when the shutdown signal is initiated. To evaluate the impact of the dynamic effects of this event, a pressure transient analysis has been performed. This analysis is performed using PTRAN, which is a computer program based on the method of characteristics. The results of this analysis are used in the stress analysis of the piping and pipe supports to ensure that the liquid injection shutdown system can withstand the pressure transient loadings. This analysis report documents the results of waterhammer analysis performed for the liquid injection shutdown system for the Wolsung nuclear power plant unit 2, 3 and 4. 4 tabs., 11 figs., 15 refs. (Author)
CANDU 6 liquid injection shutdown system waterhammer analysis using PTRAN
Energy Technology Data Exchange (ETDEWEB)
Ko, Deuk Yoon; Kim, Eun Ki; Ko, Yong Sang; Park, Byung Ho; Kim, Seok Bum [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1996-06-01
An in-core LOCA could result in flooding of the helium header in the liquid injection shutdown system. Flooding of the helium header will result in severe pressure transients (waterhammer) in the liquid injection shutdown system when the shutdown signal is initiated. To evaluate the impact of the dynamic effects of this event, a pressure transient analysis has been performed. This analysis is performed using PTRAN, which is a computer program based on the method of characteristics. The results of this analysis are used in the stress analysis of the piping and pipe supports to ensure that the liquid injection shutdown system can withstand the pressure transient loadings. This analysis report documents the results of waterhammer analysis performed for the liquid injection shutdown system for the Wolsung nuclear power plant unit 2, 3 and 4. 4 tabs., 11 figs., 15 refs. (Author).
Simulation of non-isothermal transient flow in gas pipeline
Energy Technology Data Exchange (ETDEWEB)
Ferreira Junior, Luis Carlos; Soares, Matheus; Lima, Enrique Luis; Pinto, Jose Carlos [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Quimica; Muniz, Cyro; Pires, Clarissa Cortes; Rochocz, Geraldo [ChemTech, Rio de Janeiro, RJ (Brazil)
2009-07-01
Modeling of gas pipeline usually considers that the gas flow is isothermal (or adiabatic) and that pressure changes occur instantaneously (quasi steady state approach). However, these assumptions are not valid in many important transient applications (changes of inlet and outlet flows/pressures, starting and stopping of compressors, changes of controller set points, among others). Besides, the gas properties are likely to depend simultaneously on the pipe position and on the operation time. For this reason, a mathematical model is presented and implemented in this paper in order to describe the gas flow in pipeline when pressure and temperature transients cannot be neglected. The model is used afterwards as a tool for reconciliation of available measured data. (author)
Energy Technology Data Exchange (ETDEWEB)
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.
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.
Fission product transport and behavior during two postulated loss of flow transients in the air
International Nuclear Information System (INIS)
Adams, J.P.; Carboneau, M.L.
1991-01-01
This document discusses fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradation, respectively) in the Advanced Test Reactor (ATR). These transients are designated ATR Transient LCPI5 (high-pressure) and LPP9 (low-pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be analyzed. A probabilistic risk assessment indicated that the probability of occurrence for these two transients is of the order of 10 -5 and 10 -7 per reactor year for LCP15 and LPP9, respectively
Pressurized-water-reactor station blackout
International Nuclear Information System (INIS)
Dobbe, C.A.
1983-01-01
The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority
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)
Fujishiro, Toshio
1978-03-01
The computer code PULSE-2 has been developed for the analysis of pressure pulse generation process when hot fuel particles come into contact with the coolant in a fuel rod failure accident. In the program, it is assumed that hot fuel fragments mix with the coolant instantly and homogeneously in the failure region. Then, the rapid vaporization of the coolant and transient pressure rise in failure region, and the movement of ejected coolant slugs are calculated. The effect of a fuel-particle size distribution is taken into consideration. Heat conduction in the fuel particles and heat transfer at fuel-coolant interface are calculated. Temperature, pressure and void fraction in the mixed region are calculated from the average enthalpy. With physical property subroutines for liquid sodium and water, the model is usable for both LMFBR and LWR conditions. (auth.)
Meteorological interpretation of transient LOD changes
Masaki, Y.
2008-04-01
The Earth’s spin rate is mainly changed by zonal winds. For example, seasonal changes in global atmospheric circulation and episodic changes accompanied with El Nĩ os are clearly detected n in the Length-of-day (LOD). Sub-global to regional meteorological phenomena can also change the wind field, however, their effects on the LOD are uncertain because such LOD signals are expected to be subtle and transient. In our previous study (Masaki, 2006), we introduced atmospheric pressure gradients in the upper atmosphere in order to obtain a rough picture of the meteorological features that can change the LOD. In this presentation, we compare one-year LOD data with meteorological elements (winds, temperature, pressure, etc.) and make an attempt to link transient LOD changes with sub-global meteorological phenomena.
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
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.
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
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.
Hyman, D.; Bursik, M. I.; Pitman, E. B.
2017-12-01
The collapse or explosive breakup of growing and degassing lava domes presents a significant hazard due to the generation of dense, mobile pyroclastic flows as well as the wide dispersal of dense ballistic blocks. Lava dome stability is in large part governed by the balance of transport and storage of gas within the pore space. Because pore pressurization reduces the effective stress within a dome, the transient distribution of elevated gas pressure is critically important to understanding dome break up. We combine mathematical and numerical analyses to gain a better understanding of the temporal variation in gas flow and storage within the dome system. In doing so, we develop and analyze new governing equations describing nonlinear gas pressure diffusion in a deforming dome with an evolving porosity field. By relating porosity, permeability, and pressure, we show that the flux of gas through a dome is highly sensitive to the porosity distribution and viscosity of the lava, as well as the timescale and magnitude of the gas supply. The numerical results suggest that the diffusion of pressure and porosity variations play an integral role in the cyclic growth and destruction of small domes.The nearly continuous cycles of lava dome growth, pressurization, and failure that have characterized the last two decades of eruptive history at Volcán Popocatépetl, Mexico provide excellent natural data with which to compare new models of transient dome pressurization. At Popocatépetl, periodic pressure increases brought on by changes in gas supply into the base of the dome may play a role in its cyclic growth and destruction behavior. We compare our model of cyclic pressurization with lava dome survival data from Popocatépetl. We show that transient changes in pore pressure explain how small lava domes evolve to a state of criticality before explosion or collapse. Additionally, numerical analyses presented here suggest that short-term oscillations cannot arise within the dome
Transient behavior of ASTRID with a gas power conversion system
Energy Technology Data Exchange (ETDEWEB)
Bertrand, F., E-mail: frederic.bertrand@cea.fr; Mauger, G.; Bensalah, M.; Gauthé, P.
2016-11-15
Highlights: • CATHARE2 transient calculations have been performed for ASTRID with a gas PCS. • The behavior of the reactor is close for gas and for water PCS in case of LOOP. • The gas PCS enables to cool the core for at least 10 h for pressurized transients. • The depressurization of the PCS induces an over-cooling for breaches on low pressure pipes. • The spurious opening of a by-pass line of the turbomachine can be controlled without scram. - Abstract: The present article is dedicated to preliminary transient studies carried out for the analysis of the system overall behavior of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) demonstrator developed in France by CEA and its industrial partners. ASTRID is foreseen to demonstrate the progress made in SFR technology at an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability. Among the innovative options, a gas power conversion systems (PCS) is envisaged. In this innovative PCS, the working gas is nitrogen whose flow rate delivers power to a turbine driving with the same shaft two compressors (low and high pressure) separated by an intercooler. The other part of the work delivered by the gas is used to drive the alternator that produces electricity. The main objective of such a PCS consists in avoiding physically the possibility of a sodium/water reaction with the secondary circuit but the impact of this PCS on the control of incidental and accidental transients has also been studied. The main purpose of the studies presented in the paper is to assess the dynamic behavior of ASTRID including a gas PCS with the CATHARE2 code. The first transient presented deals with a loss of off-site power and has been calculated for the gas PCS but also for a classical steam/water PCS for comparison purpose. Then typical transients of gas system have been investigated. Several families of
Transient behavior of ASTRID with a gas power conversion system
International Nuclear Information System (INIS)
Bertrand, F.; Mauger, G.; Bensalah, M.; Gauthé, P.
2016-01-01
Highlights: • CATHARE2 transient calculations have been performed for ASTRID with a gas PCS. • The behavior of the reactor is close for gas and for water PCS in case of LOOP. • The gas PCS enables to cool the core for at least 10 h for pressurized transients. • The depressurization of the PCS induces an over-cooling for breaches on low pressure pipes. • The spurious opening of a by-pass line of the turbomachine can be controlled without scram. - Abstract: The present article is dedicated to preliminary transient studies carried out for the analysis of the system overall behavior of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) demonstrator developed in France by CEA and its industrial partners. ASTRID is foreseen to demonstrate the progress made in SFR technology at an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability. Among the innovative options, a gas power conversion systems (PCS) is envisaged. In this innovative PCS, the working gas is nitrogen whose flow rate delivers power to a turbine driving with the same shaft two compressors (low and high pressure) separated by an intercooler. The other part of the work delivered by the gas is used to drive the alternator that produces electricity. The main objective of such a PCS consists in avoiding physically the possibility of a sodium/water reaction with the secondary circuit but the impact of this PCS on the control of incidental and accidental transients has also been studied. The main purpose of the studies presented in the paper is to assess the dynamic behavior of ASTRID including a gas PCS with the CATHARE2 code. The first transient presented deals with a loss of off-site power and has been calculated for the gas PCS but also for a classical steam/water PCS for comparison purpose. Then typical transients of gas system have been investigated. Several families of
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
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)
Pitot tube and drag body measurements in transient steam--water flows
International Nuclear Information System (INIS)
Fincke, J.R.; Deason, V.A.; Dacus, M.W.
1979-01-01
The use of full-flow drag devices and rakes of water-cooled Pitot tubes to measure the transient two-phase mass flow during loss-of-coolant experiments in pressurized water reactor (PWR) environments has been developed. Mass flow rate measurements have been obtained in high temperature and pressure environments, similar to PWRs, under transient conditions. Comparisons of the measured time integrated value of mass flow to the known system mass before depressurization are made
International Nuclear Information System (INIS)
Moon, S.K.; Chun, S.Y.; Choi, K.Y.; Yang, S.K.
2001-01-01
An experimental study on transient critical heat flux (CHF) under flow coast-down has been performed for water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady state CHF. The transient CHF experiments have been performed for three kinds of flow transient modes based on the coast-down data of the Kori 3/4 nuclear power plant reactor coolant pump. Most of the CHFs occurred in the annular-mist flow regime. Thus, it means that the possible CHF mechanism might be the liquid film dryout in the annular-mist flow regime. For flow transient mode with the smallest flow reduction rate, the time-to-CHF is the largest. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to-CHF becomes large as the heat flux decreases. Usually, the critical mass flux is large for slow flow reduction. There is a pressure effect on the ratio of the transient CHF data to steady state CHF data. Some conventional correlations show relatively better CHF prediction results for high system pressure, high quality and slow transient modes than for low system pressure, low quality and fast transient modes. (author)
International Nuclear Information System (INIS)
Lan, G; Jiang, J; Li, D D; Yi, W S; Zhao, Z; Nie, L N
2013-01-01
The calculation of water-hammer pressure phenomenon of single-phase liquid is already more mature for a pipeline of uniform characteristics, but less research has addressed the calculation of slurry water hammer pressure in complex pipelines with slurry flows carrying solid particles. In this paper, based on the developments of slurry pipelines at home and abroad, the fundamental principle and method of numerical simulation of transient processes are presented, and several boundary conditions are given. Through the numerical simulation and analysis of transient processes of a practical engineering of long-distance slurry transportation pipeline system, effective protection measures and operating suggestions are presented. A model for calculating the water impact of solid and fluid phases is established based on a practical engineering of long-distance slurry pipeline transportation system. After performing a numerical simulation of the transient process, analyzing and comparing the results, effective protection measures and operating advice are recommended, which has guiding significance to the design and operating management of practical engineering of longdistance slurry pipeline transportation system
Lan, G.; Jiang, J.; Li, D. D.; Yi, W. S.; Zhao, Z.; Nie, L. N.
2013-12-01
The calculation of water-hammer pressure phenomenon of single-phase liquid is already more mature for a pipeline of uniform characteristics, but less research has addressed the calculation of slurry water hammer pressure in complex pipelines with slurry flows carrying solid particles. In this paper, based on the developments of slurry pipelines at home and abroad, the fundamental principle and method of numerical simulation of transient processes are presented, and several boundary conditions are given. Through the numerical simulation and analysis of transient processes of a practical engineering of long-distance slurry transportation pipeline system, effective protection measures and operating suggestions are presented. A model for calculating the water impact of solid and fluid phases is established based on a practical engineering of long-distance slurry pipeline transportation system. After performing a numerical simulation of the transient process, analyzing and comparing the results, effective protection measures and operating advice are recommended, which has guiding significance to the design and operating management of practical engineering of longdistance slurry pipeline transportation system.
CFD simulation analysis and validation for CPR1000 pressurized water reactor
International Nuclear Information System (INIS)
Zhang Mingqian; Ran Xiaobing; Liu Yanwu; Yu Xiaolei; Zhu Mingli
2013-01-01
Background: With the rapid growth in the non-nuclear area for industrial use of Computational fluid dynamics (CFD) which has been accompanied by dramatically enhanced computing power, the application of CFD methods to problems relating to Nuclear Reactor Safety (NRS) is rapidly accelerating. Existing research data have shown that CFD methods could predict accurately the pressure field and the flow repartition in reactor lower plenum. But simulations for the full domain of the reactor have not been reported so far. Purpose: The aim is to determine the capabilities of the codes to model accurately the physical phenomena which occur in the full reactor vessel. Methods: The flow field of the CPR1000 reactor which is associated with a typical pressurized water reactor (PWR) is simulated by using ANSYS CFX. The pressure loss in reactor pressure vessel, the hydraulic loads of guide tubes and support columns, and the bypass flow of head dome were obtained by calculations for the full domain of the reactor. The results were validated by comparing with the determined reference value of the operating nuclear plant (LingAo nuclear plant), and the transient simulation was conducted in order to better understand the flow in reactor pressure vessel. Results: It was shown that the predicted pressure loss with CFD code was slightly different with the determined value (10% relative deviation for the total pressure loss), the hydraulic loads were less than the determined value with maximum relative deviation 50%, and bypass flow of head dome was approximately the same with determined value. Conclusion: This analysis practice predicts accurately the physical phenomena which occur in the full reactor vessel, and can be taken as a guidance for the nuclear plant design development and improve our understanding of reactor flow phenomena. (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.)
Investigation of transient cavitating flow in viscoelastic pipes
International Nuclear Information System (INIS)
Keramat, A; Tijsseling, A S; Ahmadi, A
2010-01-01
A study on water hammer in viscoelastic pipes when the fluid pressure drops to liquid vapour pressure is performed. Two important concepts including column separation and the effects of retarded strains in the pipe wall on the fluid response have been investigated separately in recent works, but there is some curiosity as to how the results for pressure and discharge are when column separation occurs in a viscoelastic pipe. For pipes made of plastic such as polyethylene (PE) and polyvinyl chloride (PVC), viscoelasticity is a crucial mechanical property which changes the hydraulic and structural transient responses. Based on previous developments in the analysis of water hammer, a model which is capable of analysing column separation in viscoelastic pipes is presented and used for solving the selected case studies. For the column-separation modelling the Discrete Vapour Cavity Model (DVCM) is utilised and the viscoelasticity property of the pipe wall is modelled by Kelvin-Voigt elements. The effects of viscoelasticity play an important role in the column separation phenomenon because it changes the water hammer fundamental frequency and so affects the time of opening or collapse of the cavities. Verification of the implemented computer code is performed for the effects of viscoelasticity and column separation - separately and simultaneously - using experimental results from the literature. In the provided examples the focus is placed on the simultaneous effect of viscoelasticity and column separation on the hydraulic transient response. The final conclusions drawn are that if rectangular grids are utilised the DVCM gives acceptable predictions of the phenomenon and that the pipe wall material's retarded behaviour strongly dampens the pressure spikes caused by column separation.
Investigation of transient cavitating flow in viscoelastic pipes
Keramat, A.; Tijsseling, A. S.; Ahmadi, A.
2010-08-01
A study on water hammer in viscoelastic pipes when the fluid pressure drops to liquid vapour pressure is performed. Two important concepts including column separation and the effects of retarded strains in the pipe wall on the fluid response have been investigated separately in recent works, but there is some curiosity as to how the results for pressure and discharge are when column separation occurs in a viscoelastic pipe. For pipes made of plastic such as polyethylene (PE) and polyvinyl chloride (PVC), viscoelasticity is a crucial mechanical property which changes the hydraulic and structural transient responses. Based on previous developments in the analysis of water hammer, a model which is capable of analysing column separation in viscoelastic pipes is presented and used for solving the selected case studies. For the column-separation modelling the Discrete Vapour Cavity Model (DVCM) is utilised and the viscoelasticity property of the pipe wall is modelled by Kelvin-Voigt elements. The effects of viscoelasticity play an important role in the column separation phenomenon because it changes the water hammer fundamental frequency and so affects the time of opening or collapse of the cavities. Verification of the implemented computer code is performed for the effects of viscoelasticity and column separation - separately and simultaneously - using experimental results from the literature. In the provided examples the focus is placed on the simultaneous effect of viscoelasticity and column separation on the hydraulic transient response. The final conclusions drawn are that if rectangular grids are utilised the DVCM gives acceptable predictions of the phenomenon and that the pipe wall material's retarded behaviour strongly dampens the pressure spikes caused by column separation.
TRAC-PF1 analysis of LOFT steam-generator feedwater transient test L9-1
International Nuclear Information System (INIS)
Meier, J.K.
1983-01-01
The Transient Reactor Analysis Code (TRAC-PF1) calculations were compared to test data from Loss-of-Fluid Test (LOFT) L9-1, which was a loss-of-feedwater transient. This paper includes descriptions of the test and the TRAC input and compares the TRAC-calculated results with the test data. We conclude that the code predicted the experiment well, given the uncertainties in the boundary conditions. The analysis indicates the need to model all the flow paths and heat structures, and to improve the TRAC wall condensation heat-transfer model
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
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
PWR station blackout transient simulation in the INER integral system test facility
International Nuclear Information System (INIS)
Liu, T.J.; Lee, C.H.; Hong, W.T.; Chang, Y.H.
2004-01-01
Station blackout transient (or TMLB' scenario) in a pressurized water reactor (PWR) was simulated using the INER Integral System Test Facility (IIST) which is a 1/400 volumetrically-scaled reduce-height and reduce-pressure (RHRP) simulator of a Westinghouse three-loop PWR. Long-term thermal-hydraulic responses including the secondary boil-off and the subsequent primary saturation, pressurization and core uncovery were simulated based on the assumptions of no offsite and onsite power, feedwater and operator actions. The results indicate that two-phase discharge is the major depletion mode since it covers 81.3% of the total amount of the coolant inventory loss. The primary coolant inventory has experienced significant re-distribution during a station blackout transient. The decided parameter to avoid the core overheating is not the total amount of the coolant inventory remained in the primary core cooling system but only the part of coolant left in the pressure vessel. The sequence of significant events during transient for the IIST were also compared with those of the ROSA-IV large-scale test facility (LSTF), which is a 1/48 volumetrically-scaled full-height and full-pressure (FHFP) simulator of a PWR. The comparison indicates that the sequence and timing of these events during TMLB' transient studied in the RHRP IIST facility are generally consistent with those of the FHFP LSTF. (author)
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.
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
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.
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.
Transient pattern analysis for fault detection and diagnosis of HVAC systems
International Nuclear Information System (INIS)
Cho, Sung-Hwan; Yang, Hoon-Cheol; Zaheer-uddin, M.; Ahn, Byung-Cheon
2005-01-01
Modern building HVAC systems are complex and consist of a large number of interconnected sub-systems and components. In the event of a fault, it becomes very difficult for the operator to locate and isolate the faulty component in such large systems using conventional fault detection methods. In this study, transient pattern analysis is explored as a tool for fault detection and diagnosis of an HVAC system. Several tests involving different fault replications were conducted in an environmental chamber test facility. The results show that the evolution of fault residuals forms clear and distinct patterns that can be used to isolate faults. It was found that the time needed to reach steady state for a typical building HVAC system is at least 50-60 min. This means incorrect diagnosis of faults can happen during online monitoring if the transient pattern responses are not considered in the fault detection and diagnosis analysis
International Nuclear Information System (INIS)
Chen, Mingya; Lu, Feng; Wang, Rongshan; Ren, Ai
2014-01-01
Highlights: • The regulation and the code are proved to be conservative in the integrity assessment. • This study is helpful to understand the complex influence of the parameters. • The most dangerous case is given for the reference transient. - Abstract: Fracture mechanics analysis of pressurized thermal shock (PTS) is the key element of the integrity evaluation of the nuclear reactor pressure vessel (RPV). While the regulation of 10 CFR 50.61 and the ASME Code provide the guidance for the structural integrity, the guidance has been prepared under conservative assumptions. In this paper, the effects of conservative assumptions involved in the PTS analysis were investigated. The influence of different parameters, such as crack size, cladding effect and neutron fluence, were reviewed based on 3-D finite element analyses. Also, the sensitivity study of elastic–plastic approach, crack type and cladding thickness were reviewed. It was shown that crack depth, crack type, plastic effect and cladding thickness change the safety margin (SM) significantly, and the SM at the deepest point of the crack is not always smaller than that of the surface point, indicating that both the deepest and surface points of the crack front should be considered. For the reference transient, deeper cracks always give more conservative prediction. So compared to the prescribed analyses of a set of postulated defects with varying depths in the ASME code, it only needs to assess the crack with maximum depth in the code for the reference transient according to the conclusions
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
Soil-structure interaction for transient loads due to safety relief valve discharges
International Nuclear Information System (INIS)
Tseng, W.S.; Tsai, N.C.
1978-01-01
Dynamic responses of BWR Mark II containment structures subjected to axisymmetric transient pressure loadings due to simultaneous safety relief valve discharges were investigated using finite element analysis, including the soil-structure interaction effect. To properly consider the soil-structure interaction effect, a simplified lumped parameter foundation model and axisymmetric finite element foundation model with viscous boundary impedance are used. Analytical results are presented to demonstrate the effectiveness of the simplified foundation model and to exhibit the dynamic response behavior of the structure as the transient loading frequency and the foundation rigidity vary. The impact of the dynamic structural response due to this type of loading on the equipment design is also discussed. (Auth.)
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)
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.
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
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)
Dynamic behaviour of mono bucket foundations subjected to combined transient loading
DEFF Research Database (Denmark)
Nielsen, Søren Dam; Ibsen, Lars Bo; Nielsen, Benjaminn Nordahl
2015-01-01
This article presents the results from small scale testing, investigating the effect of transient combined loading of a bucketfoundation. The tests are performed inside a pressure tank at Aalborg University, Denmark. The bucket foundation was installed in dense water saturated sand and transient ...
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
International Nuclear Information System (INIS)
Tas, Fatma Burcu; Ergun, Sule
2013-01-01
Highlights: • Fuel performance of a typical Pressurized Water Reactor rod is analyzed. • Steady state fuel rod behavior is examined to see the effects of pellet to cladding gap thickness and gap gas pressure. • Transient fuel rod behavior is examined to see the effects of pellet to cladding gap thickness and gap gas pressure. • The optimum pellet to cladding gap thickness and gap gas pressure values of the simulated fuel are determined. • The effects of pellet to cladding gap design parameters on nuclear fuel reliability are examined. - Abstract: As an important improvement in the light water nuclear reactor operations, the nuclear fuel burnup rate is increased in recent decades and this increase causes heavier duty for the nuclear fuel. Since the high burnup fuel is exposed to very high thermal and mechanical stresses and since it operates in an environment with high radiation for about 18 month cycles, it carries the risk of losing its integrity. In this study; it is aimed to determine the effects of pellet–cladding gap thickness and gap pressure on reliability of high burnup nuclear fuel in Pressurized Water Reactors (PWRs) under steady state operation conditions and suggest optimum values for the examined parameters only and validate these suggestions for a transient condition. In the presented study, fuel performance was analyzed by examining the effects of pellet–cladding gap thickness and gap pressure on the integrity of high burnup fuels. This work is carried out for a typical Westinghouse type PWR fuel. The steady state conditions were modeled and simulated with FRAPCON-3.4a steady state fuel performance code and the FRAPTRAN-1.4 fuel transient code was used to calculate transient fuel behavior. The analysis included the changes in the important nuclear fuel design limitations such as the centerline temperature, cladding stress, strain and oxidation with the change in pellet–cladding gap thickness and initial pellet–cladding gap gas
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 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
Analysis of transient thermal response in the outlet plenum of an LMFBR
International Nuclear Information System (INIS)
Yang, J.W.
1976-05-01
A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the upper outlet plenum of an LMFBR. The one-dimensional turbulent jet flow equations were solved to determine the maximum penetration of the core flow. The maximum penetration is used as the criterion for dividing the sodium region into two mixing zones. The lumped-parameter model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of bypass flow into the plenum. Numerical calculations were performed for two cases. The first case corresponds to a normal scram followed by flow coast-down. The second case represents the double-ended pipe rupture at the inlet of cold leg followed by reactor scram. The results indicate that effects of flow stratification, chimney height, metal heat capacity and bypass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas does not play any significant role on sodium temperature. This two-zone mixing model will be a part of the thermohydraulic transient code SSC
Transient 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
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.
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)
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)
Discharge models through the pressurizer valves
International Nuclear Information System (INIS)
Madeira, A.A.
1985-01-01
A reliable estimate of discharge through the pressurizer relief and safety valves is of concern to adequately predict the behaviour of RCS pressure during transients. It's investigated the discharge models used by the ALMOD code, and to implement alternative models from the available literature, which are recommended for different conditions of flow that shall exist during transients requiring discharge through the relief and safety valves. (Author) [pt
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
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
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)
Park, Yong-Chan; Song, Dong-Soo; Jun, Hwang-Yong
2006-01-01
The Main steam line break(MSLB) occurring inside a reactor containment structure may result in significant releases of high energy fluid to the containment, possibly result in high containment pressure and temperature. The MSLB accident, along with the Loss Of Coolant Accident, is a design basis accident for determining the peak containment pressure and temperature. The analysis for a MSLB for inside containment should be performed to justify the structural integrity and equipment qualification in accordance with revision 1 of Reg. Guide 1.89. Rev1(1984), which is also required as part of obtaining the extended operating license for WestingHouse(WH) 3-Loops Nuclear Power Plant(NPP). Now, the WH NPP has been performed power uprating. Therefore, all initial conditions, setpoints and uncertainties were considered with MSLB analysis for environment qualification(EQ). The transient was analyzed to determine the worst set of mass and energy releases that impact the EQ aspects of safety related equipment inside containment. The most limiting single failure in this event was determined by a sensitivity study. The MSLB event was analyzed for a full set of power conditions and break sizes
Baumeister, K. J.
1983-01-01
A time-dependent finite difference formulation to the inhomogeneous wave equation is derived for plane wave propagation with harmonic noise sources. The difference equation and boundary conditions are developed along with the techniques to simulate the Dirac delta function associated with a concentrated noise source. Example calculations are presented for the Green's function and distributed noise sources. For the example considered, the desired Fourier transformed acoustic pressures are determined from the transient pressures by use of a ramping function and an integration technique, both of which eliminates the nonharmonic pressure associated with the initial transient.
Baumeiste, K. J.
1983-01-01
A time-dependent finite difference formulation to the inhomogeneous wave equation is derived for plane wave propagation with harmonic noise sources. The difference equation and boundary conditions are developed along with the techniques to simulate the Dirac delta function associated with a concentrated noise source. Example calculations are presented for the Green's function and distributed noise sources. For the example considered, the desired Fourier transformed acoustic pressures are determined from the transient pressures by use of a ramping function and an integration technique, both of which eliminates the nonharmonic pressure associated with the initial transient.
Energy Technology Data Exchange (ETDEWEB)
Kim, Young Ae; Kim, Chang Hyun; Kweon, Gab Joo; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)
2007-10-15
The pressurizer safety valves (PSV) in Pressurized Water Reactors are required to provide the overpressure protection for the Reactor Coolant System (RCS) during the overpressure transients. Korea Hydro and Nuclear Power Company (KHNP) plans to build the PSV test facility for the purpose of providing the PSV pop-up characteristics and the loop seal dynamics for the new safety analysis. When the pressurizer safety valve is mounted in a loop seal configuration, the valve must initially pass the loop seal water prior to popping open on steam. The loop seal in the upstream of PSV prevents leakage of hydrogen gas or steam through the safety valve seat. This paper studies on the loop seal clearing dynamics using GOTHIC-7.2a code to verify the effects of loop seal purge time on the reactor coolant system overpressure.
Statistical study of particle acceleration in the core of foreshock transients
Liu, Terry Z.; Angelopoulos, Vassilis; Hietala, Heli; Wilson III, Lynn B.
2017-01-01
Several types of foreshock transients upstream of Earth's bow shock possessing a tenuous, hot core have been observed and simulated. Because of the low dynamic pressure in their cores, these phenomena can significantly disturb the bow shock and the magnetosphere-ionosphere system. Recent observations have also demonstrated that foreshock transients can accelerate particles which, when transported earthward, can affect space weather. Understanding the potential of foreshock transients to accel...
Lin, Yung-Hsu
The goal of this dissertation is to study high pressure streamers in air and apply it to diesel engine technologies. Nanosecond scale pulsed high voltage discharges in air/fuel mixtures can generate radicals which in turn have been shown to improve combustion efficiency in gasoline fueled internal combustion engines. We are exploring the possibility to extend such transient plasma generation and expected radical species generation to the range of pressures encountered in compression-ignition (diesel) engines having compression ratios of ˜20:1, thereby improving lean burning efficiency and extending the range of lean combustion. At the beginning of this dissertation, research into streamer discharges is reviewed. Then, we conducted experiments of streamer propagation at high pressures, calculated the streamer velocity based on both optical and electrical measurements, and the similarity law was checked by analyzing the streamer velocity as a function of the reduced electric field, E/P. Our results showed that the similarity law is invalid, and an empirical scaling factor, E/√P, is obtained and verified by dimensional analysis. The equation derived from the dimensional analysis will be beneficial to proper electrode and pulse generator design for transient plasma assisted internal engine experiments. Along with the high pressure study, we applied such technique on diesel engine to improve the fuel efficiency and exhaust treatment. We observed a small effect of transient plasma on peak pressure, which implied that transient plasma has the capability to improve the fuel consumption. In addition, the NO can be reduced effectively by the same technique and the energy cost is 30 eV per NO molecule.
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.
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.
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 a small break loss-of-coolant accident of pressurized water reactor by APROS
Energy Technology Data Exchange (ETDEWEB)
Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)
1995-09-01
The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.
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.
A coupled inversion of pressure and surface displacement
International Nuclear Information System (INIS)
Vasco, D.W.; Karasaki, Kenzi; Kishida, Kiyoshi
2001-01-01
A coupled inversion of transient pressure observations and surface displacement measurements provides an efficient technique for estimating subsurface permeability variations. The methodology has the advantage of utilizing surface observations, which are typically much less expensive than measurements requiring boreholes. Furthermore, unlike many other geophysical observables, the relationship between surface deformation and reservoir pore fluid volume changes is relatively well understood. Our treatment enables us to partition the estimation problem into a sequence of three linear sub-problems. An application of the approach to a set of tilt and borehole pressure data from the Raymond field site in California illustrates it's efficiency and utility. The observations are associated with a well test in which fluid is withdrawn from a shallow fracture zone. During the test thirteen tiltmeters recorded the movement of the ground surface. Simultaneously, nine transducers measured pressure changes in boreholes intersecting the fracture system. We are able to image a high permeability, north trending channel located within the fracture zone. The existence and orientation of this high permeability feature is substantiated by a semi-quantitative analysis of some 4,000 transient pressure curves. (author)
International Nuclear Information System (INIS)
Huang, W D; Fan, H G; Chen, N X
2012-01-01
To study the interaction between the transient flow in pipe and the unsteady turbulent flow in turbine, a coupled model of the transient flow in the pipe and three-dimensional unsteady flow in the turbine is developed based on the method of characteristics and the fluid governing equation in the accelerated rotational relative coordinate. The load-rejection process under the closing of guide vanes of the hydraulic power plant is simulated by the coupled method, the traditional transient simulation method and traditional three-dimensional unsteady flow calculation method respectively and the results are compared. The pressure, unit flux and rotation speed calculated by three methods show a similar change trend. However, because the elastic water hammer in the pipe and the pressure fluctuation in the turbine have been considered in the coupled method, the increase of pressure at spiral inlet is higher and the pressure fluctuation in turbine is stronger.
Huang, W. D.; Fan, H. G.; Chen, N. X.
2012-11-01
To study the interaction between the transient flow in pipe and the unsteady turbulent flow in turbine, a coupled model of the transient flow in the pipe and three-dimensional unsteady flow in the turbine is developed based on the method of characteristics and the fluid governing equation in the accelerated rotational relative coordinate. The load-rejection process under the closing of guide vanes of the hydraulic power plant is simulated by the coupled method, the traditional transient simulation method and traditional three-dimensional unsteady flow calculation method respectively and the results are compared. The pressure, unit flux and rotation speed calculated by three methods show a similar change trend. However, because the elastic water hammer in the pipe and the pressure fluctuation in the turbine have been considered in the coupled method, the increase of pressure at spiral inlet is higher and the pressure fluctuation in turbine is stronger.
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.
Directory of Open Access Journals (Sweden)
Norazlina Subani
2015-01-01
Full Text Available Water hammer on transient flow of hydrogen-natural gas mixture in a horizontal pipeline is analysed to determine the relationship between pressure waves and different modes of closing and opening of valves. Four types of laws applicable to closing valve, namely, instantaneous, linear, concave, and convex laws, are considered. These closure laws describe the speed variation of the hydrogen-natural gas mixture as the valve is closing. The numerical solution is obtained using the reduced order modelling technique. The results show that changes in the pressure wave profile and amplitude depend on the type of closing laws, valve closure times, and the number of polygonal segments in the closing function. The pressure wave profile varies from square to triangular and trapezoidal shape depending on the type of closing laws, while the amplitude of pressure waves reduces as the closing time is reduced and the numbers of polygonal segments are increased. The instantaneous and convex closing laws give rise to minimum and maximum pressure, respectively.
International Nuclear Information System (INIS)
Strydom, G.; Reitsma, F.; Ngeleka, P.T.; Ivanov, K.N.
2010-01-01
The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated, and code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MWth design and a representative set of transient exercises are defined as an OECD benchmark. The scope of the benchmark is to establish a series of well defined multi-dimensional computational benchmark problems with a common given set of cross sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. This paper describes the current status of the benchmark project and shows the results for the six transient exercises, consisting of three Loss of Cooling Accidents, two Control Rod Withdrawal transients, a power load-follow transient, and a Helium over-cooling Accident. The participants' results are compared using a statistical method and possible areas of future code improvement are identified. (authors)
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
International Nuclear Information System (INIS)
Peterson, C.E.; Gose, G.C.; McFadden, J.H.
1983-01-01
RETRAN-02 represents a significant achievement in the development of a versatile and reliable computer program for use in best estimate transient thermal-hydraulic analysis of light water reactor systems. The RETRAN-02 computer program is an extension of the RETRAN-01 program designed to provide analysis capabilities for 1) BWR and PWR transients, 2) small break loss of coolant accidents, 3) balance of plant modeling, and 4) anticipated transients without scram, while maintaining the analysis capabilities of the predecessor code. The RETRAN-02 computer code is constructed in a semimodular and dynamic dimensioned form where additions to the code can be easily carried out as new and improved models are developed. This report (the fourth of a five volume computer code manual) describes the verification and validation of RETRAN-02
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
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)
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)
Asad, Usman; Tjong, Jimi; Zheng, Ming
2014-01-01
Highlights: • Zero-dimensional EGR model for transient diesel combustion control. • Detailed analysis of EGR effects on intake, cylinder charge and exhaust properties. • Intake oxygen validated as an operating condition-independent measure of EGR. • Quantified EGR effectiveness in terms of NOx emission reduction. • Twin lambda sensor technique for estimation of EGR/in-cylinder parameters. - Abstract: The application of exhaust gas recirculation (EGR) during transient engine operation is a challenging task since small fluctuations in EGR may cause larger than acceptable spikes in NOx/soot emissions or deterioration in the combustion efficiency. Moreover, the intake charge dilution at any EGR ratio is a function of engine load and intake pressure, and typically changes during transient events. Therefore, the management of EGR during transient engine operation or advanced combustion cycles (that are inherently less stable) requires a fundamental understanding of the transient EGR behaviour and its impact on the intake charge development. In this work, a zero-dimensional EGR model is described to estimate the transient (cycle-by-cycle) progression of EGR and the time (engine cycles) required for its stabilization. The model response is tuned to a multi-cylinder engine by using an overall engine system time-constant and shown to effectively track the transient EGR changes. The impact of EGR on the actual air–fuel ratio of the cylinder charge is quantified by defining an in-cylinder excess-air ratio that accounts for the oxygen in the recycled exhaust gas. Furthermore, a twin lambda sensor (TLS) technique is implemented for tracking the intake dilution and in-cylinder excess-air ratio in real-time. The modelling and analysis results are validated against a wide range of engine operations, including transient and steady-state low temperature combustion tests
Development of MCP transient operation strategy for the SMART-P
International Nuclear Information System (INIS)
Yoo, S. E.; Choi, B. S.; Kang, H. O.; Yoon, J. H.; Ji, S. K.
2003-01-01
SMART-P MCP(Main Coolant Pump) transient operation strategies are developed. A Modular Modeling System (MMS) computer code is used for the evaluation of the developed operation strategies. In the SMART-P, normal operating modes are classified into MCP high speed(3600 rpm) mode and MCP low speed mode. Also, natural circulation mode is defined as a performance test case. MCP operation transients occur when changing modes from one to another, and system parameters(core power, system pressure, temperature) are having transients. These transients affect on system performance and, in some cases, limit system operation. In this study, MCP operation strategies are developed and obtained acceptable results
International Nuclear Information System (INIS)
Cerullo, N.; Delli Gatti, A.; Marinelli, M.; Mazzini, M.; Mazzoni, A.; Sbrana, A.; Todisco, P.
1977-01-01
The SOPRE-1 test facility is an integral model (scale 1:13) of a MARK II pressure suppression containment system. It was set up at the University of Pisa in order to study the pressure-temperature transient in pressure suppression containment systems during LOCAs. Knowledge of this transient is necessary to perform a correct structural analysis of reactor containment. The containment system behavior is studied by changing the principal parameters which affect the transient (blow-down mass and energy release, suppression pool water temperature, vent pipe number and submergence, heat transfer coefficients). The first series of tests involved: A) 13 tests with break area of 1.8 cm 2 , B) 8 tests with break area of 20.0 cm 2 . The following experimental conditions were changed: position of the simulated break (from liquid or steam zone), water pressure (20-85 Kg/cm 2 ) and mass (45-70 Kg) in the vessel model. Tests A): the CONTEMPT codes correctly forecast the pressure-temperature history, both in dry- and in wet-well. Tests B): the experimental runs have shown that increasing of blow-down flowrate produces dry-well pressure spatial differences and anomalous vent pipe behavior. This results in damped oscillations of dry- and wet-well pressure, probably due to alternating air bubble over-expansion and collapse, and in vent pipe opening and reclosing. Dry-well pressure maxima at the end of blow-down are greater than those forecasted by currently applied codes: these codes use an homogeneous model, and do not take into account the above mentioned dynamic phenomena. In some tests other interesting phenomena were observed, such as some local pressure peaks in the suppression pool greater than dry-well pessure maxima at the end of blow-down. At present, all these phenomena are under study; they could be important for the structural analysis of containment systems
Directory of Open Access Journals (Sweden)
Andrzej Rusek
2008-01-01
Full Text Available The mathematical model of cylindrical linear induction motor (C-LIM fed via frequency converter is presented in the paper. The model was developed in order to analyze numerically the transient states. Problems concerning dynamics of ac-machines especially linear induction motor are presented in [1 – 7]. Development of C-LIM mathematical model is based on circuit method and analogy to rotary induction motor. The analogy between both: (a stator and rotor windings of rotary induction motor and (b winding of primary part of C-LIM (inductor and closed current circuits in external secondary part of C-LIM (race is taken into consideration. The equations of C-LIM mathematical model are presented as matrix together with equations expressing each vector separately. A computational analysis of selected transient states of C-LIM fed via frequency converter is presented in the paper. Two typical examples of C-LIM operation are considered for the analysis: (a starting the motor at various static loads and various synchronous velocities and (b reverse of the motor at the same operation conditions. Results of simulation are presented as transient responses including