Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases
Energy Technology Data Exchange (ETDEWEB)
Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1999-12-31
A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)
Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases
Energy Technology Data Exchange (ETDEWEB)
Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1998-12-31
A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)
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.
THEBES: a thermal hydraulic code for the calculation of transient two phase flow in bundle geometry
International Nuclear Information System (INIS)
Camous, F.
1983-01-01
The three dimensional thermal hydraulic code THEBES, capable to calculate transient boiling of sodium in rod bundles is described here. THEBES, derived from the transient single phase code SABRE-2A, was developed in CADARACHE by the SIES to analyse the SCARABEE N loss of flow experiments. This paper also presents the results of tests which were performed against various types of experiments: (1) transient boiling in a 7 pin bundle simulating a partial blockage at the bottom of a subassembly (rapid transient SCARABEE 7.2 experiment), (2) transient boiling in a 7 pin bundle simulating a coolant coast down (slow transient SCARABEE 7.3 experiment), (3) steady local and generalised boiling in a 19 pin bundle (GR 19 I experiment), (4) transient boiling in a 19 pin bundle simulating a coolant coast down (GR 19 I experiment), (5) steady local boiling in a 37 pin bundle with internal blockage (MOL 7C experiment). Excellent agreement was found between calculated and experimental results for these different situations. Our conclusion is that THEBES is able to calculate transient boiling of sodium in rod bundles in a quite satisfying way
International Nuclear Information System (INIS)
Chenu, A.
2011-10-01
Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the
International Nuclear Information System (INIS)
2007-01-01
In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as current applications. (authors) Recently developed best-estimate computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present volume is a follow-up to the first two volumes. While the first described the specification of the benchmark, the second presented the results of the first exercise that identified the key parameters and important issues concerning the thermal-hydraulic system modelling of the simulated transient caused by the switching on of a main coolant pump when the other three were in operation. Volume 3 summarises the results for Exercise 2 of the benchmark that identifies the key parameters and important issues concerning the 3-D neutron kinetics modelling of the simulated transient. These studies are based on an experiment that was conducted by Bulgarian and Russian engineers during the plant-commissioning phase at the VVER-1000 Kozloduy Unit 6. The final volume will soon be published, completing Phase 1 of this study. (authors)
Thermal-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
International Nuclear Information System (INIS)
Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.
2013-01-01
Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code
Transient two-phase performance of LOFT reactor coolant pumps
International Nuclear Information System (INIS)
Chen, T.H.; Modro, S.M.
1983-01-01
Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed
Chaudhry, M Hanif
2014-01-01
This book covers hydraulic transients in a comprehensive and systematic manner from introduction to advanced level and presents various methods of analysis for computer solution. The field of application of the book is very broad and diverse and covers areas such as hydroelectric projects, pumped storage schemes, water-supply systems, cooling-water systems, oil pipelines and industrial piping systems. Strong emphasis is given to practical applications, including several case studies, problems of applied nature, and design criteria. This will help design engineers and introduce students to real-life projects. This book also: · Presents modern methods of analysis suitable for computer analysis, such as the method of characteristics, explicit and implicit finite-difference methods and matrix methods · Includes case studies of actual projects · Provides extensive and complete treatment of governed hydraulic turbines · Presents design charts, desi...
International Nuclear Information System (INIS)
Tincani, A.; Malavasi, A.; Ricapito, I.; Riccardi, B.; Di Maio, P.A.; Vella, G.
2007-01-01
In the frame of the activities related to ITER divertor R and D, ENEA CR Brasimone was charged by EFDA (European Fusion Design Agreement) to investigate the thermal-hydraulic behaviour of the full-scale divertor plasma facing components, i.e. Inner Vertical Target, Dome Liner and Outer Vertical Target, both in steady state and during draining and drying transient. More in detail, for each PFC, the first phase of the work is the steady state hydraulic characterization which consists of: - measurements of pressure drops at different temperatures; - determination of the velocity distribution in the internal channels; - check the possible insurgence of cavitation. The subsequent phase of the thermal-hydraulic characterization foresees a testing campaign of draining and drying procedure by means of a suitable gas flow. The objective of this experimental procedure is to eliminate in the most efficient way the residual amount of water after gravity discharge. In order to accomplish this experimental campaign a significant modification of CEF1 loop has been designed and realized. This paper presents, first of all, the experimental set-up, the agreed test matrix and the achieved results for both steady state and transient tests. Moreover, the level of the implementation of a predictive hydraulic model, based on RELAP 5 code, as well as its results are described, discussed and compared with the experimental ones. (orig.)
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.
HYTRAN: hydraulic transient code for investigating channel flow stability
International Nuclear Information System (INIS)
Kao, H.S.; Cardwell, W.R.; Morgan, C.D.
1976-01-01
HYTRAN is an analytical program used to investigate the possibility of hydraulic oscillations occurring in a reactor flow channel. The single channel studied is ordinarily the hot channel in the reactor core, which is parallel to other channels and is assumed to share a constant pressure drop with other channels. Since the channel of highest thermal state is studied, provision is made for two-phase flow that can cause a flow instability in the channel. HYTRAN uses the CHATA(1) program to establish a steady-state condition. A heat flux perturbation is then imposed on the channel, and the flow transient is calculated as a function of time
Energy Technology Data Exchange (ETDEWEB)
Ebert, D.
1997-07-01
This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.
International Nuclear Information System (INIS)
Ebert, D.
1997-07-01
This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts' meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes
International Nuclear Information System (INIS)
Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.
2004-01-01
An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect
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)
Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling
Domalapally, Phani; Di Caro, Marco
2018-05-01
Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.
Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition
International Nuclear Information System (INIS)
Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei
2015-01-01
Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor
Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition
Energy Technology Data Exchange (ETDEWEB)
Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)
2015-02-15
Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.
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)
Reitsma, Frederik
2007-01-01
efficiency: ≥ 41%; Emergency planning zone: 400 meters. The PBMR functions under a direct Brayton cycle with primary coolant helium flowing downward through the core and exiting at 900 C. The helium then enters the turbine relinquishing energy to drive the electric generator and compressors. After leaving the turbine, the helium then passes consecutively through the LP primary side of the recuperator, then the pre-cooler, the low pressure compressor, inter-cooler, high pressure compressor and then on to the HP secondary side of the recuperator before re-entering the reactor vessel at 500 C. Power is adjusted by regulating the mass flow rate of gas inside the primary circuit. This is achieved by a combination of compressor bypass and system pressure changes. Increasing the pressure results in an increase in mass flow rate, which results in an increase in the power removed from the core. Power reduction is achieved by removing gas from the circuit. A Helium Inventory Control System is used to provide an increase or decrease in system pressure. The benchmark is divided into Phases and Exercises as follows: Phase I: Steady State Benchmark Calculational Cases; Exercise 1: Neutronics Solution with Fixed Cross Sections ; Exercise 2: Thermal Hydraulic solution with given power / heat sources; Exercise 3: Combined neutronics thermal hydraulics calculation - starting condition for the transients. Phase II: Transient benchmark: Exercise 1: De-pressurised Loss of Forced Cooling (DLOFC) without SCRAM; Exercise 2 : De-pressurised Loss of Forced Cooling (DLOFC) with SCRAM; Exercise 3: Pressurised Loss of Forced Cooling (PLOFC) with SCRAM; Exercise 4 : 100-40-100 Load Follow; Exercise 5: Fast Reactivity Insertion - Control Rod Withdrawal (CRW) and Control Rod Ejection (CRE) scenarios at hot full power conditions; Exercise 6 : Cold Helium Inlet
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
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
Energy Technology Data Exchange (ETDEWEB)
Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Vallone, E., E-mail: eug.vallone@gmail.com [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy)
2015-10-15
Highlights: • ITER blanket cooling system hydraulic behaviour is studied under draining transient. • A computational approach based on the finite volume method has been followed. • Draining efficiency has been assessed in term of transient duration and residual water. • Transient duration ranges from ∼40 to 50 s, under the reference draining scenario. • Residual water is predicted to range from few tens of gram up to few kilograms. - Abstract: Within the framework of the research and development activities supported by the ITER Organization on the blanket system issues, an intense analysis campaign has been performed at the University of Palermo with the aim to investigate the thermal-hydraulic behaviour of the cooling system of a standard 20° sector of ITER blanket during the draining transient operational procedure. The analysis has been carried out following a theoretical-computational approach based on the finite volume method and adopting the RELAP5 system code. In a first phase, attention has been focused on the development and validation of the finite volume models of the cooling circuits of the most demanding modules belonging to the standard blanket sector. In later phase, attention has been put to the numerical simulation of the thermal-hydraulic transient behaviour of each cooling circuit during the draining operational procedure. The draining procedure efficiency has been assessed in terms of both transient duration and residual amount of coolant inside the circuit, observing that the former ranges typically between 40 and 120 s and the latter reaches at most ∼8 kg, in the case of the cooling circuit of twinned modules #6–7. Potential variations to operational parameters and/or to circuit lay-out have been proposed and investigated to optimize the circuit draining performances. In this paper, the set-up of the finite volume models is briefly described and the key results are summarized and critically discussed.
Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code
International Nuclear Information System (INIS)
Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang
2005-01-01
Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent
Proceedings of transient thermal-hydraulics and coupled vessel and piping system responses 1991
International Nuclear Information System (INIS)
Wang, G.Y.; Shin, Y.W.; Moody, F.J.
1991-01-01
This book reports on transient thermal-hydraulics and coupled vessel and piping system responses. Topics covered include: nuclear power plant containment designs; analysis of control rods; gate closure of hydraulic turbines; and shock wave solutions for steam water mixtures in piping systems
Trost, Nico; Jiménez, Javier; Imke, Uwe; Sanchez, Victor
2014-06-01
TWOPORFLOW is a thermo-hydraulic code based on a porous media approach to simulate single- and two-phase flow including boiling. It is under development at the Institute for Neutron Physics and Reactor Technology (INR) at KIT. The code features a 3D transient solution of the mass, momentum and energy conservation equations for two inter-penetrating fluids with a semi-implicit continuous Eulerian type solver. The application domain of TWOPORFLOW includes the flow in standard porous media and in structured porous media such as micro-channels and cores of nuclear power plants. In the latter case, the fluid domain is coupled to a fuel rod model, describing the heat flow inside the solid structure. In this work, detailed profiling tools have been utilized to determine the optimization potential of TWOPORFLOW. As a result, bottle-necks were identified and reduced in the most feasible way, leading for instance to an optimization of the water-steam property computation. Furthermore, an OpenMP implementation addressing the routines in charge of inter-phase momentum-, energy- and mass-coupling delivered good performance together with a high scalability on shared memory architectures. In contrast to that, the approach for distributed memory systems was to solve sub-problems resulting by the decomposition of the initial Cartesian geometry. Thread communication for the sub-problem boundary updates was accomplished by the Message Passing Interface (MPI) standard.
Steady and transient states of a two-phase counter current flow
International Nuclear Information System (INIS)
Siebert, S.
1984-06-01
The aim of this work is to estimate the efficiency of the counter current exchange between a heavy dispersed phase and a continous light phase in a pulse perforated plate column. From an experimental point, hydraulic measurements (retention ratio, droplet size) and residence time measurements (radioactive tracers). The model will be so applied to the calculation of retention ratios in steady conditions then of tracer concentrations in transient conditions. From a numerical point of view a fixed point type iteration then a method Runge Kutta are then adapted [fr
International Nuclear Information System (INIS)
Hsu, Y.Y.
1974-01-01
The following papers related to two-phase flow are summarized: current assumptions made in two-phase flow modeling; two-phase unsteady blowdown from pipes, flow pattern in Laval nozzle and two-phase flow dynamics; dependence of radial heat and momentum diffusion; transient behavior of the liquid film around the expanding gas slug in a vertical tube; flooding phenomena in BWR fuel bundles; and transient effects in bubble two-phase flow. (U.S.)
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)
Energy Technology Data Exchange (ETDEWEB)
Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)
1997-07-01
This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.
Transient behavior of natural circulation for boiling two-phase flow, 2
International Nuclear Information System (INIS)
Aritomi, Masanori; Chiang, Jing-Hsien; Mori, Michitugu.
1991-01-01
In this set of experiments, natural circulation in boiling two-phase flow has been investigated for power transients, simulating the start-up process in a natural circulation BWR. This was done in order to understand the underlying mechanism of thermo-hydraulic instability which may appear during a start-up. In this paper, geysering is dealt with especially and the driving mechanism is clarified by investigating the stability related to effects of inlet velocity, subcooling, temperature in an outlet plenum and non-heated length between heated section and the outlet plenum. Furthermore, by considering these results and the operational experience in the Dodewaard reactor, recommendations on how the thermo-hydraulic instabilities can be prevented from occurring are proposed concerning a reactor configuration and start-up procedure for natural circulation BWRs. (author)
Experimental studies on thermal hydraulic responses for transient operations of the SMART-P
International Nuclear Information System (INIS)
Choi, K.Y.; Park, H.S.; Cho, S.; Park, C.K.; Lee, S.J.; Song, C.H.; Chung, M.K.
2005-01-01
Full text of publication follows: Thermal hydraulic responses for transient operations of the SMART-P are experimentally investigated by using a integral effect test facility. This test facility (VISTA) has been constructed to simulate the SMART-P, which is a pilot plant of the SMART. The SMART-P is an advanced modular integral type pressurized water reactor (65 MWt) whose major RCS components, such as main coolant pumps, helical-coiled tube bundle steam generators and pressurizers, are contained in a reactor vessel. This integral design approach eliminates the large coolant loop piping, thus eliminates the occurrence of a large break LOCA. Passive Residual Heat Removal System (PRHRS) is installed to prevent overheating and over-pressurization of the primary system during accidental conditions. The PRHRS of the SMART-P removes the core decay heat by natural circulation of the two-phase fluid. The VISTA facility is a full height and 1/96 volume scaled test facility with respect to the SMART-P and will be used to understand the thermal-hydraulic responses following transients and finally to verify the system design of the SMART-P. The experimental data from the VISTA facility will be essential to system designers to resolve open issues relevant to the design of the SMART-P. The full functional control logics are implanted into the VISTA facility to cope with abnormal transients. The core of the facility can be selectively controlled by either a T-control or a T+N control method. The T-control method is a control method to adjust the core power according to the core exit coolant temperature and is designed to be used for high primary coolant flow conditions. On the other hand, the T+N control method is for low primary coolant flow conditions and it uses core exit temperature as well as core power itself as control inputs. The thermal hydraulic responses are carefully investigated according to different core control methods. Several experiments have been performed to
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)
Hirano, Masashi; Hada, Kazuhiko
1990-04-01
The THYDE-HTGR code has been developed for transient thermal-hydraulic analyses of high-temperature gas-cooled reactors, based on the THYDE-W code. THYDE-W is a code developed at JAERI for the simulation of Light Water Reactor plant dynamics during various types of transients including loss-of-coolant accidents. THYDE-HTGR solves the conservation equations of mass, momentum and energy for compressible gas, or single-phase or two-phase flow. The major code modification from THYDE-W is to treat helium loops as well as water loops. In parallel to this, modification has been made for the neutron kinetics to be applicable to helium-cooled graphite-moderated reactors, for the heat transfer models to be applicable to various types of heat exchangers, and so forth. In order to assess the validity of the modifications, analyses of some of the experiments conducted at the High Temperature Test Loop of ERANS have been performed. In this report, the models applied in THYDE-HTGR are described focusing on the present modifications and the results from the assessment calculations are presented. (author)
Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research
International Nuclear Information System (INIS)
Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari
2001-03-01
JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)
3D numerical simulation of transient processes in hydraulic turbines
International Nuclear Information System (INIS)
Cherny, S; Chirkov, D; Lapin, V; Eshkunova, I; Bannikov, D; Avdushenko, A; Skorospelov, V
2010-01-01
An approach for numerical simulation of 3D hydraulic turbine flows in transient operating regimes is presented. The method is based on a coupled solution of incompressible RANS equations, runner rotation equation, and water hammer equations. The issue of setting appropriate boundary conditions is considered in detail. As an illustration, the simulation results for runaway process are presented. The evolution of vortex structure and its effect on computed runaway traces are analyzed.
3D numerical simulation of transient processes in hydraulic turbines
Cherny, S.; Chirkov, D.; Bannikov, D.; Lapin, V.; Skorospelov, V.; Eshkunova, I.; Avdushenko, A.
2010-08-01
An approach for numerical simulation of 3D hydraulic turbine flows in transient operating regimes is presented. The method is based on a coupled solution of incompressible RANS equations, runner rotation equation, and water hammer equations. The issue of setting appropriate boundary conditions is considered in detail. As an illustration, the simulation results for runaway process are presented. The evolution of vortex structure and its effect on computed runaway traces are analyzed.
International Nuclear Information System (INIS)
Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.
2012-01-01
The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)
Dynamic Response in Transient Stress-Field Behavior Induced by Hydraulic Fracturing
Jenkins, Andrew
magnitude. These types of shifts are of great concern because they can impact subsequent fracture development causing non-uniform fracture propagation and the potential overlapping of fracture paths as they extend from the wellbore at the point of injection. The dynamics of stress variation that occur with respect to hydraulic fracturing is a somewhat new area of study. In order to accomplish the goals of this thesis and continue future research in this area a new transient model has been developed in order to asses these dynamic systems and determine their influence on fracture behavior. This applies the use of a fully coupled finite element method in 2-D using linear elastic fracture mechanics which is then expanded using displacement discontinuity to a cohesive zone model in 3-D. A static boundary element model was also used to determine stress fields surrounding static, predetermined fracture geometries. These models have been verified against analytical solutions for simple cases and are now being applied to more detailed case studies and analysis. These models have been briefly discussed throughout this thesis in order to give insight on their current capabilities and application as well as their future potential within this area of research. The majority of this work introduces transient stress field prediction to cases of single and multiple hydraulic fractures. The static assessment of these stresses is determined for verification of results to those found in publication which leads into these transient stress field variations. A new method has been developed and applied to the stress state prediction for the first time in a transient fracture model which is partly based upon a critical distance theory. These dynamic interactions can provide useful insight to pertinent issues within the petroleum and natural gas industry such as those to hydraulic fracturing fluid loss and induced seismic events, as well as to applications of efficiency and optimization of the
International Nuclear Information System (INIS)
Ivanov, B.; Ivanov, K.; Aniel, S.; Royer, E.; Kolev, N.; Groudev, P.
2004-01-01
The present paper describes the two phases of the OECD/DOE/CEA VVER-1000 coolant transient benchmark labeled as V1000CT. This benchmark is based on a data from the Bulgarian Kozloduy NPP Unit 6. The first phase of the benchmark was designed for the purpose of assessing neutron kinetics and thermal-hydraulic modeling for a VVER-1000 reactor, and specifically for their use in analyzing reactivity transients in a VVER-1000 reactor. Most of the results of Phase 1 will be compared against experimental data and the rest of the results will be used for code-to-code comparison. The second phase of the benchmark is planned for evaluation and improvement of the mixing computational models. Code-to-code and code-to-data comparisons will be done based on data of a mixing experiment conducted at Kozloduy-6. Main steam line break will be also analyzed in the second phase of the V1000CT benchmark. The results from it will be used for code-to-code comparison. The benchmark team has been involved in analyzing different aspects and performing sensitivity studies of the different benchmark exercises. The paper presents a comparison of selected results, obtained with two different system thermal-hydraulics codes, with the plant data for the Exercise 1 of Phase 1 of the benchmark as well as some results for Exercises 2 and 3. Overall, this benchmark has been well accepted internationally, with many organizations representing 11 countries participating in the first phase of the benchmark. (authors)
International Nuclear Information System (INIS)
Reitsma, F.; Han, J.; Ivanov, K.; Sartori, E.
2008-01-01
The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated. Since only a few pebble-bed HTR experimental facilities or plant data are available the use of code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MW design and a representative set of transient cases is defined as an OECD benchmark. The scope of the benchmark is to establish a series of well-defined multi-dimensional computational benchmark problems with a common given set of cross-sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. The OECD benchmark includes steady-state and transients cases. Although the focus of the benchmark is on the modelling of the transient behaviour of the PBMR core, it was also necessary to define some steady-state cases to ensure consistency between the different approaches before results of transient cases could be compared. This paper describes the status of the benchmark project and shows the results for the three steady state exercises defined as a standalone neutronics calculation, a standalone thermal-hydraulic core calculation, and a coupled neutronics/thermal-hydraulic simulation. (authors)
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.
International Nuclear Information System (INIS)
Toti, A.; Vierendeels, J.; Belloni, F.
2017-01-01
Highlights: • A system thermal-hydraulic/CFD coupling methodology is proposed for high-fidelity transient flow analyses. • The method is based on domain decomposition and implicit numerical scheme. • A novel interface Quasi-Newton algorithm is implemented to improve stability and convergence rate. • Preliminary validation analyses on the TALL-3D experiment. - Abstract: The paper describes the development and validation of a coupling methodology between the best-estimate system thermal-hydraulic code RELAP5-3D and the CFD code FLUENT, conceived for high fidelity plant-scale safety analyses of pool-type reactors. The computational tool is developed to assess the impact of three-dimensional phenomena occurring in accidental transients such as loss of flow (LOF) in the research reactor MYRRHA, currently in the design phase at the Belgian Nuclear Research Centre, SCK• CEN. A partitioned, implicit domain decomposition coupling algorithm is implemented, in which the coupled domains exchange thermal-hydraulics variables at coupling boundary interfaces. Numerical stability and interface convergence rates are improved by a novel interface Quasi-Newton algorithm, which is compared in this paper with previously tested numerical schemes. The developed computational method has been assessed for validation purposes against the experiment performed at the test facility TALL-3D, operated by the Royal Institute of Technology (KTH) in Sweden. This paper details the results of the simulation of a loss of forced convection test, showing the capability of the developed methodology to predict transients influenced by local three-dimensional phenomena.
Analysis of the VVER-1000 coolant transient benchmark phase 1 with the code system RELAP5/PARCS
International Nuclear Information System (INIS)
Victor Hugo Sanchez Espinoza
2005-01-01
Full text of publication follows: As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during
International Nuclear Information System (INIS)
Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman
1988-01-01
TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory
Energy Technology Data Exchange (ETDEWEB)
Muzhang, Peng; Quan, Zhang; Guoli, Wang; Yuman, Zhang
1988-03-01
TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory.
International Nuclear Information System (INIS)
Sanchez-Espinoza, Victor Hugo
2008-07-01
As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the
Energy Technology Data Exchange (ETDEWEB)
Sanchez-Espinoza, Victor Hugo
2008-07-15
As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the
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)
Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA
International Nuclear Information System (INIS)
Sugimoto, Jun
1989-01-01
This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)
Virtual Design of a Controller for a Hydraulic Cam Phasing System
Schneider, Markus; Ulbrich, Heinz
2010-09-01
Hydraulic vane cam phasing systems are nowadays widely used for improving the performance of combustion engines. At stationary operation, these systems should achieve a constant phasing angle, which however is badly disturbed by the alternating torque generated by the valve actuation. As the hydraulic system shows a non-linear characteristic over the full operation range and the inductivity of the hydraulic pipes generates a significant time delay, a full model based control emerges very complex. Therefore a simple feed-forward controller is designed, bridging the time delay of the hydraulic system and improving the system behaviour significantly.
International Nuclear Information System (INIS)
Raymond, P.; Caruge, D.; Paik, H.J.
1994-01-01
The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs
Power transients of Ghana research reactor-1 using PARET/ANL thermal hydraulic code
International Nuclear Information System (INIS)
Ampomah-Amoaka, E.; Akaho, E.H.K.; Anim-Sampong, S.; Nyarko, B.J.B.
2010-01-01
PARET/ANL(Version 7.3 of 2007) thermal-hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1.The reactivities inserted were 2.1mk and 4mk.The peak power of 5.81kW was obtained for 2.1 mk insertion whereas the peak power for 4mk insertion of reactivity was 92.32kW.These results compare closely with experiments and theoretical studies conducted previously.
International Nuclear Information System (INIS)
Chvetsov, I.; Volkov, A.
2000-01-01
For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)
Transient thermal hydraulic modeling and analysis of ITER divertor plate system
International Nuclear Information System (INIS)
El-Morshedy, Salah El-Din; Hassanein, Ahmed
2009-01-01
A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m 2 plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.
Transient thermal hydraulic modeling and analysis of ITER divertor plate system
Energy Technology Data Exchange (ETDEWEB)
El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu
2009-12-15
A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.
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)
Domanus, H.M.; Schmitt, R.C.; Sha, W.T.; Shah, V.L.
1983-12-01
The COMMIX-1A computer program is an updated and improved version of COMMIX-1 designed to analyze steady-state/transient, single-phase, three-dimensional fluid flow with heat transfer in reactor components and multicomponent systems. A new porous-media formulation via local volume averaging has been derived and employed in the COMMIX code. The concepts of volume porosity, directional surface permeability, distributed resistance, and distributed heat source or sink is used in the new porous-media formulation to model a flow domain with stationary structures. The concept of directional surface permeability is new and greatly facilitates modeling of velocity and temperature fields in anisotropic media. The new porous-media formulation represents the first unified approach to thermal-hydraulic analysis. It is now possible to perform a multidimensional thermal-hydraulic simulation of either a single component, such as a rod bundle, reactor plenum, piping system, heat exchanger, etc., or a multicomponent system that is a combination of these components. The conservation equations of mass, momentum, and energy based on the new porous-media formulation are solved as a boundary-value problem in space and an initial-value problem in time. Two other unique features provided in the COMMIX-1A code are (1) two solution procedures - a semi-implicit procedure modified from ICE and a fully-implicit procedure, named SIMPLEST-ANL, similar to the SIMPLE/SIMPLER algorithms - available a user's option and (2) a geometrical package capable of approximating many geometries. This report (Volume I) describes in detail the basic equations, formulations, solution procedures, flow charts, rebalancing scheme for faster convergence, options available to users, models to describe the auxiliary phenomena, input instructions, and two sample problems. The Volume II assembles and summarizes the results of many simulations that have been performed with COMMIX-1A computer program
International Nuclear Information System (INIS)
Hoeld, A.
2004-01-01
The thermal-hydraulic theory of single- and especially two-phase flow systems used for plant transient analysis is dominated by separate-phase models. The corresponding mostly very comprehensive codes (TRAC, RELAP, CATHARE, ATHLET etc.) are looked as to be by far more efficient than a 3 eq. mixture-fluid approach and code also if they show deficiencies in describing flow situations within inner loops as for example the distribution into parallel channels (and thus the simulation of 3D thermal-hydraulic phenomena). This may be justified if comparing them to the very simple 'homogeneous equilibrium models (HEM)', but not if looking to the more refined non-homogeneous 'separate-region' mixture-fluid approaches based on appropriate drift-flux correlation packages which can have, on the contrary, enormous advantages with respect to such separate-phase models. Especially if comparing the basic (and starting) eqs. of such theoretical models of both types the differences are remarkable. Single-phase and mixture-fluid models start from genuine conservation eqs. for mass, energy and momentum, demanding (in case of two-phase flow) additionally an adequate drift flux package (in order to get a relation for a fourth independent variable), a heat transfer coefficients package (over the whole range of the possible fields of application) and correlations for single- and two-phase friction. The other types of models are looking at each phase separately with corresponding 'field' eqs. for each phase, connected by exchange (=closure) terms which substitute the classical constitutive packages for drift, heat transfer and friction. That the drift-flux, heat transfer into a coolant channel and friction along a wall and between the phases is described better by a separate-phase approach is at least doubtful. The corresponding mixture-fluid correlations are based over a wide range on a treasure of experience and measurements, their pseudo-stationary treatment can (due to their small time
International Nuclear Information System (INIS)
Tyobeka, Bismark; Pautz, Andreas; Ivanov, Kostadin
2008-01-01
In new reactor designs that are still under review such as the PBMR, not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400 MW OECD/NEA/NSC coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate the transient scenarios in the above-mentioned benchmark problem. Steady-state calculations results are compared with selected participants' results as well as transient models in which the diffusion and transport theory solutions of the same code system are directly compared. Several sensitivity studies are also shown in order to determine how much the change in certain parameters influences the overall behaviour of a given transient. It is shown in this paper that DORT-TD/THERMIX is a versatile tool which can be deployed for design and safety analyses of high temperature reactors of pebble-bed type. (authors)
Energy Technology Data Exchange (ETDEWEB)
Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear
2017-11-01
A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)
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.
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
Numerical simulation of the two-phase flows in a hydraulic coupling by solving VOF model
International Nuclear Information System (INIS)
Luo, Y; Zuo, Z G; Liu, S H; Fan, H G; Zhuge, W L
2013-01-01
The flow in a partially filled hydraulic coupling is essentially a gas-liquid two-phase flow, in which the distribution of two phases has significant influence on its characteristics. The interfaces between the air and the liquid, and the circulating flows inside the hydraulic coupling can be simulated by solving the VOF two-phase model. In this paper, PISO algorithm and RNG k–ε turbulence model were employed to simulate the phase distribution and the flow field in a hydraulic coupling with 80% liquid fill. The results indicate that the flow forms a circulating movement on the torus section with decreasing speed ratio. In the pump impeller, the air phase mostly accumulates on the suction side of the blades, while liquid on the pressure side; in turbine runner, air locates in the middle of the flow passage. Flow separations appear near the blades and the enclosing boundaries of the hydraulic coupling
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.
Thermal hydraulic behavior of SCWR sliding pressure startup
International Nuclear Information System (INIS)
Fu Shengwei; Zhou Chong; Xu Zhihong; Yang Yanhua
2011-01-01
The modification to ATHLET-SC code is introduced in this paper, which realizes the simulation of trans-critical transients using two-phase model. With the modified code, the thermal-hydraulic dynamic behavior of the mixed SCWR core during the startup process is simulated. The startup process is similar to the design of SCLWR-H sliding pressure startup. The results show that maximum temperature of cladding-surface does not exceed 650℃ in the whole startup process, and the sudden change of water properties in the trans-critical transients will not cause harmful influence to the heat transfer of the fuel cladding. (authors)
International Nuclear Information System (INIS)
Hirano, Masashi; Sudo, Yukio
1986-01-01
Transient thermal-hydraulic behaviors of the JRR-3 which is an open-pool type research reactor has been analyzed with the THYDE-P1 code. The focal point is the thermal-hydraulic behaviors related to the core flow reversal during the transition from forced circulation downflow to natural circulation upflow. In the case of a loss-of-coolant accident (LOCA), for example, the core flow reversal is expected to occur just after the water pool isolation from the primary cooling loop with a leak. The core flow reversal should cause a sudden increase in fuel temperature and a steep decrease in the departure-from-nucleate-boiling ratio (DNBR) and the phenomenon is, therefore, very important especially for safety design and evaluation of research reactors. Major purposes of the present work are to clarify physical phenomena during the transient and to identify important parameters affecting the peak fuel temperature and the minimum DNBR. The results calculated with THYDE-P1 assuming the sequences of events of the loss-of-offsite power and LOCA help us to understand the phenomena both qualitatively and quantitatively, with respect to the safety design and evaluation. (author)
International Nuclear Information System (INIS)
Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.
2006-01-01
The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)
Energy Technology Data Exchange (ETDEWEB)
Canetta, D.; Capozza, A.; Iovino, G.
1985-01-01
The transient response following pump trip-offs and start-ups was investigated in the sea water system of a nuclear power plant. Specific care was devoted to water column separation and cavity collapse phenomena. A computer program designed for analysis of complex hydraulic networks was used. It is found that dangerous overpressures can be avoided by the use of loop seals. The design of the vacuum breaker valves of the loop seals and the optimization of overall transient behavior is discussed. 1 reference.
Ambiguous hydraulic heads and 14C activities in transient regional flow.
Schwartz, Franklin W; Sudicky, Edward A; McLaren, Robert G; Park, Young-Jin; Huber, Matthew; Apted, Mick
2010-01-01
A regional flow and transport model is used to explore the implications of significant variability in Pleistocene and Holocene climates on hydraulic heads and (14)C activity. Simulations involve a 39 km slice of the Death Valley Flow System through Yucca Mountain toward the Amargosa Desert. The long-time scale over which infiltration has changed (tens-of-thousands of years) is matched by the large physical extent of the flow system (many tens-of-kilometers). Estimated paleo-infiltration rates were estimated using a juniper pollen percentage that extends from the last interglacial (LIG) period (approximately 120 kyrbp) to present. Flow and (14)C transport simulations show that groundwater flow changes markedly as a function of paleoclimate. At the last glacial maximum (LGM, 21 kyrbp), the recharge to the flow system was about an order-of-magnitude higher than present, and water table was more than 100 m higher. With large basin time constants, flow is complicated because hydraulic heads at a given location reflect conditions of the past, but at another location the flow may reflect present conditions. This complexity is also manifested by processes that depend on flow, for example (14)C transport. Without a model that accounts for the historical transients in recharge for at least the last 20,000 years, there is no simple way to deconvolve the (14)C dates to explain patterns of flow.
International Nuclear Information System (INIS)
Di Maio, P.A.; Paradiso, D.; Dell’Orco, G.; Pitcher, C.S.; Kalish, M.
2011-01-01
Highlights: ► UPP TS hydraulic behaviour has been investigated under steady state and D and D transient conditions. ► A thermal–hydraulic system code has been adopted and a UPP TS model has been set-up and validated against results of steady state CFD analyses. ► The TS steady state hydraulic characteristic functions have been derived for two coolant flow paths showing that right plate inlet one is the most promising. ► Draining simulations indicate that the 4 MPa injection pressure is high enough to drain almost completely the circuit in a reasonable time (∼6 s). ► Results indicate that right plate inlet flow path allows the TS complete draining, eliminating the need for the drying procedure. - Abstract: The ITER diagnostic Upper Port Plug (UPP) is a water-cooled stainless steel structure aimed to integrate within vacuum vessel the plasma diagnostic systems, shielding them from neutron and photon irradiation. Due to the very intense heat loads expected, a proper cooling circuit has been designed to ensure an adequate UPP cooling with an acceptable thermal rise and an unduly high pumping power and to perform its draining and drying procedure by injection of pressurized nitrogen. A theoretical research activity has been launched at the Department of Nuclear Engineering of the University of Palermo aiming to investigate the hydraulic behaviour of the UPP Trapezoid Section cooling circuit under steady state conditions and during its draining and drying transient procedure. The research activity has been performed following a theoretical–computational approach and adopting the RELAP5 thermal–hydraulic system code. The Trapezoid Section cooling circuit characteristic functions have been derived under steady state conditions at various coolant temperatures for both the coolant flow paths at the present under consideration for this circuit. The distributions of coolant mass flow rates along the channels of the cooling circuit have been calculated too
Ahmed, A. Soueid; Jardani, A.; Revil, A.; Dupont, J. P.
2016-03-01
Transient hydraulic tomography is used to image the heterogeneous hydraulic conductivity and specific storage fields of shallow aquifers using time series of hydraulic head data. Such ill-posed and non-unique inverse problem can be regularized using some spatial geostatistical characteristic of the two fields. In addition to hydraulic heads changes, the flow of water, during pumping tests, generates an electrical field of electrokinetic nature. These electrical field fluctuations can be passively recorded at the ground surface using a network of non-polarizing electrodes connected to a high impedance (> 10 MOhm) and sensitive (0.1 mV) voltmeter, a method known in geophysics as the self-potential method. We perform a joint inversion of the self-potential and hydraulic head data to image the hydraulic conductivity and specific storage fields. We work on a 3D synthetic confined aquifer and we use the adjoint state method to compute the sensitivities of the hydraulic parameters to the hydraulic head and self-potential data in both steady-state and transient conditions. The inverse problem is solved using the geostatistical quasi-linear algorithm framework of Kitanidis. When the number of piezometers is small, the record of the transient self-potential signals provides useful information to characterize the hydraulic conductivity and specific storage fields. These results show that the self-potential method reveals the heterogeneities of some areas of the aquifer, which could not been captured by the tomography based on the hydraulic heads alone. In our analysis, the improvement on the hydraulic conductivity and specific storage estimations were based on perfect knowledge of electrical resistivity field. This implies that electrical resistivity will need to be jointly inverted with the hydraulic parameters in future studies and the impact of its uncertainty assessed with respect to the final tomograms of the hydraulic parameters.
Basic researches on thermo-hydraulic non-equilibrium phenomena related to nuclear reactor safety
International Nuclear Information System (INIS)
Sakurai, Akira; Kataoka, Isao; Aritomi, Masanori.
1989-01-01
A review was made of recent developments of fundamental researches on thermo-hydraulic non-equilibrium phenomena related to light water reactor safety, in relation to problems to be solved for the improvement of safety analysis codes. As for the problems related to flow con ditions, fundamental researches on basic conservation equations and constitutive equations for transient two-phase flow were reviewed. Regarding to the problems related to thermal non-equilibrium phenomena, fundamental researches on film boiling in pool and forced convection, transient boiling heat transfer and flow behavior caused by pressure transients were reviewed. (author)
Energy Technology Data Exchange (ETDEWEB)
Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Scari, Maria E., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear
2015-07-01
Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)
Dutta, G.; Jiang, J.; Maitri, R.; Zhang, C.
2016-01-01
The present work demonstrates the extension of a thermal-hydraulic model, THRUST, with an objective to simulate the fast transient flow dynamics in a supercritical water channel of circular cross section. THRUST is a 1-D model which solves the nonlinearly coupled mass, axial momentum and energy
Analysis of forced convective transient boiling by homogeneous model of two-phase flow
International Nuclear Information System (INIS)
Kataoka, Isao
1985-01-01
Transient forced convective boiling is of practical importance in relation to the accident analysis of nuclear reactor etc. For large length-to-diameter ratio, the transient boiling characteristics are predicted by transient two-phase flow calculations. Based on homogeneous model of two-phase flow, the transient forced convective boiling for power and flow transients are analysed. Analytical expressions of various parameters of transient two-phase flow have been obtained for several simple cases of power and flow transients. Based on these results, heat flux, velocity and time at transient CHF condition are predicted analytically for step and exponential power increases, and step, exponential and linear velocity decreases. The effects of various parameters on heat flux, velocity and time at transient CHF condition have been clarified. Numerical approach combined with analytical method is proposed for more complicated cases. Solution method for pressure transient are also described. (author)
Transient 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.
PSH Transient Simulation Modeling
Energy Technology Data Exchange (ETDEWEB)
Muljadi, Eduard [National Renewable Energy Laboratory (NREL), Golden, CO (United States)
2017-12-21
PSH Transient Simulation Modeling presentation from the WPTO FY14 - FY16 Peer Review. Transient effects are an important consideration when designing a PSH system, yet numerical techniques for hydraulic transient analysis still need improvements for adjustable-speed (AS) reversible pump-turbine applications.
Directory of Open Access Journals (Sweden)
Youwei He
2018-02-01
Full Text Available Although technical advances in hydraulically fracturing and drilling enable commercial production from tight reservoirs, oil/gas recovery remains at a low level. Due to the technical and economic limitations of well-testing operations in tight reservoirs, rate-transient analysis (RTA has become a more attractive option. However, current RTA models hardly consider the effect of the non-uniform production on rate decline behaviors. In fact, PLT results demonstrate that production profile is non-uniform. To fill this gap, this paper presents an improved RTA model of multi-fractured horizontal wells (MFHWs to investigate the effects of non-uniform properties of hydraulic fractures (production of fractures, fracture half-length, number of fractures, fracture conductivity, and vertical permeability on rate transient behaviors through the diagnostic type curves. Results indicate obvious differences on the rate decline curves among the type curves of uniform properties of fractures (UPF and non-uniform properties of fractures (NPF. The use of dimensionless production integral derivative curve magnifies the differences so that we can diagnose the phenomenon of non-uniform production. Therefore, it’s significant to incorporate the effects of NPF into the RDA models of MFHWs, and the model proposed in this paper enables us to better evaluate well performance based on long-term production data.
International Nuclear Information System (INIS)
Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.; Hughes, E.D.; Solbrig, C.W.
1975-11-01
A mathematical model and a numerical solution scheme for thermal-hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media
Overcooling transient selection and thermal hydraulic analyses of the Loviisa PTS assessments
Energy Technology Data Exchange (ETDEWEB)
Tuomisto, H [IVO Power Engineering Ltd, Vantaa (Finland)
1997-09-01
This paper describes transients selection and thermal hydraulic analyses of various PTS assessment studies performed for the pressure vessels of the Loviisa WWER-reactors. Deterministic analyses have been performed in various stages of the PTS studies and they have always made the formal basis for design and licensing of the reactor pressure vessel. The integrated, probabilistic PTS study was carried out to give an overview of the severity of all different PTS sequences, and give a quantitative estimate of the importance of the PTS issues in relation to the overall safety of the plant. Later, the sequences including external flooding of the pressure vessels were added to the PTS assessment. Thermal recovery annealing of the Loviisa 1 reactor pressure vessel took place during refuelling outage in 1996. (author). 10 refs, 4 figs, 3 tabs.
Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures
Directory of Open Access Journals (Sweden)
Alessandro Petruzzi
2008-01-01
Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.
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.)
Application of Phase-Field Techniques to Hydraulically- and Deformation-Induced Fracture.
Energy Technology Data Exchange (ETDEWEB)
Culp, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Miller, Nathan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Schweizer, Laura [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2017-08-01
Phase-field techniques provide an alternative approach to fracture problems which mitigate some of the computational expense associated with tracking the crack interface and the coalescence of individual fractures. The technique is extended to apply to hydraulically driven fracture such as would occur during fracking or CO_{2} sequestration. Additionally, the technique is applied to a stainless steel specimen used in the Sandia Fracture Challenge. It was found that the phase-field model performs very well, at least qualitatively, in both deformation-induced fracture and hydraulically-induced fracture, though spurious hourglassing modes were observed during coupled hydralically-induced fracture. Future work would include performing additional quantitative benchmark tests and updating the model as needed.
International Nuclear Information System (INIS)
Daimaru, Shuji; Takeuchi, Ryuji; Onoe, Hironori; Saegusa, Hiromitsu
2012-09-01
This report summarize the results of the single borehole hydraulic tests of 79 sections conducted as part of the Construction phase (Phase 2) in the Mizunami Underground Research Laboratory (MIU) Project. The details of each test (test interval depth, geology, etc.) as well as the interpreted hydraulic parameters and analytical method used are presented in this report. (author)
TRANTHAC-1: transient thermal-hydraulic analysis code for HTGR core of multi-channel model
International Nuclear Information System (INIS)
Sato, Sadao; Miyamoto, Yoshiaki
1980-08-01
The computer program TRANTHAC-1 is for predicting thermal-hydraulic transient behavior in HTGR's core of pin-in-block type fuel elements, taking into consideration of the core flow distribution. The program treats a multi-channel model, each single channel representing the respective column composed of fuel elements. The fuel columns are grouped in flow control regions; each region is provided with an orifice assembly. In the region, all channels are of the same shape except one channel. Core heat is removed by downward flow of the control through the channel. In any transients, for given time-dependent power, total core flow, inlet coolant temperature and coolant pressure, the thermal response of the core can be determined. In the respective channels, the heat conduction in radial and axial direction are represented. And the temperature distribution in each channel with the components is calculated. The model and usage of the program are described. The program is written in FORTRAN-IV for computer FACOM 230-75 and it is composed of about 4,000 cards. The required core memory is about 75 kilowords. (author)
International Nuclear Information System (INIS)
Nalezny, C.L.; Chapman, R.L.; Martinell, J.S.; Riordon, R.P.; Solbrig, C.W.
1979-01-01
Mass flow is an important measured variable in the Loss-of-Fluid Test (LOFT) Program. Large uncertainties in mass flow measurements in the LOFT piping during LOFT coolant experiments requires instrument testing in a transient two-phase flow loop that simulates the geometry of the LOFT piping. To satisfy this need, a transient two-phase flow loop has been designed and built. The load cell weighing system, which provides reference mass flow measurements, has been analyzed to assess its capability to provide the measurements. The analysis consisted of first performing a thermal-hydraulic analysis using RELAP4 to compute mass inventory and pressure fluctuations in the system and mass flow rate at the instrument location. RELAP4 output was used as input to a structural analysis code SAPIV which is used to determine load cell response. The computed load cell response was then smoothed and differentiated to compute mass flow rate from the system. Comparison between computed mass flow rate at the instrument location and mass flow rate from the system computed from the load cell output was used to evaluate mass flow measurement capability of the load cell weighing system. Results of the analysis indicate that the load cell weighing system will provide reference mass flows more accurately than the instruments now in LOFT
International Nuclear Information System (INIS)
Siebe, D.A.; Boyack, B.E.; Giguere, P.T.
1994-11-01
This report presents the results of the review and planning done for the United States Nuclear Regulatory Commission to identify the thermal-hydraulic phenomena that could occur in the CANDU 3 reactor design during transient conditions, plan modifications to the TRAC-PF1/MOD2 (TRAC) computer code needed to adequately predict CANDU 3 transient thermal-hydraulic phenomena, and identify an assessment program to verify the ability of TRAC, when modified, to predict these phenomena. This work builds on analyses and recommendations produced by the Idaho National Engineering Laboratory (INEL). To identify the thermal-hydraulic phenomena, a large-break loss-of-coolant accident simulation, performed as part of earlier work by INEL with an Atomic Energy of Canada, Limited (AECL) thermal-hydraulic computer code (CATHENA), was analyzed in detail. Other accident scenarios were examined for additional phenomena. A group of Los Alamos National Laboratory reactor thermal-hydraulics experts ranked the phenomena to produce a preliminary phenomena identification and ranking table (PIRT). The preliminary nature of the PIRT was a result of a lack of direct expertise with the unique processes and phenomena of the CANDU 3. Nonetheless, this PIRT provided an adequate foundation for planning a program of code modifications. We believe that this PIRT captured the most important phenomena and that refinements to the PIRT will mainly produce clarification of the relative importance (ranking) of phenomena. A plan for code modifications was developed based on this PIRT and on information about the modeling methodologies for CANDU-specific phenomena used in AECL codes. AECL thermal-hydraulic test facilities and programs were reviewed and the information used in developing an assessment plan to ensure that TRAC-PF1/MOD2, when modified, will adequately predict CANDU 3 phenomena
International Nuclear Information System (INIS)
Green, C.
1979-12-01
A set of FORTRAN subroutines is described for calculating water thermodynamic properties. These were written for use in a transient thermal-hydraulics program, where speed of execution is paramount. The choice of which subroutines to optimise depends on the primary variables in the thermal-hydraulics code. In this particular case the subroutine which has been optimised is the one which calculates pressure and specific enthalpy given the specific volume and the specific internal energy. Another two subroutines are described which complete a self-consistent set. These calculate the specific volume and the temperature given the pressure and the specific enthalpy, and the specific enthalpy and the specific volume given the pressure and the temperature (or the quality). The accuracy is high near the saturation lines, typically less than 1% relative error, and decreases as the fluid becomes more subcooled in the liquid region or more superheated in the steam region. This behaviour is inherent in the method which uses quantities defined on the saturation lines and assumes that certain derivatives are constant for excursions away from these saturation lines. The accuracy and speed of the subroutines are discussed in detail in this report. (author)
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
Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors
Energy Technology Data Exchange (ETDEWEB)
Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)
2015-05-15
Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.
How good are thermal-hydraulics codes for analyses of plant transients
International Nuclear Information System (INIS)
Fabic, S.
1996-01-01
In the early seventies, all thermal-hydraulics codes were based on the Homogeneous Equilibrium Model (HEM), represented by three conservation equations: mixture mass, momentum and energy. Various means were utilized to solve the resulting system of equations: finite differences in FLASH, SATAN, RELAP3 and RELAP4, method of characteristics in BLOWDWN2, loop momentum method in RAMONA and NORCOOL, and others. As the result the world came to regard HEM as too restrictive and the Two-Fluid model came into fashion, first featuring a six and later, a seven-equation model. New codes like KACHINA, TRAC and RELAP5 were developed also. Experience and comparisons with test data have recently forced us to wonder whether the ability to 'compute' while considering great many complexities, ran ahead of the ability to competently define various interactions between fluid phases and components that such complex codes require. The long running times are also a problem that needs to be resolved. More recent trends in the treatment of thermal-hydraulics in Power Plant Simulators and in Plant Analyzers will also be discussed
Energy Technology Data Exchange (ETDEWEB)
Reocreux, M; Rubinstein, M C [CEA-IPSN/DRS/SEMAR, Centre d' Etudes Nucleaires de Cadarache, 13108 Saint Paul-lez-Durance (France)
1992-07-01
The 4. Specialists' Meeting on Transient Two Phase Flow was organized by the Safety Research Department of the French Nuclear Safety and Protection Institute at the request of the OECD Committee for the Safety of Nuclear Installations. After Toronto in 1976, Paris in 1978 and Pasadena in 1981, the Aix-en-Provence meeting was in keeping with the course of studies initiated by the Thermalhydraulic Systems Behavior Task Group of the Principal Working Group No.2 for discussing the achievements and defining the needs of safety research in accident thermal-hydraulics. 60 Specialists from 14 Countries (Belgium, Canada, Finland, France, Germany, Italy, Japan, the Netherlands, the United Kingdom, the USA, Spain, Sweden, Switzerland, Taiwan) attended the meeting, representing a large spectrum of experts from National Safety Authorities, Research Laboratories, Universities, Vendors and Utilities. These specialists had to review the 15-year research period which had elapsed since the last meetings. This period had been characterized by the issuance of the large thermalhydraulic computer codes for LWR accidents, the performance of several hundreds of separate effect tests for the development and the qualification of the physical models, the carrying-out of the large experimental programmes on system loops (up to scale 1) for verifying the computer codes. Although this research was mainly characterized by remarkable success, limitations still exist. In a safety approach, there need to be well identified and handled, and the specialists were asked to exchange their views in order to determine which solutions they expected to be affordable in the future. Safety applications have already started which use these latest research achievements. They raise specific problems such as the use of validation matrices, the evaluation of uncertainties, the identification and the control of unavoidable users' effects. The specialists were required to exchange their experience of applications
International Nuclear Information System (INIS)
Reocreux, M.; Rubinstein, M.C.
1992-01-01
The 4. Specialists' Meeting on Transient Two Phase Flow was organized by the Safety Research Department of the French Nuclear Safety and Protection Institute at the request of the OECD Committee for the Safety of Nuclear Installations. After Toronto in 1976, Paris in 1978 and Pasadena in 1981, the Aix-en-Provence meeting was in keeping with the course of studies initiated by the Thermalhydraulic Systems Behavior Task Group of the Principal Working Group No.2 for discussing the achievements and defining the needs of safety research in accident thermal-hydraulics. 60 Specialists from 14 Countries (Belgium, Canada, Finland, France, Germany, Italy, Japan, the Netherlands, the United Kingdom, the USA, Spain, Sweden, Switzerland, Taiwan) attended the meeting, representing a large spectrum of experts from National Safety Authorities, Research Laboratories, Universities, Vendors and Utilities. These specialists had to review the 15-year research period which had elapsed since the last meetings. This period had been characterized by the issuance of the large thermalhydraulic computer codes for LWR accidents, the performance of several hundreds of separate effect tests for the development and the qualification of the physical models, the carrying-out of the large experimental programmes on system loops (up to scale 1) for verifying the computer codes. Although this research was mainly characterized by remarkable success, limitations still exist. In a safety approach, there need to be well identified and handled, and the specialists were asked to exchange their views in order to determine which solutions they expected to be affordable in the future. Safety applications have already started which use these latest research achievements. They raise specific problems such as the use of validation matrices, the evaluation of uncertainties, the identification and the control of unavoidable users' effects. The specialists were required to exchange their experience of applications
International Nuclear Information System (INIS)
AL-Yahia, Omar S.; Albati, Mohammad A.; Park, Jonghark; Chae, Heetaek; Jo, Daeseong
2013-01-01
Highlights: • Transient analyses of a slow and fast LOFA were investigated. • A reactor kinetic and thermal hydraulic coupled model was developed. • Based on force balance, the flow rate during flow inversion was determined. • Flow inversion in a hot channel occurred earlier than in an average channel. • Two temperature peaks were observed during both slow and fast LOFA. - Abstract: Transient analyses of the IAEA 10 MW MTR reactor are investigated during a fast and slow Loss of Flow Accident (LOFA) with a neutron kinetic and thermal hydraulic coupling model. A spatial-dependent thermal hydraulic technique is adopted for analyzing the local thermal hydraulic parameters and hotspot location during a flow inversion. The flow rate through the channel is determined in terms of a balance between driving and preventing forces. Friction and buoyancy forces act as resistance of the flow before a flow inversion while buoyancy force becomes the driving force after a flow inversion. By taking into account the buoyancy effect to determine the flow rate, the difference in the flow inversion time between hot and average channels is investigated: a flow inversion occurs earlier in the hot channel than in an average channel. Furthermore, the movement of the hotspot location before and after a flow inversion is investigated for a slow and fast LOFA. During a flow inversion, two temperature peaks are observed: (1) the first temperature peak is at the initiation of the LOFA, and (2) the second temperature peak is when a flow inversion occurs. The maximum temperature of the cladding is found at the second temperature peak for both LOFA analyses, and is lower than the saturation temperature
Computer simulation of thermal-hydraulic transient events in multi-circuits with multipumps
International Nuclear Information System (INIS)
Veloso, Marcelo Antonio
2003-01-01
PANTERA-2 (from Programa para Analise Termo-hidraulica de Reatores a Agua - Program for Thermal-hydraulic Analysis of Water Reactors, Version 2), whose fundamentals are described in this work, is intended to carry out rod bundle subchannel analysis in conjunction with multiloop simulation. It solves simultaneously the conservation equations of mass, axial and lateral momentum, and energy for subchannel geometry coupled with the balance equations that describe the fluid flows in any number of coolant loops connected to a pressure vessel containing the rod bundle. As far as subchannel analysis is concerned, the basic computational strategy of PANTERA-2 comes from COBRA codes, but an alternative implicit solution method oriented to the pressure field has been used to solve the finite difference approximations for the balance laws. The results provided by the subchannel model comprise the fluid density, enthalpy, flow rate, and pressure fields in the subchannels. The loop model predicts the individual loop flows, total flow through the pressure vessel, and pump rotational speeds as a function of time subsequent to the failure of any number of the coolant pumps. The flow transients in the loops may initiated by partial, total or sequential loss of electric power to the operating pumps. Transient events caused by either shaft break or rotor locking may also be simulated. The changes in rotational speed of the pumps as a function of rime are determined from a torque balance. Pump dynamic head and hydraulic torque are calculated as a function of rotational speed and volumetric flow from two polar homologous curves supplied to the code in the tabular form. In order to illustrate the analytical capability of PANTERA-2, three sample problems are presented and discussed. Comparisons between calculated and measured results indicate that the program reproduces with a good accuracy experimental data for subchannel exit temperatures and critical heat fluxes in 5x5 rod bundles. It
Analytical prediction and experimental verification of reactor safety system injection transient
International Nuclear Information System (INIS)
Roy, B.N.; Nomm, E.
1991-01-01
This paper describes the computer code that was developed for thermal hydraulic transient analysis of mixed phase fluid system and the flow tests that were carried out to validate the Code. A full scale test facility was designed to duplicate the Supplementary Shutdown System (SSS) of Savannah River Production Reactors. Several steady state and dynamic flow tests were conducted simulating the actual reactor injection transients. A dynamic multiphase fluid flow code was developed and validated with experimental results and utilized for system performance predictions and development of technical specifications for reactors. 3 refs
Applied mathematical methods in nuclear thermal hydraulics
International Nuclear Information System (INIS)
Ransom, V.H.; Trapp, J.A.
1983-01-01
Applied mathematical methods are used extensively in modeling of nuclear reactor thermal-hydraulic behavior. This application has required significant extension to the state-of-the-art. The problems encountered in modeling of two-phase fluid transients and the development of associated numerical solution methods are reviewed and quantified using results from a numerical study of an analogous linear system of differential equations. In particular, some possible approaches for formulating a well-posed numerical problem for an ill-posed differential model are investigated and discussed. The need for closer attention to numerical fidelity is indicated
Thermal hydraulic model descrition of TASS/SMR
Energy Technology Data Exchange (ETDEWEB)
Yoon, Han Young; Kim, H. C.; Chung, Y. J.; Lim, H. S.; Yang, S. H
2001-04-01
The TASS/SMR code has been developed for the safety analysis of SMART. The governing equations were applied only to the primary coolant system in TASS which had been developed at KAERI. In TASS/SMR, the solution method is improved so that the primary and secondary coolant systems are solved simultaneously. Besides the solution method, thermal-hydraulic models are incorporated, in TASS/SMR, such as non-condensible gas model, helical steam generator heat transfer model, and passive residual heat removal system (PRHRS) heat transfer model for the application to SMART. The governing equtions of TASS/SMR are based on the drift-flux model so that the accidents and transients accompaning with two-phase flow can be analized. This report describes the governing equations and solution methods used in TASS/SMR and also includes the description for the thermal hydraulic models for SMART design.
International Nuclear Information System (INIS)
Santos, G.A. dos.
1990-01-01
The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The single-phase pump characteristics are an essential feature for operational transients studies, for example, the shut-down and start-up of pump. These parameters, in terms of the homologous curves, set up the complete performance of the pump and are input for transients and accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the single-phase and two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)
Study on thermal-hydraulics during a PWR reflood phase
Energy Technology Data Exchange (ETDEWEB)
Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-10-01
In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.
Calculation of dynamic hydraulic forces in nuclear plant piping systems
International Nuclear Information System (INIS)
Choi, D.K.
1982-01-01
A computer code was developed as one of the tools needed for analysis of piping dynamic loading on nuclear power plant high energy piping systems, including reactor safety and relief value upstream and discharge piping systems. The code calculates the transient hydraulic data and dynamic forces within the one-dimensional system, caused by a pipe rupture or sudden value motion, using a fixed space and varying time grid-method of characteristics. Subcooled, superheated, homogeneous two-phase and transition flow regimes are considered. A non-equilibrium effect is also considered in computing the fluid specific volume and fluid local sonic velocity in the two-phase mixture. Various hydraulic components such as a spring loaded or power operated value, enlarger, orifice, pressurized tank, multiple pipe junction (tee), etc. are considered as boundary conditions. Comparisons of calculated results with available experimental data shows a good agreement. (Author)
Hydrological controls on transient aquifer storage in a karst watershed
Spellman, P.; Martin, J.; Gulley, J. D.
2017-12-01
While surface storage of floodwaters is well-known to attenuate flood peaks, transient storage of floodwaters in aquifers is a less recognized mechanism of flood peak attenuation. The hydraulic gradient from aquifer to river controls the magnitude of transient aquifer storage and is ultimately a function of aquifer hydraulic conductivity, and effective porosity. Because bedrock and granular aquifers tend to have lower hydraulic conductivities and porosities, their ability to attenuate flood peaks is generally small. In karst aquifers, however, extensive cave systems create high hydraulic conductivities and porosities that create low antecedent hydraulic gradients between aquifers and rivers. Cave springs can reverse flow during high discharges in rivers, temporarily storing floodwaters in the aquifer thus reducing the magnitude of flood discharge downstream. To date however, very few studies have quantified the magnitude or controls of transient aquifer storage in karst watersheds. We therefore investigate controls on transient aquifer storage by using 10 years of river and groundwater data from the Suwannee River Basin, which flows over the karstic upper Floridan aquifer in north-central Florida. We use multiple linear regression to compare the effects of three hydrological controls on the magnitude of transient aquifer storage: antecedent stage, recharge and slope of hydrograph rise. We show the dominant control on transient aquifer storage is antecedent stage, whereby lower stages result in greater magnitudes of transient aquifer storage. Our results suggest that measures of groundwater levels prior to an event can be useful in determining whether transient aquifer storage will occur and may provide a useful metric for improving predictions of flood magnitudes.
International Nuclear Information System (INIS)
Goto, Shoji; Ishizaki, Yasuo; Ohashi, Hirotada; Akiyama, Mamoru
1995-01-01
Simulation of transient two-phase flow has been performed by solving transient hydrodynamic equations. However, constitution relations used in this simulation are primarily based on steady-state experimental results. Thus it is important to understand the transient behavior of bubbles and slugs, in particular, transient behavior of the void fraction, the interfacial area and the flow pattern, to confirm the applicability of the present simulation method and to advance two-phase flow simulation further. The present study deals with measurement of transient two-phase flow. We have measured local and instantaneous void fractions using imaging techniques, and compared the experimental data with simulation results. (author)
The delay function in finite difference models for nuclear channels thermo-hydraulic transients
International Nuclear Information System (INIS)
Agazzi, A.
1977-01-01
The study of the thermo-hydraulic transients in a nuclear reactor core often requires a bi- or tri-dimensional mathematical simulation of a reactor channel. The equations involved are generally solved by means of finite-difference methods. The determination of the spatial mesh-width and the time interval is strongly conditioned by the necessity of a good accuracy in the description of the delay function which defines the transfer of thermal perturbations along the cooling channel. In this paper the effects of both space and time discretization on the delay function are considered and for the classical cases of inlet temperature step and ramp universal functions and diagrams are given in order to make possible the determination of optimal spatial mesh-width and time interval, once the requested accuracy of the model is fixed in advance
Applications of the thermit code to 3D thermal hydraulic analysis of LWR cores
International Nuclear Information System (INIS)
Reed, W.H.
1979-01-01
The THERMIT code calculates the three-dimensional transient thermal hydraulic behavior of light water reactor cores. Its two-fluid dynamics equations for two-phase flow offer improved physical modelling capability needed in the context of calculation coupled to neutron kinetics for feedback. The numerical fluid dynamics method was chosen for reliability over a wider range of transients. An improved heat transfer numerical method is presented which gives better numerical stability and accuracy. A number of example calculations are discussed which give an idea of the power and flexibility of the code
Current capabilities of transient two-phase flow instruments
International Nuclear Information System (INIS)
Solbrig, C.W.; Kondic, N.N.
1979-01-01
The measurement of two phase flow phenomena in transient conditions representative of a Loss-of-Coolant Accident requires the use of sophisticated instruments and the further development of other instruments. Measurements made in large size pipes are often flow regime dependent. The flow regimes encountered depend upon the system geometry, transient effects, heat transfer, etc. The geometries in which these measurements must be made, the instruments which are currently used, new instruments being developed, the facilities used to calibrate these instruments, and the improvements which must be made to measurement capabilities are described
Investigation on Capacitor Switching Transient Limiter with a Three phase Variable Resistance
DEFF Research Database (Denmark)
Naderi, Seyed Behzad; Jafari, Mehdi; Zandnia, Amir
2017-01-01
In this paper, a capacitor switching transient limiter based on a three phase variable resistance is proposed. The proposed structure eliminates the capacitor switching transient current and over-voltage by introducing a variable resistance to the current path with its special switching pattern...... transients on capacitor after bypassing. Analytic Analyses for this structure in transient cases are presented in details and simulations are performed by MATLAB software to prove its effectiveness....
International Nuclear Information System (INIS)
Pautz, A.; Tyobeka, B.; Ivanov, K.
2009-01-01
In new reactor designs that are still under review such as the Pebble Bed Modular Reactor (PBMR), not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400MW OECD/NEA coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate both the steady-state as well as several transient scenarios in this benchmark problem. (orig.)
Transient phases during fast crystallization of organic thin films from solution
Wan, Jing; Li, Yang; Ulbrandt, Jeffrey G.; Smilgies, Detlef-M.; Hollin, Jonathan; Whalley, Adam C.; Headrick, Randall L.
2016-01-01
We report an in situ microbeam grazing incidence X-ray scattering study of 2,7-dioctyl[1]benzothieno[3,2-b][1]benzothiophene (C8-BTBT) organic semiconductor thin film deposition by hollow pen writing. Multiple transient phases are observed during the crystallization for substrate temperatures up to ≈93 °C. The layered smectic liquid-crystalline phase of C8-BTBT initially forms and preceedes inter-layer ordering, followed by a transient crystalline phase for temperature >60 °C, and ultimately the stable phase. Based on these results, we demonstrate a method to produce extremely large grain size and high carrier mobility during high-speed processing. For high writing speed (25 mm/s), mobility up to 3.0 cm2/V-s has been observed.
International Nuclear Information System (INIS)
Lapins, J.; Seubert, A.; Buck, M.; Bader, J.; Laurien, E.
2011-01-01
Comprehensive safety studies of high temperature gas cooled reactors (HTR) require full three dimensional coupled treatments of both neutron kinetics and thermal-hydraulics. In a common effort, GRS and IKE developed the coupled code system TORT-TD/ATTICA3D for pebble bed type HTR that connects the 3-D transient discrete-ordinates transport code TORT-TD with the 3-D porous medium thermal-hydraulics code ATTICA3D. In this paper, the physical models and calculation capabilities of TORT-TD and ATTICA3D are presented, focusing on model improvements in ATTICA3D and extensions made in TORT-TD related to HTR application. For first applications, the OECD/NEA/NSC PBMR-400 benchmark has been chosen. Results obtained with TORT-TD/ATTICA3D will be shown for transient exercises, e.g. control rod withdrawal and a control rod ejection. Results are compared to other benchmark participants' solutions with special focus on fuel temperature modelling features of ATTICA3D. The provided “grey-curtain” nuclear cross section libraries have been used. First results on 3-D effects during a control rod withdrawal transient will be presented. (author)
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)
Scaling Analysis of the Single-Phase Natural Circulation: the Hydraulic Similarity
International Nuclear Information System (INIS)
Yu, Xin-Guo; Choi, Ki-Yong
2015-01-01
These passive safety systems all rely on the natural circulation to cool down the reactor cores during an accident. Thus, a robust and accurate scaling methodology must be developed and employed to both assist in the design of a scaled-down test facility and guide the tests in order to mimic the natural circulation flow of its prototype. The natural circulation system generally consists of a heat source, the connecting pipes and several heat sinks. Although many applauding scaling methodologies have been proposed during last several decades, few works have been dedicated to systematically analyze and exactly preserve the hydraulic similarity. In the present study, the hydraulic similarity analyses are performed at both system and local level. By this mean, the scaling criteria for the exact hydraulic similarity in a full-pressure model have been sought. In other words, not only the system-level but also the local-level hydraulic similarities are pursued. As the hydraulic characteristics of a fluid system is governed by the momentum equation, the scaling analysis starts with it. A dimensionless integral loop momentum equation is derived to obtain the dimensionless numbers. In the dimensionless momentum equation, two dimensionless numbers, the dimensionless flow resistance number and the dimensionless gravitational force number, are identified along with a unique hydraulic time scale, characterizing the system hydraulic response. A full-height full-pressure model is also made to see which model among the full-height model and reduced-height model can preserve the hydraulic behavior of the prototype. From the dimensionless integral momentum equation, a unique hydraulic time scale, which characterizes the hydraulic response of a single-phase natural circulation system, is identified along with two dimensionless parameters: the dimensionless flow resistance number and the dimensionless gravitational force number. By satisfying the equality of both dimensionless numbers
Scaling Analysis of the Single-Phase Natural Circulation: the Hydraulic Similarity
Energy Technology Data Exchange (ETDEWEB)
Yu, Xin-Guo; Choi, Ki-Yong [KAERI, Daejeon (Korea, Republic of)
2015-05-15
These passive safety systems all rely on the natural circulation to cool down the reactor cores during an accident. Thus, a robust and accurate scaling methodology must be developed and employed to both assist in the design of a scaled-down test facility and guide the tests in order to mimic the natural circulation flow of its prototype. The natural circulation system generally consists of a heat source, the connecting pipes and several heat sinks. Although many applauding scaling methodologies have been proposed during last several decades, few works have been dedicated to systematically analyze and exactly preserve the hydraulic similarity. In the present study, the hydraulic similarity analyses are performed at both system and local level. By this mean, the scaling criteria for the exact hydraulic similarity in a full-pressure model have been sought. In other words, not only the system-level but also the local-level hydraulic similarities are pursued. As the hydraulic characteristics of a fluid system is governed by the momentum equation, the scaling analysis starts with it. A dimensionless integral loop momentum equation is derived to obtain the dimensionless numbers. In the dimensionless momentum equation, two dimensionless numbers, the dimensionless flow resistance number and the dimensionless gravitational force number, are identified along with a unique hydraulic time scale, characterizing the system hydraulic response. A full-height full-pressure model is also made to see which model among the full-height model and reduced-height model can preserve the hydraulic behavior of the prototype. From the dimensionless integral momentum equation, a unique hydraulic time scale, which characterizes the hydraulic response of a single-phase natural circulation system, is identified along with two dimensionless parameters: the dimensionless flow resistance number and the dimensionless gravitational force number. By satisfying the equality of both dimensionless numbers
OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Third Workshop (V1000-CT3)
International Nuclear Information System (INIS)
2005-01-01
The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. The technical topics presented at this workshop were: Review of the benchmark activities after the 2. Workshop; - Discussion of participant's feedback and introduced modifications
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
Problems of heat transfer and hydraulics of two-phase media
Kutateladze, S S
1969-01-01
Problems of Heat Transfer and Hydraulics of Two-Phase Media presents the theory of heat transfer and hydrodynamics. This book discusses the various aspects of heat transfer and the flow of two-phase systems. Organized into two parts encompassing 22 chapters, this book starts with an overview of the laws of similarity for heat transfer to or from a flowing liquid with various physical properties and allowed for variation in viscosity and thermal conductivity. This book then explores the general functional relationship that exists between viscosity and thermal conductivity for thermodynamically
OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Fourth Workshop (V100-CT4)
International Nuclear Information System (INIS)
2006-01-01
The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. Since the previous coupled code benchmarks indicated that further development of the mixing computation models in the integrated codes is necessary, a coolant mixing experiment and
Transient phases during fast crystallization of organic thin films from solution
Directory of Open Access Journals (Sweden)
Jing Wan
2016-01-01
Full Text Available We report an in situ microbeam grazing incidence X-ray scattering study of 2,7-dioctyl[1]benzothieno[3,2-b][1]benzothiophene (C8-BTBT organic semiconductor thin film deposition by hollow pen writing. Multiple transient phases are observed during the crystallization for substrate temperatures up to ≈93 °C. The layered smectic liquid-crystalline phase of C8-BTBT initially forms and preceedes inter-layer ordering, followed by a transient crystalline phase for temperature >60 °C, and ultimately the stable phase. Based on these results, we demonstrate a method to produce extremely large grain size and high carrier mobility during high-speed processing. For high writing speed (25 mm/s, mobility up to 3.0 cm2/V-s has been observed.
Energy Technology Data Exchange (ETDEWEB)
Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-06-01
A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other
International Nuclear Information System (INIS)
Tyobeka, B.; Ivanov, K.; Pautz, A.
2007-01-01
In the advent of increased demand for safety and economics of nuclear power plants, nuclear engineers and designers are called upon to develop advanced computation tools. In these developments, space-time effects in the dynamics of nuclear reactors must be considered within the framework of a full 3-dimensional treatment of both neutron kinetics and thermal hydraulics. In a recent effort at the Pennsylvania State University, a time-dependent version of the discrete ordinates transport code DORT, DORT-TD was coupled to a 2-dimensional core thermal hydraulics code THERMIX-DIREKT. In the coupling process, a feedback model was developed to account for the feedback effects and was implemented into DORT-TD. During the calculation process for each spatial node of the DORT-TD core model, feedback parameters representative of this node are passed to the feedback module. Using these values, cross section tables are then interpolated for the appropriate macroscopic cross section values. The updated macroscopic cross sections are passed back to DORT-TD to perform transport core calculations, and the power distribution is transferred to THERMIX-DIREKT to obtain the relevant thermal-hydraulics data in turn, and this calculation loop continues. In this paper, DORT-TD/THERMIX is used to simulate transients of interest in the PBMR (Pebble Bed Modular Reactor) safety using established benchmark problems: load change from 100% to 40% power and fast control rod ejection (PBMR-268 benchmark problem). The results obtained are compared with those obtained using the diffusion-based module of the code. The results are only preliminary and so far show that diffusion theory is not such a bad approximation for PBMR for the prediction of integral parameters
Three-Phase Bone Scintigraphic Diagnosis of Acute Transient Synovitis
International Nuclear Information System (INIS)
Chung, Soo Kyo; Lee, Myung Hee; Kim, Choon Yul; Bahk, Yong Whee
1985-01-01
Acute transient synovitis of the hip presents clinically pain and limping. But in the majority of the cases, definite positive findings are not manifest in roentgenogram in its early phase. However radionuclide bone imaging combines with the assessment of vascularization and bone tracer uptake is of great value in solving this diagnostic problem. The materials for this study consisted of 29 children with acute transient synovitis of the hip, characterized by symptoms and physical signs of an arthritis, negative X-ray findings and disappearance of all symptoms and signs within a short period of time. They were twenty males and 9 females and age ranged from 1 to 12 years. We took pelvic roentgenogram in AP and frog-leg views. After intravenous bolus injection of 10 to 15 mCi of 99m Tc-methylene diphosphonate, 24 sequential image of the pelvis was taken at 2-second interval for blood flow study. The scintigrams were made using a gamma camera with high resolution parallel hole collimator. Blood pool imaging was obtained at 2 minutes after tracer administration. After 3 hours, static images were taken and then close-up image of the hip using pin-hole collimator was followed. The results were as follows: 1) Bone scintigram was much more sensitive than conventional roentgenogram in diagnosis of acute transient synovitis of the hip. 2) Three-phase imagings showed increased vascular activities in blood pool scintigrams in 96%. 3) Pin-hole imaging showed increased tracer uptake in the regional bones of the hip, particularly in the medial aspect of femoral head and acetabulum. 4) We confirmed that three-phase imaging reinforced with pin-hale technique were very useful in diagnose of acute transient synovitis of the hip.
International Nuclear Information System (INIS)
Yudov, Y.V.
2001-01-01
The functional part of the KORSAR computer code is based on the computational unit for the reactor system thermal-hydraulics and other thermal power systems with water cooling. The two-phase flow dynamics of the thermal-hydraulic network is modelled by KORSAR in one-dimensional two-fluid (non-equilibrium and nonhomogeneous) approximation with the same pressure of both phases. Each phase is characterized by parameters averaged over the channel sections, and described by the conservation equations for mass, energy and momentum. The KORSAR computer code relies upon a novel approach to mathematical modelling of two-phase dispersed-annular flows. This approach allows a two-fluid model to differentiate the effects of the liquid film and droplets in the gas core on the flow characteristics. A semi-implicit numerical scheme has been chosen for deriving discrete analogs the conservation equations in KORSAR. In the semi-implicit numerical scheme, solution of finite-difference equations is reduced to the problem of determining the pressure field at a new time level. For the one-channel case, the pressure field is found from the solution of a system of linear algebraic equations by using the tri-diagonal matrix method. In the branched network calculation, the matrix of coefficients in the equations describing the pressure field is no longer tri-diagonal but has a sparseness structure. In this case, the system of linear equations for the pressure field can be solved with any of the known classical methods. Such an approach is implemented in the existing best-estimate thermal-hydraulic computer codes (TRAC, RELAP5, etc.) For the KORSAR computer code, we have developed a new non-iterative method for calculating the pressure field in the network of any topology. This method is based on the tri-diagonal matrix method and performs well when solving the thermal-hydraulic network problems. (author)
Liquid metal thermal-hydraulics
International Nuclear Information System (INIS)
Kottowski-Duemenil, H.M.
1994-01-01
This textbook is a report of the 26 years activity of the Liquid Metal Boiling Working Group (LMBWG). It summarizes the state of the art of liquid metal thermo-hydraulics achieved through the collaboration of scientists concerned with the development of the Fast Breeder Reactor. The first chapter entitled ''Liquid Metal Boiling Behaviour'', presents the background and boiling mechanisms. This section gives the reader a brief but thorough survey on the superheat phenomena in liquid metals. The second chapter of the text, ''A Review of Single and Two-Phase Flow Pressure Drop Studies and Application to Flow Stability Analysis of Boiling Liquid Metal Systems'' summarizes the difficulty of pressure drop simulation of boiling sodium in core bundles. The third chapter ''Liquid Metal Dry-Out Data for Flow in Tubes and Bundles'' describes the conditions of critical heat flux which limits the coolability of the reactor core. The fourth chapter dealing with the LMFBR specific topic of ''Natural Convection Cooling of Liquid Metal Systems''. This chapter gives a review of both plant experiments and out-of-pile experiments and shows the advances in the development of computing power over the past decade of mathematical modelling ''Subassembly Blockages Suties'' are discussed in chapter five. Chapter six is entitled ''A Review of the Methods and Codes Available for the Calculation on Thermal-Hydraulics in Rod-Cluster and other Geometries, Steady state and Transient Boiling Flow Regimes, and the Validation achieves''. Codes available for the calculation of thermal-hydraulics in rod-clusters and other geometries are reviewed. Chapter seven, ''Comparative Studies of Thermohydraulic Computer Code Simulations of Sodium Boiling under Loss of Flow Conditions'', represents one of the key activities of the LMBWG. Several benchmark exercises were performed with the aim of transient sodium boiling simulation in single channels and bundle blockages under steady state conditions and loss of
Overpower transient in the first wall cooling system of NET/ITER
International Nuclear Information System (INIS)
Komen, E.M.J.; Koning, H.
1993-09-01
The overpower transient from a plasma power excursion. The overpower transient considered in this report results from a postulated linear increase of the plasma power from the nominal generated power to four times this nominal power in 30 s. The Next European Torus (NET) design or the International Thermonuclear Experimental Reactor (ITER) design will be cooled by a number of separate cooling systems. The most important cooling systems are: The first wall cooling system, the blanket cooling system, the divertor cooling system, and the shield cooling system. In this report, the thermal-hydraulic analysis of the above-mentioned overpower transient will be presented for the first wall cooling system of NET/ITER. During overpower transients, the fusion power will increase to less than four times the nominal power. For this reason, the overpower transient considered in this report is the worst case scenario. The analysis of the thermal-hydraulic system behaviour during the considered overpower transient has been performed for a coolant temperature of 333 K (60 C) in the first wall inlet manifolds and 433 K (160 C) in the first wall outlet manifolds. The analysis has been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analysis, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall. (orig.)
Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel
Energy Technology Data Exchange (ETDEWEB)
Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)
2016-01-01
Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm^{2} and temporary heat flux limit of 600 W/cm^{2}. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.
International Nuclear Information System (INIS)
Pietro Alessandro Di Maio; Valerio Tomarchio; Giuseppe Vella; Irene Zammuto; Giovanni Dell'Orco
2005-01-01
Full text of publication follows: The divertor is one of the most challenging components of the next step ITER nuclear fusion reactor. It is aimed at controlling the characteristics of boundary plasma, reducing the impurities in the plasma and sustaining the heat and particle fluxes arising from it, during normal and transient operations as well as during disruption events. The ITER divertor consists of 54 cassettes, each one mainly composed of three Plasma-Facing Components (PFCs), namely the inner vertical target, the outer vertical target and the dome-liner, actively cooled by subcooled pressurized water. Each PFC consists in a number of plasma facing units, cooled in parallel and assembled onto a supporting structure. The water maximum total flow rate, for the whole divertor, should be 1000 kg/s, with 100-150 deg. C inlet/outlet temperatures, 4.2 MPa inlet pressure and a maximum pressure drop of 1.4 MPa. The PFCs are cooled in series, with a maximum water velocity in the channel of 11 m/s, whilst the water coolant is routed via the cassette body. Due to the extremely high heat loads expected onto the PFCs (up to 20 MW/m 2 over 20 s), the hydraulic design of the divertor is particularly demanding. It shall ensure that the foreseen flow rate actually reaches each plasma-facing unit to ensure an adequate cooling and to prevent any risk of Critical Heat Flux (CHF). Sufficient margin ( > 40 %) to avoid the reaching of a CHR limit on the PFCs could be obtained by using hypervapotron design inside the flat channels and swirl flow turbulence tape promoters inside the vertical target cooling tubes. Furthermore the overall pressure drop and flow rate shall be within the specified design limit to avoid an unduly high pumping power. Another important issue is the definition of a proper procedure to drain the coolant and dry the divertor components prior to the maintenance operations as well as to refill them with water after maintenance, ensuring a complete elimination of
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
11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)
International Nuclear Information System (INIS)
Lemonnier, H.
2005-01-01
; aerosol transport, deposition and re-entrainment; steam generators thermal-hydraulics; system codes development and assessment; uncertainties analysis; diffuse interface methods and interface tracking methods; C - severe accidents and fires: molten core natural convection and physico-chemical phenomena, modeling and experiments; fuel coolant interaction, modeling and experiments; debris bed cooling; combustion and fires, modeling and experiments; molten corium concrete interaction; D - advanced code developments: fast transient modelling and experiments; multidimensional single-phase or two-phase flow and heat transfer modeling; neutronics and thermal-hydraulics coupling; fluid and structures mechanical interactions; coupled thermal-hydraulics of fluids and structures; thermal-hydraulic dependent corrosion and ablation; E - operation and safety of existing reactors: instabilities and nonlinear dynamics; NPP transients and accidents analysis; RBMK and VVER safety analysis, including the OECD benchmark; F - experimental thermal-hydraulics: boiling heat transfer; CHF and post-CHF heat transfer; condensation heat transfer; integral testing; vibrations, wear and thermal fatigue phenomena; fuel design and performance; G - advanced reactors thermal-hydraulics (gen IV, INPRO, fusion, hydrogen production): accelerator driven reactors; advanced pressurized water reactors thermal-hydraulics; gas cooled fast reactors; gas cooled high temperature reactors; lead and lead-bismuth cooled reactors; future and existing sodium cooled reactors; molten salt reactors; H - waste management thermal-hydraulics: thermal-hydraulics problems related to waste processing and storage; I - thermal-hydraulics of non electricity generating nuclear equipment: sono-fusion (cavitation induced bubble fusion; hydrogen producing nuclear reactors
International Nuclear Information System (INIS)
Leavell, W.H.; Mullens, J.A.
1981-01-01
A computational algorithm has been developed to measure transient, phase-interface velocity in two-phase, steam-water systems. The algorithm will be used to measure the transient velocity of steam-water mixture during simulated PWR reflood experiments. By utilizing signals produced by two, spatially separated impedance probes immersed in a two-phase mixture, the algorithm computes the average transit time of mixture fluctuations moving between the two probes. This transit time is computed by first, measuring the phase shift between the two probe signals after transformation to the frequency domain and then computing the phase shift slope by a weighted least-squares fitting technique. Our algorithm, which has been tested with both simulated and real data, is able to accurately track velocity transients as fast as 4 m/s/s
Kozlovcev, Petr; Přikryl, Richard; Přikrylová, Jiřina
2015-04-01
In contrast to modern ordinary Portland cement production from finely ground raw material blends, ancient burning of hydraulic lime was conducted by burning larger pieces of natural raw material. Due to natural variability of raw material composition, exploitation of different beds from even one formation can result the product with significantly different composition and/or properties. Prague basin (Neoproterozoic to pre-Variscan Palaeozoic of the central part of the Bohemian Massif - the so-called Barrandian area, Czech Republic) represents a classical example of the limestone-rich region with long-term history of limestone burning for quick lime and/or various types of hydraulic binders. Due to the fact that burning of natural hydraulic lime has been abandoned in this region at the turn of 19th/20th c., significant gap in knowledge on the behavior of various limestone types and on the influence of minor variance in composition on the quality of burned product is encountered. Moreover, the importance of employment of larger pieces of raw material for burning for the development of proper phase-to-phase relationships (i.e. development of hydraulic phases below sintering temperature at mutual contacts of minerals) has not been examined before. To fill this gap, a representative specimens of major limestone types from the Prague basin have been selected for experimental study: Upper Silurian limestone types (Přídolí and Kopanina Lms.), and Lower Devonian limestones (Radotín, Kotýs, Řeporyje, Dvorce-Prokop, and Zlíchov Lms.). Petrographic character of the experimental material was examined by polarizing microscopy, cathodoluminescence, scanning electron microscopy with an energy dispersive spectrometer (SEM-EDS), and X-ray diffraction (XRD) of insoluble residue. Based on the data from wet silicate analyses, modal composition of studied impure limestones was computed. Experimental raw material was burned in laboratory electric furnace at 1000 and 1200°C for 3
International Nuclear Information System (INIS)
Fischer, K.; Schall, M.; Wolf, L.
1993-01-01
The present final report comprises the major results of Phase II of the CEC thermal-hydraulic benchmark exercise on Fiploc verification experiment F2 in the Battelle model containment, experimental phases 2, 3 and 4, which was organized and sponsored by the Commission of the European Communities for the purpose of furthering the understanding and analysis of long-term thermal-hydraulic phenomena inside containments during and after severe core accidents. This benchmark exercise received high European attention with eight organizations from six countries participating with eight computer codes during phase 2. Altogether 18 results from computer code runs were supplied by the participants and constitute the basis for comparisons with the experimental data contained in this publication. This reflects both the high technical interest in, as well as the complexity of, this CEC exercise. Major comparison results between computations and data are reported on all important quantities relevant for containment analyses during long-term transients. These comparisons comprise pressure, steam and air content, velocities and their directions, heat transfer coefficients and saturation ratios. Agreements and disagreements are discussed for each participating code/institution, conclusions drawn and recommendations provided. The phase 2 CEC benchmark exercise provided an up-to-date state-of-the-art status review of the thermal-hydraulic capabilities of present computer codes for containment analyses. This exercise has shown that all of the participating codes can simulate the important global features of the experiment correctly, like: temperature stratification, pressure and leakage, heat transfer to structures, relative humidity, collection of sump water. Several weaknesses of individual codes were identified, and this may help to promote their development. As a general conclusion it may be said that while there is still a wide area of necessary extensions and improvements, the
Multiphase flow models for hydraulic fracturing technology
Osiptsov, Andrei A.
2017-10-01
The technology of hydraulic fracturing of a hydrocarbon-bearing formation is based on pumping a fluid with particles into a well to create fractures in porous medium. After the end of pumping, the fractures filled with closely packed proppant particles create highly conductive channels for hydrocarbon flow from far-field reservoir to the well to surface. The design of the hydraulic fracturing treatment is carried out with a simulator. Those simulators are based on mathematical models, which need to be accurate and close to physical reality. The entire process of fracture placement and flowback/cleanup can be conventionally split into the following four stages: (i) quasi-steady state effectively single-phase suspension flow down the wellbore, (ii) particle transport in an open vertical fracture, (iii) displacement of fracturing fluid by hydrocarbons from the closed fracture filled with a random close pack of proppant particles, and, finally, (iv) highly transient gas-liquid flow in a well during cleanup. The stage (i) is relatively well described by the existing hydralics models, while the models for the other three stages of the process need revisiting and considerable improvement, which was the focus of the author’s research presented in this review paper. For stage (ii), we consider the derivation of a multi-fluid model for suspension flow in a narrow vertical hydraulic fracture at moderate Re on the scale of fracture height and length and also the migration of particles across the flow on the scale of fracture width. At the stage of fracture cleanaup (iii), a novel multi-continua model for suspension filtration is developed. To provide closure relationships for permeability of proppant packings to be used in this model, a 3D direct numerical simulation of single phase flow is carried out using the lattice-Boltzmann method. For wellbore cleanup (iv), we present a combined 1D model for highly-transient gas-liquid flow based on the combination of multi-fluid and
Energy Technology Data Exchange (ETDEWEB)
Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)
2017-03-15
Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.
International Nuclear Information System (INIS)
Badea, Aurelian F.; Cacuci, Dan G.
2017-01-01
Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.
Learning from anticipated and abnormal plant transients
International Nuclear Information System (INIS)
Varnado, B.
1983-01-01
A report is given of the American Nuclear Society topical meeting on Anticipated and Abnormal Transients in Light Water Reactors held in Jackson, Wyoming in September 1983. Industry involvement in the evaluation of operating experience, human error contributions, transient management, thermal hydraulic modelling, the role of probabilistic risk assessment and the cost of transient incidents are discussed. (U.K.)
Energy Technology Data Exchange (ETDEWEB)
Azad, Hamed Moslehi; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering
2017-07-15
Thermal hydraulic analysis of sodium boiling in fuel assemblies is an important issue in safety of sodium cooled reactors and subchannel method is an efficient approach in transient two phase flow analyses. Almost all of the subchannel codes which use two-fluid model in two phase flow analysis, are based on semi implicit algorithm. With the full implicit method it is possible to use larger time steps. In order to compare the semi implicit algorithm with full implicit algorithm, two transient subchannel numerical programs which one is based on semi implicit algorithm and the other is based on full implicit algorithm have been written in FORTRAN in this work for simulation of transients in sodium cooled Kompakter-Natriumsiede-Kreislauf (KNS) at the former Kernforschungszentrum Karlsruhe (KfK) in Germany.
Zhao, Zhanfeng; Illman, Walter A.
2018-04-01
Previous studies have shown that geostatistics-based transient hydraulic tomography (THT) is robust for subsurface heterogeneity characterization through the joint inverse modeling of multiple pumping tests. However, the hydraulic conductivity (K) and specific storage (Ss) estimates can be smooth or even erroneous for areas where pumping/observation densities are low. This renders the imaging of interlayer and intralayer heterogeneity of highly contrasting materials including their unit boundaries difficult. In this study, we further test the performance of THT by utilizing existing and newly collected pumping test data of longer durations that showed drawdown responses in both aquifer and aquitard units at a field site underlain by a highly heterogeneous glaciofluvial deposit. The robust performance of the THT is highlighted through the comparison of different degrees of model parameterization including: (1) the effective parameter approach; (2) the geological zonation approach relying on borehole logs; and (3) the geostatistical inversion approach considering different prior information (with/without geological data). Results reveal that the simultaneous analysis of eight pumping tests with the geostatistical inverse model yields the best results in terms of model calibration and validation. We also find that the joint interpretation of long-term drawdown data from aquifer and aquitard units is necessary in mapping their full heterogeneous patterns including intralayer variabilities. Moreover, as geological data are included as prior information in the geostatistics-based THT analysis, the estimated K values increasingly reflect the vertical distribution patterns of permeameter-estimated K in both aquifer and aquitard units. Finally, the comparison of various THT approaches reveals that differences in the estimated K and Ss tomograms result in significantly different transient drawdown predictions at observation ports.
Gulping phenomena in transient countercurrent two-phase flow
International Nuclear Information System (INIS)
Tehrani, Ali A.K.
2001-04-01
Apart from previous work on countercurrent gas-liquid flow, transient tank drainage through horizontal off-take pipes is described, including experimental procedure, flow pattern on observations and countercurrent flow limitation results. A separate chapter is devoted to countercurrent two-phase flow in a pressurised water reactor hot-leg scaled model. Results concerning low head flooding, high head and loss of bowl flooding, transient draining of the steam generator and pressure variation and bubble detachment are presented. The following subjects are covered as well: draining of sealed tanks of vertical pipes, unsteady draining of closed vessel via vertical tube, unsteady filling of a closed vessel via vertical tube from a constant head reservoir. Practical significance of the results obtained is discussed
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
NALAP: an LMFBR system transient code
International Nuclear Information System (INIS)
Martin, B.A.; Agrawal, A.K.; Albright, D.C.; Epel, L.G.; Maise, G.
1975-07-01
NALAP is a LMFBR system transient code. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic response of sodium cooled fast breeder reactors when subjected to postulated accidents such as a massive pipe break as well as a variety of other upset conditions that do not disrupt the system geometry. Various components of the plant are represented by control volumes. These control volumes are connected by junctions some of which may be leak or fill junctions. The fluid flow equations are modeled as compressible, single-stream flow with momentum flux in one dimension. The transient response is computed by integrating the thermal-hydraulic conservation equations from user-initialized operating conditions by an implicit numerical scheme. Point kinetics approximation is used to represent the time dependent heat generation in the reactor core
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.)
Adaptive PID and Model Reference Adaptive Control Switch Controller for Nonlinear Hydraulic Actuator
Directory of Open Access Journals (Sweden)
Xin Zuo
2017-01-01
Full Text Available Nonlinear systems are modeled as piecewise linear systems at multiple operating points, where the operating points are modeled as switches between constituent linearized systems. In this paper, adaptive piecewise linear switch controller is proposed for improving the response time and tracking performance of the hydraulic actuator control system, which is essentially piecewise linear. The controller composed of PID and Model Reference Adaptive Control (MRAC adaptively chooses the proportion of these two components and makes the designed system have faster response time at the transient phase and better tracking performance, simultaneously. Then, their stability and tracking performance are analyzed and evaluated by the hydraulic actuator control system, the hydraulic actuator is controlled by the electrohydraulic system, and its model is built, which has piecewise linear characteristic. Then the controller results are compared between PID and MRAC and the switch controller designed in this paper is applied to the hydraulic actuator; it is obvious that adaptive switch controller has better effects both on response time and on tracking performance.
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
Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition
International Nuclear Information System (INIS)
Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.
1996-12-01
Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76 o C which is much less than the limiting maximum value of fuel temperature of 1150 0 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000 o C, the calculated outer cladding wall temperature was observed to be 702.39 0 C compared to a value of 760 o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)
Thermally Actuated Hydraulic Pumps
Jones, Jack; Ross, Ronald; Chao, Yi
2008-01-01
Thermally actuated hydraulic pumps have been proposed for diverse applications in which direct electrical or mechanical actuation is undesirable and the relative slowness of thermal actuation can be tolerated. The proposed pumps would not contain any sliding (wearing) parts in their compressors and, hence, could have long operational lifetimes. The basic principle of a pump according to the proposal is to utilize the thermal expansion and contraction of a wax or other phase-change material in contact with a hydraulic fluid in a rigid chamber. Heating the chamber and its contents from below to above the melting temperature of the phase-change material would cause the material to expand significantly, thus causing a substantial increase in hydraulic pressure and/or a substantial displacement of hydraulic fluid out of the chamber. Similarly, cooling the chamber and its contents from above to below the melting temperature of the phase-change material would cause the material to contract significantly, thus causing a substantial decrease in hydraulic pressure and/or a substantial displacement of hydraulic fluid into the chamber. The displacement of the hydraulic fluid could be used to drive a piston. The figure illustrates a simple example of a hydraulic jack driven by a thermally actuated hydraulic pump. The pump chamber would be a cylinder containing encapsulated wax pellets and containing radial fins to facilitate transfer of heat to and from the wax. The plastic encapsulation would serve as an oil/wax barrier and the remaining interior space could be filled with hydraulic oil. A filter would retain the encapsulated wax particles in the pump chamber while allowing the hydraulic oil to flow into and out of the chamber. In one important class of potential applications, thermally actuated hydraulic pumps, exploiting vertical ocean temperature gradients for heating and cooling as needed, would be used to vary hydraulic pressures to control buoyancy in undersea research
Contribution to the theoretical study of transient two-phase flows
International Nuclear Information System (INIS)
Achard, J.L.
1978-12-01
The work presented in this paper has been given rise from the existence of violent boiling phenomena of the coolant that have been revealed by reactor safety studies with water and sodium. The aim as to describe in a basic mammer, one of these phenomena called ''chugging'' or ''choucage''. The experimental part of this work concerns two original works concerning the temperature measurement at the wall; a device is proposed to evaluate the contact resistance and the thermal inertia of the thermocouple; from the measurements that have been obtained, the flux the wall transfers to the flow and the temperature of the internal wall surface are deduced. A statistical method is developed for dispersed two-phase flow study, to establish: 1) a mass transfer law, 2) a law of change of the flow configuration. The proposed model contains: 1) for the dispersed phase (vapor bubbles), the basic momentum transport equations; 2) for the continuous phase (liquid), the transport equations of the classical formulation. The statistical formulation introduces the interaction phenomenon between the phases before applying the operation of the average (homogenization method); it allows to introduce the coalescence phenomena of bubbles. Finally, structures of exchange laws for transient laminar flows are proposed: transient linear momentum exchange law; possible structures of heat exchange laws [fr
Transient phases during crystallization of solution-processed organic thin films
Wan, Jing; Li, Yang; Ulbrandt, Jeffery; Smilgies, Detlef-M.; Hollin, Jonathan; Whalley, Adam; Headrick, Randall
We report an in-situ study of 2,7-dioctyl[1]benzothieno[3,2-b][1]benzothiophene (C8-BTBT) organic semiconductor thin film deposition from solution via hollow pen writing, which exhibits multiple transient phases during crystallization. Under high writing speed (25 mm/s) the films have an isotropic morphology, although the mobilities range up to 3.0 cm2/V.s. To understand the crystallization in this highly non-equilibrium regime, we employ in-situ microbeam grazing incidence wide-angle X-ray scattering combined with optical video microscopy at different deposition temperatures. A sequence of crystallization was observed in which a layered liquid-crystalline (LC) phase of C8-BTBT precedes inter-layer ordering. For films deposited above 80ºC, a transition from LC phase to a transient crystalline state that we denote as Cr1 occurs after a temperature-dependent incubation time, which is consistent with classical nucleation theory. After an additional ~ 0.5s, Cr1 transforms to the final stable structure Cr2. Based on these results, we demonstrate a method to produce large crystalline grain size and high carrier mobility during high-speed processing by controlling the nucleation rate during the transformation from the LC phase. Nsf DMR-1307017, NSF DMR-1332208.
One-dimensional two-phase thermal hydraulics (ENSTA course)
International Nuclear Information System (INIS)
Olive, J.
1995-11-01
This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends
Transient analysis and thermal hydraulic margins of GHARR-1 using the PARET/NAL code
International Nuclear Information System (INIS)
Adoo, N.A.
2009-06-01
The PARET code has been adapted by the IAEA for testing transient behaviour in research reactors. The PARET code provides a coupled thermal hydrodynamic and point kinetics capability with a continuous reactivity feedback and an optional voiding model that estimates the voiding produced by the subcooled boiling. The present version of the PARET/ANL 73 code provides a convenient means of assessing the various models and correlations proposed for the use in the analysis of research reactor behaviour. The Monte Carlo N-Particle code (MCNP) has been used to obtain power peaking profile for a two channel PARET/ANL model. A PARET model with the corresponding neutronics and thermal hydraulic characteristics for the miniature neutron source reactor (MNSR) has been used to simulate reactivity accidents for the Ghana Research Reactor - 1(GHARR-1) under the MNSR operation conditions of natural circulation, normal operation and reactivity insertion accidents. The simulation results via the insertion of large reactivity demonstrated the high inherent safety features of the MNSR for which the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The hot channel peaking factors for both radial and axial were found to be 1.17 and 1.44 respectively. Thermal hydraulic performance characteristics were investigated and the safety margins determined. The peak clad and coolant temperatures ranged from 59.18 0 C to 106.75 0 C and 42.95 0 C to 178.44 0 C respectively at which nucleate boiling will occur within the flow channels of the core. (au)
Results of LWR core transient benchmarks
International Nuclear Information System (INIS)
Finnemann, H.; Bauer, H.; Galati, A.; Martinelli, R.
1993-10-01
LWR core transient (LWRCT) benchmarks, based on well defined problems with a complete set of input data, are used to assess the discrepancies between three-dimensional space-time kinetics codes in transient calculations. The PWR problem chosen is the ejection of a control assembly from an initially critical core at hot zero power or at full power, each for three different geometrical configurations. The set of problems offers a variety of reactivity excursions which efficiently test the coupled neutronic/thermal - hydraulic models of the codes. The 63 sets of submitted solutions are analyzed by comparison with a nodal reference solution defined by using a finer spatial and temporal resolution than in standard calculations. The BWR problems considered are reactivity excursions caused by cold water injection and pressurization events. In the present paper, only the cold water injection event is discussed and evaluated in some detail. Lacking a reference solution the evaluation of the 8 sets of BWR contributions relies on a synthetic comparative discussion. The results of this first phase of LWRCT benchmark calculations are quite satisfactory, though there remain some unresolved issues. It is therefore concluded that even more challenging problems can be successfully tackled in a suggested second test phase. (authors). 46 figs., 21 tabs., 3 refs
Krishnan, S.; Garimella, S V
2004-01-01
A transient thermal analysis is performed to investigate thermal control of power semiconductors using phase change materials, and to compare the performance of this approach to that of copper heat sinks. Both the melting of the phase change material under a transient power spike input, as well as the resolidification process, are considered. Phase change materials of different kinds (paraffin waxes and metallic alloys) are considered, with and without the use of thermal conductivity enhancer...
International Nuclear Information System (INIS)
Griggs, D.P.; Kazimi, M.S.; Henry, A.F.
1982-01-01
The initial development of TITAN, a three-dimensional coupled neutronics/thermal-hydraulics code for LWR safety analysis, has been completed. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code THERMIT with the appropriate feedback mechanisms modeled. A detailed steady-state and transient coupling scheme based on the tandem technique was implemented in accordance with the important structural and operational characteristics of QUANDRY and THERMIT. A two channel sample problem formed the basis for steady-state and transient analyses performed with TITAN. TITAN steady-state results were compared with those obtained with MEKIN and showed good agreement. Null transients, simulated turbine trip transients, and a rod withdrawal transient were analyzed with TITAN and reasonable results were obtained
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
Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core
International Nuclear Information System (INIS)
Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar
2005-01-01
Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)
Modelling of Moving Coil Actuators in Fast Switching Valves Suitable for Digital Hydraulic Machines
DEFF Research Database (Denmark)
Nørgård, Christian; Roemer, Daniel Beck; Bech, Michael Møller
2015-01-01
an estimation of the eddy currents generated in the actuator yoke upon current rise, as they may have significant influence on the coil current response. The analytical model facilitates fast simulation of the transient actuator response opposed to the transient electro-magnetic finite element model which......The efficiency of digital hydraulic machines is strongly dependent on the valve switching time. Recently, fast switching have been achieved by using a direct electromagnetic moving coil actuator as the force producing element in fast switching hydraulic valves suitable for digital hydraulic...... machines. Mathematical models of the valve switching, targeted for design optimisation of the moving coil actuator, are developed. A detailed analytical model is derived and presented and its accuracy is evaluated against transient electromagnetic finite element simulations. The model includes...
Parallel Computing Characteristics of Two-Phase Thermal-Hydraulics code, CUPID
International Nuclear Information System (INIS)
Lee, Jae Ryong; Yoon, Han Young
2013-01-01
Parallelized CUPID code has proved to be able to reproduce multi-dimensional thermal hydraulic analysis by validating with various conceptual problems and experimental data. In this paper, the characteristics of the parallelized CUPID code were investigated. Both single- and two phase simulation are taken into account. Since the scalability of a parallel simulation is known to be better for fine mesh system, two types of mesh system are considered. In addition, the dependency of the preconditioner for matrix solver was also compared. The scalability for the single-phase flow is better than that for two-phase flow due to the less numbers of iterations for solving pressure matrix. The CUPID code was investigated the parallel performance in terms of scalability. The CUPID code was parallelized with domain decomposition method. The MPI library was adopted to communicate the information at the interface cells. As increasing the number of mesh, the scalability is improved. For a given mesh, single-phase flow simulation with diagonal preconditioner shows the best speedup. However, for the two-phase flow simulation, the ILU preconditioner is recommended since it reduces the overall simulation time
Numerical flow analyses of a two-phase hydraulic coupling
Energy Technology Data Exchange (ETDEWEB)
Hur, N.; Kwak, M.; Moshfeghi, M. [Sogang University, Seoul (Korea, Republic of); Chang, C.-S.; Kang, N.-W. [VS Engineering, Seoul (Korea, Republic of)
2017-05-15
We investigated flow characteristics in a hydraulic coupling at different charged water conditions and speed ratios. Hence, simulations were performed for three-dimensional two-phase flow by using the VOF method. The realizable k-ε turbulence model was adopted. To resolve the interaction of passing blades of the primary and secondary wheels, simulations were conducted in the unsteady framework using a sliding grid technique. The results show that the water-air distribution inside the wheel is strongly dependent upon both amount of charged water and speed ratio. Generally, air is accumulated in the center of the wheel, forming a toroidal shape wrapped by the circulating water. The results also show that at high speed ratios, the solid-body-like rotation causes dry areas on the periphery of the wheels and, hence, considerably decreases the circulating flow rate and the transmitted torque. Furthermore, the momentum transfer was investigated through the concept of a mass flux triangle based on the local velocity multiplied by the local mixture density instead of the velocity triangle commonly used in a single-phase turbomachine analysis. Also, the mass fluxes along the radius of the coupling in the partially charged and fully charged cases were found to be completely different. It is shown that the flow rate at the interfacial plane and also the transmitted torque are closely related and are strongly dependent upon both the amount of charged water and speed ratio. Finally, a conceptual categorization together with two comprehensive maps was provided for the torque transmission and also circulating flow rates. These two maps in turn exhibit valuable engineering information and can serve as bases for an optimal design of a hydraulic coupling.
Energy Technology Data Exchange (ETDEWEB)
Hartmann, Christoph Oliver
2016-06-13
Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS
Simulation model for the dynamic behavior of the hydraUlic circuito of PWR reactors
International Nuclear Information System (INIS)
Hirdes, V.R.T.R.
1987-01-01
The present work consist of the development of a computer code for the simulations of hydraulic transients caused by stoppages of the primary coolant pumps of nuclear reactors and it applied to the hydraulic circuits typical of PWR reactor. The code calculates the time-histories of the mass flux, rotation speed, electric and hydraulic torque and dynamic head of the pumps. It can be used for any combination of active and inactive pumps. Several transients were analysed and the results were compared with comparared with data from the Angra-I nuclear power plant. The results were considered satisfactory. (author) [pt
Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system
International Nuclear Information System (INIS)
Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong
2004-01-01
As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis
Modeling of two-phase flow instabilities during startup transients utilizing RAMONA-4B methodology
International Nuclear Information System (INIS)
Paniagua, J.; Rohatgi, U.S.; Prasad, V.
1996-01-01
RAMONA-4B code is currently under development for simulating thermal hydraulic instabilities that can occur in Boiling Water Reactors (BWRs) and the Simplified Boiling Water Reactor (SBWR). As one of the missions of RAMONA-4B is to simulate SBWR startup transients, where geysering or condensation-induced instability may be encountered, the code needs to be assessed for this application. This paper outlines the results of the assessments of the current version of RAMONA-4B and the modifications necessary for simulating the geysering or condensation-induced instability. The test selected for assessment are the geysering tests performed by Prof Aritomi (1993)
International Nuclear Information System (INIS)
Garner, P.L.; Blomquist, R.N.; Gelbard, E.M.
1992-09-01
The COMMIX-1AR/P computer program is designed for analyzing the steady-state and transient aspects of single-phase fluid flow and heat transfer in three spatial dimensions. This version is an extension of the modeling in COMMIX-1A to include multiple fluids in physically separate regions of the computational domain, modeling descriptions for pumps, radiation heat transfer between surfaces of the solids which are embedded in or surround the fluid, a k-var-epsilon model for fluid turbulence, and improved numerical techniques. The porous-medium formulation in COMMIX allows the program to be applied to a wide range of problems involving both simple and complex geometrical arrangements. The input preparation and execution procedures are presented for the COMMIX-1AR/P program and several postprocessor programs which produce graphical displays of the calculated results
Energy Technology Data Exchange (ETDEWEB)
Garner, P.L.; Blomquist, R.N.; Gelbard, E.M.
1992-09-01
The COMMIX-1AR/P computer program is designed for analyzing the steady-state and transient aspects of single-phase fluid flow and heat transfer in three spatial dimensions. This version is an extension of the modeling in COMMIX-1A to include multiple fluids in physically separate regions of the computational domain, modeling descriptions for pumps, radiation heat transfer between surfaces of the solids which are embedded in or surround the fluid, a k-[var epsilon] model for fluid turbulence, and improved numerical techniques. The porous-medium formulation in COMMIX allows the program to be applied to a wide range of problems involving both simple and complex geometrical arrangements. The input preparation and execution procedures are presented for the COMMIX-1AR/P program and several postprocessor programs which produce graphical displays of the calculated results.
Transient CFD simulation of a Francis turbine startup
International Nuclear Information System (INIS)
Nicolle, J; Morissette, J F; Giroux, A M
2012-01-01
To assess the life expectancy of hydraulic turbines, it is essential to obtain the loading on the blades, especially during transient operations known to be the most damaging. This paper presents a simplified CFD setup to model the startup phase of a Francis turbine while it goes from rest to speed no-load condition. The fluid domain included one distributor sector coupled with one runner passage. The guide vane motion and change in the angular velocity were included in a commercial code with user functions. Comparisons between numerical results and measurements acquired on a full-size turbine showed that most of the flow physics occurring during startup were captured.
An animal experimental study of transient synovitis of hip using three phase bone imaging
International Nuclear Information System (INIS)
Liang Jiugen; Lu Bing; Lu Xiaohu; Liu Shangli
1994-01-01
A model of transient synovitis was established by means of injecting noradrenaline (NA) into the joint cavity of young dogs. Radionuclide three phase bone imaging was then used to observe the local blood supply of femoral head and histological examination was used to understand the natural course of the disease process. The result showed that there were transient synovitis of the hip and decrease of blood supply in the affected femoral head after NA injection, but the changes gradually returned to normal after 4 weeks. No evidence of femoral head necrosis had been noticed. It is suggested that serial quantitative analysis of three phase bone imaging may have good clinical value in the early diagnosis transient hip synovitis, as well as in the assessment of the stage of the disease etc
Development of numerical simulation technology for high resolution thermal hydraulic analysis
International Nuclear Information System (INIS)
Yoon, Han Young; Kim, K. D.; Kim, B. J.; Kim, J. T.; Park, I. K.; Bae, S. W.; Song, C. H.; Lee, S. W.; Lee, S. J.; Lee, J. R.; Chung, S. K.; Chung, B. D.; Cho, H. K.; Choi, S. K.; Ha, K. S.; Hwang, M. K.; Yun, B. J.; Jeong, J. J.; Sul, A. S.; Lee, H. D.; Kim, J. W.
2012-04-01
A realistic simulation of two phase flows is essential for the advanced design and safe operation of a nuclear reactor system. The need for a multi dimensional analysis of thermal hydraulics in nuclear reactor components is further increasing with advanced design features, such as a direct vessel injection system, a gravity driven safety injection system, and a passive secondary cooling system. These features require more detailed analysis with enhanced accuracy. In this regard, KAERI has developed a three dimensional thermal hydraulics code, CUPID, for the analysis of transient, multi dimensional, two phase flows in nuclear reactor components. The code was designed for use as a component scale code, and/or a three dimensional component, which can be coupled with a system code. This report presents an overview of the CUPID code development and preliminary assessment, mainly focusing on the numerical solution method and its verification and validation. It was shown that the CUPID code was successfully verified. The results of the validation calculations show that the CUPID code is very promising, but a systematic approach for the validation and improvement of the physical models is still needed
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...
Energy Technology Data Exchange (ETDEWEB)
NONE
2004-07-01
More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.
International Nuclear Information System (INIS)
2004-01-01
More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling
Energy Technology Data Exchange (ETDEWEB)
NONE
2004-07-01
More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.
International Nuclear Information System (INIS)
2004-01-01
More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling
COBRA-WC: a version of COBRA for single-phase multiassembly thermal hydraulic transient analysis
International Nuclear Information System (INIS)
George, T.L.; Basehore, K.L.; Wheeler, C.L.; Prather, W.A.; Masterson, R.E.
1980-07-01
The objective of this report is to provide the user of the COBRA-WC (Whole Core) code a basic understanding of the code operation and capabilities. Included in this manual are the equations solved and the assumptions made in their derivations, a general description of the code capabilities, an explanation of the numerical algorithms used to solve the equations, and input instructions for using the code. Also, the auxiliary programs GEOM and SPECSET are described and input instructions for each are given. Input for COBRA-WC sample problems and the corresponding output are given in the appendices. The COBRA-WC code has been developed from the COBRA-IV-I code to analyze liquid Metal Fast Breeder Reactor (LMFBR) assembly transients. It was specifically developed to analyze a core flow coastdown to natural circulation cooling
The detection of transient directional couplings based on phase synchronization
International Nuclear Information System (INIS)
Wagner, T; Fell, J; Lehnertz, K
2010-01-01
We extend recent approaches based on the concept of phase synchronization to enable the time-resolved investigation of directional relationships between coupled dynamical systems from short and transient noisy time series. For our approach, we consider an observed ensemble of a sufficiently large number of time series as multiple realizations of a process. We derive an index that quantifies the direction of transient interactions and assess its statistical significance using surrogate techniques. Analysing time series from noisy and chaotic systems, we demonstrate numerically the applicability and limitations of our approach. Our findings from an exemplary application to event-related brain activities underline the importance of our method for improving knowledge about the mechanisms underlying memory formation in humans.
The detection of transient directional couplings based on phase synchronization
Energy Technology Data Exchange (ETDEWEB)
Wagner, T; Fell, J; Lehnertz, K, E-mail: twagner@uni-bonn.d [Department of Epileptology, University of Bonn, Sigmund-Freud-Strasse 25, 53127 Bonn (Germany)
2010-05-15
We extend recent approaches based on the concept of phase synchronization to enable the time-resolved investigation of directional relationships between coupled dynamical systems from short and transient noisy time series. For our approach, we consider an observed ensemble of a sufficiently large number of time series as multiple realizations of a process. We derive an index that quantifies the direction of transient interactions and assess its statistical significance using surrogate techniques. Analysing time series from noisy and chaotic systems, we demonstrate numerically the applicability and limitations of our approach. Our findings from an exemplary application to event-related brain activities underline the importance of our method for improving knowledge about the mechanisms underlying memory formation in humans.
DYNREL - the reference calculation (coupled code utilization on analysis of RIA-transient)
International Nuclear Information System (INIS)
Strmensky, C.; Darilek, P.
2003-01-01
DYNREL is coupled code, comprising DYN3D and RELAP5 programs. The coupled code has been developed during four years. Now DYNREL is tested on selected RIA and thermo-hydraulic transient calculations. This material describes some results from selected RIA transient calculation (initiated by control rod movement). DYNREL modelled the whole nuclear reactors. The core is modeled as 313 or 349 independent thermo-hydraulic channels with 10 or 20 axial layers. Thermo-hydraulic part contains about 700 components that covered the six loops' model of nuclear power plant in detail. The calculated results are compared with DYN3D/M3, DYN3D/H1.1 results (Authors)
Energy Technology Data Exchange (ETDEWEB)
Imanbaew, D.; Nosenko, Y. [Fachbereich Chemie, Technische Universität Kaiserslautern, Erwin-Schrödinger-Str. 52–54, 67663 Kaiserslautern (Germany); Forschungszentrum OPTIMAS, Erwin-Schrödinger-Str. 46, 67663 Kaiserslautern (Germany); Kerner, C. [Fachbereich Chemie, Technische Universität Kaiserslautern, Erwin-Schrödinger-Str. 52–54, 67663 Kaiserslautern (Germany); Chevalier, K.; Rupp, F. [Fachbereich Physik, Technische Universität Kaiserslautern, Erwin-Schrödinger-Str. 46, 67663 Kaiserslautern (Germany); Riehn, C., E-mail: riehn@chemie.uni-kl.de [Fachbereich Chemie, Technische Universität Kaiserslautern, Erwin-Schrödinger-Str. 52–54, 67663 Kaiserslautern (Germany); Forschungszentrum OPTIMAS, Erwin-Schrödinger-Str. 46, 67663 Kaiserslautern (Germany); Thiel, W.R. [Fachbereich Chemie, Technische Universität Kaiserslautern, Erwin-Schrödinger-Str. 52–54, 67663 Kaiserslautern (Germany); Diller, R. [Fachbereich Physik, Technische Universität Kaiserslautern, Erwin-Schrödinger-Str. 46, 67663 Kaiserslautern (Germany)
2014-10-17
Graphical abstract: - Highlights: • Ultrafast dynamics of new Ru(II) catalysts investigated in gas phase and solution. • Catalyst activation (HCl loss) achieved in ion trap by UV photoexcitation. • Electronic relaxation proceeds by IVR and IC followed by ground state dissociation. • No triplet formation in contrast to other Ru-polypyridine complexes. • Solvent prohibits catalyst activation in solution by fast vibrational cooling. - Abstract: We report studies on the excited state dynamics of new ruthenium(II) complexes [(η{sup 6}-cymene)RuCl(apypm)]PF{sub 6} (apypm=2-NR{sub 2}-4-(pyridine-2-yl)-pyrimidine, R=CH{sub 3} (1)/H (2)) which, in their active form [1{sup +}-HCl] and [2{sup +}-HCl], catalyze the transfer hydrogenation of arylalkyl ketones in the absence of a base. The investigations encompass femtosecond pump–probe transient mass spectrometry under isolated conditions and transient absorption spectroscopy in acetonitrile solution, both on the cations [(η{sup 6}-cymene)RuCl(apypm)]{sup +} (1{sup +}, 2{sup +}). Gas phase studies on mass selected ions were performed in an ESI ion trap mass spectrometer by transient photofragmentation, unambiguously proving the formation of the activated catalyst species [1{sup +}-HCl] or [2{sup +}-HCl] after photoexcitation being the only fragmentation channel. The primary excited state dynamics in the gas phase could be fitted to a biexponential decay, yielding time constants of <100 fs and 1–3 ps. Transient absorption spectroscopy performed in acetonitrile solution using femtosecond UV/Vis and IR probe laser pulses revealed additional deactivation processes on longer time scales (∼7–12 ps). However, the formation of the active catalyst species after photoexcitation could not be observed in solution. The results from both studies are compared to former CID investigations and DFT calculations concerning the activation mechanism.
Simulation of Thermal-hydraulic Process in Reactor of HTR-PM
International Nuclear Information System (INIS)
Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle
2014-01-01
This paper provides the physical process in the reactor of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) and introduces the standard operation conditions. The FORTRAN code developed for the thermal hydraulic module of Full-Scale Simulator (FSS) of HTR-PM is used to simulate two typical operation transients including cold startup process and cold shutdown process. And the results were compared to the safety analysis code, namely TINTE. The good agreement indicates that the code is applicable for simulating the thermal-hydraulic process in reactor of HTR-PM. And for long time transient process, the code shows good stability and convergence. (author)
Villar, V. Ashley; Berger, Edo; Metzger, Brian D.; Guillochon, James
2017-11-01
The duration-luminosity phase space (DLPS) of optical transients is used, mostly heuristically, to compare various classes of transient events, to explore the origin of new transients, and to influence optical survey observing strategies. For example, several observational searches have been guided by intriguing voids and gaps in this phase space. However, we should ask, do we expect to find transients in these voids given our understanding of the various heating sources operating in astrophysical transients? In this work, we explore a broad range of theoretical models and empirical relations to generate optical light curves and to populate the DLPS. We explore transients powered by adiabatic expansion, radioactive decay, magnetar spin-down, and circumstellar interaction. For each heating source, we provide a concise summary of the basic physical processes, a physically motivated choice of model parameter ranges, an overall summary of the resulting light curves and their occupied range in the DLPS, and how the various model input parameters affect the light curves. We specifically explore the key voids discussed in the literature: the intermediate-luminosity gap between classical novae and supernovae, and short-duration transients (≲ 10 days). We find that few physical models lead to transients that occupy these voids. Moreover, we find that only relativistic expansion can produce fast and luminous transients, while for all other heating sources events with durations ≲ 10 days are dim ({M}{{R}}≳ -15 mag). Finally, we explore the detection potential of optical surveys (e.g., Large Synoptic Survey Telescope) in the DLPS and quantify the notion that short-duration and dim transients are exponentially more difficult to discover in untargeted surveys.
International Nuclear Information System (INIS)
2001-06-01
This report deals with an internationally agreed experimental test facility matrix for the validation of best estimate thermal-hydraulic computer codes applied for the analysis of VVER reactor primary systems in accident and transient conditions. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities that supplement the CSNI CCVMs and are suitable for reproducing these aspects are selected. Secondly, a list of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. The construction of VVER Thermal-Hydraulic Code Validation Matrix follows the logic of the CSNI Code Validation Matrices (CCVM). Similar to the CCVM it is an attempt to collect together in a systematic way the best sets of available test data for VVER specific code validation, assessment and improvement, including quantitative assessment of uncertainties in the modelling of phenomena by the codes. In addition to this objective, it is an attempt to record information which has been generated in countries operating VVER reactors over the last 20 years so that it is more accessible to present and future workers in that field than would otherwise be the case. (authors)
Advanced modelling and numerical strategies in nuclear thermal-hydraulics
International Nuclear Information System (INIS)
Staedtke, H.
2001-01-01
The first part of the lecture gives a brief review of the current status of nuclear thermal hydraulics as it forms the basis of established system codes like TRAC, RELAP5, CATHARE or ATHLET. Specific emphasis is given to the capabilities and limitations of the underlying physical modelling and numerical solution strategies with regard to the description of complex transient two-phase flow and heat transfer conditions as expected to occur in PWR reactors during off-normal and accident conditions. The second part of the lecture focuses on new challenges and future needs in nuclear thermal-hydraulics which might arise with regard to re-licensing of old plants using bestestimate methodologies or the design and safety analysis of Advanced Light Water Reactors relying largely on passive safety systems. In order to meet these new requirements various advanced modelling and numerical techniques will be discussed including extended wellposed (hyperbolic) two-fluid models, explicit modelling of interfacial area transport or higher order numerical schemes allowing a high resolution of local multi-dimensional flow processes.(author)
PWR boron dilution transients. Thermal-hydraulic analyses of PKL-E experiments
International Nuclear Information System (INIS)
Pietro Alessandro Di Maio; Antonino Tomasello; Giuseppe Vella
2005-01-01
Full text of publication follows: In the field of PWR reactor safety, one of the topics that is currently of major interest worldwide is that of inadvertent boron dilution events. The safety issue involved in such scenarios is that inadvertent transport into the reactor core of un-borated water - or water having only a low boron concentration - can lead to local recriticality and possibly to power excursions. Studies on various accidental sequences that could initiate boron dilution events revealed that some SBLOCAs, occurring in the primary system, lead to reflux condenser conditions and subsequent re-establishment of natural circulation are of particular significance. In this work, the first field of analysis is related to the investigation of the thermal - hydraulic conditions that could lead to boron dilution events, such as the stop of natural circulation within primary system and the subsequent start of reflux condenser functioning mode. The investigation of the primary thermal - hydraulic conditions has been performed using the experimental results obtained in the PKL test integral facility in which some SBLOCA sequences have been carried out. Particular useful were the PKL III E experiments data whose results have been numerically reproduced using the code Relap5/MOD3.3/Beta code, contributing to understand the complex thermalhydraulic phenomena related to a PWR boron dilution event. The second field of analysis is related to the effects that possible displacements of un-borated water slugs towards the Reactor Pressure Vessel (RPV) could have on the core reactivity. A numerical approach using the Relap5 reactor kinetics model has been adopted to integrate the experimental thermal - hydraulic data obtained in the PKL III E tests. A careful analysis has been performed in order to establish which core conditions at incident start could produce the largest reactivity increase as a consequence of restarting of natural circulation during the primary system
PWR boron dilution transients. Thermal-hydraulic analyses of PKL-E experiments
Energy Technology Data Exchange (ETDEWEB)
Pietro Alessandro Di Maio; Antonino Tomasello; Giuseppe Vella [Dipartimento di Ingegneria Nucleare, Viale delle Scienze, 90128 Palermo (Italy)
2005-07-01
Full text of publication follows: In the field of PWR reactor safety, one of the topics that is currently of major interest worldwide is that of inadvertent boron dilution events. The safety issue involved in such scenarios is that inadvertent transport into the reactor core of un-borated water - or water having only a low boron concentration - can lead to local recriticality and possibly to power excursions. Studies on various accidental sequences that could initiate boron dilution events revealed that some SBLOCAs, occurring in the primary system, lead to reflux condenser conditions and subsequent re-establishment of natural circulation are of particular significance. In this work, the first field of analysis is related to the investigation of the thermal - hydraulic conditions that could lead to boron dilution events, such as the stop of natural circulation within primary system and the subsequent start of reflux condenser functioning mode. The investigation of the primary thermal - hydraulic conditions has been performed using the experimental results obtained in the PKL test integral facility in which some SBLOCA sequences have been carried out. Particular useful were the PKL III E experiments data whose results have been numerically reproduced using the code Relap5/MOD3.3/Beta code, contributing to understand the complex thermalhydraulic phenomena related to a PWR boron dilution event. The second field of analysis is related to the effects that possible displacements of un-borated water slugs towards the Reactor Pressure Vessel (RPV) could have on the core reactivity. A numerical approach using the Relap5 reactor kinetics model has been adopted to integrate the experimental thermal - hydraulic data obtained in the PKL III E tests. A careful analysis has been performed in order to establish which core conditions at incident start could produce the largest reactivity increase as a consequence of restarting of natural circulation during the primary system
Thermal-hydraulic analysis for wire-wrapped PWR cores
Energy Technology Data Exchange (ETDEWEB)
Diller, P. [General Electric Company, 3901 Castle Hayne Rd., Wilmington, NC 28401 (United States)], E-mail: pdiller@gmail.com; Todreas, N. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)], E-mail: todreas@mit.edu; Hejzlar, P. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)
2009-08-15
This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH{sub 1.6} (referred to as U-ZrH{sub 1.6}) or UO{sub 2} fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA). The thermal-hydraulic performance of U-ZrH{sub 1.6} and UO{sub 2} were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.
Design of a welltest for determining two-phase hydraulic properties
International Nuclear Information System (INIS)
Finsterle, S.
1995-01-01
This report describes the design of a well test to determine two-phase hydraulic properties of a low permeability, low porosity formation. Estimation of gas-related parameters in such formations is difficult using standard pumping tests mainly because of the strong fluctuations in the pressure and flow rate data which are a consequence of gas bubbles evolving in the test interval. Even more important is the fact that the data do not allow distinguishing among alternative conceptual models. The estimated parameters are therefore uncertain, highly correlated, and ambiguous. In this study we examine a test sequence that could be appended to a standard hydraulic testing program. It is shown that performing a series of water and gas injection tests significantly reduces parameter correlations, thus decreasing the estimation error. Moreover, the extended test sequence makes possible the identification of the model that describes relative permeabilities and capillary pressures. This requires, however, that data of high accuracy are collected under controlled test conditions. The purpose of this report is to describe the modeling approach, assumptions and limitations of the procedure, and to provide practical recommendations for future testing
Transitioning from interpretive to predictive in thermal hydraulic codes
International Nuclear Information System (INIS)
Mousseau, V.A.
2004-01-01
simulation tools. These different computer codes have then been loosely coupled to represent the transient. It is important to note that the accuracy of the transient is determined by the largest error in the system. For example, if neutron diffusion and thermal conduction are each solved in a second order in time accurate manner, but they are only exchange information every five time steps, then the system is first order in time since the coupling is first order in time. Therefore, it is important to ensure that the level of accuracy used to couple two different computer codes is as accurate as the two codes being coupled. Focusing in on the reactor cooling system (the two phase flow and the heat conduction) it is important to solve the coupling between phases and the wall as accurately as possible. In RELAP these two pieces of nonlinearly coupled physics are solved in separate linear systems. The RELAP solution procedure is to take a nonlinear system of equations, split them into two separate pieces, linearize and solve the fluid flow and heat conduction separately, then couple them together in an explicit fashion. It should be noted that different versions of RELAP allow for linearly implicit coupling of the fluid flow and wall heat transfer under special conditions. This manuscript will examine the reactor cooling system and present an analysis of the error caused by linearizing and splitting nonlinearly coupled physics. Modern computers and numerical methods allow for this system of nonlinear equations to be solved without linearizing and splitting. This coupled approach is referred to as an implicitly balanced solution method. Because of the fact that future thermal hydraulic codes will be required to move from the low accuracy requirements of data interpretation to the high accuracy requirements of data prediction, the accuracy of current thermal hydraulic codes need to be increased. This manuscript provides some of the first steps required to analyze the temporal
Dynamic modelling for two-phase flow systems
International Nuclear Information System (INIS)
Guerra, M.A.
1991-06-01
Several models for two-phase flow have been studied, developing a thermal-hydraulic analysis code with one of these models. The program calculates, for one-dimensional cases with variable flow area, the transient behaviour of system process variables, when the boundary conditions (heat flux, flow rate, enthalpy and pressure) are functions of time. The modular structure of the code, eases the program growth. In fact, the present work is the basis for a general purpose accident and transient analysis code in nuclear reactors. Code verification has been made against RETRAN-02 results. Satisfactory results have been achieved with the present version of the code. (Author) [es
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
Transient behaviour of main coolant pump in nuclear power plants
International Nuclear Information System (INIS)
Delja, A.
1986-01-01
A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)
Wang, Hang; Felder, Stefan; Chanson, Hubert
2014-07-01
Intense turbulence develops in the two-phase flow region of hydraulic jump, with a broad range of turbulent length and time scales. Detailed air-water flow measurements using intrusive phase-detection probes enabled turbulence characterisation of the bubbly flow, although the phenomenon is not a truly random process because of the existence of low-frequency, pseudo-periodic fluctuating motion in the jump roller. This paper presents new measurements of turbulent properties in hydraulic jumps, including turbulence intensity, longitudinal and transverse integral length and time scales. The results characterised very high turbulent levels and reflected a combination of both fast and slow turbulent components. The respective contributions of the fast and slow motions were quantified using a triple decomposition technique. The decomposition of air-water detection signal revealed "true" turbulent characteristics linked with the fast, microscopic velocity turbulence of hydraulic jumps. The high-frequency turbulence intensities were between 0.5 and 1.5 close to the jump toe, and maximum integral turbulent length scales were found next to the bottom. Both decreased in the flow direction with longitudinal turbulence dissipation. The results highlighted the considerable influence of hydrodynamic instabilities of the flow on the turbulence characterisation. The successful application of triple decomposition technique provided the means for the true turbulence properties of hydraulic jumps.
Thermal-hydraulic analyses for in-pile SCWR fuel qualification test loops and SCWR material loop
Energy Technology Data Exchange (ETDEWEB)
Vojacek, A.; Mazzini, G.; Zmitkova, J.; Ruzickova, M. [Research Centre Rez (Czech Republic)
2014-07-01
One of the R&D directions of Research Centre Rez is dedicated to the supercritical water-cooled reactor concept (SCWR). Among the developed experimental facilities and infrastructure in the framework of the SUSEN project (SUStainable ENergy) is construction and experimental operation of the supercritical water loop SCWL focusing on material tests. At the first phase, this SCWL loop is assembled and operated out-of-pile in the dedicated loop facilities hall. At this out-of-pile operation various operational conditions are tested and verified. After that, in the second phase, the SCWL loop will be situated in-pile, in the core of the research reactor LVR-15, operated at CVR. Furthermore, it is planned to carry out a test of a small scale fuel assembly within the SuperCritical Water Reactor Fuel Qualification Test (SCWR-FQT) loop, which is now being designed. This paper presents the results of the thermal-hydraulic analyses of SCWL loop out-of-pile operation using the RELAP5/MOD3.3. The thermal-hydraulic modeling and the performed analyses are focused on the SCWL loop model validation through a comparison of the calculation results with the experimental results obtained at various operation conditions. Further, the present paper focuses on the transient analyses for start-up and shut-down of the FQT loop, particularly to explore the ability of system codes ATHLET 3.0A to simulate the transient between subcritical conditions and supercritical conditions. (author)
Energy Technology Data Exchange (ETDEWEB)
Huy, Ngo Quang; Khang, Ngo Phu; An, Tran Khac; Nghiem, Huynh Ton [Nuclear Research Inst., Da Lat (Viet Nam)
1994-10-01
Some experimental and theoretical thermal hydraulic characteristics of the Dalat Nuclear Research Reactor are presented, together with some general assessments, from a thermal hydraulic point of view, of its safety under transient conditions. (author). 3 refs., 9 figs., 7 tabs.
RETRAN nonequilibrium two-phase flow model for operational transient analyses
International Nuclear Information System (INIS)
Paulsen, M.P.; Hughes, E.D.
1982-01-01
The field balance equations, flow-field models, and equation of state for a nonequilibrium two-phase flow model for RETRAN are given. The differential field balance model equations are: (1) conservation of mixture mass; (2) conservation of vapor mass; (3) balance of mixture momentum; (4) a dynamic-slip model for the velocity difference; and (5) conservation of mixture energy. The equation of state is formulated such that the liquid phase may be subcooled, saturated, or superheated. The vapor phase is constrained to be at the saturation state. The dynamic-slip model includes wall-to-phase and interphase momentum exchanges. A mechanistic vapor generation model is used to describe vapor production under bulk subcooling conditions. The speed of sound for the mixture under nonequilibrium conditions is obtained from the equation of state formulation. The steady-state and transient solution methods are described
Efficiency limit factor analysis for the Francis-99 hydraulic turbine
Zeng, Y.; Zhang, L. X.; Guo, J. P.; Guo, Y. K.; Pan, Q. L.; Qian, J.
2017-01-01
The energy loss in hydraulic turbine is the most direct factor that affects the efficiency of the hydraulic turbine. Based on the analysis theory of inner energy loss of hydraulic turbine, combining the measurement data of the Francis-99, this paper calculates characteristic parameters of inner energy loss of the hydraulic turbine, and establishes the calculation model of the hydraulic turbine power. Taken the start-up test conditions given by Francis-99 as case, characteristics of the inner energy of the hydraulic turbine in transient and transformation law are researched. Further, analyzing mechanical friction in hydraulic turbine, we think that main ingredients of mechanical friction loss is the rotation friction loss between rotating runner and water body, and defined as the inner mechanical friction loss. The calculation method of the inner mechanical friction loss is given roughly. Our purpose is that explore and research the method and way increasing transformation efficiency of water flow by means of analysis energy losses in hydraulic turbine.
Thermal-hydraulic tests with out-of-pile test facility for BOCA development
International Nuclear Information System (INIS)
Kitagishi, Shigeru; Aoyama, Masashi; Tobita, Masahiro; Inaba, Yoshitomo; Yamaura, Takayuki
2012-01-01
The fuel transient test facility was prepared for power ramping tests of light-water-reactor (LWR) fuels in the Japan Materials Testing Reactor (JMTR) under a contract project with the Nuclear Industrial Safety Agent (NISA) of the Ministry of Economy, Trade and Industry (METI). It is necessary to develop high accuracy analysis procedure for power ramping tests after restart of the JMTR. The out-of-pile test facility to simulate thermal-hydraulic conditions of the fuel transient test facility was therefore developed. Applicability of the analysis code ACE-3D was examined for thermal-hydraulic analysis of power ramping tests for 10x10 BWR fuels by the fuel transient test facility. As the results, the calculated temperature was 304°C in comparison with measured value of 304.9-317.4°C in the condition of 600 W/cm. There is a bright prospect of high accuracy power ramping tests by the fuel transient test facility in JMTR. (author)
Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results
International Nuclear Information System (INIS)
Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A.; Beelman, R.J.
1987-11-01
The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs
Transient analysis of air-water two-phase flow in channels and bends
International Nuclear Information System (INIS)
Khan, H.J.; Ye, W.; Pertmer, G.A.
1992-01-01
The algorithm used in this paper is the Newton Block Gauss Seidel method, which has been applied to both simple and complex flow conditions in two-phase flow. This paper contains a description of difference techniques and an iterative solution algorithm that is used to solve the field and constitutive equations of the two-fluid model. In practice, this solution procedure has been proven to be stable and capable of generating solutions in problems where other schemes have failed. The method converges rapidly for reasonable error tolerances and is easily extended to three-dimensional geometries. Using air-water as the two-phase medium, transient flow behavior in several geometries of interest are shown. Flow through a vertical channel with flow obstruction, large U bends, and 90-deg bends are being demonstrated with variation of inlet void fraction and slip ratio. Significant changes in the velocity and void distribution profiles have been observed. Various regions of flow recirculation are obtained in the flow domain for each phase. The phasic velocity and void distributions are dominated by gravity-induced phase separation causing air to accumulate in the upper region. The influence of inlet slip ratio and interfacial momentum transfer on the transient flow profile has been demonstrated in detail
Phase control of the transient resonance of the automatic ball balancer
Michalczyk, Jerzy; Pakuła, Sebastian
2016-05-01
Hazards related to undesired increases of vibration amplitudes in transient resonance of vibroinsulated rotor systems with automatic ball balancer (ABB) are discussed in the paper. The application of the phase control method with taking into account the limited drive power is proposed for these amplitudes reduction. The high efficiency of this approach is indicated.
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.
Single-phase sodium pump model for LMFBR thermal-hydraulic analysis
International Nuclear Information System (INIS)
Madni, I.K.; Cazzoli, E.G.; Agrawal, A.K.
1979-01-01
A single-phase, homologous pump model has been developed for simulation of safety-related transients in LMFBR systems. Pump characteristics are modeled by homologous head and torque relations encompassing all regimes of operation. These relations were derived from independent model test results with a centrifugal pump of specific speed equal to 35 (SI units) or 1800 (gpm units), and are used to analyze the steady-state and transient behavior of sodium pumps in a number of LMFBR plants. Characteristic coefficients for the polynomials in all operational regimes are provided in a tabular form. The speed and flow dependence of head is included through solutions of the impeller and coolant dynamic equations. Results show the model to yield excellent agreement with experimental data in sodium for the FFTF prototype pump, and with vendor calculations for the CRBR pump. A sample pipe rupture calculation is also performed to demonstrate the necessity for modeling the complete pump characteristics
Measurements of flow-rate transients in one-phase liquid flow
International Nuclear Information System (INIS)
Mueller-Roos, J.
1975-01-01
A report is given on a method to determine flow-rate transients in a one-phase flow. Periodic temperature signals are superposed on the flow, from which flow times are calculated through correlation each over a half period. The evaluation is carried out according to the digitalization 'off-line' on a large computer. Rate peaks of over 100% within 1.9 s were qualitatively and quantitatively well represented. (orig./LH) [de
Transient Structures and Possible Limits of Data Recording in Phase-Change Materials.
Hu, Jianbo; Vanacore, Giovanni M; Yang, Zhe; Miao, Xiangshui; Zewail, Ahmed H
2015-07-28
Phase-change materials (PCMs) represent the leading candidates for universal data storage devices, which exploit the large difference in the physical properties of their transitional lattice structures. On a nanoscale, it is fundamental to determine their performance, which is ultimately controlled by the speed limit of transformation among the different structures involved. Here, we report observation with atomic-scale resolution of transient structures of nanofilms of crystalline germanium telluride, a prototypical PCM, using ultrafast electron crystallography. A nonthermal transformation from the initial rhombohedral phase to the cubic structure was found to occur in 12 ps. On a much longer time scale, hundreds of picoseconds, equilibrium heating of the nanofilm is reached, driving the system toward amorphization, provided that high excitation energy is invoked. These results elucidate the elementary steps defining the structural pathway in the transformation of crystalline-to-amorphous phase transitions and describe the essential atomic motions involved when driven by an ultrafast excitation. The establishment of the time scales of the different transient structures, as reported here, permits determination of the possible limit of performance, which is crucial for high-speed recording applications of PCMs.
Energy Technology Data Exchange (ETDEWEB)
Bianchi, Paulo Henrique
2008-07-01
This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)
Hydraulic testing in crystalline rock
International Nuclear Information System (INIS)
Almen, K.E.; Andersson, J.E.; Carlsson, L.; Hansson, K.; Larsson, N.A.
1986-12-01
Swedish Geolocical Company (SGAB) conducted and carried out single-hole hydraulic testing in borehole Fi 6 in the Finnsjoen area of central Sweden. The purpose was to make a comprehensive evaluation of different methods applicable in crystalline rocks and to recommend methods for use in current and scheduled investigations in a range of low hydraulic conductivity rocks. A total of eight different methods of testing were compared using the same equipment. This equipment was thoroughly tested as regards the elasticity of the packers and change in volume of the test section. The use of a hydraulically operated down-hole valve enabled all the tests to be conducted. Twelve different 3-m long sections were tested. The hydraulic conductivity calculated ranged from about 5x10 -14 m/s to 1x10 -6 m/s. The methods used were water injection under constant head and then at a constant rate-of-flow, each of which was followed by a pressure fall-off period. Water loss, pressure pulse, slug and drill stem tests were also performed. Interpretation was carried out using standard transient evaluation methods for flow in porous media. The methods used showed themselves to be best suited to specific conductivity ranges. Among the less time-consuming methods, water loss, slug and drill stem tests usually gave somewhat higher hydraulic conductivity values but still comparable to those obtained using the more time-consuming tests. These latter tests, however, provided supplementary information on hydraulic and physical properties and flow conditions, together with hydraulic conductivity values representing a larger volume of rock. (orig./HP)
The phenomenology of a small break LOCA in a complex thermal hydraulic loop
International Nuclear Information System (INIS)
Di Marzo, M.; Almenas, K.K.; Hsu, Y.Y.; Wang, Z.
1988-01-01
A phenomenological description of the thermal hydraulics events that take place during a simulated Small Break Loss of Coolant Accident (SB-LOCA) is presented. The SB-LOCA transient is described in detail and the various mass and energy transport modes are identified. Similar behavior is observed in other facilities designed for the simulation of this type of accidents. Previous investigations suggest a simple modelling of the phenomena based on fluid mechanic considerations. An extensive experimental program conducted at the experimental facility of the University of Maryland reveals that condensation is a dominant driving force for this type of transients. This finding has significant implications in the modelling of enthalpy transport for some of the flow modes which occur during the transient. In particular it affects the Interruption and Resumption Mode (IRM) during which enthalpy is transported by periodic flow of a two phase mixture. The efforts to predict the flow interruption based on fluid mechanic criteria of phase separation in the hot leg are shown to be misdirected since thermodynamic phenomena taking place in the horizontal portion of the cold legs and in the reactor vessel downcomer are mostly responsible for that transition. For flow resumption to occur the liquid-vapor mixture swelling in the vertical portion of the hot leg determines the occurrence of the liquid spill over the top of the candy cane. (orig.)
Development of three dimensional transient analysis code STTA for SCWR core
International Nuclear Information System (INIS)
Wang, Lianjie; Zhao, Wenbo; Chen, Bingde; Yao, Dong; Yang, Ping
2015-01-01
Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation
International standard problem ISP-47 on containment thermal hydraulics - Final report
International Nuclear Information System (INIS)
Allelein, H. J.; Schwarz, S.; Fischer, K.; Vendel, J.; Malet, J.; Bentaib, A.; Studer, E.; Paillere, H.; Houkema, M.
2007-01-01
The main objective of the ISP-47 is to assess the capabilities of Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes in the area of containment thermal-hydraulics. Following the recommendations made in the state-of-the-art report on 'Containment Thermal-hydraulics and Hydrogen Distribution' this ISP was based on the application of different complementary experimental facilities and a progressive modelling difficulty. The three experimental facilities TOSQAN, MISTRA and ThAI have shown the quality of the provided experimental data suitable for CFD and LP code benchmarking in steady-state and transient conditions (control of the initial and boundary conditions, and the accuracy of the measurement techniques). This mainly includes pressure transients and gas temperature field as in former exercises. Detailed gas velocity and gas concentration (air, steam and helium) fields were obtained for the first time in such an exercise. ISP-47 was executed in two main steps: Step 1 was dedicated to the validation of the codes in the separate effects facility TOSQAN (7 m 3 ). Wall condensation, steam injection in air or air / helium atmospheres, and buoyancy were addressed under well-controlled initial conditions in a simple geometry. Furthermore, the interactions of phenomena such as condensation / stratification, turbulence / buoyancy, etc. were addressed using the larger scale of MISTRA (100 m 3 ) facility. Both TOSQAN and MISTRA were specifically designed to produce data for CFD codes with state-of-the-art instrumentation. The TOSQAN benchmark was open, whereas the MISTRA benchmark was blind. Step 2 addressed the code assessment using an experiment in the multi-compartment ThAI (60 m 3 ) facility with different steam and helium injection phases, transient stratification and mixing conditions in the atmosphere, development of natural convection, wall condensate distribution, fog formation, and transient thermal response of heat conducting walls. Detailed
Two and Three-Phases Fractal Models Application in Soil Saturated Hydraulic Conductivity Estimation
Directory of Open Access Journals (Sweden)
ELNAZ Rezaei abajelu
2017-03-01
Full Text Available Introduction: Soil Hydraulic conductivity is considered as one of the most important hydraulic properties in water and solutionmovement in porous media. In recent years, variousmodels as pedo-transfer functions, fractal models and scaling technique are used to estimate the soil saturated hydraulic conductivity (Ks. Fractal models with two subset of two (solid and pore and three phases (solid, pore and soil fractal (PSF are used to estimate the fractal dimension of soil particles. The PSF represents a generalization of the solid and pore mass fractal models. The PSF characterizes both the solid and pore phases of the porous material. It also exhibits self-similarity to some degree, in the sense that where local structure seems to be similar to the whole structure.PSF models can estimate interface fractal dimension using soil pore size distribution data (PSD and soil moisture retention curve (SWRC. The main objective of this study was to evaluate different fractal models to estimate the Ksparameter. Materials and Methods: The Schaapetal data was used in this study. The complex consists of sixty soil samples. Soil texture, soil bulk density, soil saturated hydraulic conductivity and soil particle size distribution curve were measured by hydrometer method, undistributed soil sample, constant head method and wet sieve method, respectively for all soil samples.Soil water retention curve were determined by using pressure plates apparatus.The Ks parameter could be estimated by Ralws model as a function of fractal dimension by seven fractal models. Fractal models included Fuentes at al. (1996, Hunt and Gee (2002, Bird et al. (2000, Huang and Zhang (2005, Tyler and Wheatcraft (1990, Kutlu et al. (2008, Sepaskhah and Tafteh (2013.Therefore The Ks parameter can be estimated as a function of the DS (fractal dimension by seven fractal models (Table 2.Sensitivity analysis of Rawls model was assessed by making changes±10%, ±20% and±30%(in input parameters
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
Analysis of steady state and transient two-phase flows in downwardly inclined lines
International Nuclear Information System (INIS)
Crawford, T.J.
1983-01-01
A study of steady-state and transient two-phase flows in downwardly inclined lines is described. Steady-state flow patterns maps are presented using Freon-113 as the working fluid to provide new high density vapors. These flow maps with high density vapor serve to significantly extend the investigations of steady-state downward two-phase flow patterns. Physical models developed which successfully predicted the onset or location of various flow pattern transitions. A new simplified criterion that would be useful to designers and experimenters is offered for the onset of dispersed flow. A new empirical holdup correlation and a new bubble diameter/flow rate correlation are also proposed. Flow transients in vertical downward lines were studied to investigate the possible formation of intermediate or spurious flow patterns that would not be seen at steady-state conditions. Void fraction behavior during the transients was modeled by using the dynamic slip equation from the transient analysis code RETRAN. Physical models of interfacial area were developed and compared with models and data from literature. There was satisfactory agreement between the models of the present study and the literature models and data. The concentration parameter of the drift flux model was evaluated for vertical downward flow. These new values of the flow dependent parameter were different from those previously proposed in the literature for use in upward flows, and made the drift flux model suitable for use in upward or downward flow lines
Thermal hydraulic and neutronic interaction in the rotating bed reactor
International Nuclear Information System (INIS)
Lee, C.C.
1986-01-01
Power transient characteristics in a rotating fluidized bed reactor (RBR) are investigated theoretically. A propellant flow perturbation is assumed to occur in an initially equilibrium state of the core. Transfer functions representing quasi-one-dimensional mutual feedback between thermal hydraulics and neutronics are developed and analyzed in the frequency domain. Neutronic responses are determined by Fermi-age theory for slowing down of fast neutrons and diffusion theory for thermal neutron distribution. Neutron leakage through the exhaust nozzle is accounted for by applying diffuse view factors similar to those applied in radiative heat transfer. The bed expansion behavior is described by a kinematic wave equation derived from the continuity of the gas phase. The drift flux approach is used to determine the yield fractions in the equilibrium bed. Thermal responses of fuel are evaluated by dividing it into several volume-averaged zones to better account for the transient effects over single zone models. Sample calculations are undertaken for the various operation conditions and design parameters of the RBR based on 250 MW/sub t/, 1000 MW/sub t/, and 5000 MW/sub t/ power reactors. The results show that power transients are dependent on the parametric changes of optical thickness and view factors
A Stochastic model for two-station hydraulics exhibiting transient impact
DEFF Research Database (Denmark)
Jacobsen, Judith L.; Madsen, Henrik; Harremoës, Poul
1997-01-01
The objective of the paper is to interpret data on water level variation in a river affected by overflow from a sewer system during rain. The simplest possible, hydraulic description is combined with stochastic methods for data analysis and model parameter estimation. This combination...
Energy Technology Data Exchange (ETDEWEB)
Lee, Youho, E-mail: euo@kaist.ac.kr; Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr
2016-03-15
Highlights: • Use of constant heat transfer coefficient for fracture analysis is not sound. • On-time heat transfer coefficient should be used for thermal fracture prediction. • ∼90% of the actual fracture stresses were predicted with the on-time transient h. • Thermal-hydraulic codes can be used to better predict brittle cladding fracture. • Effects of surface oxides on thermal shock fracture should be accounted by h. - Abstract: This study presents the importance of coherency in modeling thermal-hydraulics and mechanical behavior of a solid for an advanced prediction of cladding thermal shock fracture. In water quenching, a solid experiences dynamic heat transfer rate evolutions with phase changes of the fluid over a short quenching period. Yet, such a dynamic change of heat transfer rates has been overlooked in the analysis of thermal shock fracture. In this study, we are presenting quantitative evidence against the prevailing use of a constant heat transfer coefficient for thermal shock fracture analysis in water. We conclude that no single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials. The presented results show a remarkable stress prediction improvement up to 80–90% of the actual stress with the use of the surface temperature dependent heat transfer coefficient. For thermal shock fracture analysis of brittle fuel cladding such as oxidized zirconium-based alloy or silicon carbide during LWR reflood, transient subchannel heat transfer coefficients obtained from a thermal-hydraulics code should be used as input for stress analysis. Such efforts will lead to a fundamental improvement in thermal shock fracture predictability over the current experimental empiricism for cladding fracture analysis during reflood.
International Nuclear Information System (INIS)
Martin, R.P.; Nassersharif, B.
1987-01-01
TAMUS (Transient Analysis of MUltiple-failure Simulations) is a prototype expert system which is the result of a project investigating and implementing event confidence-levels (used by reactor safety experts in reactor transient analysis) in the form of an expert system. Currently, TAMUS is designed to diagnose reactor transients by analyzing simulated sensor and plant thermal hydraulic information from a system simulation. TAMUS uses a knowledge base of existing emergency nuclear plant operating guidelines and detailed thermal-hydraulic calculating results correlated to confidence-levels. TAMUS can diagnose a number of reactor transients (for example, loss-of-coolant accidents, steam-generator-tube ruptures, loss-of-offsite power, etc.). Future work includes the expansion of the knowledge base and improvement of the deep-knowledge qualitative models
Current and anticipated uses of thermal-hydraulic codes in NFI
Energy Technology Data Exchange (ETDEWEB)
Tsuda, K. [Nuclear Fuel Industries, Ltd., Tokyo (Japan); Takayasu, M. [Nuclear Fuel Industries, Ltd., Sennann-gun (Japan)
1997-07-01
This paper presents the thermal-hydraulic codes currently used in NFI for the LWR fuel development and licensing application including transient and design basis accident analyses of LWR plants. The current status of the codes are described in the context of code capability, modeling feature, and experience of code application related to the fuel development and licensing. Finally, the anticipated use of the future thermal-hydraulic code in NFI is briefly given.
International Nuclear Information System (INIS)
1976-06-01
A discussion is presented of the use of the RELAP4/MOD5 computer program in simulating the thermal-hydraulic behavior of light-water reactor systems when subjected to postulated transients such as a LOCA, pump failure, or nuclear excursion. The volume is divided into main sections which cover: (1) program description, (2) input data, (3) problem initialization, (4) user guidelines, (5) output discussion, (6) source program description, (7) implementation requirements, (8) data files, (9) description of PLOTR4M, (10) description of STH20, (11) summary flowchart, (12) sample problems, (13) problem definition, and (14) problem input
International Nuclear Information System (INIS)
Beyer, M.; Carl, H.; Schuetz, H.; Pietruske, H.; Lenk, S.
2004-07-01
The Forschungszentrum Rossendorf (FZR) e. V. is constructing a new large-scale test facility, TOPFLOW, for thermalhydraulic single effect tests. The acronym stands for transient two phase flow test facility. It will mainly be used for the investigation of generic and applied steady state and transient two phase flow phenomena and the development and validation of models of computational fluid dynamic (CFD) codes. The manual of the test facility must always be available for the staff in the control room and is restricted condition during operation of personnel and also reconstruction of the facility. (orig./GL)
Transient boiling in two-phase helium natural circulation loops
Furci, H.; Baudouy, B.; Four, A.; Meuris, C.
2014-01-01
Two-phase helium natural circulation loops are used for cooling large superconducting magnets, as CMS for LHC. During normal operation or in the case of incidents, transients are exerted on the cooling system. Here a cooling system of this type is studied experimentally. Sudden power changes are operated on a vertical-heated-section natural convection loop, simulating a fast increase of heat deposition on magnet cooling pipes. Mass flow rate, heated section wall temperature and pressure drop variations are measured as a function of time, to assess the time behavior concerning the boiling regime according to the values of power injected on the heated section. The boiling curves and critical heat flux (CHF) values have been obtained in steady state. Temperature evolution has been observed in order to explore the operating ranges where heat transfer is deteriorated. Premature film boiling has been observed during transients on the heated section in some power ranges, even at appreciably lower values than the CHF. A way of attenuating these undesired temperature excursions has been identified through the application of high enough initial heating power.
International Nuclear Information System (INIS)
Wang, Yi-Hsien; Yang, Yue-Tzu
2011-01-01
Transient three-dimensional heat transfer numerical simulations were conducted to investigate a hybrid PCM (phase change materials) based multi-fin heat sink. Numerical computation was conducted with different amounts of fins (0 fin, 3 fins and 6 fins), various heating power level (2 W, 3 W and 4 W), different orientation tests (vertical/horizontal/slanted), and charge and discharge modes. Calculating time step (0.03 s, 0.05 s, and 0.07 s) size was discussed for transient accuracy as well. The theoretical model developed is validated by comparing numerical predictions with the available experimental data in the literature. The results showed that the transient surface temperatures are predicted with a maximum discrepancy within 10.2%. The operation temperature can be controlled well by the attendance of phase change material and the longer melting time can be conducted by using a multi-fin hybrid heat sink respectively. -- Highlights: → Electronic device cooling use phase change materials. → N-eicosane is adapted as phase change materials. → Present surface transient temperatures prediction error is within 10.2%. → Hybrid PCM-heat sink system provides stable operation temperature. → Orientation effects show independent on the phase change performance.
Proceedings of the 8. Brazilian Meeting on Reactor Physics and Thermal Hydraulics
International Nuclear Information System (INIS)
1991-01-01
Some papers about pressurized light water reactors, fast reactors, accident analysis, transients, research reactors, nuclear data collection, thermal hydraulics, reactor monitoring, neutronics are presented. (E.G.)
Tree shoot bending generates hydraulic pressure pulses: a new long-distance signal?
Lopez, Rosana; Badel, Eric; Peraudeau, Sebastien; Leblanc-Fournier, Nathalie; Beaujard, François; Julien, Jean-Louis; Cochard, Hervé; Moulia, Bruno
2014-05-01
When tree stems are mechanically stimulated, a rapid long-distance signal is induced that slows down primary growth. An investigation was carried out to determine whether the signal might be borne by a mechanically induced pressure pulse in the xylem. Coupling xylem flow meters and pressure sensors with a mechanical testing device, the hydraulic effects of mechanical deformation of tree stem and branches were measured. Organs of several tree species were studied, including gymnosperms and angiosperms with different wood densities and anatomies. Bending had a negligible effect on xylem conductivity, even when deformations were sustained or were larger than would be encountered in nature. It was found that bending caused transient variation in the hydraulic pressure within the xylem of branch segments. This local transient increase in pressure in the xylem was rapidly propagated along the vascular system in planta to the upper and lower regions of the stem. It was shown that this hydraulic pulse originates from the apoplast. Water that was mobilized in the hydraulic pulses came from the saturated porous material of the conduits and their walls, suggesting that the poroelastic behaviour of xylem might be a key factor. Although likely to be a generic mechanical response, quantitative differences in the hydraulic pulse were found in different species, possibly related to differences in xylem anatomy. Importantly the hydraulic pulse was proportional to the strained volume, similar to known thigmomorphogenetic responses. It is hypothesized that the hydraulic pulse may be the signal that rapidly transmits mechanobiological information to leaves, roots, and apices.
Summary of ROSA-4 LSTF first phase test program and station blackout (TMLB) test results
International Nuclear Information System (INIS)
Tasaka, K.; Kukita, Y.; Anoda, Y.
1990-01-01
This paper summarizes major test results obtained at the ROSA-4 Large Scale Test Facility (LSTF) during the first phase of the test program. The results from a station blackout (TMLB) test conducted at the end of the first-phase program are described in some detail. The LSTF is an integral test facility being operated by the Japan Atomic Energy Research Institute for simulation of pressurized water reactor (PWR) thermal-hydraulic responses during small-break loss-of-coolant accidents (SBLOCAs) and operational/abnormal transients. It is a 1/48 volumetrically scaled, full-height, full-pressure simulator of a Westinghouse-type 4-loop PWR. The facility includes two symmetric primary loops each one containing an active inverted-U tube steam generator and an active reactor coolant pump. The loop horizontal legs are sized to conserve the scaled (1/24) volumes as well as the length to the square root of the diameter ratio in order to simulate the two-phase flow regime transitions. The primary objective of the LSTF first-phase program was to define the fundamental PWR thermal-hydraulic responses during SBLOCAs and transients. Most of the tests were conducted with simulated component/operator failures, including unavailability of the high pressure injection system and auxiliary feedwater system, as well as operator failure to take corrective actions. The forty-two first phase tests included twenty-nine SBLOCA tests conducted mainly for cold leg breaks, three abnormal transient tests and ten natural circulation tests. Attempts were made in several of the SBLOCA tests to simulate the plant recovery procedures as well as candidate accident management measures for prevention of high-pressure core melt situation. The natural circulation tests simulated the single-phase and two-phase natural circulation as well as reflux condensation behavior in the primary loops in steady or quasi-steady states
Phase-amplitude reduction of transient dynamics far from attractors for limit-cycling systems
Shirasaka, Sho; Kurebayashi, Wataru; Nakao, Hiroya
2017-02-01
Phase reduction framework for limit-cycling systems based on isochrons has been used as a powerful tool for analyzing the rhythmic phenomena. Recently, the notion of isostables, which complements the isochrons by characterizing amplitudes of the system state, i.e., deviations from the limit-cycle attractor, has been introduced to describe the transient dynamics around the limit cycle [Wilson and Moehlis, Phys. Rev. E 94, 052213 (2016)]. In this study, we introduce a framework for a reduced phase-amplitude description of transient dynamics of stable limit-cycling systems. In contrast to the preceding study, the isostables are treated in a fully consistent way with the Koopman operator analysis, which enables us to avoid discontinuities of the isostables and to apply the framework to system states far from the limit cycle. We also propose a new, convenient bi-orthogonalization method to obtain the response functions of the amplitudes, which can be interpreted as an extension of the adjoint covariant Lyapunov vector to transient dynamics in limit-cycling systems. We illustrate the utility of the proposed reduction framework by estimating the optimal injection timing of external input that efficiently suppresses deviations of the system state from the limit cycle in a model of a biochemical oscillator.
Response time verification of in situ hydraulic pressure sensors in a nuclear reactor
International Nuclear Information System (INIS)
Foster, C.G.
1978-01-01
A method and apparatus for verifying response time in situ of hydraulic pressure and pressure differential sensing instrumentation in a nuclear circuit is disclosed. Hydraulic pressure at a reference sensor and at an in situ process sensor under test is varied according to a linear ramp. Sensor response time is then determined by comparison of the sensor electrical analog output signals. The process sensor is subjected to a relatively slowly changing and a relatively rapidly changing hydraulic pressure ramp signal to determine an upper bound for process sensor response time over the range of all pressure transients to which the sensor is required to respond. Signal linearity is independent of the volumetric displacement of the process sensor. The hydraulic signal generator includes a first pressurizable gas reservoir, a second pressurizable liquid and gas reservoir, a gate for rapidly opening a gas communication path between the two reservoirs, a throttle valve for regulating rate of gas pressure equalization between the two reservoirs, and hydraulic conduit means for simultaneously communicating a ramp of hydraulic pressure change between the liquid/gas reservoir and both a reference and a process sensor. By maintaining a sufficient pressure differential between the reservoirs and by maintaining a sufficient ratio of gas to liquid in the liquid/gas reservoir, excellent linearity and minimal transient effects can be achieved for all pressure ranges, magnitudes, and rates of change of interest
Thermal-hydraulic code selection for modular high temperature gas-cooled reactors
Energy Technology Data Exchange (ETDEWEB)
Komen, E M.J.; Bogaard, J.P.A. van den
1995-06-01
In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).
Nuclear reactors transients identification and classification system
International Nuclear Information System (INIS)
Bianchi, Paulo Henrique
2008-01-01
This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)
International Nuclear Information System (INIS)
Maki, Akira; Mori, Michitsugu; Kanemoto, Shigeru; Konishi, Hideo
1997-01-01
A study has been conducted to evaluate how process parameters will exhibit the change in the event of the troubles related to reactor internal by using transient thermal-hydraulic analysis codes (RETRAN3D-MOD002, etc.). In the present study, the following six events are analytically investigated: 1) a leak from the upper plenum; 2) a leak from the middle part of a shroud; 3) a leak from the lower plenum; 4) a leak from the riser pipe for the jet-pump; 5) the blockage of the jet-pump nozzle; and 6) a leak from the jet-pump diffuser. The results by analyses indicated that the leak from the upper plenum resulted in increasing in the inlet temperature of primary loop recirculation (PLR) and in the differential pressure at the core support plate, and decreasing in the neutron flux (reactor power). Similar analyses were made for the five other events to identify the pattern of relevant process parameter variation in each event. (author)
Development and calibration of instruments for measurements in transient two-phase flow
International Nuclear Information System (INIS)
Banerjee, S.; Heidrick, T.R.
1981-01-01
For validation and development of theoretical models for transient two-phase flow, it is necessary to measure local and cross-sectionally averaged thermalhydraulic parameters. Of these parameters, void fraction and mass velocity are the most difficult to measure. In this paper, we present our recent work on various techniques for determining these quantities. The possibility of determining flow regime by using fast neutron transmission is discussed. The development of a miniaturized electrical resistivity probe for measuring local void fraction is described, together with calibrations obtained by integrating the void fraction profile and comparing the cross-sectionally averaged void fraction with direct measurements using two quick closing valves. Results on the calibration of combinations of full-flow turbine meters, Pitot tube rakes and gamma densitometers for measuring cross-sectionally averaged mass velocity in steady steam-water flow are presented. The results are interpreted with a simple model using single-phase calibration factors for the Pitot tube rakes and turbine meters. Calibration experiments were also done in transient steam-water flows and interpretation of the results with the steady state models is also discussed
Bohling, Geoffrey C.; Butler, James J.; Zhan, Xiaoyong; Knoll, Michael D.
2007-01-01
Hydraulic tomography is a promising approach for obtaining information on variations in hydraulic conductivity on the scale of relevance for contaminant transport investigations. This approach involves performing a series of pumping tests in a format similar to tomography. We present a field‐scale assessment of hydraulic tomography in a porous aquifer, with an emphasis on the steady shape analysis methodology. The hydraulic conductivity (K) estimates from steady shape and transient analyses of the tomographic data compare well with those from a tracer test and direct‐push permeameter tests, providing a field validation of the method. Zonations based on equal‐thickness layers and cross‐hole radar surveys are used to regularize the inverse problem. The results indicate that the radar surveys provide some useful information regarding the geometry of the K field. The steady shape analysis provides results similar to the transient analysis at a fraction of the computational burden. This study clearly demonstrates the advantages of hydraulic tomography over conventional pumping tests, which provide only large‐scale averages, and small‐scale hydraulic tests (e.g., slug tests), which cannot assess strata connectivity and may fail to sample the most important pathways or barriers to flow.
Directory of Open Access Journals (Sweden)
Nenad Marković
2016-01-01
Full Text Available The paper presents a new approach in the analysis of a transient state in a system where the feeding source is a transducer-IGBT inverter and load is introduced through the induction motor with its R-L parameters. Induction motors with different parameters of powers and power factors are tested. MATLAB simulation of the three-phase inverter that feeds the induction machine has replaced the missing lab equipment with which mathematical model of this system was verified. According to the selected parameters of the inverter and induction machine and through the simulation in the MATLAB program, the results are obtained in the form of diagrams that verify the model of a transient state of the induction machine operation when it operates as a motor which is presented as a variable R-L load. The transient process of the system three-phase bridge inverter whose active-inductive load is the induction machine in the conditions of the change of the load parameters is analyzed. The model of the transient process in the system formed by the inverter in PWM (Pulse Width Modulation converter and induction machine is developed in the time domain and phase coordinates.
Transient performance simulation of aircraft engine integrated with fuel and control systems
International Nuclear Information System (INIS)
Wang, C.; Li, Y.G.; Yang, B.Y.
2017-01-01
Highlights: • A new performance simulation method for engine hydraulic fuel systems is introduced. • Time delay of engine performance due to fuel system model is noticeable but small. • The method provides details of fuel system behavior in engine transient processes. • The method could be used to support engine and fuel system designs. - Abstract: A new method for the simulation of gas turbine fuel systems based on an inter-component volume method has been developed. It is able to simulate the performance of each of the hydraulic components of a fuel system using physics-based models, which potentially offers more accurate results compared with those using transfer functions. A transient performance simulation system has been set up for gas turbine engines based on an inter-component volume (ICV) method. A proportional-integral (PI) control strategy is used for the simulation of engine controller. An integrated engine and its control and hydraulic fuel systems has been set up to investigate their coupling effect during engine transient processes. The developed simulation system has been applied to a model aero engine. The results show that the delay of the engine transient response due to the inclusion of the fuel system model is noticeable although relatively small. The developed method is generic and can be applied to any other gas turbines and their control and fuel systems.
Nuclear power plant thermal-hydraulic performance research program plan
International Nuclear Information System (INIS)
1988-07-01
The purpose of this program plan is to present a more detailed description of the thermal-hydraulic research program than that provided in the NRC Five-Year Plan so that the research plan and objectives can be better understood and evaluated by the offices concerned. The plan is prepared by the Office of Nuclear Regulatory Research (RES) with input from the Office of Nuclear Reactor Regulation (NRR) and updated periodically. The plan covers the research sponsored by the Reactor and Plant Systems Branch and defines the major issues (related to thermal-hydraulic behavior in nuclear power plants) the NRC is seeking to resolve and provides plans for their resolution; relates the proposed research to these issues; defines the products needed to resolve these issues; provides a context that shows both the historical perspective and the relationship of individual projects to the overall objectives; and defines major interfaces with other disciplines (e.g., structural, risk, human factors, accident management, severe accident) needed for total resolution of some issues. This plan addresses the types of thermal-hydraulic transients that are normally considered in the regulatory process of licensing the current generation of light water reactors. This process is influenced by the regulatory requirements imposed by NRC and the consequent need for technical information that is supplied by RES through its contractors. Thus, most contractor programmatic work is administered by RES. Regulatory requirements involve the normal review of industry analyses of design basis accidents, as well as the understanding of abnormal occurrences in operating reactors. Since such transients often involve complex thermal-hydraulic interactions, a well-planned thermal-hydraulic research plan is needed
Jayachandran, S.; Prithiviraajan, R. N.; Reddy, K. S.
2017-07-01
This paper presents the thermal conductivity of various two-phase materials using modified transient plane source (MTPS) technique. The values are determined by using commercially available C-Therm TCi apparatus. It is specially designed for testing of low to high thermal conductivity materials in the range of 0.02 to 100 Wm-1K-1 within a temperature range of 223-473 K. The results obtained for the two-phase materials (solids, powders and liquids) are having an accuracy better than 5%. The transient method is one of the easiest and less time consuming method to determine the thermal conductivity of the materials compared to steady state methods.
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
ATWS thermal-hydraulic analysis for Krsko Full Scope Simulator validation
International Nuclear Information System (INIS)
Parzer, I.; Kljenak, I.
2005-01-01
The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for annual ANSI/ANS validation of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD3.3 code and the input model for NPP Krsko, delivered by NPP Krsko, was used. In the presented paper the most severe ATWS scenario has been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied assuming AMSAC availability. (author)
Energy Technology Data Exchange (ETDEWEB)
Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.
2017-05-15
Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.
International Nuclear Information System (INIS)
Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.
2017-01-01
Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.
International Nuclear Information System (INIS)
Suh, Jae Seung
2004-08-01
The NSSS (Nuclear Steam Supply System) thermal-hydraulic programs adopted in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited computational capability at that time, they usually used very simplified physical models for the real-time simulation of Ness thermal-hydraulic transients, which entails inaccurate results and, thus, the possibility of so-called 'negative training', especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, a realistic NSSS thermal-hydraulic program ARTS has been developed, it was based on the RETRAN code for the improvement of the Nuclear Power Plant full-scope simulator. Since ARTS is a generalized code to solve a simultaneous equation system, the smaller time-step size should be used if converged solution could not obtain even in a single volume. Therefore, dedicated models which do not force to reduce the time-step size are sometimes more suitable in terms of a real-time calculation and robustness. The PRT(Pressurizer Relief Tank) is a good example, which requires a dedicated model. The PRT consists of subcooled water in bottom and non-condensable gas in top. The sparger merged under subcooled water enhances condensation. The complicated thermal-hydraulic phenomena such as condensation, phase separation with existence of non-condensable gas makes difficult to simulate. Therefore, the PRT volume may limit the time-step size if it is modeled with a general control volume. To mitigate the time-step size reduction due to convergence failure at this component using RETRAN, the PRT model was developed as a dedicated model. The dedicated model was expected to provide reasonable results without convergence problem in the analysis of the system transients. The ARTS code guarantees the real-time calculations of almost all transients and ensures the robustness of simulations. However, there are some possibilities of calculation failure in the
TRACE/VALKIN: a neutronics-thermohydraulics coupled code to analyze strong 3D transients
Energy Technology Data Exchange (ETDEWEB)
Rafael Miro; Gumersindo Verdu; Ana Maria Sanchez [Chemical and Nuclear Engineering Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain); Damian Ginestar [Applied Mathematics Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain)
2005-07-01
Full text of publication follows: A nuclear reactor simulator consists mainly of two different blocks, which solve the models used for the basic physical phenomena taking place in the reactor. In this way, there is a neutronic module which simulates the neutron balance in the reactor core, and a thermal-hydraulics module, which simulates the heat transfer in the fuel, the heat transfer from the fuel to the water, and the different condensation and evaporation processes taking place in the reactor core and in the condenser systems. TRACE is a two-phase, two-fluid thermal-hydraulic reactor systems analysis code. The TRACE acronym stands for TRAC/RELAP Advanced Computational Engine, reflecting its ability to run both RELAP5 and TRAC legacy input models. It includes a three-dimensional kinetics module called PARCS for performing advanced analysis of coupled core thermal-hydraulic/kinetics problems. TRACE-VALKIN code is a new time domain analysis code to study transients in LWR reactors. This code uses the best estimate code TRACE to give account of the heat transfer and thermal-hydraulic processes, and a 3D neutronics module. This module has two options, the MODKIN option that makes use of a modal method based on the assumption that the neutronic flux can be approximately expanded in terms of the dominant lambda modes associated with a static configuration of the reactor core, and the NOKIN option that uses a one-step backward discretization of the neutron diffusion equation. The lambda modes are obtained using the Implicit Restarted Arnoldi approach or the Jacob-Davidson algorithm. To check the performance of the coupled code TRACE-VALKIN against complex 3D neutronic transients, using the cross-sections tables generated with the translator SIMTAB from SIMULATE to TRACE/VALKIN, the Cofrentes NPP SCRAM-61 transient is simulated. Cofrentes NPP is a General Electric BWR-6 design located in Valencia-land (Spain). It is in operation since 1985 and currently in its fifteenth
Two phase flow arising in hydraulics
Czech Academy of Sciences Publication Activity Database
Straškraba, Ivan
2015-01-01
Roč. 60, č. 1 (2015), s. 21-33 ISSN 0862-7940 R&D Projects: GA ČR GA201/08/0012 Institutional support: RVO:67985840 Keywords : compressible fluid * Navier-Stokes equations * hydraulic systems Subject RIV: BA - General Mathematics Impact factor: 0.507, year: 2015 http://link.springer.com/article/10.1007/s10492-015-0083-9
Selected hydraulic test analysis techniques for constant-rate discharge tests
International Nuclear Information System (INIS)
Spane, F.A. Jr.
1993-03-01
The constant-rate discharge test is the principal field method used in hydrogeologic investigations for characterizing the hydraulic properties of aquifers. To implement this test, the aquifer is stressed by withdrawing ground water from a well, by using a downhole pump. Discharge during the withdrawal period is regulated and maintained at a constant rate. Water-level response within the well is monitored during the active pumping phase (i.e., drawdown) and during the subsequent recovery phase following termination of pumping. The analysis of drawdown and recovery response within the stress well (and any monitored, nearby observation wells) provides a means for estimating the hydraulic properties of the tested aquifer, as well as discerning formational and nonformational flow conditions (e.g., wellbore storage, wellbore damage, presence of boundaries, etc.). Standard analytical methods that are used for constant-rate pumping tests include both log-log type-curve matching and semi-log straight-line methods. This report presents a current ''state of the art'' review of selected transient analysis procedures for constant-rate discharge tests. Specific topics examined include: analytical methods for constant-rate discharge tests conducted within confined and unconfined aquifers; effects of various nonideal formation factors (e.g., anisotropy, hydrologic boundaries) and well construction conditions (e.g., partial penetration, wellbore storage) on constant-rate test response; and the use of pressure derivatives in diagnostic analysis for the identification of specific formation, well construction, and boundary conditions
DEFF Research Database (Denmark)
Gong, M.; Zhang, Y.; Weschler, Charles J.
2014-01-01
A transient model is developed to predict dermal absorption of gas-phase chemicals via direct air-to-skin-to-blood transport under non-steady-state conditions. It differs from published models in that it considers convective mass-transfer resistance in the boundary layer of air adjacent to the skin....... Results calculated with this transient model are in good agreement with the limited experimental results that are available for comparison. The sensitivity of the modeled estimates to key parameters is examined. The model is then used to estimate air-to-skin-to-blood absorption of six phthalate esters...... and less absorbed into blood than would a steady-state model. In the 7-day scenario, results calculated by the transient and steady-state models converge over a time period that varies between 3 and 4days for all but the largest phthalate (DEHP). Dermal intake is comparable to or larger than inhalation...
The effect of the virtual mass force term on the stability of transient two-phase flow analysis
International Nuclear Information System (INIS)
Watanabe, Tadashi; Hirano, Masashi; Tanabe, Fumiya
1989-08-01
The effect of the virtual mass force term on the stability of transient two-phase flow analysis is studied. The objective form of the virtual mass acceleration is used. The virtual mass coefficient is determined from the stability condition of basic equations against infinitesimal high wave-number perturbations. The parameter is chosen so that a reasonable agreement between the analytical and experimental sound speed in two-phase flows can be obtained. A one-dimensional sedimentation problem is simulated by the MINCS code which is a tool for transient two-phase flow analysis. The stability analysis is performed for the numerical procedure. It is shown that calculated results are stabilized so long as the virtual mass coefficient satisfies the stability condition of differential equations. (author)
International Nuclear Information System (INIS)
Plas, Roger.
1975-05-01
This computer program describes the flow and heat transfer in steady and transient state in two-phase flows. It is the present stage of the evolution about FLICA, FLICA II and FLICA II B codes which have been used and developed at CEA for the thermal-hydraulic analysis of reactors and experimental loops with heating rod bundles. In the mathematical model all the significant terms of the fundamental hydrodynamic equations are taken into account with the approximations of turbulent viscosity and conductivity. The two-phase flow is calculated by the homogeneous model with slip. In the flow direction an implicit resolution scheme is available, which make possible to study partial or total flow blockage, with upstream and downstream effects. A special model represents the helical wire effects in out-of pile experimental rod bundles [fr
Transient three-phase three-component flow. Pt. 3
International Nuclear Information System (INIS)
Kolev, N.I.
1986-05-01
A mathematical model of a transient three-dimensional three-phase three-component flow described by three-velocity fields in porous body is presented. A combination of separated mass and energy equations together with mixture momentum equations for the flow is used. The mixture equations are used in diffusion form with the assumption that the diffusion velocity can be calculated from empirical correlations. An analytical coupling between the governing equations is developed for calculation of the pressure field. The system is discretized semiimplicitly in 3D-cylindrical space and different solution methods for the algebraic problem are presented. Finally, numerical examples and comparisons with experimental data demonstrate that the method presented is a powerful tool for numerical multiphase flow simulation. (orig.) [de
Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors
International Nuclear Information System (INIS)
Baek, Won Pil; Song, C. H.; Kim, Y. S.
2007-02-01
The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted
International Nuclear Information System (INIS)
Nishi, Yoshihisa; Ueda, Nobuyuki; Kinoshita, Izumi; Yoshida, Kazuo
1999-01-01
An innovative plant concept (ARES), which eliminates intermediate heat-transfer system, was proposed in order to reduce the construction cost of the liquid metal cooled fast breeder reactor (LMFBR). In the present study, representative events of the ARES were picked up and plant transient analysis was conducted in order to reveal the thermal hydraulic characteristics of the ARES. The main results of the analysis are as follows: (1) The core was sufficiently cooled without sodium boiling in an event of mis-operation of the siphon break system. (2) The disturbances of the SG transfer to the core within about 30 seconds in a event of the feed water increasing. This transition period is less than half that of the conventional LMFBR. However, the transient characteristics of the ARES are very similar to those of the conventional LMFBR. Based on these results, it is only necessary to consider the short transition period in order to design the control system. (3) Based on the results of the calculation of a total blackout accident, it is concluded that the use of the SG-side cooling together is profitable in order to satisfy the design safety criteria in design basis events. In the case of full natural-circulation cooling, the use of the SG-side cooling together is profitable in order to satisfy the design safety criteria. (author)
International Nuclear Information System (INIS)
Staalek, Mathias
2008-03-01
Coupled calculations are important for the simulation of nuclear power plants when there is a strong feedback between the neutron kinetics and the thermal-hydraulics. A general coupled model of the Ringhals-3 Pressurized Water Reactor has been developed for this purpose. The development is outlined in the thesis with details given in the appended papers. A PARCS model was developed for the core calculations and a RELAP5 model for the thermal-hydraulic calculations. The RELAP5 model has 157 channels for modelling the flow in the fuel assemblies. This means that there is a one-one correspondence radially between the neutronic and thermal-hydraulic nodalization. This detailed mapping between the neutron kinetics and the thermal-hydraulics makes it possible to use the model for all kinds of transient. To provide realistic material data to the PARCS model, a cross-section interface was developed. With this interface one can import material data from a binary CASMO-4 library file into PARCS. Due to the one-to-one mapping, any any core loading can easily be considered. The PARCS model was benchmarked against measurements of the steady-state power distribution of Ringhals-3. The power shape was well reproduced by the model. Validational work for steady-state conditions of the thermal-hydraulic was also successfully performed. The most challenging part of the validation of a coupled model is for transients. This is much more difficult since the dynamics of the system becomes very important. Two transients that occurred at Ringhals-3 were chosen for the validational work. The first transient was a Load Rejection Transient. In general the model gave good results but some problems were experienced, e.g. the pressurizer pressure turned out to be more difficult to be correctly simulated. The second transient was a Loss of Feed Water transient. A malfunctioning feed water control valve closed, and therefore shut down the feed water supply to the steam generator in one of the
Energy Technology Data Exchange (ETDEWEB)
Staalek, Mathias
2008-03-15
Coupled calculations are important for the simulation of nuclear power plants when there is a strong feedback between the neutron kinetics and the thermal-hydraulics. A general coupled model of the Ringhals-3 Pressurized Water Reactor has been developed for this purpose. The development is outlined in the thesis with details given in the appended papers. A PARCS model was developed for the core calculations and a RELAP5 model for the thermal-hydraulic calculations. The RELAP5 model has 157 channels for modelling the flow in the fuel assemblies. This means that there is a one-one correspondence radially between the neutronic and thermal-hydraulic nodalization. This detailed mapping between the neutron kinetics and the thermal-hydraulics makes it possible to use the model for all kinds of transient. To provide realistic material data to the PARCS model, a cross-section interface was developed. With this interface one can import material data from a binary CASMO-4 library file into PARCS. Due to the one-to-one mapping, any any core loading can easily be considered. The PARCS model was benchmarked against measurements of the steady-state power distribution of Ringhals-3. The power shape was well reproduced by the model. Validational work for steady-state conditions of the thermal-hydraulic was also successfully performed. The most challenging part of the validation of a coupled model is for transients. This is much more difficult since the dynamics of the system becomes very important. Two transients that occurred at Ringhals-3 were chosen for the validational work. The first transient was a Load Rejection Transient. In general the model gave good results but some problems were experienced, e.g. the pressurizer pressure turned out to be more difficult to be correctly simulated. The second transient was a Loss of Feed Water transient. A malfunctioning feed water control valve closed, and therefore shut down the feed water supply to the steam generator in one of the
Strength and reliability of low temperature transient liquid phase bonded Cu-Sn-Cu interconnects
DEFF Research Database (Denmark)
Brincker, Mads; Söhl, Stefan; Eisele, Ronald
2017-01-01
As power electronic devices have tendencies to operate at higher temperatures and current densities, the demand for reliable and efficient packaging technologies are ever increasing. This paper reports the studies on application of transient liquid phase (TLP) bonding of CuSnCu systems...
Thermal-hydraulic methods in fast reactor safety
International Nuclear Information System (INIS)
Weber, D.P.; Briggs, L.L.
1985-01-01
Methods for the solution of thermal-hydraulic problems in liquid metal fast breeder reactors (LMFBRs) arising primarily from transient accident analysis are reviewed. Principal emphasis is given to the important phenomenological issues of sodium boiling and fuel motion. Descriptions of representative phenomenological and mathematical models, computational algorithms, advantages and limitations of the approaches, and current research needs and directions are provided
International Nuclear Information System (INIS)
Gilli, L.; Lathouwers, D.; Kloosterman, J.L.; Van der Hagen, T.H.J.J.
2011-01-01
In this paper a method to perform sensitivity analysis for a simplified multi-physics problem is presented. The method is based on the Adjoint Sensitivity Analysis Procedure which is used to apply first order perturbation theory to linear and nonlinear problems using adjoint techniques. The multi-physics problem considered includes a neutronic, a thermo-kinetics, and a thermal-hydraulics part and it is used to model the time dependent behavior of a sodium cooled fast reactor. The adjoint procedure is applied to calculate the sensitivity coefficients with respect to the kinetic parameters of the problem for two reference transients using two different model responses, the results obtained are then compared with the values given by a direct sampling of the forward nonlinear problem. Our first results show that, thanks to modern numerical techniques, the procedure is relatively easy to implement and provides good estimation for most perturbations, making the method appealing for more detailed problems. (author)
Lumped parameter modeling of a two-phase thermal-hydraulic channel with interface tracking
International Nuclear Information System (INIS)
Jo, J.H.; Kaufman, J.M.; Ruger, C.J.; Stein, S.
1978-01-01
A nonhomogenous, thermal nonequilibrium model for one-dimensional two-phase flow in a heated channel has been formulated in lumped parameter form. The channel is divided into a variable number of flow regimes separated by moving interfaces. The model can be used to predict the behavior of a LWR core and both primary and secondary sides of a steam generator under transient conditions. (author)
Real time thermal hydraulic model for high temperature gas-cooled reactor core
International Nuclear Information System (INIS)
Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng
2013-01-01
A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model. (authors)
Model and simulation of the hydraulic turbine speed regulator of the Atucha I nuclear power plant
International Nuclear Information System (INIS)
Copparoni, G.; Etchepareborda, A.; Urrutia, G.
1992-01-01
The hydraulics turbines of Atucha I Nuclear Power Plant takes advantage of condenser cooling water level difference between the plant and the river to recover about 2,5 MW e. It also supplies emergency power until diesel generators start up. Speed regulation is needed due to the transients that during this process occur. The purpose is to minimize the diesels start up time, and to avoid overshoots on the internal grid frequency. The hydraulic turbine, its speed regulator and the electric system associated with this transient have been modeled. The models and some simulation results are presented in this work. (author)
Self-potential observations during hydraulic fracturing
Energy Technology Data Exchange (ETDEWEB)
Moore, Jeffrey R.; Glaser, Steven D.
2007-09-13
The self-potential (SP) response during hydraulic fracturing of intact Sierra granite was investigated in the laboratory. Excellent correlation of pressure drop and SP suggests that the SP response is created primarily by electrokinetic coupling. For low pressures, the variation of SP with pressure drop is linear, indicating a constant coupling coefficient (Cc) of -200 mV/MPa. However for pressure drops >2 MPa, the magnitude of the Cc increases by 80% in an exponential trend. This increasing Cc is related to increasing permeability at high pore pressures caused by dilatancy of micro-cracks, and is explained by a decrease in the hydraulic tortuosity. Resistivity measurements reveal a decrease of 2% prior to hydraulic fracturing and a decrease of {approx}35% after fracturing. An asymmetric spatial SP response created by injectate diffusion into dilatant zones is observed prior to hydraulic fracturing, and in most cases this SP variation revealed the impending crack geometry seconds before failure. At rupture, injectate rushes into the new fracture area where the zeta potential is different than in the rock porosity, and an anomalous SP spike is observed. After fracturing, the spatial SP distribution reveals the direction of fracture propagation. Finally, during tensile cracking in a point load device with no water flow, a SP spike is observed that is caused by contact electrification. However, the time constant of this event is much less than that for transients observed during hydraulic fracturing, suggesting that SP created solely from material fracture does not contribute to the SP response during hydraulic fracturing.
Thermal-hydraulic modeling needs for passive reactors
Energy Technology Data Exchange (ETDEWEB)
Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)
1997-07-01
The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.
Thermal-hydraulic modeling needs for passive reactors
International Nuclear Information System (INIS)
Kelly, J.M.
1997-01-01
The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken
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-hydraulic
Shock analysis on hydraulic drive control rod during scram
International Nuclear Information System (INIS)
Song Wei; Qin Benke; Bo Hanliang
2013-01-01
Control rod hydraulic drive mechanism (CRHDM) is a new invention of Institute of Nuclear and New Energy Technology of Tsinghua University. The hydraulic absorber buffers the control rod when it scrams. The control rod fast drop impact experiment was conducted and the key parameters of control rod hydraulic buffering performance were obtained. Based on the test results and according to D'Alembert principle, the maximum inertial impact force on the control rod during the fast drop period was applied as equivalent static load force on the control rod. The deformations and stress distributions on the control rod in this worst case were calculated by using finite element software ABAQUS. Calculation results were compared with the experiment results, and it was verified that nonlinear transient dynamics analysis in this problem can be simplified as static analysis. Damage criterion of the control rod fast drop impact process was also given. And it lays foundation for optimal design of the control rod and hydraulic absorber. (authors)
PUMP: analog-hybrid reactor coolant hydraulic transient model
International Nuclear Information System (INIS)
Grandia, M.R.
1976-03-01
The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated
Directory of Open Access Journals (Sweden)
Daqing Zhou
2018-04-01
Full Text Available The transient characteristic of the power-off process is investigated due to its close relation to hydraulic facilities’ safety in a pumped storage hydropower (PSH. In this paper, power-off transient characteristics of a PSH station in pump mode was studied using a three-dimensional (3D unsteady numerical method based on a single-phase and volume of fluid (SP-VOF coupled model. The computational domain covered the entire flow system, including reservoirs, diversion tunnel, surge tank, pump-turbine unit, and tailrace tunnel. The fast changing flow fields and dynamic characteristic parameters, such as unit flow rate, runner rotate speed, pumping lift, and static pressure at measuring points were simulated, and agreed well with experimental results. During the power-off transient process, the PSH station underwent pump mode, braking mode, and turbine mode, with the dynamic characteristics and inner flow configurations changing significantly. Intense pressure fluctuation occurred in the region between the runner and guide vanes, and its frequency and amplitude were closely related to the runner’s rotation speed and pressure gradient, respectively. While the reversed flow rate of the PSH unit reached maximum, some parameters, such as static pressure, torque, and pumping lift would suddenly jump significantly, due to the water hammer effect. The moment these marked jumps occurred was commonly considered as the most dangerous moment during the power-off transient process, due to the blade passages being clogged by vortexes, and chaos pressure distribution on the blade surfaces. The results of this study confirm that 3D SP-VOF hybrid simulation is an effective method to reveal the hydraulic mechanism of the PSH transient process.
COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 2, User's manual
International Nuclear Information System (INIS)
Rector, D.R.; Cuta, J.M.; Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.
1986-11-01
COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code
Hydraulic characterization of aquifers, reservoir rocks, and soils: A history of ideas
Narasimhan, T. N.
1998-01-01
Estimation of the hydraulic properties of aquifers, petroleum reservoir rocks, and soil systems is a fundamental task in many branches of Earth sciences and engineering. The transient diffusion equation proposed by Fourier early in the 19th century for heat conduction in solids constitutes the basis for inverting hydraulic test data collected in the field to estimate the two basic parameters of interest, namely, hydraulic conductivity and hydraulic capacitance. Combining developments in fluid mechanics, heat conduction, and potential theory, the civil engineers of the 19th century, such as Darcy, Dupuit, and Forchheimer, solved many useful problems of steady state seepage of water. Interest soon shifted towards the understanding of the transient flow process. The turn of the century saw Buckingham establish the role of capillary potential in governing moisture movement in partially water-saturated soils. The 1920s saw remarkable developments in several branches of the Earth sciences; Terzaghi's analysis of deformation of watersaturated earth materials, the invention of the tensiometer by Willard Gardner, Meinzer's work on the compressibility of elastic aquifers, and the study of the mechanics of oil and gas reservoirs by Muskat and others. In the 1930s these led to a systematic analysis of pressure transients from aquifers and petroleum reservoirs through the work of Theis and Hurst. The response of a subsurface flow system to a hydraulic perturbation is governed by its geometric attributes as well as its material properties. In inverting field data to estimate hydraulic parameters, one makes the fundamental assumption that the flow geometry is known a priori. This approach has generally served us well in matters relating to resource development primarily concerned with forecasting fluid pressure declines. Over the past two decades, Earth scientists have become increasingly concerned with environmental contamination problems. The resolution of these problems
Simulating the transient regime for main condensate system at Cernavoda NPP
International Nuclear Information System (INIS)
Nita, Iulian; Gheorghiu, Mihai; Prisecaru, Ilie; Dupleac, Daniel
2005-01-01
The purpose of this project is to make a Thermal Hydraulic Analysis of Main Condensate System for getting real-time answer of installation during regimes occurring during normal and abnormal operation. To obtain the analyses the MMS code was used. The boundaries of the systems analysis are extended to Main Feedwater System in order to get a realistic response of Deaerator equipment which are situated between those two systems and have entrances from both systems. In this way we made a complex analysis with main condenser and steam generators as boundaries. We obtained a model for the entire chain of condensate and feedwater preheater with interface just turbine bleed steam. From that we could reduce the number of assumptions necessary to make the analysis. The analyses consist in hydraulics and thermal hydraulics analyses, respectively. For the first case analysed are: - the nominal operation regime with main condensate pumps; - start-up regime with total circulate of condensate to condenser; - 25% MCR (Maximum Continuous Rate) regime (this regime was used in designing the condensate regulating valves at low flow; - 40% MCR regime (with circulate of some condensate flow to condenser); - operating regime of 60% MCR with one main condensate pump operating; - operating regime with auxiliary condensate pump; - operating regime with discharging a condensate flow to condensate storage tank. The thermal hydraulic analyses deal with normal and abnormal operating regimes, respectively. In the first case analysed are the following regimes: - nominal operating regime with main condensate pump operating 100% MCR; - transient regime, 100-80% MCR; - transient regime, 100-80-60% MCR with two pumps in operation and 60 % MCR with one main condensate pump in operation; - transient regime, 100-80-60-60-40 % MCR; - shut-down regime; - start-up regime from Hot zero power to rated power regime. Finally, for the abnormal operating regimes the analyses concerned: - transient regime 100
Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor
International Nuclear Information System (INIS)
Vaiana, F.
2009-11-01
This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)
Steam generator thermal-hydraulics
International Nuclear Information System (INIS)
Inch, W.W.; Scott, D.A.; Carver, M.B.
1980-01-01
This paper discusses a code for detailed numerical modelling of steam generator thermal-hydraulics, and describes related experimental programs designed to promote in-depth understanding of three-dimensional two-phase flow. (auth)
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)
Transient analysis of a thermal storage unit involving a phase change material
Griggs, E. I.; Pitts, D. R.; Humphries, W. R.
1974-01-01
The transient response of a single cell of a typical phase change material type thermal capacitor has been modeled using numerical conductive heat transfer techniques. The cell consists of a base plate, an insulated top, and two vertical walls (fins) forming a two-dimensional cavity filled with a phase change material. Both explicit and implicit numerical formulations are outlined. A mixed explicit-implicit scheme which treats the fin implicity while treating the phase change material explicitly is discussed. A band algorithmic scheme is used to reduce computer storage requirements for the implicit approach while retaining a relatively fine grid. All formulations are presented in dimensionless form thereby enabling application to geometrically similar problems. Typical parametric results are graphically presented for the case of melting with constant heat input to the base of the cell.
Di Prima, Simone; Bagarello, Vincenzo; Iovino, Massimo
2017-04-01
Simple infiltration experiments carried out in the field allow an easy and inexpensive way of characterizing soil hydraulic behavior, maintaining the functional connection of the sampled soil volume with the surrounding soil. The beerkan method consists of a three-dimensional (3D) infiltration experiment at zero pressure head (Haverkamp et al., 1996). It uses a simple annular ring inserted to a depth of about 0.01 m to avoid lateral loss of the ponded water. Soil disturbance is minimized by the limited ring insertion depth. Infiltration time of small volumes of water repeatedly poured on the confined soil are measured to determine the cumulative infiltration. Different algorithms based on this methodology (the so-called BEST family of algorithms) were developed for the determination of soil hydraulic characteristic parameters (Bagarello et al., 2014a; Lassabatere et al., 2006; Yilmaz et al., 2010). Recently, Bagarello et al. (2014b) developed a Simplified method based on a Beerkan Infiltration run (SBI method) to determine saturated soil hydraulic conductivity, Ks, by only the transient phase of a beerkan infiltration run and an estimate of the α* parameter, expressing the relative importance of gravity and capillary forces during an infiltration process (Reynolds and Elrick, 1990). However, several problems yet arise with the existing BEST-algorithms and the SBI method, including (i) the need of supplementary field and laboratory measurements (Bagarello et al., 2013); (ii) the difficulty to detect a linear relationship between I / √t and √t in the early stage of the infiltration process (Bagarello et al., 2014b); (iii) estimation of negative Ks values for hydrophobic soils (Di Prima et al., 2016). In this investigation, a new Simplified method based on the analysis of the Steady-state Beerkan Infiltration run (SSBI method) was proposed and tested. In particular, analytical data were generated to simulate beerkan infiltration experiments for six contrasting
Modelling transient 3D multi-phase criticality in fluidised granular materials - the FETCH code
International Nuclear Information System (INIS)
Pain, C.C.; Gomes, J.L.M.A.; Eaton, M.D.; Ziver, A.K.; Umpleby, A.P.; Oliveira, C.R.E. de; Goddard, A.J.H.
2003-01-01
The development and application of a generic model for modelling criticality in fluidised granular materials is described within the Finite Element Transient Criticality (FETCH) code - which models criticality transients in spatial and temporal detail from fundamental principles, as far as is currently possible. The neutronics model in FETCH solves the neutron transport in full phase space with a spherical harmonics angle of travel representation, multi-group in neutron energy, Crank Nicholson based in time stepping, and finite elements in space. The fluids representation coupled with the neutronics model is a two-fluid-granular-temperature model, also finite element fased. A separate fluid is used to represent the liquid/vapour gas and the solid fuel particle phases, respectively. Particle-particle, particle-wall interactions are modelled using a kinetic theory approach on an analogy between the motion of gas molecules subject to binary collisions and granular flows. This model has been extensively validated by comparison with fluidised bed experimental results. Gas-fluidised beds involve particles that are often extremely agitated (measured by granular temperature) and can thus be viewed as a particularly demanding application of the two-fluid model. Liquid fluidised systems are of criticality interest, but these can become demanding with the production of gases (e.g. radiolytic and water vapour) and large fluid/particle velocities in energetic transients. We present results from a test transient model in which fissile material ( 239 Pu) is presented as spherical granules subsiding in water, located in a tank initially at constant temperature and at two alternative over-pressures in order to verify the theoretical model implemented in FETCH. (author)
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.
Validation and applicability of the 3D core kinetics and thermal hydraulics coupled code SPARKLE
International Nuclear Information System (INIS)
Miyata, Manabu; Maruyama, Manabu; Ogawa, Junto; Otake, Yukihiko; Miyake, Shuhei; Tabuse, Shigehiko; Tanaka, Hirohisa
2009-01-01
The SPARKLE code is a coupled code system based on three individual codes whose physical models have already been verified and validated. Mitsubishi Heavy Industries (MHI) confirmed the coupling calculation, including data transfer and the total reactor coolant system (RCS) behavior of the SPARKLE code. The confirmation uses the OECD/NEA MSLB benchmark problem, which is based on Three Mile Island Unit 1 (TMI-1) nuclear power plant data. This benchmark problem has been used to verify coupled codes developed and used by many organizations. Objectives of the benchmark program are as follows. Phase 1 is to compare the results of the system transient code using point kinetics. Phase 2 is to compare the results of the coupled three-dimensional (3D) core kinetics code and 3D core thermal-hydraulics (T/H) code, and Phase 3 is to compare the results of the combined coupled system transient code, 3D core kinetics code, and 3D core T/H code as a total validation of the coupled calculation. The calculation results of the SPARKLE code indicate good agreement with other benchmark participants' results. Therefore, the SPARKLE code is validated through these benchmark problems. In anticipation of applying the SPARKLE code to licensing analyses, MHI and Japanese PWR utilities have established a safety analysis method regarding the calculation conditions such as power distributions, reactivity coefficients, and event-specific features. (author)
Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments
Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride
The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.
International Nuclear Information System (INIS)
Choi, J.D.; Yoon, H.Y.; Um, K.S.; Kim, H.C.; Sim, S.K.
1997-01-01
Best estimate transient analysis code, TASS, has been developed for the normal and transient simulation of the Westinghouse and ABB-CE type PWRs. TASS thermal hydraulic model is based on the non-homogeneous, non-equilibrium two-phase continuity, energy and mixture momentum equations with constitutive relations for closure. Core neutronics model employs both the point kinetics and one-dimensional neutron diffusion model. Semi-implicit numerical scheme is used to solve the discretized finite difference equations. TASS one dimensional neutronics core model has been assessed through the reactivity induced transient analyses for the KORI-3, three loop Westinghouse PWR, and Younggwang-3 (YGN-3), two-loop ABB-CE PWR, nuclear power plants currently operating in Korea. The assessment showed that the TASS one dimensional neutronics core model can be applied for the Westinghouse and ABB-CE type PWRs to gain thermal margin which is necessary for a potential use of the high fuel burnup, extended fuel cycle, power upgrading and for the plant life extension
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
Parametric Assessment of Perchloroethylene Hydraulic Behaviour in a Two-Phase System
International Nuclear Information System (INIS)
Chatrenour, M.; Homaee, M.; Asadi Kapourchal, S.; Mahmoodian Shoshtari, M.
2016-01-01
Quantitative description of soil hydraulic properties is crucial for preventing organic contamination entering into the soil and groundwater. In order to assess the hydraulic behaviour of Perchloroethylene as a toxic chlorinated contaminant in soil, the retention curves for Perchloroethylene and water were determined. The Saturated hydraulic conductivity of both fluids examined was determined by the constant head method. The Perchloroethylene and water hydraulic conductivities obtained were 492.84 and 450.27 cm day-1, respectively. The porous medium retention parameters were obtained based on the van Genuchten, Brooks-Corey and Kosugi retention models. Further, the unsaturated hydraulic conductivity for both fluids was obtained based on the Mualem-Brooks-Corey, Mualem-van Genuchten and Mualem-Kosugi models. The accuracy performance of the models was assessed using some statistics including ME, RMSE, EF, CD and CRM. Results indicated that the van Genuchten model provided better estimations than other models when the fluid studied was Perchloroethylene. The results further indicated that the magnitudes of the pore-size distribution parameters and the bubbling pressure parameters are reduced more in a water-air system compared to a Perchloroethylene-air system. This can be attributed to the high viscosity of water and its considerable resistance against flow. This implies that more suction is needed to drain water out from a porous medium than Perchloroethylene. Consequently, a porous medium provides less retention for Perchloroethylene at a given quantity of fluid than water. Owing to lower Perchloroethylene viscosity, the saturated and unsaturated porous medium hydraulic conductivity of Perchloroethylene was greater than that of water. Since Perchloroethylene has lower retention and larger hydraulic conductivity than water, its infiltration into a porous medium would lead to its faster movement towards groundwater.
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
Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems
International Nuclear Information System (INIS)
Song, C. H.; Chung, M. K.; Park, C. K. and others
2005-04-01
The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved
Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems
Energy Technology Data Exchange (ETDEWEB)
Song, C. H.; Chung, M. K.; Park, C. K. and others
2005-04-15
The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.
RAMONA-3B/MINET composite representation of BWR thermal-hydraulic systems
International Nuclear Information System (INIS)
Van Tuyle, G.J.; Slovik, G.; Cazzoli, E.G.; Nepsee, T.C.; Guppy, J.G.
1985-01-01
The modification and interfacing of two computer codes, RAMONA-3B and MINET, for the thermal hydraulic transient analysis of a Boiling Water Reactor nuclear steam supply system, is described. The RAMONA-3B code provides for multi-channel thermal hydraulics and three-dimensional (or one-dimensional) neutron kinetics analysis of a boiling water reactor core. The RAMONA-3B system representation terminates at the end of the steam line and at the junction of the feedwater line at the vessel inlet. By interfacing RAMONA-3B with MINET, a generic balance-of-plant systems analysis code, a complete BWR systems code with detailed core modeling was obtained. The result is a code of particular importance to the analysis of transients such as ATWS. A comparison between the 3-D and 1-D neutronics representation is provided, along with a test case utilizing the composite RAMONA-3B/MINET code
BR2 reactor core steady state transient modeling
International Nuclear Information System (INIS)
Makarenko, A.; Petrova, T.
2000-01-01
A coupled neutronics/hydraulics/heat-conduction model of the BR2 reactor core is under development at SCK-CEN. The neutron transport phenomenon has been implemented as steady state and time dependent nodal diffusion. The non-linear heat conduction equation in-side fuel elements is solved with a time dependent finite element method. To allow coupling between functional modules and to simulate subcooled regimes, a simple single-phase hydraulics has been introduced, while the two-phase hydraulics is under development. Multiple tests, general benchmark cases as well as calculation/experiment comparisons demonstrated a good accuracy of both neutronic and thermal hydraulic models, numerical reliability and full code portability. A refinement methodology has been developed and tested for better neutronic representation in hexagonal geometry. Much effort is still needed to complete the development of an extended cross section library with kinetic data and two-phase flow representation. (author)
One-dimensional transient unequal velocity two-phase flow by the method of characteristics
International Nuclear Information System (INIS)
Rasouli, F.
1981-01-01
An understanding of two-phase flow is important when one is analyzing the accidental loss of coolant or when analyzing industrial processes. If a pipe in the steam generator of a nuclear reactor breaks, the flow will remain critical (or choked) for almost the entire blowdown. For this reason the knowledge of the two-phase maximum (critical) flow rate is important. A six-equation model--consisting of two continuity equations, two energy equations, a mixture momentum equation, and a constitutive relative velocity equation--is solved numerically by the method of characteristics for one-dimensional, transient, two-phase flow systems. The analysis is also extended to the special case of transient critical flow. The six-equation model is used to study the flow of a nonequilibrium sodium-argon system in a horizontal tube in which the nonequilibrium sodium-argon system in a horizontal tube in which the critical flow condition is at the entrance. A four-equation model is used to study the pressure-pulse propagation rate in an isothermal air-water system, and the results that are found are compared with the experimental data. Proper initial and boundary conditions are obtained for the blowdown problem. The energy and mass exchange relations are evaluated by comparing the model predictions with results of void-fraction and heat-transfer experiments. A simplified two-equation model is obtained for the special case of two incompressible phases. This model is used in the preliminary analysis of batch sedimentation. It is also used to predict the shock formation in the gas-solid fluidized bed
Evaluation of hot spot factors for thermal and hydraulic design of HTTR
International Nuclear Information System (INIS)
Maruyama, So; Yamashita, Kiyonobu; Fujimoto, Nozomu; Murata, Isao; Sudo, Yukio; Murakami, Tomoyuki; Fujii, Sadao.
1993-01-01
High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950degC in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly. (author)
A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS
Energy Technology Data Exchange (ETDEWEB)
D’Auria, F; Rohatgi, Upendra S.
2017-01-12
The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.
Transient thermal-hydraulic simulations of direct cycle gas cooled reactors
International Nuclear Information System (INIS)
Tauveron, Nicolas; Saez, Manuel; Marchand, Muriel; Chataing, Thierry; Geffraye, Genevieve; Bassi, Christophe
2005-01-01
This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to 'active' heat extraction systems
Macroscopic balance equations for two-phase flow models
International Nuclear Information System (INIS)
Hughes, E.D.
1979-01-01
The macroscopic, or overall, balance equations of mass, momentum, and energy are derived for a two-fluid model of two-phase flows in complex geometries. These equations provide a base for investigating methods of incorporating improved analysis methods into computer programs, such as RETRAN, which are used for transient and steady-state thermal-hydraulic analyses of nuclear steam supply systems. The equations are derived in a very general manner so that three-dimensional, compressible flows can be analysed. The equations obtained supplement the various partial differential equation two-fluid models of two-phase flow which have recently appeared in the literature. The primary objective of the investigation is the macroscopic balance equations. (Auth.)
Transient behaviour of small HTR for cogeneration
International Nuclear Information System (INIS)
Verkerk, E.C.; Van Heek, A.I.
2000-01-01
The Dutch market for combined generation of heat and power identifies a unit size of 40 MW thermal for the conceptual design of a nuclear cogeneration plant. The ACACIA system provides 14 MWe electricity combined with 17 t/h of high temperature steam (220 deg C, 10 bar) with a pebble-bed high temperature reactor directly coupled with a helium compressor and a helium turbine. The design of this small CHP unit that is used for industrial applications is mainly based on a pre-feasibility study in 1996, performed by a joint working group of five Dutch organisations, in which technical feasibility was shown. Thermal hydraulic and reactor physics analyses show favourable control characteristics during normal operation and a benign response to loss of helium coolant and loss of flow conditions. Throughout the response on these highly infrequent conditions, ample margin exists between the highest fuel temperatures and the temperature above which fuel degradation will occur. To come to quantitative statements about the ACACIA transient behaviour, a calculational coupling between the high temperature reactor core analysis code package PANTHER/DIREKT and the thermal hydraulic code RELAP5 for the energy conversion system has been made. This coupling offers a more realistic simulation of the entire system, since it removes the necessity of forcing boundary conditions on the simulation models at the data transfer points. In this paper, the models used for the dynamic components of the energy conversion system are described, and the results of the calculation for two operational transients in order to demonstrate the effects of the interaction between reactor core and its energy conversion system are shown. Several transient cases that are representative as operational transients for an HTR will be discussed, including one representing a load rejection case that shows the functioning of the control system, in particular the bypass valve. Another transient is a load following
The 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
On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application
International Nuclear Information System (INIS)
Freels, J.D.
1993-01-01
This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ''the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed
AIREKMOD-RR, Reactivity Transients in Nuclear Research Reactors
International Nuclear Information System (INIS)
Baggoura, B.; Mazrou, H.
2001-01-01
1 - Description of program or function: AIREMOD-RR is a point kinetics code which can simulate fast transients in nuclear research reactor cores. It can also be used for theoretical reactor dynamics studies. It is used for research reactor kinetic analysis and provides a point neutron kinetic capability. The thermal hydraulic behavior is governed by a one-dimensional heat balance equation. The calculations are restricted to a single equivalent unit cell which consists of fuel, clad and coolant. 2 - Method of solution: For transient reactor kinetic calculations a modified Runge Kutta numerical method is used. The external reactivity insertion, specified as a function of time, is converted in dollar ($) unit. The neutron density, energy release and feedback variables are given at each time step. The two types of reactivity feedback considered are: Doppler effect and moderator effect. A new expression for the reactivity dependence on the feedback variables has been introduced in the present version of the code. The feedback reactivities are fitted in power series expression. 3 - Restrictions on the complexity of the problem: The number of delayed neutron groups and the total number of equations are limited only by computer storage capabilities. - Coolant is always in liquid phase. - Void reactivity feedback is not considered
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
Light-water-reactor coupled neutronic and thermal-hydraulic codes
International Nuclear Information System (INIS)
Diamond, D.J.
1982-01-01
An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented
International Nuclear Information System (INIS)
Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.
1986-11-01
COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations: however, the transient capability has not yet been validated. This volume describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods
Energy Technology Data Exchange (ETDEWEB)
Yu, C.; Matray, J.-M. [Institut de Radioprotection et de Sûreté Nucléaire, Fontenay-aux-Roses, (France); Yu, C.; Gonçalvès, J. [Aix Marseille Université UMR 6635 CEREGE Technopôle Environnement Arbois-Méditerranée Aix-en-Provence, Cedex 4 (France); and others
2017-04-15
The deep borehole (DB) experiment gave the opportunity to acquire hydraulic parameters in a hydraulically undisturbed zone of the Opalinus Clay at the Mont Terri rock laboratory (Switzerland). Three methods were used to estimate hydraulic conductivity and specific storage values of the Opalinus Clay formation and its bounding formations through the 248 m deep borehole BDB-1: application of a Poiseuille-type law involving petrophysical measurements, spectral analysis of pressure time series and in situ hydraulic tests. The hydraulic conductivity range in the Opalinus Clay given by the first method is 2 × 10{sup -14}-6 × 10{sup -13} m s{sup -1} for a cementation factor ranging between 2 and 3. These results show low vertical variability whereas in situ hydraulic tests suggest higher values up to 7 × 10{sup -12} m s{sup -1}. Core analysis provides economical estimates of the homogeneous matrix hydraulic properties but do not account for heterogeneities at larger scale such as potential tectonic conductive features. Specific storage values obtained by spectral analysis are consistent and in the order of 10{sup -6} m{sup -1}, while formulations using phase shift and gain between pore pressure signals were found to be inappropriate to evaluate hydraulic conductivity in the Opalinus Clay. The values obtained are globally in good agreement with the ones obtained previously at the rock laboratory. (authors)
Advanced numerical methods for three dimensional two-phase flow calculations in PWR
International Nuclear Information System (INIS)
Toumi, I.; Gallo, D.; Royer, E.
1997-01-01
This paper is devoted to new numerical methods developed for three dimensional two-phase flow calculations. These methods are finite volume numerical methods. They are based on an extension of Roe's approximate Riemann solver to define convective fluxes versus mean cell quantities. To go forward in time, a linearized conservative implicit integrating step is used, together with a Newton iterative method. We also present here some improvements performed to obtain a fully implicit solution method that provides fast running steady state calculations. This kind of numerical method, which is widely used for fluid dynamic calculations, is proved to be very efficient for the numerical solution to two-phase flow problems. This numerical method has been implemented for the three dimensional thermal-hydraulic code FLICA-4 which is mainly dedicated to core thermal-hydraulic transient and steady-state analysis. Hereafter, we will also find some results obtained for the EPR reactor running in a steady-state at 60% of nominal power with 3 pumps out of 4, and a thermal-hydraulic core analysis for a 1300 MW PWR at low flow steam-line-break conditions. (author)
Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems
International Nuclear Information System (INIS)
Song, C. H.; Baek, W. P.; Chung, M. K.
2007-06-01
The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base
Dynamic thermo-hydraulic model of district cooling networks
International Nuclear Information System (INIS)
Oppelt, Thomas; Urbaneck, Thorsten; Gross, Ulrich; Platzer, Bernd
2016-01-01
Highlights: • A dynamic thermo-hydraulic model for district cooling networks is presented. • The thermal modelling is based on water segment tracking (Lagrangian approach). • Thus, numerical errors and balance inaccuracies are avoided. • Verification and validation studies proved the reliability of the model. - Abstract: In the present paper, the dynamic thermo-hydraulic model ISENA is presented which can be applied for answering different questions occurring in design and operation of district cooling networks—e.g. related to economic and energy efficiency. The network model consists of a quasistatic hydraulic model and a transient thermal model based on tracking water segments through the whole network (Lagrangian method). Applying this approach, numerical errors and balance inaccuracies can be avoided which leads to a higher quality of results compared to other network models. Verification and validation calculations are presented in order to show that ISENA provides reliable results and is suitable for practical application.
Modelling of an ULOF transient in a sodium fast reactor
International Nuclear Information System (INIS)
Droin, Jean-Baptiste
2016-01-01
Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R and D program of CEA (French Commissariat a l'Energie Atomique et aux Energies Alternatives), safety in case of severe accidents is assessed.Such transients are usually simulated with mechanistic codes (such as SAS-SFR and SIMMER III). as a complement to these codes, which give reference accidental transient calculations, a new physico-statistical approach is currently followed by the CEA; its final objective being to derive the variability of the main results of interest for safety. This approach involves a fast-running description of extended accident sequences coupling physical models for the main phenomena to advanced statistical analysis techniques. It enables to perform a large number of simulations in a reasonable computational time and to describe all the possible bifurcations of the accident transient.In this context, this PhD work presents the physical tool (models and results assessment) dedicated to the initiation and primary phases of an Unprotected Loss Of Flow accident (i.e. until the end of sub-assemblies degradation and before large molten pools formation). The accident phenomenology during these phases is described and illustrated by numerous experimental evidences.It is underlined that the features of the new heterogeneous core concept (called CFV of the French ASTRID prototype) leads to different kinds of ULOF transients than those occurring in the previous past homogeneous cores (SuperPhenix, Phenix...). Indeed, its negative void effect drops the nuclear power when sodium heats-up and possibly boils. This enables three types of ULOF transients characterized by various core final states; the first two types leading to final coolable core states in natural circulation flow (the first one in single phase, the second one in stabilized two-phase flow) whereas the core undergoes a flow excursion followed by sub-assemblies degradation in the last type. In this study, a
International Nuclear Information System (INIS)
1976-05-01
Results are presented of analyses of the transient thermal-hydraulic conditions and radiological release consequences which would occur in power plants which employ a Combustion Engineering Nuclear Steam Supply System during Anticipated Transients Without Scram due to a lack of insertion of the Control Element Assemblies upon signals for automatic or manual reactor shutdown. The transients analyzed include all events which meet the criterion to be considered as anticipated at least once in the plant lifetime with automatic reactor shutdown
Energy Technology Data Exchange (ETDEWEB)
Garner, P.L.; Blomquist, R.N.; Gelbard, E.M.
1992-09-01
The COMMIX-1AR/P computer program is designed for analyzing the steady-state and transient aspects of single-phase fluid flow and heat transfer in three spatial dimensions. This version is an extension of the modeling in COMMIX-1A to include multiple fluids in physically separate regions of the computational domain, modeling descriptions for pumps, radiation heat transfer between surfaces of the solids which are embedded in or surround the fluid, a k-{var_epsilon} model for fluid turbulence, and improved numerical techniques. The porous-medium formulation in COMMIX allows the program to be applied to a wide range of problems involving both simple and complex geometrical arrangements. The input preparation and execution procedures are presented for the COMMIX-1AR/P program and several postprocessor programs which produce graphical displays of the calculated results.
Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis
Hoogenboom, J. Eduard; Sjenitzer, Bart L.
2014-06-01
To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.
Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis
International Nuclear Information System (INIS)
Hoogenboom, J.E.
2013-01-01
To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branch-less collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires the coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail. (authors)
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
Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios
Directory of Open Access Journals (Sweden)
Khedr Ahmed
2005-01-01
Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.
Transient analyses for a molten salt fast reactor with optimized core geometry
Energy Technology Data Exchange (ETDEWEB)
Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)
2015-10-15
Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.
Thermal modeling of a hydraulic hybrid vehicle transmission based on thermodynamic analysis
International Nuclear Information System (INIS)
Kwon, Hyukjoon; Sprengel, Michael; Ivantysynova, Monika
2016-01-01
Hybrid vehicles have become a popular alternative to conventional powertrain architectures by offering improved fuel efficiency along with a range of environmental benefits. Hydraulic Hybrid Vehicles (HHV) offer one approach to hybridization with many benefits over competing technologies. Among these benefits are lower component costs, more environmentally friendly construction materials, and the ability to recover a greater quantity of energy during regenerative braking which make HHVs partially well suited to urban environments. In order to further the knowledge base regarding HHVs, this paper explores the thermodynamic characteristics of such a system. A system model is detailed for both the hydraulic and thermal components of a closed circuit hydraulic hybrid transmission following the FTP-72 driving cycle. Among the new techniques proposed in this paper is a novel method for capturing rapid thermal transients. This paper concludes by comparing the results of this model with experimental data gathered on a Hardware-in-the-Loop (HIL) transmission dynamometer possessing the same architecture, components, and driving cycle used within the simulation model. This approach can be used for several applications such as thermal stability analysis of HHVs, optimal thermal management, and analysis of the system's thermodynamic efficiency. - Highlights: • Thermal modeling for HHVs is introduced. • A model for the hydraulic and thermal system is developed for HHVs. • A novel method for capturing rapid thermal transients is proposed. • The thermodynamic system diagram of a series HHV is predicted.
Optimization of Classical Hydraulic Engine Mounts Based on RMS Method
Directory of Open Access Journals (Sweden)
J. Christopherson
2005-01-01
Full Text Available Based on RMS averaging of the frequency response functions of the absolute acceleration and relative displacement transmissibility, optimal parameters describing the hydraulic engine mount are determined to explain the internal mount geometry. More specifically, it is shown that a line of minima exists to define a relationship between the absolute acceleration and relative displacement transmissibility of a sprung mass using a hydraulic mount as a means of suspension. This line of minima is used to determine several optimal systems developed on the basis of different clearance requirements, hence different relative displacement requirements, and compare them by means of their respective acceleration and displacement transmissibility functions. In addition, the transient response of the mount to a step input is also investigated to show the effects of the optimization upon the time domain response of the hydraulic mount.
Energy Technology Data Exchange (ETDEWEB)
Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)
2013-07-01
A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics
International Nuclear Information System (INIS)
Jordan, K. A.; Schubring, D.; Girardin, G.; Pautz, A.
2013-01-01
A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics
Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor
Energy Technology Data Exchange (ETDEWEB)
Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq Rede), Rio de Janeiro, RJ (Brazil)
2013-07-01
The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)
International Nuclear Information System (INIS)
Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.
1998-03-01
This document is the User's Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code's capabilities and limitations; Chapter 2 describes the code's structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs
TRAC-B thermal-hydraulic analysis of the Black Fox boiling water reactor
International Nuclear Information System (INIS)
Martin, R.P.
1993-05-01
Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electric Black Fox Nuclear Project boiling water reactor were performed using the TRAC-BF1 computer code. This work is sponsored by the US Nuclear Regulatory Commission and is being done in conjunction with future analysis work at the US Nuclear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. These accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient conditions
Outage Risk Assessment and Management (ORAM) thermal-hydraulics toolkit
International Nuclear Information System (INIS)
Denny, V.E.; Wassel, A.T.; Issacci, F.; Pal Kalra, S.
2004-01-01
A PC-based thermal-hydraulic toolkit for use in support of outage optimization, management and risk assessment has been developed. This mechanistic toolkit incorporates simple models of key thermal-hydraulic processes which occur during an outage, such as recovery from or mitigation of outage upsets; this includes heat-up of water pools following loss of shutdown cooling, inadvertent drain down of the RCS, boiloff of coolant inventory, heatup of the uncovered core, and reflux cooling. This paper provides a list of key toolkit elements, briefly describes the technical basis and presents illustrative results for RCS transient behavior during reflux cooling, peak clad temperatures for an uncovered core and RCS response to loss of shutdown cooling. (author)
Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD
Energy Technology Data Exchange (ETDEWEB)
Trambauer, K. [GRS, Garching (Germany)
1997-07-01
The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.
Some new directions in system transient simulation
International Nuclear Information System (INIS)
Ransom, V.H.
1986-01-01
The current research in system transient simulation at the Idaho National Engineering Laboratory (INEL) is summarized in this paper and three new directions that are emerging from this work are discussed. The new directions are: development of an Advanced Thermal Hydraulic Energy Network Analyzer (ATHENA) having new modeling capability, use of expert systems for enhancing simulation methods, and the trend to individual workstations for simulation
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
Overview of SPH-ALE applications for hydraulic turbines in ANDRITZ Hydro
Rentschler, M.; Marongiu, J. C.; Neuhauser, M.; Parkinson, E.
2018-02-01
Over the past 13 years, ANDRITZ Hydro has developed an in-house tool based on the SPH-ALE method for applications in flow simulations in hydraulic turbines. The initial motivation is related to the challenging simulation of free surface flows in Pelton turbines, where highly dynamic water jets interact with rotating buckets, creating thin water jets traveling inside the housing and possibly causing disturbances on the runner. The present paper proposes an overview of industrial applications allowed by the developed tool, including design evaluation of Pelton runners and casings, transient operation of Pelton units and free surface flows in hydraulic structures.
Directory of Open Access Journals (Sweden)
Masahiro eKawasaki
2014-03-01
Full Text Available Electroencephalogram (EEG phase synchronization analyses can reveal large-scale communication between distant brain areas. However, it is not possible to identify the directional information flow between distant areas using conventional phase synchronization analyses. In the present study, we applied transcranial magnetic stimulation (TMS to the occipital area in subjects who were resting with their eyes closed, and analyzed the spatial propagation of transient TMS-induced phase resetting by using the transfer entropy (TE, to quantify the causal and directional flow of information. The time-frequency EEG analysis indicated that the theta (5 Hz phase locking factor (PLF reached its highest value at the distant area (the motor area in this study, with a time lag that followed the peak of the transient PLF enhancements of the TMS-targeted area at the TMS onset. PPI (phase-preservation index analyses demonstrated significant phase resetting at the TMS-targeted area and distant area. Moreover, the TE from the TMS-targeted area to the distant area increased clearly during the delay that followed TMS onset. Interestingly, the time lags were almost coincident between the PLF and TE results (152 vs. 165 ms, which provides strong evidence that the emergence of the delayed PLF reflects the causal information flow. Such tendencies were observed only in the higher-intensity TMS condition, and not in the lower-intensity or sham TMS conditions. Thus, TMS may manipulate large-scale causal relationships between brain areas in an intensity-dependent manner. We demonstrated that single-pulse TMS modulated global phase dynamics and directional information flow among synchronized brain networks. Therefore, our results suggest that single-pulse TMS can manipulate both incoming and outgoing information in the TMS-targeted area associated with functional changes.
Microstructural development in NiAl/Ni-Si-B/Ni transient liquid phase bonds
International Nuclear Information System (INIS)
Gale, W.F.; Orel, S.V.
1996-01-01
A transmission electron microscopy (TEM) based investigation of microstructural development during transient liquid phase bonding of near-stoichiometric NiAl to commercial purity nickel is presented in this article. The work described employed Ni-4.5 wt pct Si-3.2 wt pct B (BNi-3) melt-spun interlayers. The precipitation of both Ni-Al based phases and borides within the joint and adjacent substrate regions is discussed. The article considers martensite formation (within the NiAl substrate) and the precipitation of L1 2 type phases (both within the joint and at the interface with the NiAl substrate). The relative roles of the two substrate materials (NiAl and Ni) in the isothermal resolidification process are identified. The preferential formation of Ni 3 B boride phases in the Ni substrate near the original location of the Ni substrate-joint interface is discussed and contrasted with the absence of similar events in the NiAl substrate
Multimodel Robust Control for Hydraulic Turbine
Osuský, Jakub; Števo, Stanislav
2014-01-01
The paper deals with the multimodel and robust control system design and their combination based on M-Δ structure. Controller design will be done in the frequency domain with nominal performance specified by phase margin. Hydraulic turbine model is analyzed as system with unstructured uncertainty, and robust stability condition is included in controller design. Multimodel and robust control approaches are presented in detail on hydraulic turbine model. Control design approaches are compared a...
International Nuclear Information System (INIS)
Marais, D.; Greyvenstein, G. P.
2008-01-01
TINTE is a well established reactor analysis code which models the transient behaviour of pebble bed reactor cores but it does not include the capabilities to model a power conversion unit (PCU). This raises the issue that TINTE cannot model full system transients. One way to overcome this problem is to supply TINTE with time-dependant thermal-hydraulic boundary conditions which are obtained from PCU simulations. This study investigates a method to provide boundary conditions for the nuclear code TINTE during full system transients. This was accomplished by creating a high level interface between the systems CFD code Flownex and TINTE. An indirect coupling method is explored whereby characteristics of the PCU are matched to characteristics of the nuclear core. This method eliminates the need to iterate between the two codes. A number of transients are simulated using the coupled code and then compared against stand-alone Flownex simulations. The coupling method introduces relatively small errors when reproducing mass flow, temperature and pressure in steady state analysis, but become more pronounced when dealing with fast thermal-hydraulic transients. Decreasing the maximum time step length of TINTE reduces this problem, but increases the computational time. Copyright ASME 2008. (authors)
International Nuclear Information System (INIS)
Seubert, A.; Velkov, K.; Langenbuch, S.
2008-01-01
This paper describes the time-dependent 3D discrete ordinates transport code TORT-TD. Thermal-hydraulic feedback is considered by coupling TORT-TD with the thermal-hydraulics system code ATHLET. The coupled code TORT-TD/ATHLET allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. The nuclear cross sections are interpolated between pre-calculated table values of fuel temperature, moderator density and boron concentration. For verification of the implementation, selected test cases have been calculated by TORT-TD/ATHLET. They include a control rod ejection transient in a small PWR fuel assembly arrangement and a local boron concentration change in a single PWR fuel assembly. In the latter, special attention has been paid to study the influence of the thermal-hydraulic feedback modelling in ATHLET. The results obtained for a control rod ejection accident in a PWR quarter core demonstrate the applicability of TORT-TD/ATHLET. (authors)
International Nuclear Information System (INIS)
Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.
2010-10-01
Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)
Energy Technology Data Exchange (ETDEWEB)
Molina, M. C.; Prieto, J.; Olmedo, J.; Mota, M.
2013-07-01
The present paper presents the study of the possible transient hydraulic that they could occur in the essential service water system due to changes in modes of operation, as well as replacement of components or failure of these within the same operating mode. For a complete analysis, it has created a computer model of the system through software EcosimPro, whereby different models have been corresponding to each division's system, making the check that in any mode of operation, and in any event, the values be exceeded the design for the system and its components.
International Nuclear Information System (INIS)
Ohtaka, Masahiko; Ohshima, Hiroyuki
1998-10-01
A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including inter-wrapper flow under various reactor operation conditions. In this work, the core module as a main part of the ACT developed last year, which simulates thermal-hydraulics in the subassemblies and the inter-subassembly gaps, was coupled with an one dimensional plant system thermal-hydraulic analysis code LEDHER to simulate transients in the primary heat transport system and to give appropriate boundary conditions to the core model. The effective algorithm to couple these two calculation modules was developed, which required minimum modification of them. In order to couple these two calculation modules on the computing system, parallel computing technique using PVM (Parallel Virtual Machine) programming environment was applied. The code system was applied to analyze an out-of-pile sodium experiment simulating core with 7 subassemblies under transient condition for code verification. It was confirmed that the analytical results show a similar tendency of experimental results. (author)
International Nuclear Information System (INIS)
Ondrej Slezak; Milan Kalal; Hon Jin Kong
2010-01-01
Complete text of publication follows. Analytical description of an experimentally verified scheme leading to a phase-locked stimulated Brillouin scattering (SBS), used in a laser beam combination systems, is presented. The essential condition for the phase-locking effect for SBS is the fixation of the starting position and time of the acoustic Brillouin wave. It is shown that the starting position fixation of this acoustic wave may have its origin in a transient acoustic standing wave initiated by an arising optical interference field produced by the back-seeding concave mirror. This interference field leads to a stationary density modulation of the medium. However, the way to the formation of this density modulation leads via the acoustic standing wave. An appropriate solution, in the form of the standing wave, was obtained from solving the acoustic wave-equation using the electrostriction as a driving force. As a consequence of the damping term included in this equation the acoustic standing wave becomes gradually attenuated and contrary to the undamped solution published earlier, thus constitutes a truly transient phenomenon. Using a mathematical formalism similar to that which is used for the SBS description in the case of a random phase, the coupled equations describing the phase-locked SBS were derived. Contrary to the case without the back-seeding mirror, where the wave chosen from the thermal noise background subsequently plays the role of a trigger of the stimulated process, in this case it is replaced by the transient standing wave produced as a consequence of the presence of an optical interference field arisen in the focal region of the back-seeding concave mirror.
Energy Technology Data Exchange (ETDEWEB)
Jaeger, Wadim; Manes, Jorge Perez; Imke, Uwe; Escalante, Javier Jimenez; Espinoza, Victor Sanchez, E-mail: victor.sanchez@kit.edu
2013-10-15
Highlights: • Simulation of BFBT turbine and pump transients at multiple scales. • CFD, sub-channel and system codes are used for the comparative study. • Heat transfer models are compared to identify difference between the code predictions. • All three scales predict results in good agreement to experiment. • Sub cooled boiling models are identified as field for future research. -- Abstract: The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in the validation and qualification of modern thermo hydraulic simulations tools at various scales. In the present paper, the prediction capabilities of four codes from three different scales – NEPTUNE{sub C}FD as fine mesh computational fluid dynamics code, SUBCHANFLOW and COBRA-TF as sub channels codes and TRACE as system code – are assessed with respect to their two-phase flow modeling capabilities. The subject of the investigations is the well-known and widely used data base provided within the NUPEC BFBT benchmark related to BWRs. Void fraction measurements simulating a turbine and a re-circulation pump trip are provided at several axial levels of the bundle. The prediction capabilities of the codes for transient conditions with various combinations of boundary conditions are validated by comparing the code predictions with the experimental data. In addition, the physical models of the different codes are described and compared to each other in order to explain the different results and to identify areas for further improvements.
Mitigation of thermal transients by tube bundle inlet plenum design
International Nuclear Information System (INIS)
Oras, J.J.; Kasza, K.E.
1984-06-01
A multiphase program aimed at investigating the importance of thermal buoyancy to LMFBR steam-generator and heat-exchanger thermal hydraulics under low-flow transient conditions is being conducted in the Argonne Mixing Components Test Facility (MCTF) on a 60 0 sector shell-side flow model of the Westinghouse straight-tube steam generator being developed under the US/DOE large-component development program. A series of shell-side constant-flow thermal-downramp transient tests have been conducted focusing on the phenomenon of thermal-buoyancy-induced-flow channeling. In addition, it was discovered that a shell-inlet flow-distribution plenum can play a significant role in mitigating the severity of a thermal transient entering a steam generator or heat exchanger
International Nuclear Information System (INIS)
Hall, P.; Hutt, P.
1994-01-01
This paper describes Nuclear Electric's (NE) development of an integrated code package in support of all its reactors including Sizewell B, designed for the provision of fuel management design, core performance studies, operational support and fault transient analysis. The package uses the NE general purpose three-dimensional transient reactor physics code PANTHER with cross-sections derived in the PWR case from the LWRWIMS LWR lattice neutronics code. The package also includes ENIGMA a generic fuel performance code and for PWR application VIPRE-01 a subchannel thermal hydraulics code, RELAP5 the system thermal hydraulics transient code and SCORPIO an on-line surveillance system. The paper describes the capabilities and validation of the elements of this package for PWR, how they are coupled within the package and the way in which they are being applied for Sizewell B to on-line surveillance and fault transient analysis. (Author)
Energy Technology Data Exchange (ETDEWEB)
Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.
1998-03-01
This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.
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
SCANAIR: A transient fuel performance code
International Nuclear Information System (INIS)
Moal, Alain; Georgenthum, Vincent; Marchand, Olivier
2014-01-01
Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under
SCANAIR: A transient fuel performance code
Energy Technology Data Exchange (ETDEWEB)
Moal, Alain, E-mail: alain.moal@irsn.fr; Georgenthum, Vincent; Marchand, Olivier
2014-12-15
Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under
Analysis of loss of normal feedwater transient using RELAP5/MOD1/NSC; KNU1 plant simulation
International Nuclear Information System (INIS)
Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo
1986-01-01
Simulation of the system thermal-hydraulic parameters was carried out following the KNU1(Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on november 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS(Reactor Coolant System) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018. (Author)
Transient analysis of multicavity klystrons
International Nuclear Information System (INIS)
Lavine, T.L.; Miller, R.H.; Morton, P.L.; Ruth, R.D.
1988-09-01
We describe a model for analytic analysis of transients in multicavity klystron output power and phase. Cavities are modeled as resonant circuits, while bunching of the beam is modeled using linear space-charge wave theory. Our analysis has been implemented in a computer program which we use in designing multicavity klystrons with stable output power and phase. We present as examples transient analysis of a relativistic klystron using a magnetic pulse compression modulator, and of a conventional klystron designed to use phase shifting techniques for RF pulse compression. 4 refs., 4 figs
Transient cooling of electronics using phase change material (PCM)-based heat sinks
International Nuclear Information System (INIS)
Kandasamy, Ravi; Wang Xiangqi; Mujumdar, Arun S.
2008-01-01
Use of a phase change material (PCM)-based heat sink in transient thermal management of plastic quad flat package (QFP) electronic devices was investigated experimentally and numerically. Results show that increased power inputs enhance the melting rate as well as the thermal performance of the PCM-based heat sinks until the PCM is fully melted. A three-dimensional computational fluid dynamics model was proposed to simulate the problem and demonstrated good agreement with experimental data. Results indicate the potential for PCM-based heat sinks for use in intermittent-use devices
A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers
Energy Technology Data Exchange (ETDEWEB)
Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1998-12-31
A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)
A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers
Energy Technology Data Exchange (ETDEWEB)
Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1997-12-31
A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)
Dynamics of gas-phase transient species studied by dissociative photodetachment of molecular anions
Lu, Zhou
2007-01-01
Gas-phase transient species, such as the CH₃CO₂ and HOCO free radicals, play important roles in combustion and environment chemistry. In this thesis work, the dynamics of these two radicals were studied by dissociative photodetachment (DPD) of the negative ions, CH₃CO₂-С and HOCO⁻, respectively. The experiments were carried out with a fast-ion-beam photoelectron-photofragment coincidence (PPC) spectrometer. Mass-selected molecular anions in a fast ion beam were intercepted by a linearly polar...
OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the First Workshop (V1000-CT1)
International Nuclear Information System (INIS)
2003-01-01
The first workshop for the VVER-1000 Coolant Transient Benchmark TT Benchmark was hosted by the Commissariat a l'Energie Atomique, Centre d'Etudes de Saclay, France. The V1000CT benchmark defines standard problems for validation of coupled three-dimensional (3-D) neutron-kinetics/system thermal-hydraulics codes for application to Soviet-designed VVER-1000 reactors using actual plant data without any scaling. The overall objective is to access computer codes used in the safety analysis of VVER power plants, specifically for their use in reactivity transient simulations in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 - simulation of the switching on of one main coolant pump (MCP) while the other three MCP are in operation, and V1000CT- 2 - calculation of coolant mixing tests and Main Steam Line Break (MSLB) scenario. Further background information on this benchmark can be found at the OECD/NEA benchmark web site . The purpose of the first workshop was to review the benchmark activities after the Starter Meeting held last year in Dresden, Germany: to discuss the participants' feedback and modifications introduced in the Benchmark Specifications on Phase 1; to present and to discuss modelling issues and preliminary results from the three exercises of Phase 1; to discuss the modelling issues of Exercise 1 of Phase 2; and to define work plan and schedule in order to complete the two phases
Rossi-Pisa, P.
1978-01-01
This paper describes a laboratory metbod used to determine both the soil moisture retention curve and the unsaturated hydraulic conductivity in soil columns under transient flow conditions during evaporation.
International Nuclear Information System (INIS)
Kasai, Masao; Niikura, Setsuo; Ueda, Koju
1985-07-01
This report corresponds to Chapter V of Japanese contribution report to IAEA INTOR Workshop, Phase Two A, Part 2. Simulation results are shown for feedback control of plasma position, electromagnetic forces at disruptions, penetration of electric and magnetic fields, and benchmark tests for transient electromagnetics. Design guide lines for feedback control system and database assessments are also reported. (author)
Thermal-hydraulic code qualification: ATHOS2 and data from Bugey 4 and Tricastin 1. Final report
International Nuclear Information System (INIS)
Masiello, P.J.
1983-02-01
Measured data from steam generators at the Bugey 4 and Tricastin 1 nuclear power plants operated by Electricite de France (EdF) have been used in the qualification of the ATHOS2 computer code. ATHOS2 is a three-dimensional, two-phase thermal-hydraulic code for the steady-state and transient analysis of recirculating-type steam generators. Predicted data for circulation ratio and secondary fluid temperature just above the tube sheet have been compared with corresponding data measured by EdF during on-site testing of Westinghouse Model 51A (Bugey 4) and 51M (Tricastin 1) steam generators. Comparative analyses have been performed for steady-state operating conditions at five power levels for each plant installation. The transient capabilities of the ATHOS2 code were examined in the simulation of an open-grid (load reject from 100% power) test conducted at Bugey 4. Results show that predicted data for secondary fluid temperature at eight locations just above the tube sheet are typically within 1.5 0 C of measured data
Views on the future of thermal hydraulic modeling
Energy Technology Data Exchange (ETDEWEB)
Ishii, M. [Purdue Univ., West Lafayette, IN (United States)
1997-07-01
It is essential for the U.S. NRC to sustain the highest level of the thermal-hydraulics and reactor safety research expertise and continuously improve their accident analysis capability. Such expertise should span over four different areas which are strongly related to each other. These are: (1) Reactor Safety Code Development, (2) Two-phase Flow Modeling, (3) Instrumentation and Fundamental Experimental Research, and (4) Separate Effect and Integral Test. The NRC is already considering a new effort in the area of advanced thermal-hydraulics effort. Its success largely depends on the availability of a significantly improved two-phase flow formulation and constitutive relations supported by detailed experimental data. Therefore, it is recommended that the NRC start significant research efforts in the areas of two-phase flow modeling, instrumentation, basic and separate effect experiments which should be pursued systematically and with clearly defined objectives. It is desirable that some international program is developed in this area. This paper is concentrated on those items in the thermal-hydraulic area which eventually determine the quality of future accident analysis codes.
Views on the future of thermal hydraulic modeling
International Nuclear Information System (INIS)
Ishii, M.
1997-01-01
It is essential for the U.S. NRC to sustain the highest level of the thermal-hydraulics and reactor safety research expertise and continuously improve their accident analysis capability. Such expertise should span over four different areas which are strongly related to each other. These are: (1) Reactor Safety Code Development, (2) Two-phase Flow Modeling, (3) Instrumentation and Fundamental Experimental Research, and (4) Separate Effect and Integral Test. The NRC is already considering a new effort in the area of advanced thermal-hydraulics effort. Its success largely depends on the availability of a significantly improved two-phase flow formulation and constitutive relations supported by detailed experimental data. Therefore, it is recommended that the NRC start significant research efforts in the areas of two-phase flow modeling, instrumentation, basic and separate effect experiments which should be pursued systematically and with clearly defined objectives. It is desirable that some international program is developed in this area. This paper is concentrated on those items in the thermal-hydraulic area which eventually determine the quality of future accident analysis codes
Thermal-hydraulics associated with nuclear education and research
International Nuclear Information System (INIS)
Yokobori, Seiichi
2011-01-01
This article was the rerecording of the author's lecture at the fourth 'Future Energy Forum' (aiming at improving nuclear safety and economics) held in December 2010. The lecture focused on (1) importance of thermal hydraulics associated with nuclear education and research (critical heat flux, two-phase flow and multiphase flow), (2) emerging trend of maintenance engineering (fluid induced vibration, flow accelerated corrosion and stress corrosion cracks), (3) fostering sensible nuclear engineer with common engineering sense, (4) balanced curriculum of basics and advanced research, (5) computerized simulation and fluid mechanics, (6) crucial point of thermo hydraulics education (viscosity, flux, steam and power generation), (7) safety education and human resources development (indispensable technologies such as defence in depth) and (8) topics of thermo hydraulics research (vortices of curbed pipes and visualization of two-phase flow). (T. Tanaka)
Energy Technology Data Exchange (ETDEWEB)
Veloso, Marcelo Antonio
2003-07-01
PANTERA-2 (from Programa para Analise Termo-hidraulica de Reatores a Agua - Program for Thermal-hydraulic Analysis of Water Reactors, Version 2), whose fundamentals are described in this work, is intended to carry out rod bundle subchannel analysis in conjunction with multiloop simulation. It solves simultaneously the conservation equations of mass, axial and lateral momentum, and energy for subchannel geometry coupled with the balance equations that describe the fluid flows in any number of coolant loops connected to a pressure vessel containing the rod bundle. As far as subchannel analysis is concerned, the basic computational strategy of PANTERA-2 comes from COBRA codes, but an alternative implicit solution method oriented to the pressure field has been used to solve the finite difference approximations for the balance laws. The results provided by the subchannel model comprise the fluid density, enthalpy, flow rate, and pressure fields in the subchannels. The loop model predicts the individual loop flows, total flow through the pressure vessel, and pump rotational speeds as a function of time subsequent to the failure of any number of the coolant pumps. The flow transients in the loops may initiated by partial, total or sequential loss of electric power to the operating pumps. Transient events caused by either shaft break or rotor locking may also be simulated. The changes in rotational speed of the pumps as a function of rime are determined from a torque balance. Pump dynamic head and hydraulic torque are calculated as a function of rotational speed and volumetric flow from two polar homologous curves supplied to the code in the tabular form. In order to illustrate the analytical capability of PANTERA-2, three sample problems are presented and discussed. Comparisons between calculated and measured results indicate that the program reproduces with a good accuracy experimental data for subchannel exit temperatures and critical heat fluxes in 5x5 rod bundles. It
Whole-core thermal-hydraulic transient code development and verification for LMFBR analysis
International Nuclear Information System (INIS)
Spencer, D.R.
1979-04-01
Predicted performance during both steady state and transient reactor operation determines the steady state operating limits on LMFBRs. Unnecessary conservatism in performance predictions will not contribute to safety, but will restrict the reactor to more conservative, less economical steady state operation. The most general method for reducing analytical conservatism in LMFBR's without compromising safety is to develop, validate and apply more sophisticated computer models to the limiting performance analyses. The purpose of the on-going Natural Circulation Verification Program (NCVP) is to develop and validate computer codes to analyze natural circulation transients in LMFBRs, and thus, replace unnecessary analytical conservatism with demonstrated calculational capability
Numerical method for three dimensional steady-state two-phase flow calculations
International Nuclear Information System (INIS)
Raymond, P.; Toumi, I.
1992-01-01
This paper presents the numerical scheme which was developed for the FLICA-4 computer code to calculate three dimensional steady state two phase flows. This computer code is devoted to steady state and transient thermal hydraulics analysis of nuclear reactor cores 1,3 . The first section briefly describes the FLICA-4 flow modelling. Then in order to introduce the numerical method for steady state computations, some details are given about the implicit numerical scheme based upon an approximate Riemann solver which was developed for calculation of flow transients. The third section deals with the numerical method for steady state computations, which is derived from this previous general scheme and its optimization. We give some numerical results for steady state calculations and comparisons on required CPU time and memory for various meshing and linear system solvers
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
Almomani, Fares
2016-01-01
This study investigates the effect of hydraulic mixing on anaerobic digestion and sludge accumulation in a septic tank. The performance of a septic tank equipped with a hydraulic mixer was compared with that of a similar standard septic tank over a period of 10 months. The study was conducted in two phases: Phase-I--from May to November 2013 (6 months); Phase-II--from January to May 2014 (4 months). Hydraulic mixing effectively reduced the effluent biological oxygen demand (BOD) and total suspended solids, and reduced the sludge accumulation rate in the septic tank. The BOD removal efficiencies during Phase-II were 65% and 75% in the standard septic tank and a septic tank equipped with hydraulic mixer (Smart Digester™), respectively. The effect of hydraulic mixing reduced the rate of sludge accumulation from 0.64 cm/day to 0.27 cm/day, and increased the pump-out interval by a factor of 3.
Test results of the new NSSS thermal-hydraulics program of the KNPEC-2 simulator
International Nuclear Information System (INIS)
Jeong, J. Z.; Kim, K. D.; Lee, M. S.; Hong, J. H.; Lee, Y. K.; Seo, J. S.; Kweon, K. J.; Lee, S. W.
2001-01-01
As a part of the KNPEC-2 Simulator Upgrade Project, KEPRI and KAERI have developed a new NSSS thermal-hydraulics program, which is based on the best-estimate system code, RETRAN. The RETRAN code was originally developed for realistic simulation of thermal-hydraulic transient in power plant systems. The capability of 'real-time simulation' and robustness' should be first developed before being implemented in full-scope simulators. For this purpose, we have modified the RETRAN code by (i) eliminating the correlations' discontinuities between flow regime maps, (ii) simplifying physical correlations, (iii) correcting errors in the original program, and (iv) others. This paper briefly presents the test results fo the new NSSS thermal-hydraulics program
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
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)
Determination of thermal-hydraulic loads on reactor internals in a DBA-situation
International Nuclear Information System (INIS)
Ville Lestinen; Timo Toppila
2005-01-01
Full text of publication follows: According to Finnish regulatory requirements, reactor internals have to stay intact in a design basis accident (DBA) situation, so that control rods can still penetrate into the core. To fulfill this demand some criteria must be followed in periodical in-service inspections. This is the motivation for studying and developing more detailed methods for analysis of thermal-hydraulic loads on reactor internals during the DBA-situation for the Loviisa NPP in Finland. The objective of this research program is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. This paper concentrates mainly on thermal-hydraulic part of the research, but also the mechanical aspects are discussed. Firstly, the paper includes a short literary review of the available methods to analyse the described problem including both thermal-hydraulic and structural analysis parts. Secondly, different possibilities to carry out thermal-hydraulic analyses have been studied. The DBA-case includes complex physical phenomena and therefore modelling is difficult. The accident situation can be for example LLOCA. When the pipe has broken, the pressure decreases and water starts to evaporate, which consumes energy and that way limits the pressure decrease. After some period of time, the system reaches a new equilibrium state. To perform exact thermal-hydraulic analysis also two phase phenomena must be included. Therefore CFD codes are not capable of modelling the DBA situation very well, but the use of CFD codes requires that the effect of two phase flow must be added somehow. One method to calculate two phase phenomena with CFD codes is to use thermal-hydraulic system codes to calculate
International Nuclear Information System (INIS)
Ribando, R.J.
1979-01-01
A comparison is made between computed results and experimental data for single-phase natural convection in an experimental sodium loop. The tests were conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility, an engineering-scale high temperature sodium facility at the Oak Ridge National Laboratory used for thermal-hydraulic testing of simulated LMFBR subassemblies at normal and off-normal operating conditions. Heat generation in the 19 pin assembly during these tests was typical of decay heat levels. Tests were conducted both with zero initial forced flow and with a small initial forced flow. The bypass line was closed in most tests, but open in one. The computer code used to analyze these tests [LONAC (LOw flow and NAtural Convection)] is an ORNL-developed, fast running, one-dimensional, single-phase finite difference model for simulating forced and free convection transients in the THORS loop
International Nuclear Information System (INIS)
Bemansour, A.
2006-01-01
The present work concerns the numerical and experimental study of the transient response of a packed bed latent heat thermal energy storage system. Experiments were carried out to measures the transient temperature distributions inside a cylindrical bed, which is randomly packed with spheres having uniform sizes and encapsulated the paraffin wax as a phase change material (PCM), with air as a working fluid. A two-dimensional separate phases formulation is used to develop a numerical analysis of the transient response of the bed, considering the influence of both axial and radial thermal dispersion. The fluid energy equation was transformed by finite difference approximation and solved by alternating direction implicit scheme, while the PCM energy equation was solved using fully explicit scheme. This analysis can be applied for both charging and recovery modes and a broad range of Reynolds numbers. Measurements of both fluid and PCM temperature were conducted at different axial and radial positions and at different operating parameters. Experimental measurements of temperature distribution compare favorably with the numerical results over a broad range of Reynolds numbers.(Author)
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
Energy Technology Data Exchange (ETDEWEB)
Olive, J
1995-11-01
This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends.
International Nuclear Information System (INIS)
Hossain, K.; Buck, M.; Bernnat, W.; Lohnert, G.
2008-01-01
The institute of nuclear engineering and energy systems (IKE), Univ. of Stuttgart (Germany)) has developed a new thermal hydraulic tool which can be used for three-dimensional thermal hydraulic analysis of pebble bed as well as block type HTRs. During nominal operation, the flow inside the gas-cooled High Temperature Reactor is essentially single-phase, impressible, and non-isothermal. So, at least one gas phase has to be considered beside the solid phase for thermal hydraulic analysis of HTRs. Each phase (e.g. solid, gas) is considered as a continuum which occupies only its respective fraction of. the control volume. Thermal non-equilibrium is considered between phases and time dependent energy conservation equations for solid and gas phases are solved. Simplified momentum conservation equation for gas obtained from porous media approximation is solved along with the time dependent mass conservation equation. Pro visions for simulating more than one gas component are available in this newly developed code TH3D which could be required for simulating some accident situations (e.g air / water ingress by pipes break). The interaction between phases is made by a set of constitutive equations which re/v on semi-empirical correlations obtained from different experiments. Finite volume method with a staggered grid approach is used for spatial discretization and a fully implicit, time adaptive, multi step method is used for time-dependent discretization. A benchmark calculation which is oriented to the pebble i fuel reactor PBMR-400 and a 3D calculation were presented in HTR -2006 conference and will also be published in Nuclear Engineering and Design (NED) journal. In order to demonstrate the capabilities of TH3D for simulating all block type HTRs. A benchmark calculation which is proposed by IAEA CRP-3 and oriented to the Gas Turbine Modular Helium Reactor (GT-MHR) is performed. calculations are performed for the steady state case (nominal operation) as well as for Loss
Energy Technology Data Exchange (ETDEWEB)
Haemaelaeinen, H.
2005-07-01
As a part of the site investigations for the disposal of spent nuclear fuel, hydraulic conductivity measurements were carried out in borehole OL-KR15 at Eurajoki, Olkiluoto. The objective was to investigate the distribution of the hydraulic conductivity in the surrounding bedrock volume. Measurements were carried out during 2003-2004 in two phases. The total length of the borehole OL-KR15 is 518,85 m and 158 45,14 m. Of the 471 ,5 m + 44,5 m total measurable length 414 m was covered with 237 standard tests with 2 m packer separation as specified in the research plan, partly with 1 m overlaps. 259 tests were initiated, but some of them ended to hardware or software errors or unsuitable parameter values. Double-packer constant-head method was used throughout with nominal 200 kPa overpressure. Injection stage lasted normally 20 minutes and fall-off stage 10 minutes. The tests were often shortened if there were clear indications that the hydraulic conductivity is below the measuring range of the system. The pressure in the test section was let to stabilise at least 5 min before injection. In some test sections the stabilisation or injection stage lasted several hours. Two transient (Horner and 1/Q) interpretations and one stationary-state (Moye) interpretation were made in-situ immediately after the test. The Hydraulic Testing Unit (HTU-system) is owned by Posiva Oy and it was operated by Geopros Oy. (orig.)
International Nuclear Information System (INIS)
Haemaelaeinen, H.
2005-01-01
As a part of the site investigations for the disposal of spent nuclear fuel, hydraulic conductivity measurements were carried out in borehole OL-KR15 at Eurajoki, Olkiluoto. The objective was to investigate the distribution of the hydraulic conductivity in the surrounding bedrock volume. Measurements were carried out during 2003-2004 in two phases. The total length of the borehole OL-KR15 is 518,85 m and 158 45,14 m. Of the 471 ,5 m + 44,5 m total measurable length 414 m was covered with 237 standard tests with 2 m packer separation as specified in the research plan, partly with 1 m overlaps. 259 tests were initiated, but some of them ended to hardware or software errors or unsuitable parameter values. Double-packer constant-head method was used throughout with nominal 200 kPa overpressure. Injection stage lasted normally 20 minutes and fall-off stage 10 minutes. The tests were often shortened if there were clear indications that the hydraulic conductivity is below the measuring range of the system. The pressure in the test section was let to stabilise at least 5 min before injection. In some test sections the stabilisation or injection stage lasted several hours. Two transient (Horner and 1/Q) interpretations and one stationary-state (Moye) interpretation were made in-situ immediately after the test. The Hydraulic Testing Unit (HTU-system) is owned by Posiva Oy and it was operated by Geopros Oy. (orig.)
Energy Technology Data Exchange (ETDEWEB)
NONE
1998-03-01
The paper described the fiscal 1997 result of the fracture hydraulic exploration method as the variation exploration method of geothermal reservoirs. By elucidating hydraulic characteristics of the fracture system forming reservoir, technologies are established which are effective for the reservoir evaluation in early stages of development, maintenance of stable power after operational start-up, and extraction of peripheral reservoirs. As for the pressure transient test method, a test supporting system was basically designed to obtain high accuracy hydraulic parameters. As to the tiltmeter fracture monitoring method, a simulation was made for distribution of active fractures and evaluation of hydraulic constants without drilling wells. In relation to the two-phase flow measuring method, for stable steam production, the use of the orifice plate, the existing flow measuring method, etc. was forecast as a simple measuring method of the two-phase state of reservoir. Concerning the hydrophone VSP method, a feasibility study was made of the practical VSP for high temperature which can analyze hydraulic characteristics and geological structures around the well at the same time which the existing methods were unable to grasp, and brought the results. Moreover, to make high accuracy reservoir modeling possible, Doppler borehole televiewer was made in each reservoir. 80 refs., 147 figs., 22 tabs.
An integrated transient model for simulating the operation of natural gas transport systems
Pambour, Kwabena Addo; Bolado-Lavin, Ricardo; Dijkema, Gerard P. J.
This paper presents an integrated transient hydraulic model that describes the dynamic behavior of natural gas transport systems (GTS). The model includes sub models of the most important facilities comprising a GTS, such as pipelines, compressor stations, pressure reduction stations, underground
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
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.
Study of transitory regimes in hydraulic cooling circuits
International Nuclear Information System (INIS)
Bonnin, Jacques; Fanelli, Michel.
1975-01-01
The problem of transient regimes operated voluntary or not in hydraulic circuits is posed and the risks they cause are shown. As for the case of coolant flow loss the various methods for studying the problem are examined: numerical simulation (explicit and implicit), physical model, on-site testing. The numerical methods that not yet fully satisfying or economic, are still very badly representative for hollow closures. Physical models, expensive in the case of a first facility, are not still fully representative (inconsistent similitudes, difficulties in pump picturing). Site test recordings are often a trouble for exploitation and always limited to nondestructive tests. Comparison between the three methods, already satisfying, will have to be improved to allow remedies to the over pressures due to the transients to be developed [fr
Climate variability and vadose zone controls on damping of transient recharge
Corona, Claudia R.; Gurdak, Jason J.; Dickinson, Jesse; Ferré, T.P.A.; Maurer, Edwin P.
2017-01-01
Increasing demand on groundwater resources motivates understanding of the controls on recharge dynamics so model predictions under current and future climate may improve. Here we address questions about the nonlinear behavior of flux variability in the vadose zone that may explain previously reported teleconnections between global-scale climate variability and fluctuations in groundwater levels. We use hundreds of HYDRUS-1D simulations in a sensitivity analysis approach to evaluate the damping depth of transient recharge over a range of periodic boundary conditions and vadose zone geometries and hydraulic parameters that are representative of aquifer systems of the conterminous United States (U.S). Although the models were parameterized based on U.S. aquifers, findings from this study are applicable elsewhere that have mean recharge rates between 3.65 and 730 mm yr–1. We find that mean infiltration flux, period of time varying infiltration, and hydraulic conductivity are statistically significant predictors of damping depth. The resulting framework explains why some periodic infiltration fluxes associated with climate variability dampen with depth in the vadose zone, resulting in steady-state recharge, while other periodic surface fluxes do not dampen with depth, resulting in transient recharge. We find that transient recharge in response to the climate variability patterns could be detected at the depths of water levels in most U.S. aquifers. Our findings indicate that the damping behavior of transient infiltration fluxes is linear across soil layers for a range of texture combinations. The implications are that relatively simple, homogeneous models of the vadose zone may provide reasonable estimates of the damping depth of climate-varying transient recharge in some complex, layered vadose zone profiles.
Sharma, Meena Kumari; Kazmi, Absar Ahmad
2015-01-01
A laboratory-scale study was carried out to investigate the effects of physical properties of the supporting media and variable hydraulic shock loads on the hydraulic characteristics of an advanced onsite wastewater treatment system. The system consisted of two upflow anaerobic reactors (a septic tank and an anaerobic filter) accommodated within a single unit. The study was divided into three phases on the basis of three different supporting media (Aqwise carriers, corrugated ring and baked clay) used in the anaerobic filter. Hydraulic loadings were based on peak flow factor (PFF), varying from one to six, to simulate the actual conditions during onsite wastewater treatment. Hydraulic characteristics of the system were identified on the basis of residence time distribution analyses. The system showed a very good hydraulic efficiency, between 0.86 and 0.93, with the media of highest porosity at the hydraulic loading of PFF≤4. At the higher hydraulic loading of PFF 6 also, an appreciable hydraulic efficiency of 0.74 was observed. The system also showed good chemical oxygen demand and total suspended solids removal efficiency of 80.5% and 82.3%, respectively at the higher hydraulic loading of PFF 6. Plug-flow dispersion model was found to be the most appropriate one to describe the mixing pattern of the system, with different supporting media at variable loading, during the tracer study.
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
Thermal hydraulics analysis of the Advanced High Temperature Reactor
Energy Technology Data Exchange (ETDEWEB)
Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)
2015-12-01
Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.
Yi, Guodong; Li, Jin
2018-03-01
The master cylinder hydraulic system is the core component of the fineblanking press that seriously affects the machine performance. A key issue in the design of the master cylinder hydraulic system is dealing with the heavy shock loads in the fineblanking process. In this paper, an equivalent model of the master cylinder hydraulic system is established based on typical process parameters for practical fineblanking; then, the response characteristics of the master cylinder slider to the step changes in the load and control current are analyzed, and lastly, control strategies for the proportional valve are studied based on the impact of the control parameters on the kinetic stability of the slider. The results show that the kinetic stability of the slider is significantly affected by the step change of the control current, while it is slightly affected by the step change of the system load, which can be improved by adjusting the flow rate and opening time of the proportional valve.
Modeling multidomain hydraulic properties of shrink-swell soils
Stewart, Ryan D.; Abou Najm, Majdi R.; Rupp, David E.; Selker, John S.
2016-10-01
Shrink-swell soils crack and become compacted as they dry, changing properties such as bulk density and hydraulic conductivity. Multidomain models divide soil into independent realms that allow soil cracks to be incorporated into classical flow and transport models. Incongruously, most applications of multidomain models assume that the porosity distributions, bulk density, and effective saturated hydraulic conductivity of the soil are constant. This study builds on a recently derived soil shrinkage model to develop a new multidomain, dual-permeability model that can accurately predict variations in soil hydraulic properties due to dynamic changes in crack size and connectivity. The model only requires estimates of soil gravimetric water content and a minimal set of parameters, all of which can be determined using laboratory and/or field measurements. We apply the model to eight clayey soils, and demonstrate its ability to quantify variations in volumetric water content (as can be determined during measurement of a soil water characteristic curve) and transient saturated hydraulic conductivity, Ks (as can be measured using infiltration tests). The proposed model is able to capture observed variations in Ks of one to more than two orders of magnitude. In contrast, other dual-permeability models assume that Ks is constant, resulting in the potential for large error when predicting water movement through shrink-swell soils. Overall, the multidomain model presented here successfully quantifies fluctuations in the hydraulic properties of shrink-swell soil matrices, and are suitable for use in physical flow and transport models based on Darcy's Law, the Richards Equation, and the advection-dispersion equation.
Elimination of numerical diffusion in 1 - phase and 2 - phase flows
Energy Technology Data Exchange (ETDEWEB)
Rajamaeki, M. [VTT Energy (Finland)
1997-07-01
The new hydraulics solution method PLIM (Piecewise Linear Interpolation Method) is capable of avoiding the excessive errors, numerical diffusion and also numerical dispersion. The hydraulics solver CFDPLIM uses PLIM and solves the time-dependent one-dimensional flow equations in network geometry. An example is given for 1-phase flow in the case when thermal-hydraulics and reactor kinetics are strongly coupled. Another example concerns oscillations in 2-phase flow. Both the example computations are not possible with conventional methods.
Elimination of numerical diffusion in 1 - phase and 2 - phase flows
International Nuclear Information System (INIS)
Rajamaeki, M.
1997-01-01
The new hydraulics solution method PLIM (Piecewise Linear Interpolation Method) is capable of avoiding the excessive errors, numerical diffusion and also numerical dispersion. The hydraulics solver CFDPLIM uses PLIM and solves the time-dependent one-dimensional flow equations in network geometry. An example is given for 1-phase flow in the case when thermal-hydraulics and reactor kinetics are strongly coupled. Another example concerns oscillations in 2-phase flow. Both the example computations are not possible with conventional methods
Hydraulic Transients Caused by Air Expulsion During Rapid Filling of Undulating Pipelines
Directory of Open Access Journals (Sweden)
Ciro Apollonio
2016-01-01
Full Text Available One of the main issues arising during the rapid filling of a pipeline is the pressure transient which originates after the entrapped air has been expelled at the air release valve. Because of the difference in density between water and air, a pressure transient originates at the impact of the water column. Many authors have analyzed the problem, both from the theoretical and the experimental standpoint. Nevertheless, mainly vertical or horizontal pipelines have been analyzed, whereas in real field applications, the pipe profile is a sequence of ascending and descending pipes, with air release/vacuum valves at high points. To overcome lack of knowledge regarding this latter case, laboratory experiments were carried out to simulate the filling of an undulating pipeline, initially empty at atmospheric pressure. The pipe profile has a high point where an orifice is installed for air venting, so as to simulate the air release valve at intermediate high point of a supply pipeline. In the experiments, the diameter of the orifice and the opening degree of both upstream and downstream valves were varied, in order to analyze their effect on the pressure transient. The experiments were also carried out with a longer descending pipe, in order to assess the effects on the pressure surge of the air volume downstream of the orifice.
Thermodynamic and kinetic simulation of transient liquid-phase bonding
Lindner, Brad
The use of numeric computational methods for the simulation of materials systems is becoming more prevalent and an understanding of these tools may soon be a necessity for Materials Engineers and Scientists. The applicability of numerical simulation methods to transient liquid-phase (TLP) bonding is evaluated using a type 316L/MBF-51 material system. The comparisons involve the calculation of bulk diffusivities, tracking of interface positions during dissolution, widening, and isothermal solidification stages, as well as comparison of elemental composition profiles. The simulations were performed with Thermo-Calc and DICTRA software packages and the experiments with differential scanning calorimetry (DSC), scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and optical microscopic methods. Analytical methods are also discussed to enhance understanding. The results of the investigation show that while general agreement between simulations and experiments can be obtained, assumptions made with the simulation programs may cause difficulty in interpretation of the results unless the user has sufficient, mathematical, thermodynamic, kinetic, and simulation background.
International Nuclear Information System (INIS)
Lahm, C.E.; Koenig, J.F.; Seidel, B.R.
1990-01-01
Metallic fuel elements irradiated in EBR-II at temperatures significantly higher than design, causing liquid phase attack of the cladding, were subsequently irradiated at normal operating temperatures to first breach. The fuel element lifetime was compared to that for elements not subjected to the over-temperature transient and found to be equivalent. 1 ref., 3 figs
International Nuclear Information System (INIS)
Guillermin, Roger
1973-01-01
After some generalities on desalination by distillation processes and on problems due to scaling, this research thesis aims at a better understanding of the formation of deposits and their effects in the case of precipitation of calcium sulphate hemihydrate which precipitates above 100 C in pure solution and between 100 and 300 C in brine. Deposit growth, influence of impurities, and influence of the deposit on heat exchange and loss of load are the main issues of this research. The author addresses general principles of salt crystallization and the thermodynamic and kinetic aspects of calcium sulphate precipitation, studies the growth kinetics of the calcium sulphate hemihydrate in a transient concentration regime, reports the study of thermal and hydraulic effects of the formation of a hemihydrate deposit, notably by discussing the different phases involved in the covering of a metallic surface
A thermal-hydraulic code for transient analysis in a channel with a rod bundle
International Nuclear Information System (INIS)
Khodjaev, I.D.
1995-01-01
The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident
A thermal-hydraulic code for transient analysis in a channel with a rod bundle
Energy Technology Data Exchange (ETDEWEB)
Khodjaev, I.D. [Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)
1995-09-01
The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.
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.
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
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
International Nuclear Information System (INIS)
Guyot, Maxime
2014-01-01
This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and re-criticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios. During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. In the multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level. In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling. (author) [fr
International Nuclear Information System (INIS)
Fischer, K.; Schall, M.; Wolf, L.
1991-01-01
The major objective of the F2 experiment was to investigate the thermal-hydraulic long-term phenomena with special emphasis on natural convection phenomena in a loop-type geometry affected by variations of steam and air injections at different locations as well as dry energy supply into various compartments. The open post-test exercise is being performed in two consecutive phases, with Phase I covering the initial long-term heat-up phase. The exercise received widespread international attention with nine organizations from six European countries participating with seven different computer codes (FUMO, Jericho2, Fiploc, Wavco, Contain, Melcor, Cobra/Fathoms). These codes cover a broad spectrum of presently known European computational tools in severe accident containment analyses. The participants used either the specified mass flow or pressure control boundary conditions. Some exercised their codes for both. In total, 14 different computations were officially provided by the participants indicating strong interests and cooperative efforts by various institutions
Acceptance test report for the Westinghouse 100 ton hydraulic trailer
International Nuclear Information System (INIS)
Barrett, R.A.
1995-01-01
The SY-101 Equipment Removal System 100 Ton Hydraulic Trailer was designed and built by KAMP Systems, Inc. Performance of the Acceptance Test Procedure at KAMP's facility in Ontario, California (termed Phase 1 in this report) was interrupted by discrepancies noted with the main hydraulic cylinder. The main cylinder was removed and sent to REMCO for repair while the trailer was sent to Lampson's facility in Pasco, Washington. The Acceptance Test Procedure was modified and performance resumed at Lampson (termed Phase 2 in this report) after receipt of the repaired cylinder. At the successful conclusion of Phase 2 testing the trailer was accepted as meeting all the performance criteria specified
A Computational Model of Hydraulic Volume Displacement Drive
Directory of Open Access Journals (Sweden)
V. N. Pil'gunov
2014-01-01
Full Text Available The paper offers a computational model of industrial-purpose hydraulic drive with two hydraulic volume adjustable working chamber machines (pump and motor. Adjustable pump equipped with the pressure control unit can be run together with several adjustable hydraulic motors on the principle of three-phase hydraulic socket-outlet with high-pressure lines, drain, and drainage system. The paper considers the pressure-controlled hydrostatic transmission with hydraulic motor as an output link. It shows a possibility to create a saving hydraulic drive using a functional tie between the adjusting parameters of the pump and hydraulic motor through the pressure difference, torque, and angular rate of the hydraulic motor shaft rotation. The programmable logic controller can implement such tie. The Coulomb and viscous frictions are taken into consideration when developing a computational model of the hydraulic volume displacement drive. Discharge balance considers external and internal leakages in equivalent clearances of hydraulic machines, as well as compression loss volume caused by hydraulic fluid compressibility and deformation of pipe walls. To correct dynamic properties of hydraulic drive, the paper offers that in discharge balance are included the additional regulated external leakages in the open circuit of hydraulic drive and regulated internal leakages in the closed-loop circuit. Generalized differential equations having functional multipliers and multilinked nature have been obtained to describe the operation of hydraulic positioning and speed drive with two hydraulic volume adjustable working chamber machines. It is shown that a proposed computational model of hydraulic drive can be taken into consideration in development of LS («Load-Sensing» drives, in which the pumping pressure is tuned to the value required for the most loaded slave motor to overcome the load. Results attained can be used both in designing the industrial-purpose heavy
FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results
Energy Technology Data Exchange (ETDEWEB)
Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.
1997-07-01
The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime.
FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results
International Nuclear Information System (INIS)
Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.
1997-07-01
The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime
The three-dimensional transient two-phase flow computer programme BACCHUS-3D/TP
International Nuclear Information System (INIS)
Bottoni, M.; Dorr, B.; Homann, C.
1992-04-01
The three-dimensional single-phase flow version of the BACCHUS code, which describes the thermal behaviour of a coolant in hexagonal bundle geometry, developed earlier, provided the basis for the development of the two-phase flow version documented in this report. A detailed description is given of the two-phase Slip Model (SM), and of the Homogeneous Equilibrium Model (HEM) as a subcase, which presents several improvements from both viewpoints of physical modelling and numerical treatment, with respect to usual models found in the literature. The most advanced Separated Phases Model (SPM) is then described in all analytical details necessary to fully understand its implementation in the code. Poblems related to the link between the two above models into an integrated code version are then discussed. The code provides an additional option for modelling of active or passive, permeable or impermeable blockages. This option is documented separately. New numerical methods for solving the algebraic systems of equations derived from the linearization of the fundamental equations have completely superseded previous ones and are explained in detail. Eventually a section is dedicated to an overview of the code verification, made over several years, which goes from steady state single-phase unheated bundle experiments up to fast transient two-phase flow experiments in electrically heated 37-pin bundles. (orig.) [de
International Nuclear Information System (INIS)
Shin, Y.W.; Wiedermann, A.H.
1979-10-01
A solution method is presented for transient, homogeneous, equilibrium, two-phase flows of a single-component fluid in one space dimension. The method combines a direct finite-difference procedure and the method of characteristics. The finite-difference procedure solves the interior points of the computing domain; the boundary information is provided by a separate procedure based on the characteristics theory. The solution procedure for boundary points requires information in addition to the physical boundary conditions. This additional information is obtained by a new procedure involving integration of characteristics in the hodograph plane. Sample problems involving various combinations of basic boundary types are calculated for two-phase water/steam mixtures and single-phase nitrogen gas, and compared with independent method-of-characteristics solutions using very fine characteristic mesh. In all cases, excellent agreement is demonstrated
The design concept of the 6-degree-of-freedom hydraulic shaker at ESTEC
Brinkman, P. W.; Kretz, D.
1992-01-01
The European Space Agency (ESA) has decided to extend its test facilities at the European Space and Technology Center (ESTEC) at Noordwijk, The Netherlands, by implementing a 6-degree-of-freedom hydraulic shaker. This shaker will permit vibration testing of large payloads in the frequency range from 0.1 Hz to 100 Hz. Conventional single axis sine and random vibration modes can be applied without the need for a configuration change of the test set-up for vertical and lateral excitations. Transients occurring during launch and/or landing of space vehicles can be accurately simulated in 6-degrees-of-freedom. The performance requirements of the shaker are outlined and the results of the various trade-offs, which are investigated during the initial phase of the design and engineering program are provided. Finally, the resulting baseline concept and the anticipated implementation plan of the new test facility are presented.
Photoisomerization of ethyl ferulate: A solution phase transient absorption study
Horbury, Michael D.; Baker, Lewis A.; Rodrigues, Natércia D. N.; Quan, Wen-Dong; Stavros, Vasilios G.
2017-04-01
Ethyl ferulate (ethyl 4-hydroxy-3-methoxycinnamate) is currently used as a sunscreening agent in commercial sunscreen blends. Recent time-resolved gas-phase measurements have demonstrated that it possesses long-lived (>ns) electronic excited states, counterintuitive to what one might anticipate for an effective sunscreening agent. In the present work, the photodynamics of ethyl ferulate in cyclohexane, are explored using time-resolved transient electronic absorption spectroscopy, upon photoexcitation to the 11ππ∗ and 21ππ∗ states. We demonstrate that ethyl ferulate undergoes efficient non-radiative decay to repopulate the electronic ground state, mediated by trans-cis isomerization. These results strongly suggest that even mild perturbations induced by a non-polar solvent, as may be found in a closer-to-market sunscreen blend, may contribute to our understanding of ethyl ferulate's role as a sunscreening agent.
Energy Technology Data Exchange (ETDEWEB)
López, R., E-mail: ralope1@ing.uc3m.es; Lecuona, A., E-mail: lecuona@ing.uc3m.es; Nogueira, J., E-mail: goriba@ing.uc3m.es; Vereda, C., E-mail: cvereda@ing.uc3m.es
2017-03-15
Highlights: • A two-phase flows numerical algorithm with high order temporal schemes is proposed. • Transient solutions route depends on the temporal high order scheme employed. • ESDIRK scheme for two-phase flows events exhibits high computational performance. • Computational implementation of the ESDIRK scheme can be done in a very easy manner. - Abstract: An extension for 1-D transient two-phase flows of the SIMPLE-ESDIRK method, initially developed for incompressible viscous flows by Ijaz is presented. This extension is motivated by the high temporal order of accuracy demanded to cope with fast phase change events. This methodology is suitable for boiling heat exchangers, solar thermal receivers, etc. The methodology of the solution consist in a finite volume staggered grid discretization of the governing equations in which the transient terms are treated with the explicit first stage singly diagonally implicit Runge-Kutta (ESDIRK) method. It is suitable for stiff differential equations, present in instant boiling or condensation processes. It is combined with the semi-implicit pressure linked equations algorithm (SIMPLE) for the calculation of the pressure field. The case of study consists of the numerical reproduction of the Bartolomei upward boiling pipe flow experiment. The steady-state validation of the numerical algorithm is made against these experimental results and well known numerical results for that experiment. In addition, a detailed study reveals the benefits over the first order Euler Backward method when applying 3rd and 4th order schemes, making emphasis in the behaviour when the system is subjected to periodic square wave wall heat function disturbances, concluding that the use of the ESDIRK method in two-phase calculations presents remarkable accuracy and computational advantages.
International Nuclear Information System (INIS)
Ekrami, A.
2003-01-01
Transient liquid phase diffusion bonding procedure was used to join an ODS Ma 758 superalloy in two conditions, wrought fine grains, and recrystallised grains. An Ni-Cr-B-Si alloy was used as an interlayer. Bonding was carried out at 1100 d ig C for bonding hold times of 15,30, and 60 minutes under vacuum of 6x10 -4 torr. Bonded samples were homogenized at 1360 d ig C for one hour and then cooled with a rate of 15 d ig C /min. Shear and fatigue strengths of bonds were determined. The results showed that there is no effect of bonding hold times on shear strength after bonding hold time of 30 minutes. At a given bonding hold time, the shear strength of bonds made in the recrystallized condition was greater than the shear strength of bonds made in the fine grain condition. The same was true for fatigue strength at a given cycle to fracture. Transient liquid phase bonding was also carried out under pressure of 0.1 Mpa under the same temperature and bonding hold time for fine grain material. Microstructure studies of bonds made under pressure showed no effects of pressure on bond region grain size. Shear tests results also demonstrate little effects of pressure on shear strength of bonds
Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis
Energy Technology Data Exchange (ETDEWEB)
Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu [Korea Atomic Energy Research Institute, T/H Safety Research Team, Yusung, Daejeon (Korea)
2000-10-01
MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)
Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis
International Nuclear Information System (INIS)
Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu
2000-01-01
MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)
Hysteresis, phase transitions, and dangerous transients in electrical power distribution systems.
Duclut, Charlie; Backhaus, Scott; Chertkov, Michael
2013-06-01
The majority of dynamical studies in power systems focus on the high-voltage transmission grids where models consider large generators interacting with crude aggregations of individual small loads. However, new phenomena have been observed indicating that the spatial distribution of collective, nonlinear contribution of these small loads in the low-voltage distribution grid is crucial to the outcome of these dynamical transients. To elucidate the phenomenon, we study the dynamics of voltage and power flows in a spatially extended distribution feeder (circuit) connecting many asynchronous induction motors and discover that this relatively simple 1+1 (space+time) dimensional system exhibits a plethora of nontrivial spatiotemporal effects, some of which may be dangerous for power system stability. Long-range motor-motor interactions mediated by circuit voltage and electrical power flows result in coexistence and segregation of spatially extended phases defined by individual motor states, a "normal" state where the motors' mechanical (rotation) frequency is slightly smaller than the nominal frequency of the basic ac flows and a "stalled" state where the mechanical frequency is small. Transitions between the two states can be initiated by a perturbation of the voltage or base frequency at the head of the distribution feeder. Such behavior is typical of first-order phase transitions in physics, and this 1+1 dimensional model shows many other properties of a first-order phase transition with the spatial distribution of the motors' mechanical frequency playing the role of the order parameter. In particular, we observe (a) propagation of the phase-transition front with the constant speed (in very long feeders) and (b) hysteresis in transitions between the normal and stalled (or partially stalled) phases.
International Nuclear Information System (INIS)
Wulff, W.
1977-01-01
A review is presented on the development of analyses and computer codes for the prediction of thermohydraulic transients in nuclear reactor systems. Models for the dynamics of two-phase mixtures are summarized. Principles of process, reactor component and reactor system modeling are presented, as well as the verification of these models by comparing predicted results with experimental data. Codes of major importance are described, which have recently been developed or are presently under development. The characteristics of these codes are presented in terms of governing equations, solution techniques and code structure. Current efforts and problems of code verification are discussed. A summary is presented of advances which are necessary for reducing the conservatism currently implied in reactor hydraulics codes for safety assessment
Energy Technology Data Exchange (ETDEWEB)
Page, R.; Jones, J.R.
1997-07-01
Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.
International Nuclear Information System (INIS)
Page, R.; Jones, J.R.
1997-01-01
Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell 'B' Loss of offsite power fault transient
International Nuclear Information System (INIS)
Niklaus, F.; Korteniemi, V.
1996-01-01
At the Department of Energy Technology at Lappeenranta University of Technology a CATHARE model of one unit of the St. Petersburg (RBMK) nuclear power plant has been generated. The investigations have been done in order to understand better the thermal-hydraulic behaviour of RBMK type reactors and in order to see how far the French thermal-hydraulic safety code CATHARE can predict the physical phenomena during various RBMK transients. (12 refs.)
Synthetical optimization of hydraulic radius and acoustic field for thermoacoustic cooler
International Nuclear Information System (INIS)
Kang Huifang; Li Qing; Zhou Gang
2009-01-01
It is well known that the acoustic field and the hydraulic radius of the regenerator play key roles in thermoacoustic processes. The optimization of hydraulic radius strongly depends on the acoustic field in the regenerator. This paper investigates the synthetical optimization of hydraulic radius and acoustic field which is characterized by the ratio of the traveling wave component to the standing wave component. In this paper, we discussed the heat flux, cooling power, temperature gradient and coefficient of performance of thermoacoustic cooler with different combinations of hydraulic radiuses and acoustic fields. The calculation results show that, in the cooler's regenerator, due to the acoustic wave, the heat is transferred towards the pressure antinodes in the pure standing wave, while the heat is transferred in the opposite direction of the wave propagation in the pure traveling wave. The better working condition for the regenerator appears in the traveling wave phase region of the like-standing wave, where the directions of the heat transfer by traveling wave component and standing wave component are the same. Otherwise, the small hydraulic radius is not a good choice for acoustic field with excessively high ratio of traveling wave, and the small hydraulic radius is only needed by the traveling wave phase region of like-standing wave.
International Nuclear Information System (INIS)
Pleskunas, R.J.
2015-01-01
In response to the Fukushima Dai-ichi beyond design basis accident in March 2011, the Nuclear Regulatory Commission (NRC) issued Order EA-12-049, 'Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies Beyond-Design-Basis-External-Events'. To outline the process to be used by individual licensees to define and implement site-specific diverse and flexible mitigation strategies (FLEX) that reduce the risks associated with beyond design basis conditions, Nuclear Energy Institute document NEI 12-06, 'Diverse and Flexible Coping Strategies (FLEX) Implementation Guide', was issued. A beyond design basis external event (BDBEE) is postulated to cause an Extended Loss of AC Power (ELAP), which will result in a loss of ventilation which has the potential to impact room habitability and equipment operability. During the ELAP, portable FLEX equipment will be used to achieve and maintain safe shutdown, and only a minimal set of instruments and controls will be available. Given these circumstances, analysis is required to determine the environmental conditions in several vital areas of the Nuclear Power Plant. The BDBEE mitigating strategies require certain room environments to be maintained such that they can support the occupancy of personnel and the functionality of equipment located therein, which is required to support the strategies associated with compliance to NRC Order EA-12-049. Three thermal-hydraulic analyses of vital areas during an extended loss of AC power using the GOTHIC computer code will be presented: 1) Safety-related pump and instrument room transient analysis; 2) Control Room transient analysis; and 3) Auxiliary/Control Building transient analysis. GOTHIC (Generation of Thermal-Hydraulic Information for Containment) is a general purpose thermal-hydraulics software package for the analysis of nuclear power plant containments, confinement buildings, and system components. It is a volume/path/heat sink
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)
Thermal-hydraulic criteria for the APT tungsten neutron source design
International Nuclear Information System (INIS)
Pasamehmetoglu, K.
1998-03-01
This report presents the thermal-hydraulic design criteria (THDC) developed for the tungsten neutron source (TNS). The THDC are developed for the normal operations, operational transients, and design-basis accidents. The requirements of the safety analyses are incorporated into the design criteria, consistent with the integrated safety management and the safety-by-design philosophy implemented throughout the APT design process. The phenomenology limiting the thermal-hydraulic design and the confidence level requirements for each limit are discussed. The overall philosophy of the uncertainty analyses and the confidence level requirements also are presented. Different sets of criteria are developed for normal operations, operational transients, anticipated accidents, unlikely accidents, extremely unlikely accidents, and accidents during TNS replacement. In general, the philosophy is to use the strictest criteria for the high-frequency events. The criteria is relaxed as the event frequencies become smaller. The THDC must be considered as a guide for the design philosophy and not as a hard limit. When achievable, design margins greater than those required by the THDC must be used. However, if a specific event sequence cannot meet the THDC, expensive design changes are not necessary if the single event sequence results in sufficient margin to safety criteria and does not challenge the plant availability or investment protection considerations
Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors
Energy Technology Data Exchange (ETDEWEB)
Baek, W. P.; Song, C. H.; Kim, Y. S. and others
2005-02-15
The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.
Stream hydraulics and temperature determine the metabolism of geothermal Icelandic streams
Directory of Open Access Journals (Sweden)
Demars B. O.L.
2011-07-01
Full Text Available Stream ecosystem metabolism plays a critical role in planetary biogeochemical cycling. Stream benthic habitat complexity and the available surface area for microbes relative to the free-flowing water volume are thought to be important determinants of ecosystem metabolism. Unfortunately, the engineered deepening and straightening of streams for drainage purposes could compromise stream natural services. Stream channel complexity may be quantitatively expressed with hydraulic parameters such as water transient storage, storage residence time, and water spiralling length. The temperature dependence of whole stream ecosystem respiration (ER, gross primary productivity (GPP and net ecosystem production (NEP = GPP − ER has recently been evaluated with a “natural experiment” in Icelandic geothermal streams along a 5–25 °C temperature gradient. There remained, however, a substantial amount of unexplained variability in the statistical models, which may be explained by hydraulic parameters found to be unrelated to temperature. We also specifically tested the additional and predicted synergistic effects of water transient storage and temperature on ER, using novel, more accurate, methods. Both ER and GPP were highly related to water transient storage (or water spiralling length but not to the storage residence time. While there was an additional effect of water transient storage and temperature on ER (r2 = 0.57; P = 0.015, GPP was more related to water transient storage than temperature. The predicted synergistic effect could not be confirmed, most likely due to data limitation. Our interpretation, based on causal statistical modelling, is that the metabolic balance of streams (NEP was primarily determined by the temperature dependence of respiration. Further field and experimental work is required to test the predicted synergistic effect on ER. Meanwhile, since higher metabolic activities allow for higher pollutant degradation or uptake
User effects on the thermal-hydraulic transient system code calculations
International Nuclear Information System (INIS)
Aksan, S.N.; D'Auria, F.; Staedtke, H.
1993-01-01
In the paper, the results of the investigations on the user effects for the thermalhydraulic transient system codes will be presented and discussed on the basis of some case studies. The general findings of the investigations show that in addition to user effects, there are other reasons that affect the results of the calculations and which are hidden under user effects. Both the hidden factors and the direct user effects will be discussed in detail and general recommendations and conclusions will be presented to control and limit them. (orig.)
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)
A Hydraulic Motor-Alternator System for Ocean-Submersible Vehicles
Aintablian, Harry O.; Valdez, Thomas I.; Jones, Jack A.
2012-01-01
An ocean-submersible vehicle has been developed at JPL that moves back and forth between sea level and a depth of a few hundred meters. A liquid volumetric change at a pressure of 70 bars is created by means of thermal phase change. During vehicle ascent, the phase-change material (PCM) is melted by the circulation of warm water and thus pressure is increased. During vehicle descent, the PCM is cooled resulting in reduced pressure. This pressure change is used to generate electric power by means of a hydraulic pump that drives a permanent magnet (PM) alternator. The output energy of the alternator is stored in a rechargeable battery that powers an on-board computer, instrumentation and other peripherals.The focus of this paper is the performance evaluation of a specific hydraulic motor-alternator system. Experimental and theoretical efficiency data of the hydraulic motor and the alternator are presented. The results are used to evaluate the optimization of the hydraulic motor-alternator system. The integrated submersible vehicle was successfully operated in the Pacific Ocean near Hawaii. A brief overview of the actual test results is presented.
International Nuclear Information System (INIS)
D'Auria, F.; Galassi, G.; Spadoni, A.; Hassan, Y.
2001-01-01
The Relap5-3D, the latest in the series of the Relap5 code, distinguishes from the previous versions by the fully integrated, multi-dimensional thermalhydraulic and kinetic modeling capability. It has been applied to Phase I and III of OECD-CSNI/ NSC PWR MSLB Benchmark adopting the same thermalhydraulic input deck already used with Relap5/Parcs and Relap5/Quabbox coupled codes during the previous MSLB analysis. The OECD jointly with the US NRC proposed the PWR MSLB Benchmark in order to gather a common understanding about the coupling between thermal hydraulics and neutronics, and evaluating the behavior of this transient with different coupled codes, giving emphasis to the 3-D modeling. This paper deals with the application of Relap5-3D code to phase I and III of the PWR MSLB Benchmark. The Relap5-3D is a thermal hydraulics-neutronics internally coupled code, the thermal hydraulics module is the INEEL version of Relap and the neutronics module is derived from NESTLE multi-dimension kinetics code. (author)
Comparison of EPRI safety valve test data with analytically determined hydraulic results
International Nuclear Information System (INIS)
Smith, L.C.; Howe, K.S.
1983-01-01
NUREG-0737 (November 1980) and all subsequent U.S. NRC generic follow-up letters require that all operating plant licensees and applicants verify the acceptability of plant specific pressurizer safety valve piping systems for valve operation transients by testing. To aid in this verification process, the Electric Power Research Institute (EPRI) conducted an extensive testing program at the Combustion Engineering Test Facility. Pertinent tests simulating dynamic opening of the safety valves for representative upstream environments were carried out. Different models and sizes of safety valves were tested at the simulated operating conditions. Transducers placed at key points in the system monitored a variety of thermal, hydraulic and structural parameters. From this data, a more complete description of the transient can be made. The EPRI test configuration was analytically modeled using a one-dimensional thermal hydraulic computer program that uses the method of characteristics approach to generate key fluid parameters as a function of space and time. The conservative equations are solved by applying both the implicit and explicit characteristic methods. Unbalanced or wave forces were determined for each straight run of pipe bounded on each side by a turn or elbow. Blowdown forces were included, where appropriate. Several parameters were varied to determine the effects on the pressure, hydraulic forces and timings of events. By comparing these quantities with the experimentally obtained data, an approximate picture of the flow dynamics is arrived at. Two cases in particular are presented. These are the hot and cold loop seal discharge tests made with the Crosby 6M6 spring-loaded safety valve. Included in the paper is a description of the hydraulic code, modeling techniques and assumptions, a comparison of the numerical results with experimental data and a qualitative description of the factors which govern pipe support loading. (orig.)
National Laboratory of Hydraulics. 1996 progress report
International Nuclear Information System (INIS)
1996-01-01
This progress report of the National Laboratory of Hydraulics (LNH) of Electricite de France (EdF) summarizes, first, the research and development studies carried out in 1996 for the development of research tools for industrial fluid mechanics and environmental hydraulics and for the development of computer tools (computer codes and softwares for fluid mechanics modeling, modeling of reactive, compressible, two-phase and turbulent flows and of complex chemical kinetics using finite elements and finite volume methods). A second parts describes the research studies performed for other services of EdF, concerning: the functioning of nuclear reactors (thermohydraulic studies of the reactor vessel and of the primary coolant circuit, gas flows following severe accidents, fluid-structure thermal coupling etc...), fossil fuel power plants, the equipment and operation of thermal power plants and hydraulic power plants, the use of electric power. A third part summarizes the river and marine hydraulic studies carried out for other companies. (J.S.)
Applications for coupled core neutronics and thermal-hydraulic models
International Nuclear Information System (INIS)
Eller, J.
1996-01-01
The unprecedented increases in computing capacity that have occurred during the last decade have affected our sciences, and thus our lives, to an extent that is difficult to overstate. All indications are that this trend will continue for years to come. Nuclear reactor systems analysis is one of many areas of engineering that has changed dramatically as a result of this evolution. Our ability to model the various mechanical and physical systems in greater and greater detail has allowed significant improvements in operational efficiency in spite of increasing regulatory requirements. Many of these efficiencies result from the use of more complex and geometrically detailed computer modeling, which is used to justify a reduction or elimination of some of the conservatisms required by earlier, less sophisticated analyses. And more recently, as our industries open-quotes downsize,close quotes efforts are being made to find ways to use the ever-increasing computing capacity to design systems that accomplish more work, in less time, and with fewer people. The balance of this paper discusses some of the visions that Duke Power Company feels would most benefit their particular methodologies. One of the concepts receiving a lot of attention involves an automated coupling of a thermal-hydraulic plant systems analysis model to a three-dimensional core neutronics program. The thermal-hydraulic analysis of several postulated system transients incorporates large conservatisms because of limited ability to model complex time-dependent asymmetric heat sources in adequate geometric detail. For these transients, the core behavior is closely coupled with the thermal-hydraulic behavior of the total plant system and vice versa. Steam-line break, uncontrolled rod withdrawal, and rod drop anayses are likely to benefit most from this type of linked process
Acceptance test report for the Westinghouse 100 ton hydraulic trailer
Energy Technology Data Exchange (ETDEWEB)
Barrett, R.A.
1995-03-06
The SY-101 Equipment Removal System 100 Ton Hydraulic Trailer was designed and built by KAMP Systems, Inc. Performance of the Acceptance Test Procedure at KAMP`s facility in Ontario, California (termed Phase 1 in this report) was interrupted by discrepancies noted with the main hydraulic cylinder. The main cylinder was removed and sent to REMCO for repair while the trailer was sent to Lampson`s facility in Pasco, Washington. The Acceptance Test Procedure was modified and performance resumed at Lampson (termed Phase 2 in this report) after receipt of the repaired cylinder. At the successful conclusion of Phase 2 testing the trailer was accepted as meeting all the performance criteria specified.
Scaling of Thermal-Hydraulic Phenomena and System Code Assessment
International Nuclear Information System (INIS)
Wolfert, K.
2008-01-01
In the last five decades large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Many separate effects tests and integral system tests were carried out to establish a data base for code development and code validation. In this context the question has to be answered, to what extent the results of down-scaled test facilities represent the thermal-hydraulic behaviour expected in a full-scale nuclear reactor under accidental conditions. Scaling principles, developed by many scientists and engineers, present a scientific technical basis and give a valuable orientation for the design of test facilities. However, it is impossible for a down-scaled facility to reproduce all physical phenomena in the correct temporal sequence and in the kind and strength of their occurrence. The designer needs to optimize a down-scaled facility for the processes of primary interest. This leads compulsorily to scaling distortions of other processes with less importance. Taking into account these weak points, a goal oriented code validation strategy is required, based on the analyses of separate effects tests and integral system tests as well as transients occurred in full-scale nuclear reactors. The CSNI validation matrices are an excellent basis for the fulfilling of this task. Separate effects tests in full scale play here an important role.
International Nuclear Information System (INIS)
Unal, C.; Nelson, R.
1991-01-01
After completion of the thermal-hydraulic model developed in a companion paper, the authors performed developmental assessment calculation of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The overall interfacial drag model predicted reasonable drag coefficients for both the nucleate boiling and the inverted annular flow (IAF) regimes. The predicted pressure drops agreed reasonably well with the measured data of two transient experiments, CCTF Run 14 and a Lehigh reflood test. The thermal-hydraulic model for post-CHF convective heat transfer predicted the rewetting velocities reasonably well for both experiments. The predicted average slope of the wall temperature traces for these tests showed reasonable agreement with the measured data, indicating that the transient-calculated precursory cooling rates agreed with measured data. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. The interfacial heat-transfer model tended to slightly underpredict the vapor temperatures. The maximum difference between calculated and measured vapor temperatures was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall temperatures were in reasonable agreement with measured data with a maximum relative error of less than 13%
Energy Technology Data Exchange (ETDEWEB)
Seo, Jae Kwang; Yoon, J
2003-12-01
The concept of a Multi-cavity Cold Gas PressuriZeR (MCGPZR) is applied to the SMART: The pressurizer system includes in-vessel cavities and out-of-vessel gas cylinders holding the gas supply/vent system. The gas cylinders are connected to the one of the in-vessel cavities via piping with valves. A pressurizer is maintained at a cold temperature of less than about 100 .deg. C, which is realized with coolers installed in and with wet thermal insulators installed on one of the cavities located inside the hot reactor vessel, to minimize the contribution of a steam partial pressure and is filled with nitrogen gas as a pressure-absorbing medium. The working medium and working temperature of the MCGPZR is totally different from that of a hot steam pressurizer of the commercial PWR. In addition, the MCGPZR is intended to be designed to meet a pressure transient during normal power operation (by its gas volume capacity) without using an active control system and during plant heatup/cooldown operation by using an active gas control (filling/venting) system. Therefore in order to evaluate the feasibility of the concept of the MCGPZR and its intended design goal, the thermal hydraulic behaviors and controllability of the MCGPZR during transients especially a heatup/cooldown operation must be analyzed. In this study, a thermal hydraulic transient analysis computer code, PZRTR, for the Reactor Coolant System (RCS) of an integral reactor composed of the MCGPZR, modular Once-Through Steam Generators (OTSGs), a core and a reactor coolant loop is developed. The pressurizer module (MCGPZR module) of the PZRTR code is based on a two-fluid, nonhomogeneous, nonequilibrium model for the two-phase system behavior and the OTSG module is based on a homogeneous equilibrium model of the two-phase flow process. The core module is simply based on the axial power distributions and the reactor coolant loop is based on the temperature distributions. The code is currently dedicated for the
Advantages of Oscillatory Hydraulic Tomography
Kitanidis, P. K.; Bakhos, T.; Cardiff, M. A.; Barrash, W.
2012-12-01
Characterizing the subsurface is significant for most hydrogeologic studies, such as those involving site remediation and groundwater resource explo¬ration. A variety of hydraulic and geophysical methods have been developed to estimate hydraulic conductivity and specific storage. Hydraulic methods based on the analysis of conventional pumping tests allow the estimation of conductivity and storage without need for approximate petrophysical relations, which is an advantage over most geophysical methods that first estimate other properties and then infer values of hydraulic parameters. However, hydraulic methods have the disadvantage that the head-change signal decays with distance from the pumping well and thus becomes difficult to separate from noise except in close proximity to the source. Oscillatory hydraulic tomography (OHT) is an emerging technology to im¬age the subsurface. This method utilizes the idea of imposing sinusoidally varying pressure or discharge signals at several points, collecting head observations at several other points, and then processing these data in a tomographic fashion to estimate conductivity and storage coefficients. After an overview of the methodology, including a description of the most important potential advantages and challenges associated with this approach, two key promising features of the approach will be discussed. First, the signal at an observation point is orthogonal to and thus can be separated from nuisance inputs like head fluctuation from production wells, evapotranspiration, irrigation, and changes in the level of adjacent streams. Second, although the signal amplitude may be weak, one can extract the phase and amplitude of the os¬cillatory signal by collecting measurements over a longer time, thus compensating for the effect of large distance through longer sampling period.
Status and subjects of thermal-hydraulic analysis for next-generation LWRs
International Nuclear Information System (INIS)
2000-03-01
The status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were surveyed through about 5 years until March 1999 by subcommittee on improvement of reactor thermal-hydraulic analysis codes under the nuclear code committee in Japan Atomic Energy Research Institute. Based on the survey results and discussion, the status and subjects on system analysis for various types of proposed reactor were summarized in 1998 and those on multidimensional two-phase flow analysis were also reviewed, since the multidimensional analysis was recognized as one of the most important subjects through the investigation on system analysis. In this report, the status and subjects for the following were summarized from the survey results and discussion in 1998 and 1999; (1) BWR neutronic/thermal-hydraulic coupled analysis, (2) Evaluation of passive safety system performance and (3) Gas-liquid two-phase flow analysis. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs including test results from several large-scale facilities. We expect that the contents can offer a guideline to improve reactor thermal-hydraulic analysis codes in future. (author)
Energy Technology Data Exchange (ETDEWEB)
Baghban, Ghonche [Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of). Nuclear Science and Technology Research Inst.; Shayesteh, Mohsen [Imam Hussein Univ., Tehran (Iran, Islamic Republic of). Dept. of Physics; Bahonar, Majid [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering
2017-03-15
In view of the importance of studying coolant transient behavior in a nuclear reactor, this work is devoted to the thermal-hydraulic analysis of protected and unprotected loss of flow transients in a WWER-1000 reactor. A series of corresponding mathematical and physical models based on the four-equation Drift-Flux model has been applied. Based on a multi-channel approach, the core has been divided into different regions. Each region has different characteristics as represented in a single fuel pin with its associated coolant channel. Appropriate initial and boundary conditions have been considered and two situations of tripping four and two primary pumps in a protected core in addition to situation of tripping all four pumps in an unprotected core have been analyzed. For each transient, a full range of thermal-hydraulic parameters has been obtained. For verification of the proposed model, the results have been compared with those of the RELAP5/MOD3 and Bushehr nuclear power plant Final Safety Analysis Report (FSAR). A good agreement between results has been attained for the aforementioned transients.
International Nuclear Information System (INIS)
Yao, Jianyong; Jiao, Zongxia; Yao, Bin
2014-01-01
High performance robust force control of hydraulic load simulator with constant but unknown hydraulic parameters is considered. In contrast to the linear control based on hydraulic linearization equations, hydraulic inherent nonlinear properties and uncertainties make the conventional feedback proportional-integral-derivative (PID) control not yield to high performance requirements. Furthermore, the hydraulic system may be subjected to non-smooth and discontinuous nonlinearities due to the directional change of valve opening. In this paper, based on a nonlinear system model of hydraulic load simulator, a discontinuous projection-based nonlinear adaptive robust back stepping controller is developed with servo valve dynamics. The proposed controller constructs a novel stable adaptive controller and adaptation laws with additional pressure dynamic related unknown parameters, which can compensate for the system nonlinearities and uncertain parameters, meanwhile a well-designed robust controller is also synthesized to dominate the model uncertainties coming from both parametric uncertainties and uncertain nonlinearities including unmodeled and ignored system dynamics. The controller theoretically guarantee a prescribed transient performance and final tracking accuracy in presence of both parametric uncertainties and uncertain nonlinearities; while achieving asymptotic output tracking in the absence of unstructured uncertainties. The implementation issues are also discussed for controller simplification. Some comparative results are obtained to verify the high-performance nature of the proposed controller.
Energy Technology Data Exchange (ETDEWEB)
Yao, Jianyong [Nanjing University of Science and Technology, Nanjing (China); Jiao, Zongxia [Beihang University, Beijing (China); Yao, Bin [Purdue University, West Lafayette (United States)
2014-04-15
High performance robust force control of hydraulic load simulator with constant but unknown hydraulic parameters is considered. In contrast to the linear control based on hydraulic linearization equations, hydraulic inherent nonlinear properties and uncertainties make the conventional feedback proportional-integral-derivative (PID) control not yield to high performance requirements. Furthermore, the hydraulic system may be subjected to non-smooth and discontinuous nonlinearities due to the directional change of valve opening. In this paper, based on a nonlinear system model of hydraulic load simulator, a discontinuous projection-based nonlinear adaptive robust back stepping controller is developed with servo valve dynamics. The proposed controller constructs a novel stable adaptive controller and adaptation laws with additional pressure dynamic related unknown parameters, which can compensate for the system nonlinearities and uncertain parameters, meanwhile a well-designed robust controller is also synthesized to dominate the model uncertainties coming from both parametric uncertainties and uncertain nonlinearities including unmodeled and ignored system dynamics. The controller theoretically guarantee a prescribed transient performance and final tracking accuracy in presence of both parametric uncertainties and uncertain nonlinearities; while achieving asymptotic output tracking in the absence of unstructured uncertainties. The implementation issues are also discussed for controller simplification. Some comparative results are obtained to verify the high-performance nature of the proposed controller.
International Nuclear Information System (INIS)
Chen Zhao; Chen, Xue-Nong; Rineiski, Andrei; Zhao Pengcheng; Chen Hongli
2014-01-01
Safety analysis is an important tool for justifying the safety of nuclear reactors. The traditional method for nuclear reactor safety analysis is performed by means of system codes, which use one-dimensional lumped-parameter method to model real reactor systems. However, there are many multi-dimensional thermal-hydraulic phenomena cannot be predicated using traditional one-dimensional system codes. This problem is extremely important for pool-type nuclear systems. Computational fluid dynamics (CFD) codes are powerful numerical simulation tools to solve multi-dimensional thermal-hydraulics problems, which are widely used in industrial applications for single phase flows. In order to use general CFD codes to solve nuclear reactor transient problems, some additional models beyond general ones are required. Neutron kinetics model for power calculation and fuel pin model for fuel pin temperature calculation are two important models of these additional models. The motivation of this work is to develop an advance numerical simulation method for nuclear reactor safety analysis by implementing neutron kinetics model and fuel pin model into general CFD codes. In this paper, the Point Kinetics Model (PKM) and Fuel Pin Model (FPM) are implemented into a general CFD code FLUENT. The improved FLUENT was called as FLUENT/PK. The mathematical models and implementary method of FLUENT/PK are descripted and two demonstration application cases, e.g. the unprotected transient overpower (UTOP) accident of a Liquid Metal cooled Fast Reactor (LMFR) and the unprotected beam overpower (UBOP) accident of an Accelerator Driven System (ADS), are presented. (author)
Synchronizing noisy nonidentical oscillators by transient uncoupling
Energy Technology Data Exchange (ETDEWEB)
Tandon, Aditya, E-mail: adityat@iitk.ac.in; Mannattil, Manu, E-mail: mmanu@iitk.ac.in [Department of Physics, Indian Institute of Technology Kanpur, Kanpur, Uttar Pradesh 208016 (India); Schröder, Malte, E-mail: malte@nld.ds.mpg.de [Network Dynamics, Max Planck Institute for Dynamics and Self-Organization (MPIDS), 37077 Göttingen (Germany); Timme, Marc, E-mail: timme@nld.ds.mpg.de [Network Dynamics, Max Planck Institute for Dynamics and Self-Organization (MPIDS), 37077 Göttingen (Germany); Department of Physics, Technical University of Darmstadt, 64289 Darmstadt (Germany); Chakraborty, Sagar, E-mail: sagarc@iitk.ac.in [Department of Physics, Indian Institute of Technology Kanpur, Kanpur, Uttar Pradesh 208016 (India); Mechanics and Applied Mathematics Group, Indian Institute of Technology Kanpur, Kanpur, Uttar Pradesh 208016 (India)
2016-09-15
Synchronization is the process of achieving identical dynamics among coupled identical units. If the units are different from each other, their dynamics cannot become identical; yet, after transients, there may emerge a functional relationship between them—a phenomenon termed “generalized synchronization.” Here, we show that the concept of transient uncoupling, recently introduced for synchronizing identical units, also supports generalized synchronization among nonidentical chaotic units. Generalized synchronization can be achieved by transient uncoupling even when it is impossible by regular coupling. We furthermore demonstrate that transient uncoupling stabilizes synchronization in the presence of common noise. Transient uncoupling works best if the units stay uncoupled whenever the driven orbit visits regions that are locally diverging in its phase space. Thus, to select a favorable uncoupling region, we propose an intuitive method that measures the local divergence at the phase points of the driven unit's trajectory by linearizing the flow and subsequently suppresses the divergence by uncoupling.
Energy Technology Data Exchange (ETDEWEB)
Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Escalante, Javier Jimenez; Espinoza, Victor Sanchez
2015-07-15
Highlights: • Simulation of BFBT-BWR steady-state and transient tests with ATHLET. • Validation of thermal-hydraulic models based on pressure drops and void fraction measurements. • TRACE system code is used for the comparative study. • Predictions result in a good agreement with the experiments. • Discrepancies are smaller or comparable with respect to the measurements uncertainty. - Abstract: Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena. Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes.
Energy Technology Data Exchange (ETDEWEB)
Raschke, E. (BASF AG, Ludwigshafen am Rhein (Germany)); Seelinger, P. (BASF AG, Ludwigshafen am Rhein (Germany)); Sperber, A. (BASF AG, Ludwigshafen am Rhein (Germany)); Strassburger, R. (BASF AG, Ludwigshafen am Rhein (Germany))
1994-05-01
A knowledge of transient flow processes is becoming increasingly important for the reliable design and control of process engineering plant. Transient flow processes occur, of example, in long liquid-carrying pipes: on start-up and shut-down of plant, in emergency shut-downs and fast closure, i.e. when liquid is rapidly decelerated or accelerated. The consequence of such an event is a hammer effect, i.e. a short, often violent change of pressure placing considerable stress on structures. Such hammer effects are readily calculated by numerical methods in single phase media. Technical devices for prevention of inadmissibly high pressure surges can also be designed by means of simulation calculations. However, hammer effects also occur by sudden condensation of vapours. A number of systems in which condensation hammer effects can occur are considered at the end of this contribution. Two special damping measures are presented. (orig.)
International Nuclear Information System (INIS)
Jung, J. J.; Joo, H. K.; Lee, W. J.; Ji, S. K.; Jung, B. D.
2002-01-01
KAERI has developed the coupled 'system thermal-hydraulics - 3 dimensional reactor kinetics' code, MARS/MASTER since 1998. However, there is a limitation in the existing MARS/MASTER code; that is, to perform the coupled calculations using MARS/MASTER, we have to utilize the hydrodynamic model and the heat structure model of the MARS '3D module'. In some transients, reactor kinetics behavior is strongly multi-dimensional, but core thermal-hydraulic behavior remains in one-dimensional manner. For efficient analysis of such transients, we coupled the MARS 1D module with MASTER. The new feature has been assessed by the 'OECD NEA Main Steam Line Break (MSLB) benchmark exercise III' simulations
Modeling transient streaming potentials in falling-head permeameter tests.
Malama, Bwalya; Revil, André
2014-01-01
We present transient streaming potential data collected during falling-head permeameter tests performed on samples of two sands with different physical and chemical properties. The objective of the work is to estimate hydraulic conductivity (K) and the electrokinetic coupling coefficient (Cl ) of the sand samples. A semi-empirical model based on the falling-head permeameter flow model and electrokinetic coupling is used to analyze the streaming potential data and to estimate K and Cl . The values of K estimated from head data are used to validate the streaming potential method. Estimates of K from streaming potential data closely match those obtained from the associated head data, with less than 10% deviation. The electrokinetic coupling coefficient was estimated from streaming potential vs. (1) time and (2) head data for both sands. The results indicate that, within limits of experimental error, the values of Cl estimated by the two methods are essentially the same. The results of this work demonstrate that a temporal record of the streaming potential response in falling-head permeameter tests can be used to estimate both K and Cl . They further indicate the potential for using transient streaming potential data as a proxy for hydraulic head in hydrogeology applications. © 2013, National Ground Water Association.
Energy Technology Data Exchange (ETDEWEB)
Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.
1998-03-01
This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.
Analysis of an controller design for an electro-hydraulic servo pressure regulator
DEFF Research Database (Denmark)
Pedersen, Henrik C.; Andersen, Torben Ole; Madsen, A. M.
2009-01-01
Mobile hydraulics is in a transition phase, where electronic sensors and digital signal processors are starting to become standard on a high number of machines, hereby replacing hydraulic pilot lines and oering new possibilities with regard to both control and feasibility. For controlling some...... of the existing hydraulic components there are, however, still a need for being able to generate a hydraulic pilot pressure, as e.g. almost all open-circuit pumps are hydraulically controlled. The focus of the current paper is therefore on the analysis and controller design an electro-hydraulic servo pressure...... regulator, which generates a hydraulic LS-pressure based on an electrical reference, hereby synergistically integrating knowledge from all parts of the mechatronics area. The servo pressure regulator is used to generate the LS-signal for a variable displacement pump, and the paper rst presents...
CRISSUE-S, Neutronics/Thermal-hydraulics Coupling in LWR Technology
International Nuclear Information System (INIS)
D'Auria, Francesco; Bousbia Salah, Anis; Galassi, G.M.; Vedovi, Juswald; Van Goethem, Georges; Hadek, Jan; Macek, Jiri; Rindelhardt, Udo; Rohde, Ulrich; Ahnert Iglesias, Carol; Aragones Beltran, Jose Maria; Reventos, Francesc; Cuadra, Arantxa; Gago, Jose Luis; Verdu, Gumersindo; Miro, Rafael; Ginestar, Damian; Sanchez, Ana Maria; Sjoberg, Anders; Yitbarek, M.; Sandervag, Oddbjoern; Garis, Ninos; Frid, Wiktor; Panayotov, Dobromir; Ivanov, Kostadin; Uddin, Rizwan; Sartori, Enrico
2004-01-01
expertise and findings and gathering together expert scientists from various areas relevant to the issues addressed. Added value for the CRISSUE-S activity consists of proposing and making available a list of transients to be analysed by coupled neutron kinetics/thermal-hydraulic techniques and of defining ?acceptability? (or required precision) thresholds for the results of the analyses. The list of transients is specific to the different NPP types such as PWR, BWR and VVER. The acceptability thresholds for calculation precision are general in nature and are applicable to all LWRs. The creation of a database including the main results from coupled 3-D neutron kinetics/thermal-hydraulic calculations and their analysis should also be noted. The CRISSUE-S project is organised into three work packages (WPs). The first WP includes activities related to obtaining and documenting relevant data. The second WP is responsible for the state-of-the-art report (SOAR), while the third WP concerns the evaluation of the findings from the SOAR and includes outcomes of the entire project formulated as recommendations, mainly to the nuclear power industry and to the regulatory authorities. The present report is the result of the first WP and discusses the type of transients that are of interest in relation to reactivity-initiated accidents in LWR NPPs and elaborates on the data required for coupled 3-D neutron kinetics/thermal hydraulic analysis and also on data needed to perform associated validations. The reports have been written to accomplish the objectives established in the contract between the EU and its partners. Expected beneficiaries include institutions and organisations involved with nuclear technology (e.g. utilities, regulators, research, fuel industry). In addition, specific expected beneficiaries are junior- or senior-level researchers and technologists working in the considered field of research and development and application of coupled neutron kinetics/thermal-hydraulics
VIPRE-01: A thermal-hydraulic code for reactor cores
International Nuclear Information System (INIS)
Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.; Nomura, K.K.
1989-08-01
The VIPRE-01 thermal hydraulics code for PWR and BWR analysis has undergone significant modifications and error correction. This manual for the updated code, designated as VIPRE-01 Mod-02, describes improvements that eliminate problems of slow convergence with the drift flux model in transient simulation. To update the VIPRE-01 code and its documentation the drift flux model of two-phase flow was implemented and error corrections developed during VIPRE-01 application were included. The project team modified the existing VIPRE-01 equations into drift flux model equations by developing additional terms. They also developed and implemented corrections for the errors identified during the last four years. They then validated the modified code against standard test data using selected test cases. The project team prepared documentation revisions reflecting code improvements and corrections to replace the corresponding sections in the original VIPRE documents. The revised VIPRE code, designated VIPRE-01 Mod-02, incorporates improvements that eliminate many shortcomings of the previous version. During the validation, the code produced satisfactory output compared with test data. The revised documentation is in the form of binder pages to replace existing pages in three of the original manuals
Permeability, storage and hydraulic diffusivity controlled by earthquakes
Brodsky, E. E.; Fulton, P. M.; Xue, L.
2016-12-01
Earthquakes can increase permeability in fractured rocks. In the farfield, such permeability increases are attributed to seismic waves and can last for months after the initial earthquake. Laboratory studies suggest that unclogging of fractures by the transient flow driven by seismic waves is a viable mechanism. These dynamic permeability increases may contribute to permeability enhancement in the seismic clouds accompanying hydraulic fracking. Permeability enhancement by seismic waves could potentially be engineered and the experiments suggest the process will be most effective at a preferred frequency. We have recently observed similar processes inside active fault zones after major earthquakes. A borehole observatory in the fault that generated the M9.0 2011 Tohoku earthquake reveals a sequence of temperature pulses during the secondary aftershock sequence of an M7.3 aftershock. The pulses are attributed to fluid advection by a flow through a zone of transiently increased permeability. Directly after the M7.3 earthquake, the newly damaged fault zone is highly susceptible to further permeability enhancement, but ultimately heals within a month and becomes no longer as sensitive. The observation suggests that the newly damaged fault zone is more prone to fluid pulsing than would be expected based on the long-term permeability structure. Even longer term healing is seen inside the fault zone of the 2008 M7.9 Wenchuan earthquake. The competition between damage and healing (or clogging and unclogging) results in dynamically controlled permeability, storage and hydraulic diffusivity. Recent measurements of in situ fault zone architecture at the 1-10 meter scale suggest that active fault zones often have hydraulic diffusivities near 10-2 m2/s. This uniformity is true even within the damage zone of the San Andreas fault where permeability and storage increases balance each other to achieve this value of diffusivity over a 400 m wide region. We speculate that fault zones
International Nuclear Information System (INIS)
Chen, S.R.; Lin, W.C.; Ferng, Y.M.; Chieng, C.C.; Pei, B.S.
2014-01-01
Highlights: • A 3-D CFD is adopted to simulate transient behaviors in an SFP under the accident. • This model realistically simulates a 17 × 17 bundle, rid of porous media approach. • The loss of external cooling system accident for an SFP is assumed in this paper. • Thermal–hydraulic characteristics in a bundle are strongly influenced by grids. • The results confirm temperature rising rate used in Maanshan NPP is conservative. - Abstract: This paper develops a three-dimensional (3-D) transient computational fluid dynamics (CFD) model to simulate the thermal–hydraulic characteristics in a fuel bundle located in a spent fuel pool (SFP) under the loss of external cooling system accident. The SFP located in the Maanshan nuclear power plant (NPP) is selected herein. Without adopting the porous media approach usually used in the previous CFD works, this model uses a real-geometry simulation of a 17 × 17 fuel bundle, which can obtain the localized distributions of the flow and heat transfer during the accident. These distribution characteristics include several peaks in the axial distributions of flow, pressure, temperature, and Nusselt number (Nu) near the support grids, the non-uniform distribution of secondary flow, and the non-uniform temperature distribution due to flow mixing between rods, etc. According to the conditions adopted in the Procedure 597.1 (MNPP Plant Procedure 597.1, 2010) for the management of the loss-of-cooling event of the spent fuel pool in the Maanshan NPP, the temperature rising rate predicted by the present model can be equivalent to 1.26 K/h, which is the same order as that of 3.5 K/h in the this procedure. This result also confirms that the temperature rising rate used in the Procedure 597.1 for the Maanshan NPP is conservative. In addition, after the loss of external cooling system, there are about 44 h for the operator to repair the malfunctioning system or provide the alternative water source for the pool inventory to
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
DESIGNING HYDRAULIC AIR CHAMBER IN WATER TRANSMISSION SYSTEMS USING GENETIC ALGORITHM
Directory of Open Access Journals (Sweden)
Abdorahim Jamal
2016-09-01
Full Text Available Transient flow control in Water Transmission Systems (WTS is one of the requirements of designing these systems. Hence, among control equipment, air chambers offer the best solution to control transient flow effects, i.e. both prevents water column separation and absorbs pressure increase. It is essential to carry out an accurate and optimized design of air chambers, not only due to high costs of their manufacturing but also their important protective role. Accordingly, hydraulic design parameters comprise tank volume, diameter of nozzle and coefficients of inflow and outflow of nozzle. In this paper, it is intended to optimize these parameters in order to minimize manufacturing costs. On the other hand, maximum and minimum pressures in main pipeline are considered as constraints which shall fall in allowed range. Therefore, a model has been developed which is a combination of a hydraulic simulation model of WTS and an optimization model based on genetic algorithm. This model is first applied to WTS of Dehgolan-Ghorveh plain as a case study. Results of this research demonstrate that based on suggested model, negative wave creation and pressure increase in pipeline is prevented as well as decrease in manufacturing costs of air chamber.
Gas exchange and hydraulics in seedlings of Hevea brasiliensis during water stress and recovery.
Chen, Jun-Wen; Zhang, Qiang; Li, Xiao-Shuang; Cao, Kun-Fang
2010-07-01
The response of plants to drought has received significant attention, but far less attention has been given to the dynamic response of plants during recovery from drought. Photosynthetic performance and hydraulic capacity were monitored in seedlings of Hevea brasiliensis under water stress and during recovery following rewatering. Leaf water relation, gas exchange rate and hydraulic conductivity decreased gradually after water stress fell below a threshold, whereas instantaneous water use efficiency and osmolytes increased significantly. After 5 days of rewatering, leaf water relation, maximum stomatal conductance (g(s-max)) and plant hydraulic conductivity had recovered to the control levels except for sapwood area-specific hydraulic conductivity, photosynthetic assimilation rate and osmolytes. During the phase of water stress, stomata were almost completely closed before water transport efficiency decreased substantially, and moreover, the leaf hydraulic pathway was more vulnerable to water stress-induced embolism than the stem hydraulic pathway. Meanwhile, g(s-max) was linearly correlated with hydraulic capacity when water stress exceeded a threshold. In addition, a positive relationship was shown to occur between the recovery of g(s-max) and of hydraulic capacity during the phase of rewatering. Our results suggest (i) that stomatal closure effectively reduces the risk of xylem dysfunction in water-stressed plants at the cost of gas exchange, (ii) that the leaf functions as a safety valve to protect the hydraulic pathway from water stress-induced dysfunction to a larger extent than does the stem and (iii) that the full drought recovery of gas exchange is restricted by not only hydraulic factors but also non-hydraulic factors.
International Nuclear Information System (INIS)
Ohtsu, Iwao; Murata, Hideo; Kukita, Yutaka; Kumamaru, Hiroshige.
1996-07-01
JAERI, the University of Tokyo, the Central Research Institute of Electric Power Industry and Shimizu Corporation jointing performed and experimental study on two-phase flow in the hydraulically-compensated Compressed Air Energy Storage (CAES) system with a large-diameter vertical pipe two-phase flow test facility from 1993 to 1995. A hydraulically-compensated CAES system is a proposed, conceptual energy storage system where energy is stored in the form of compressed air in an underground cavern which is sealed by a deep (several hundred meters) water shaft. The shaft water head maintains a constant pressure in the cavern, of several mega Pascals, even during loading or unloading of the cavern with air. The dissolved air in the water, however, may create voids in the shaft when the water rises through the shaft during the loading, being forced by the air flow into the cavern. The voids may reduce the effective head of the shaft, and thus the seal may fail, if significant bubbling should occur in the shaft. This bubbling phenomenon (termed 'Champaign effect') and potential failure of the water seal ('blowout') are simulated in a scaled-height, scaled-diameter facility. Carbon dioxide is used to simulate high solubility of air in the full-height, full-pressure system. This report describes the expected and potential two-phase flow phenomena in a hydraulically-compensated CAES system, the test facility and the test procedure, a method to estimate quantities which are not directly measured by using measured quantities and hydrodynamic basic equations, and desirable additional instrumentation. (author)
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.
Phase I Project: Fiber Optic Distributed Acoustic Sensing for Periodic Hydraulic Tests
Energy Technology Data Exchange (ETDEWEB)
Becker, Matthew
2017-12-31
The extraction of heat from hot rock requires circulation of fluid through fracture networks. Because the geometry and connectivity of these fractures determines the efficiency of fluid circulation, many tools are used to characterize fractures before and after development of the reservoir. Under this project, a new tool was developed that allows hydraulic connectivity between geothermal boreholes to be identified. Nanostrain in rock fractures is measured using fiber optic distributed acoustic sensing (DAS). This strain is measured in one borehole in response to periodic pressure pulses induced in another borehole. The strain in the fractures represents hydraulic connectivity between wells. DAS is typically used at frequencies of Hz to kHz, but strain at mHz frequencies were measured for this project. The tool was demonstrated in the laboratory and in the field. In the laboratory, strain in fiber optic cables was measured in response to compression due to oscillating fluid pressure. DAS recorded strains as small as 10 picometer/m in response to 1 cm of water level change. At a fractured crystalline rock field site, strain was measured in boreholes. Fiber-optic cable was mechanically coupled borehole walls using pressured flexible liners. In one borehole 30 m from the oscillating pumping source, pressure and strain were measured simultaneously. The DAS system measured fracture displacement at frequencies of less than 1 mHz (18 min periods) and amplitudes of less than 1 nm, in response to fluid pressure changes of less 20 Pa (2 mm of water). The attenuation and phase shift of the monitored strain signal is indicative of the permeability and storage (compliance) of the fracture network that connects the two wells. The strain response as a function of oscillation frequency is characteristic of the hydraulic structure of the formation. This is the first application of DAS to the measurement of low frequency strain in boreholes. It has enormous potential for monitoring
Multi-level adaptive simulation of transient two-phase flow in heterogeneous porous media
Chueh, C.C.
2010-10-01
An implicit pressure and explicit saturation (IMPES) finite element method (FEM) incorporating a multi-level shock-type adaptive refinement technique is presented and applied to investigate transient two-phase flow in porous media. Local adaptive mesh refinement is implemented seamlessly with state-of-the-art artificial diffusion stabilization allowing simulations that achieve both high resolution and high accuracy. Two benchmark problems, modelling a single crack and a random porous medium, are used to demonstrate the robustness of the method and illustrate the capabilities of the adaptive refinement technique in resolving the saturation field and the complex interaction (transport phenomena) between two fluids in heterogeneous media. © 2010 Elsevier Ltd.
Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code
International Nuclear Information System (INIS)
Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan
2010-01-01
Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.
Energy Technology Data Exchange (ETDEWEB)
Seo, Jae Kwang; Kang, H. O.; Yoon, J.; Kim, K. K
2006-12-15
The concept of a Multi-cavity Cold Gas PressuriZeR(MCGPZR) is applied to the SMART: The pressurizer system includes in-vessel cavities and out-of-vessel gas cylinders holding the gas supply/vent system. The gas cylinders are connected to the one of the in-vessel cavities via piping with valves. A pressurizer is maintained at a cold temperature of less than about 120 .deg. C which is realized with coolers installed in and with wet thermal insulators installed on one of the cavities located inside the hot reactor vessel, to minimize the contribution of a steam partial pressure and is filled with nitrogen gas as a pressure-absorbing medium. The working medium and working temperature of the MCGPZR is totally different from that of a hot steam pressurizer of the commercial PWR. In addition, the MCGPZR is intended to be designed to meet a pressure transient during normal power operation (by its gas volume capacity) without using an active control system and during plant heatup/cooldown operation by using an active gas control (filling/venting) system. Therefore in order to evaluate the feasibility of the concept of the MCGPZR and its intended design goal, the thermal hydraulic behaviors and controllability of the MCGPZR during transients especially a heatup/cooldown operation must be analyzed. In this study, a thermal hydraulic transient analysis computer code, PZRTR rev 1, for the Reactor Coolant System(RCS) of an integral reactor composed of the MCGPZR, modular Once-Through Steam Generators(OTSGs), a core and a reactor coolant loop is developed. The pressurizer module (MCGPZR module) of the PZRTR rev 1 code is based on a two-fluid, nonhomogeneous, nonequilibrium model for the two-phase system behavior and the OTSG module is based on a homogeneous equilibrium model of the two-phase flow process. The core module is simply based on the axial power distributions and the reactor coolant loop is based on the temperature distributions. The code is currently dedicated for the
Thermal-hydraulics Analysis of a Radioisotope-powered Mars Hopper Propulsion System
International Nuclear Information System (INIS)
O'Brien, Robert C.; Klein, Andrew C.; Taitano, William T.; Gibson, Justice; Myers, Brian; Howe, Steven D.
2011-01-01
Thermal-hydraulics analyses results produced using a combined suite of computational design and analysis codes are presented for the preliminary design of a concept Radioisotope Thermal Rocket (RTR) propulsion system. Modeling of the transient heating and steady state temperatures of the system is presented. Simulation results for propellant blow down during impulsive operation are also presented. The results from this study validate the feasibility of a practical thermally capacitive RTR propulsion system.
International Nuclear Information System (INIS)
Carlson, K.E.; Ransom, V.H.; Roth, P.A.
1987-03-01
The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems that may be found in fusion reactors, space reactors, and other advanced systems. As an assessment of current capability the code was applied to a number of physical problems, both conceptual and actual experiments. Results indicate that the numerical solution to the basic conservation equations is technically sound, and that generally good agreement can be obtained when modeling relevant hydrodynamic experiments. The assessment also demonstrates basic fusion system modeling capability and verifies compatibility of the code with both CDC and CRAY mainframes. Areas where improvements could be made include constitutive modeling, which describes the interfacial exchange term. 13 refs., 84 figs
International Nuclear Information System (INIS)
Nikonov, S.; Velkov, K.
2008-01-01
The ATHLET-BIPR-VVER coupled system code is applied for performing of safety analysis for different WWER reactors. During the last years its validation matrix is continuously being enlarged. The measurements performed during the commissioning phase of NPP Kalinin Unit 3 for the transient 'Switching-off of one Main Circulation Pump at nominal power' are very well documented and have a variety of recorded integral and local thermo-hydraulic and neutron-physic parameters including the measurements' errors. This data is being used for further validation of the coupled code system ATHLET-BIPR-VVER. In the paper are discussed the problems and our solutions by the correct interpretation of the measured thermocouples' records at NPP Kalinin-3 and the comparison with the predicted results by the coupled thermal-hydraulic/neutron-kinetic code ATHLET-BIPR-VVER. Of primary importance by such comparisons is the correct accounting of the fluid mixing process that take place in the surrounding of the measuring sensors and also the consideration of the time delay (inertia term) of the measuring devices. On the bases of previous experience and many simulations of the defined transient a method is discussed and proposed to consider correctly the inertia term of the thermocouples' measurements. The new modelling is implemented in the coupled system code ATHLET-BIPR-VVER for further validation. (Author)
Steady and transient regimes in hydropower plants
Gajic, A.
2013-12-01
Hydropower plant that has been in operation for about 30 years has to be reconstructed. They have already installed 12 Kaplan turbines, the largest in the world at that time. The existing CAM relationship was determined based on hydraulic model tests and checked by efficiency on-site tests. It was also tested based on turbine bearing vibrations. In order to discover vibrations and long cracks on stay vanes detailed on-site measurements were performed. Influence of the modification of the trailing edges on the dynamic stresses of the stay vanes is also shown. In order to improve power output transient regimes were analyzed, both experimentally and numerically. Reversible hydropower plant, a pioneer in Europe since it was the first Pump storage power plant constructed with the highest head pump-turbines in the world. Analyses of transient regimes discover some problems with S-shaped characteristics coupled with non-symmetrical penstock.
Steady and transient regimes in hydropower plants
International Nuclear Information System (INIS)
Gajic, A
2013-01-01
Hydropower plant that has been in operation for about 30 years has to be reconstructed. They have already installed 12 Kaplan turbines, the largest in the world at that time. The existing CAM relationship was determined based on hydraulic model tests and checked by efficiency on-site tests. It was also tested based on turbine bearing vibrations. In order to discover vibrations and long cracks on stay vanes detailed on-site measurements were performed. Influence of the modification of the trailing edges on the dynamic stresses of the stay vanes is also shown. In order to improve power output transient regimes were analyzed, both experimentally and numerically. Reversible hydropower plant, a pioneer in Europe since it was the first Pump storage power plant constructed with the highest head pump-turbines in the world. Analyses of transient regimes discover some problems with S-shaped characteristics coupled with non-symmetrical penstock
Neutron and thermo - hydraulic model of a reactivity transient in a nuclear power plant fuel element
International Nuclear Information System (INIS)
Oliva, Jose de Jesus Rivero
2012-01-01
A reactivity transient without reactor scram was modeled and calculated using analytical expressions for the space distributions of the temperature fields, combined with discrete numerical calculations for the time dependences of thermal power and temperatures. The transient analysis covered the time dependencies of reactivity, global thermal power, fuel heat flux and temperatures in fuel, cladding and cooling water. The model was implemented in Microsoft Office Excel, dividing the Excel file in several separated worksheets for input data, initial steady-state calculations, calculation of parameters non-depending on eigenvalues, eigenvalues determination, calculation of parameters depending on eigenvalues, transient calculation and graphical representation of intermediate and final results. The results show how the thermal power reaches a new equilibrium state due to the negative reactivity feedback derived from the fuel temperature increment. Nevertheless, the reactor mean power increases 40% during the first second and, in the hottest channel, the maximum fuel temperature goes to a significantly high value, slightly above 2100 deg C, after 8 seconds of transient. Consequently, the results confirm that certain degree of fuel damage could be expected in case of a reactor scram failure. Once the basic model has being established the scope of accidents for future analyses can be extended, modifying the nuclear power behavior (reactivity) during transient and the boundary conditions for coolant temperature. A more complex model is underway for an annular fuel element. (author)
International Nuclear Information System (INIS)
Spitz, P.B.; Lemoine, F.; Tirini, S.
1997-01-01
IPSN (Nuclear Protection and Safety Institute) and GRS (Gesellschaft fur Anlagen und Reaktorsicherheit Schwertnergasse 1) are developing the ESCADRE-ASTEC systems of codes devoted to the prediction of the behaviour of water-cooled reactors during a severe accident. The RALOC-mod 4 code belongs to this system and is specifically devoted to containment thermal-hydraulics studies. IPSN has designed a Thermal Hydraulic Containment Test Program in support to the Phebus Fission Product Test Program/2/. Evaporation tests have been recently performed in the Phebus containment test facility. The objective of this work is to assess against these tests the capability of the RALOC -mod 4 code to capture the phenomena observed in these experiments and more particularly the evaporation heat transfer and wall heat transfers. (DM)
Self Referencing Heterodyne Transient Grating Spectroscopy with Short Wavelength
Directory of Open Access Journals (Sweden)
Jakob Grilj
2015-04-01
Full Text Available Heterodyning by a phase stable reference electric field is a well known technique to amplify weak nonlinear signals. For short wavelength, the generation of a reference field in front of the sample is challenging because of a lack of suitable beamsplitters. Here, we use a permanent grating which matches the line spacing of the transient grating for the creation of a phase stable reference field. The relative phase among the two can be changed by a relative translation of the permanent and transient gratings in direction orthogonal to the grating lines. We demonstrate the technique for a transient grating on a VO2 thin film and observe constructive as well as destructive interference signals.
Blowdown hydraulic influence on core thermal response in LOFT nuclear experiment L2-3
International Nuclear Information System (INIS)
Reeder, D.L.
1979-01-01
Experimental research into pressurized water reactor (PWR) loss-of-coolant phenomena conducted in the Loss-of-Fluid Test (LOFT) facility has given results indicating that for very large pipe breaks the core thermal response is tightly coupled to the fluid hydraulic phenomena during the blowdown phase of the loss-of-coolant transient. This summary presents and discusses data supporting this conclusion. LOFT Loss-of-Coolant Experiment (LOCE) L2-3 simulated a complete double-ended offset shear break of a primary coolant reactor vessel inlet pipe in a commercial PWR. The LOFT system conditions at experiment initiation were: fuel rod maximum linear heat generation rate (MLHGR) of 39.4 +- 3 kW/m, hot leg temperature of 593 +- 3 K, core ΔT of 32.2 +- 4 K, system pressure of 15.06 +- 0.03 MPa, and flow rate/system volume of 25.6 +- 0.8 kg/m 3 . These conditions are typical of those in commercial PWR systems at normal operating conditions
International Nuclear Information System (INIS)
Biaty, Patricia Andrea Paladino; Sabundjian, Gaiane
2005-01-01
The thermal hydraulic study in accidents and transients analyses in nuclear power plants is realized with some special tools. These programs use the best estimate analyses and have been developed to simulate accidents and transients in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code has been used as tool to licensing the nuclear facilities in our country, which is the objective of this study. The main problem when RELAP5 code is used is a lot of information necessary to simulate thermal hydraulic accidents. Moreover, there is the necessity of a reasonable amount of mathematical operations to calculation of the geometry of the components existents. Therefore, in order to facilitate the manipulation of this information, it is necessary the developing a friendly preprocessor for attainment of the mathematical calculations for RELAP5 code. One of the tools used for some of these calculations is the MS-EXCEL, which will be used in this work. (author)
Energy Technology Data Exchange (ETDEWEB)
Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)
1997-12-31
Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.
International Nuclear Information System (INIS)
Marguet, S.D.
1997-01-01
Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF's neutronic code COCCINELLE uses the Rowland's formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission's products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on 'low' configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.)
Von Euw, Stanislas; Ajili, Widad; Chan-Chang, Tsou-Hsi-Camille; Delices, Annette; Laurent, Guillaume; Babonneau, Florence; Nassif, Nadine; Azaïs, Thierry
2017-09-01
The presence of an amorphous surface layer that coats a crystalline core has been proposed for many biominerals, including bone mineral. In parallel, transient amorphous precursor phases have been proposed in various biomineralization processes, including bone biomineralization. Here we propose a methodology to investigate the origin of these amorphous environments taking the bone tissue as a key example. This study relies on the investigation of a bone tissue sample and its comparison with synthetic calcium phosphate samples, including a stoichiometric apatite, an amorphous calcium phosphate sample, and two different biomimetic apatites. To reveal if the amorphous environments in bone originate from an amorphous surface layer or a transient amorphous precursor phase, a combined solid-state nuclear magnetic resonance (NMR) experiment has been used. The latter consists of a double cross polarization 1 H→ 31 P→ 1 H pulse sequence followed by a 1 H magnetization exchange pulse sequence. The presence of an amorphous surface layer has been investigated through the study of the biomimetic apatites; while the presence of a transient amorphous precursor phase in the form of amorphous calcium phosphate particles has been mimicked with the help of a physical mixture of stoichiometric apatite and amorphous calcium phosphate. The NMR results show that the amorphous and the crystalline environments detected in our bone tissue sample belong to the same particle. The presence of an amorphous surface layer that coats the apatitic core of bone apatite particles has been unambiguously confirmed, and it is certain that this amorphous surface layer has strong implication on bone tissue biogenesis and regeneration. Questions still persist on the structural organization of bone and biomimetic apatites. The existing model proposes a core/shell structure, with an amorphous surface layer coating a crystalline bulk. The accuracy of this model is still debated because amorphous calcium
International Nuclear Information System (INIS)
Mazzini, M.; D'Auria, F.; Vigni, P.
1981-01-01
This paper deals with the state of art of the research performed at the Instituto di Impianti Nucleari of Pisa University, aiming at construction of PIPER-ONE experimental facility. PIPER-ONE program is devoted to acquire direct experience on some basic phenomena, arising in BWR plants subsequently to small breaks, and on the use of the main thermal-hydraulic codes. The research has been planned taking into consideration recent trends of the studies all over the world of small LOCA thermal-hydraulics and particular needs of nuclear safety in Italy. Cost limitations and availability of some components, already installed at the Institute Laboratory, have influenced the design of the loop. The development steps of PIPER-ONE project are presented. Particularly, the overall flowsheet of the apparatus is reported. Some results of preliminary calculation, executed by RELAP4-Mod 6 code concerning both the experimental loop and the reference BWR are shown, too. A comparison with similar facilities in the world closes the paper
DRUFAN-01/MOD2, Transient Thermohydraulics of PWR Primary System LOCA
Energy Technology Data Exchange (ETDEWEB)
Burwell, M J; Lerchl, G; Steinhoff, F; Wolfert, K [Gesellschaft fuer Reaktorsicherheit (GRS) mbH, Forschungsgelaende, 8046 Garching (Germany)
1982-12-13
1 - Description of problem or function: DRUFAN is an advanced best estimate code for simulation of the transient thermal hydraulic behaviour during PWR-blowdown with large break size. 2 - Method of solution: The code is based on the lumped parameter approach and allows flexible control volume configurations. The physical model takes into account thermodynamic nonequilibrium. Using finite difference techniques a 1-dimensional representation of the discharge flow path including geometrical influences is possible. The physical model is based on separated field equations for liquid and vapour mass and overall field equations for energy and momentum. The mass transfer rates between phases during evaporation and condensation are based on correlations for the controlled growth and shrinkage of vapour bubbles or liquid droplets, respectively. A heat conductor model based on the energy transport equation is available for simulation of structures, electrical heater rods and fuel rods. For the heat transfer between solid structures and the fluid a comprehensive package of flow regime dependent heat transfer and critical heat flux correlations can be used. Simulation of components (valve, pressurizer, accumulator, pump, steam generator) is possible with functions or models. Power generation in solid structures may be simulated by an input time function, an electrical heater model or a neutron kinetics models. As a result of the lumped parameter approach a set of ordinary differential equations is obtained from the field equations. These equations, together with those resulting from the simulation of critical discharge flow near the outlet by a finite difference method, are solved by an explicit/implicit integration method with automatic time step, order and error control. The ordinary differential equations representing heat conductors are solved by an essentially implicit integration method. 3 - Restrictions on the complexity of the problem: - Vapour or liquid phase are
Energy Technology Data Exchange (ETDEWEB)
Prasser, H.M. [ed.
1997-12-01
There is hardly another area of physics which has a comparable multiplicity of phenomena, like flow in multi-phase mixtures. The wishes of experimenters regarding measurement technique are correspondingly great: Apart from the conventional parameters of pressure, temperature and speed of flow, as great a collection with resolution of the instantaneous phase distribution is required. Also, the phases themselves frequently consists of several components, whose concentration should also be measured. The enormous progress which has recently been made with laser optics and tomographic processes, must be compared with a long list of unsolved problems, above all where non-contact measurement is concerned. The attempts at solutions are multifarious, the need for the exchange of experience is great and the comparson of measurement processes with one another must be strengthened. The workshop has set itself these targets. (orig.) [Deutsch] Es gibt kaum ein anderes Gebiet der Physik, das eine vergleichbare Vielfalt der Erscheinungen aufweist wie Stroemungen von Mehrphasengemischen. Entsprechend gross sind die Wuensche der Experimentatoren hinsichtlich der Messtechnik: Neben den klassischen Parametern Druck, Temperatur und Stroemungsgeschwindigkeit wird eine moeglichst hoch aufloesende Erfassung der momentanen Phasenverteilung benoetigt. Ausserdem bestehen die Phasen selbst haeufig aus mehreren Komponenten, deren Konzentration ebenfalls gemessen werden soll. Den enormen Fortschritten, ie mit laseroptischen und tomographischen Verfahren in letzter Zeit gemacht wurden, steht nach wie vor eine lange Liste bisher ungeloester Aufgaben gegenueber, vor allen Dingen, wenn beruehrungslos gemessen werden soll. Die Loesungsansaetze sind vielfaeltig, der Bedarf an Erfahrungsaustausch ist gross, der Vergleich der Messverfahren untereinander muss verstaerkt werden. Diesen Zielen hatte sich der Workshop ``Messtechnik fuer tationaere und transiente Mehrphasenstroemungen`` verschrieben.
Analysis of loss of offsite power transient using RELAP5/MOD1/NSC
International Nuclear Information System (INIS)
Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo
1986-01-01
System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAP5/MOD1/NSC), based upon the sequence of events for the KNU1( Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximate the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV setting etc. are recognized to be important in the transient analyses on a bestestimate basis. (Author)
International Nuclear Information System (INIS)
Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio
2012-01-01
The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-sca