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)

State space model extraction of thermohydraulic systems – Part I: A linear graph approach

International Nuclear Information System (INIS)

Uren, K.R.; Schoor, G. van

2013-01-01

Thermohydraulic simulation codes are increasingly making use of graphical design interfaces. The user can quickly and easily design a thermohydraulic system by placing symbols on the screen resembling system components. These components can then be connected to form a system representation. Such system models may then be used to obtain detailed simulations of the physical system. Usually this kind of simulation models are too complex and not ideal for control system design. Therefore, a need exists for automated techniques to extract lumped parameter models useful for control system design. The goal of this first paper, in a two part series, is to propose a method that utilises a graphical representation of a thermohydraulic system, and a lumped parameter modelling approach, to extract state space models. In this methodology each physical domain of the thermohydraulic system is represented by a linear graph. These linear graphs capture the interaction between all components within and across energy domains – hydraulic, thermal and mechanical. These linear graphs are analysed using a graph-theoretic approach to derive reduced order state space models. These models capture the dominant dynamics of the thermohydraulic system and are ideal for control system design purposes. The proposed state space model extraction method is demonstrated by considering a U-tube system. A non-linear state space model is extracted representing both the hydraulic and thermal domain dynamics of the system. The simulated state space model is compared with a Flownex ® model of the U-tube. Flownex ® is a validated systems thermal-fluid simulation software package. - Highlights: • A state space model extraction methodology based on graph-theoretic concepts. • An energy-based approach to consider multi-domain systems in a common framework. • Allow extraction of transparent (white-box) state space models automatically. • Reduced order models containing only independent state

Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC

International Nuclear Information System (INIS)

Gomez Garcia-Torano, I.; Jimenez, G.

2013-01-01

The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

Thermohydraulic and safety analysis on China advanced research reactor under station blackout accident

International Nuclear Information System (INIS)

Tian Wenxi; Qiu Suizheng; Su Guanghui; Jia Dounan; Liu Xingmin; Zhang Jianwei

2007-01-01

A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules

Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

Energy Technology Data Exchange (ETDEWEB)

Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

1985-07-01

This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

A benchmark for coupled thermohydraulics system/three-dimensional neutron kinetics core models

International Nuclear Information System (INIS)

Kliem, S.

1999-01-01

During the last years 3D neutron kinetics core models have been coupled to advanced thermohydraulics system codes. These coupled codes can be used for the analysis of the whole reactor system. Although the stand-alone versions of the 3D neutron kinetics core models and of the thermohydraulics system codes generally have a good verification and validation basis, there is a need for additional validation work. This especially concerns the interaction between the reactor core and the other components of a nuclear power plant (NPP). In the framework of the international 'Atomic Energy Research' (AER) association on VVER Reactor Physics and Reactor Safety, a benchmark for these code systems was defined. (orig.)

Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall

International Nuclear Information System (INIS)

Kim, Byoung-Yoon; Ahn, Hee-Jae

2011-01-01

The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.

Thermohydraulic analysis of pressurized water reactors

International Nuclear Information System (INIS)

Veloso, M.A.

1980-01-01

The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

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)

Development of a model for the primary system CAREM reactor's stationary thermohydraulic calculation

International Nuclear Information System (INIS)

Gaspar, C.; Abbate, P.

1990-01-01

The ESCAREM program oriented to CAREM reactors' stationary thermohydraulic calculation is presented. As CAREM gives variations in relation to models for BWR (Boiling Water Reactors)/PWR (Pressurized Water Reactors) reactors, it was decided to develop a suitable model which allows to calculate: a) if the Steam Generator design is adequate to transfer the power required; b) the circulation flow that occurs in the Primary System; c) the temperature at the entrance (cool branch) and d) the contribution of each component to the pressure drop in the circulation connection. Results were verified against manual calculations and alternative numerical models. An experimental validation at the Thermohydraulic Essays Laboratory is suggested. A parametric analysis series is presented on CAREM 25 reactor, demonstrating operating conditions, at different power levels, as well as the influence of different design aspects. (Author) [es

Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

1998-03-01

The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

The TOPFLOW multi-purpose thermohydraulic test facility

International Nuclear Information System (INIS)

Schaffrath, Andreas; Kruessenberg, A.-K.; Weiss, F.-P.; Prasser, H.-M.

2002-01-01

The TOPFLOW (Transient Two Phase Flow Test Facility) multi-purpose thermohydraulic test facility is being built for studies of steady-state and transient flow phenomena in two-phase flows, and for the development and validation of the models contained in CFD (Computational Fluid Dynamics) codes. The facility is under construction at the Institute for Safety Research of the Rossendorf Research Center (FZR). It will be operated together with the Dresden Technical University and the Zittau/Goerlitz School for Technology, Economics and Social Studies within the framework of the Nuclear Technology Competence Preservation Program. TOPFLOW, with its test sections and its flexible concept, is available as an attractive facility also to users from all European countries. Experiments are planned in these fields, among others: - Transient two-phase flows in vertical and horizontal pipes and pipes of any inclination as well as in geometries typical of nuclear reactors (annulus, hot leg). - Boiling in large vessels and water pools (measurements of steam generation, 3D steam content distribution, turbulence, temperature stratification). - Test of passive components and safety systems. - Condensation in horizontal pipes in the absence and presence of non-condensable gases. The construction phase of TOPFLOW has been completed more or less on schedule. Experiments can be started after a commissioning phase in the 3rd quarter of 2002. (orig.) [de

Parallelization of TWOPORFLOW, a Cartesian Grid based Two-phase Porous Media Code for Transient Thermo-hydraulic Simulations

Science.gov (United States)

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.

Simulation of thermohydraulic phenomena and model test for FBR

International Nuclear Information System (INIS)

Satoh, Kazuziro

1994-01-01

This paper summarizes the major thermohydraulic phenomena of FBRs and the conventional ways of their model tests, and introduces the recent findings regarding measurement technology and computational science. In the future commercial stage of FBRs, the design optimization will becomes important to improve economy and safety more and more. It is indispensable to use computational science to the plant design and safety evaluation. The most of the model tests will be replaced by the simulation analyses based on computational science. The measurement technology using ultrasonic and the numerical simulation with super parallel computing are considered to be the key technology to realize the design by analysis method. (author)

A thermohydraulic analysis for LOCA accident of a CANDU 600 reactor core charged with SEU 43 fuel by means of FIREBIRD code

International Nuclear Information System (INIS)

Serbanel, M.; Catana, A.

2001-01-01

This report presents a comparative analysis of the behaviour of primary circuit during a LOCA 20% RIH accident for two types of reactor core, namely, normally charged, i.e., with clusters of 37 rods and charged with clusters of 43 rods, respectively. This type of accident was chosen since Canadian analyses showed that the associated transient regime stress the fuel elements. The void reactivity as a function of coolant average density was calibrated for a reference regime (LOCA 20% RIH) so that the results of the model be able to reproduce the average distribution in the reference transient regime. The computation makes use of CERBERUS and FIREBIRD codes externally coupled by files. The void reactivity of the hot pencil was obtained this way. An extremely conservative hypothesis was used, namely that the momentary power of the cluster hosting the pencil is the maximal power over the cluster for the corresponding half reactor core. To carry out this work the following steps were covered: 1. The scenario for the LOCA 20% RIH accident was worked out and the input data corresponding to the thermohydraulic and neutronic modules, for the complex model and the 37 rod clusters, were checked; 2. The input data corresponding to the thermohydraulic module for the complex model and the 43 rod cluster were checked; 3. The kinetic parameters corresponding to the 37 rod cluster were computed; 4. The kinetic parameters corresponding to the 43 rod cluster were computed and the file for the input data in the neutronic module was built; 5. A sub-routine for writing files with the thermohydraulic and neutronic quantities, in a format adequate to the other programs, was implemented; 6. The two transient regimes considered were implemented and the archives containing the quantities were built ;7. The results obtained were analyzed. The conclusion of this work is that in case of LOCA 20% RIH accident the 43 bar clusters have a better behaviour than the 37 bar clusters

Several new thermo-hydraulic test facilities in NPIC

International Nuclear Information System (INIS)

Ye Shurong; Sun Yufa; Ji Fuyun; Zong Guifang; Guo Zhongchuan

1997-01-01

Several new thermo-hydraulic test facilities are under construction in Nuclear Power Institute of Chinese (NPIC) at Chengdu. These facilities include: 1. Nuclear Power Component Comprehensive Test Facility. 2. Reactor Hydraulic Modeling Test Facility. 3. Control Rod Drive Line Hydraulic Test Facility. 4. Large Scale Thermo-Hydraulic Test Facility. The construction of these facilities will make huge progress in the research and development capability of nuclear power technology in CHINA. The author will present a brief description of the design parameters flowchart and test program of these facilities

Confirmatory simulation of safety and operational transients in LMFBR systems

International Nuclear Information System (INIS)

Guppy, J.G.; Agrawal, A.K.

1978-01-01

Operational and safety transients that may originate anywhere in an LMFBR system must be adequately simulated to assist in safety evaluation and plant design efforts. This paper describes an advanced thermohydraulic transient code, the Super System Code (SSC), that may be used for confirmatory safety evaluations of plant wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions, and presents results obtained with SSC illustrating the degree of modelling detail present in the code as well as the computing efficiency. (author)

Thermohydraulic relationships for advanced water cooled reactors

International Nuclear Information System (INIS)

2001-04-01

This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

Thermohydraulic relationships for advanced water cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-04-01

This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

Point kinetics improvements to evaluate three-dimensional effects in transients calculation

International Nuclear Information System (INIS)

Castellotti, U.

1987-01-01

A calculation method, which considers the flux axial perturbations in the parameters related to the reactivity within a point kinetics model, is described. The method considered uses axial factors of consideration which act on the thermohydraulic variables included in the reactivity calculation. The PUMA three-dimensional code as reference model for the comparisons, is used. The limitations inherent to the reactivity balance of the point models used in the transients calculation, are given. (Author)

Evaluation of two-fluid and drift flux thermohydraulics in APROS code environment

International Nuclear Information System (INIS)

Miettinen, J.; Karppinen, I.; Haenninen, M.; Ylijoki, J.

1999-01-01

The characteristics of the thermohydraulic solutions in APROS are considered for the nuclear power plant modelling. The thermohydraulic model of the APROS plant analyzer includes three levels of solutions, homogeneous 3-equation model, 5-equation drift flux model and 6-equation two-fluid model. In practical modelling of versatile process systems different approaches are selected for different types of the power plant sections. The 3-equation model is used for turbines and auxiliary systems. The 5-equation model and 6-equation model are alternative models for main process sections of the primary and secondary sides. The 5-equation model has been typically selected for the real time applications and the 6-equation model for the safety analysis applications. The validation needs for both approaches are the same. Because the change of the solution mode is an easy task in APROS, the validation tasks are typically performed in parallel for 5-equation and 6-equation models. By calculating in parallel with both models systematic errors in solutions may be pointed out. The testing against both separate effects tests and integral tests is an essential part in the thermohydraulics. In different plant applications different physical features are important. The analysis requirements vary from one application to another. When nodalizations together with increased computer speed are growing up, the earlier validation cases may be insufficient. That is why the content of the code has to be known in detail. Such an expertise in the code development has to be gained that properties of the code against other thermohydraulics codes are known. (author)

Comparison of the results of the fifth dynamic AER benchmark-a benchmark for coupled thermohydraulic system/three-dimensional hexagonal kinetic core models

International Nuclear Information System (INIS)

Kliem, S.

1998-01-01

The fifth dynamic benchmark was defined at seventh AER-Symposium, held in Hoernitz, Germany in 1997. It is the first benchmark for coupled thermohydraulic system/three-dimensional hexagonal neutron kinetic core models. In this benchmark the interaction between the components of a WWER-440 NPP with the reactor core has been investigated. The initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one control rod group stucking. This break causes an overcooling of the primary circuit. During this overcooling the scram reactivity is compensated and the scrammed reactor becomes re critical. The calculation was continued until the highly-borated water from the high pressure injection system terminated the power excursion. Each participant used own best-estimate nuclear cross section data. Only the initial subcriticality at the beginning of the transient was given. Solutions were received from Kurchatov Institute Russia with the code BIPR8/ATHLET, VTT Energy Finland with HEXTRAN/SMABRE, NRI Rez Czech Republic with DYN3/ATHLET, KFKI Budapest Hungary with KIKO3D/ATHLET and from FZR Germany with the code DYN3D/ATHLET.In this paper the results are compared. Beside the comparison of global results, the behaviour of several thermohydraulic and neutron kinetic parameters is presented to discuss the revealed differences between the solutions.(Authors)

Fine numerical modelling of thermohydraulic phenomena in EDF PWR reactors

International Nuclear Information System (INIS)

Boulot, F.

1993-01-01

Over the last 20 years, EDF has developed a family of 2D and 3D industrial thermohydraulics software to solve problems encountered in existing PWR power plants and to design new reactors for the future. The equations used in the models are the averaged Navier-Stokes and energy equations. A brief description is given of the four main codes developed for single-phase and two-phase water-steam flows, some of which use finite differences or finite volumes methods, while others make use of finite elements methods. An example of application is given for each code. (author). 4 figs., 4 refs

A model development for a thermohydraulic calculation material convection of MTR (Materials Testing Reactors)

International Nuclear Information System (INIS)

Abbate, P.

1990-01-01

The CONVEC program developed for the thermohydraulic calculation under a natural convection regime for MTR type reactors is presented. The program is based on a stationary, one dimensional model of finite differences that allow to calculate the temperatures of cooler, cladding and fuel as well as the flow for a power level specified by the user. This model has been satisfactorily validated by a water cooling (liquid phase) and air system. (Author) [es

Thermohydraulic model of WWER-1000 core

International Nuclear Information System (INIS)

Maroti, L.; Szabados, L.

1987-11-01

Safe and economic operation of the WWER-1000 type reactor requires more accurate calculation of the thermohydraulic processes than the one satisfactory for the 440 type cores. The high degree of accuracy is needed both for reactor physics calculations and for the determination of the operational safety limits of the core. The paper illustrates the most important differences between the 1000 and 440 type reactors and presents the main fields of the development work necessary to reach the required accuracy. A prediction for the capability of the computer programs after the proposed development is also given and some suggestions for the further improvement is outlined. (author) 7 refs

Research activity on thermohydraulic problems of PWR reactors

International Nuclear Information System (INIS)

Szabados, L.

1976-06-01

The general review of the experimental and theoretical research works on thermohydraulic investigation of pressurized water type reactors being done in the Central Research Institute for Physics is given. The main results of the theoretical and theoretical-numerical research are summarized. The most important result of the past years is the construction of the High Pressure Water Cooled Loop (NVH) thermohydraulic loop. Another significant achievement was the development of the reactor thermohydraulic program system. (Sz.N.Z.)

THYDE-P, PWR LOCA Thermohydraulic Transient Analysis

International Nuclear Information System (INIS)

Asahi, Yoshiro

2001-01-01

1 - Description of problem or function: THYDE-P1 analyzes the behaviour of LWR plants in response to various disturbances, including the thermal hydraulic transient following a break of the primary coolant pipe system, generally referred to as a loss-of-coolant-accident (LOCA). LOCA can be considered as the most critical condition for testing the methods and models for plant dynamics, since thermal hydraulic conditions in the system change drastically during the transient. THYDE-P is capable of a complete LOCA calculation from start to complete reflooding of the core by subcooled water. The program performs steady-state adjustment, which is complete in the sense that the steady state obtained is a set of exact solutions of all the transient equations without time derivatives, not only for plant hydraulics but also for all the other phenomena in the simulation of a PWR plant. THYDE-P2 contains among others the following improvements over THYDE-P1: (1) not only the mass and momentum equations but also the energy equation are included in the non-linear implicit scheme; (2) the valve model is implemented; (3) the relaxation equation for void fraction is theoretically derived; (4) vectorized programming is implemented; (5) both EM (evaluation mode) and BE (best estimate) calculations are possible. THYDE-W is an improved version of THYDE-P2 and contains the following additional features: (a) analysis of multiple number of disjoint loops is possible; (b) a control system simulation model is included; (c) the trip model has been improved; (d) heavy water is allowed as coolant; (e) the effect of drift flux is accounted for in the steady state calculation; (f) boron transport is included; (g) to obtain steady state loop heat balance, the option of adjusting the enthalpy distribution is prepared included in addition to that of adjusting heat exchanger areas; (h) to obtain steady state pressure distribution, three other options are prepared in addition to the original ones

ALARM, Thermohydraulics of BWR with Jet Pumps During LOCA

International Nuclear Information System (INIS)

Araya, F.; Akimoto, M.

1985-01-01

1 - Nature of physical problem solved: ALARM-B2 which is an improved version of ALARM-B1 is a computer program to analyze thermo-hydraulic phenomena of BWR during a blowdown period under a large-break loss-of-coolant accident condition with special emphasis on the heat transfer phenomena in the core region. 2 - Method of solution: A so called volume-junction method is used to present fluid conservations. The primary system is divided into a number of special elements called 'control-volumes'. The system of partial differential equations describing fluid conservations for a stream-tube are integrated over a number of control volumes. The resulting set of simultaneous differential equations that is based on the assumptions of one-dimensional, homogeneous and thermal- equilibrium flow is linearized and solved for a small time increment by a simple explicit numerical technique. The one-dimensional heat conduction equations describing temperature profiles within solid material are written in finite difference forms which are linearized and solved by the Crank-Nicholson implicit method. In order to simulate the blowdown heat transfer phenomena, the code has correlation packages for heat transfer coefficient and critical heat flux. The heat generation in the core is given by a point reactor kinetics model with six groups of delayed neutrons and decay of eleven groups of fission products and actinides. The solution technique of the reactor kinetics is based on the Runge-Kutta method. ALARM-B2 has the models to simulate various components incorporated in BWRs such as jet pumps, recirculation pumps, steam separators, valves, and so on. The discharge and injection systems are modeled by leak and fill systems, respectively. 3 - Restrictions on the complexity of the problem: As this has been developed to simulate a blowdown thermo-hydraulic transient during a large break LOCA, users must pay attention when applying the code to any medium or small break LOCAs or to later phases

Development of a computer code for transients simulation in PWR type reactors

International Nuclear Information System (INIS)

Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de

1981-01-01

A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt

Three-dimensional reactor model for the Paks NPP full-scope simulator

International Nuclear Information System (INIS)

Gyori, C.; Hegyi, G.; Hozer, Z.; Kereszturi, A.; Maraczy, C.

1993-01-01

The reactor model includes thermohydraulic and neutron-physical components. The thermohydraulic model is based on the SMABRE code developed at the Technical Research Centre of Finland for the analysis of loss-of-coolant transients in PWRs. The fuel rod model will be replaced by a new software module providing a comprehensive description of the behavior of fuel rods during reactor transients and hypothetical accidents. The calculation is performed in four individual models: fuel rod temperature model, fuel rod internal pressure model, fuel rod deformation model and fuel rod failure model. In the neutron-physical model the core is calculated with nodes for all of the 349 fuel assemblies, and each assembly is calculated in ten layers. (Z.S.) 1 fig., 5 refs

Thermo-hydraulic analysis of the generic equatorial port plug design

International Nuclear Information System (INIS)

Rodríguez, E.; Guirao, J.; Ordieres, J.; Cortizo, J.L.; Iglesias, S.

2012-01-01

Highlights: ► Thermo-hydraulic transient performance evaluation and optimization of the GEPP structure cooling/heating system under neutronic heating and baking conditions. ► The optimization of the GEPP box structure's cooling system includes positioning and minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions. - Abstract: The port-based ITER diagnostic systems are housed primarily in two locations, the equatorial and upper port plugs. The port plug structure provides confinement function, maintains ultra-high vacuum quality and the first confinement barrier for radioactive materials at the ports. The port plug structure design, from the ITER International Organisation (IO), is cooled and heated by pressurized water which flows through a series of gun-drilled water channels and water pipes. The cooling function is required to remove nuclear heating due to radiation during operation of ITER, while the heating function is intended to heat up uniformly the machine during baking condition. The work presented provides coupled thermo-hydraulic analysis and optimization of a Generic Equatorial Port Plug (GEPP) structure cooling and heating system. The optimization performed includes positioning, minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions.

Thermohydraulic modeling and simulation of breeder reactors

International Nuclear Information System (INIS)

Agrawal, A.K.; Khatib-Rahbar, M.; Curtis, R.T.; Hetrick, D.L.; Girijashankar, P.V.

1982-01-01

This paper deals with the modeling and simulation of system-wide transients in LMFBRs. Unprotected events (i.e., the presumption of failure of the plant protection system) leading to core-melt are not considered in this paper. The existing computational capabilities in the area of protected transients in the US are noted. Various physical and numerical approximations that are made in these codes are discussed. Finally, the future direction in the area of model verification and improvements is discussed

Computational fluid dynamic model for thermohydraulic calculation for the steady-state of the real scale HTR-1

Energy Technology Data Exchange (ETDEWEB)

Gamez, Abel; Rojas, Leorlen; Rosales, Jesus; Castro, Landy Y.; Gonzalez, Daniel; Garcia, Carlos, E-mail: agamezgmf@gmail.com, E-mail: leored1984@gmail.com, E-mail: jrosales@instec.cu, E-mail: lcastro@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgr@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Oliveira, Carlos B. de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil)

2015-07-01

The high temperature gas cooled reactor (HTGR) is one of candidates of next generation of nuclear reactor according to IAEA report 2013. Evaluation of thermohydraulic performance and an experimental comparison results were proposed to the international research community. In this article, the tree dimensional CFD thermohydraulic modelation of steady state of HTR-10 modular reactor, using ANSYS CFX v14.0, has been done. Code-to-code and Code-to-experiment benchmark analyses, related to the testing program of the HTR-10 plant including steady state temperature distribution with the reactor at full power, were developed. The 3D real scale representation of reflector zone and fluid path flow inner and outer reflector blocks and cold helium cavity were carried out. The porous medium model was used to simulate the core zone in the reactor. The power distribution of the initial core published by IAEA-TECDOC-1694 obtained by Chief Scientific Investigators (CSIs) from China was used as heat sources in the core zone. (author)

Atucha I nuclear power plant transients analysis

International Nuclear Information System (INIS)

Castano, J.; Schivo, M.

1987-01-01

A program for the transients simulation thermohydraulic calculation without loss of coolant (KWU-ENACE development) to evaluate Atucha I nuclear power plant behaviour is used. The program includes systems simulation and nuclear power plants control bonds with real parameters. The calculation results show a good agreement with the output 'protocol' of various transients of the nuclear power plant, keeping the error, in general, lesser than ± 10% from the variation of the nuclear power plant's state variables. (Author)

Influence of taking into account in-pressurizer convective heat- and mass transfer influence effects at the transients in VVER with code RELAP 5/MOD 3.2

International Nuclear Information System (INIS)

Konovalyuk, L.N.; Shevelev, D.V.; Kravchenko, V.G.

2003-01-01

PRZ model is proposed which allows taking into account in pressurizer convective heat- and mass transfer influence effects at the transients in VVER (PWR) Type Reactors case when calculations performed with using 1D thermohydraulic codes. The theoretical backgrounds are given to define the transients with the convective coolant instability in PRZ. The instability threshold is given for real PRZ geometry

Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC; Desarrollo de un modelo del NSSS de un reactor PWR con el codigo termo-hidraulico GOTHIC

Energy Technology Data Exchange (ETDEWEB)

Gomez Garcia-Torano, I.; Jimenez, G.

2013-07-01

The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

Library thermohydraulic components for training simulators

International Nuclear Information System (INIS)

Castelao Caruana, M. J.; Di Benedetto, A.; Pierini, J.P.

2013-01-01

The thermohydraulic components Library was modeled in MatLab/Simulink®. This library owns Pipe type components (pump, control valve and / or heaters), storage tanks (Open, Closed and Equilibrium Water Vapor-Air) and Heat Exchangers (Co-Current, Counter-Current and U-tubes). Each component can be attached to other components through the component library Header, in order to create a more complex thermal-hydraulic system which in turn can interact with other thermal-hydraulic systems. (author)

Reactivity transient calculatios in research reactor

International Nuclear Information System (INIS)

Santos, R.S. dos

1986-01-01

A digital program for reactivity transient analysis in research reactor and cylindrical geometry was showed quite efficient when compared with methods and programs of the literature, as much in the solution of the neutron kinetics equation as in the thermohydraulic. An improvement in the representation of the feedback reactivity adopted on the program reduced markedly the computation time, with some accuracy. (Author) [pt

Computer simulation of WWER - 440 normal and emergency transient operating conditions

International Nuclear Information System (INIS)

Izbeshesku, M.; Rajka, V.; Untaru, S.; Dumitresku, A.; Paneh, M.; Turku, I.

1976-01-01

Results of computer realization of a model for studying transient process in the nuclear system of vapour production at the WWER - 40 peactor nuclear power plant are presented. The first circuit model consists of a number of modules, corresponding to its main parts: for each module derived were the equations describing neutron and thermohydraulic parameters. The second circuit effect is taken into account by heat quantity accepted from a steam generator. The equations are mostly differential with constant coefficients. Coefficient values and initial values of physical quantities are evaluated according to the technical literature. Both manual and automatic operations are modelled [ru

Thermohydraulic analysis for power increase of IEAR-1 reactor

International Nuclear Information System (INIS)

Umbehaun, Pedro E.; Bastos, Jose L.F.

1996-01-01

In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30

Adaptation and implementation of the TRACE code for transient analysis on designs of cooled lead fast reactors

International Nuclear Information System (INIS)

Lazaro, A.; Ammirabile, L.; Martorell, S.

2014-01-01

The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)

Computation of 3D thermohydraulics in partially blocked bundles during the reflood phase of a LOCA

International Nuclear Information System (INIS)

Cicero, G.M.; Briere, E.; Fornaciari, G.

1994-06-01

In Pressurized Water Reactors (PWR), ballooning of the fuel rod claddings may occur during a LOCA, since the fuel rod claddings are heated up, and the system pressure is low. The severe blockages that may result induce cross-flow diversion and three-dimensional effects on thermohydraulics in the core bundle, during the reflood phase. To improve the knowledge of these phenomena and their physical modelling in the code CATHARE, 3D computer codes are needed. In 1990, EDF has started up a development and validation program of the 3D THYC computer code to analyze the thermohydraulics of the flow during the reflood phase, in partially blocked bundles. The main objective is to calculate the temperatures of the rods above the quench front, when they are cooled by superheated steam with saturated droplets. First, this paper introduces the THYC model developed for reflood studies. Secondly, we report the first qualification results on a Flooding Experiments with Blocked Array (FEBA) test. Thirdly, we analyze the model predictions on a large break LOCA transient, in a 900 MW PWR 11x11 core area with a 3x3 central blockage. THYC simulates the transient in the bundle around and above the blockage, until the quench front enters the computational domain. Previously, a 1D CATHARE simulation gives the boundary conditions and, in the reactor core case, the deformation of the blocked fuel rods. The results analysis focused on the time evolution of the clad temperatures in the blocked and in the bypass region. In the FEBA test simulation, the main observations are properly predicted within the blockage. Temperatures are lower in blocked rod sleeves than in unblocked rod claddings since the steam gap reduces the power transmitted by the heater rod to the sleeve. In the core case, the model predicts the opposite result. Within the blockage, ballooned rod temperatures are higher than non-ballooned rod ones. We show by sensitivity studies that these behaviour difference between FEBA rods

Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

International Nuclear Information System (INIS)

2014-02-01

The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

Nuclear reactors transients identification and classification system; Sistema de identificacao e classificacao de transientes em reatores nucleares

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)

RAP-2A Computer code for transients analysis in fast reactors

International Nuclear Information System (INIS)

Iftode, I.; Popescu, C.; Turcu, I.; Biro, L.

1975-10-01

The RAP-2A computer code is designed for analyzing thermohydraulic transients and/or steady state problems for large LMFBR cores. Physical and mathematical models, main input-output data, the flow chart of the code and a sample problem are given. RAP-2A calculates the power and the thermoydraulic transients initiated by a flow or reactivity changes, from a normal operating state of the reactor up to core disassembly. In this analysis a representative fuel pin is considered: a one-group space-independent (point) kinetics model to describe the neutron kinetics and a one-dimensional model describing the heat transfer (radial in the fuel and axial in the coolant) are used. Mechanical deformations due to temperature gradient, pressure losses, fuel melting, etc., are also calculated. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer

An overview of IPPE research on liquid metal fast reactor thermohydraulics

International Nuclear Information System (INIS)

Sorokin, A. P.; Efanov, A. D.; Zhukov, A. V.; Bogoslovskaia, G. P.

2003-01-01

The paper presents brief information on the most significant researches in the fields of liquid metal hydrodynamics and heat transfer performed in the State Scientific Center of Russian Federation 'Institute for Physics and Power Engineering' named after A.I.Leypunski applied to sodium-cooled fast reactors. Experimental methods for studying liquid metal thermohydraulics and applied measurement techniques are overviewed briefly in the paper. Some results of fundamental thermohydraulic investigations, such as quasi-universal character of velocity and temperature profile in liquid metals, if considered normally to the channel wall etc. are presented. Specific features of heat transfer in liquid metal cooled fuel subassembly are mentioned, among them there are: high level of coolant temperature; significant influence of an interchannel exchange on velocity and temperature distribution; an availability of contact thermal resistance; large azimuthal non-uniformity of velocity and temperature; 'conjugate' problem of heat transfer in combined geometry of fuel pin; an absence of stabilization of heat transfer in non-standard channels; an influence of non-uniform heat generation. Special attention is given to the temperature fields in fuel subassembly subjected to deformation because of radioactive swelling and creeping, as well as in case of blockage of a part of subassembly cross section. Some results of thermohydraulic investigation are demonstrated for intermediate heat exchangers, pressurized head collectors. Also the developed methods and codes of thermohydraulic calculations applied to fast reactor core are considered: subchannel approach, porous body model

EBaLM-THP - A neural network thermohydraulic prediction model of advanced nuclear system components

International Nuclear Information System (INIS)

Ridluan, Artit; Manic, Milos; Tokuhiro, Akira

2009-01-01

In lieu of the worldwide energy demand, economics and consensus concern regarding climate change, nuclear power - specifically near-term nuclear power plant designs are receiving increased engineering attention. However, as the nuclear industry is emerging from a lull in component modeling and analyses, optimization for example using ANN has received little research attention. This paper presents a neural network approach, EBaLM, based on a specific combination of two training algorithms, error-back propagation (EBP), and Levenberg-Marquardt (LM), applied to a problem of thermohydraulics predictions (THPs) of advanced nuclear heat exchangers (HXs). The suitability of the EBaLM-THP algorithm was tested on two different reference problems in thermohydraulic design analysis; that is, convective heat transfer of supercritical CO 2 through a single tube, and convective heat transfer through a printed circuit heat exchanger (PCHE) using CO 2 . Further, comparison of EBaLM-THP and a polynomial fitting approach was considered. Within the defined reference problems, the neural network approach generated good results in both cases, in spite of highly fluctuating trends in the dataset used. In fact, the neural network approach demonstrated cumulative measure of the error one to three orders of magnitude smaller than that produce via polynomial fitting of 10th order

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)

Modelling of an ULOF transient in a sodium fast reactor

International Nuclear Information System (INIS)

Droin, Jean-Baptiste

2016-01-01

particular attention is paid to stabilize boiling occurrence which leads to minimize severe accident consequences.The phenomena occurring during the various ULOF transients are modelled in accordance to the level of details required to catch all the possible bifurcations of the transient. The tool coupled different (2D, 1D and 0D) models of thermics, thermo-hydraulics, core degradation (material melting and motions) and neutronics. The assumptions associated to these models are highlighted, discussed and validated. The physical tool capability of simulating the various realistic ULOF transients (without boiling, with stabilized boiling or flow excursion after boiling) is demonstrated by comparisons to experimental results (GR19, SCARABEE experiments) and to mechanistic simulations (CATHARE2 and SIMMER III).Parametric studies are then carried out on two variables: the fuel burn-up and the model of neutronic feedbacks. They underline the important influence of these parameters on the transient and the final core state. Finally, a preliminary sensitivity analysis (2000 simulations) is performed on 26 uncertain parameters (linked to initial core configuration, accident features, model uncertainties and radial nodalization). The variability of the final core state is underlined and quantified; only around 25% of cases lead to core degradation. The main influent parameters on transient phenomena are also identified, enabling to prioritize core design and safety studies. In the future, this tool will be used for safety-informed design and stability analyses of fast reactor systems, allowing to emphasize the main dominant phenomena and trends of significance for safety assessment. (author) [fr

CAREM reactor thermohydraulic essays laboratory

International Nuclear Information System (INIS)

Horro, R.; Mazzi, R.; Rossini, A.

1990-01-01

The main characteristics, essays projected and the present state of the Thermohydraulic Essays Laboratory -under construction at present- prepared to meet the experimental needs resulting from a power reactor design of the CAREM type, are herein described. (Author) [es

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)

Condensation front modelling in a moisture separator reheater by application of SICLE numerical model

International Nuclear Information System (INIS)

Grange, J.L.; Caremoli, C.; Eddi, M.

1988-01-01

This paper presents improvements performed on SICLE numerical model in order to analyse the condensation front that occurs in the moisture separator reheaters (MSR) of nuclear power plants. Modifications of SICLE numerical model architecture and a fine modelling of reheater have allowed to correctly simulate the MSR thermohydraulic behaviour during a severe transient (plant islanding) [fr

Thermo-Hydraulic behaviour of dual-channel superconducting Cable-In-Conduit Conductors for ITER

International Nuclear Information System (INIS)

Renard, B.

2006-09-01

In an effort to optimise the cryogenics of large superconducting coils for fusion applications (ITER), dual channel Cable-In-Conduit Conductors (CICC) are designed with a central channel spiral to provide low hydraulic resistance and faster helium circulation. The qualitative and economic rationale of the conductor central channel is here justified to limit the superconductor temperature increase, but brings more complexity to the conductor cooling characteristics. The pressure drop of spirals is experimentally evaluated in nitrogen and water and an explicit hydraulic friction model is proposed. Temperatures in the cable must be quantified to guarantee superconductor margin during coil operation under heat disturbance and set adequate inlet temperature. Analytical one-dimensional thermal models, in steady state and in transient, allow to better understand the thermal coupling of CICC central and annular channels. The measurement of a heat transfer characteristic space and time constants provides cross-checking experimental estimations of the internal thermal homogenization. A simple explicit model of global inter-channel heat exchange coefficient is proposed. The risk of thermosyphon between the two channels is considered since vertical portions of fusion coils are subject to gravity. The new hydraulic model, heat exchange model and gravitational risk ratio allow the thermohydraulic improvement of CICC central spirals. (author)

Analysis of transient thermal response in the outlet plenum of an LMFBR

International Nuclear Information System (INIS)

Yang, J.W.

1976-05-01

A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the upper outlet plenum of an LMFBR. The one-dimensional turbulent jet flow equations were solved to determine the maximum penetration of the core flow. The maximum penetration is used as the criterion for dividing the sodium region into two mixing zones. The lumped-parameter model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of bypass flow into the plenum. Numerical calculations were performed for two cases. The first case corresponds to a normal scram followed by flow coast-down. The second case represents the double-ended pipe rupture at the inlet of cold leg followed by reactor scram. The results indicate that effects of flow stratification, chimney height, metal heat capacity and bypass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas does not play any significant role on sodium temperature. This two-zone mixing model will be a part of the thermohydraulic transient code SSC

Confirmatory simulation of safety and operational transients in LMFBR systems

International Nuclear Information System (INIS)

Guppy, J.G.; Agrawal, A.K.

1978-01-01

Operational and safety transients (anticipated, unlikely, or extremely unlikely) that may originate anywhere in a liquid-metal fast breeder reactor (LMFBR) system must be adequately simulated to assist in safety evaluation and plant design efforts. An advanced thermohydraulic transient code, the Super System Code (SSC), is described that may be used for confirmatory safety evaluations of plant-wide events, such as assurance of adequate decay heat removal capability under natural circulation conditions. Results obtained with SSC illustrating the degree of modeling detail present in the code as well as the computing efficiency are presented. A version of the SSC code, SSC-L, applicable to any loop-type LMFBR design, has been developed at Brookhaven. The scope of SSC-L is to enable the simulation of all plant-wide transients covered by Plant Protection System (PPS) action, including sodium pipe rupture and coastdown to natural circulation conditions. The computations are stopped when loss of core integrity (i.e., clad melting temperature exceeded) is indicated

IBIS, FBR 3-D Steady-State and Kinetics with Thermohydraulic Feedback

International Nuclear Information System (INIS)

Konomura, Mamoru; Tada, Nobuo; Oka, Yoshiaki; An, Shigehiro

1987-01-01

1 - Description of program or function: The IBIS code performs steady state and kinetics calculations based on a three-dimensional nuclear diffusion kinetics with thermal hydraulic feedback. It can calculate the following values in hexagonal-Z geometry of a fast breeder reactor core through the progress of transient: (1) Net reactivity; (2) Total and group-wise delayed neutron fraction; (3) Group-wise delayed neutron precursor concentration; (4) Total power and energy; (5) Space dependent neutron flux in each energy group; (6) Space dependent temperature of each material; (7) Maximum temperature of each material and its location. 2 - Method of solution: The quasi-static method is adopted to solve the three-dimensional nuclear diffusion kinetics problem. The method is the same as employed in the code QX1. The shape function equation is solved with the finite difference treatment as used in the codes CITATION and HONEYCOMB. One-dimensional thermo-hydraulics is solved with a model similar to that given in the code SASLA. Sodium boiling can be taken into account. 3 - Restrictions on the complexity of the problem: The number of neutron energy groups is fixed to 3 groups in the present version of the code

A novel thermo-hydraulic coupling model to investigate the crater formation in electrical discharge machining

Science.gov (United States)

Tang, Jiajing; Yang, Xiaodong

2017-09-01

A novel thermo-hydraulic coupling model was proposed in this study to investigate the crater formation in electrical discharge machining (EDM). The temperature distribution of workpiece materials was included, and the crater formation process was explained from the perspective of hydrodynamic characteristics of the molten region. To better track the morphology of the crater and the movement of debris, the level-set method was introduced in this study. Simulation results showed that the crater appears shortly after the ignition of the discharge, and the molten material is removed by vaporizing in the initial stage, then by splashing at the following time. The driving force for the detachment of debris in the splashing removal stage comes from the extremely large pressure difference in the upper part of the molten region, and the morphology of the crater is also influenced by the shearing flow of molten material. It was found that the removal ratio of molten material is only about 7.63% under the studied conditions, leaving most to form the re-solidification layer on the surface of the crater. The size of the crater reaches the maximum at the end of discharge duration then experiences a slight reduction because of the reflux of molten material after the discharge. The results of single pulse discharge experiments showed that the morphologies and sizes between the simulation crater and actual crater are good at agreement, verifying the feasibility of the proposed thermo-hydraulic coupling model in explaining the mechanisms of crater formation in EDM.

Comparative Analysis of Thermohydraulic Margins in Embalse Power Station, CARA Vs. CANDU with Cobra IV-HW

International Nuclear Information System (INIS)

Daverio, H; Juanico, L

2000-01-01

Comparative analysis of thermohydraulic margins were studied of the CANDU 37 and CARA fuel bundles (FB) in Embalse power station with COBRA IV-HW code ., the geometry of the bundle laying on the channel was particularly modeled and discussing the results in comparison with former calculations with 1/6 simetry .The CARA design with enriched uranium (0.9 %) and extended burn up lets maintain the current thermohydraulic nominal margins , while compared with CANDU 37 rods FB enriched , the CARA design permits widely improve the current margins

Prediction of the thermohydraulic performance of porous-media reservoirs for compressed-air energy storage

Energy Technology Data Exchange (ETDEWEB)

Wiles, L.E.; McCann, R.A.

1981-09-01

The numerical modeling capability that has been developed at the Pacific Northwest Laboratory (PNL) for the prediction of the thermohydraulic performance of porous media reservoirs for compressed air energy storage (CAES) is described. The capability of the numerical models was demonstrated by application to a variety of parametric analyses and the support analyses for the CAES porous media field demonstration program. The demonstration site analyses include calculations for the displacement of aquifer water to develop the air storage zone, the potential for water coning, thermal development in the reservoir, and the dehydration of the near-wellbore region. Unique features of the demonstration site reservoir that affect the thermohydraulic performance are identified and contrasted against the predicted performance for conditions that would be considered more typical of a commercial CAES site.

Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility

International Nuclear Information System (INIS)

Coragem, Helio Boemer de Oliveira

1980-01-01

A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

Limitations of transient power loads on DEMO and analysis of mitigation techniques

Energy Technology Data Exchange (ETDEWEB)

Maviglia, F., E-mail: francesco.maviglia@euro-fusion.org [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Consorzio CREATE, University Napoli Federico II – DIETI, 80125 Napoli (Italy); Federici, G. [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Strohmayer, G. [Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Wenninger, R. [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Bachmann, C. [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Albanese, R. [Consorzio CREATE, University Napoli Federico II – DIETI, 80125 Napoli (Italy); Ambrosino, R. [Consorzio CREATE University Napoli Parthenope, Naples (Italy); Li, M. [Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Loschiavo, V.P. [Consorzio CREATE, University Napoli Federico II – DIETI, 80125 Napoli (Italy); You, J.H. [Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Zani, L. [CEA, IRFM, F-13108 St Paul-Lez-Durance (France)

2016-11-01

Highlights: • A parametric thermo-hydraulic analysis of the candidate DEMO divertor is presented. • The operational space assessment is presented under static and transient heat loads. • Strike points sweeping is analyzed as a divertor power exhaust mitigation technique. • Results are presented on sweeping installed power required, AC losses and thermal fatigue. - Abstract: The present European standard DEMO divertor target technology is based on a water-cooled tungsten mono-block with a copper alloy heat sink. This paper presents the assessment of the operational space of this technology under static and transient heat loads. A transient thermo-hydraulic analysis was performed using the code RACLETTE, which allowed a broad parametric scan of the target geometry and coolant conditions. The limiting factors considered were the coolant critical heat flux (CHF), and the temperature limits of the materials. The second part of the work is devoted to the study of the plasma strike point sweeping as a mitigation technique for the divertor power exhaust. The RACLETTE code was used to evaluate the impact of a large range of sweeping frequencies and amplitudes. A reduced subset of cases, which complied with the constraints, was benchmarked with a 3D FEM model. A reduction of the heat flux to the coolant, up to a factor ∼4, and lower material temperatures were found for an incident heat flux in the range (15–30) MW/m{sup 2}. Finally, preliminary assessments were performed on the installed power required for the sweeping, the AC losses in the superconductors and thermal fatigue analysis. No evident show stoppers were found.

Utilization of Relap 5 computer code for analyzing thermohydraulic projects

International Nuclear Information System (INIS)

Silva Filho, E.

1987-01-01

This work deals with the design of a scaled test facility of a typical pressurized water reactor plant of the 1300 MW (electric) class. A station blackout has been choosen to investigate the thermohydraulic behaviour of the the test facility in comparison to the reactor plant. The computer code RELAPS/MOD1 has been utilized to simulate the blackout and to compare the test facility behaviour with the reactor plant one. The results demonstrate similar thermohydraulic behaviours of the two systems. (author) [pt

Thermohydraulic tests of 3x3-rod bundle maquette

International Nuclear Information System (INIS)

Ladeira, L.C.D.

1986-10-01

The results of a 3x3-rod bundle thermohydraulic research program, performed in the Thermohydraulics Laboratory of NUCLEBRAS' Nuclear Technology Development Center, are briefly described. This program included measurements of pressure drops in one and two-phase flows, heat transfer coefficients, mixing between interconnected subchannels in one-phase flow conditions and critical heat fluxes. The measurements covered the following parameter ranges: heat fluxes from zero to the critical values, pressure ranging from 1 to 15 ata, inlet temperature from 25 to 150 sup(0)C and flow rate from 20 to 300l/min. (author)

Observation and control system of the thermohydraulic assays laboratory

International Nuclear Information System (INIS)

Santome, D.; Hualde, R.

1990-01-01

The Thermohydraulic Assays Laboratory (L.E.T.) is an installation whose purpose will be the components testing and the CAREM-25 reactor thermohydraulic processes operation dynamics. This plant is located at Pilcaniyeu, province of Rio Negro. Part of the tests which will be carried out consist in the use of different control strategies. The control of the systems by digital processors (control by software) has been decided to proceed with a maximum flexibility and capacity to make changes in the algorithms. This work describes the design and implementation of a digital control system to command the three circuits of the installation. (Author) [es

Application of CFD methods in research of SCWR thermo-hydraulics

International Nuclear Information System (INIS)

Zeng Xiaokang; Li Yongliang; Yan Xiao; Xiao Zejun; Huang Yanping

2013-01-01

The CFD method has been an important tool in the research of SCWR thermo- hydraulics. Currently, the CFD methods uses commonly the subcritical turbulence models, which can not accurately simulate the gravity and thermal expansion acceleration effect, and CFD numerical method is not applicable when the heat flux is large. The paper summarizes the application status of the CFD methods in the research of SCWR thermo-hydraulics in RETH. (authors)

Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 4. Numerical analysis of 1/10 scaled water experiment with the AQUA code

International Nuclear Information System (INIS)

Muramatu, Toshiharu; Yamaguchi, Akira

2004-01-01

A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. A numerical analysis was carried out with a multi-dimensional code AQUA to confirm an applicability to the evaluations for the in-vessel thermohydraulic phenomena using a 1/10 scaled water experiment simulating the large-scale fast breeder reactor in the feasibility studies. From the analysis, the following results were obtained. (1) In-vessel thermohydraulics characterized by a radiated flow pattern to the reactor vessel wall and a strong upward flow through a slit of the upper core structures were evaluated. These characteristics agreed approximately with the water experiment. (2) The upward velocity values at the slit agreed well with the experimental data under a condition of γ z = 0.3 and ξ z = 0.5, though overall evaluations of the in-vessel thermohydraulics were failed to predict quantitatively. (3) The AQUA code is applicable to the in-vessel thermohydraulics evaluations in the feasibility studies, though it is necessary to make further modifications of the calculational models for accurate evaluations. On the one hand, it was confirmed that calculated results for the 1/10 water experimental model and the 1/1 actual-scaled model agreed quantitatively for the in-vessel thermohydraulics characteristics indicated above. (author)

FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

International Nuclear Information System (INIS)

Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

1984-01-01

1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

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)

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

Study on the transient behaviours of MNSR reactor for control rod withdrawal

International Nuclear Information System (INIS)

Yang Shunhai

1995-10-01

The transient behaviours of Miniature Neutron Source Reactor MNSR are analyzed and calculated with the reactor thermohydraulics RETRAN-02 program and the reactor physics MARIA program. The obtained event sequence and consequence from the calculation are compared with the experiments. The effective resonance integral for study on Doppler effect is taken into account. The reactivity temperature coefficient weighting factors are computed. The transient parameters related to reactor power peaking, coolant inlet temperatures, outlet temperatures and coolant mass flow, etc. are computed and compared with the experimental results. (6 refs., 2 figs., 5 tabs.)

Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

International Nuclear Information System (INIS)

March-Leuba, J.; Rey, J.M.

1992-01-01

This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of 'unexpected' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a 'new and improved' state of the art has emerged recently. (authors). 6 figs., 57 refs., 1 appendix

Assessment of TRAC-PD2 reflood core thermo-hydraulic model by CCTF Test C1-16

International Nuclear Information System (INIS)

Sugimoto, Jun

1982-11-01

The TRAC-PD2 reflood core thermo-hydraulic model was assessed by CCTF Test C1-16. The measured data were utilized as core boundary conditions in the TRAC calculations. The results indicate that the core inlet liquid temperature and the core heater rod temperatures are in reasonable agreement with data, but the pressure distribution in the core and water pool formation in the upper plenum are not in good agreement. The parametric effects of the droplet critical Weber number, the material properties of the heater rod, the noding of the upper plenum, and the minimum stable film boiling temperature are also discussed. (author)

The applicability of CFD to simulate and study the mixing process and the thermo-hydraulic consequences of a main steam line break in PWR model

Directory of Open Access Journals (Sweden)

Farkas Istvan

2017-01-01

Full Text Available This paper focuses on the validation and applicability of CFD to simulate and analyze the thermo-hydraulic consequences of a main steam line break. Extensive validation data come from experiments performed using the Rossendorf coolant mixing model facility. For the calculation, the range of 9 to 12 million hexahe¬dral cells was constructed to capture all details in the interrogation domain in the system. The analysis was performed by running a time-dependent calculation, Detailed analyses were made at different cross-sections in the system to evaluate not only the value of the maximum and minimum temperature, but also the loca¬tion and the time at which it occurs during the transient which is considered to be indicator for the quality of mixing in the system. CFD and experimental results were qualitatively compared; mixing in the cold legs with emergency core cooling systems was overestimated. This could be explained by the sensitivity to the bound¬ary conditions. In the downcomer, the experiments displayed higher mixing: by our assumption this related to the dense measurement grid (they were not modelled. The temperature distribution in the core inlet plane agreed with the measurement results. Minor deviations were seen in the quantitative comparisons: the maximum temperature difference was 2ºC.

Transient Analysis of Generation IV quick reactors; Analisis de Transitorios en Reactores Rapidos de Generacion IV

Energy Technology Data Exchange (ETDEWEB)

Vazquez, M.; Martin-Fuertes, F.

2013-07-01

As a complement to the attached code 3D neutron-CIEMAT thermohydraulic added a module to simulate transient. Temporary kinetics is resolved by factoring flow in a spatial part and another storm. MCNP provides the reactivity and updated spatial function and COBRA-IV calculates the temperature distribution. Temporary dependence of amplitude is calculated using time delayed neutron Kinetic equations. As an example of application, examines a transient loss of flow in MYRRHA, a lead-cooled experimental reactor.

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.

Analysis of steam generator plugging on core thermohydraulic performance of NPP Krsko; Analiza vpliva cepljenja cevi v upravljaniku na termohidravliko sredice JE Krsko

Energy Technology Data Exchange (ETDEWEB)

Kostadinov, V; Petelin, S; Sarler, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)

1988-07-01

Nuclear safety analysis of NPP Krsko core operating at full power with 4% steam generator tubes plugged have been performed. Influence of individual parameters on core thermohydraulic performance have been evaluated. Using COBRA-III-C computer code we have analysed a core design (evaluation) model. The DNBR change was calculated as a consequence of 4% plugging. The influence of thermohydraulic parameters change on DNBR was analysed. (author)

The state of art of the methods for thermohydraulics design of LMFBR fuel elements

International Nuclear Information System (INIS)

Fernandez y Fernandez, E.; Carajilescov, P.

1981-09-01

The present (experimental and analytical) state of art of the methods for thermohydraulics design of LMFBR fuel elements is analyzed. A development program is suggested, in order to obtain a computer code for modelling the distribution of coolant enthalpy in reactor core. This computer code is in development. (Author) [pt

Neutronics and thermohydraulics of the reactor C.E.N.E. Part II

International Nuclear Information System (INIS)

Caro, R.

1976-01-01

In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

Investigation on in-vessel thermal transients in a fast breeder reactor

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Kasahara, Naoto

1999-01-01

Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of the phenomena in the design of the internal structures in an LMFBR plenum. To evaluate thermal stress characteristics for the inner barrel in a typical LMFBR upper plenum, numerical analysis was carried out with a multi-dimensional thermohydraulics code AQUA for a scram condition from full power operation conditions. Thereafter, thermal stress conditions for the inner barrel were evaluated by the use of a structural analysis code FINAS with the thermohydraulic results calculated by the AQUA code as boundary conditions. From the thermohydraulic analysis and the thermal stress analysis, the following results have been obtained. (1) A large axial temperature gradient was calculated at the region between the upper and lower flow holes located on the inner barrel. The axial position of the thermal stratification interface was fixed in the various circumferential directions. As for the comparison with a 40% operation condition, maximum temperature gradients at the lower flow hole region indicated a 2 times value of that in the 40% operation condition. (2) Transient thermal stratification phenomena were observed after 120 sec from the reactor scram in the numerical results. These tendencies on thermal stratification phenomena were sameness with the transient results from the 40% operation condition. (3) During the reactor trip from full power operation, large temperature gradient in both vertical and sectional direction are enforced around the lower flow hole, since there exists flow pass of low temperature sodium through this hole. As a result, the maximum thermal stress within 32.6 kg/mm 2 was predicted at the lower flow hole when considering stress concentration at the hole edge. (J.P.N.)

Adaptation and implementation of the TRACE code for transient analysis on designs of cooled lead fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

Energy Technology Data Exchange (ETDEWEB)

Lazaro, A.; Ammirabile, L.; Martorell, S.

2014-07-01

The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)

On three-dimensional nuclear thermo-hydraulic computation techniques for ATR

International Nuclear Information System (INIS)

1997-08-01

The three-dimensional computation code for nuclear thermo-hydraulic combination core LAYMON-2A is used for the calculation of the power distribution and the control rod reactivity value of the ATR. This code possesses various functions which are required for planning the core operation such as the search function for critical boric acid concentration, and can do various simulation calculations such as core burning calculation. Further, the three-dimensional analysis code for xenon dynamic characteristics in the core LAYMON-2C, in which the dynamic characteristic equation of xenon-samarium was incorporated into the LAYMON-2A code can take the change with time lapse of xenon-samarium concentration accompanying the change of power level and power distribution into account, and it is used for the analysis of the spatial vibration characteristics of power and the regional power control characteristics due to xenon in the core. As to the LAYMON-2A, the computation flow, power distribution and thermo-hydraulic computation models, and critical search function are explained. As to the LAYMON-2C, the computation flow is described. The comparison of the calculated values by using the LAYMON-2A code and the operation data of the Fugen is reported. (K.I.)

Validation and verification of the MTRPC thermohydraulic package

International Nuclear Information System (INIS)

Doval, Alicia

1998-01-01

The MTR P C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)

Fundamental study on thermo-hydraulics during start-up in natural circulation boiling water reactors, (1)

International Nuclear Information System (INIS)

Aritomi, Masanori; Chiang Jing-Hsien; Takahashi, Tohru; Wataru, Masumi; Mori, Michitsugu.

1992-01-01

Recently, many concepts, in which passive and simplified functions are actively adapted, have been proposed for the next generation LWRs. The natural circulation BWR is one such considered from the requirements for next generation LWRs as compared with current BWRs. It is pointed out from this consideration that a thermo-hydraulic instability, which may appear during start-up, greatly influences concept feasibility because its occurence makes operation for raising power output difficult. Thermo-hydraulic instabilities are investigated experimentally under conditions simulating normal and abnormal start-up processes. It is clarified that three kinds of thermo-hydraulic instabilities may occur during start-up in the natural circulation BWR according to its procedure and reactor configuration, which are (1) geysering induced by condensation, (2) natural circulation instability induced by hydrostatic head fluctuation in steam separators and (3) density wave instability. Driving mechanisms of the geysering and the natural circulation instability, which have never understood enough, are inferred from the results. Finally, the difference of thermo-hydraulic behavior during start-up processes between thermal natural circulation boilers and the Dodewaard reactor is discussed. (author)

Effect of thermohydraulic parameter on the flux distribution and the effective multiplication factor

International Nuclear Information System (INIS)

Mello, J.C.; Valladares, G.L.

1990-01-01

The influence of two thermohydraulics parameters; the coolant flow velocity along the reactor channels and the increase of the average water temperature through the core, on the thermal flux distribution and on the effective multiplication factor, was studied in a radioisotopes production reactor. The results show that, for a fixed values of the thermohydraulics parameters reffered above, there are limits for the reactor core volume reduction for each value of the V sub(mod)/V sub(comb) ratio. These thermohydraulics conditions determine the higher termal flux value in the flux-trap and the lower value of the reactor effective multiplication factor. It is also show that there is a V sub(mod)/V sub(comb) ratio value that correspond to the higher value of the lower effective multiplication factor. These results was interpreted and comment using fundamentals concepts and relations of reactor physics. (author)

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

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

Effects of thermohydraulics on clad ballooning, flow blockage and coolability in a LOCA

International Nuclear Information System (INIS)

Erbacher, F.J.; Neitzel, H.J.; Wiehr, K.

1983-01-01

Thermohydraulic boundary conditions have a dominating effect on clad ballooning, flow blockage and coolability: Increasing heat transfer to the fluid decreases the total circumferential strain; Countercurrent flow in a combined injection leads to a relatively small flow blockage; Burst claddings exhibit premature quenching. Differences in the test results obtained in several countries are mainly due to different thermohydraulic test conditions; all test data are consistent with the understanding elaborated within the REBEKA program. Core coolability in a LOCA can be maintained. (author)

Thermohydraulic characteristics under some transient conditions of the Dalat Nuclear Research Reactor

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.

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

Energy Technology Data Exchange (ETDEWEB)

Elina Syrjaelahti; Anitta Haemaelaeinen [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

2005-07-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

International Nuclear Information System (INIS)

Elina Syrjaelahti; Anitta Haemaelaeinen

2005-01-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Numerical modeling of secondary side thermohydraulics of horizontal steam generator

Energy Technology Data Exchange (ETDEWEB)

Melikhov, V.I.; Melikhov, O.I.; Nigmatulin, B.I. [Research and Engineering Centre of LWR Nuclear Plants Safety, Moscow (Russian Federation)

1995-12-31

A mathematical model for the transient three-dimensional secondary side thermal hydraulics of the horizontal steam generator has been developed. The calculations of the steam generator PGV-1000 and PGV-4 nominal regimes and comparison of numerical and experimental results have been carried out. 7 refs.

Numerical modeling of secondary side thermohydraulics of horizontal steam generator

Energy Technology Data Exchange (ETDEWEB)

Melikhov, V I; Melikhov, O I; Nigmatulin, B I [Research and Engineering Centre of LWR Nuclear Plants Safety, Moscow (Russian Federation)

1996-12-31

A mathematical model for the transient three-dimensional secondary side thermal hydraulics of the horizontal steam generator has been developed. The calculations of the steam generator PGV-1000 and PGV-4 nominal regimes and comparison of numerical and experimental results have been carried out. 7 refs.

Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

International Nuclear Information System (INIS)

Kamide, H.; Ieda, Y.; Toda, S.; Isozaki, T.; Sugawara, S.

1993-01-01

Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor core during natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

Energy Technology Data Exchange (ETDEWEB)

Kamide, H; Ieda, Y; Toda, S; Isozaki, T; Sugawara, S [Reactor Engineering Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, Narita, O-arai, Ibaraki-ken (Japan)

1993-02-01

Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor coreduring natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA

International Nuclear Information System (INIS)

Badulescu, A.; Groeneveld, D.C.

2000-01-01

Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)

Control of the neutronic and thermohydraulic conditions of power ramps in an irradiation loop for PWR fuel rod; Controle des conditions neutroniques et thermohydrauliques des rampes de puissance dans une boucle d`irradiation de combustibles de reacteur a eau pressurisee

Energy Technology Data Exchange (ETDEWEB)

Moulin, D J.F.

1993-09-10

In order to study the power transients effects on PWR fuel rod clad, ramp tests in a pressurized water loop, are carried out at OSIRIS reactor. The present thesis deals with the on-line control of the device, during power ramp and conditioning irradiation. Based on a convolution-type resolution of the kinetics equations, a dynamic compensation of the Silver self-powered neutron detector was developed. With this method, the uncertainty of the ramp end-point is lower than 1%, thus it is very suited for monitoring both transient, as well as steady state conditions. Furthermore, a thermohydraulic model of the irradiation device is described: heat transfer equations, including gamma heating in materials, are solved to obtain temperatures and thermal fluxes of steady states. Results from the model and temperature measurements of the coolant are used together for fuel power determination, in real time. The clad external temperature profile is also calculated and displayed, to improve the irradiation monitoring. (author), 51 refs., 12 annexes, 66 figs.

Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

International Nuclear Information System (INIS)

Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

1983-01-01

The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

The problems of thermohydraulics of prospective fast reactor concepts

International Nuclear Information System (INIS)

Sedov, A.A.

2000-01-01

In this report the main requirements to fast reactors in system of future multicomponent Nuclear Power with closed U-Pu fuel cycle are regarded. The peculiarities of different liquid-metal (sodium and lead-alloyed) coolants as well as the thermohydraulics problems of prospective fast reactors (FR) concepts are discussed. (author)

Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies

International Nuclear Information System (INIS)

Sorokin, G.A.; Avdeev, E.F.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P.

2000-01-01

An investigation into the reactor core accident cooling, which are associated with the power grow up or switch off circulation pumps in the event of the protective equipment comes into action, results in the problem of liquid metal boiling heat transfer. Considerable study has been given over the last 30 years to alkaline metal boiling including researches of heat transfer, boiling patterns, hydraulic resistance, crisis of heat transfer, initial heating up, boiling onset and instability of boiling. The results of these investigations have shown that the process of liquid metal boiling has substantial features in comparison with water boiling. Mathematical modeling of two phase flows in fast reactor fuel subassemblies have been developed intensively. Significant success has been achieved in formulation of two phase flow through the pin bundle and in their numerical realization. Currently a set of codes for thermohydraulic analysis of two phase flows in fast reactor subassembly have been developed with 3D macrotransfer governing equations. These codes are used for analysis of boiling onset and liquid metals boiling in fuel subassemblies during loss-of-coolant accidents, of warming up of reactor core, of blockage of some part of flow cross section in fuel subassembly. (author)

Modelling of thermohydraulic emergency core cooling phenomena

International Nuclear Information System (INIS)

Yadigaroglu, G.; Andreani, M.; Lewis, M.J.

1990-10-01

The codes used in the early seventies for safety analysis and licensing were based either on the homogeneous model of two-phase flow or on the so-called separate-flow models, which are mixture models accounting, however, for the difference in average velocity between the two phases. In both cases the behavior of the mixture is prescribed a priori as a function of local parameters such as the mass flux and the quality. The modern best-estimate codes used for analyzing LWR LOCA's and transients are often based on a two-fluid or 6-equation formulation of the conservation equations. In this case the conservation equations are written separately for each phase; the mixture is allowed to evolve on its own, governed by the interfacial exchanges of mass, momentum and energy between the phases. It is generally agreed that such relatively sophisticated 6-equation formulations of two-phase flow are necessary for the correct modelling of a number of phenomena and situations arising in LWR accidental situations. They are in particular indispensible for the analysis of stratified or countercurrent flows and of situations in which large departures from thermal and velocity equilibrium exist. This report will be devoted to a discussion of the need for, the capacity and the limitations of the two-phase flow models (with emphasis on the 6-equation formulations) in modelling these two-phase flow and heat transfer phenomena and/or different core cooling situations. 18 figs., 1 tab., 72 refs

Thermohydraulics in rod bundles and critical heat flux in transient conditions in a tube

International Nuclear Information System (INIS)

Courtaud, M.; Roumy, R.

1975-01-01

After the determination of the scaling factor of Stevens's similitude for the pressure range of pressurized water vectors by comparison of critical heat flux data in from and in water, some examples of studies performed with freon are shown. The efficiency of the mixing vanes of spacer grids has been determined on the mixing phenomenon in single phase on critical heat flux. A calculation performed with the code FLICA using subchannel analysis on freon data transposed in water is in good agreement with the experiment. The influence of the number of spacer grids has been also shown. Critical heat fluxes have been determined in water at 140 bar in steady state and transient conditions on two tubular test sections. During the transient tests the flow rate was reduced by half in 0.5 seconds and the reincreased heat flux and inlet temperature remaining constant. These tests have shown the validity of the method which consists in using a critical heat flux correlation determined in steady state conditions applied with local transient conditions of enthalpy and mass velocity computed with the FLICA code [fr

Numerical simulation of fuel assembly thermohydraulics of fast reactors with the partial blockage of cross section under the coolant

International Nuclear Information System (INIS)

Zhukov, A.V.; Sorokin, A.P.

2000-01-01

The problems of numerical modeling of thermohydraulics in assembly of fuel elements of fast reactors with the partial blockage of cross-section under the coolant are considered. The information about existing codes constructed on use of subchannel technique and model of porous body are presented. The results of calculation obtained by these codes are presented. (author)

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

International Nuclear Information System (INIS)

2014-08-01

The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-08-15

The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

International Nuclear Information System (INIS)

Wiles, L.E.; McCann, R.A.

1981-09-01

This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis

International Nuclear Information System (INIS)

Arien, B.; Daniels, J.

1986-12-01

CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)

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)

On closure strategy for 1-D thermohydraulics models and closure relationships of two-phase flows in simple and subchannel geometry for NPP accident conditions

International Nuclear Information System (INIS)

Kornienko, Y.; Kornienko, E.; Ninokata, H.

2001-01-01

One-dimensional mathematical models are extensively used in thermohydraulics assessment of Nuclear Power Plant (NPP) transients and accidents, because specifically 1-D system of the conservation laws allows to reduce computing time and required memory, especially in ''best estimate'' code calculations. This work is generalization of the well-known Zuber-Findley and Hancox-Nicoll methods for two-phase flow distribution parameters Cs taking into account the non-monotonous void fraction distribution in the transverse direction in terms of two superimposed monotonous profiles. The method is very useful in evaluating the saddle-shape void fraction profile effects. In this work two-phase flow distribution parameters Cs were developed for simple circular and rectangular pipes, and subchannel geometry in a rod bundle. Basic assumptions were power-mode approximations for describing the profiles of local volume flux density, phase velocity and temperature. The general analytical (quadrature) relationships for Cs were obtained and their 3-D illustrations are proposed. Also, we propose generalized formulation and simple approach to construct friction factor, heat and mass transfer coefficients within the gradient hypothesis and boundary layer assumptions. The contribution of momentum, heat and mass transfer as well as their sources and sinks in the channel cross-section are taken into account. In the same way, the friction factor, heat and mass transfer coefficients with the transversal and azimuthal variations being taken into account are proposed for subchannel geometry as well. (author)

Seminar on experimental thermohydraulics

International Nuclear Information System (INIS)

Katsaounis, A.

1979-06-01

Considerations on reactor safety are made. Problems related to the project and assembling of facilities for experiments in steady state and transient conditions are discussed. The advantages of using model fluids are finally, analysed. (Author) [pt

Systems for neutronic, thermohydraulic and shielding calculation in personal computers

International Nuclear Information System (INIS)

Villarino, E.A.; Abbate, P.; Lovotti, O.; Santini, M.

1990-01-01

The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author) [es

Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

International Nuclear Information System (INIS)

Kobayashi, Michiyuki; Aya, Izuo; Inasaka, Fujio; Murata, Hiroyuki; Odano, Naoteru; Shiozaki, Koki

1998-01-01

A research project from 1995-1999 had a plan to make experimental studies on (1) safety of nuclear ship loaded with an integral ship propulsion reactor (2) effects of pulsating flow on the thermo-hydraulic characteristics of ship propulsion reactor and (3) thermo-hydraulic behaviors of the reactor container at the time of accident in a passively safe ship propulsion reactor. Development of a data base for ship propulsion reactor was attempted using previous experimental data on the thermo-hydraulic characteristics of the reactor in the institute in addition to the present results aiming to make general analytical evaluation for the safety of the engineering-simulation system for nuclear ship. A general data base was obtained by integrating the data list and the analytical program for static characteristics. A test equipment which allows to visualize the pulsating flow was produced and visualization experiments have started. (M.N.)

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.

Thermohydraulic stability coupled to the neutronic in a BWR

International Nuclear Information System (INIS)

Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.

2006-01-01

In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the Laguna Verde

Methodology to carry out a sensitivity and uncertainty analysis for cross sections using a coupled model Trace-Parcs

International Nuclear Information System (INIS)

Reyes F, M. C.; Del Valle G, E.; Gomez T, A. M.; Sanchez E, V.

2015-09-01

A methodology was implemented to carry out a sensitivity and uncertainty analysis for cross sections used in a coupled model for Trace/Parcs in a transient of control rod fall of a BWR-5. A model of the reactor core for the neutronic code Parcs was used, in which the assemblies located in the core are described. Thermo-hydraulic model in Trace was a simple model, where only a component type Chan was designed to represent all the core assemblies, which it was within a single vessel and boundary conditions were established. The thermo-hydraulic part was coupled with the neutron part, first for the steady state and then a transient of control rod fall was carried out for the sensitivity and uncertainty analysis. To carry out the analysis of cross sections used in the coupled model Trace/Parcs during the transient, the Probability Density Functions for 22 parameters selected from the total of neutronic parameters that use Parcs were generated, obtaining 100 different cases for the coupled model Trace/Parcs, each one with a database of different cross sections. All these cases were executed with the coupled model, obtaining in consequence 100 different output files for the transient of control rod fall doing emphasis in the nominal power, for which an uncertainty analysis was realized at the same time generate the band of uncertainty. With this analysis is possible to observe the ranges of results of the elected responses varying the selected uncertainty parameters. The sensitivity analysis complements the uncertainty analysis, identifying the parameter or parameters with more influence on the results and thus focuses on these parameters in order to better understand their effects. Beyond the obtained results, because is not a model with real operation data, the importance of this work is to know the application of the methodology to carry out the sensitivity and uncertainty analyses. (Author)

Thermal-hydraulic model of the primary coolant circuits for the full-scale training facility with WWER-1000

International Nuclear Information System (INIS)

Kroshilin, A.E.; Zhukavin, A.P.; Pryakhin, V.N.

1992-01-01

The mathematical model realized in the full-scale educational facility for NPP operator training is described. The RETACT computational complex providing real time process simulation for all regimes including the maximum credible accident is used for calculation of thermohydraulic parameters of the primary coolant circuits and steam generator under stationary and transient conditions. The two-velocity two-temperature model of one-dimensional steam-water flow containing uncondensed gases is realized in the program

Fracture mechanical analysis of relevant transients in the pressure vessel of Atucha I reactor

International Nuclear Information System (INIS)

Saavedra, Fernando M.

2001-01-01

The evolution of the applied stress intensity factor K I for 10 relevant transients of the nuclear power station Atucha I obtained from thermohydraulic data is analyzed according to the methodology proposed in Section XI of ASME Boiler and Pressure Vessel Code. Vast knowledge was thus obtained about basic concepts of fracture mechanics and its application to remanent life of nuclear components. Basic knowledge which commands the performance of nuclear power stations was also obtained, especially that related to the Atucha I utility [es

3. Workshop for IAEA ICSP on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents. Presentations

International Nuclear Information System (INIS)

2012-04-01

Most advanced nuclear power plant designs adopted several kinds of passive systems. Natural circulation is used as a key driving force for many passive systems and even for core heat removal during normal operation such as NuScale, CAREM, ESBWR and Indian AHWR designs. Simulation of natural circulation phenomena is very challenging since the driving force of it is weak compared to forced circulation and involves a coupling between primary system and containment for integral type reactor. The IAEA ICSP (International Collaborative Standard Problem) on 'Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents' was proposed within the CRP on 'Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems that utilize Natural Circulation'. Oregon State University (OSU) of USA offered to host this ICSP. This ICSP plans to conduct the following experiments and blind/open simulations with system codes: 1. Quasi-steady state operation with different core power levels: Conduct quasi-steady state operation with step-wise increase of core power level in order to observe single phase natural circulation flow according to power level. The experimental facility and operating conditions for an integral PWR will be used. 2. Thermo-hydraulic Coupling between Primary system and Containment: Conduct a loss of feedwater transient with subsequent ADS blowdown and long term cooling to determine the progression of a loss of feedwater transient by natural circulation through primary and containment systems. These tests would examine the blowdown phase as well as the long term cooling using sump natural circulation by coupling the primary to containment systems. This data could be used for the evaluation of system codes to determine if they model specific phenomena in an accurate manner. OSU completed planned two ICSP tests in July 2011 and real initial and boundary conditions measured from the

Identifying the optimal supply temperature in district heating networks - A modelling approach

DEFF Research Database (Denmark)

Mohammadi, Soma; Bojesen, Carsten

2014-01-01

of this study is to develop a model for thermo-hydraulic calculation of low temperature DH system. The modelling is performed with emphasis on transient heat transfer in pipe networks. The pseudo-dynamic approach is adopted to model the District Heating Network [DHN] behaviour which estimates the temperature...... dynamically while the flow and pressure are calculated on the basis of steady state conditions. The implicit finite element method is applied to simulate the transient temperature behaviour in the network. Pipe network heat losses, pressure drop in the network and return temperature to the plant...... are calculated in the developed model. The model will serve eventually as a basis to find out the optimal supply temperature in an existing DHN in later work. The modelling results are used as decision support for existing DHN; proposing possible modifications to operate at optimal supply temperature....

Influence of the outlet air temperature on the thermohydraulic behaviour of air coolers

Directory of Open Access Journals (Sweden)

Đorđević Emila M.

2003-01-01

Full Text Available The determination of the optimal process conditions for the operation of air coolers demands a detailed analysis of their thermohydraulic behaviour on the one hand, and the estimation of the operating costs, on the other. One of the main parameters of the thermohydraulic behaviour of this type of equipment, is the outlet air temperature. The influence of the outlet air temperature on the performance of air coolers (heat transfer coefficient overall heat transfer coefficient, required surface area for heat transfer air-side pressure drop, fan power consumption and sound pressure level was investigated in this study. All the computations, using AirCooler software [1], were applied to cooling of the process fluid and the condensation of a multicomponent vapour mixture on two industrial devices of known geometries.

Accounting for the inertia of the thermocouples' measurements by modelling of a NPP Kalinin-3 transient with the coupled system code ATHLET-BIPR-VVER

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)

Thermohydraulic Design Analysis Modeling for Korea Advanced NUclear Thermal Engine Rocket for Space Application

Energy Technology Data Exchange (ETDEWEB)

Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

2015-05-15

NTR engines have continued as a main stream based on the mature technology. The typical core design of the NERVA derived engines uses hexagonal shaped fuel elements with circular cooling channels and structural tie-tube elements for supporting the fuel elements, housing moderator and regeneratively cooling the moderator. The state-of-the-art NTR designs mostly use a fast or epithermal neutron spectrum core utilizing a HEU fuel to make a high power reactor with small and simple core geometry. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The KANUTER-HEU and -LEU are the innovative and futuristic NTR engines to reduce the reactor size and to implement a LEU fuel in the reactor by using thermal neutron spectrum. The KANUTERs have some features in the reactor design such as the integrated fuel element and the regeneratively cooling channels to increase room for moderator and heat transfer in the core, and ensuing rocket performance. To study feasible design points in terms of thermo-hydraulics and to estimate rocket performance of the KANUTERs, the NSES is under development. The model of the NSES currently focuses on thermo-hydraulic analysis of the peculiar and complex EHTGR design during the propulsion mode in steady-state. The results indicate comparable performance for future applications, even though it uses the heavier LEU fuel. In future, the NSES will be modified to obtain temperature distribution of the entire reactor components and then more extensive design analysis of neutronics, thermohydraulics and their coupling will be conducted to validate design feasibility and to optimize the reactor design enhancing the rocket performance.

Thermo-hydraulic consequence of pressure suppression containment vessel during blowdown, 2

International Nuclear Information System (INIS)

Aya, Izuo; Nariai, Hideki; Kobayashi, Michiyuki

1980-01-01

As a part of the safety research works for the integral-type marine reactor, an analytical code SUPPAC-2V was developed to simulate the thermo-hydraulic consequence of a pressure suppression containment system during blowdown and the code was applied to the Model Experimental Facility of the Safety of Integral Type Marine Reactors (explained already in Part 1). SUPPAC-2V is much different from existing codes in the following points. A nonhomogeneous model for the gaseous region in the drywell, a new correlation for condensing heat transfer coefficient at drywell wall based on existing data and approximation of air bubbles in wetwell water by one dimensional bubble rising model are adopted in this code. In comparing calculational results with experimental results, values of predominant input parameters were evaluated and discussed. Moreover, the new code was applied also to the NSR-7 marine reactor, conceptually designed at the Shipbuilding Research Association in Japan, of which suppression system had been already analysed by CONTEMPT-PS. (author)

Validation and verification of the MTR{sub P}C thermohydraulic package

Energy Technology Data Exchange (ETDEWEB)

Doval, Alicia [INVAP S.E., Bariloche, Rio Negro (Argentina). Nuclear Engineering Dept.]. E-mail: doval@invap.com.ar

1998-07-01

The MTR{sub P}C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)

Method of relative comparison of the thermohydraulic efficiency of heat exchange intensification in channels of heat-exchange surfaces

International Nuclear Information System (INIS)

Dubrovskij, E.V.; Vasil'ev, V.Ya.

2002-01-01

One introduces a technique to compare relatively thermohydraulic efficiency of heat transfer intensification in channels of heat exchange surfaces of any design types. It is shown that one should compare thermohydraulic efficiency of heat exchange intensification as to the thermal power of heat exchangers and pressure losses in channels with turbulators and in polished channels of heat exchange surfaces on the basis of dimensions of heat exchangers, their heat exchange surfaces and at similar (as to Re numbers) modes of coolant flow [ru

Simulation of nonlinear dynamics of a PWR core by an improved lumped formulation for fuel heat transfer

International Nuclear Information System (INIS)

Su, Jian; Cotta, Renato M.

2000-01-01

In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant. An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling. Transient temperature response of fuel, cladding and coolant is analysed. (author)

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.

Thermohydraulic tests in the area of reactor safety done in CDTN

International Nuclear Information System (INIS)

Ladeira, L.C.D.

1990-01-01

The main experimental works performed in the last five years at the Thermohydraulics Laboratory of the Nuclear Technology Development Center, in the field of reactor safety are briefly described. This paper cover the performing and analysis of pressure drop, heat transfer and mixing tests in 3X3 rod bundle and rewetting tests in single tube section. (autor) [pt

Modeling of Transients in an Enrichment Circuit

International Nuclear Information System (INIS)

Fernandino, Maria; Delmastro, Dario; Brasnarof, Daniel

2003-01-01

In the present work a mathematical model is presented in order to describe the dynamic behavior inside a closed enrichment loop, the latter representing a single stage of an uranium gaseous diffusion enrichment cascade.The analytical model is turned into a numerical model, and implemented through a computational code.Transients of two species separation were numerically analyzed, including setting times of each magnitude, behavior of each one of them during different transients, and redistribution of concentrations along the closed loop

Deformation modeling and the strain transient dip test

International Nuclear Information System (INIS)

Jones, W.B.; Rohde, R.W.; Swearengen, J.C.

1980-01-01

Recent efforts in material deformation modeling reveal a trend toward unifying creep and plasticity with a single rate-dependent formulation. While such models can describe actual material deformation, most require a number of different experiments to generate model parameter information. Recently, however, a new model has been proposed in which most of the requisite constants may be found by examining creep transients brought about through abrupt changes in creep stress (strain transient dip test). The critical measurement in this test is the absence of a resolvable creep rate after a stress drop. As a consequence, the result is extraordinarily sensitive to strain resolution as well as machine mechanical response. This paper presents the design of a machine in which these spurious effects have been minimized and discusses the nature of the strain transient dip test using the example of aluminum. It is concluded that the strain transient dip test is not useful as the primary test for verifying any micromechanical model of deformation. Nevertheless, if a model can be developed which is verifiable by other experimentts, data from a dip test machine may be used to generate model parameters

Single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3. Input data description

Energy Technology Data Exchange (ETDEWEB)

Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1998-08-01

This report explains the numerical methods and the set-up method of input data for a single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3 (Direct Numerical Simulation using a 3rd-order upwind scheme). The code was developed to simulate non-stationary temperature fluctuation phenomena related to thermal striping phenomena, developed at Power Reactor and Nuclear Fuel Development Corporation (PNC). The DINUS-3 code was characterized by the use of a third-order upwind scheme for convection terms in instantaneous Navier-Stokes and energy equations, and an adaptive control system based on the Fuzzy theory to control time step sizes. Author expect this report is very useful to utilize the DINUS-3 code for the evaluation of various non-stationary thermohydraulic phenomena in reactor applications. (author)

Thermohydraulic behaviour and heat transfer in the molten core

International Nuclear Information System (INIS)

Reineke, H.H.

1977-01-01

Increasing the application of nuclear reactors to produce electrical power extremely unprobable accidents should be investigated too. In the Federal Republic of Germany, a research program is performed for some years engaged in accidents at light water reactors in which the melting of the reactor core is presumed. A part of this program is to investigate the thermohydraulic and the heat transfer behavior in an accumulation of molten core material. The knowledge of these events is necessary to analyse the accident exactly. Further on the results of this work are of great importance to build a catcher for the molten core material. As a result of the decay heat the molten material is heated up and the density differences induce a free convection motion. In this work the thermohydraulic behavior and the distribution of the escaping heat fluxes for several accumulations of molten core material were determined. The numerical methods for solving the system of partial differential equation were used to develop computer codes, able to compute the average and local heat fluxes at the walls enclosing the molten core material and the inside increase of the temperature. The numerical computations were confirmed and verified by experimental investigations. In these investigations the molten core material was always assumed as a homogeneous fluid. In this case, the results could be reproduced by simple power laws

Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

Energy Technology Data Exchange (ETDEWEB)

Moreira, Uebert G.; Dominguez, Dany S. [Universidade Estadual de Santa Cruz (UESC), Ilh´eus, BA (Brazil). Programa de P´os-Graduacao em Modelagem Computacional em Ciencia e Tecnologia; Mazaira, Leorlen Y.R.; Lira, Carlos A.B.O. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Hernandez, Carlos R.G., E-mail: uebert.gmoreira@gmail.com, E-mail: dsdominguez@gmail.com, E-mail: leored1984@gmail.com, E-mail: cabol@ufpe.br, E-mail: cgh@instec.cu [Instituto Superior de Tecnologas y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

2017-07-01

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

International Nuclear Information System (INIS)

Moreira, Uebert G.; Dominguez, Dany S.

2017-01-01

The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

Simulation Model of a Transient

DEFF Research Database (Denmark)

Jauch, Clemens; Sørensen, Poul; Bak-Jensen, Birgitte

2005-01-01

This paper describes the simulation model of a controller that enables an active-stall wind turbine to ride through transient faults. The simulated wind turbine is connected to a simple model of a power system. Certain fault scenarios are specified and the turbine shall be able to sustain operati...

Containment severe accident thermohydraulic phenomena

International Nuclear Information System (INIS)

Frid, W.

1991-08-01

This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

Uncertainty and Sensitivity Studies with TRACE-SUSA and TRACE-DAKOTA by Means of Transient BFBT Data

Directory of Open Access Journals (Sweden)

Wadim Jaeger

2013-01-01

Full Text Available In the present paper, an uncertainty and sensitivity study is performed for transient void fraction and pressure drop measurements. Two transients have been selected from the NUPEC BFBT database. The first one is a turbine trip without bypass and the second one is a trip of a recirculation pump. TRACE (version 5.0 patch 2 is used for the thermohydraulic study and SUSA and DAKOTA are used for the quantification of the model uncertainties and the evaluation of the sensitivities. As uncertain parameters geometrical values, hydraulic diameter, and wall roughness are considered while mass flow rate, power, pressure, and inlet subcooling (inlet temperature are chosen as boundary and input conditions. Since these parameters change with time, it is expected that the importance of them on pressure drop and void fraction will change, too. The results show that the pressure drop is mostly sensitive to geometrical variations like the hydraulic diameter and the form loss coefficient of the spacer grid. For low void fractions, the parameter of the highest importance is the inlet temperature/subcooling while at higher void fraction the power is also of importance.

Thermal, thermo-hydraulic and thermo-mechanic analysis for fuel elements of IEA-R1 reactor at 5MW

International Nuclear Information System (INIS)

Teixeira e Silva, A.; Silva Macedo, L.V. da

1989-01-01

In connection with the on going conversion of IEA-R1 Research Reactor, operated by IPEN-CNEN/SP, from the use of highly enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel, steady-state thermal and thermo-hydraulic analysis of both existing HEU and proposed LEU cores under 2 MW operating conditions have been carried out. Keeping in mind the possibility of power upgrading, steady-state thermal, thermo-hydraulic and thermomechanical analysis of proposed LEU core under 5 MW operating conditions have also been carried out. The thermal and thermo-hydraulic analysis at 2 MW show that the conversion of the existing HEU core to be proposed LEU core will not change the reactor safety margins. Although the upgrading of the reactor power to 5 MW will result in safety margins lower than in case of 2MW, these will be still sufficient for optimum operation and safe behaviour. The thermomechanical analysis at 5 MW show that the thermal stresses induced in the fuel element will satisfy the design limits for mechanical strenght and elastic stability. (author) [pt

Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

1998-03-01

The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

Performance of neutron kinetics models for ADS transient analyses

International Nuclear Information System (INIS)

Rineiski, A.; Maschek, W.; Rimpault, G.

2002-01-01

Within the framework of the SIMMER code development, neutron kinetics models for simulating transients and hypothetical accidents in advanced reactor systems, in particular in Accelerator Driven Systems (ADSs), have been developed at FZK/IKET in cooperation with CE Cadarache. SIMMER is a fluid-dynamics/thermal-hydraulics code, coupled with a structure model and a space-, time- and energy-dependent neutronics module for analyzing transients and accidents. The advanced kinetics models have also been implemented into KIN3D, a module of the VARIANT/TGV code (stand-alone neutron kinetics) for broadening application and for testing and benchmarking. In the paper, a short review of the SIMMER and KIN3D neutron kinetics models is given. Some typical transients related to ADS perturbations are analyzed. The general models of SIMMER and KIN3D are compared with more simple techniques developed in the context of this work to get a better understanding of the specifics of transients in subcritical systems and to estimate the performance of different kinetics options. These comparisons may also help in elaborating new kinetics models and extending existing computation tools for ADS transient analyses. The traditional point-kinetics model may give rather inaccurate transient reaction rate distributions in an ADS even if the material configuration does not change significantly. This inaccuracy is not related to the problem of choosing a 'right' weighting function: the point-kinetics model with any weighting function cannot take into account pronounced flux shape variations related to possible significant changes in the criticality level or to fast beam trips. To improve the accuracy of the point-kinetics option for slow transients, we have introduced a correction factor technique. The related analyses give a better understanding of 'long-timescale' kinetics phenomena in the subcritical domain and help to evaluate the performance of the quasi-static scheme in a particular case. One

Modelling of void formation in the subcooled boiling regime in the ATHLET code to simulate flow instability for research reactors

International Nuclear Information System (INIS)

Hainoun, A.

1996-01-01

The ATHLET thermohydraulic code was developed at the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Society for Plant and Reactor Safety) to analyse leaks and transients for power reactors. In order to extend the code's range of application to the safety analysis of research reactors, a model was implemented permitting a description of the thermodynamic non-equilibrium effects in the subcooled boiling regime. The aim of the extension is, on one hand, to cover the thermohydraulic instability which is particularly characteristic of research reactors owing to their high power densities and low system pressures and, on the other hand, to provide a consideration of the influence of the steam formed in this boiling regime on the neutron balance. The model developed takes into consideration the competing evaporation and condensation effects in the subcooled boiling regime. It describes the bubble production rate at the superheated heating surfaces as well as the subsequent condensation of the bubbles in the subcooled core flow. The installed model is validated by the recalculation of two extensive series of experiments. In the first series the McMaster experiments on axial void distribution in the subcooled boiling regime are recalculated. The recalculation shows that the extended programme is capable of calculating the axial void distribution in the subcooled boiling regime with good agreement with the data. The second series deals with KFA experiments on thermohydraulic instability (flow excursion) in the subcooled boiling regime, comprising a broad parameter range of heat flow density, inlet temperature and channel width. Recalculation of this experimental series shows that the programme extension ensures simulation of thermohydraulic instability. (orig.)

Analysis code for medium and small rupture accidents in ATR. LOTRAC/HEATUP

International Nuclear Information System (INIS)

1997-08-01

In the evaluation of thermo-hydraulic and fuel temperature transient changes in the events which are classified in medium and small rupture accidents of reactor coolant loss that is the safety evaluation event of the ATR, the analysis code for synthetic thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC and the detailed analysis code for fuel temperature HEATUP are used, respectively. By using the LOTAC, the thermo-hydraulic behavior of reactor cooling facility and the temperature behavior of fuel at the time of blow-down are analyzed, and also the characteristics of changing reactor thermal output is analyzed, considering the functioning characteristics of emergency core cooling system. Based on the data of thermo-hydraulic behavior obtained by the LOTRAC, the time of beginning the turn-around of fuel cladding tube temperature obtained by the data of ECCS pouring characteristics, the heat transfer rate after the turn-around and so on, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The LOTRAC code, the HEATUP code, various analysis models, and rupture simulation experiment are reported. (K.I.)

Physical models and numerical methods of the reactor dynamic computer program RETRAN

International Nuclear Information System (INIS)

Kamelander, G.; Woloch, F.; Sdouz, G.; Koinig, H.

1984-03-01

This report describes the physical models and the numerical methods of the reactor dynamic code RETRAN simulating reactivity transients in Light-Water-Reactors. The neutron-physical part of RETRAN bases on the two-group-diffusion equations which are solved by discretization similar to the TWIGL-method. An exponential transformation is applied and the inner iterations are accelerated by a coarse-mesh-rebalancing procedure. The thermo-hydraulic model approximates the equation of state by a built-in steam-water-table and disposes of options for the calculation of heat-conduction coefficients and heat transfer coefficients. (Author) [de

A model for steady-state and transient determination of subcooled boiling for calculations coupling a thermohydraulic and a neutron physics calculation program for reactor core calculation

International Nuclear Information System (INIS)

Mueller, R.G.

1987-06-01

Due to the strong influence of vapour bubbles on the nuclear chain reaction, an exact calculation of neutron physics and thermal hydraulics in light water reactors requires consideration of subcooled boiling. To this purpose, in the present study a dynamic model is derived from the time-dependent conservation equations. It contains new methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. Furthermore, it enables the complete two-phase flow region to be treated in a consistent manner. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement was reached. The results from the coupling of the new calculation model with a neutron kinetics program proved its suitability for the steady-state and transient calculation of reactor cores. (orig.) [de

New transient-flow modelling of a multiple-fractured horizontal well

International Nuclear Information System (INIS)

Jia, Yong-Lu; Wang, Ben-Cheng; Nie, Ren-Shi; Wang, Dan-Ling

2014-01-01

A new transient-flow modelling of a multiple-fractured horizontal well is presented. Compared to conventional modelling, the new modelling considered more practical physical conditions, such as various inclined angles for different fractures, different fracture intervals, different fracture lengths and partially penetrating fractures to formation. A kind of new mathematical method, including a three-dimensional eigenvalue and orthogonal transform, was created to deduce the exact analytical solutions of pressure transients for constant-rate production in real space. In order to consider a wellbore storage coefficient and skin factor, we used a Laplace-transform approach to convert the exact analytical solutions to the solutions in Laplace space. Then the numerical solutions of pressure transients in real space were gained using a Stehfest numerical inversion. Standard type curves were plotted to describe the transient-flow characteristics. Flow regimes were clearly identified from type curves. Furthermore, the differences between the new modelling and the conventional modelling in pressure transients were especially compared and discussed. Finally, an example application to show the accordance of the new modelling with real conditions was implemented. Our new modelling is different from, but more practical than, conventional modelling. (paper)

Methodology to carry out a sensitivity and uncertainty analysis for cross sections using a coupled model Trace-Parcs; Metodologia para realizar un analisis de sensibilidad e incertidumbre para las secciones eficaces usando un modelo acoplado TRACE-PARCS

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico); Gomez T, A. M. [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Sanchez E, V., E-mail: rf.melisa@gmail.com [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

2015-09-15

A methodology was implemented to carry out a sensitivity and uncertainty analysis for cross sections used in a coupled model for Trace/Parcs in a transient of control rod fall of a BWR-5. A model of the reactor core for the neutronic code Parcs was used, in which the assemblies located in the core are described. Thermo-hydraulic model in Trace was a simple model, where only a component type Chan was designed to represent all the core assemblies, which it was within a single vessel and boundary conditions were established. The thermo-hydraulic part was coupled with the neutron part, first for the steady state and then a transient of control rod fall was carried out for the sensitivity and uncertainty analysis. To carry out the analysis of cross sections used in the coupled model Trace/Parcs during the transient, the Probability Density Functions for 22 parameters selected from the total of neutronic parameters that use Parcs were generated, obtaining 100 different cases for the coupled model Trace/Parcs, each one with a database of different cross sections. All these cases were executed with the coupled model, obtaining in consequence 100 different output files for the transient of control rod fall doing emphasis in the nominal power, for which an uncertainty analysis was realized at the same time generate the band of uncertainty. With this analysis is possible to observe the ranges of results of the elected responses varying the selected uncertainty parameters. The sensitivity analysis complements the uncertainty analysis, identifying the parameter or parameters with more influence on the results and thus focuses on these parameters in order to better understand their effects. Beyond the obtained results, because is not a model with real operation data, the importance of this work is to know the application of the methodology to carry out the sensitivity and uncertainty analyses. (Author)

Studies on an Electromagnetic Transient Model of Offshore Wind Turbines and Lightning Transient Overvoltage Considering Lightning Channel Wave Impedance

Directory of Open Access Journals (Sweden)

Li Zhang

2017-12-01

Full Text Available In recent years, with the rapid development of offshore wind turbines (WTs, the problem of lightning strikes has become more and more prominent. In order to reduce the failure rate caused by the transient overvoltage of lightning struck offshore WTs, the influencing factors and the response rules of transient overvoltage are analyzed. In this paper, a new integrated electromagnetic transient model of offshore WTs is established by using the numerical calculation method of the electromagnetic field first. Then, based on the lightning model and considering the impedance of the lightning channel, the transient overvoltage of lightning is analyzed. Last, the electromagnetic transient model of offshore WTs is simulated and analyzed by using the alternative transients program electro-magnetic transient program (ATP-EMTP software. The influence factors of lightning transient overvoltage are studied. The main influencing factors include the sea depth, the blade length, the tower height, the lightning flow parameters, the lightning strike point, and the blade rotation position. The simulation results show that the influencing factors mentioned above have different effects on the lightning transient overvoltage. The results of the study have some guiding significance for the design of the lightning protection of the engine room.

Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility; Estudo termohidraulico de um elemento combustivel tipo MTR visando a construcao de um dispositivo de irradiacao

Energy Technology Data Exchange (ETDEWEB)

Coragem, Helio Boemer de Oliveira

1980-07-01

A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility; Estudo termohidraulico de um elemento combustivel tipo MTR visando a construcao de um dispositivo de irradiacao

Energy Technology Data Exchange (ETDEWEB)

Coragem, Helio Boemer de Oliveira

1980-07-01

A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

Thermo-hydraulic modelling of the South East Gas Pipeline System - an integrated model; Modelagem termo-hidraulica do Sistema de Gasodutos do Sudeste : um modelo integrado

Energy Technology Data Exchange (ETDEWEB)

Vianna Neto, Armando M.; Santos, Arnaldo M.; Mercon, Eduardo G. [TRANSPETRO - PETROBRAS Transportes, Rio de Janeiro, RJ (Brazil)

2003-07-01

This paper presents the development of an integrated simulation model, for the numerical calculation of thermal-hydraulic behaviors in the Brazilian southeast onshore gas pipeline flow system, remotely operated by TRANSPETRO's Gas Pipeline Control Centre (CCG). In its final application, this model is supposed to provide simulated results at the closer range to reality, in order to improve gas pipeline simulation studies and evaluations for the system in question. Considering the fact that numerical thermo-hydraulic simulation becomes the CCG's most important tool to analyze the boundary conditions to adjust the mentioned gas flow system, this paper seeks and takes aim to the optimization of the following prime attributions of a gas pipeline control centre: verification of system behaviors, face to some unit maintenance stop or procedure, programmed or not, or to some new gas outlet or inlet connection to the system; daily operational compatibility analysis between programmed and realized gas volumes; gas technical expedition and delivery analysis. Finally, all this work was idealized and carried out within the one-phase flow domain (dry gas) (author)

The simulation of CANDU fuel channel behavior in thermal transient conditions

International Nuclear Information System (INIS)

Mihalache, M.; Roth, M.; Radu, V.; Dumitrescu, I.

2005-01-01

In certain LOCA conditions into the CANDU fuel channel, is possible the ballooning of the pressure tube and the contact with the calandria tube. After the contact moment, a radial heat transfer to the moderator through the contact area is occurs. When the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. Thus, the fuel channel could lose its integrity. This paper present a computer code, DELOCA, developed in INR, which simulate the transient thermo-mechanical behaviour of CANDU fuel channel before and after contact. The code contains few models: alloy creep, heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. It was verified step by step by Contact1 and Cathena codes. In this paper, the results obtained at different temperature increasing rates are presented. Also, the contact moment for a RIH 5% postulated accident was presented. The input data was furnished by the Cathena thermo-hydraulic code. (author)

Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code

International Nuclear Information System (INIS)

Martin, L.; Saenz Tejada, P.

1993-01-01

When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

Modeling of Transient Response of the Wickless Heat Pipes

International Nuclear Information System (INIS)

Hussien, A.K.A.

2013-01-01

Thermosyphons transient response for startup from ambient temperature to steady state until shutdown conditions, is considered a stringent necessity for applications such as electronic, solar, geothermal and even nuclear reactors safety systems. This typically returns to the need to keep the temperature within certain limits before reaching critical conditions. A simple network model is derived for describing the transient response of closed two-phase thermosyphon (CTPT) at startup and shutdown states. In addition, for predicting the effect of operational characteristics of water/copper closed two-phase thermosyphon such as thermal load, filling ratio, evaporator length, and thermosyphon tube diameter. The thermosyphons operation was considered a thermal network of various components with different thermal resistances and dynamic responses. The network model consists of six sub-models. These models are pure conduction in walls of evaporator, adiabatic and condenser, and convection in evaporator pool, evaporator film, and condenser film. So, an energy balance for each sub-model was done to estimate temperatures, heat transfer coefficients, thermal resistances, time constant, and other thermal characteristics that describe the required transient response of the closed two-phase thermosyphon. Governing equations of the transient thermosyphon behavior can be simplified into a set of first-order linear ordinary differential equations. The Runge-Kutta method can be used to obtain transient thermosyphon temperatures from these equations.

Transient accelerating scalar models with exponential potentials

International Nuclear Information System (INIS)

Cui Wen-Ping; Zhang Yang; Fu Zheng-Wen

2013-01-01

We study a known class of scalar dark energy models in which the potential has an exponential term and the current accelerating era is transient. We find that, although a decelerating era will return in the future, when extrapolating the model back to earlier stages (z ≳ 4), scalar dark energy becomes dominant over matter. So these models do not have the desired tracking behavior, and the predicted transient period of acceleration cannot be adopted into the standard scenario of the Big Bang cosmology. When couplings between the scalar field and matter are introduced, the models still have the same problem; only the time when deceleration returns will be varied. To achieve re-deceleration, one has to turn to alternative models that are consistent with the standard Big Bang scenario.

Thermohydraulic behavior of liquid metal pool submitted to electronic bombardment

International Nuclear Information System (INIS)

Brun, Patrice

1998-01-01

This thesis deals with the thermohydraulics of liquid metal molten by an electron beam. We study the relationship between the liquid metal pool and the vapor rate. The aim is to find good conditions increasing the metal vapor rate. In first place, energy losses are identified. Mains are convection (buoyancy and thermo-capillary) strengthen by the deformation of the molten pool. The first action is to reduce the liquid interface deformation with a transient spot realized by scanning the electron beam. I find that in this case, the optimum vapor rate is obtained when the crossing time of the beam is smaller than characteristic time of formation of the cavity, but greater than the heating time of the surface. Secondly, I impose forces to change the morphology of the flow. Two actions are tried: magnetic field application and rotating motion of the crucible. External magnetic field application may reduce convective flow, by the creation of a magnetic brake. But in my experiment, magnetic field deteriorates electron beam before to be effective. Results obtained by the rotating motion of the crucible approve this choice to reduce energy losses and increase vapor rate. This growth of vapor rate is due to an expansion of the emitted vapor source and an increase of the central temperature of the molten pool. Nevertheless with the increase of the rotation velocity and after the optimum vapor rate, I note that the flow is not axisymmetric. My observation give to think about instabilities that are developed by baroclinic waves. The comparison of my works with the Eady's linear theory gives good results. (author) [fr

Thermohydraulic calculations in rectangular channels for RA-6 type reactors with transition regime

International Nuclear Information System (INIS)

Sillin, N; Vertullo, A.; Masson, V.; Hilal, R

2009-01-01

In August 2000 and within the framework of the RA-6 core conversion from high to low enrichment (20%), a preliminary analysis was performed to evaluate the maximum power that the reactor could operate with the new kernel without makeing substantial changes. This meant keeping intact, for example, the concrete shield of the pool and the nucleus inlet and outlet pipes embedded in the walls. Preliminary results indicated that for these boundary conditions a maximum power of about 3 MWt could be achieved. In August 2005 the project was resumed and new calculations performed taking as a starting point the ECBE plate fuel element(U3O8-Al). A core was developed with cooling channle widths of 2.6 mm for the control fuel elements and 2.7 mm for standard fuel elements. The thermo-hydraulic calculation puts in evidence that coolant flow into the core was in the transitional regime for the vast majority of configurations. While TERMIC code, used for thermo-hydraulic design, has been extensively tested and validated for use in research reactors under turbulent and laminar flows, this is not so for transition conditions. The transition regime is strongly dependent on conditions such as flow inlet characteristics, channel geometry, etc.. and therefore there are no reliable correlations for general use. For this reason we found it convenient to carry out experiments simulating the working conditions in order to adjust the code results with experimental data. In the present work we show the experimental results, the simulation of the experiences using the TERMIC code, and the adjustments made to the correlations used by the code so that it can be applied to the thermo-hydraulic design of the new core. [es

Experimental study and modelling of transient boiling

International Nuclear Information System (INIS)

Baudin, Nicolas

2015-01-01

A failure in the control system of the power of a nuclear reactor can lead to a Reactivity Initiated Accident in a nuclear power plant. Then, a power peak occurs in some fuel rods, high enough to lead to the coolant film boiling. It leads to an important increase of the temperature of the rod. The possible risk of the clad failure is a matter of interest for the Institut de Radioprotection et de Securite Nucleaire. The transient boiling heat transfer is not yet understood and modelled. An experimental set-up has been built at the Institut de Mecanique des Fluides de Toulouse (IMFT). Subcooled HFE-7000 flows vertically upward in a semi annulus test section. The inner half cylinder simulates the clad and is made of a stainless steel foil, heated by Joule effect. Its temperature is measured by an infrared camera, coupled with a high speed camera for the visualization of the flow topology. The whole boiling curve is studied in steady state and transient regimes: convection, onset of boiling, nucleate boiling, critical heat flux, film boiling and rewetting. The steady state heat transfers are well modelled by literature correlations. Models are suggested for the transient heat flux: the convection and nucleate boiling evolutions are self-similar during a power step. This observation allows to model more complex evolutions, as temperature ramps. The transient Hsu model well represents the onset of nucleate boiling. When the intensity of the power step increases, the film boiling begins at the same temperature but with an increasing heat flux. For power ramps, the critical heat flux decreases while the corresponding temperature increases with the heating rate. When the wall is heated, the film boiling heat transfer is higher than in steady state but it is not understood. A two-fluid model well simulates the cooling film boiling and the rewetting. (author)

Thermo-hydraulic analysis of the cool-down of the EDIPO test facility

Science.gov (United States)

Lewandowska, Monika; Bagnasco, Maurizio

2011-09-01

The first cool-down of the EDIPO (European DIPOle) test facility is foreseen to take place in 2011 by means of the existing 1.2 kW cryoplant at EPFL-CRPP Villigen. In this work, the thermo-hydraulic analysis of the EDIPO cool-down is performed in order both to assess the its duration and to optimize the procedure. The cool-down is driven by the helium flowing in both the outer cooling channel and in the windings connected hydraulically in parallel. We take into account limitations due to the pressure drop in the cooling circuit and the refrigerator capacity as well as heat conduction in the iron yoke. Two schemes of the hydraulic cooling circuit in the EDIPO windings are studied (coils connected in series and coils connected in parallel). The analysis is performed by means of an analytical model complemented by and numerical model. The results indicate that the cool-down to 5 K can be achieved in about 12 days.

Convergence analysis of neutronic/thermohydraulic coupling behavior of SCWR

International Nuclear Information System (INIS)

Liu, Shichang; Cai, Jiejin

2013-01-01

The neutronic/thermohydraulic coupling (N–T coupling) calculations play an important role in core design and stability analysis. The traditional iterative method is not applicable for some new reactors (such as supercritical water-cooled reactor) which have intense N–T coupling behavior. In this paper, the mathematical model of N–T coupling based on fixed point theory is established firstly, with the convergent criterion, which can show the real-time convergence situation of iteration. Secondly, the self-adaptive relaxation factor and corresponding algorithm are proposed. Thirdly, the convergence analysis of the method of self-adaptive relaxation factor and common relaxation iteration has been performed, based on three calculation examples of SCWR fuel assembly. The results show that the proposed algorithm can efficiently reduce the calculation time and be adapted to different coupling cases and different initial distribution. It is easy to program, providing convenience for reactor design and analysis. This research also provides the theoretical basis for further study of N–T coupling behavior of new reactors such as SCWR

Aeroelastic Modeling of a Nozzle Startup Transient

Science.gov (United States)

Wang, Ten-See; Zhao, Xiang; Zhang, Sijun; Chen, Yen-Sen

2014-01-01

Lateral nozzle forces are known to cause severe structural damage to any new rocket engine in development during test. While three-dimensional, transient, turbulent, chemically reacting computational fluid dynamics methodology has been demonstrated to capture major side load physics with rigid nozzles, hot-fire tests often show nozzle structure deformation during major side load events, leading to structural damages if structural strengthening measures were not taken. The modeling picture is incomplete without the capability to address the two-way responses between the structure and fluid. The objective of this study is to develop a tightly coupled aeroelastic modeling algorithm by implementing the necessary structural dynamics component into an anchored computational fluid dynamics methodology. The computational fluid dynamics component is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, while the computational structural dynamics component is developed under the framework of modal analysis. Transient aeroelastic nozzle startup analyses at sea level were performed, and the computed transient nozzle fluid-structure interaction physics presented,

Experimental study on thermo-hydraulic instability on reduced-moderation natural circulation BWR concept

International Nuclear Information System (INIS)

Watanabe, Noriyuki; Subki, M.H.; Kikura, Hiroshige; Aritomi, Masanori

2003-01-01

Reduced-moderation natural circulation BWR has been promoted to solve the recent challenges in BWR nuclear power technology problems as one of advanced small and medium-sized reactors equipped with the passive safety features in conformity with the natural law. However, the elimination of recirculation pumps and a high-density core due to the increase of conversion ratio could cause various thermo-hydraulic instabilities especially during the start-up stage. The occurrences of the thermo-hydraulic instabilities are not desirable and it is one of the main challenges in establishing reduced-moderation natural circulation BWR as a commercial reactor. The purpose of this present study is to experimentally investigate the driving mechanism of the thermo-hydraulic instabilities and the effect of system pressure on the unstable flow patterns. Hence, as the fundamental research for this study, a natural circulation loop that carries boiling fluid with parallel boiling channel has been constructed. Channel gap that has been set at 2 mm in order to simulate reduced-moderation reactor core. Pressure ranges of 0.1 up to 0.7 MPa, input heat flux range of 0 ou to 577 kW/m 2 , and inlet subcooling temperatures of 5, 10, and 15 K respectively, are imposed in the experiments. This experiment clarifies that changes in unstable flow patterns with increase in heat flux can be classified into two in response to system pressure range. In case of atmospheric pressure, unstable flow patters has been classified in beyond order, (1) in-phase geysering, (2) transition oscillation combined with both features of in-phase geysering and natural circulation oscillation, (3) natural circulation oscillation induced by hydrostatic head fluctuation, (4) density wave oscillation, and finally (5) stable boiling two-phase flow. On the other hand, in the system pressure range from 0.2 to 0.7 MPa, unstable patters have been dramatically changed in the following order (1) out-of-phase geysering, (2

Model tests in RAMONA and NEPTUN

International Nuclear Information System (INIS)

Hoffmann, H.; Ehrhard, P.; Weinberg, D.; Carteciano, L.; Dres, K.; Frey, H.H.; Hayafune, H.; Hoelle, C.; Marten, K.; Rust, K.; Thomauske, K.

1995-01-01

In order to demonstrate passive decay heat removal (DHR) in an LMR such as the European Fast Reactor, the RAMONA and NEPTUN facilities, with water as a coolant medium, were used to measure transient flow data corresponding to a transition from forced convection (under normal operation) to natural convection under DHR conditions. The facilities were 1:20 and 1:5 models, respectively, of a pool-type reactor including the IHXs, pumps, and immersed coolers. Important results: The decay heat can be removed from all parts of the primary system by natural convection, even if the primary fluid circulation through the IHX is interrupted. This result could be transferred to liquid metal cooling by experiments in models with thermohydraulic similarity. (orig.)

Transient freezing of molten salts in pipe-flow systems: Application to the direct reactor auxiliary cooling system (DRACS)

International Nuclear Information System (INIS)

Le Brun, N.; Hewitt, G.F.; Markides, C.N.

2017-01-01

Highlights: • A thermo-hydraulic model has been proposed to simulate the transient freezing of molten salts in complex piping systems. • The passive safety system DRACS in Generation-IV, molten salt reactor is susceptible to failure due to salt freezing. • For the prototypical 0.2 MW reactor considered in this study considerable freezing occurs after 20 minutes leading to reactor temperatures above 900 °C within 4 hours. • Conservative criteria for the most important/least known variables in the design of DRACS have been discussed. • Over-conservative approaches in designing the NDHX should be used with caution as they can promote pipe clogging due to freezing. - Abstract: The possibility of molten-salt freezing in pipe-flow systems is a key concern for the solar-energy industry and a safety issue in the new generation of molten-salt reactors, worthy of careful consideration. This paper tackles the problem of coolant solidification in complex pipe networks by developing a transient thermohydraulic model and applying it to the ‘Direct Reactor Auxiliary Cooling System’ (DRACS), the passive-safety system proposed for the Generation-IV molten-salt reactors. The results indicate that DRACS, as currently envisioned, is prone to failure due to freezing in the air/molten-salt heat exchanger, which can occur after approximately 20 minutes, leading to reactor temperatures above 900 °C within 4 hours. The occurrence of this scenario is related to an unstable behaviour mode of DRACS in which newly formed solid-salt deposit on the pipe walls acts to decrease the flow-rate in the secondary loop, facilitating additional solid-salt deposition. Conservative criteria are suggested to facilitate preliminary assessments of early-stage DRACS designs. The present study is, to the knowledge of the authors, the first of its kind in serving to illustrate possible safety concerns in molten-salt reactors, which are otherwise considered very safe in the literature. Furthermore

Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

International Nuclear Information System (INIS)

1997-08-01

In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

Solving linear systems in FLICA-4, thermohydraulic code for 3-D transient computations

International Nuclear Information System (INIS)

Allaire, G.

1995-01-01

FLICA-4 is a computer code, developed at the CEA (France), devoted to steady state and transient thermal-hydraulic analysis of nuclear reactor cores, for small size problems (around 100 mesh cells) as well as for large ones (more than 100000), on, either standard workstations or vector super-computers. As for time implicit codes, the largest time and memory consuming part of FLICA-4 is the routine dedicated to solve the linear system (the size of which is of the order of the number of cells). Therefore, the efficiency of the code is crucially influenced by the optimization of the algorithms used in assembling and solving linear systems: direct methods as the Gauss (or LU) decomposition for moderate size problems, iterative methods as the preconditioned conjugate gradient for large problems. 6 figs., 13 refs

Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

International Nuclear Information System (INIS)

Kobayashi, Michiyuki; Murata, Hiroyuki; Sawada, Kenichi; Inasaka, Fujio; Aya, Izuo; Shiozaki, Koki

1999-01-01

By inputting the experimental data, information and others on thermo-hydraulic characteristics of integrated ship propulsion reactor accumulated hitherto by the Ship Research Institute and some recent cooperation results into the nuclear ship engineering simulation system, it was conducted not only to contribute an improvement study on next ship reactor by executing general analysis and evaluation on motion characteristics under ship body motion conditions, safety at accidents, and others of the integrated ship reactor but also to investigate and prepare some measures to apply fundamental experiment results based on obtained here information to safety countermeasure of the nuclear ships. In 1997 fiscal year, on safety of the integrated ship propulsion reactor loading nuclear ship, by adding experimental data on unstable flow analysis and information on all around of the analysis to general data base fundamental program, development to intellectual data base program was intended; on effect of pulsation flow on thermo-hydraulic characteristics of ship propulsion reactor; after pulsation flow visualization experiment, experimental equipment was reconstructed into heat transfer type to conduct numerical analysis of pulsation flow by confirming validity of numerical analysis code under comparison with the visualization experiment results; and on thermo-hydraulic behavior in storage container at accident of active safety type ship propulsion reactor; a flashing vibration test using new apparatus finished on its higher pressurization at last fiscal year to examine effects of each parameter such as radius and length of exhausting nozzle and pool water temperature. (G.K.)

Thermomechanical analysis of a SPX 1 type sodium-air heat exchanger. Analisi termomeccanica di scambiatore sodio-aria (tipo SPX 1): comportamento della macchina e dei relativi tubi alettati

Energy Technology Data Exchange (ETDEWEB)

Cesari, F; Menghini, S; Piolanti, G

1981-08-15

A description is given of a mathematical model for simulating the thermohydraulics of a heat exchanger. Temperature distribution is determined for stationary and transient function. An analysis is made of the pipelines in which the sodium circulates. 9 references.

Computational model for transient studies of IRIS pressurizer behavior

International Nuclear Information System (INIS)

Rives Sanz, R.; Montesino Otero, M.E.; Gonzalez Mantecon, J.; Rojas Mazaira, L.

2014-01-01

International Reactor Innovative and Secure (IRIS) excels other Small Modular Reactor (SMR) designs due to its innovative characteristics regarding safety. IRIS integral pressurizer makes the design of larger pressurizer system than the conventional PWR, without any additional cost. The IRIS pressurizer volume of steam can provide enough margins to avoid spray requirement to mitigate in-surge transient. The aim of the present research is to model the IRIS pressurizer's dynamic using the commercial finite volume Computational Fluid Dynamic code CFX 14. A symmetric tridimensional model equivalent to 1/8 of the total geometry was adopted to reduce mesh size and minimize processing time. The model considers the coexistence of three phases: liquid, steam, and vapor bubbles in liquid volume. Additionally, it takes into account the heat losses between the pressurizer and primary circuit. The relationships for interfacial mass, energy, and momentum transport are programmed and incorporated into CFX by using expressions in CFX Command Language (CCL) format. Moreover, several additional variables are defined for improving the convergence and allow monitoring of boron dilution sequences and condensation-evaporation rate in different control volumes. For transient states a non - equilibrium stratification in the pressurizer is considered. This paper discusses the model developed and the behavior of the system for representative transients sequences such as the in/out-surge transients and boron dilution sequences. The results of analyzed transients of IRIS can be applied to the design of pressurizer internal structures and components. (author)

The modelling of BLEVE fireball transients

Energy Technology Data Exchange (ETDEWEB)

Shield, S.R. [Shell Research Ltd., Chester (United Kingdom). Thornton Research Centre

1995-12-31

An existing physically based BLEVE fireball model has been developed to predict the shape of the transient heat pulse to a receiver, to model ``cold`` BLEVEs, and to assess the consequences to structures and people. This has been achieved by finding a correlation to predict the size of a characteristic liquid drop within the flashing cloud. These drops burn out to predict the time to fireball break-up and extinction, and their height is tracked to find the rise of the fireball, agreement with small and large scale data being satisfactory. The model is semi-empirical in that, when the level of pre-heating is low, a heat balance cannot predict the development of fireball temperature with time since the fuel is pyrolising. Also, the modelling of the transition from ``cold`` BLEVE fireball to pool fire is a cautious best estimate. Using both traditional and more recent models, the prediction of pain, burns and fatality have been incorporated into the model. Whatever criterion for a safety distance is used, the effect of modelling the transient is to reduce the calculated safety distances by up to a factor of two, which brings fatality predictions for real incidents much more in line with the historical record. (author)

A transient single particle model under FCI conditions

Institute of Scientific and Technical Information of China (English)

LI Xiao-Yan; SHANG Zhi; XU Ji-Jun

2005-01-01

The paper is focused on the coupling effect between film boiling heat transfer and evaporation drag around a hot-particle in cold liquid. Based on the continuity, momentum and energy equations of the vapor film, a transient two-dimensional single particle model has been established. This paper contains a detailed description of HPMC (High-temperature Particle Moving in Coolant) model for studying some aspects of the premixing stage of fuel-coolant interactions (FCIs). The transient process of high-temperature particles moving in coolant can be simulated. Comparisons between the experiment results and the calculations using HPMC model demonstrate that HPMC model achieves a good agreement in predicting the time-varying characteristic of high-temperature spheres moving in coolant.

Plant dynamics and safety analysis of fast breeder reactors

International Nuclear Information System (INIS)

Ertel, V.

1982-01-01

Some general features of sodium cooled fast breeders which influence the thermohydraulics and differ from LWR'S are discussed. Using the SNR-300 as a reference, some thermohydraulic transients from normal operation and from design accidents are presented. (orig.)

Possibilities of optimizing non-nuclear simulation of pressurized water reactor transients

International Nuclear Information System (INIS)

Silva Filho, E.

1985-01-01

The GKSS-Forschungszentrum Geesthacht GmbH has instituted the concept of a scaled test facility (volume scale factor of 1/100) of a typical PWR of the 1 300 MWe class for the purpose of studying small breaks Loss-of-Coolant Accidents (LOCA) and transients. Having in mind the goal of an optimization of this concept has been choosen a station blackout with and without reactor shutdown and a small break LOCA in a primary loop piping to investigate the thermohydraulic behaviour of the test facility in comparison to the reactor plant. The computer code RELAP 5/MOD 1 has been utilized to compare the test facility behaviour with the reactor plant one. Recommendations are given for minimization of distortions between test facility and reactor plant. (orig./HP) [de

Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

International Nuclear Information System (INIS)

Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

1988-04-01

As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

Thermohydraulic calculations of PWR primary circuits

International Nuclear Information System (INIS)

Botelho, D.A.

1984-01-01

Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

Thermo-hydraulic design of earth-air heat exchangers

Energy Technology Data Exchange (ETDEWEB)

Paepe, M. de [Ghent University (Belgium). Department of Flow, Heat and Combustion Mechanics; Janssens, A. [Ghent University (Belgium). Department of Architecture and Urbanism

2003-05-01

Earth-air heat exchangers, also called ground tube heat exchangers, are an interesting technique to reduce energy consumption in a building. They can cool or heat the ventilation air, using cold or heat accumulated in the soil. Several papers have been published in which a design method is described. Most of them are based on a discretisation of the one-dimensional heat transfer problem in the tube. Three-dimensional complex models, solving conduction and moisture transport in the soil are also found. These methods are of high complexity and often not ready for use by designers. In this paper, a one-dimensional analytical method is used to analyse the influence of the design parameters of the heat exchanger on the thermo-hydraulic performance. A relation is derived for the specific pressure drop, linking thermal effectiveness with pressure drop of the air inside the tube. The relation is used to formulate a design method which can be used to determine the characteristic dimensions of the earth-air heat exchanger in such a way that optimal thermal effectiveness is reached with acceptable pressure loss. The choice of the characteristic dimensions, becomes thus independent of the soil and climatological conditions. This allows designers to choose the earth-air heat exchanger configuration with the best performance. (author)

Thermo-hydraulic design of earth-air heat exchangers

Energy Technology Data Exchange (ETDEWEB)

De Paepe, M. [Department of Flow, Heat and Combustion Mechanics, Ghent University, Ghent (Belgium); Janssens, A. [Department of Architecture and Urbanism, Ghent University, Ghent (Belgium)

2003-07-01

Earth-air heat exchangers, also called ground tube heat exchangers, are an interesting technique to reduce energy consumption in a building. They can cool or heat the ventilation air, using cold or heat accumulated in the soil. Several papers have been published in which a design method is described. Most of them are based on a discretisation of the one-dimensional heat transfer problem in the tube. Three-dimensional complex models, solving conduction and moisture transport in the soil are also found. These methods are of high complexity and often not ready for use by designers. In this paper, a one-dimensional analytical method is used to analyse the influence of the design parameters of the heat exchanger on the thermo-hydraulic performance. A relation is derived for the specific pressure drop, linking thermal effectiveness with pressure drop of the air inside the tube. The relation is used to formulate a design method which can be used to determine the characteristic dimensions of the earth-air heat exchanger in such a way that optimal thermal effectiveness is reached with acceptable pressure loss. The choice of the characteristic dimensions, becomes thus independent of the soil and climatological conditions. This allows designers to choose the earth-air heat exchanger configuration with the best performance. (author)

Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

2004-01-01

Requirements of neutron, thermohydraulic and safety analysis calculation are very important because of issuing new version of SAR for DNRR, research on construction of new research reactor and nuclear power plant. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor has been done in the frame work of research theme in the year 2002-2003. The purposes of the research are maintaining safety operation of the DNRR and enhancement of man power and calculation and safety analysis tool potential. (author)

Flow inversion and natural convection in a MTR (Materials Testing Reactor)

International Nuclear Information System (INIS)

Gimenez, M.O.; Clausse, A.

1990-01-01

The thermohydraulic evolution of a refrigerating channel of the MTR (Materials Testing Reactors) RA-6 reactor's core, at the Bariloche Atomic Center, has been studied during the transient caused by the primary system's pump decommissioning. This transient constitutes one of the reactor's operating power boundaries due to the maximum temperature permissible in fuel plates. The problem regarding the thermohydraulic code altered for the rectangular geometry calculation characteristic of the MTR design is analyzed. (Author) [es

The SABRE code for fuel rod cluster thermohydraulics

International Nuclear Information System (INIS)

Macdougall, J.D.; Lillington, J.N.

1984-01-01

This paper describes the capabilities of the SABRE code for the calculation of single phase and two phase fluid flow and temperature in fuel pin bundles, discusses the methods used in the modelling and solution of the problem, and presents some results including comparison with experiments. The SABRE code permits calculation of steady-state or transient, single or two phase flows and the geometrical options include general representation of grids, wire wraps, multiple blockages, bowed pins, etc. The derivation and solution of the difference equations is discussed. Emphasis is given to the derivation of the spatial differences in triangular subchannel geometry, and the use of central, upward or vector upwind schemes. The method of solution of the difference equations is described for both steady state and transient problems. Together with these topics we consider the problems involved in turbulence modelling and how it is implemented in SABRE. This includes supporting work with a fine scale curvilinear coordinate programme to provide turbulence source data. The problem of modelling boiling flows is discussed, with particular reference to the numerical problems caused by the rapid density change on boiling. The final part of the paper presents applications of the code to the analysis of blockage situations, the study of flow and power transients and analysis of natural circulation within clusters to demonstrate the scope of the code and compare with available experimental results. The comparisons include the calculation of a flow pressure drop characteristic of a boiling channel showing the Ledinegg instability, examples of overpower and flow rundown transients which lead to coolant boiling, and calculation of natural circulation within a rod cluster. (orig./GL)

Development and verification of a three-dimensional core model for WWR type reactors and its coupling with the accident code ATHLET. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Lucas, D.; Mittag, S.; Rohde, U.

1995-04-01

The main goal of the project was the coupling of the 3D core model DYN3D for Russian VVER-type reactors, which has been developed in the RCR, with the thermohydraulic code ATHLET. The coupling has been realized on two basically different ways: - The implementation of only the neutron kinetics model of DYN3D into ATHLET (internal coupling), - the connection of the complete DYN3D core model including neutron kinetics, thermohydraulics and fuel rod model via data interfaces at the core top and bottom (external coupling). For the test of the coupling, comparative calculations between internal and external coupling versions have been carried out for a LOCA and a reactivity transient. Complementary goals of the project were: - The development of a DYN3D version for burn-up calculations, - the verification of DYN3D on benchmark tasks and experimental data on fuel rod behaviour, - a study on the extension of the neutron-physical data base. The project contributed to the development of advanced tools for the safety analysis of VVER-type reactors. Future work is aimed to the verification of the coupled code complex DYN3D-ATHLET. (orig.) [de

Preliminary study of the thermo-hydraulic behaviour of the binary breeder reactor

International Nuclear Information System (INIS)

Silveira Luz, M. da; Ferreira, W.J.

1984-06-01

Continuing the development of the Binary Breeder Reactor, its physical configuration and the advantages of differents types of spacers are analysed. In order to simulate the thermo-hydraulic behaviour and obtain data for a preliminary evaluation of the core geometry, the COBRA III C code was used to study the effects of the lenght and diameter of the fuel element, the coolant inlet temperature, the system pressure, helicoidal pitch and the pitch to diameter ratio. (Author) [pt

Neutronics and thermohydraulics of the reactor C.E.N.E. Part II; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte II

Energy Technology Data Exchange (ETDEWEB)

Caro, R

1976-07-01

In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

Energy Technology Data Exchange (ETDEWEB)

Bradley K. Heath

2014-03-01

This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

Modeling of environmentally induced transients within satellites

Science.gov (United States)

Stevens, N. John; Barbay, Gordon J.; Jones, Michael R.; Viswanathan, R.

1987-01-01

A technique is described that allows an estimation of possible spacecraft charging hazards. This technique, called SCREENS (spacecraft response to environments of space), utilizes the NASA charging analyzer program (NASCAP) to estimate the electrical stress locations and the charge stored in the dielectric coatings due to spacecraft encounter with a geomagnetic substorm environment. This information can then be used to determine the response of the spacecraft electrical system to a surface discharge by means of lumped element models. The coupling into the electronics is assumed to be due to magnetic linkage from the transient currents flowing as a result of the discharge transient. The behavior of a spinning spacecraft encountering a severe substorm is predicted using this technique. It is found that systems are potentially vulnerable to upset if transient signals enter through the ground lines.

Accelerating transient simulation of linear reduced order models.

Energy Technology Data Exchange (ETDEWEB)

Thornquist, Heidi K.; Mei, Ting; Keiter, Eric Richard; Bond, Brad

2011-10-01

Model order reduction (MOR) techniques have been used to facilitate the analysis of dynamical systems for many years. Although existing model reduction techniques are capable of providing huge speedups in the frequency domain analysis (i.e. AC response) of linear systems, such speedups are often not obtained when performing transient analysis on the systems, particularly when coupled with other circuit components. Reduced system size, which is the ostensible goal of MOR methods, is often insufficient to improve transient simulation speed on realistic circuit problems. It can be shown that making the correct reduced order model (ROM) implementation choices is crucial to the practical application of MOR methods. In this report we investigate methods for accelerating the simulation of circuits containing ROM blocks using the circuit simulator Xyce.

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)

Modeling transient radiation effects in power MOSFETS

International Nuclear Information System (INIS)

Hoffman, J.R.; Hall, W.E.; Dunn, D.E.

1987-01-01

Using standard device specifications and simple assumptions, the transient radiation response of VDMOS MOSFETs can be modeled in a standard circuit analysis program. The device model consists of a body diode, a parasitic bipolar transistor, and elements to simulate high-current reduced breakdown. The attached photocurrent model emulates response to any pulse shape and accounts for bias-dependent depletion regions. The model can be optimized to best fit available test data

On the empirical relevance of the transient in opinion models

International Nuclear Information System (INIS)

Banisch, Sven; Araujo, Tanya

2010-01-01

While the number and variety of models to explain opinion exchange dynamics is huge, attempts to justify the model results using empirical data are relatively rare. As linking to real data is essential for establishing model credibility, this Letter develops an empirical confirmation experiment by which an opinion model is related to real election data. The model is based on a representation of opinions as a vector of k bits. Individuals interact according to the principle that similarity leads to interaction and interaction leads to still more similarity. In the comparison to real data we concentrate on the transient opinion profiles that form during the dynamic process. An artificial election procedure is introduced which allows to relate transient opinion configurations to the electoral performance of candidates for which data are available. The election procedure based on the well-established principle of proximity voting is repeatedly performed during the transient period and remarkable statistical agreement with the empirical data is observed.

On the empirical relevance of the transient in opinion models

Energy Technology Data Exchange (ETDEWEB)

Banisch, Sven, E-mail: sven.banisch@universecity.d [Mathematical Physics, Physics Department, Bielefeld University, 33501 Bielefeld (Germany); Institute for Complexity Science (ICC), 1249-078 Lisbon (Portugal); Araujo, Tanya, E-mail: tanya@iseg.utl.p [Research Unit on Complexity in Economics (UECE), ISEG, TULisbon, 1249-078 Lisbon (Portugal); Institute for Complexity Science (ICC), 1249-078 Lisbon (Portugal)

2010-07-12

While the number and variety of models to explain opinion exchange dynamics is huge, attempts to justify the model results using empirical data are relatively rare. As linking to real data is essential for establishing model credibility, this Letter develops an empirical confirmation experiment by which an opinion model is related to real election data. The model is based on a representation of opinions as a vector of k bits. Individuals interact according to the principle that similarity leads to interaction and interaction leads to still more similarity. In the comparison to real data we concentrate on the transient opinion profiles that form during the dynamic process. An artificial election procedure is introduced which allows to relate transient opinion configurations to the electoral performance of candidates for which data are available. The election procedure based on the well-established principle of proximity voting is repeatedly performed during the transient period and remarkable statistical agreement with the empirical data is observed.

Development of LILAC-meltpool for the thermo-hydraulic analysis of core melt relocated in a reactor vessel

International Nuclear Information System (INIS)

Kim, Jong Tae; Kim, Sang Baik; Kim, Hee Dong

2002-03-01

LILAC-meltpool has been developed to study thermo-hydraulic behavior of molten pool and thermal behavior of vessel wall during severe accident. To validate LILAC-meltpool code several two and three dimensional thermo-hydraulic problems were selected and solved. The benchmark problems have experimental results or verified numerical results. Through the validation it was found that LILAC-meltpool reproduces very accurate numerical results. Two-layered semicircular pool was solved to study thermal and hydraulic characteristics of pool stratification. The LAVA experiment using alumina/ferrite molten pool was calculated and compared with computed results. Cooling of alumina/ferrite two-layered pool was affected by stratification. In the numerical results temperature of vessel inner was highest at a location below the interface. Crust was developed from upper surface and lower outer surface, but in the area near the interface corium simulant existed as molten state for long time. LAVA-4 experiment was studied using gap-cooling model in LILAC-meltpool code. Temperature increase of LAVA vessel after alumina melt relocation was strongly dependent on gap formation mechanism. Calculated cooling rates of the vessel were very similar to experimental results. For LAVA experiments which do not have heat generation coolant penetrates easily into a gap and it is found that gap-cooling is very effective for cooling of vessel, but it is thought that coolant penetration could be limited near upper part of gap because of decay heat and high temperature of corium crust

A transient model to the thermal detonation

International Nuclear Information System (INIS)

Karachalios, K.

1987-04-01

The model calculates the escalation dynamics and the long time behavior of thermal detonation waves depending on the initial and boundary conditions (data of the premixture, ignition at a solid wall or at an open end, etc.). Especially, for a given mixture and a certain fragmentation behavior more than one stable steady-state cases resulted, depending on the applied ignition energy. Investigations showed a very good consistency between the transient model and a steady-state model which is based on the same physical description and includes an additional stability criterion. Also the influence of effects such as e.g. non-homogeneous coolant heating, spherical instead of plane wave propagation and inhomogeneities of the premixture on the development of the wave were investigated. Comparison calculations with large scale experiments showed that they can be well explained by means of the thermal detonation theory, especially considering the transient phase of the wave development. (orig./HP) [de

Complete wind farm electromagnetic transient modelling for grid integration studies

International Nuclear Information System (INIS)

Zubia, I.; Ostolaza, X.; Susperregui, A.; Tapia, G.

2009-01-01

This paper presents a modelling methodology to analyse the impact of wind farms in surrounding networks. Based on the transient modelling of the asynchronous generator, the multi-machine model of a wind farm composed of N generators is developed. The model incorporates step-up power transformers, distribution lines and surrounding loads up to their connection to the power network. This model allows the simulation of symmetric and asymmetric short-circuits located in the distribution network and the analysis of transient stability of wind farms. It can be also used to study the islanding operation of wind farms

The development of the thermohydraulic analysis code for the passive containment cooling system: PCCSAC

International Nuclear Information System (INIS)

Wang Jianyu; Zhang Shenru; Min Yuanyou

1994-01-01

To estimate the performance of the passive containment cooling system (PCCS) of the AC-600 nuclear power plant, the PCCSAC code is developed currently by the jointed efforts between Tsinghua University and NPIC. Different features on the passive behavior of the system and the main components of the containment are considered in the code which is needed by the further AC-600 R and D Program. With a brief description of the AC-600 passive containment cooling system and components, the main thermohydraulic models and numerical scheme used in the PCCSAC code are introduced and the selected results of the verification and the prediction for the performance of the AC-600 passive containment cooling system under LOCA and a steam line break accident are presented to preliminarily demonstrate the applicability and reliability of the PCCSAC model. The current PCCSAC model is conservative and a further 2-D PCCSAC version is under consideration in addition to provide the database for models from some tests associated with the components and systems unique to AC-600 nuclear power plant to meet the requirement of the more realistic modelization for the AC-600 passive containment cooling system. (author)

Application of an estimation model to predict future transients at US nuclear power plants

International Nuclear Information System (INIS)

Hallbert, B.P.; Blackman, H.S.

1987-01-01

A model developed by R.A. Fisher was applied to a set of Licensee Event Reports (LERs) summarizing transient initiating events at US commercial nuclear power plants. The empirical Bayes model was examined to study the feasibility of estimating the number of categories of transients which have not yet occurred at nuclear power plants. An examination of the model's predictive ability using an existing sample of data provided support for use of the model to estimate future transients. The estimate indicates that an approximate fifteen percent increase in the number of categories of transient initiating events may be expected during the period 1983--1993, assuming a stable process of transients. Limitations of the model and other possible applications are discussed. 10 refs., 1 fig., 3 tabs

Modifications of the bubble rise model and heat transfer model used in RELAP 4/Mod 5 computer code for transient analysis

International Nuclear Information System (INIS)

Scharfmann, E.; Silva, D.E. da

1981-01-01

The modifications on the phase separation model and heat tranfer model in Relap4/Mod 5 computer code, in order to make more realistic estimates of the core thermohydraulic behavior submitted to a loss of coolant accident. This research is directed to the accident analysis caused by small breaks in the primary circuits of PWR plants, where two-phase flow occurs most of the time. Calculation have been performed with the help of the original version of Relap code, as well as the version containing the proposed modifications on this work. Comparing one results with the original ones, we arrive at the conclusion that our results show more conservative values of core pressure and coolant temperature, while the peak values of fuel temperature are not exceeded. (Author) [pt

Reliability analysis of PWR thermohydraulic design by the Monte Carlo method

International Nuclear Information System (INIS)

Silva Junior, H.C. da; Berthoud, J.S.; Carajilescov, P.

1977-01-01

The operating power level of a PWR is limited by the occurence of DNB. Without affecting the safety and performance of the reactor, it is possible to admit failure of a certain number of core channels. The thermohydraulic design, however, is affect by a great number of uncertainties of deterministic or statistical nature. In the present work, the Monte Carlo method is applied to yield the probability that a number F of channels submitted to boiling crises will not exceed a number F* previously given. This probability is obtained as function of the reactor power level. (Author) [pt

DRESDYN: A new platform for liquid metal thermohydraulic studies and measurement technique developments

International Nuclear Information System (INIS)

Gerbeth, Gunter; Eckert, Sven; Stefani, Frank; Gundrum, Thomas

2013-01-01

DRESDYN: General features. DRESDYN: DREsden Sodium facility for DYNamo and thermohydraulic studies. A large-scale new infrastructure for liquid metal experiments. Features: • New building ~ 500 m 2 ; • Total sodium inventory: 12-15 tons; • Precession driven experiment with separate strong basement and containment for Argon flooding; • Big hall for SFR related experiments, including ISI, a sodium loop, X-ray lab; • Financing is given, construction will start soon in spring 2013; • First experiments 2015 (hopefully...)

Thermohydraulic characteristics analysis of natural convective cooling mode on the steady state condition of upgraded JRR-3 core, using COOLOD-N code

International Nuclear Information System (INIS)

Kaminaga, Masanori; Watanabe, Shukichi; Ando, Hiroei; Sudo, Yukio; Ikawa, Hiromasa.

1987-03-01

This report describes the results of the steady state thermohydraulic analysis of upgraded JRR-3 core under natural convective cooling mode, using COOLOD-N code. In the code, function to calculate flow-rate under natural convective cooling mode, and a heat transfer package have been newly added to the COOLOD code which has been developed in JAERI. And this report describes outline of the COOLOD-N code. The results of analysis show that the thermohydraulics of upgraded JRR-3 core, under natural convective cooling mode have enough margine to ONB temperature, DNB heat flux and occurance of blisters in fuel meats, which are design criterion of upgraded JRR-3. (author)

Model tests in RAMONA and NEPTUN; Modellversuche in RAMONA und NEPTUN

Energy Technology Data Exchange (ETDEWEB)

Hoffmann, H.; Ehrhard, P.; Weinberg, D.; Carteciano, L.; Dres, K.; Frey, H.H.; Hayafune, H.; Hoelle, C.; Marten, K.; Rust, K.; Thomauske, K.

1995-08-01

In order to demonstrate passive decay heat removal (DHR) in an LMR such as the European Fast Reactor, the RAMONA and NEPTUN facilities, with water as a coolant medium, were used to measure transient flow data corresponding to a transition from forced convection (under normal operation) to natural convection under DHR conditions. The facilities were 1:20 and 1:5 models, respectively, of a pool-type reactor including the IHXs, pumps, and immersed coolers. Important results: The decay heat can be removed from all parts of the primary system by natural convection, even if the primary fluid circulation through the IHX is interrupted. This result could be transferred to liquid metal cooling by experiments in models with thermohydraulic similarity. (orig.)

Thermonuclear model for high energy transients

International Nuclear Information System (INIS)

Woosley, S.E.

1982-01-01

The thermonuclear model for x- and γ-ray bursts is discussed. Different regimes of nuclear burning are reviewed, each appropriate to a given range of (steady state) accretion rate. Accretion rates in the range 10 -14 to 10 -8 Msub solar y -1 all appear capable of producing x-ray transients of various durations and intervals. Modifications introduced by radiatively driven mass loss, the thermal inertia of the envelope, different burning mechanisms, and two-dimensional considerations are discussed as are difficulties encountered when the thermonuclear model is confronted with observations of rapidly recurrent bursts (less than or equal to 10 min), and super-Eddington luminosities and temperatures. Results from a numerical simulation of a combined hydrogen-helium runaway initiated at pycnonuclear density are presented for the first time. The thermonuclear model for γ-ray bursts is also reviewed and updated, particularly with regard to the breakdown of the steady state hypothesis employed in previous work. Solely on the basis of nuclear instability, γ-ray bursts of various types appear possible for a very broad variety of accretion rates (approx. 10 -17 to approx. 10 -11 Msub solar y -1 ) although other considerations may restrict this range. The thermonuclear model appears capable of yielding a great diversity of high energy transient phenomena for various accretion rates, magnetic field configurations, and neutron star envelope histories

Design, construction and evaluation of solar flat-plate collector simulator based on the thermohydraulic coefficient

Directory of Open Access Journals (Sweden)

H Rahmati Aidinlou

2017-05-01

Full Text Available Introduction Increasing the area of absorber plate between the flowed air through the duct can be accomplished by corrugating the absorber plate or by using the artificial roughness underside of the absorber plate as the commercial methods for enhancing the thermohydraulic performance of the flat plate solar air heaters. Evaluation of this requires the construction of separated solar air heater which is costly and time consuming. The constructed solar flat-plate collector simulator can be a sufficient solution for obtaining the heat transfer and thermodynamic parameters for evaluating the absorber plate. The inclined broken roughness was chosen as the optimum roughness which is surrounded by three aluminum smooth walls. Materials and Methods The duct for both smooth and roughened plate have been constructed based on the ASHRAE 93-2010 standard. In order to achieve a fully thermal and hydraulic developed flow, the plenum is constructed. The centrifugal fan is considered by applying the required air volume at the pressure drop obtained by the duct, plenum and the orifice meter. The TSI velocity-meter 8355 is used to measure the velocity of air crossing through the pipe connected to the centrifugal fan. The micro manometer Kimo CPE310-s with the resolution of 0.1 Pa is used to measure the pressure drop across the test section of the smooth and roughened duct. The LM35 sensors are used to measure the absorber plate and air temperature through the test section. Obtained parameters are used to calculate the Nusselt number and friction factor across the test section for smooth and roughened absorber plate. The Nusselt number and friction factor parameters which is obtained for smooth absorber plate based on experimental set-up, is compared with Dittus-Bolter and Blasius equations, respectively, for validating the simulator. By calculating the Nusselt number and friction factor, Stanton number is obtained based on the equation (6, and thermohydraulic

Inductive analysis of failure patterns and of their impact on thermohydraulic circuits of nuclear power plants

International Nuclear Information System (INIS)

Limnios, N.

1983-01-01

The APACHE code (Automatic Analysis of Failures of Hydraulic and Thermohydraulic Circuits more particularly of Water) situates in an important program of computer codes development in the field of studies on reliability and safety of systems in nuclear power plants. APACHE is an automatic generation code of failure pattern and of their effects. After a presentation of the theoretical basis, the methodological principles of the theory of networks are developed. Then, the model of the code is developed: model of individual behavior of each classical model component of normal behavior and model of failure pattern with specifications. The global model of hydraulic systems and the resolution systems are then developed. More particularly, some aspects of the theory of graphs, and the algorithms developed for the automatic construction of the equation systems and especially the algorithm of the research of meshes are presented. The computer aspect of the code and the programming of the code with its limits and some specifications are described. The practical aspect of utilization is finally presented [fr

Validation of the coupled neutron kinetic thermohydraulic code ATHLET/DYN3D with help of measured data of the OECD Turbine Trip Benchmarks. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.

2003-12-01

various calculations for phase II. So, differences of the power and void distribution in single fuel assemblies can be explained by the differences in the neutron kinetic and thermo-hydraulic models. Comparisons between ATHLET/DYN3D (parallel coupling) and ATHLET/QUABOX-CUBBOX (internal coupling) show only small deviations for spatially averaged parameters. The deviations in the local parameters can be explained mainly by the differences in the modeling of the reactor core (reduced number of coolant channels, disregard of the ADF and a different two-phase model in the calculations with ATHLET/QUABOX-CUBBOX). The calculations for the extreme scenarios in phase III demonstrate the applicability of the coupled code ATHLET/DYN3D to transients with conditions beyond the experiment. (orig.) [de

ASCOT-1, Thermohydraulics of Axisymmetric PWR Core with Homogeneous Flow During LOCA

International Nuclear Information System (INIS)

1978-01-01

1 - Nature of the physical problem solved: ASCOT-1 is used to analyze the thermo-hydraulic behaviour in a PWR core during a loss-of-coolant accident. 2 - Method of solution: The core is assumed to be axisymmetric two-dimensional and the conservation laws are solved by the method of characteristics. For the temperature response of fuel in the annular regions into which the core is divided, the heat conduction equations are solved by an explicit method with averaged flow conditions. 3 - Restrictions on the complexity of the problem: Axisymmetric two-dimensional homogeneous flows

Validation of thermohydraulic codes by comparison of experimental results with computer simulations

International Nuclear Information System (INIS)

Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

1989-01-01

The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

Runaway transient simulation of a model Kaplan turbine

Energy Technology Data Exchange (ETDEWEB)

Liu, S; Liu, D; Wu, Y [State Key Laboratory of Hydroscience and Engineering, Department of Thermal Eng., Tsinghua University, Beijing, 100084 (China); Zhou, D [Water Conservancy and Hydropower Eng., Hohai University, Nanjing. 210098 (China); Nishi, M, E-mail: liushuhong@tsinghua.edu.c [Kyushu Inst. Tech. Senior Academy, Kitakyushu, 804-8550 (Japan)

2010-08-15

The runaway transient is a typical transient process of a hydro power unit, where the rotational speed of a turbine runner rapidly increases up to the runaway speed under a working head as the guide vanes cannot be closed due to some reason at the load rejection. In the present paper, the characteristics of the runaway transient of a model Kaplan turbine having ns = 479(m-kW) is simulated by using a time-dependent CFD technique where equation of rotational motion of runner, continuity equation and unsteady RANS equations with RNG k-{epsilon} turbulence model are solved iteratively. In the calculation, unstructured mesh is used to the whole flow passage, which consists of several sub-domains: entrance, casing, stay vanes + guide vanes, guide section, runner and draft tube. And variable speed sliding mesh technique is used to exchange interface flow information between moving part and stationary part, and three-dimensional unstructured dynamic mesh technique is also adopted to ensure mesh quality. Two cases were treated in the simulation of runaway transient characteristics after load rejection: one is the rated operating condition as the initial condition, and the other is the condition at the maximum head. Regarding the runaway speed, the experimental speed is 1.45 times the initial speed and the calculation is 1.47 times the initial for the former case. In the latter case, the experiment and the calculation are 1.67 times and 1.69 times respectively. From these results, it is recognized that satisfactorily prediction will be possible by using the present numerical method. Further, numerical results show that the swirl in the draft-tube flow becomes stronger in the latter part of the transient process so that a vortex rope will occur in the draft tube and its precession will cause the pressure fluctuations which sometimes affect the stability of hydro power system considerably.

Runaway transient simulation of a model Kaplan turbine

Science.gov (United States)

Liu, S.; Zhou, D.; Liu, D.; Wu, Y.; Nishi, M.

2010-08-01

The runaway transient is a typical transient process of a hydro power unit, where the rotational speed of a turbine runner rapidly increases up to the runaway speed under a working head as the guide vanes cannot be closed due to some reason at the load rejection. In the present paper, the characteristics of the runaway transient of a model Kaplan turbine having ns = 479(m-kW) is simulated by using a time-dependent CFD technique where equation of rotational motion of runner, continuity equation and unsteady RANS equations with RNG k-epsilon turbulence model are solved iteratively. In the calculation, unstructured mesh is used to the whole flow passage, which consists of several sub-domains: entrance, casing, stay vanes + guide vanes, guide section, runner and draft tube. And variable speed sliding mesh technique is used to exchange interface flow information between moving part and stationary part, and three-dimensional unstructured dynamic mesh technique is also adopted to ensure mesh quality. Two cases were treated in the simulation of runaway transient characteristics after load rejection: one is the rated operating condition as the initial condition, and the other is the condition at the maximum head. Regarding the runaway speed, the experimental speed is 1.45 times the initial speed and the calculation is 1.47 times the initial for the former case. In the latter case, the experiment and the calculation are 1.67 times and 1.69 times respectively. From these results, it is recognized that satisfactorily prediction will be possible by using the present numerical method. Further, numerical results show that the swirl in the draft-tube flow becomes stronger in the latter part of the transient process so that a vortex rope will occur in the draft tube and its precession will cause the pressure fluctuations which sometimes affect the stability of hydro power system considerably.

Runaway transient simulation of a model Kaplan turbine

International Nuclear Information System (INIS)

Liu, S; Liu, D; Wu, Y; Zhou, D; Nishi, M

2010-01-01

The runaway transient is a typical transient process of a hydro power unit, where the rotational speed of a turbine runner rapidly increases up to the runaway speed under a working head as the guide vanes cannot be closed due to some reason at the load rejection. In the present paper, the characteristics of the runaway transient of a model Kaplan turbine having ns = 479(m-kW) is simulated by using a time-dependent CFD technique where equation of rotational motion of runner, continuity equation and unsteady RANS equations with RNG k-ε turbulence model are solved iteratively. In the calculation, unstructured mesh is used to the whole flow passage, which consists of several sub-domains: entrance, casing, stay vanes + guide vanes, guide section, runner and draft tube. And variable speed sliding mesh technique is used to exchange interface flow information between moving part and stationary part, and three-dimensional unstructured dynamic mesh technique is also adopted to ensure mesh quality. Two cases were treated in the simulation of runaway transient characteristics after load rejection: one is the rated operating condition as the initial condition, and the other is the condition at the maximum head. Regarding the runaway speed, the experimental speed is 1.45 times the initial speed and the calculation is 1.47 times the initial for the former case. In the latter case, the experiment and the calculation are 1.67 times and 1.69 times respectively. From these results, it is recognized that satisfactorily prediction will be possible by using the present numerical method. Further, numerical results show that the swirl in the draft-tube flow becomes stronger in the latter part of the transient process so that a vortex rope will occur in the draft tube and its precession will cause the pressure fluctuations which sometimes affect the stability of hydro power system considerably.

A Semi-implicit Numerical Scheme for a Two-dimensional, Three-field Thermo-Hydraulic Modeling

International Nuclear Information System (INIS)

Hwang, Moonkyu; Jeong, Jaejoon

2007-07-01

The behavior of two-phase flow is modeled, depending on the purpose, by either homogeneous model, drift flux model, or separated flow model, Among these model, in the separated flow model, the behavior of each flow phase is modeled by its own governing equation, together with the interphase models which describe the thermal and mechanical interactions between the phases involved. In this study, a semi-implicit numerical scheme for two-dimensional, transient, two-fluid, three-field is derived. The work is an extension to the previous study for the staggered, semi-implicit numerical scheme in one-dimensional geometry (KAERI/TR-3239/2006). The two-dimensional extension is performed by specifying a relevant governing equation set and applying the related finite differencing method. The procedure for employing the semi-implicit scheme is also described in detail. Verifications are performed for a 2-dimensional vertical plate for a single-phase and two-phase flows. The calculations verify the mass and energy conservations. The symmetric flow behavior, for the verification problem, also confirms the momentum conservation of the numerical scheme

Modelling transient energy release from molten fuel coolant interaction debris

International Nuclear Information System (INIS)

Fletcher, D.F.

1984-05-01

A simple model of transient energy release in a Molten Fuel Coolant Interaction is presented. A distributed heat transfer model is used to examine the effect of heat transfer coefficient, time available for rapid energy heat transfer and particle size on transient energy release. The debris is assumed to have an Upper Limit Lognormal distribution. Model predictions are compared with results from the SUW series of experiments which used thermite-generated uranium dioxide molybdenum melts released below the surface of a pool of water. Uncertainties in the physical principles involved in the calculation of energy transfer rates are discussed. (author)

BWR thermohydraulics simulation on the AD-10 peripheral processor

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Lekach, S.V.; Mallen, A.N.

1983-01-01

This presentation demonstrates the feasibility of simulating plant transients and severe abnormal transients in nuclear power plants at much faster than real-time computing speeds in a low-cost, dedicated, interactive minicomputer. This is achieved by implementing advanced modeling techniques in modern, special-purpose peripheral processors for high-speed system simulation. The results of this demonstration will impact safety analyses and parametric studies, studies on operator responses and control system failures and it will make possible the continuous on-line monitoring of plant performance and the detection and diagnosis of system or component failures

Transient Heat Transfer Model for Car Body Primer Curing

OpenAIRE

D. Zabala; N. Sánchez; J. Pinto

2010-01-01

A transient heat transfer mathematical model for the prediction of temperature distribution in the car body during primer baking has been developed by considering the thermal radiation and convection in the furnace chamber and transient heat conduction governing equations in the car framework. The car cockpit is considered like a structure with six flat plates, four vertical plates representing the car doors and the rear and front panels. The other two flat plates are the...

Computer Models for IRIS Control System Transient Analysis

International Nuclear Information System (INIS)

Gary D Storrick; Bojan Petrovic; Luca Oriani

2007-01-01

This report presents results of the Westinghouse work performed under Task 3 of this Financial Assistance Award and it satisfies a Level 2 Milestone for the project. Task 3 of the collaborative effort between ORNL, Brazil and Westinghouse for the International Nuclear Energy Research Initiative entitled 'Development of Advanced Instrumentation and Control for an Integrated Primary System Reactor' focuses on developing computer models for transient analysis. This report summarizes the work performed under Task 3 on developing control system models. The present state of the IRIS plant design--such as the lack of a detailed secondary system or I and C system designs--makes finalizing models impossible at this time. However, this did not prevent making considerable progress. Westinghouse has several working models in use to further the IRIS design. We expect to continue modifying the models to incorporate the latest design information until the final IRIS unit becomes operational. Section 1.2 outlines the scope of this report. Section 2 describes the approaches we are using for non-safety transient models. It describes the need for non-safety transient analysis and the model characteristics needed to support those analyses. Section 3 presents the RELAP5 model. This is the highest-fidelity model used for benchmark evaluations. However, it is prohibitively slow for routine evaluations and additional lower-fidelity models have been developed. Section 4 discusses the current Matlab/Simulink model. This is a low-fidelity, high-speed model used to quickly evaluate and compare competing control and protection concepts. Section 5 describes the Modelica models developed by POLIMI and Westinghouse. The object-oriented Modelica language provides convenient mechanisms for developing models at several levels of detail. We have used this to develop a high-fidelity model for detailed analyses and a faster-running simplified model to help speed the I and C development process. Section

Experimental Investigation of Thermohydraulic Performance of a Rectangular Solar Air Heater Duct Equipped with V-Shaped Perforated Blocks

Directory of Open Access Journals (Sweden)

Tabish Alam

2014-01-01

Full Text Available This paper presents the thermohydraulic performance of rectangular solar air heater duct equipped with V-shaped rectangular perforated blocks attached to the heated surface. The V-shaped perforated blocks are tested for downstream (V-down to the air flow at Reynolds number from 2000 to 20000. The perforated blocks have relative pitch ratio (P/e from 4 to 12, relative blockage height ratio (e/H from 0.4 to 1.0, and open area ration from 5% to 25% at a fixed value of angle of attack of 60∘ in a rectangular duct having duct aspect ratio (W/H of 12. Thermohydraulic performance is compared at different geometrical parameters of V-shaped perforated blocks for equal pumping power which shows that maximum performance is observed at a relative pitch of 8, relative rib height of 0.8, and open area ration of 20%. It is also observed that the performance of V-shaped perforated blocks was better than transverse-perforated blocks.

Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

International Nuclear Information System (INIS)

Scheuerer, Martina; Weis, Johannes

2012-01-01

Highlights: ► Pressurized thermal shocks are important phenomena for plant life extension and aging. ► The thermal-hydraulics of PTS have been studied experimentally and numerically. ► In the Large Scale Test Facility a loss of coolant accident was investigated. ► CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

Improved lumped parameter for annular fuel element thermohydraulic analysis

International Nuclear Information System (INIS)

Duarte, Juliana Pacheco; Su, Jian; Alvim, Antonio Carlos Marques

2011-01-01

Annular fuel elements have been intensively studied for the purpose of increasing power density in light water reactors (LWR). This paper presents an improved lumped parameter model for the dynamics of a LWR core with annular fuel elements, composed of three sub-models: the fuel dynamics model, the neutronics model, and the coolant energy balance model. The transient heat conduction in radial direction is analyzed through an improved lumped parameter formulation. The Hermite approximation for integration is used to obtain the average temperature of the fuel and cladding and also to obtain the average heat flux. The volumetric heat generation in fuel rods was obtained with the point kinetics equations with six delayed neutron groups. The equations for average temperature of fuel and cladding are solved along with the point kinetic equations, assuming linear reactivity and coolant temperature in cases of reactivity insertion. The analytical development of the model and the numeric solution of the ordinary differential equation system were obtained by using Mathematica 7.0. The dynamic behaviors for average temperatures of fuel, cladding and coolant in transient events as well as the reactor power were analyzed. (author)

Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, S.; Koonen, E. [SCK-CEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium); Olson, A. P. [RERTR Program, Nuclear Engineering Div., Argonne National Laboratory, Cass Avenue, Argonne, IL 60439 (United States)

2012-07-01

Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)

Application of a qualified RETRAN model to plant transient evaluation support

International Nuclear Information System (INIS)

Sedano, P.G.; Mata, P.; Alcantud, F.; Serra, J.; Castrillo, F.

1989-01-01

This paper presents the applicability and usefulness of a complete and well qualified plant transient code and model to support in depth evaluation of anomalous plant transients. Analyses of several operational and abnormal transients that ocurred during the first three cycles of Cofrentes (BWR-6) NPP are presented. This application demonstrated the need of a very detailed and adjusted simulation of the control systems as well as the convenience of having as complete as possible data adquisition system. (orig.)

Application of a qualified RETRAN model to plant transient evaluation support

International Nuclear Information System (INIS)

Sedano, P.G.; Mata, P.; Alcantud, F.; Serra, J.

1989-01-01

This paper presents the applicability and usefulness of a complete and well qualified plant transient code and model to support in depth evaluation of anomalous plant transients. Analyses of several operational and abnormal transients occurred during the first three cycles of Cofrentes (BWR-6) NPP are presented. This application remarked the need of a very detailed and adjusted simulation of the control systems as well as the convenience of having an as complete as possible data acquisition system

Validation of the coupled TRACE /PARCS model of KKL through data of plant from the event shooting turbine 2001; Validacion del modelo acoplado TRACE/PARCS de KKL por medio de los datos de planta del evento disparo de turbina 2001

Energy Technology Data Exchange (ETDEWEB)

Hidalga, P.; Sekhri, A.; Baumann, P.; Miro, R.; Barrachina, T.; Morera, D.; Verdu, G.

2014-07-01

In order to improve the modeling of the nuclear power station Leibstadt (KKL), has been used neutronic code PARCS 3D mesh and the Thermo-hydraulic TRACE. This work is part of the development of a multi-scale methodology and multi-physic allowing the analysis of transient in the reactor, using the available simulation tools. To check the validity of the model it has been simulated a transient complex type turbine trigger the same actual event in plant-based and the results have been compared. At the same time, have been generated the necessary effective sections corresponding to the cycle that is happened the event using the SIMTAB methodology, developed by the ISIRYM jointly with IBERINCO. The model used in TRACE is based on the prior model existing TRAC-BF1. The comparison with plant data shows a good agreement with the data from the simulation. As a result, it can be concluded that KKL TRACE/PARCS coupled model with the generation of effective sections SIMTAB provides a satisfactory analysis of transient turbine trigger complex. (Author)

Thermohydraulic behaviour of the hot channel in a PWR type reactor under loss-of-coolant accident conditions (LOCA)

International Nuclear Information System (INIS)

Costa, J.R.

1978-12-01

An analysis is done of the core behavior for a 1861 MW(th) pressurized water reactor with two coolant loops, during the blowdown phase of a double-ended cold leg rupture, between the main feedwater pump, and the pressure vessel. The analysis is done through a detailed thermohydraulic study of the hot pin channel with RELAP4/MOD 5 code, including the Evaluatin Model options. The problem is solved separately for two values of discharge coefficient (C sub(D)= 1,0 and 0,4). The results show that the maximum clad temperature is lower than the limit value for licensing purposes. Concerning clad material oxidation, the maximum value obtained is also under the limit of acceptance. (author) [pt

Testing the ontogenetic base for the transient model of inflorescence development.

Science.gov (United States)

Bull-Hereñu, Kester; Claßen-Bockhoff, Regine

2013-11-01

Current research in plant science has concentrated on revealing ontogenetic processes of key attributes in plant evolution. One recently discussed model is the 'transient model' successful in explaining some types of inflorescence architectures based on two main principles: the decline of the so called 'vegetativeness' (veg) factor and the transient nature of apical meristems in developing inflorescences. This study examines whether both principles find a concrete ontogenetic correlate in inflorescence development. To test the ontogenetic base of veg decline and the transient character of apical meristems the ontogeny of meristematic size in developing inflorescences was investigated under scanning electron microscopy. Early and late inflorescence meristems were measured and compared during inflorescence development in 13 eudicot species from 11 families. The initial size of the inflorescence meristem in closed inflorescences correlates with the number of nodes in the mature inflorescence. Conjunct compound inflorescences (panicles) show a constant decrease of meristematic size from early to late inflorescence meristems, while disjunct compound inflorescences present an enlargement by merging from early inflorescence meristems to late inflorescence meristems, implying a qualitative change of the apical meristems during ontogeny. Partial confirmation was found for the transient model for inflorescence architecture in the ontogeny: the initial size of the apical meristem in closed inflorescences is consistent with the postulated veg decline mechanism regulating the size of the inflorescence. However, the observed biphasic kinetics of the development of the apical meristem in compound racemes offers the primary explanation for their disjunct morphology, contrary to the putative exclusive transient mechanism in lateral axes as expected by the model.

Model for transient creep of southeastern New Mexico rock salt

International Nuclear Information System (INIS)

Herrmann, W.; Wawersik, W.R.; Lauson, H.S.

1980-11-01

In a previous analysis, existing experimental data pertaining to creep tests on rock salt from the Salado formation of S.E. New Mexico were fitted to an exponential transient creep law. While very early time portions of creep strain histories were not fitted very well for tests at low temperatures and stresses, initial creep rates in particular generally being underestimated, the exponential creep law has the property that the transient creep strain approaches a finite limit with time, and is therefore desirable from a creep modelling point of view. In this report, an analysis of transient creep is made. It is found that exponential transient creep can be related to steady-state creep through a universal creep curve. The resultant description is convenient for creep analyses where very early time behavior is not important

3-dimensional thermohydraulic analysis of KALIMER reactor pool during unprotected accidents

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Bum; Hahn Do Hee

2003-01-01

During a normal reactor scram, the heat generation is reduced almost instantaneously while the coolant flow rate follows the pump coastdown. This mismatch between power and flow results in a situation where the core flow entering the hot pool is at a lower temperature than the temperature of the bulk pool sodium. This temperature difference leads to thermal stratification. Thermal stratification can occur in the hot pool region if the entering coolant is colder than the existing hot pool coolant and the flow momentum is not large enough to overcome the negative buoyancy force. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response. Hence, it is necessary to predict the pool coolant temperature distribution with sufficient accuracy to determine the inlet temperature conditions for the IHXs and its contribution to the net buoyancy head. Therefore, two-dimensional hot pool thermohydraulic model named HP2D has been developed. In this report code-to-code comparison analysis between HP2D and COMMIX-1AR/P has been performed in the case of steady-state and UTOP.

Application of a film flow model to predicting burnout under transient conditions

International Nuclear Information System (INIS)

Leslie, D.C.; Kirby, G.J.

1967-08-01

The film flow model developed previously has been generalised to transient situations by assuming that only convection is changed by the transient; evaporation, deposition and entrainment are assumed to be unaffected. A computer code TRABUT computes the time behaviour of the mass velocity and the quality by the method of characteristics, and then integrates the film flow equations along the same characteristics until the point of burn-out or zero film flow is reached. The time delay between the onset of a transient and burn-out has been computed both for flux and flow transients. These computations have been compared with those made using the standard local conditions hypothesis. The film flow model gives shorter delays in almost all cases, but the difference would not be detectable with present experimental techniques. (author)

Predictive modeling of transient storage and nutrient uptake: Implications for stream restoration

Science.gov (United States)

O'Connor, Ben L.; Hondzo, Miki; Harvey, Judson W.

2010-01-01

This study examined two key aspects of reactive transport modeling for stream restoration purposes: the accuracy of the nutrient spiraling and transient storage models for quantifying reach-scale nutrient uptake, and the ability to quantify transport parameters using measurements and scaling techniques in order to improve upon traditional conservative tracer fitting methods. Nitrate (NO3–) uptake rates inferred using the nutrient spiraling model underestimated the total NO3– mass loss by 82%, which was attributed to the exclusion of dispersion and transient storage. The transient storage model was more accurate with respect to the NO3– mass loss (±20%) and also demonstrated that uptake in the main channel was more significant than in storage zones. Conservative tracer fitting was unable to produce transport parameter estimates for a riffle-pool transition of the study reach, while forward modeling of solute transport using measured/scaled transport parameters matched conservative tracer breakthrough curves for all reaches. Additionally, solute exchange between the main channel and embayment surface storage zones was quantified using first-order theory. These results demonstrate that it is vital to account for transient storage in quantifying nutrient uptake, and the continued development of measurement/scaling techniques is needed for reactive transport modeling of streams with complex hydraulic and geomorphic conditions.

TRACE/PARCS modelling of rips trip transients for Lungmen ABWR

Energy Technology Data Exchange (ETDEWEB)

Chang, C. Y. [Inst. of Nuclear Engineering and Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Lin, H. T.; Wang, J. R. [Inst. of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Shih, C. [Inst. of Nuclear Engineering and Science, Dept. of Engineering and System Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China)

2012-07-01

The objectives of this study are to examine the performances of the steady-state results calculated by the Lungmen TRACE/PARCS model compared to SIMULATE-3 code, as well as to use the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen TRACE/PARCS model. In this study, three power generation methods in TRACE were utilized to analyze the three reactor internal pumps (RIPs) trip transient for the purpose of validating the TRACE/PARCS model. In general, the comparisons show that the transient responses of key system parameters agree well with the FSAR results, including core power, core inlet flow, reactivity, etc. Further studies will be performed in the future using Lungmen TRACE/PARCS model. After the commercial operation of Lungmen nuclear power plant, TRACE/PARCS model will be verified. (authors)

Unified fluid flow model for pressure transient analysis in naturally fractured media

International Nuclear Information System (INIS)

Babak, Petro; Azaiez, Jalel

2015-01-01

Naturally fractured reservoirs present special challenges for flow modeling with regards to their internal geometrical structure. The shape and distribution of matrix porous blocks and the geometry of fractures play key roles in the formulation of transient interporosity flow models. Although these models have been formulated for several typical geometries of the fracture networks, they appeared to be very dissimilar for different shapes of matrix blocks, and their analysis presents many technical challenges. The aim of this paper is to derive and analyze a unified approach to transient interporosity flow models for slightly compressible fluids that can be used for any matrix geometry and fracture network. A unified fractional differential transient interporosity flow model is derived using asymptotic analysis for singularly perturbed problems with small parameters arising from the assumption of a much smaller permeability of the matrix blocks compared to that of the fractures. This methodology allowed us to unify existing transient interporosity flow models formulated for different shapes of matrix blocks including bounded matrix blocks, unbounded matrix cylinders with any orthogonal crossection, and matrix slabs. The model is formulated using a fractional order diffusion equation for fluid pressure that involves Caputo derivative of order 1/2 with respect to time. Analysis of the unified fractional derivative model revealed that the surface area-to-volume ratio is the key parameter in the description of the flow through naturally fractured media. Expressions of this parameter are presented for matrix blocks of the same geometrical shape as well as combinations of different shapes with constant and random sizes. Numerical comparisons between the predictions of the unified model and those obtained from existing transient interporosity ones for matrix blocks in the form of slabs, spheres and cylinders are presented for linear, radial and spherical flow types for

Analysis of the influences of thermal correlations on neutronic–thermohydraulic coupling calculation of SCWR

International Nuclear Information System (INIS)

Xu, Weifeng; Cai, Jiejin; Liu, Shichang; Tang, Qi

2015-01-01

Highlights: • Different thermal correlations for supercritical water are summarized. • Influences of thermal correlations on neutronic–thermohydraulic coupling calculation are analyzed. • Sensitivity analysis has been done for the thermal correlations. - Abstract: The neutronic–thermohydraulic coupling (N–T coupling) calculation is important on core design, security and stability analysis of supercritical water-coolant reactor (SCWR), and a suitable thermal correlation is also necessary for the N–T coupling calculation. In this paper, the scheme of the U.S. SCWR design and the process of the N–T coupling will be introduced as well as some of different thermal correlations firstly. Then, based on the N–T coupling system ARNT, the U.S. SCWR design is simulated to analyze the influences of thermal correlations on N–T coupling calculation of SCWR so as to find out which correlation is best. The result shows that all thermal correlations are suitable. However, using different correlations for calculation leads to a great difference in safety margin of SCWR. What's more, the Bishop and Jackson correlations are more suitable and conservative, but the Griem correlation is not very precise. And the effect of buoyancy lift makes little influence on the calculation of heat transfer of SCWR. This research is also of great significance for the further study of N–T coupling of SCWR

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)

Review of the SCDAP/RELAP5/MOD3.1 code structure and core T/H model before core damage

International Nuclear Information System (INIS)

Kim, See Darl; Kim, Dong Ha

1998-04-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code is being developed at the INEL under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. NRC. As The current time, the SCDAP/RELAP5/MOD3.1 code is the result of merging the RELAP5/MOD3 and SCDAP models. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. Major purpose of the report is to provide information about the characteristics of SCDAP/RELAP5/MOD3.1 core T/H models for an integrated severe accident computer code being developed under the mid/long-term project. This report analyzes the overall code structure which consists of the input processor, transient controller, and plot file handler. The basic governing equations to simulate the thermohydraulics of the primary system are also described. As the focus is currently concentrated in the core, core nodalization parameters of the intact geometry and the phenomenological subroutines for the damaged core are summarized for the future usage. In addition, the numerical approach for the heat conduction model is investigated along with heat convection model. These studies could provide a foundation for input preparation and model improvement. (author). 6 refs., 3 tabs., 4 figs

Modifications in Compacted MX-80 Bentonite Due to Thermo-Hydraulic Treatment

International Nuclear Information System (INIS)

Gomez-Espina, R.; Villar, M. V.

2013-01-01

The thermo-hydraulic tests reproduce the thermal and hydraulic conditions to which bentonite is subjected in the engineered barrier of a deep geological repository of radioactive waste. The results of thermo-hydraulic test TBT1500, which was running for approximately 1500 days, are presented. This is a continuation to the Technical Report Ciemat 1199, which presented results of test TBT500, performed under similar conditions but with duration of 500 days. In both tests the MX-80 bentonite was used with initial density and water content similar to those of the large-scale test TBT. The bentonite column was heated at the bottom at 140 degree centigrade and hydrated on top with deionized water. At the end of the test a sharp water content gradient was observed along the column, as well as an inverse dry density gradient. Hydration modified also the bentonite microstructure. Besides, an overall decrease of the smectite content with respect to the initial value took place, especially in the most hydrated areas where the percentage of interest ratified illite increased and in the longer test. On the other hand, the content of cristobalite, feldspars and calcite increased. Smectite dissolution processes (probably colloidal) occurred, particularly in the more hydrated areas and in the longer test. Due to the dissolution of low-solubility species and to the loss of exchangeable positions in the smectite, the content of soluble salts in the pore water increased with respect to the original one, especially in the longer test. The solubilized ions were transported; sodium, calcium, magnesium and sulphate having a similar mobility, which was in turn lower than that of potassium and chloride. The cationic exchange complex was also modified. (Author)

Development of computer models for fuel element behaviour in water reactors

International Nuclear Information System (INIS)

Gittus, J.H.

1987-03-01

Description of fuel behaviour during normal operation transients and accident conditions has always represented a most challenging and important problem. Reliable predictions constitute a basic demand for safety based calculations, for design purposes and for fuel performance. Therefore, computer codes based on deterministic and probabilistic models were developed. Possibility of comprehensive descriptions of the phenomena is precluded in view of the great number of individual processes, involving physical, chemical, thermohydraulical and mechanical parameters, to be considered in a wide range of situations. In case of fast thermal transients predictive capability is limited by the kinetics of evolution of the system and its eventual dynamic behaviour. Evidently, probabilistic approaches are also limited by the sparcity and limited breadth of the impirical data base. Code predictions have to be evaluated against power reactor data, results from simulation experiments and, if possible, include cross validation of different codes and validation of sub-models. Progress on this subject is reviewed in this report, which completes the co-ordinated research programme on 'Development of Computer Models for Fuel Element Behaviour in Water Reactors' (D-COM), initiated under the auspices of the IAEA in 1981

Simplified drive system models for power system transient studies in industrial plants

DEFF Research Database (Denmark)

Chen, Peiyuan; Sannino, Ambra

2007-01-01

In order to simulate industrial plants for different power system transient studies, simplified adjustable speed drive (ASD) models are needed. For power system transient studies such as assessing the voltage dip ride-through capability of ASDs, detailed representation of semiconductor valve...... switching can be avoided, thereby making possible to increase the time step of the simulation. In this paper, simplified ASD models are developed and compared with corresponding detailed models. The performance of the simplified models is assessed when increasing the simulation step as much as possible...

Modelling structural systems for transient response analysis

International Nuclear Information System (INIS)

Melosh, R.J.

1975-01-01

This paper introduces and reports success of a direct means of determining the time periods in which a structural system behaves as a linear system. Numerical results are based on post fracture transient analyses of simplified nuclear piping systems. Knowledge of the linear response ranges will lead to improved analysis-test correlation and more efficient analyses. It permits direct use of data from physical tests in analysis and simplication of the analytical model and interpretation of its behavior. The paper presents a procedure for deducing linearity based on transient responses. Given the forcing functions and responses of discrete points of the system at various times, the process produces evidence of linearity and quantifies an adequate set of equations of motion. Results of use of the process with linear and nonlinear analyses of piping systems with damping illustrate its success. Results cover the application to data from mathematical system responses. The process is successfull with mathematical models. In loading ranges in which all modes are excited, eight digit accuracy of predictions are obtained from the equations of motion deduced. Small changes (less than 0.01%) in the norm of the transfer matrices are produced by manipulation errors for linear systems yielding evidence that nonlinearity is easily distinguished. Significant changes (greater than five %) are coincident with relatively large norms of the equilibrium correction vector in nonlinear analyses. The paper shows that deducing linearity and, when admissible, quantifying linear equations of motion from transient response data for piping systems can be achieved with accuracy comparable to that of response data

Unification of three linear models for the transient visual system

NARCIS (Netherlands)

Brinker, den A.C.

1989-01-01

Three different linear filters are considered as a model describing the experimentally determined triphasic impulse responses of discs. These impulse responses arc associated with the transient visual system. Each model reveals a different feature of the system. Unification of the models is

Predictive Modeling of Transient Storage and Nutrient Uptake: Implications for Stream Restoration

Energy Technology Data Exchange (ETDEWEB)

O’Connor, Ben L.; Hondzo, Miki; Harvey, Judson W.

2010-12-01

This study examined two key aspects of reactive transport modeling for stream restoration purposes: the accuracy of the nutrient spiraling and transient storage models for quantifying reach-scale nutrient uptake, and the ability to quantify transport parameters using measurements and scaling techniques in order to improve upon traditional conservative tracer fitting methods. Nitrate (NO-3)(NO3-) uptake rates inferred using the nutrient spiraling model underestimated the total NO-3NO3- mass loss by 82%, which was attributed to the exclusion of dispersion and transient storage. The transient storage model was more accurate with respect to the NO-3NO3- mass loss (±20%) and also demonstrated that uptake in the main channel was more significant than in storage zones. Conservative tracer fitting was unable to produce transport parameter estimates for a riffle-pool transition of the study reach, while forward modeling of solute transport using measured/scaled transport parameters matched conservative tracer breakthrough curves for all reaches. Additionally, solute exchange between the main channel and embayment surface storage zones was quantified using first-order theory. These results demonstrate that it is vital to account for transient storage in quantifying nutrient uptake, and the continued development of measurement/scaling techniques is needed for reactive transport modeling of streams with complex hydraulic and geomorphic conditions.

Physical modelling of a rapid boron dilution transient

Energy Technology Data Exchange (ETDEWEB)

Anderson, N.G.; Hemstroem, B.; Karlsson, R. [Vattenfall Utveckling AB, Aelvkarleby (Sweden); Jacobson, S. [Vattenfall AB, Ringhals, Vaeroebacka (Sweden)

1995-09-01

The analysis of boron dilution accidents in pressurised water reactors has traditionally assumed that mixing is instantaneous and complete everywhere, eliminating in this way the possibility of concentration inhomogeneities. Situations can nevertheless arise where a volume of coolant with a low boron concentration may eventually enter the core and generate a severe reactivity transient. The work presented in this paper deals with a category of Rapid Boron Dilution Events characterised by a rapid start of a Reactor Coolant Pump (RCP) with a plug of relatively unborated water present in the RCS pipe. Model tests have been made at Vattenfall Utveckling AB in a simplified 1:5 scale model of a Westinghouse PWR. Conductivity measurements are used to determine dimensionless boron concentration. The main purpose of this experimental work is to define an experimental benchmark against which a mathematical model can be tested. The final goal is to be able to numerically predict Boron Dilution Transients. This work has been performed as a part of a Co-operative Agreement with Electricite` de France (EDF).

Thermo-hydraulic Quench Propagation at the LHC Superconducting Magnet String

CERN Document Server

Rodríguez-Mateos, F; Serio, L

1998-01-01

The superconducting magnets of the LHC are protected by heaters and cold by-pass diodes. If a magnet quenches, the heaters on this magnet are fired and the magnet chain is de-excited in about two minu tes by opening dump switches in parallel to a resistor. During the time required for the discharge, adjacent magnets might quench due to thermo-hydraulic propagation in the helium bath and/or heat con duction via the bus bar. The number of quenching magnets depends on the mechanisms for the propagation. In this paper we report on quench propagation experiments from a dipole magnet to an adjacent ma gnet. The mechanism for the propagation is hot helium gas expelled from the first quenching magnet. The propagation changes with the pressure opening settings of the quench relief valves.

Experiment on thermohydraulics of simulated control rod

International Nuclear Information System (INIS)

Ogawa, Masuro; Ouchi, Mitsuo; Akino, Norio; Fujimura, Kaoru; Shiina, Yasuaki; Kawamura, Hiroshi

1984-10-01

A thermohydraulic study of a control rod channel is required for the core design of the Very High Temperature Gas Cooled Reactor (VHTR). A non-heating experiment with air flow was performed prior to heating experiment with helium flow. Experimental results on stability of flow, flow rate distribution and pressure drop of the control rod channel are reported. In a test section of the experimental apparatus, five simulated control subrods were suspended vertically in a circular duct. Their dimension was in coincide with those of the Detailed Disign (I) of the VHTR. Air of atomospheric pressure was used as a coolant gas, which flowed in inner and outer paths of the subrods. Total flow rate ranged from 0.0011 to 0.0062 kg/s. Flow rate distribution and pressure drop were obtained for various flow rates. Velocity fluctuation in the channel was also observed using a hot wire anemometer. From these experiments, it was found that the flow rate distribution was nearly the same as a disigned value and that turbulent and laminar flows were simultaneously realized in outer and inner paths respectively. These observations supported a feasibility of the present design. (author)

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.

Development of a Transient Model of a Stirling-Based CHP System

Directory of Open Access Journals (Sweden)

Antón Cacabelos

2013-06-01

Full Text Available Although the Stirling engine was invented in 1816, this heat engine still continues to be investigated due to the variety of energy sources that can be used to power it (e.g., solar energy, fossil fuels, biomass, and geothermal energy. To study the performance of these machines, it is necessary to develop and simulate models under different operating conditions. In this paper, we present a one-dimensional dynamic model based on components from Trnsys: principally, a lumped mass and a heat exchanger. The resulting model is calibrated using GenOpt. Furthermore, the obtained model can be used to simulate the machine both under steady-state operation and during a transient response. The results provided by the simulations are compared with data measured in a Stirling engine that has been subjected to different operating conditions. This comparison shows good agreement, indicating that the model is an appropriate method for transient thermal simulations. This new proposed model requires few configuration parameters and is therefore easily adaptable to a wide range of commercial models of Stirling engines. A detailed analysis of the system results reveals that the power is directly related to the difference of temperatures between the hot and cold sources during the transient and steady-state processes.

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

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.

Evaluation of core modeling effect on transients for multi-flow zone design of SFR

International Nuclear Information System (INIS)

Shin, Andong; Choi, Yong Won

2016-01-01

SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

Evaluation of core modeling effect on transients for multi-flow zone design of SFR

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Choi, Yong Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

Predicting dermal absorption of gas-phase chemicals: transient model development, evaluation, and application

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...

Analysis of transients aimed at assessing the feasibility of eliminating the HO-2 accident protection and the ''moderate leak'' SOB signal

International Nuclear Information System (INIS)

Sommer, J.

1993-12-01

Accidents and transient processes were analyzed in order to assess the feasibility of eliminating the 2nd level accident protection (HO-2). All analyses were performed in 3 alternatives, viz. for the normal performance of HO-2, for the HO-2 signals being transferred to the 1st level accident protection (HO-1), and for a complete elimination of HO-2. Transfer of HO-2 signals to HO-1 definitely brings about an improvement of the nuclear power plant operation safety. There is no evidence indicating that the safety would decrease intolerably if HO-2 were eliminated altogether. Elimination of the ''moderate leak'' safety system does not require any thermohydraulic analysis to be performed. 18 refs

Theoretical Models of Optical Transients. I. A Broad Exploration of the Duration-Luminosity Phase Space

Science.gov (United States)

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.

Models development for natural circulation and its transition process in nuclear power plant

International Nuclear Information System (INIS)

Yu Lei; Cai Qi; Cai Zhangsheng; Xie Haiyan

2008-01-01

On the basis of nuclear power plant (NPP) best-estimate transient analysis code RELAP5/MOD3, the point reactor kinetics model in RELAP5/MOD3 was replaced by the two-group, 3-D space and time dependent neutron kinetic model, in order to exactly analyze the responses of key parameters in natural circulation and its transition process considering the reactivity feedback. The coupled model for three-dimensional physics and thermohydraulics was established and corresponding computing code was developed. Using developed code, natural circulation of NPP and its transiton process were calculated and analyzed. Compared with the experiment data, the calculated results show that its high precise avoids the shortage that the point reactor equation can not reflect the reactivity exactly. This code can be a computing and analysis tool for forced circulation and natural circulation and their transitions. (authors)

Impact of Model Detail of Synchronous Machines on Real-time Transient Stability Assessment

DEFF Research Database (Denmark)

Weckesser, Johannes Tilman Gabriel; Jóhannsson, Hjörtur; Østergaard, Jacob

2013-01-01

In this paper, it is investigated how detailed the model of a synchronous machine needs to be in order to assess transient stability using a Single Machine Equivalent (SIME). The results will show how the stability mechanism and the stability assessment are affected by the model detail. In order...... of the machine models is varied. Analyses of the results suggest that a 4th-order model may be sufficient to represent synchronous machines in transient stability studies....

Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

Energy Technology Data Exchange (ETDEWEB)

Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

2014-07-01

In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

Development of a Two-dimensional Thermohydraulic Hot Pool Model and ITS Effects on Reactivity Feedback during a UTOP in Liquid Metal Reactors

International Nuclear Information System (INIS)

Lee, Yong Bum; Jeong, Hae Yong; Cho, Chung Ho; Kwon, Young Min; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong; Hahn, Do Hee

2009-01-01

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect

Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

Energy Technology Data Exchange (ETDEWEB)

Scheuerer, Martina, E-mail: Martina.Scheuerer@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany); Weis, Johannes, E-mail: Johannes.Weis@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Pressurized thermal shocks are important phenomena for plant life extension and aging. Black-Right-Pointing-Pointer The thermal-hydraulics of PTS have been studied experimentally and numerically. Black-Right-Pointing-Pointer In the Large Scale Test Facility a loss of coolant accident was investigated. Black-Right-Pointing-Pointer CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

International Nuclear Information System (INIS)

Massoud, M.

1987-01-01

Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients

Challenges in mechanical modeling of SFR fuel rod transient behavior

Energy Technology Data Exchange (ETDEWEB)

Feria, F.; Herranz, L. E.

2013-07-01

Modeling of SFR fuel rod mechanical behavior under transient conditions entails the development of a creep law to predict cladding viscoplastic strain. In this regard, this work is focused on defining a proper clad creep law structure as the basis to set a suitable model under SFR off-normal conditions as transient overpower and loss of fluid. To do so, a review of in-codes clad creep models has been done by using SAS-SFR, SCANAIR and ASTEC. The proposed creep model has been structured in two parts: viscoplastic behaviour before the failure (primary and secondary creep) and the failure due to viscoplastic collapse (tertiary creep). In order to model the first part, Norton creep law has been proposed as a conservative option. An irradiation hardening factor should be included for best estimate calculations. The recommendation for the second part is to apply a failure criterion based on strain limit or rupture time, which allows achieving conservative results.

Modeling the effect of transient populations on epidemics in Washington DC.

Science.gov (United States)

Parikh, Nidhi; Youssef, Mina; Swarup, Samarth; Eubank, Stephen

2013-11-06

Large numbers of transients visit big cities, where they come into contact with many people at crowded areas. However, epidemiological studies have not paid much attention to the role of this subpopulation in disease spread. We evaluate the effect of transients on epidemics by extending a synthetic population model for the Washington DC metro area to include leisure and business travelers. A synthetic population is obtained by combining multiple data sources to build a detailed minute-by-minute simulation of population interaction resulting in a contact network. We simulate an influenza-like illness over the contact network to evaluate the effects of transients on the number of infected residents. We find that there are significantly more infections when transients are considered. Since much population mixing happens at major tourism locations, we evaluate two targeted interventions: closing museums and promoting healthy behavior (such as the use of hand sanitizers, covering coughs, etc.) at museums. Surprisingly, closing museums has no beneficial effect. However, promoting healthy behavior at the museums can both reduce and delay the epidemic peak. We analytically derive the reproductive number and perform stability analysis using an ODE-based model.

Modeling the effect of transient populations on epidemics in Washington DC

Science.gov (United States)

Parikh, Nidhi; Youssef, Mina; Swarup, Samarth; Eubank, Stephen

2013-11-01

Large numbers of transients visit big cities, where they come into contact with many people at crowded areas. However, epidemiological studies have not paid much attention to the role of this subpopulation in disease spread. We evaluate the effect of transients on epidemics by extending a synthetic population model for the Washington DC metro area to include leisure and business travelers. A synthetic population is obtained by combining multiple data sources to build a detailed minute-by-minute simulation of population interaction resulting in a contact network. We simulate an influenza-like illness over the contact network to evaluate the effects of transients on the number of infected residents. We find that there are significantly more infections when transients are considered. Since much population mixing happens at major tourism locations, we evaluate two targeted interventions: closing museums and promoting healthy behavior (such as the use of hand sanitizers, covering coughs, etc.) at museums. Surprisingly, closing museums has no beneficial effect. However, promoting healthy behavior at the museums can both reduce and delay the epidemic peak. We analytically derive the reproductive number and perform stability analysis using an ODE-based model.

Exhaust gas recirculation – Zero dimensional modelling and characterization for transient diesel combustion control

International Nuclear Information System (INIS)

Asad, Usman; Tjong, Jimi; Zheng, Ming

2014-01-01

Highlights: • Zero-dimensional EGR model for transient diesel combustion control. • Detailed analysis of EGR effects on intake, cylinder charge and exhaust properties. • Intake oxygen validated as an operating condition-independent measure of EGR. • Quantified EGR effectiveness in terms of NOx emission reduction. • Twin lambda sensor technique for estimation of EGR/in-cylinder parameters. - Abstract: The application of exhaust gas recirculation (EGR) during transient engine operation is a challenging task since small fluctuations in EGR may cause larger than acceptable spikes in NOx/soot emissions or deterioration in the combustion efficiency. Moreover, the intake charge dilution at any EGR ratio is a function of engine load and intake pressure, and typically changes during transient events. Therefore, the management of EGR during transient engine operation or advanced combustion cycles (that are inherently less stable) requires a fundamental understanding of the transient EGR behaviour and its impact on the intake charge development. In this work, a zero-dimensional EGR model is described to estimate the transient (cycle-by-cycle) progression of EGR and the time (engine cycles) required for its stabilization. The model response is tuned to a multi-cylinder engine by using an overall engine system time-constant and shown to effectively track the transient EGR changes. The impact of EGR on the actual air–fuel ratio of the cylinder charge is quantified by defining an in-cylinder excess-air ratio that accounts for the oxygen in the recycled exhaust gas. Furthermore, a twin lambda sensor (TLS) technique is implemented for tracking the intake dilution and in-cylinder excess-air ratio in real-time. The modelling and analysis results are validated against a wide range of engine operations, including transient and steady-state low temperature combustion tests

Simplified distributed parameters BWR dynamic model for transient and stability analysis

International Nuclear Information System (INIS)

Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro

2006-01-01

This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR

State, space relay modeling and simulation using the electromagnetic Transients Program and its transient analysis of control systems capability

International Nuclear Information System (INIS)

Domijan, A.D. Jr.; Emami, M.V.

1990-01-01

This paper reports on a simulation of a MHO distance relay developed to study the effect of its operation under various system conditions. Simulation is accomplished using a state space approach and a modeling technique using ElectroMagnetic Transient Program (Transient Analysis of Control Systems). Furthermore, simulation results are compared with those obtained in another independent study as a control, to validate the results. A data code for the practical utilization of this simulation is given

Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

International Nuclear Information System (INIS)

Mendler, O.J.; Takeuchi, K.; Young, M.Y.

1986-10-01

The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results

Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

Energy Technology Data Exchange (ETDEWEB)

Mendler, O J; Takeuchi, K; Young, M Y

1986-10-01

The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

Energy Technology Data Exchange (ETDEWEB)

Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

2006-07-01

In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

Analysis of natural circulation stability in a low pressure thermohydraulic test loop

International Nuclear Information System (INIS)

Jafari, J.; D'Auria, F.; Kazeminejad, H.; Davilu, H.

2002-01-01

This paper discusses an instability study of a natural circulation (NC) loop performed with the aid of Relap5 thermal-hydraulic system code. This loop has been designed and constructed for the analysis of relevant thermohydraulic parameters of a nuclear reactor. In this study, the main parameters for the stability of NC are identified and characterized through the execution of proper code runs. The obtained stability boundary (SB) in the dimensionless Zuber- Sub-cooling plane is compared with the SB reported in referenced literature. The agreement of predicted NC stability boundaries with the results of independent studies demonstrates both the capability of the mentioned code in assessing NC loop stability and the quality of the performed calculations.(author)

Thermohydraulic status and component behavior in the PWR during the selected meltdown scenario station blackout (SBO); Thermohydraulisches Verhalten und Komponentenverhalten eines DWR bei ausgewaehltem Kernschmelzszenarium infolge Station Blackout (SBO). Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Band, Sebastian; Blaesius, Christoph; Scheuerer, Martina; Steinroetter, Thomas

2017-09-15

The report on the thermohydraulic status and component behavior in the PWR during the selected meltdown scenario station blackout (SBO) includes the following issues: status of science and technology on this topic, analysis of a high-pressure meltdown scenario using ATHLET-CD for a German PWR starting from the initiating event station blackout, three-dimensional computational fluid dynamic (CFD) analyses of the pressurizer coolant loop in a generic German PWR, evaluation of the thermohydraulic steam generator behavior and its effect on the involved primary circuit components.

An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Massoud, M

1987-01-01

Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

Modeling and Mitigation for High Frequency Switching Transients Due to Energization in Offshore Wind Farms

Directory of Open Access Journals (Sweden)

Yanli Xin

2016-12-01

Full Text Available This paper presents a comprehensive investigation on high frequency (HF switching transients due to energization of vacuum circuit breakers (VCBs in offshore wind farms (OWFs. This research not only concerns the modeling of main components in collector grids of an OWF for transient analysis (including VCBs, wind turbine transformers (WTTs, submarine cables, but also compares the effectiveness between several mainstream switching overvoltage (SOV protection methods and a new mitigation method called smart choke. In order to accurately reproduce such HF switching transients considering the current chopping, dielectric strength (DS recovery capability and HF quenching capability of VCBs, three models are developed, i.e., a user–defined VCB model, a HF transformer terminal model and a three-core (TC frequency dependent model of submarine cables, which are validated through simulations and compared with measurements. Based on the above models and a real OWF configuration, a simulation model is built and several typical switching transient cases are investigated to analyze the switching transient process and phenomena. Subsequently, according to the characteristics of overvoltages, appropriate parameters of SOV mitigation methods are determined to improve their effectiveness. Simulation results indicate that the user–defined VCB model can satisfactorily simulate prestrikes and the proposed component models display HF characteristics, which are consistent with onsite measurement behaviors. Moreover, the employed protection methods can suppress induced SOVs, which have a steep front, a high oscillation frequency and a high amplitude, among which the smart choke presents a preferable HF damping effect.

The significance of thermohydraulic conditions for the corrosion safety of PWR steam generators

International Nuclear Information System (INIS)

Gulich, J.F.

1975-04-01

In several PWR nuclear power plants leakages have occurred in the steam generator which were caused by localised corrosion attack. While the attention of manufacturers and operators is focused on the influences of feedwater chemistry and tube material, the present work highlights the fact that the damage always occurred in those places where flow regimed are poorly defined. The investigation leads to the result that local dry out of the heating surface can be contributing cause of damage. A method is indicated for estimating the thermohydraulic conditions in the inflow region over the tube plate and measures to improve corrosion safety are discussed. (author)

Modeling of the transient mobility in disordered organic semiconductors

NARCIS (Netherlands)

Germs, W.C.; Van der Holst, J.M.M.; Van Mensfoort, S.L.M.; Bobbert, P.A.; Coehoorn, R.

2011-01-01

In non-steady-state experiments, the electrical response of devicesbased on disordered organic semiconductors often shows a large transient contribution due to relaxation of the out-of-equilibrium charge-carrier distribution. We have developed a model describing this process, based only on the

Modification of the bubble rise model used in RELAP4/Mod5 computer code for transients analysis

International Nuclear Information System (INIS)

Scharfmann, E.

1981-01-01

To improve the separation phase and heat transfer models in RELAP4/MOD5 computer code, in order to make more realistic estimates of the thermohydraulic behavior of the core submitted to a loss-of-coolant accident, is the objective of this work. This research is directed to the accident analysis caused by small breaks in the primary circuit of PWR plants, where two-phase flow occurs most of the time. Calculation have been performed with the help of the original version of RELAP code, and the version containing the proposed modifications on this work. Comparing one results with the original ones, we arrive at the conclusion that our results show more conservative values of core pressure and coolant temperature, while the peak values of fuel temperature are not exceeded. (Author) [pt

A simplified model for tritium permeation transient predictions when trapping is active

International Nuclear Information System (INIS)

Longhurst, G.R.

1994-01-01

This report describes a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. Comparison calculations with the verified and validated TMAP4 transient code show good agreement. ((orig.))

A simplified model for tritium permeation transient predictions when trapping is active

Energy Technology Data Exchange (ETDEWEB)

Longhurst, G.R. (Fusion Safety Program, Idaho National Engineering Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States))

1994-09-01

This report describes a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. Comparison calculations with the verified and validated TMAP4 transient code show good agreement. ((orig.))

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

2016-08-15

Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

International Nuclear Information System (INIS)

Shi, Chengbin; Cheng, Maosong; Liu, Guimin

2016-01-01

Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

MODELLING OF NON-ROAD TRANSIENT CYCLE

Directory of Open Access Journals (Sweden)

Martin Kotus

2013-12-01

Full Text Available The paper describes the modeling of NRTC (Non-Road Transient Cycle test procedure based on previously measured characteristics of fuel consumption, carbon monoxide (CO, carbon dioxide (CO2, hydrocarbons (HC, nitrogen oxides (NOx and particulates (PM production. It makes possible to compare the current technical condition of an internal combustion engine of an agricultural tractor with its previous state or other tractor’s engine. Based on measured characteristics, it is also possible to model any other cycle without further measurements (NRSC test procedure, cycle for specific conditions – mountain tractor, etc.. The result may thus contribute to improving the environment by reducing the production of harmful substances emitted into the air and save money due to reduced fuel consumption.

General and preliminary thermohydraulic, hydrogen and aerosol instrumentation plan for the Phebus Fp-project

International Nuclear Information System (INIS)

Hampel, G.; Poss, G.; Frohlich, H.K.

1989-10-01

The objective of the project was to draw up an instrumentation plan for the French core melting programme PHEBUS FP. This instrumentation plan essentially was to include proven and reliable instruments for recording various thermohydraulic, aerosol and hydrogen phenomena. The candidate measuring methods, which are known mainly from reactor safety programmes, have been described and examined for their usefulness in PHEBUS. Each method and instrument has been described in detail under various aspects such as measuring principle, measuring range, technical design, evaluation model, calibration procedure, accuracy, previous experience, commercial availability, etc. Special attention has been paid to the behaviour of the measuring transducers when exposed to radiation. First, the performance of the instruments was compared with the requirements of PHEBUS. The results of this comparison served as the basis for a measuring concept in tabular form, giving the locations of the measurements, the measuring tasks, and the number and kind of instruments that are recommended. Redundancy and cost-benefit aspects have been taken into account in qualitative terms

State space model extraction of thermohydraulic systems – Part II: A linear graph approach applied to a Brayton cycle-based power conversion unit

International Nuclear Information System (INIS)

Uren, Kenneth Richard; Schoor, George van

2013-01-01

This second paper in a two part series presents the application of a developed state space model extraction methodology applied to a Brayton cycle-based PCU (power conversion unit) of a PBMR (pebble bed modular reactor). The goal is to investigate if the state space extraction methodology can cope with larger and more complex thermohydraulic systems. In Part I the state space model extraction methodology for the purpose of control was described in detail and a state space representation was extracted for a U-tube system to illustrate the concept. In this paper a 25th order nonlinear state space representation in terms of the different energy domains is extracted. This state space representation is solved and the responses of a number of important states are compared with results obtained from a PBMR PCU Flownex ® model. Flownex ® is a validated thermo fluid simulation software package. The results show that the state space model closely resembles the dynamics of the PBMR PCU. This kind of model may be used for nonlinear MIMO (multi-input, multi-output) type of control strategies. However, there is still a need for linear state space models since many control system design and analysis techniques require a linear state space model. This issue is also addressed in this paper by showing how a linear state space model can be derived from the extracted nonlinear state space model. The linearised state space model is also validated by comparing the state space model to an existing linear Simulink ® model of the PBMR PCU system. - Highlights: • State space model extraction of a pebble bed modular reactor PCU (power conversion unit). • A 25th order nonlinear time varying state space model is obtained. • Linearisation of a nonlinear state space model for use in power output control. • Non-minimum phase characteristic that is challenging in terms of control. • Models derived are useful for MIMO control strategies

Improved lumped models for transient combined convective and radiative cooling of multi-layer composite slabs

International Nuclear Information System (INIS)

An Chen; Su Jian

2011-01-01

Improved lumped parameter models were developed for the transient heat conduction in multi-layer composite slabs subjected to combined convective and radiative cooling. The improved lumped models were obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of three-layer composite slabs was analyzed to illustrate the applicability of the proposed lumped models, with respect to different values of the Biot numbers, the radiation-conduction parameter, the dimensionless thermal contact resistances, the dimensionless thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the higher order lumped model (H 1,1 /H 0,0 approximation) yielded significant improvement of average temperature prediction over the classical lumped model. In addition, the higher order (H 1,1 /H 0,0 ) model was applied to analyze the transient heat conduction problem of steel-concrete-steel sandwich plates. - Highlights: → Improved lumped models for convective-radiative cooling of multi-layer slabs were developed. → Two-point Hermite approximations for integrals were employed. → Significant improvement over classical lumped model was achieved. → The model can be applied to high Biot number and high radiation-conduction parameter. → Transient heat conduction in steel-concrete-steel sandwich pipes was analyzed as an example.

Heat-pipe transient model for space applications

International Nuclear Information System (INIS)

Tournier, J.; El-Genk, M.S.; Juhasz, A.J.

1991-01-01

A two-dimensional model is developed for simulating heat pipes transient performance following changes in the input/rejection power or in the evaporator/condenser temperatures. The model employs the complete form of governing equations and momentum and energy jump conditions at the liquid-vapor interface. Although the model is capable of handling both cylindrical and rectangular geometries, the results reported are for a circular heat pipe with liquid lithium as the working fluid. The model incorporates a variety of other working fluids, such as water, ammonia, potassium, sodium, and mercury, and offers combinations of isothermal, isoflux, convective and radiative heating/cooling conditions in the evaporator and condenser regions of the heat pipe. Results presented are for lithium heat pipes with exponential heating of the evaporator and isothermal cooling of the condenser

Compositional Abstraction of PEPA Models for Transient Analysis

DEFF Research Database (Denmark)

Smith, Michael James Andrew

2010-01-01

- or interval - Markov chains allow us to aggregate states in such a way as to safely bound transient probabilities of the original Markov chain. Whilst we can apply this technique directly to a PEPA model, it requires us to obtain the CTMC of the model, whose state space may be too large to construct......Stochastic process algebras such as PEPA allow complex stochastic models to be described in a compositional way, but this leads to state space explosion problems. To combat this, there has been a great deal of work in developing techniques for abstracting Markov chains. In particular, abstract...

Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

International Nuclear Information System (INIS)

Silva, Alexandro S.; Dominguez, Dany S.; Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G.; Lira, Carlos Alberto Brayner de Oliveira

2015-01-01

High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

Steady-state and transient fission gas release and swelling model for LIFE-4

International Nuclear Information System (INIS)

Villalobos, A.; Liu, Y.Y.; Rest, J.

1984-06-01

The fuel-pin modeling code LIFE-4 and the mechanistic fission gas behavior model FASTGRASS have been coupled and verified against gas release data from mixed-oxide fuels which were transient tested in the TREAT reactor. Design of the interface between LIFE-4 and FASTGRASS is based on an earlier coupling between an LWR version of LIFE and the GRASS-SST code. Fission gas behavior can significantly affect steady-state and transient fuel performance. FASTGRASS treats fission gas release and swelling in an internally consistent manner and simultaneously includes all major mechanisms thought to influence fission gas behavior. The FASTGRASS steady-state and transient analysis has evolved through comparisons of code predictions with fission-gas release and swelling data from both in- and ex-reactor experiments. FASTGRASS was chosen over other fission-gas behavior models because of its availability, its compatibility with the LIFE-4 calculational framework, and its predictive capability

MELCOR based severe accident simulation for WWER-440 type nuclear power plants

International Nuclear Information System (INIS)

Vegh, E.; Buerger, L.; Gacs, A.; Gyenes, F.G.; Hozer, Z.; Makovi, P.

1997-01-01

SUBA is a MELCOR based severe accident simulator, installed this summer at the Hungarian Nuclear Safety Directorate. In this simulator the thermohydraulics, chemical reactions and material transport in the primary and secondary systems are calculated by the MELCOR code, but the containment, except the cavity, is modelled by the HERMET code, developed in our Institute. The instrumentation and control, the safety systems and the plant logic, are calculated by our models. This paper describes the main features of the used models and presents three different test transients. The presented transients are as follows: a small break LOCA, a cold leg large break LOCA, and the station blackout, without Diesel generators. In each treated transients the most important parameters are presented as time functions and the most significant events are analysed. (author)

Transient non-isothermal model of a polymer electrolyte fuel cell

Energy Technology Data Exchange (ETDEWEB)

Shah, A.A. [Queen' s-RMC Fuel Cell Research Centre, 945 Princess Street, Kingston, Ont. K7L 5L9 (Canada); Kim, G.-S.; Harvey, D. [Ballard Power Systems, 4343 North Fraser Way, Burnaby, BC V5J 5J9 (Canada); Sui, P.C. [Institute for Integrated Energy Systems, University of Victoria, Victoria, BC V8W 3P6 (Canada)

2007-01-01

In this paper we present a one-dimensional transient model for the membrane electrode assembly of a polymer-electrolyte fuel cell. In earlier work we established a framework to describe the water balance in a steady-state, non-isothermal cathode model that explicitly included an agglomerate catalyst layer component. This paper extends that work in several directions, explicitly incorporating components of the anode, including a micro-porous layer, and accounting for electronic potential variations, gas convection and time dependance. The inclusion of temperature effects, which are vital to the correct description of condensation and evaporation, is new to transient modelling. Several examples of the modelling results are given in the form of potentiostatic sweeps and compared to experimental results. Excellent qualitative agreement is demonstrated, particularly in regard to the phenomenon of hysteresis, a manifestation of the sensitive response of the system to the presence of water. Results pertaining to pore size, contact angle and the presence of a micro-porous layer are presented and future work is discussed. (author)

Transient Model of Hybrid Concentrated Photovoltaic with Thermoelectric Generator

DEFF Research Database (Denmark)

Mahmoudi Nezhad, Sajjad; Qing, Shaowei; Rezaniakolaei, Alireza

2017-01-01

Transient performance of a concentrated photovoltaic thermoelectric (CPV-TEG) hybrid system is modeled and investigated. A heat sink with water, as the working fluid has been implemented as the cold reservoir of the hybrid system to harvest the heat loss from CPV cell and to increase the efficiency...

URGAP: A gap conductance model for transient conditions

International Nuclear Information System (INIS)

Lassmann, K.; Pazdera, F.

1983-01-01

A gap conductance model, URGAP, has been developed with contributions from solid, fluid and radiation heat transfer components. Model parameters are easily available, independent of different combinations of material surfaces. The model parameters were fitted to 388 data points under reactor conditions. For model verification, another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Materials data were carefully checked over a wide range of temperatures. Special attention was paid to the contact term for high temperatures. Thus, the model can be applied to transients. The URGAP model is being used successfully in several codes (e.g. URANUS, SSYST). (author)

Development of gas-cooled fast reactor and its thermo-hydraulics

International Nuclear Information System (INIS)

Kawamura, Hiroshi

1977-10-01

Development, thermo-hydraulics and safety of GCFR are reviewed. The Development of Gas-Cooled Fast Reactor (GCFR) utilizes helium technology of HTGR and fuel technology of LMFBR. The breeding ratio of GCFR will be larger than that of LMFBR by about 0.2. Features of GCFR are a fuel with roughened surface to raise the heat transfer and vent system for the pressure equalization in the fuel rod. Helium as coolant of GCFR is chemically stable and stays in the single phase. So, there is no fuel-coolant interaction unlike the case of LMFBR. Since the helium must be pressurized, possibility of a depressurization accident is not negligible. In the United States, a 300MWe demonstration plant program is about to start; the collaboration with European countries is now quite active in this field. Though the development of GCFR started behind that of LMFBR, GCFR is equally promising as a fast breeder reactor. When realized, it will present possibility of a choice between these two. (auth.)

Fabrication, characterization, and modeling of a biodegradable battery for transient electronics

Science.gov (United States)

Edupuganti, Vineet; Solanki, Raj

2016-12-01

Traditionally, emphasis has been placed on durable, long-lasting electronics. However, electronics that are meant to intentionally degrade over time can actually have significant practical applications. Biodegradable, or transient, electronics would open up opportunities in the field of medical implants, where the need for surgical removal of devices could be eliminated. Environmental sensors and, eventually, consumer electronics would also greatly benefit from this technology. An essential component of transient electronics is the battery, which serves as a biodegradable power source. This work involves the fabrication, characterization, and modeling of a magnesium-based biodegradable battery. Galvanostatic discharge tests show that an anode material of magnesium alloy AZ31 extends battery lifetime by over six times, as compared to pure magnesium. With AZ31, the maximum power and capacity of the fabricated device are 67 μW and 5.2 mAh, respectively, though the anode area is just 0.8 cm2. The development of an equivalent circuit model provided insight into the battery's behavior by extracting fitting parameters from experimental data. The model can accurately simulate device behavior, taking into account its intentional degradation. The size of the device and the power it produces are in accordance with typical levels for low-power transient systems.

Transient combustion modeling of an oscillating lean premixed methane/air flam

NARCIS (Netherlands)

Withag, J.A.M.; Kok, Jacobus B.W.; Syed, Khawar

2009-01-01

The main objective of the present study is to demonstrate accurate low frequency transient turbulent combustion modeling. For accurate flame dynamics some improvements were made to the standard TFC combustion model for lean premixed combustion. With use of a 1D laminar flamelet code, predictions

A transient multi-scale model for direct methanol fuel cells

International Nuclear Information System (INIS)

Jahnke, T.; Zago, M.; Casalegno, A.; Bessler, W.G.; Latz, A.

2017-01-01

The DMFC is a promising option for backup power systems and for the power supply of portable devices. However, from the modeling point of view liquid-feed DMFC are challenging systems due to the complex electrochemistry, the inherent two-phase transport and the effect of methanol crossover. In this paper we present a physical 1D cell model to describe the relevant processes for DMFC performance ranging from electrochemistry on the surface of the catalyst up to transport on the cell level. A two-phase flow model is implemented describing the transport in gas diffusion layer and catalyst layer at the anode side. Electrochemistry is described by elementary steps for the reactions occurring at anode and cathode, including adsorbed intermediate species on the platinum and ruthenium surfaces. Furthermore, a detailed membrane model including methanol crossover is employed. The model is validated using polarization curves, methanol crossover measurements and impedance spectra. It permits to analyze both steady-state and transient behavior with a high level of predictive capabilities. Steady-state simulations are used to investigate the open circuit voltage as well as the overpotentials of anode, cathode and electrolyte. Finally, the transient behavior after current interruption is studied in detail.

Radiation heat transfer model in a spent fuel pool by TRACE code

International Nuclear Information System (INIS)

Sanchez-Saez, F.; Carlos, S.; Villanueva, J.F.; Martorell, S.

2014-01-01

Nuclear policies have experienced an important change since Fukushima Daiichi nuclear plant accident and the safety of spent fuels has been in the spot issue among all the safety concerns. The work presented consists of the thermohydraulic simulation of spent fuel pool behavior after a loss of coolant throughout transfer channel with loss of cooling transient is produced. The simulation is done with the TRACE code. One of the most important variables that define the behavior of the pool is cladding temperature, which evolution depends on the heat emission. In this work convection and radiation heat transfer is considered. When both heat transfer models are considered, a clear delay in achieving the maximum peak cladding temperature (1477 K) is observed compared with the simulation in which only convection heat transfer is considered. (authors)

Grounding line transient response in marine ice sheet models

Directory of Open Access Journals (Sweden)

A. S. Drouet

2013-03-01

Full Text Available Marine ice-sheet stability is mostly controlled by the dynamics of the grounding line, i.e. the junction between the grounded ice sheet and the floating ice shelf. Grounding line migration has been investigated within the framework of MISMIP (Marine Ice Sheet Model Intercomparison Project, which mainly aimed at investigating steady state solutions. Here we focus on transient behaviour, executing short-term simulations (200 yr of a steady ice sheet perturbed by the release of the buttressing restraint exerted by the ice shelf on the grounded ice upstream. The transient grounding line behaviour of four different flowline ice-sheet models has been compared. The models differ in the physics implemented (full Stokes and shallow shelf approximation, the numerical approach, as well as the grounding line treatment. Their overall response to the loss of buttressing is found to be broadly consistent in terms of grounding line position, rate of surface elevation change and surface velocity. However, still small differences appear for these latter variables, and they can lead to large discrepancies (> 100% observed in terms of ice sheet contribution to sea level when cumulated over time. Despite the recent important improvements of marine ice-sheet models in their ability to compute steady state configurations, our results question the capacity of these models to compute short-term reliable sea-level rise projections.

Analytical model for transient fluid mixing in upper outlet plenum of an LMFBR

International Nuclear Information System (INIS)

Yang, J.W.; Agrawal, A.K.

1976-01-01

A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the outlet plenum of an LMFBR. The maximum penetration of core flow is used as the criterion for dividing the sodium region into two mixing zones. The model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of by-pass flow into the plenum. The results of numerical calculations indicate that effects of flow stratification, chimney height, metal heat capacity and by-pass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas do not play any significant role on sodium temperature

BWR regional instability model and verification on ringhals-1 test

International Nuclear Information System (INIS)

Hotta, Akitoshi; Suzawa, Yojiro

1996-01-01

Regional instability is known as one type of the coupled neutronic-thermohydraulic phenomena of boiling water reactors (BWRs), where the thermohydraulic density wave propagation mechanism is predominant. Historically, it has been simulated by the three-dimensional time domain code in spite of its significant computing time. On the other hand, there have been proposals to apply the frequency domain models in regional instability considering the subcriticality of the higher neutronic mode. However, their application still remains in corewide instability mainly because of the lack of more detailed methodological and empirical studies. In this study, the current version of the frequency domain model was extended and verified based on actual core regional instability measurement data. The mathematical model LAPUR, the well-known frequency domain stability code, was reviewed from the standpoint of pure thermohydraulics and neutronic-thermohydraulic interaction mechanisms. Based on the ex-core loop test data, the original LAPUR mixed friction and local pressure loss model was modified, taking into account the different dynamic behavior of these two pressure-loss mechanisms. The perturbation term of the two-phase friction multiplier, which is the sum of the derivative of void fraction and subcool enthalpy, was adjusted theoretically. The adequacy of the instability evaluation system was verified based on the Ringhals unit 1 test data, which were supplied to participants of the Organization for Economic Cooperation and Development/Nuclear Energy Agency BWR Stability Benchmark Project

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)

A simplified model for tritium permeation transient predictions when trapping is active*1

Science.gov (United States)

Longhurst, G. R.

1994-09-01

This report describes a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. Comparison calculations with the verified and validated TMAP4 transient code show good agreement.

Transient Model Validation of Fixed-Speed Induction Generator Using Wind Farm Measurements

DEFF Research Database (Denmark)

Rogdakis, Georgios; Garcia-Valle, Rodrigo; Arana Aristi, Iván

2012-01-01

In this paper, an electromagnetic transient model for fixed-speed wind turbines equipped with induction generators is developed and implemented in PSCAD/EMTDC. The model is comprised by: an induction generator, aerodynamic rotor, and a two-mass representation of the shaft system. Model validation...

PWR hybrid computer model for assessing the safety implications of control systems

International Nuclear Information System (INIS)

Smith, O.L.; Booth, R.S.; Clapp, N.E.; DiFilippo, F.C.; Renier, J.P.; Sozer, A.

1985-01-01

The ORNL study of safety-related aspects of control systems consists of two interrelated tasks, (1) a failure mode and effects analysis that, in part, identifies single and multiple component failures that may lead to significant plant upsets, and (2) a hybrid computer model that uses these failures as initial conditions and traces the dynamic impact on the control system and remainder of the plant. The second task is reported here. The initial step in model development was to define a suitable interface between the FMEA and computer simulation tasks. This involved identifying primary plant components that must be simulated in dynamic detail and secondary components that can be treated adequately by the FMEA alone. The FMEA in general explores broader spectra of initiating events that may collapse into a reduced number of computer runs. A portion of the FMEA includes consideration of power supply failures. Consequences of the transients may feedback on the initiating causes, and there may be an interactive relationship between the FMEA and the computer simulation. Since the thrust of this program is to investigate control system behavior, the controls are modeled in detail to accurately reproduce characteristic response under normal and off-normal transients. The balance of the model, including neutronics, thermohydraulics and component submodels, is developed in sufficient detail to provide a suitable support for the control system

Investigation of transient models and performances for a doubly fed wind turbine under a grid fault

DEFF Research Database (Denmark)

Wang, M.; Zhao, B.; Li, H.

2011-01-01

fed induction generator (DFIG), the assessments of the impact on the electrical transient performances were investigated for the doubly fed wind turbine with different representations of wind turbine drive-train dynamics models, different initial operational conditions and different active crowbar...... crowbar on the transient performances of the doubly fed wind turbine were also investigated, with the possible reasonable trip time of crowbar. The investigation have shown that the transient performances are closely correlated with the wind turbine drive train models, initial operational conditions, key...

Mechanistic insights into a hydrate contribution to the Paleocene-Eocene carbon cycle perturbation from coupled thermohydraulic simulations

Science.gov (United States)

Minshull, T. A.; Marín-Moreno, H.; Armstrong McKay, D. I.; Wilson, P. A.

2016-08-01

During the Paleocene-Eocene Thermal Maximum (PETM), the carbon isotopic signature (δ13C) of surface carbon-bearing phases decreased abruptly by at least 2.5 to 3.0‰. This carbon isotope excursion (CIE) has been attributed to widespread methane hydrate dissociation in response to rapid ocean warming. We ran a thermohydraulic modeling code to simulate hydrate dissociation due to ocean warming for various PETM scenarios. Our results show that hydrate dissociation in response to such warming can be rapid but suggest that methane release to the ocean is modest and delayed by hundreds to thousands of years after the onset of dissociation, limiting the potential for positive feedback from emission-induced warming. In all of our simulations at least half of the dissociated hydrate methane remains beneath the seabed, suggesting that the pre-PETM hydrate inventory needed to account for all of the CIE is at least double that required for isotopic mass balance.

Thermo-Hydraulic Analysis of Heat Storage Filled with the Ceramic Bricks Dedicated to the Solar Air Heating System.

Science.gov (United States)

Nemś, Magdalena; Nemś, Artur; Kasperski, Jacek; Pomorski, Michał

2017-08-12

This article presents the results of a study into a packed bed filled with ceramic bricks. The designed storage installation is supposed to become part of a heating system installed in a single-family house and eventually to be integrated with a concentrated solar collector adapted to climate conditions in Poland. The system's working medium is air. The investigated temperature ranges and air volume flow rates in the ceramic bed were dictated by the planned integration with a solar air heater. Designing a packed bed of sufficient parameters first required a mathematical model to be constructed and heat exchange to be analyzed, since heat accumulation is a complex process influenced by a number of material properties. The cases discussed in the literature are based on differing assumptions and different formulas are used in calculations. This article offers a comparison of various mathematical models and of system operating parameters obtained from these models. The primary focus is on the Nusselt number. Furthermore, in the article, the thermo-hydraulic efficiency of the investigated packed bed is presented. This part is based on a relationship used in solar air collectors with internal storage.

Thermo-Hydraulic Analysis of Heat Storage Filled with the Ceramic Bricks Dedicated to the Solar Air Heating System

Science.gov (United States)

Nemś, Magdalena; Nemś, Artur; Kasperski, Jacek; Pomorski, Michał

2017-01-01

This article presents the results of a study into a packed bed filled with ceramic bricks. The designed storage installation is supposed to become part of a heating system installed in a single-family house and eventually to be integrated with a concentrated solar collector adapted to climate conditions in Poland. The system’s working medium is air. The investigated temperature ranges and air volume flow rates in the ceramic bed were dictated by the planned integration with a solar air heater. Designing a packed bed of sufficient parameters first required a mathematical model to be constructed and heat exchange to be analyzed, since heat accumulation is a complex process influenced by a number of material properties. The cases discussed in the literature are based on differing assumptions and different formulas are used in calculations. This article offers a comparison of various mathematical models and of system operating parameters obtained from these models. The primary focus is on the Nusselt number. Furthermore, in the article, the thermo-hydraulic efficiency of the investigated packed bed is presented. This part is based on a relationship used in solar air collectors with internal storage. PMID:28805703

Modeling and Controlling Flow Transient in Pipeline Systems: Applied for Reservoir and Pump Systems Combined with Simple Surge Tank

Directory of Open Access Journals (Sweden)

Itissam ABUIZIAH

2014-03-01

Full Text Available When transient conditions (water hammer exist, the life expectancy of the system can be adversely impacted, resulting in pump and valve failures and catastrophic pipe rupture. Hence, transient control has become an essential requirement for ensuring safe operation of water pipeline systems. To protect the pipeline systems from transient effects, an accurate analysis and suitable protection devices should be used. This paper presents the problem of modeling and simulation of transient phenomena in hydraulic systems based on the characteristics method. Also, it provides the influence of using the protection devices to control the adverse effects due to excessive and low pressure occuring in the transient. We applied this model for two main pipeline systems: Valve and pump combined with a simple surge tank connected to reservoir. The results obtained by using this model indicate that the model is an efficient tool for water hammer analysis. Moreover, using a simple surge tank reduces the unfavorable effects of transients by reducing pressure fluctuations.

Analysis Thermo-hydraulic of trajectories related to procedures for operation of Emergency (POE). Application to the loss of a train of the DTH; Analisis termohidraulico de trayectorias vinculadas a Procedimientos de Operacion de emergencia (POE). Aplicacion a la perdida de un tren de RHR

Energy Technology Data Exchange (ETDEWEB)

Sanchez-Saez, F.; Martorell Alsina, S.; Carlos Alberola, A.; Villanueva Lopez, J. F.; Martorell Aygues, P.

2012-07-01

This work explores different possible sequences at the loss of a train of the DTH when the plant is lowering power. The study of the different possible trajectories has been done through the collapse tool and study thermo-hydraulic each of these paths is done by the code TRACE Thermo-hydraulic.

Variable thickness transient ground-water flow model. Volume 1. Formulation

International Nuclear Information System (INIS)

Reisenauer, A.E.

1979-12-01

Mathematical formulation for the variable thickness transient (VTT) model of an aquifer system is presented. The basic assumptions are described. Specific data requirements for the physical parameters are discussed. The boundary definitions and solution techniques of the numerical formulation of the system of equations are presented

STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

Energy Technology Data Exchange (ETDEWEB)

Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

2016-05-01

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

Mathematical Model and Computational Analysis of Selected Transient States of Cylindrical Linear Induction Motor Fed via Frequency Converter

Directory of Open Access Journals (Sweden)

Andrzej Rusek

2008-01-01

Full Text Available The mathematical model of cylindrical linear induction motor (C-LIM fed via frequency converter is presented in the paper. The model was developed in order to analyze numerically the transient states. Problems concerning dynamics of ac-machines especially linear induction motor are presented in [1 – 7]. Development of C-LIM mathematical model is based on circuit method and analogy to rotary induction motor. The analogy between both: (a stator and rotor windings of rotary induction motor and (b winding of primary part of C-LIM (inductor and closed current circuits in external secondary part of C-LIM (race is taken into consideration. The equations of C-LIM mathematical model are presented as matrix together with equations expressing each vector separately. A computational analysis of selected transient states of C-LIM fed via frequency converter is presented in the paper. Two typical examples of C-LIM operation are considered for the analysis: (a starting the motor at various static loads and various synchronous velocities and (b reverse of the motor at the same operation conditions. Results of simulation are presented as transient responses including transient electromagnetic force, transient linear velocity and transient phase current.

Equilibrium and kinetic models for colloid release under transient solution chemistry conditions

Science.gov (United States)

We present continuum models to describe colloid release in the subsurface during transient physicochemical conditions. Our modeling approach relates the amount of colloid release to changes in the fraction of the solid surface area that contributes to retention. Equilibrium, kinetic, equilibrium and...

An integrated transient model for simulating the operation of natural gas transport systems

NARCIS (Netherlands)

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

Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

International Nuclear Information System (INIS)

Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio

2002-10-01

An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

Thermohydraulic design of saturated temperature capsule for IASCC irradiation test

Energy Technology Data Exchange (ETDEWEB)

Ide, Hiroshi; Matsui, Yoshinori; Itabashi, Yukio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

2002-10-01

An advanced water chemistry controlled irradiation research device is being developed in JAERI, to perform irradiation tests for irradiation assisted stress corrosion cracking (IASCC) research concerned with aging of LWR. This device enables the irradiation tests under the water chemistry condition and the temperature, which simulate the conditions for BWR core internals. The advanced water chemistry controlled irradiation research device is composed of saturated temperature capsule inserted into the JMTR core and the water chemistry control unit installed in the reactor building. Regarding the saturated temperature capsule, the Thermohydraulic design of capsule structure was done, aimed at controlling the specimen's temperature, feeding water velocity on specimen's surface to the environment of BWR nearer. As the result of adopting the new capsule structure based on the design study, it was found out that feeding water velocity at the surface of specimen's is increased to about 10 times as much as before, and nuclear heat generated in the capsule components can be removed safely even in the abnormal event such as the case of loss of feeding water. (author)

Comparisons with measured data of the simulated local core parameters by the coupled code ATHLET-BIPR-VVER applying a new enhanced model of the reactor pressure vessel

International Nuclear Information System (INIS)

Nikonov, S.; Pasichnyk, I.; Velkov, K.; Pautz, A.

2011-01-01

The paper describes the performed comparisons of measured and simulated local core data based on the OECD/NEA Benchmark on Kalinin-3 NPP: 'Switching off of one of the four operating main circulation pumps at nominal reactor power'. The local measurements of in core self-powered neutron detectors (SPND) in 64 fuel assemblies on 7 axial levels are used for the comparisons of the assemblies axial power distributions and the thermocouples readings at 93 fuel assembly heads are applied for the fuel assembly coolant temperature comparisons. The analyses are done on the base of benchmark transient calculations performed with the coupled system code ATHLET/BIPR-VVER. In order to describe more realistically the fluid mixing phenomena in a reactor pressure vessel a new enhanced nodalization scheme is being developed. It could take into account asymmetric flow behaviour in the reactor pressure vessel structures like downcomer, reactor core inlet and outlet, control rods' guided tubes, support grids etc. For this purpose details of the core geometry are modelled. About 58000 control volumes and junctions are applied. Cross connection are used to describe the interaction between the fluid objects. The performed comparisons are of great interest because they show some advantages by performing coupled code production pseudo-3D analysis of NPPs applying the parallel thermo-hydraulic channel methodology (or 1D thermo-hydraulic system code modeling). (Authors)

Transient Inverse Calibration of Site-Wide Groundwater Model to Hanford Operational Impacts from 1943 to 1996-Alternative Conceptual Model Considering Interaction with Uppermost Basalt Confined Aquifer; FINAL

International Nuclear Information System (INIS)

Vermeul, Vince R; Cole, Charles R; Bergeron, Marcel P; Thorne, Paul D; Wurstner, Signe K

2001-01-01

The baseline three-dimensional transient inverse model for the estimation of site-wide scale flow parameters, including their uncertainties, using data on the transient behavior of the unconfined aquifer system over the entire historical period of Hanford operations, has been modified to account for the effects of basalt intercommunication between the Hanford unconfined aquifer and the underlying upper basalt confined aquifer. Both the baseline and alternative conceptual models (ACM-1) considered only the groundwater flow component and corresponding observational data in the 3-Dl transient inverse calibration efforts. Subsequent efforts will examine both groundwater flow and transport. Comparisons of goodness of fit measures and parameter estimation results for the ACM-1 transient inverse calibrated model with those from previous site-wide groundwater modeling efforts illustrate that the new 3-D transient inverse model approach will strengthen the technical defensibility of the final model(s) and provide the ability to incorporate uncertainty in predictions related to both conceptual model and parameter uncertainty

Transient Model of a 10 MW Supercritical CO{sub 2} Brayton Cycle for Light Water Reactors by using MARS Code

Energy Technology Data Exchange (ETDEWEB)

Park, Joo-Hyun; Park, Hyun Sun; Kim, Moo Hwan [POSTECH, Pohang (Korea, Republic of); Bae, Sung Won; Cha, Jae-Eun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this study, recuperation cycle was chosen as a reference loop design and the MARS code was chosen as the transient cycle analysis code. Cycle design condition is focus on operation point of the light-water reactor. Development of a transient model was performed for 10MW-electron SCO{sub 2} coupled with light water reactors. In order to perform transient analysis, cycle transient model was developed and steady-state run was performed and presented in the paper. In this study, the transient model of SCO{sub 2} recuperation Brayton cycle was developed and implemented in MARS to study the steady-state simulation. We performed nodalization of the transient model using MARS code and obtained steady-state results. This study is shown that the supercritical CO{sub 2} Brayton cycle can be used as a power conversion system for light water reactors. Future work will include transient analysis such as partial road operation, power swing, start-up, and shutdown. Cycle control strategy will be considered for various control method.

An efficient approach to transient turbulent dispersion modeling by CFD-statistical analysis of a many-puff system

International Nuclear Information System (INIS)

Ching, W-H; K H Leung, Michael; Leung, Dennis Y C

2009-01-01

Transient turbulent dispersion phenomena can be found in various practical problems, such as the accidental release of toxic chemical vapor and the airborne transmission of infectious droplets. Computational fluid dynamics (CFD) is an effective tool for analyzing such transient dispersion behaviors. However, the transient CFD analysis is often computationally expensive and time consuming. In the present study, a computationally efficient CFD-statistical hybrid modeling method has been developed for studying transient turbulent dispersion. In this method, the source emission is represented by emissions of many infinitesimal puffs. Statistical analysis is performed to obtain first the statistical properties of the puff trajectories and subsequently the most probable distribution of the puff trajectories that represent the macroscopic dispersion behaviors. In two case studies of ambient dispersion, the numerical modeling results obtained agree reasonably well with both experimental measurements and conventional k-ε modeling results published in the literature. More importantly, the proposed many-puff CFD-statistical hybrid modeling method effectively reduces the computational time by two orders of magnitude.

A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector

Directory of Open Access Journals (Sweden)

Saas Laurent

2017-01-01

Full Text Available In the context of the simulation of the Severe Accidents (SA in Light Water Reactors (LWR, we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate…. During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4.

Response of Compacted Bentonites to Thermal and Thermo-Hydraulic Loadings at High Temperatures

Directory of Open Access Journals (Sweden)

Snehasis Tripathy

2017-07-01

Full Text Available The final disposal of high-level nuclear waste in many countries is preferred to be in deep geological repositories. Compacted bentonites are proposed for use as the buffer surrounding the waste canisters which may be subjected to both thermal and hydraulic loadings. A significant increase in the temperature is anticipated within the buffer, particularly during the early phase of the repository lifetime. In this study, several non-isothermal and non-isothermal hydraulic tests were carried on compacted MX80 bentonite. Compacted bentonite specimens (water content = 15.2%, dry density = 1.65 Mg/m3 were subjected to a temperature of either 85 or 150 °C at one end, whereas the temperature at the opposite end was maintained at 25 °C. During the non-isothermal hydraulic tests, water was supplied from the opposite end of the heat source. The temperature and relative humidity were monitored along predetermined depths of the specimens. The profiles of water content, dry density, and degree of saturation were established after termination of the tests. The test results showed that thermal gradients caused redistribution of the water content, whereas thermo-hydraulic gradients caused both redistribution and an increase in the water content within compacted bentonites, both leading to development of axial stress of various magnitudes. The applied water injection pressures (5 and 600 kPa and temperature gradients appeared to have very minimal impact on the magnitude of axial stress developed. The thickness of thermal insulation layer surrounding the testing devices was found to influence the temperature and relative humidity profiles thereby impacting the redistribution of water content within compacted bentonites. Under the influence of both the applied thermal and thermo-hydraulic gradients, the dry density of the bentonite specimens increased near the heat source, whereas it decreased at the opposite end. The test results emphasized the influence of

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)

Research on the behaviour of pressure suppression containment systems carried out at the University of Pisa

International Nuclear Information System (INIS)

Bigi, R.; Bovalini, R.; Mazzini, M.; Micheletti, E.

1978-01-01

A research programme has been carried out at the University of Pisa to study the thermo-hydraulic transient in pressure suppression containment systems during a LOCA. In the first series of experimental tests remarkable oscillations of pressure were observed both in dry and in wet-well. In order to describe these dynamic phenomena, a mathematical model has been set up; the main out-lines of this model are briefly described and the comparison between the calculated and experimental results is reported. (author)

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

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.

Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

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.

Coupling a transient solvent extraction module with the separations and safeguards performance model.

Energy Technology Data Exchange (ETDEWEB)

DePaoli, David W. (Oak Ridge National Laboratory, Oak Ridge, TN); Birdwell, Joseph F. (Oak Ridge National Laboratory, Oak Ridge, TN); Gauld, Ian C. (Oak Ridge National Laboratory, Oak Ridge, TN); Cipiti, Benjamin B.; de Almeida, Valmor F. (Oak Ridge National Laboratory, Oak Ridge, TN)

2009-10-01

A number of codes have been developed in the past for safeguards analysis, but many are dated, and no single code is able to cover all aspects of materials accountancy, process monitoring, and diversion scenario analysis. The purpose of this work was to integrate a transient solvent extraction simulation module developed at Oak Ridge National Laboratory, with the Separations and Safeguards Performance Model (SSPM), developed at Sandia National Laboratory, as a first step toward creating a more versatile design and evaluation tool. The SSPM was designed for materials accountancy and process monitoring analyses, but previous versions of the code have included limited detail on the chemical processes, including chemical separations. The transient solvent extraction model is based on the ORNL SEPHIS code approach to consider solute build up in a bank of contactors in the PUREX process. Combined, these capabilities yield a more robust transient separations and safeguards model for evaluating safeguards system design. This coupling and initial results are presented. In addition, some observations toward further enhancement of separations and safeguards modeling based on this effort are provided, including: items to be addressed in integrating legacy codes, additional improvements needed for a fully functional solvent extraction module, and recommendations for future integration of other chemical process modules.

Computer code HYDRO-ACE for analyzing thermo-hydraulic phenomena in the BWR core

International Nuclear Information System (INIS)

Abe, Kiyoharu; Naito, Yoshitaka

1979-10-01

A computer code HYDRO-ACE has been developed for analyzing thermo-hydraulic phenomena in the BWR core under forced or natural circulation of cooling water. The code is composed of two main calculation routines for single channels such as riser, separator, and downcommer and multiple channels such as the reactor core with a heated zone. Functionally the code is divided into many subroutines to be connected straightforwardly, and so that the user can choose a given course freely by simply arranging the subroutines. In the program, void fraction is calculated by Maurer's method, two-phase frictional pressure drop by Maltinelli-Nelson's, and critical heat flux ratio by Hench-Levy's. The coolant flow distributions in the JPDR-II core calculated by the code are in good agreement with those measured. (author)

Research on Model-Based Fault Diagnosis for a Gas Turbine Based on Transient Performance

Directory of Open Access Journals (Sweden)

Detang Zeng

2018-01-01

Full Text Available It is essential to monitor and to diagnose faults in rotating machinery with a high thrust–weight ratio and complex structure for a variety of industrial applications, for which reliable signal measurements are required. However, the measured values consist of the true values of the parameters, the inertia of measurements, random errors and systematic errors. Such signals cannot reflect the true performance state and the health state of rotating machinery accurately. High-quality, steady-state measurements are necessary for most current diagnostic methods. Unfortunately, it is hard to obtain these kinds of measurements for most rotating machinery. Diagnosis based on transient performance is a useful tool that can potentially solve this problem. A model-based fault diagnosis method for gas turbines based on transient performance is proposed in this paper. The fault diagnosis consists of a dynamic simulation model, a diagnostic scheme, and an optimization algorithm. A high-accuracy, nonlinear, dynamic gas turbine model using a modular modeling method is presented that involves thermophysical properties, a component characteristic chart, and system inertial. The startup process is simulated using this model. The consistency between the simulation results and the field operation data shows the validity of the model and the advantages of transient accumulated deviation. In addition, a diagnostic scheme is designed to fulfill this process. Finally, cuckoo search is selected to solve the optimization problem in fault diagnosis. Comparative diagnostic results for a gas turbine before and after washing indicate the improved effectiveness and accuracy of the proposed method of using data from transient processes, compared with traditional methods using data from the steady state.

Modeling and Analyzing Transient Military Air Traffic Control

Science.gov (United States)

2010-12-01

arrive and be serviced. In general, for n flights, the number of ways that flights can enter and leave the ATC is given by the nth Catalan number ...collection of information if it does not display a currently valid OMB control number . 1. REPORT DATE DEC 2010 2. REPORT TYPE 3. DATES COVERED 00-00...2010 to 00-00-2010 4. TITLE AND SUBTITLE Modeling and Analyzing Transient Military Air Traffic Control 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c

Developing semi-analytical solution for multiple-zone transient storage model with spatially non-uniform storage

Science.gov (United States)

Deng, Baoqing; Si, Yinbing; Wang, Jia

2017-12-01

Transient storages may vary along the stream due to stream hydraulic conditions and the characteristics of storage. Analytical solutions of transient storage models in literature didn't cover the spatially non-uniform storage. A novel integral transform strategy is presented that simultaneously performs integral transforms to the concentrations in the stream and in storage zones by using the single set of eigenfunctions derived from the advection-diffusion equation of the stream. The semi-analytical solution of the multiple-zone transient storage model with the spatially non-uniform storage is obtained by applying the generalized integral transform technique to all partial differential equations in the multiple-zone transient storage model. The derived semi-analytical solution is validated against the field data in literature. Good agreement between the computed data and the field data is obtained. Some illustrative examples are formulated to demonstrate the applications of the present solution. It is shown that solute transport can be greatly affected by the variation of mass exchange coefficient and the ratio of cross-sectional areas. When the ratio of cross-sectional areas is big or the mass exchange coefficient is small, more reaches are recommended to calibrate the parameter.

Numerical modeling of optical coherent transient processes with complex configurations-III: Noisy laser source

International Nuclear Information System (INIS)

Chang Tiejun; Tian Mingzhen

2007-01-01

A previously developed numerical model based on Maxwell-Bloch equations was modified to simulate optical coherent transient and spectral hole burning processes with noisy laser sources. Random walk phase noise was simulated using laser-phase sequences generated numerically according to the normal distribution of the phase shift. The noise model was tested by comparing the simulated spectral hole burning effect with the analytical solution. The noise effects on a few typical optical coherence transient processes were investigated using this numerical tool. Flicker and random walk frequency noises were considered in accumulation process

Implementation of a methodology to perform the uncertainty and sensitivity analysis of the control rod drop in a BWR

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. del C.

2015-07-01

A methodology to perform uncertainty and sensitivity analysis for the cross sections used in a Trace/PARCS coupled model for a control rod drop transient of a BWR-5 reactor was implemented with the neutronics code PARCS. A model of the nuclear reactor detailing all assemblies located in the core was developed. However, the thermohydraulic model designed in Trace was a simple model, where one channel representing all the types of assemblies located in the core, it was located inside a simple vessel model and boundary conditions were established. The thermohydraulic model was coupled with the neutronics model, first for the steady state and then a Control Rod Drop (CRD) transient was performed, in order to carry out the uncertainty and sensitivity analysis. To perform the analysis of the cross sections used in the Trace/PARCS coupled model during the transient, Probability Density Functions (PDFs) were generated for the 22 parameters cross sections selected from the neutronics parameters that PARCS requires, thus obtaining 100 different cases for the Trace/PARCS coupled model, each with a database of different cross sections. All these cases were executed with the coupled model, therefore obtaining 100 different outputs for the CRD transient with special emphasis on 4 responses per output: 1) The reactivity, 2) the percentage of rated power, 3) the average fuel temperature and 4) the average coolant density. For each response during the transient an uncertainty analysis was performed in which the corresponding uncertainty bands were generated. With this analysis it is possible to observe the results ranges of the responses chose by varying the uncertainty parameters selected. This is very useful and important for maintaining the safety in the nuclear power plants, also to verify if the uncertainty band is within of safety margins. The sensitivity analysis complements the uncertainty analysis identifying the parameter or parameters with the most influence on the

Implementation of a methodology to perform the uncertainty and sensitivity analysis of the control rod drop in a BWR

International Nuclear Information System (INIS)

Reyes F, M. del C.

2015-01-01

A methodology to perform uncertainty and sensitivity analysis for the cross sections used in a Trace/PARCS coupled model for a control rod drop transient of a BWR-5 reactor was implemented with the neutronics code PARCS. A model of the nuclear reactor detailing all assemblies located in the core was developed. However, the thermohydraulic model designed in Trace was a simple model, where one channel representing all the types of assemblies located in the core, it was located inside a simple vessel model and boundary conditions were established. The thermohydraulic model was coupled with the neutronics model, first for the steady state and then a Control Rod Drop (CRD) transient was performed, in order to carry out the uncertainty and sensitivity analysis. To perform the analysis of the cross sections used in the Trace/PARCS coupled model during the transient, Probability Density Functions (PDFs) were generated for the 22 parameters cross sections selected from the neutronics parameters that PARCS requires, thus obtaining 100 different cases for the Trace/PARCS coupled model, each with a database of different cross sections. All these cases were executed with the coupled model, therefore obtaining 100 different outputs for the CRD transient with special emphasis on 4 responses per output: 1) The reactivity, 2) the percentage of rated power, 3) the average fuel temperature and 4) the average coolant density. For each response during the transient an uncertainty analysis was performed in which the corresponding uncertainty bands were generated. With this analysis it is possible to observe the results ranges of the responses chose by varying the uncertainty parameters selected. This is very useful and important for maintaining the safety in the nuclear power plants, also to verify if the uncertainty band is within of safety margins. The sensitivity analysis complements the uncertainty analysis identifying the parameter or parameters with the most influence on the

Experimental study of core thermohydraulics in fast reactors during transition from forced to natural circulation. Influence of inter-wrapper flow

International Nuclear Information System (INIS)

Kamide, H.; Hayashi, K.; Momoi, K.

1997-01-01

The evaluation of core thermohydraulics under natural circulation conditions is important to utilize inherent safety features of Fast Reactors. When heat exchangers of a decay heat removal system are operated in an upper plenum of reactor vessel, cold sodium is provided by the heat exchangers. Core-plenum interactions will occur during a natural circulation condition due to this cold sodium in the upper plenum, e.g., it can penetrate into gap regions between fuel subassemblies (inter-wrapper flow, IWF) and the flow may reverse in low power core channels. These interactions will significantly modify the flow and temperature distributions in the core. Sodium experiments were carried out to study these phenomena. In a test section, seven subassemblies are housed and connected to an upper plenum. The influences of core-plenum interactions on the core thermohydraulics were investigated under steady state conditions and also in transitions from forced to natural circulation. Cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop due to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core. (author)

Qualification of the core model DYN3D coupled with the code ATHLET as an advanced tool for the accident analysis of VVER type reactors. Pt. 2. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Rohde, U.

2002-10-01

Benchmark calculations for the validation of the coupled neutron kinetics/thermohydraulic code complex DYN3D-ATHLET are described. Two benchmark problems concerning hypothetical accident scenarios with leaks in the steam system for a VVER-440 type reactor and the TMI-1 PWR have been solved. The first benchmark task has been defined by FZR in the frame of the international association 'Atomic Energy Research' (AER), the second exercise has been organized under the auspices of the OECD. While in the first benchmark the break of the main steam collector in the sub-critical hot zero power state of the reactor was considered, the break of one of the two main steam lines at full reactor power was assumed in the OECD benchmark. Therefore, in this exercise the mixing of the coolant from the intact and the defect loops had to be considered, while in the AER benchmark the steam collector break causes a homogeneous overcooling of the primary circuit. In the AER benchmark, each participant had to use its own macroscopic cross section libraries. In the OECD benchmark, the cross sections were given in the benchmark definition. The main task of both benchmark problems was to analyse the re-criticality of the scrammed reactor due to the overcooling. For both benchmark problems, a good agreement of the DYN3D-ATHLET solution with the results of other codes was achieved. Differences in the time of re-criticality and the height of the power peak between various solutions of the AER benchmark can be explained by the use of different cross section data. Significant differences in the thermohydraulic parameters (coolant temperature, pressure) occurred only at the late stage of the transient during the emergency injection of highly borated water. In the OECD benchmark, a broader scattering of the thermohydraulic results can be observed, while a good agreement between the various 3D reactor core calculations with given thermohydraulic boundary conditions was achieved. Reasons for the

Transient Mathematical Modeling for Liquid Rocket Engine Systems: Methods, Capabilities, and Experience

Science.gov (United States)

Seymour, David C.; Martin, Michael A.; Nguyen, Huy H.; Greene, William D.

2005-01-01

The subject of mathematical modeling of the transient operation of liquid rocket engines is presented in overview form from the perspective of engineers working at the NASA Marshall Space Flight Center. The necessity of creating and utilizing accurate mathematical models as part of liquid rocket engine development process has become well established and is likely to increase in importance in the future. The issues of design considerations for transient operation, development testing, and failure scenario simulation are discussed. An overview of the derivation of the basic governing equations is presented along with a discussion of computational and numerical issues associated with the implementation of these equations in computer codes. Also, work in the field of generating usable fluid property tables is presented along with an overview of efforts to be undertaken in the future to improve the tools use for the mathematical modeling process.

Interaction forces model on a bubble growing for nuclear best estimate computer codes

International Nuclear Information System (INIS)

Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Martinez-Mendez, Elizabeth J.

2005-01-01

This paper presents a mathematical model that takes into account the bubble radius variation that take place in a boiling water nuclear reactor during transients with changes in the pressure vessel, changes in the inlet core mass flow rate, density-wave phenomena or flow regime instability. The model with expansion effects was developed considering the interaction force between a dilute dispersion of gas bubbles and a continuous liquid phase. The closure relationships were formulated as an associated problem with the spatial deviation around averaging variables as a function of known variables. In order to solve the closure problem, a geometric model given by an eccentric unit cell was applied as an approach of heterogeneous structure of the two-phase flow. The closure relationship includes additional terms that represent combined effects between translation and pulsation due to displacement and size variation of the bubbles, respectively. This result can be implanted straightforward in best estimate thermo-hydraulics models. An example, the implementation of the closure relationships into TRAC best estimate computer code is presented

International benchmark for coupled codes and uncertainty analysis in modelling: switching-Off of one of the four operating main circulation pumps at nominal reactor power at NPP Kalinin unit 3

International Nuclear Information System (INIS)

Tereshonok, V. A.; Nikonov, S. P.; Lizorkin, M. P.; Velkov, K; Pautz, A.; Ivanov, V.

2008-01-01

The paper briefly describes the Specification of an international NEA/OECD benchmark based on measured plant data. During the commissioning tests for nominal power at NPP Kalinin Unit 3 a lot of measurements of neutron and thermo-hydraulic parameters have been carried out in the reactor pressure vessel, primary and the secondary circuits. One of the measured data sets for the transient 'Switching-off of one Main Circulation Pump (MCP) at nominal power' has been chosen to be applied for validation of coupled thermal-hydraulic and neutron-kinetic system codes and additionally for performing of uncertainty analyses as a part of the NEA/OECD Uncertainty Analysis in Modeling Benchmark. The benchmark is opened for all countries and institutions. The experimental data and the final specification with the cross section libraries will be provided to the participants from NEA/OECD only after official declaration of real participation in the benchmark and delivery of the simulated results of the transient for comparison. (Author)

A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector - 15271

International Nuclear Information System (INIS)

Saas, L.; Le Tellier, R.; Bajard, S.

2015-01-01

In this document, we present a simplified geometrical model (0D model) for both the in-core corium propagation transient and the characterization of the mode of corium transfer from the core to the vessel. A degraded core with a formed corium pool is used as an initial state. This initial state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate...). During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4

Electromagnetic transients in power cables

CERN Document Server

da Silva, Filipe Faria

2013-01-01

From the more basic concepts to the most advanced ones where long and laborious simulation models are required, Electromagnetic Transients in Power Cables provides a thorough insight into the study of electromagnetic transients and underground power cables. Explanations and demonstrations of different electromagnetic transient phenomena are provided, from simple lumped-parameter circuits to complex cable-based high voltage networks, as well as instructions on how to model the cables.Supported throughout by illustrations, circuit diagrams and simulation results, each chapter contains exercises,

Observation and control system of the thermohydraulic assays laboratory; Sistema de observacion y control del laboratorio de ensayos termohidraulicos

Energy Technology Data Exchange (ETDEWEB)

Santome, D; Hualde, R

1991-12-31

The Thermohydraulic Assays Laboratory (L.E.T.) is an installation whose purpose will be the components testing and the CAREM-25 reactor thermohydraulic processes operation dynamics. This plant is located at Pilcaniyeu, province of Rio Negro. Part of the tests which will be carried out consist in the use of different control strategies. The control of the systems by digital processors (control by software) has been decided to proceed with a maximum flexibility and capacity to make changes in the algorithms. This work describes the design and implementation of a digital control system to command the three circuits of the installation. (Author). [Espanol] El Laboratorio de Ensayos Termohidraulicos (L.E.T.) es una instalacion cuyo objeto sera el ensayo de componentes y de la dinamica de operacion de los procesos termohidraulicos del reactor CAREM-25. Esta planta esta localizada en Pilcaniyeu, provincia de Rio Negro. Parte de las pruebas que se efectuaran en el L.E.T. consisten en el empleo de distintas estrategias de control. Para disponer de una maxima flexibilidad y capacidad de efectuar cambios en los algoritmos, se decidio realizar el control de los sistemas por medio de procesadores digitales (control por software). Este trabajo consistio en el diseno e implementacion de un sistema de control digital distribuido para el comando de los tres circuitos con que cuenta la instalacion. (Autor).

Experimental investigations of heat exchange and hydrodynamics on models of a VG-400 steam generator tube bundle made up of small diameter helicoils

International Nuclear Information System (INIS)

Golovko, V.F; Ivaskov, N.A.; Obukhov, P.I.; Pospelov, V.N.; Sergeev, A.I.

1988-01-01

Features of HTGR steam generators having heat exchange surface made up of small diameter helicoils are discussed in the paper. A general approach to optimization of thermohydraulic characteristics BΓW-400 steam generator design backed by calculation and experiment are given. Main results of steam generator assembly's model aerodynamic test are presented. Data of thermohydraulic tests of a single tube model in a helium heated test rig are discussed. (author)

Comparison of transient PCRV model test results with analysis

International Nuclear Information System (INIS)

Marchertas, A.H.; Belytschko, T.B.

1979-01-01

Comparisons are made of transient data derived from simple models of a reactor containment vessel with analytical solutions. This effort is a part of the ongoing process of development and testing of the DYNAPCON computer code. The test results used in these comparisons were obtained from scaled models of the British sodium cooled fast breeder program. The test structure is a scaled model of a cylindrically shaped reactor containment vessel made of concrete. This concrete vessel is prestressed axially by holddown bolts spanning the top and bottom slabs along the cylindrical walls, and is also prestressed circumferentially by a number of cables wrapped around the vessel. For test purposes this containment vessel is partially filled with water, which comes in direct contact with the vessel walls. The explosive charge is immersed in the pool of water and is centrally suspended from the top of the vessel. The load history was obtained from an ICECO analysis, using the equations of state for the source and the water. A detailed check of this solution was made to assure that the derived loading did provide the correct input. The DYNAPCON code was then used for the analysis of the prestressed concrete containment model. This analysis required the simulation of prestressing and the response of the model to the applied transient load. The calculations correctly predict the magnitudes of displacements of the PCRV model. In addition, the displacement time histories obtained from the calculations reproduce the general features of the experimental records: the period elongation and amplitude increase as compared to an elastic solution, and also the absence of permanent displacement. However, the period still underestimates the experiment, while the amplitude is generally somewhat large

A simple heat transfer model for a heat flux plate under transient conditions

International Nuclear Information System (INIS)

Ryan, L.; Dale, J.D.

1985-01-01

Heat flux plates are used for measuring rates of heat transfer through surfaces under steady state and transient conditions. Their usual construction is to have a resistive layer bounded by thermopiles and an exterior layer for protection. If properly designed and constructed a linear relationship between the thermopile generated voltage and heat flux results and calibration under steady state conditions is straight forward. Under transient conditions however the voltage output from a heat flux plate cannot instantaneously follow the heat flux because of the thermal capacitance of the plate and the resulting time lag. In order to properly interpret the output of a heat flux plate used under transient conditions a simple heat transfer model was constructed and tested. (author)

Coupling a Transient Solvent Extraction Module with the Separations and Safeguards Performance Model

Energy Technology Data Exchange (ETDEWEB)

de Almeida, Valmor F [ORNL; Birdwell Jr, Joseph F [ORNL; DePaoli, David W [ORNL; Gauld, Ian C [ORNL

2009-10-01

A past difficulty in safeguards design for reprocessing plants is that no code existed for analysis and evaluation of the design. A number of codes have been developed in the past, but many are dated, and no single code is able to cover all aspects of materials accountancy, process monitoring, and diversion scenario analysis. The purpose of this work was to integrate a transient solvent extraction simulation module developed at Oak Ridge National Laboratory, with the SSPM Separations and Safeguards Performance Model, developed at Sandia National Laboratory, as a first step toward creating a more versatile design and evaluation tool. The SSPM was designed for materials accountancy and process monitoring analyses, but previous versions of the code have included limited detail on the chemical processes, including chemical separations. The transient solvent extraction model is based on the ORNL SEPHIS code approach to consider solute build up in a bank of contactors in the PUREX process. Combined, these capabilities yield a much more robust transient separations and safeguards model for evaluating safeguards system design. This coupling and the initial results are presented. In addition, some observations toward further enhancement of separations and safeguards modeling based on this effort are provided, including: items to be addressed in integrating legacy codes, additional improvements needed for a fully functional solvent extraction module, and recommendations for future integration of other chemical process modules.

Modeling transient streaming potentials in falling-head permeameter tests.

Science.gov (United States)

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.

Cable system transients theory, modeling and simulation

CERN Document Server

Ametani, Akihiro; Nagaoka, Naoto

2015-01-01

A systematic and comprehensive introduction to electromagnetic transient in cable systems, written by the internationally renowned pioneer in this field Presents a systematic and comprehensive introduction to electromagnetic transient in cable systems Written by the internationally renowned pioneer in the field Thorough coverage of the state of the art on the topic, presented in a well-organized, logical style, from fundamentals and practical applications A companion website is available

Understanding Transient Forcing with Plasma Instability Model, Ionospheric Propagation Model and GNSS Observations

Science.gov (United States)

Deshpande, K.; Zettergren, M. D.; Datta-Barua, S.

2017-12-01

Fluctuations in the Global Navigation Satellite Systems (GNSS) signals observed as amplitude and phase scintillations are produced by plasma density structures in the ionosphere. Phase scintillation events in particular occur due to structures at Fresnel scales, typically about 250 meters at ionospheric heights and GNSS frequency. Likely processes contributing to small-scale density structuring in auroral and polar regions include ionospheric gradient-drift instability (GDI) and Kelvin-Helmholtz instability (KHI), which result, generally, from magnetosphere-ionosphere interactions (e.g. reconnection) associated with cusp and auroral zone regions. Scintillation signals, ostensibly from either GDI or KHI, are frequently observed in the high latitude ionosphere and are potentially useful diagnostics of how energy from the transient forcing in the cusp or polar cap region cascades, via instabilities, to small scales. However, extracting quantitative details of instabilities leading to scintillation using GNSS data drastically benefits from both a model of the irregularities and a model of GNSS signal propagation through irregular media. This work uses a physics-based model of the generation of plasma density irregularities (GEMINI - Geospace Environment Model of Ion-Neutral Interactions) coupled to an ionospheric radio wave propagation model (SIGMA - Satellite-beacon Ionospheric-scintillation Global Model of the upper Atmosphere) to explore the cascade of density structures from medium to small (sub-kilometer) scales. Specifically, GEMINI-SIGMA is used to simulate expected scintillation from different instabilities during various stages of evolution to determine features of the scintillation that may be useful to studying ionospheric density structures. Furthermore we relate the instabilities producing GNSS scintillations to the transient space and time-dependent magnetospheric phenomena and further predict characteristics of scintillation in different geophysical

Development of the MARS input model for Kori nuclear units 1 transient analyzer

International Nuclear Information System (INIS)

Hwang, M.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Jeong, J. J.

2004-11-01

KAERI has been developing the 'NSSS transient analyzer' based on best-estimate codes for Kori Nuclear Units 1 plants. The MARS and RETRAN codes have been used as the best-estimate codes for the NSSS transient analyzer. Among these codes, the MARS code is adopted for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. So it is necessary to develop the MARS input model for Kori Nuclear Units 1 plants. This report includes the input model (hydrodynamic component and heat structure models) requirements and the calculation note for the MARS input data generation for Kori Nuclear Units 1 plant analyzer (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Kori Nuclear Units 1

Transient Analysis and Dosimetry of the Tokaimura Criticality Incident

International Nuclear Information System (INIS)

Pain, Christopher C.; Oliveira, Cassiano R.E. de; Goddard, Antony J. H.; Eaton, Matthew D.; Gundry, Sarah; Umpleby, Adrian P.

2003-01-01

This paper describes research on the application of the finite element transient criticality (FETCH) code to modeling and neutron dosimetry of the Tokaimura criticality incident. FETCH has been developed to model criticality transients in single and multiphase media and is applied here to fissile solution transient criticality. Since the initial transient behavior has different time scales and physics to the longer transient behavior, the transient modeling is divided into two parts: modeling the initial transient over a time scale of seconds in which radiolytic gases and free-surface sloshing play an important role in the transient - this provides information about the dose to workers; and modeling the long-term transient behavior following the initial transient that has a time scale over hours.The neutron dosimetry of worker A who received the largest dose during the Tokaimura criticality incident is also investigated here. This dose was received mainly in the first few seconds of the ensuing nuclear criticality transient. In addition to the multiorgan dosimetry of worker A, this work provides a method of helping to evaluate the yield in the initial phase of the criticality incident; it also shows how kinetic simulations can be calibrated so that they can be applied to investigate the physics behind the incident

Linear and quadrature models for data from treshold measurements of the transient visual system

NARCIS (Netherlands)

Brinker, den A.C.

1986-01-01

III this paper two models are considered for the transient visual system at threshold. One is a linear model and the other a model contain ing a quadrature element. Both models are commonly used on evidence from different experimental sources. It is shown that both models act in a similar fashion

A novel approach to model the transient behavior of solid-oxide fuel cell stacks

Science.gov (United States)

Menon, Vikram; Janardhanan, Vinod M.; Tischer, Steffen; Deutschmann, Olaf

2012-09-01

This paper presents a novel approach to model the transient behavior of solid-oxide fuel cell (SOFC) stacks in two and three dimensions. A hierarchical model is developed by decoupling the temperature of the solid phase from the fluid phase. The solution of the temperature field is considered as an elliptic problem, while each channel within the stack is modeled as a marching problem. This paper presents the numerical model and cluster algorithm for coupling between the solid phase and fluid phase. For demonstration purposes, results are presented for a stack operated on pre-reformed hydrocarbon fuel. Transient response to load changes is studied by introducing step changes in cell potential and current. Furthermore, the effect of boundary conditions and stack materials on response time and internal temperature distribution is investigated.

Modeling of Transient Nectar Flow in Hummingbird Tongues

Science.gov (United States)

Rico-Guevara, Alejandro; Fan, Tai-Hsi; Rubega, Margaret

2015-11-01

We demonstrate that hummingbirds do not pick up floral nectar via capillary action. The long believed capillary rise models were mistaken and unable to predict the dynamic nectar intake process. Instead, hummingbird's tongue acts as an elastic micropump. Nectar is drawn into the tongue grooves during elastic expansion after the grooves are squeezed flat by the beak. The new model is compared with experimental data from high-speed videos of 18 species and tens of individuals of wild hummingbirds. Self-similarity and transitions of short-to-long time behaviours have been resolved for the nectar flow driven by expansive filling. The transient dynamics is characterized by the relative contributions of negative excess pressure and the apparent area modulus of the tongue grooves.

Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2000-05-01

Worldwide there is considerable experience in nuclear power technology, especially in water cooled reactor technology. Of the operating plants, in September 1998, 346 were light water reactors (LWRs) totalling 306 GW(e) and 29 were heavy water reactors (HWRs) totalling 15 GW(e). The accumulated experience and lessons learned from these plants are being incorporated into new advanced reactor designs. Utility requirements documents have been formulated to guide these design activities by incorporating this experience, and results from research and development programmes, with the aim of reducing costs and licensing uncertainties by establishing the technical bases for the new designs. Common goals for advanced designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. Large water cooled reactors with power outputs of 1300 MW(e) and above, which possess inherent safety characteristics (e.g. negative Doppler moderator temperature coefficients, and negative moderator void coefficient) and incorporate proven, active engineered systems to accomplish safety functions are being developed. Other designs with power outputs from, for example, 220 MW(e) up to about 1300 MW(e) which also possess inherent safety characteristics and which place more emphasis on utilization of passive safety systems are being developed. Passive systems are based on natural forces and phenomena such as natural convection and gravity, making safety functions less dependent on active systems and components like pumps and diesel generators. In some cases, further experimental tests for the thermohydraulic conditions of interest in advanced designs can provide improved understanding of the phenomena. Further, analytical methods to predict reactor thermohydraulic behaviour can be qualified for use by comparison with the experimental results. These activities should ultimately result in more economical designs. The

Latest developments for a computer aided thermohydraulic network

International Nuclear Information System (INIS)

Alemberti, A.; Graziosi, G.; Mini, G.; Susco, M.

1999-01-01

Thermohydraulic networks are I-D systems characterized by a small number of basic components (pumps, valves, heat exchangers, etc) connected by pipes and limited spatially by a defined number of boundary conditions (tanks, atmosphere, etc). The network system is simulated by the well known computer program RELAPS/mod3. Information concerning the network geometry component behaviour, initial and boundary conditions are usually supplied to the RELAPS code using an ASCII input file by means of 'input cards'. CATNET (Computer Aided Thermalhydraulic NETwork) is a graphically user interface that, under specific user guidelines which completely define its range of applicability, permits a very high level of standardization and simplification of the RELAPS/mod3 input deck development process as well as of the output processing. The characteristics of the components (pipes, valves, pumps etc), defining the network system can be entered through CATNET. The CATNET interface is provided by special functions to compute form losses in the most typical bending and branching configurations. When the input of all system components is ready, CATNET is able to generate the RELAPS/mod3 input file. Finally, by means of CATNET, the RELAPS/mod3 code can be run and its output results can be transformed to an intuitive display form. The paper presents an example of application of the CATNET interface as well as the latest developments which greatly simplified the work of the users and allowed to reduce the possibility of input errors. (authors)

Transient recovery voltage analysis for various current breaking mathematical models: shunt reactor and capacitor bank de-energization study

Directory of Open Access Journals (Sweden)

Oramus Piotr

2015-09-01

Full Text Available Electric arc is a complex phenomenon occurring during the current interruption process in the power system. Therefore performing digital simulations is often necessary to analyse transient conditions in power system during switching operations. This paper deals with the electric arc modelling and its implementation in simulation software for transient analyses during switching conditions in power system. Cassie, Cassie-Mayr as well as Schwarz-Avdonin equations describing the behaviour of the electric arc during the current interruption process have been implemented in EMTP-ATP simulation software and presented in this paper. The models developed have been used for transient simulations to analyse impact of the particular model and its parameters on Transient Recovery Voltage in different switching scenarios: during shunt reactor switching-off as well as during capacitor bank current switching-off. The selected simulation cases represent typical practical scenarios for inductive and capacitive currents breaking, respectively.

A COMETHE version with transient capability

International Nuclear Information System (INIS)

Vliet, J. van; Lebon, G.; Mathieu, P.

1980-01-01

A version of the COMETHE code is under development to simulate transient situations. This paper focuses on some aspects of the transient heat transfer models. Initially the coupling between transient heat transfer and other thermomechanical models is discussed. An estimation of the thermal characteristic times shows that the cladding temperatures are often in quasi-steady state. In order to reduce the computing time, calculations are therefore switched from a transient to a quasi-static numerical procedure as soon as such a quasi-equilibrium is detected. The temperature calculation is performed by use of the Lebon-Lambermont restricted variational principle, with piecewise polynoms as trial functions. The method has been checked by comparison with some exact results and yields good agreement for transient as well as for quasi-static situations. This method therefore provides a valuable tool for the simulation of the transient behaviour of nuclear reactor fuel rods. (orig.)

Application of transient burning rate model of solid propellant in electrothermal-chemical launch simulation

Directory of Open Access Journals (Sweden)

Yan-jie Ni

2016-04-01

Full Text Available A 30 mm electrothermal-chemical (ETC gun experimental system is employed to research the burning rate characteristics of 4/7 high-nitrogen solid propellant. Enhanced gas generation rates (EGGR of propellants during and after electrical discharges are verified in the experiments. A modified 0D internal ballistic model is established to simulate the ETC launch. According to the measured pressure and electrical parameters, a transient burning rate law including the influence of EGGR coefficient by electric power and pressure gradient (dp/dt is added into the model. The EGGR coefficient of 4/7 high-nitrogen solid propellant is equal to 0.005 MW−1. Both simulated breech pressure and projectile muzzle velocity accord with the experimental results well. Compared with Woodley's modified burning rate law, the breech pressure curves acquired by the transient burning rate law are more consistent with test results. Based on the parameters calculated in the model, the relationship among propellant burning rate, pressure gradient (dp/dt and electric power is analyzed. Depending on the transient burning rate law and experimental data, the burning of solid propellant under the condition of plasma is described more accurately.

Models for Type Ia Supernovae and Related Astrophysical Transients

Science.gov (United States)

Röpke, Friedrich K.; Sim, Stuart A.

2018-06-01

We give an overview of recent efforts to model Type Ia supernovae and related astrophysical transients resulting from thermonuclear explosions in white dwarfs. In particular we point out the challenges resulting from the multi-physics multi-scale nature of the problem and discuss possible numerical approaches to meet them in hydrodynamical explosion simulations and radiative transfer modeling. We give examples of how these methods are applied to several explosion scenarios that have been proposed to explain distinct subsets or, in some cases, the majority of the observed events. In case we comment on some of the successes and shortcoming of these scenarios and highlight important outstanding issues.

COMETHE III-M for transient fuel rod behaviour prediction

International Nuclear Information System (INIS)

Billaux, M.; Vliet, J. van

1983-01-01

The COMETHE III-M version is being developed in order to provide fuel rod behaviour prediction capability both in steady-state and in transient situations. It also allows to estimate the fuel rod enthalpy evolution versus time or burnup which may be important in core-related safety studies. This paper describes the transient heat transfer models, including transient heat conduction inside the fuel rod, and a subchannel model providing transient flow as well as enthalpy calculation capability. Transient fission gas release is also modelled on basis of the change rate of oxide temperature. The models are illustrated by a few calculation examples. (author)

Multiphysics model of thermomechanical and helium-induced damage of tungsten during plasma heat transients

Energy Technology Data Exchange (ETDEWEB)

Crosby, Tamer, E-mail: tcrosby@ucla.edu; Ghoniem, Nasr M., E-mail: ghoniem@ucla.edu

2013-11-15

A combination of transient heating and bombardment by helium and hydrogen atoms has been experimentally proven to lead to severe surface and sub-surface damage. We developed a computational model to determine the relationship between the thermomechanical loading conditions and the onset of damage and failure of tungsten surfaces. The model is based on a thermoelasticity fracture damage approach that was developed using the phase field method. The model simulates the distribution of helium bubbles inside the grains and on grain boundaries using space-dependent rate theory. In addition, the model is coupled with a transient heat conduction analysis for temperature distributions inside the material. The results show the effects of helium bubbles on reducing tungsten surface energy. Further, a temperature gradient in the material equals to 10 K/μm, resulted in deep cracks propagating from the tungsten surface.

Dual Transformer Model based on Standard Circuit Elements for the Study of Low- and Mid-frequency Transients

Science.gov (United States)

Jazebi, Saeed

This thesis is a step forward toward achieving the final objective of creating a fully dual model for transformers including eddy currents and nonlinearities of the iron core using the fundamental electrical components already available in the EMTP-type programs. The model is effective for the study of the performance of transformers during power system transients. This is very important for transformer designers, because the insulation of transformers is determined with the overvoltages caused by lightning or switching operations. There are also internally induced transients that occur when a switch is actuated. For example switching actions for reconfiguration of distribution systems that offers economic advantages, or protective actions to clear faults and large short-circuit currents. Many of the smart grid concepts currently under development by many utilities rely heavily on switching to optimize resources that produce transients in the system. On the other hand, inrush currents produce mechanical forces which deform transformer windings and cause malfunction of the differential protection. Also, transformer performance under ferroresonance and geomagnetic induced currents are necessary to study. In this thesis, a physically consistent dual model applicable to single-phase two-winding transformers is proposed. First, the topology of a dual electrical equivalent circuit is obtained from the direct application of the principle of duality. Then, the model parameters are computed considering the variations of the transformer electromagnetic behavior under various operating conditions. Current modeling techniques use different topological models to represent diverse transient situations. The reversible model proposed in this thesis unifies the terminal and topological equivalent circuits. The model remains invariable for all low-frequency transients including deep saturation conditions driven from any of the two windings. The very high saturation region of the

Proceedings of the 4. General Congress on Nuclear Energy. v. 1

International Nuclear Information System (INIS)

1992-01-01

The first volume of the 4. General Congress on Nuclear Energy presents works about transient and accident analysis; thermohydraulics; radioisotopes uses; fuel cycle; shield; material technology; nuclear installation construction and reactor and nuclear physics. (C.G.C.)

Bifurcation structures and transient chaos in a four-dimensional Chua model

Energy Technology Data Exchange (ETDEWEB)

Hoff, Anderson, E-mail: hoffande@gmail.com; Silva, Denilson T. da; Manchein, Cesar, E-mail: cesar.manchein@udesc.br; Albuquerque, Holokx A., E-mail: holokx.albuquerque@udesc.br

2014-01-10

A four-dimensional four-parameter Chua model with cubic nonlinearity is studied applying numerical continuation and numerical solutions methods. Regarding numerical solution methods, its dynamics is characterized on Lyapunov and isoperiodic diagrams and regarding numerical continuation method, the bifurcation curves are obtained. Combining both methods the bifurcation structures of the model were obtained with the possibility to describe the shrimp-shaped domains and their endoskeletons. We study the effect of a parameter that controls the dimension of the system leading the model to present transient chaos with its corresponding basin of attraction being riddled.

Development of an Aeroelastic Modeling Capability for Transient Nozzle Side Load Analysis

Science.gov (United States)

Wang, Ten-See; Zhao, Xiang; Zhang, Sijun; Chen, Yen-Sen

2013-01-01

Lateral nozzle forces are known to cause severe structural damage to any new rocket engine in development during test. While three-dimensional, transient, turbulent, chemically reacting computational fluid dynamics methodology has been demonstrated to capture major side load physics with rigid nozzles, hot-fire tests often show nozzle structure deformation during major side load events, leading to structural damages if structural strengthening measures were not taken. The modeling picture is incomplete without the capability to address the two-way responses between the structure and fluid. The objective of this study is to develop a coupled aeroelastic modeling capability by implementing the necessary structural dynamics component into an anchored computational fluid dynamics methodology. The computational fluid dynamics component is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, while the computational structural dynamics component is developed in the framework of modal analysis. Transient aeroelastic nozzle startup analyses of the Block I Space Shuttle Main Engine at sea level were performed. The computed results from the aeroelastic nozzle modeling are presented.

Sensitiveness Analysis of Neutronic Parameters Due to Uncertainty in Thermo-hydraulic parameters on CAREM-25 Reactor

International Nuclear Information System (INIS)

Serra, Oscar

2000-01-01

Some studies were done about the effect of the uncertainty in the values of several thermo-hydraulic parameters on the core behaviour of the CAREM-25 reactor.By using the chain codes CITVAP-THERMIT and the perturbation the reference states, it was found that concerning to the total power, the effects were not very important, but were much bigger for the pressure.Furthermore were hardly significant in the presence of any perturbation on the void fraction calculation and the fuel temperature.The reactivity and the power peaking factor had highly important changes in the case of the coolant flow.We conclude that the use of this procedure is adequate and useful to our purpose

Historical survey of the qualifying process of Furnas calculus methodology in the areas of rods, neutronics, thermohydraulic accidents and transients

International Nuclear Information System (INIS)

Conti, C.F.S.; Silva Galetti, M.R. da.

1990-02-01

As Furnas intends to assume in the future the responsibility of performing Safety Analyses associated to Reload and Operation questions to Angra 1, it was figured out the necessity of qualifying its methodology by CNEN. The Methodology Qualification Process is based on guidelines proposed by CNEN at NT-DR-N o 02/87, where it was divided in four steps. This Technical Note aims to present the follow up of FURNAS Methodology Qualification Process and to bring it up to date in the areas of Core Physics (Neutronics), Core Thermal-Hydraulics, Fuel Rod Behaviour, Transient and Large Break Loss of Coolant Accident Analyses (LBLOCA). (author)

Development of safety analysis codes for light water reactor

International Nuclear Information System (INIS)

Akimoto, Masayuki

1985-01-01

An overview is presented of currently used major codes for the prediction of thermohydraulic transients in nuclear power plants. The overview centers on the two-phase fluid dynamics of the coolant system and the assessment of the codes. Some of two-phase phenomena such as phase separation are not still predicted with engineering accuracy. MINCS-PIPE are briefly introduced. The MINCS-PIPE code is to assess constitutive relations and to aid development of various experimental correlations for 1V1T model to 2V2T model. (author)

Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 1. Numerical investigation for the rationalization of hydrodynamics in the upper plenum

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Yamaguchi, Akira

2002-02-01

A large-scale sodium-cooled fast breeder reactor in feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for a gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, a thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermohydraulic characteristics in the upper plenum, and to investigate suitable in-vessel structure for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) It is difficult to rationale in-vessel flow patterns through adjustments of porous ratio and pressure loss for a hold down plate and baffle plates installed in an upper core structure. (2) Dummy plug insertion to a slit of the upper core structure is one of effective measures to stabilize in-vessel flow patterns. (3) Flow guide devices such as a baffle ring and a partial inner barrel are also effective measures to eliminate impinging jet to a dipped plate (D/P) and to reduce horizontal flow velocity components at free surface. (4) Installations of labyrinth structures to a R/V - D/P gap is successful for decreasing of free surface horizontal flows. (5) Gap closing of an in-vessel fuel pot and two cold trap components has the effects of reductions for free surface horizontal flows and for the difference of free surface levels. Following future investigations are important preventive measures against the gas entrainment from the free surface. (1) Flattening of spatial axial velocity distributions at the R/V - D/P gap. (2) Alleviation measures of vortex concentration at free surface. (3) Separation measures of 3-dimensional vortex

Nonlinear modeling and dynamic analysis of hydro-turbine governing system in the process of load rejection transient

International Nuclear Information System (INIS)

Zhang, Hao; Chen, Diyi; Xu, Beibei; Wang, Feifei

2015-01-01

Graphical abstract: Nonlinear dynamic transfer coefficients are introduced to the hydro-turbine governing system. In the process of load reject ion transient, the nonlinear dynamical behaviors of the system are studied in detail. - Highlights: • A novel mathematical model of a hydro-turbine governing system is established. • The process of load rejection transient is considered. • Nonlinear dynamic transfer coefficients are introduced to the system. • The bifurcation diagram with the variable t has better engineering significance. • The nonlinear dynamical behaviors of the system are studied in detail. - Abstract: This article pays attention to the mathematical modeling of a hydro-turbine governing system in the process of load rejection transient. As a pioneer work, the nonlinear dynamic transfer coefficients are introduced in a penstock system. Considering a generator system, a turbine system and a governor system, we present a novel nonlinear dynamical model of a hydro-turbine governing system. Fortunately, for the unchanged of PID parameters, we acquire the stable regions of the governing system in the process of load rejection transient by numerical simulations. Moreover, the nonlinear dynamic behaviors of the governing system are illustrated by bifurcation diagrams, Poincare maps, time waveforms and phase orbits. More importantly, these methods and analytic results will present theoretical groundwork for allowing a hydropower station in the process of load rejection transient

Mountain scale modeling of transient, coupled gas flow, heat transfer and carbon-14 migration

International Nuclear Information System (INIS)

Lu, Ning; Ross, B.

1993-01-01

We simulate mountain-scale coupled heat transfer and gas flow at Yucca Mountain. A coupled rock-gas flow and heat transfer model, TGIF2, is used to simulate mountain-scale two-dimensional transient heat transfer and gas flow. The model is first verified against an analytical solution for the problem of an infinite horizontal layer of fluid heated from below. Our numerical results match very well with the analytical solution. Then, we obtain transient temperature and gas flow distributions inside the mountain. These distributions are used by a transient semianalytical particle tracker to obtain carbon-14 travel times for particles starting at different locations within the repository. Assuming that the repository is filled with 30-year-old waste at an initial areal power density of 57 kw/acre, we find that repository temperatures remain above 60 degrees C for more than 10,000 years. Carbon-14 travel times to the surface are mostly less than 1000 years, for particles starting at any time within the first 10,000 years

Equilibrium and kinetic models for colloid release under transient solution chemistry conditions.

Science.gov (United States)

Bradford, Scott A; Torkzaban, Saeed; Leij, Feike; Simunek, Jiri

2015-10-01

We present continuum models to describe colloid release in the subsurface during transient physicochemical conditions. Our modeling approach relates the amount of colloid release to changes in the fraction of the solid surface area that contributes to retention. Equilibrium, kinetic, equilibrium and kinetic, and two-site kinetic models were developed to describe various rates of colloid release. These models were subsequently applied to experimental colloid release datasets to investigate the influence of variations in ionic strength (IS), pH, cation exchange, colloid size, and water velocity on release. Various combinations of equilibrium and/or kinetic release models were needed to describe the experimental data depending on the transient conditions and colloid type. Release of Escherichia coli D21g was promoted by a decrease in solution IS and an increase in pH, similar to expected trends for a reduction in the secondary minimum and nanoscale chemical heterogeneity. The retention and release of 20nm carboxyl modified latex nanoparticles (NPs) were demonstrated to be more sensitive to the presence of Ca(2+) than D21g. Specifically, retention of NPs was greater than D21g in the presence of 2mM CaCl2 solution, and release of NPs only occurred after exchange of Ca(2+) by Na(+) and then a reduction in the solution IS. These findings highlight the limitations of conventional interaction energy calculations to describe colloid retention and release, and point to the need to consider other interactions (e.g., Born, steric, and/or hydration forces) and/or nanoscale heterogeneity. Temporal changes in the water velocity did not have a large influence on the release of D21g for the examined conditions. This insensitivity was likely due to factors that reduce the applied hydrodynamic torque and/or increase the resisting adhesive torque; e.g., macroscopic roughness and grain-grain contacts. Our analysis and models improve our understanding and ability to describe the amounts

Nonlinear time-domain cochlear model for transient stimulation and human otoacoustic emission

DEFF Research Database (Denmark)

Verhulst, Sarah; Dau, Torsten; Shera, Christopher A.

2012-01-01

This paper describes the implementation and performance of a nonlinear time-domain model of the cochlea for transient stimulation and human otoacoustic emission generation. The nonlinearity simulates compressive growth of measured basilar-membrane impulse responses. The model accounts...... for reflection and distortion-source otoacoustic emissions (OAEs) and simulates spontaneous OAEs through manipulation of the middle-ear reflectance. The model was calibrated using human psychoacoustical and otoacoustic tuning parameters. It can be used to investigate time-dependent properties of cochlear...

A model for transient analysis of a multiple-medium confinement filter system

International Nuclear Information System (INIS)

Hyder, M.L.; Ellison, P.G.; Leonard, M.T.; Louie, D.L.Y.; Donbroski, E.L.; Wagner, K.C.

1990-01-01

A computational model is described that calculates the transient behavior of aerosol and vapor (adsorption) filter compartments such as those used in the Savannah River Site (SRS) production reactor confinement system. The principal application of the model is in the analysis of confinement response to hypothetical severe (core melt) accidents. Under these conditions, aerosol and radio-iodine deposition on filter compartments may be substantial. Attendant filter degradation mechanisms are modeled. Sample calculations are included to illustrate model performance. 6 refs., 14 figs., 1 tab

Transient analysis of DTT rakes

International Nuclear Information System (INIS)

Kamath, P.S.; Lahey, R.T. Jr.

1981-01-01

This paper presents an analytical model for the determination of the cross-sectionally averaged transient mass flux of a two-phase fluid flowing in a conduit instrumented by a Drag-Disk Turbine Transducer (DTT) Rake and a multibeam gamma densitometer. Parametric studies indicate that for a typical blowdown transient, dynamic effects such as rotor inertia can be important for the turbine-meter. In contrast, for the drag-disk, a frequency response analysis showed that the quasisteady solution is valid below a forcing frequency of about 10 Hz, which is faster than the time scale normally encountered during blowdowns. The model showed reasonably good agreement with full scale transient rake data, where the flow regimes were mostly homogeneous or stratified, thus indicating that the model is suitable for the analysis of a DTT rake. (orig.)

Modelling of phenomena associated with high burnup fuel behaviour during overpower transients

International Nuclear Information System (INIS)

Sills, H.E.; Langman, V.J.; Iglesias, F.C.

1995-01-01

Phenomena of importance to the behaviour of high burnup fuel subjected to conditions of rapid overpower (i.e., LWR RIAs) include the change in cladding material properties due to irradiation, pellet-clad interaction (PCI) and 'rim' effects associated with the periphery of high burnup fuel. 'Rim' effects are postulated to be caused by changes in fuel morphology at high burnup. Typical discharge burnups for CANDU fuel are low compared to LWRs. Maximum linear ratings for CANDU fuel are higher than those for LWRs. However, under normal operating conditions, the Zircaloy-4 clad of the CANDU fuel is collapsed onto the fuel stack. Thus, the CANDU fuel performance codes model the transient behaviour of the fuel-to-clad interface and are capable of assessing the potential for pellet-clad mechanical interaction (PCMI) failures for a wide range of overpower conditions. This report provides a discussion of the modelling of the phenomena of importance to high burnup fuel behaviour during rapid overpower transients. (author)

Development of the MARS input model for Ulchin 1/2 transient analyzer

International Nuclear Information System (INIS)

Jeong, J. J.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Chung, B. D.; Hwang, M.

2003-03-01

KAERI has been developing the NSSS transient analyzer based on best-estimate codes for Ulchin 1/2 plants. The MARS and RETRAN code are used as the best-estimate codes for the NSSS transient analyzer. Among the two codes, the MARS code is to be used for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. This report includes the input model requirements and the calculation note for the Ulchin 1/2 MARS input data generation (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Ulchin 1/2

Development of the MARS input model for Ulchin 3/4 transient analyzer

International Nuclear Information System (INIS)

Jeong, J. J.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Hwang, M. G.

2003-12-01

KAERI has been developing the NSSS transient analyzer based on best-estimate codes.The MARS and RETRAN code are adopted as the best-estimate codes for the NSSS transient analyzer. Among these two codes, the MARS code is to be used for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. This report includes the MARS input model requirements and the calculation note for the MARS input data generation (see the Appendix) for Ulchin 3/4 plant analyzer. In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Ulchin 3/4

Modeling steady state and transient fission gas behaviour with the Karlsruhe code LAKU

International Nuclear Information System (INIS)

Vaeth, L.

1984-08-01

The programme LAKU models the behaviour of gaseous fission products in reactor fuel under steady state and transient conditions, including molten fuel. A presentation of the full model is given, starting with gas behaviour in the grains and on grain faces and including the treatment of release from porosity. The results of some recent calculations are presented. (orig.) [de

RETRAN sensitivity studies of light water reactor transients. Final report

International Nuclear Information System (INIS)

Burrell, N.S.; Gose, G.C.; Harrison, J.F.; Sawtelle, G.R.

1977-06-01

This report presents the results of sensitivity studies performed using the RETRAN/RELAP4 transient analysis code to identify critical parameters and models which influence light water reactor transient predictions. Various plant transients for both boiling water reactors and pressurized water reactors are examined. These studies represent the first detailed evaluation of the RETRAN/RELAP4 transient code capability in predicting a variety of plant transient responses. The wide range of transients analyzed in conjunction with the parameter and modeling studies performed identify several sensitive areas as well as areas requiring future study and model development

A simulation model for transient response of a gas separation module using a hollow fiber membrane

Energy Technology Data Exchange (ETDEWEB)

Sugiyama, Takahiko, E-mail: t-sugiyama@nucl.nagoya-u.ac.jp [Nagoya University, Fro-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Miyahara, Naoya [Nagoya University, Fro-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Tanaka, Masahiro [National Institute for Fusion Science, Oroshi-cho 322-6, Toki 509-5292 (Japan); Munakata, Kenzo [Akita University, Tegata Gakuen-cho 1-1, Akita-shi, Akita 010-8502 (Japan); Yamamoto, Ichiro [Nagoya University, Fro-cho, Chikusa-ku, Nagoya 464-8603 (Japan)

2011-10-15

A simulation model has been developed for transient response of a gas separation module using a hollow fiber membrane for the removal of tritium from the atmosphere of the confinement space. The mass transfer process such as sorption and desorption of gases at the surface of the dense layer and the porous support layer, diffusive transfer in the both layers are treated in the model. Sorption isotherm, mass transfer rate and permeance are estimated through step-wise transient response experiments. The present model represents well not only separation factors and recovery ratio at the steady state but also responses to the multi-step wise change in the sweep gas rate.

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

Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model

Energy Technology Data Exchange (ETDEWEB)

Baudron, Anne-Marie, E-mail: anne-marie.baudron@cea.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex (France); Lautard, Jean-Jacques, E-mail: jean-jacques.lautard@cea.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex (France); Maday, Yvon, E-mail: maday@ann.jussieu.fr [Sorbonne Universités, UPMC Univ Paris 06, UMR 7598, Laboratoire Jacques-Louis Lions and Institut Universitaire de France, F-75005, Paris (France); Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); Brown Univ, Division of Applied Maths, Providence, RI (United States); Riahi, Mohamed Kamel, E-mail: riahi@cmap.polytechnique.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CMAP, Inria-Saclay and X-Ecole Polytechnique, Route de Saclay, 91128 Palaiseau Cedex (France); Salomon, Julien, E-mail: salomon@ceremade.dauphine.fr [CEREMADE, Univ Paris-Dauphine, Pl. du Mal. de Lattre de Tassigny, F-75016, Paris (France)

2014-12-15

In this paper we present a time-parallel algorithm for the 3D neutrons calculation of a transient model in a nuclear reactor core. The neutrons calculation consists in numerically solving the time dependent diffusion approximation equation, which is a simplified transport equation. The numerical resolution is done with finite elements method based on a tetrahedral meshing of the computational domain, representing the reactor core, and time discretization is achieved using a θ-scheme. The transient model presents moving control rods during the time of the reaction. Therefore, cross-sections (piecewise constants) are taken into account by interpolations with respect to the velocity of the control rods. The parallelism across the time is achieved by an adequate use of the parareal in time algorithm to the handled problem. This parallel method is a predictor corrector scheme that iteratively combines the use of two kinds of numerical propagators, one coarse and one fine. Our method is made efficient by means of a coarse solver defined with large time step and fixed position control rods model, while the fine propagator is assumed to be a high order numerical approximation of the full model. The parallel implementation of our method provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch–Maurer–Werner benchmark.

A transient model of a cesium-barium diode

International Nuclear Information System (INIS)

Luke, J.R.; El-Genk, M.S.

1995-01-01

In this work a transient model of a Cs-Ba diode is developed, and a series of experiments is performed using a diode equipped with Langmuir probes. The Langmuir probe data show that the electron energy distribution is non-Maxwellian at low discharge currents, indicating the presence of an electron beam from the emitter. Experimental results also showed that the plasma properties are non-homogeneous across the 1 mm diode gap; the electron temperature and plasma potential were higher near the emitter and the plasma density was higher near the collector. Experimental evidence is presented to show that the discharge contracts to a filament below the maximum thermal emission current

Coupled code analysis of uncertainty and sensitivity of Kalinin-3 benchmark

Energy Technology Data Exchange (ETDEWEB)

Pasichnyk, Ihor; Zwermann, Winfried; Velkov, Kiril [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Nikonov, Sergey [VNIIAES, Moscow (Russian Federation)

2016-09-15

An uncertainty and sensitivity analysis is performed for the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). A switch off of one main coolant pump (MCP) at nominal reactor power is calculated using a coupled thermohydraulic and neutron-kinetic ATHLET-PARCS code. The objectives are to study uncertainty of total reactor power and to identify the main sources of reactor power uncertainty. The GRS uncertainty and sensitivity software package XSUSA is applied to propagate uncertainties in nuclear data libraries to the full core coupled transient calculations. A set of most important thermal-hydraulic parameters of the primary circuit is identified and a total of 23 thermohydraulic parameters are statistically varied using GRS code SUSA. The ATHLET model contains also a balance-of-plant (BOP) model which is simulated using ATHLET GCSM module. In particular the operation of the main steam generator regulators is modelled in detail. A set of 200 varied coupled ATHLET-PARCS calculations is analyzed. The results obtained show a clustering effect in the behavior of global reactor parameters. It is found that the GCSM system together with varied input parameters strongly influence the overall nuclear power plant behavior and can even lead to a new scenario. Possible reasons of the clustering effect are discussed in the paper. This work is a step forward in establishing a ''best-estimate calculations in combination with performing uncertainty analysis'' methodology for coupled full core calculations.

A large deviations approach to the transient of the Erlang loss model

NARCIS (Netherlands)

Mandjes, M.R.H.; Ridder, Annemarie

2001-01-01

This paper deals with the transient behavior of the Erlang loss model. After scaling both arrival rate and number of trunks, an asymptotic analysis of the blocking probability is given. Apart from that, the most likely path to blocking is given. Compared to Shwartz and Weiss [Large Deviations for

Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code; Implicacion de las capacidades de union fenosa dentro del area de termohidraulica en el APS de la C.N. Jose Cabrera. Aplicaciones del codigo RELAP5/MOD2

Energy Technology Data Exchange (ETDEWEB)

Martin, L; Saenz Tejada, P [Empresarios Agrupados, A.I.E., Madrid (Spain)

1993-12-15

When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

Transient stratification modelling of a corium pool in a LWR vessel lower head

International Nuclear Information System (INIS)

Le Tellier, R.; Saas, L.; Bajard, S.

2015-01-01

Highlights: • A kinetic stratification model is proposed for the simulation of the in-vessel corium behaviour during a LWR severe accident. • The different associated “modes” of vessel failure by thermal focusing effect are highlighted and discussed. • A sensitivity study for a 1650 MWe GenIII PWR is presented with this model in order to illustrate the associated R&D issues. - Abstract: In the context of light water reactor severe accidents analysis, this paper is focused on one key parameter of in-vessel corium phenomenology: the immiscible phases stratification and its impact on the heat flux distribution at the corium pool lateral boundary with the so-called focusing effect related to a “thin” top metal phase and the potential vessel failure at that point. More particularly, based on the limited knowledge of the stratification transient phenomenon derived from the MASCA-RCW experiment, a basic model is proposed that can be used for corium in lower head sensitivity analyses. It has been implemented in the PROCOR platform developed at CEA Cadarache. A short parametric study on a simple hypothetical transient is presented in order to highlight the different focusing effect “modes” that can be encountered based on this in-vessel corium pool model. An early mode may occur during the formation of the top metal layer while two other modes may appear later during the thinning of this top metal layer because of thermochemically induced mass transfers. Some associated relevant parameters (model or scenario-dependent) and modelling issues are mentioned and illustrated with some results of a Monte-Carlo based sensitivity calculation on the transient behaviour of the corium in the lower head of a 1650 MWe GenIII PWR. Within the limiting modelling hypotheses, the thermal modelling of the steel layer for small (centimetre) heights and the mass diffusivity (limited in this case to the uranium diffusivity in the oxidic layer) are main sensitive parameters

Thermonuclear model for x-ray transients

International Nuclear Information System (INIS)

Wallace, R.K.; Woosley, S.E.; Weaver, T.A.

1982-01-01

The thermonuclear evolution of a 1.41 M sub solar neutron star accreting both solar and metal-deficient mixtures of hydrogen, helium, and heavy elements at rates ranging from about 10 -11 to 10 -10 M sub solar per year is examined using a one-dimensional numerical model. The metal deficient compositions may result either from placement of the neutron star in a binary system with a Population II red giant or from gravitational settling of heavy ions in the accreted material. For such accretion rates and metallicities, hydrogen burning, mediated by the β-limited CNO cycle, is stable and leads to the accumulation of a thick helium layer with mass 10 23 to 10 25 g and temperature 0.7 less than or equal to T 8 less than or equal to 1.2. Helium ignition occurs under extremely degenerate circumstances and is catastrophically violent. In the lower t helium shells this runaway is propagated as a convective deflagration, for the thicker layers a detonation front is set up which steepens into a strong relativistic shock wave in the neutron star envelope. In all models greatly super-Eddington luminosities in the outer layers of the neutron star lead to a sustained epoch of radiatively driven mass loss. Observationally, such models may correspond to rapid x-ray transients. The hopeless prospect for constructing a one-dimensional model for γ-ray bursts without magnetic field confinement is discussed and uncertainties pointed out in the strong screening correction for helium burning reaction

Implicit thermohydraulic coupling of two-phause flow calculations

International Nuclear Information System (INIS)

Lekach, S.; Kaufman, J.M.

1980-01-01

A numerical scheme that implicitly couples the hydraulic variables with thermal variables during a one or two-phase transient calculation in a one-dimensional pipe is presented. The transients are performed to achieve a steady-state condition. It is shown that by preserving the strong interdependence that exists between the hydraulic and thermal variables with an implicit flux treatment, it is possible to achieve a greater degree of numerical stability and in less computer time than with an explicit treatment. The method is slightly more complex but the large time step advantage more than pays for the overhead

FDTD modelling of induced polarization phenomena in transient electromagnetics

Science.gov (United States)

Commer, Michael; Petrov, Peter V.; Newman, Gregory A.

2017-04-01

The finite-difference time-domain scheme is augmented in order to treat the modelling of transient electromagnetic signals containing induced polarization effects from 3-D distributions of polarizable media. Compared to the non-dispersive problem, the discrete dispersive Maxwell system contains costly convolution operators. Key components to our solution for highly digitized model meshes are Debye decomposition and composite memory variables. We revert to the popular Cole-Cole model of dispersion to describe the frequency-dependent behaviour of electrical conductivity. Its inversely Laplace-transformed Debye decomposition results in a series of time convolutions between electric field and exponential decay functions, with the latter reflecting each Debye constituents' individual relaxation time. These function types in the discrete-time convolution allow for their substitution by memory variables, annihilating the otherwise prohibitive computing demands. Numerical examples demonstrate the efficiency and practicality of our algorithm.

Winnerless competition principle and prediction of the transient dynamics in a Lotka-Volterra model

Science.gov (United States)

Afraimovich, Valentin; Tristan, Irma; Huerta, Ramon; Rabinovich, Mikhail I.

2008-12-01

Predicting the evolution of multispecies ecological systems is an intriguing problem. A sufficiently complex model with the necessary predicting power requires solutions that are structurally stable. Small variations of the system parameters should not qualitatively perturb its solutions. When one is interested in just asymptotic results of evolution (as time goes to infinity), then the problem has a straightforward mathematical image involving simple attractors (fixed points or limit cycles) of a dynamical system. However, for an accurate prediction of evolution, the analysis of transient solutions is critical. In this paper, in the framework of the traditional Lotka-Volterra model (generalized in some sense), we show that the transient solution representing multispecies sequential competition can be reproducible and predictable with high probability.

Description of a heat transfer model suitable to calculate transient processes of turbocharged diesel engines with one-dimensional gas-dynamic codes

Energy Technology Data Exchange (ETDEWEB)

Galindo, J.; Lujan, J.M.; Serrano, J.R.; Dolz, V. [CMT-Motores Termicos, Universidad Politecnica de Valencia, Valencia (Spain); Guilain, S. [Renault s.a.s., Lardy (France)

2006-01-15

This paper describes a heat transfer model to be implemented in a global engine 1-D gas-dynamic code to calculate reciprocating internal combustion engine performance in steady and transient operations. A trade off between simplicity and accuracy has been looked for, in order to fit with the stated objective. To validate the model, the temperature of the exhaust manifold wall in a high-speed direct injection (HSDI) turbocharged diesel engine has been measured during a full load transient. In addition, an indirect assessment of the exhaust gas temperature during this transient process has been carried out. The results show good agreement between the measured and modelled data with good accuracy to predict the engine performance. A dual-walled air gap exhaust manifold has been tested in order to quantify the potential of exhaust gas thermal energy saving on engine transient performance. The experimental results together with the heat transfer model have been used to analyse the influence of thermal energy saving on dynamic performance during the load transient of an HSDI turbocharged diesel engine. (author)

Modelling of thermalhydraulics and reactor physics in simulators

International Nuclear Information System (INIS)

Miettinen, J.

1994-01-01

The evolution of thermalhydraulic analysis methods for analysis and simulator purposes has brought closer the thermohydraulic models in both application areas. In large analysis codes like RELAP5, TRAC, CATHARE and ATHLET the accuracy for calculating complicated phenomena has been emphasized, but in spite of large development efforts many generic problems remain unsolved. For simulator purposes fast running codes have been developed and these include only limited assessment efforts. But these codes have more simulator friendly features than large codes, like portability and modular code structure. In this respect the simulator experiences with SMABRE code are discussed. Both large analysis codes and special simulator codes have their advances in simulator applications. The evolution of reactor physical calculation methods in simulator applications has started from simple point kinetic models. For analysis purposes accurate 1-D and 3-D codes have been developed being capable for fast and complicated transients. For simulator purposes capability for simulation of instruments has been emphasized, but the dynamic simulation capability has been less significant. The approaches for 3-dimensionality in simulators requires still quite much development, before the analysis accuracy is reached. (orig.) (8 refs., 2 figs., 2 tabs.)

Modeling the release of E. coli D21g with transients in water content

Science.gov (United States)

Transients in water content are well known to mobilize colloids that are retained in the vadose zone. However, there is no consensus on the proper model formulation to simulate colloid release during drainage and imbibition. We present a model that relates colloid release to changes in the air-water...

Modeling photocurrent transients in organic solar cells

International Nuclear Information System (INIS)

Hwang, I; Greenham, N C

2008-01-01

We investigate the transient photocurrents of organic photovoltaic devices in response to a sharp turn-on of illumination, by numerical modeling of the drift-diffusion equations. We show that the photocurrent turn-on dynamics are determined not only by the transport dynamics of free charges, but also by the time required for the population of geminate charge pairs to reach its steady-state value. The dissociation probability of a geminate charge pair is found to be a key parameter in determining the device performance, not only by controlling the efficiency at low intensities, but also in determining the fate of charge pairs formed by bimolecular recombination at high intensities. Bimolecular recombination is shown to reduce the turn-on time at high intensities, since the typical distance traveled by a charge pair is reduced.

Recent development of transient electronics

Directory of Open Access Journals (Sweden)

Huanyu Cheng

2016-01-01

Full Text Available Transient electronics are an emerging class of electronics with the unique characteristic to completely dissolve within a programmed period of time. Since no harmful byproducts are released, these electronics can be used in the human body as a diagnostic tool, for instance, or they can be used as environmentally friendly alternatives to existing electronics which disintegrate when exposed to water. Thus, the most crucial aspect of transient electronics is their ability to disintegrate in a practical manner and a review of the literature on this topic is essential for understanding the current capabilities of transient electronics and areas of future research. In the past, only partial dissolution of transient electronics was possible, however, total dissolution has been achieved with a recent discovery that silicon nanomembrane undergoes hydrolysis. The use of single- and multi-layered structures has also been explored as a way to extend the lifetime of the electronics. Analytical models have been developed to study the dissolution of various functional materials as well as the devices constructed from this set of functional materials and these models prove to be useful in the design of the transient electronics.

Application of Thermal Network Model to Transient Thermal Analysis of Power Electronic Package Substrate

Directory of Open Access Journals (Sweden)

Masaru Ishizuka

2011-01-01

Full Text Available In recent years, there is a growing demand to have smaller and lighter electronic circuits which have greater complexity, multifunctionality, and reliability. High-density multichip packaging technology has been used in order to meet these requirements. The higher the density scale is, the larger the power dissipation per unit area becomes. Therefore, in the designing process, it has become very important to carry out the thermal analysis. However, the heat transport model in multichip modules is very complex, and its treatment is tedious and time consuming. This paper describes an application of the thermal network method to the transient thermal analysis of multichip modules and proposes a simple model for the thermal analysis of multichip modules as a preliminary thermal design tool. On the basis of the result of transient thermal analysis, the validity of the thermal network method and the simple thermal analysis model is confirmed.

Flow and heat transfer thermohydraulic modelisation during the reflooding phase of a P.W.R.'s core

International Nuclear Information System (INIS)

Raymond, Patrick

1978-04-01

Some generalities about L.O.C.A. are first recalled. The French experimental studies about Emergency Core Cooling System are briefly described. The different heat transfer mechanisms to take into account, according to the flow pattern in the dry zone, and the correlations or methods to calculate them, are defined. Then the Thermohydraulic code computer: FLIRA, which describe the reflooding phase, and a modelisation taking into account the different flow patterns are setted. A first interpretation of ERSEC experiments with a tubular test section shows that it is possible, with this modelisation and some classical heat transfer correlations, to describe the reflooding phase. [fr

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

Modelling the transient behaviour of pulsed current tungsten-inert-gas weldpools

Science.gov (United States)

Wu, C. S.; Zheng, W.; Wu, L.

1999-01-01

A three-dimensional model is established to simulate the pulsed current tungsten-inert-gas (TIG) welding process. The goal is to analyse the cyclic variation of fluid flow and heat transfer in weldpools under periodic arc heat input. To this end, an algorithm, which is capable of handling the transience, nonlinearity, multiphase and strong coupling encountered in this work, is developed. The numerical simulations demonstrate the transient behaviour of weldpools under pulsed current. Experimental data are compared with numerical results to show the effectiveness of the developed model.

Transient modelling of a natural circulation loop under variable pressure

International Nuclear Information System (INIS)

Vianna, Andre L.B.; Faccini, Jose L.H.; Su, Jian; Instituto de Engenharia Nuclear

2017-01-01

The objective of the present work is to model the transient operation of a natural circulation loop, which is one-tenth scale in height to a typical Passive Residual Heat Removal system (PRHR) of an Advanced Pressurized Water Nuclear Reactor and was designed to meet the single and two-phase flow similarity criteria to it. The loop consists of a core barrel with electrically heated rods, upper and lower plena interconnected by hot and cold pipe legs to a seven-tube shell heat exchanger of countercurrent design, and an expansion tank with a descending tube. A long transient characterized the loop operation, during which a phenomenon of self-pressurization, without self-regulation of the pressure, was experimentally observed. This represented a unique situation, named natural circulation under variable pressure (NCVP). The self-pressurization was originated in the air trapped in the expansion tank and compressed by the loop water dilatation, as it heated up during each experiment. The mathematical model, initially oriented to the single-phase flow, included the heat capacity of the structure and employed a cubic polynomial approximation for the density, in the buoyancy term calculation. The heater was modelled taking into account the different heat capacities of the heating elements and the heater walls. The heat exchanger was modelled considering the coolant heating, during the heat exchanging process. The self-pressurization was modelled as an isentropic compression of a perfect gas. The whole model was computationally implemented via a set of finite difference equations. The corresponding computational algorithm of solution was of the explicit, marching type, as for the time discretization, in an upwind scheme, regarding the space discretization. The computational program was implemented in MATLAB. Several experiments were carried out in the natural circulation loop, having the coolant flow rate and the heating power as control parameters. The variables used in the

Transient modelling of a natural circulation loop under variable pressure

Energy Technology Data Exchange (ETDEWEB)

Vianna, Andre L.B.; Faccini, Jose L.H.; Su, Jian, E-mail: avianna@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br, E-mail: faccini@ien.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Termo-Hidraulica Experimental

2017-07-01

The objective of the present work is to model the transient operation of a natural circulation loop, which is one-tenth scale in height to a typical Passive Residual Heat Removal system (PRHR) of an Advanced Pressurized Water Nuclear Reactor and was designed to meet the single and two-phase flow similarity criteria to it. The loop consists of a core barrel with electrically heated rods, upper and lower plena interconnected by hot and cold pipe legs to a seven-tube shell heat exchanger of countercurrent design, and an expansion tank with a descending tube. A long transient characterized the loop operation, during which a phenomenon of self-pressurization, without self-regulation of the pressure, was experimentally observed. This represented a unique situation, named natural circulation under variable pressure (NCVP). The self-pressurization was originated in the air trapped in the expansion tank and compressed by the loop water dilatation, as it heated up during each experiment. The mathematical model, initially oriented to the single-phase flow, included the heat capacity of the structure and employed a cubic polynomial approximation for the density, in the buoyancy term calculation. The heater was modelled taking into account the different heat capacities of the heating elements and the heater walls. The heat exchanger was modelled considering the coolant heating, during the heat exchanging process. The self-pressurization was modelled as an isentropic compression of a perfect gas. The whole model was computationally implemented via a set of finite difference equations. The corresponding computational algorithm of solution was of the explicit, marching type, as for the time discretization, in an upwind scheme, regarding the space discretization. The computational program was implemented in MATLAB. Several experiments were carried out in the natural circulation loop, having the coolant flow rate and the heating power as control parameters. The variables used in the

Design base transient analysis using the real-time nuclear reactor simulator model

International Nuclear Information System (INIS)

Tien, K.K.; Yakura, S.J.; Morin, J.P.; Gregory, M.V.

1987-01-01

A real-time simulation model has been developed to describe the dynamic response of all major systems in a nuclear process reactor. The model consists of a detailed representation of all hydraulic components in the external coolant circulating loops consisting of piping, valves, pumps and heat exchangers. The reactor core is described by a three-dimensional neutron kinetics model with detailed representation of assembly coolant and moderator thermal hydraulics. The models have been developed to support a real-time training simulator, therefore, they reproduce system parameters characteristic of steady state normal operation with high precision. The system responses for postulated severe transients such as large pipe breaks, loss of pumping power, piping leaks, malfunctions in control rod insertion, and emergency injection of neutron absorber are calculated to be in good agreement with reference safety analyses. Restrictions were imposed by the requirement that the resulting code be able to run in real-time with sufficient spare time to allow interfacing with secondary systems and simulator hardware. Due to hardware set-up and real plant instrumentation, simplifications due to symmetry were not allowed. The resulting code represents a coarse-node engineering model in which the level of detail has been tailored to the available computing power of a present generation super-minicomputer. Results for several significant transients, as calculated by the real-time model, are compared both to actual plant data and to results generated by fine-mesh analysis codes

Design base transient analysis using the real-time nuclear reactor simulator model

International Nuclear Information System (INIS)

Tien, K.K.; Yakura, S.J.; Morin, J.P.; Gregory, M.V.

1987-01-01

A real-time simulation model has been developed to describe the dynamic response of all major systems in a nuclear process reactor. The model consists of a detailed representation of all hydraulic components in the external coolant circulating loops consisting of piping, valves, pumps and heat exchangers. The reactor core is described by a three-dimensional neutron kinetics model with detailed representation of assembly coolant and mode-rator thermal hydraulics. The models have been developed to support a real-time training simulator, therefore, they reproduce system parameters characteristic of steady state normal operation with high precision. The system responses for postulated severe transients such as large pipe breaks, loss of pumping power, piping leaks, malfunctions in control rod insertion, and emergency injection of neutron absorber are calculated to be in good agreement with reference safety analyses. Restrictions were imposed by the requirement that the resulting code be able to run in real-time with sufficient spare time to allow interfacing with secondary systems and simulator hardware. Due to hardware set-up and real plant instrumentation, simplifications due to symmetry were not allowed. The resulting code represents a coarse-node engineering model in which the level of detail has been tailored to the available computing power of a present generation super-minicomputer. Results for several significant transients, as calculated by the real-time model, are compared both to actual plant data and to results generated by fine-mesh analysis codes

Characterizing SI Engine Transient Fuel Consumption in ALPHA

Science.gov (United States)

Examine typical transient engine operation encountered over the EPA's vehicle and engine testing drive cycles to characterize that transient fuel usage, and then describe the changes made to ALPHA to better model transient engine operation.

Transient modeling of an air conditioner with a rapid cycling compressor and multi-indoor units

Energy Technology Data Exchange (ETDEWEB)

Zhang, Wei-Jiang [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Zhang, Chun-Lu [College of Mechanical Engineering, Tongji University, 4800 Cao An Highway, Shanghai 201804 (China)

2011-01-15

Rapid cycling the compressor is an alternative of the variable speed compressor to modulate the capacity of refrigeration systems for the purpose of energy saving at part-load conditions. The multi-evaporator air conditioner combined with the rapid cycling compressor brings difficulties in control design because of the sophisticated system physics and dynamics. In this paper the transient model of a multi-split air conditioner with a digital scroll compressor is developed for predicting the system transients under performance modulations. The predicted cycling dynamics are in good agreement with the experimental data. Based on the validated model, the impact of compressor idle power and cycle period to the part load performance is discussed. (author)

Transient modeling of an air conditioner with a rapid cycling compressor and multi-indoor units

Energy Technology Data Exchange (ETDEWEB)

Zhang Weijiang [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Zhang Chunlu, E-mail: chunlu.zhang@carrier.utc.co [College of Mechanical Engineering, Tongji University, 4800 Cao An Highway, Shanghai 201804 (China)

2011-01-15

Rapid cycling the compressor is an alternative of the variable speed compressor to modulate the capacity of refrigeration systems for the purpose of energy saving at part-load conditions. The multi-evaporator air conditioner combined with the rapid cycling compressor brings difficulties in control design because of the sophisticated system physics and dynamics. In this paper the transient model of a multi-split air conditioner with a digital scroll compressor is developed for predicting the system transients under performance modulations. The predicted cycling dynamics are in good agreement with the experimental data. Based on the validated model, the impact of compressor idle power and cycle period to the part load performance is discussed.

Transient modeling of an air conditioner with a rapid cycling compressor and multi-indoor units

International Nuclear Information System (INIS)

Zhang Weijiang; Zhang Chunlu

2011-01-01

Rapid cycling the compressor is an alternative of the variable speed compressor to modulate the capacity of refrigeration systems for the purpose of energy saving at part-load conditions. The multi-evaporator air conditioner combined with the rapid cycling compressor brings difficulties in control design because of the sophisticated system physics and dynamics. In this paper the transient model of a multi-split air conditioner with a digital scroll compressor is developed for predicting the system transients under performance modulations. The predicted cycling dynamics are in good agreement with the experimental data. Based on the validated model, the impact of compressor idle power and cycle period to the part load performance is discussed.

Mathematical modelling of plant transients in the PWR for simulator purposes

International Nuclear Information System (INIS)

Hartel, K.

1984-01-01

This chapter presents the results of the testing of anticipated and abnormal plant transients in pressurized water reactors (PWRs) of the type WWER 440 by means of the numerical simulation of 32 different, stationary and nonstationary, operational regimes. Topics considered include the formation of the PWR mathematical model, the physical approximation of the reactor core, the structure of the reactor core model, a mathematical approximation of the reactor model, the selection of numerical methods, and a computerized simulation system. The necessity of a PWR simulator in Czechoslovakia is justified by the present status and the outlook for the further development of the Czechoslovak nuclear power complex

Model of Transient Process Where Three-Phase Transducer Feeds Induction Motor Equivalent as a Variable Active-Inductive Load

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.

A distributed parameter wire model for transient electrical discharges

International Nuclear Information System (INIS)

Maier, W.B. II; Kadish, A.; Sutherland, C.D.; Robiscoe, R.T.

1990-01-01

A model for freely propagating transient electrical discharges, such as lightning and punch-through arcs, is developed in this paper. We describe the electromagnetic fields by Maxwell's equations and we represent the interaction of electric fields with the medium to produce current by ∂J/∂t=ω 2 (E-E*J)/4π, where ω and E* are parameters characteristic of the medium, J≡current density, and J≡J/|J|. We illustrate the properties of this model for small-diameter, guided, cylindrically symmetric discharges. Analytic, numerical, and approximate solutions are given for special cases. The model describes, in a new and comprehensive fashion, certain macroscopic discharge properties, such as threshold behavior, quenching and reignition, path tortuosity, discharge termination with nonzero charge density remaining along the discharge path, and other experimentally observed discharge phenomena. Fields, current densities, and charge densities are quantitatively determined from given boundary and initial conditions. We suggest that many macroscopic discharge properties are properly explained by the model as electromagnetic phenomena, and we discuss extensions of the model to include chemistry, principally ionization and recombination

Transients in the Vivitron

International Nuclear Information System (INIS)

Cooke, C.M.; Frick, G.; Roumie, M.

1993-01-01

Electrical measurements are presented for the construction of a model for the study of transients in the Vivitron. Observation of the transmission of electrical pulses in the porticos clearly shows transmission-line behaviour. Measurements of the vector impedance of the outer porticos show the same transmission-line properties, but also gives a description of the modification from a pure transmission line due to the circular electrodes. The results of this investigation should allow the construction of a computer model which predicts the evolution of the transients in the case of a spark in the Vivitron. (orig.)

Fuel rod modelling during transients: The TOUTATIS code

International Nuclear Information System (INIS)

Bentejac, F.; Bourreau, S.; Brochard, J.; Hourdequin, N.; Lansiart, S.

2001-01-01

The TOUTATIS code is devoted to the PCI local phenomena simulation, in correlation with the METEOR code for the global behaviour of the fuel rod. More specifically, the TOUTATIS objective is to evaluate the mechanical constraints on the cladding during a power transient thus predicting its behaviour in term of stress corrosion cracking. Based upon the finite element computation code CASTEM 2000, TOUTATIS is a set of modules written in a macro language. The aim of this paper is to present both code modules: The axisymmetric bi-dimensional module, modeling a unique block pellet; The tri dimensional module modeling a radially fragmented pellet. Having shown the boundary conditions and the algorithms used, the application will be illustrated by: A short presentation of the bidimensional axisymmetric modeling performances as well as its limits; The enhancement due to the three dimensional modeling will be displayed by sensitivity studies to the geometry, in this case the pellet height/diameter ratio. Finally, we will show the easiness of the development inherent to the CASTEM 2000 system by depicting the process of a modeling enhancement by adding the possibility of an axial (horizontal) fissuration of the pellet. As conclusion, the future improvements planned for the code are depicted. (author)

Transient modeling of electrochemically assisted CO2 capture and release

DEFF Research Database (Denmark)

Singh, Shobhana; Stechel, Ellen B.; Buttry, Daniel A.

2017-01-01

to analyze the time-dependent behavior of CO2 capture and electro-migration transport across the cell length. Given high nonlinearity of the system, we used a finite element method (FEM) to numerically solve the coupled mass transport equations. The model describes the concentration profiles by taking......The present work aims to develop a model of a new electrochemical CO2 separation and release technology. We present a one-dimensional transient model of an electrochemical cell for point source CO2 capture and release, which mainly focuses on the simultaneous mass transport and complex chemical...... reactions associated with the separation process. For concreteness, we use an ionic liquid (IL) with 2 M thiolate anion (RS−) in 1 M disulfide (RSSR) as an electrolyte in the electrochemical cell to capture, transport and release CO2 under standard operating conditions. We computationally solved the model...

Assessment of the TASS 1-D neutronics model for the westinghouse and ABB-CE type PWR reactivity induced transients

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

Equivalent modeling of PMSG-based wind power plants considering LVRT capabilities: electromechanical transients in power systems.

Science.gov (United States)

Ding, Ming; Zhu, Qianlong

2016-01-01

Hardware protection and control action are two kinds of low voltage ride-through technical proposals widely used in a permanent magnet synchronous generator (PMSG). This paper proposes an innovative clustering concept for the equivalent modeling of a PMSG-based wind power plant (WPP), in which the impacts of both the chopper protection and the coordinated control of active and reactive powers are taken into account. First, the post-fault DC link voltage is selected as a concentrated expression of unit parameters, incoming wind and electrical distance to a fault point to reflect the transient characteristics of PMSGs. Next, we provide an effective method for calculating the post-fault DC link voltage based on the pre-fault wind energy and the terminal voltage dip. Third, PMSGs are divided into groups by analyzing the calculated DC link voltages without any clustering algorithm. Finally, PMSGs of the same group are equivalent as one rescaled PMSG to realize the transient equivalent modeling of the PMSG-based WPP. Using the DIgSILENT PowerFactory simulation platform, the efficiency and accuracy of the proposed equivalent model are tested against the traditional equivalent WPP and the detailed WPP. The simulation results show the proposed equivalent model can be used to analyze the offline electromechanical transients in power systems.

Transient multivariable sensor evaluation

Energy Technology Data Exchange (ETDEWEB)

Vilim, Richard B.; Heifetz, Alexander

2017-02-21

A method and system for performing transient multivariable sensor evaluation. The method and system includes a computer system for identifying a model form, providing training measurement data, generating a basis vector, monitoring system data from sensor, loading the system data in a non-transient memory, performing an estimation to provide desired data and comparing the system data to the desired data and outputting an alarm for a defective sensor.

A simplified transient three-dimensional model for estimating the thermal performance of the vapor chambers

International Nuclear Information System (INIS)

Chen, Y.-S.; Chien, K.-H.; Wang, C.-C.; Hung, T.-C.; Pei, B.-S.

2006-01-01

The vapor chambers (flat plate heat pipes) have been applied on the electronic cooling recently. To satisfy the quick-response requirement of the industries, a simplified transient three-dimensional linear model has been developed and tested in this study. In the proposed model, the vapor is assumed as a single interface between the evaporator and condenser wicks, and this assumption enables the vapor chamber to be analyzed by being split into small control volumes. Comparing with the previous available results, the calculated transient responses have shown good agreements with the existing results. For further validation of the proposed model, a water-cooling experiment was conducted. In addition to the vapor chamber, the heating block is also taken into account in the simulation. It is found that the inclusion of the capacitance of heating block shows a better agreement with the measurements

Transient thermal model of passenger car's cabin and implementation to saturation cycle with alternative working fluids

International Nuclear Information System (INIS)

Lee, Hoseong; Hwang, Yunho; Song, Ilguk; Jang, Kilsang

2015-01-01

A transient thermal model of a passenger car's cabin is developed to investigate the dynamic behavior of cabin thermal conditions. The model is developed based on a lumped-parameter model and solved using integral methods. Solar radiation, engine heat through the firewall, and engine heat to the air ducts are all considered. Using the thermal model, transient temperature profiles of the interior mass and cabin air are obtained. This model is used to investigate the transient behavior of the cabin under various operating conditions: the recirculation mode in the idling state, the fresh air mode in the idling state, the recirculation mode in the driving state, and fresh air mode in the driving state. The developed model is validated by comparing with experimental data and is within 5% of deviation. The validated model is then applied for evaluating the mobile air conditioning system's design. The study found that a saturation cycle concept (four-stage cycle with two-phase refrigerant injection) could improve the system efficiency by 23.9% and reduce the power consumption by 19.3%. Lastly, several alternative refrigerants are applied and their performance is discussed. When the saturation cycle concept is applied, R1234yf MAC (mobile air conditioning) shows the largest COP (coefficient of performance) improvement and power consumption reduction. - Highlights: • The transient thermal model of the passenger car cabin is developed. • The developed model is validated with experimental data and showed 5% deviation. • Saturation cycle concept is applied to the developed cabin model. • There is 24% COP improvement by applying the saturation cycle concept. • R1234yf showed the highest potential when it is applied to the saturation cycle.

RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part I: Theory and description of model capabilities

Science.gov (United States)

Raffray, A. René; Federici, Gianfranco

1997-04-01

RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under 'slow' high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case.

RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Pt. I. Theory and description of model capabilities

International Nuclear Information System (INIS)

Raffray, A.R.; Federici, G.

1997-01-01

For pt.II see ibid., p.101-30, 1997. RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under 'slow' high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case. (orig.)

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.)

Development of a transient criticality evaluation method

International Nuclear Information System (INIS)

Pain, C.C.; Eaton, M.D.; Miles, B.; Ziver, A.K.; Gomes, J.L.M.A.; Umpleby, A.P.; Piggott, M.D.; Goddard, A.J.H.; Oliveira, C.R.E. de

2005-01-01

In developing a transient criticality evaluation method we model, in full spatial/temporal detail, the neutron fluxes and consequent power and the evolving material properties - their flows, energies, phase changes etc. These methods are embodied in the generic method FETCH code which is based as far as possible on basic principles and is capable of use in exploring safety-related situations somewhat beyond the range of experiment. FETCH is a general geometry code capable of addressing a range of criticality issues in fissile materials. The code embodies both transient radiation transport and transient fluid dynamics. Work on powders, granular materials, porous media and solutions is reviewed. The capability for modelling transient criticality for chemical plant, waste matrices and advanced reactors is also outlined. (author)

Modeling of boron control during power transients in a pressurized water reactor

International Nuclear Information System (INIS)

Mathieu, P.; Distexhe, E.

1986-01-01

Accurate control instructions in a reactor control aid computer are included in order to realize the boron makeup throughput, which is required to obtain the boron concentration in the primary coolant loop, predicted by a neutronic code. A modeling of the transfer function between the makeup and the primary loop is proposed. The chemical and volumetric control system, the pressurizer, and the primary loop are modeled as instantaneous diffusion cells. The pipes are modeled as time lag lines. The model provides the unstationary boron distributions in the different elements of the setup. A numerical code is developed to calculate the time evolutions of the makeup throughput during power transients

Transient two-phase flow

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.)

Transient tire force and its modeling. Part 2. Camber angle input; Katoteki tire hasseiryoku to sono modeling. 2. Camber kaku nyuryoku

Energy Technology Data Exchange (ETDEWEB)

Higuchi, A [Toyota Motor Corp., Aichi (Japan); Pacejka, H

1997-10-01

The objective of this study is to understand the transient force and moment characteristics of tires involving side slip and camber and to develop the tire model which is capable of describing those characteristics. Some particular characteristics are found such as non-lagging part in side force response and an apparent peak in aligning torque response to a stepwise camber change. Some aspects on the analogy of turn slip to camber are also discussed. The tire model combines a dynamic part based on the physical aspect using the relaxation length concept, and a steady state part. In combined situations of side slip and camber, an estimation method to determine the transient slip quantities is introduced. 12 refs., 14 figs.

Improved lumped models for transient combined convective and radiative cooling of a two-layer spherical fuel element

International Nuclear Information System (INIS)

Silva, Alice Cunha da; Su, Jian

2013-01-01

The High Temperature Gas cooled Reactor (HTGR) is a fourth generation thermal nuclear reactor, graphite-moderated and helium cooled. The HTGRs have important characteristics making essential the study of these reactors, as well as its fuel element. Examples of these are: high thermal efficiency,low operating costs and construction, passive safety attributes that allow implication of the respective plants. The Pebble Bed Modular Reactor (PBMR) is a HTGR with spherical fuel elements that named the reactor. This fuel element is composed by a particulate region with spherical inclusions, the fuel UO2 particles, dispersed in a graphite matrix and a convective heat transfer by Helium happens on the outer surface of the fuel element. In this work, the transient heat conduction in a spherical fuel element of a pebble-bed high temperature reactor was studied in a transient situation of combined convective and radiative cooling. Improved lumped parameter model was developed for the transient heat conduction in the two-layer composite sphere subjected to combined convective and radiative cooling. The improved lumped model was obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of the two-layer spherical fuel element was analyzed to illustrate the applicability of the proposed lumped model, with respect to die rent values of the Biot number, the radiation-conduction parameter, the dimensionless thermal contact resistance, the dimensionless inner diameter and coating thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the improved lumped model, with H2,1/H1,1/H0,0 approximation yielded significant improvement of average temperature prediction over the classical lumped model. (author)