Last edited by Mizilkree
Monday, July 20, 2020 | History

3 edition of Scaled thermal hydraulic simulation of open lattice PWR cores found in the catalog.

Scaled thermal hydraulic simulation of open lattice PWR cores

Columbia University. Heat Transfer Research Facility.

Scaled thermal hydraulic simulation of open lattice PWR cores

a scoping study

by Columbia University. Heat Transfer Research Facility.

  • 123 Want to read
  • 37 Currently reading

Published by EPRI in Palo Alto, Calif .
Written in English

    Subjects:
  • Pressurized water reactors -- Cores -- Simulation methods.,
  • Presurized water reators -- Thermodynamics -- Simulation methods.

  • Edition Notes

    Statementprepared by Heat Transfer Research Facility, Department of Chemical Engineering, Columbia University ; principal investigators, Sastry Sreepada ... [et al.]. ; project supervisor, Charles F. Bonilla ; prepared for Electric Power Research Institute.
    SeriesFinal report - Electric Power Research Institute
    ContributionsSreepada, Sastry., Bonilla, Charles F. 1909-1987., Electric Power Research Institute.
    Classifications
    LC ClassificationsTK9203.P7 C64 1977
    The Physical Object
    Paginationca. 150 p. in various pagings :
    Number of Pages150
    ID Numbers
    Open LibraryOL4232548M
    LC Control Number80512999

    Then, we run the lattice Boltzmann simulation based on the 3D digital core of shale to obtain the real and whole CO 2 –CH 4 displacement process. 5 Conclusions. CO 2-formation fluid displacement at the pore scale is a key issue in CO 2 sequestration and . Presenter’s data Peter Fritzson is Professor and research director of the Programming Environment Laboratory, at Linköping University. He is also director of the Open Source Modelica Consortium, director of the MODPROD center for model-based product development, and vice chairman of the Modelica Association.

    Abstract In order to perform safety oriented transient analyses, the entire nuclear power plant has to be simulated in most cases. However, a high level of resolution is only needed where the safety parameters may reach the safety criteria. In this paper, three scale resolutions are defined (NPP level, core level, fuel pin level). Thermal-hydraulic and neutron kinetic codes are already Author: Y. Périn, M. Klein, I. Pasichnyk, K. Velkov, A. Seubert, A. Pautz. From the present results, the high prospect was acquired of the possibility of establishment of a new thermal design method for the advanced light-water reactor cores by the large-scale simulation only. Keywords: fluid dynamics, large-scale simulation, two-phase flow, fuel assembly, nuclear reactors, thermal design, three-dimensional : K. Takase, H. Yoshida, Y. Ose, H. Akimoto.

    R. Vaghetto, D. De Luca, Y. A. Hassan Analysis of Thermal-Hydraulic System Response of a Pressurized Water Reactor during Hypothetical Core Blockage Scenarios The 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH), Convention Hall, Pisa, Italy – May The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of kW; average linear heat generation rate of W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is C.


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Scaled thermal hydraulic simulation of open lattice PWR cores by Columbia University. Heat Transfer Research Facility. Download PDF EPUB FB2

Heat Transfer Research Facility. (1) Scaled thermal hydraulic simulation of open lattice PWR cores: a scoping study / prepared by Heat Transfer Research Facility, Department of Chemical Engineering, Columbia University ; principal investigators, Sastry Sreepada. The LSTF is a volumetrically scaled, full-height, full-pressure simulator of a Westinghouse-type four-loop ( MW thermal power) pressurized water reactor (PWR).

An immersed body method for coupled neutron transport and thermal hydraulic simulations of PWR assemblies. however provide a more detailed description of the flow and heat transfer within the nuclear core by modelling the core as a collection of 1D thermal hydraulic channels.

Typically, each channel could represent a single subchannel and Cited by: 2. Pin-by-pin thermal-hydraulic calculation model. The multiple channel method is employed in pin-by-pin thermal–hydraulic calculation to simulate the coolant flow in the bundle.

The coolant flow area is defined as the area between fuel rods or guide tubes. Each channel can be divided into a number of control bodies in axial direction. Flow and heat transfer Single phase flow Experiments. A tight lattice reactor core with a narrow flow channel has a hydraulic diameter less than half of the regular reactor core, which might result in considerable adverse local flow redistributions within a typical coolant channel, particularly under boiling conditions ().Moreover, substantial increase in channel Cited by: 2.

Coupled neutronic thermo-hydraulic analysis of full PWR core with Monte-Carlo based BGCore system Article in Nuclear Engineering and Design (9).

APPLICATION OF THE COUPLED THREE DIMENSIONAL THERMAL-HYDRAULICS AND NEUTRON KINETICS MODELS TO PWR STEAM based on single assembly lattice calculations. FUEL THERMAL MODEL The SMART fuel pin model solves the one-dimensional, time-dependent, radial heat conduction core power level and the DNB risk is.

A detailed analytical investigation of the feasibility of simulating large open lattice pressurized water reactor cores using to electrically heated pins was conducted.

Figure 1 shows the 3x3 lattice of infinite fuel pin cells used in the simulations. One fuel rod containing the burnable poison (UO 2 -Gd 2O 3) is located at the centre of the lattice and it is surrounded by seven rods containing standard fuel (UO 2) Cited by: 2.

A summary is described about nuclear power reactors analyses and simulations in the last decades with emphasis in recent developments for full 3D reactor core simulations using highly advanced computing techniques.

The development of the computer code AZKIND is presented as a practical exercise. AZKIND is based on multi-group time dependent neutron diffusion : Andrés Rodríguez Hernández, Armando Miguel Gómez-Torres, Edmundo del Valle-Gallegos. In contrast to all COBRA-versions, it is based on SI system of units.

SUBCHANFLOW contains water, lead, helium, lead-bismuth and sodium as working fluid. Using thermal-hydraulic modeling based on 3 equations approach, SUBCHANFLOW can simulate a problem with sub-channels within two by: The thermal–hydraulic (TH) status of PWR core has strong 3D effect on flow velocity (Conner et al., ), swirling status (Bhattacharjee et al., ), turbulent intensity (Chang et al., ), turbulent anisotropy (Nguyen et al., ), cross flow status (Qu et al., a), boron mixing (Yu et al., ), turbulent mixing (Ju et al., Author: Guangliang Chen, Jijun Wang, Zhijian Zhang, Zhaofei Tian, Lei Li, Huilun Kang, Yuguan Jin.

Modelling of Nuclear Reactor Multi-physics 1st Edition The book addresses the frontier discipline of neutronic/thermal-hydraulic modelling of nuclear reactor cores, presenting the main techniques in a generic manner and for practical reactor calculations.

Engineers and scientists from the nuclear industry who use simulation tools for. emergency core coolingsystems,was designed to model the thermal-hydraulic behaviour ofVVERtype reactors.

The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP PACTEL has been upgraded and modified on a case-by-case basis. In order to describe the thermal-hydraulic processes undergoing at the various spatial scales, many thermal-hydraulic simulation codes were developed and.

Abstract. This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVERtype by: 9.

With SimScale, you can test multiple design versions in parallel and quickly identify the best-performing one. Even in the case of large or complex designs, access to up to 96 cores and real-time simulation allows you to get your results faster than ever before.

In parallel to these safety analysis efforts, the following thermal-hydraulic tests are to be performed: (1) thermal-hydraulic test of a horizontal steam generator; (2) integrated thermal-hydraulic test using a simulation loop for the innovative MS-PWR (SLIM).}, doi = {}, journal = {Transactions of the American Nuclear Society; (United States)}, issn = {X}, number.

This paper describes a multi-scale thermal-hydraulic coupling system, which combines the capabilities of the open-source TrioCFD code and of the. Targets of multiphysics approach • Investigate the coupling between neutronic code (at lattice level and full core model) and a thermal-hydraulic CFD code • Extended use of Hyerarchical Data Format structures (HDF5) • Implementation of multiphysics approach into preliminary model of a Fast Small Modular Reactor (SMR) • Choosing as case study a lead cooled reactor.

The MIT- PCCL simulates a single PWR flow cell: a channel between fuel pins in a PWR core and one steam generator tube. This modeling results in a MIT-PCCL to PWR flow cell scale of approximately 1/3 even though the power scale-down ratio is about 1/, * Make each experimental, run with a "clean" well- characterized system.Digital core simulation technology, as an emerging numerical simulation method, has gradually come to play a significant role in the study of petrophysical properties.

By using this numerical simulation method, the influence of micro factors on seepage properties of reservoir rock is taken into consideration, making up the shortcomings of the traditional physical by: 6.The large-scale test facility (LSTF) of the Rig of Safety Assessment No.

4 (ROSA-IV) program is a volumetrically scaled (1/48) pressurized water reactor (PWR) system with an electrically heated core used for integral simulation of small break loss-of-coolant accidents (LOCAs) and operational transients.