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3 edition of Simulation of in-reactor experiments with the ELOCA.Mk5 code found in the catalog.

Simulation of in-reactor experiments with the ELOCA.Mk5 code

M. E. Klein

Simulation of in-reactor experiments with the ELOCA.Mk5 code

by M. E. Klein

  • 337 Want to read
  • 33 Currently reading

Published by Fuel Engineering Branch, Chalk River Laboratories in Chalk River, Ont .
Written in English

    Subjects:
  • ELOCA.Mk5 (Computer program),
  • Nuclear fuel elements -- Computer programs.,
  • Nuclear fuel elements -- Thermal properties.,
  • Thermal stresses -- Computer programs.

  • Edition Notes

    Other titlesSimulation d"essais en réacteur avec le programme de calcul ELOCA.Mk5.
    Statementby M.E. Klein, L.N. Carlucci and V.I. Arimescu.
    SeriesAECL -- 11133, AECL (Series) -- 11133.
    ContributionsArimescu, V. I., Carlucci, L. N., AECL Research., Chalk River Laboratories. Fuel Engineering Branch.
    Classifications
    LC ClassificationsTK9207 .K55 1994
    The Physical Object
    Pagination15 p. :
    Number of Pages15
    ID Numbers
    Open LibraryOL20657609M
    ISBN 100660157888
    OCLC/WorldCa32544793

    16 Fast Reactors: Training Courses and Workshops • IAEA Workshops and Schools on Innovative Nuclear Energy Systems • Recent Course: Joint ICTP-IAEA Workshop on the Physics and Technology of Innovative Nuclear Energy Systems for Sustainable Development, 29 Aug - 02 Sept , Trieste, Italy. A list of the nuclear reactors designed by Argonne National Laboratory for basic and applied science research. Argonne designed, built, and operated several reactors that collectively produced a vast amount of data and analysis that drove advancements in radiation protection, biosciences, reactor design, and evolution of materials used in reactor construction.

    Abstract Plate fuel assembly is widely used in reactor cores due to benefits of the narrow rectangular channels. Compared with rod assembly, the heat transfer area and power density can be increased in narrow rectangular channel, where little researches on post-CHF thermal-hydraulic characteristics have been done. Once the loss of coolant accident happens, post . The US Department of Energy's Argonne National Laboratory is taking its nuclear energy research into new territory -- virtual territory that is. With the .

      where y is the solution vector that is dependent on both space and time, respectively, and, where F is the nonlinear operator representing the coupled system and n is the total number of unknowns. For ease of comprehension, we can write F as in the second equality of equation (), where N is also a nonlinear operator and b is the load vector. It helps to . A new computer algorithm developed by researchers at the US Department of Energy's Argonne National Laboratory allows scientists to view nuclear fission in much finer detail than ever before.


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Simulation of in-reactor experiments with the ELOCA.Mk5 code by M. E. Klein Download PDF EPUB FB2

ISBN: OCLC Number: Language Note: Abstracts in English and French. Notes: "AECL Research." " August." "Based on a paper presented at the CNA/CNS Annual Conference Montréal, Québec, June ". just this purpose, and BPGs remain a central pillar of the simulation material accepted at this current workshop, as it was at its predecessors.

In order to assess the maturity of CFD codes for use in reactor safety and design, it is necessary to establish a database of CFD-grade experimental material; this remains. Giese H, Caravec C, Granget G, Palmiotti G () The RACINE-1E critical experiments for control rod method and data calculation.

Experimental and calculation results. In: Proceedings of the topical management advances in reactor physics and safety, Saratoga Springs, 17–19 Sept Google Scholar. csni ACHILLES, Heat Transfer in PWR Core During LOCA Reflood Phase: nea ANL-BPB, Argonne National Laboratory Code Center Benchmark Problem Book: csni BETHSY/C, Loss of residual heat removal system during mid-loop operation: csni BETHSY/B, Cold Leg Break Test: csni BIP PROJECT, Behaviour of Iodine Project:.

A series of three LMFBR unprotected loss of flow (LOF) accidents has been simulated in the Sodium Loop Safety Facility (SLSF). The results of these in-reactor experiments verify that the dynamics of sodium voiding are, in general, well represented by current single channel slug expulsion models.

Including the warm‐up phase, the entire experiment contained h of operation (simulation time was s for all runs).

From the TE simulator, the process operating cost ($/h) can be extracted, and we have the operating cost for every 12 min. Figure 5 illustrates the impact of the experimental factors on the process operating cost during.

Silde, A, Hostikka, S, Kankkunen, A, Hyvärinen, J & Hakola, IExperimental and Numerical Studies of Liquid Dispersal from a Soft Projectile Impacting a Wall. in Proceedings: 19th International Conference on Structural Mechanics in Reactor TechnologySMiRT vol.

3, Transactions of SMiRT, vol. 19, pp.19th International. The review specializes to the modelling of plasmas in a particular type of fusion experiment, namely the tokamak. Simulation is taken to imply the use of a model which involves variation in at least two coordinate directions and is nonlinear.

The purpose of this guide is to describe in complete detail a FORTRAN code named Program SCAT 4 written by the UCLA group in order to analyze elastic scattering of various particle.

Topics covered includes: Mathematical Description, Program Description, Description of Input Data, Symbolic Listing of the Program. In this work, the VENUS-3 benchmark has been analyzed using SuperMC code, with the intention of validating SuperMC for accurate reactor neutronics; dosimetry response calculations for in-core/ex-core structural components, particularly with respect to the VENUS-3 configuration type pressurized water reactors (PWRs).

The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design.

Effects in Reactor Pressure Vessel Steels No. NP-T uides IAEA Nuclear Energy Series No. NP-T Integrity of Reactor Pressure Vessels in Nuclear Power Plants: Assessment of Irradiation Embrittlement Effects in Reactor Pressure Vessel Steels INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA ISBN –92–0––3 ISSN –   To address such failures, efforts are underway to define the mechanisms responsible for the in-reactor pillowing, and suggest improvements to the fuel plate design and operational envelope.

For this purpose, selected plates from previous experiments were simulated to understand the thermo-mechanical response of the plates to the fission density. Integral codes (also called “system” codes or, in the past “engineering-level” codes): these codes simulate the overall NPP response, i.e.

the response of the RCS, the containment, and the source term to the environment, using "integrated" models for a self-consistent thorough analysis of the accident. They include a well-balanced. Preliminary simulation of the more extreme tests using the computer code upon which CAGR fault studies are based has demonstrated agreement to within 5 - 10% in both power and temperature transients.

This agreement, for situations much more closely simulating faults than has previously been possible, greatly increases confidence in the CAGR. The bubble shape change and the bubble rise velocity were compared with the newly performed experiments, which used solid particulate glasses of mm in diameter, liquid silicone and air.

The two-phase flow simulation of a single bubble rising in a stagnant liquid pool reproduced measured bubble shape and bubble rise velocity reasonably. Sean Cabaniss, David Park, Maxim Slivinsky, and Julianne Wagoner [Winter ] Neil Dalvie, KC Anderson, Natalia Majewska [Winter ] The center of any chemical process is the reactor, where chemical reactions are carried out to transform feeds into products.

Reactor design is a vital step in the overall design of a process. Experiments with Couette autoclaves for the investigation of activity uptake in the oxide layer stainless steel under boiling water reactor conditions An Assessment of the blanket coolant chemistries in the SEAFP-M2 fusion device by the CORA-NNC code to reduce operational radiation exposure Evaluation of Zircaloy Corrosion Under Various.

ND simulation •If you have n atoms of the radio nuclide in a sample or container, how many gamma-rays emitted. •Reverse of experiment and analysis.

•Want the number of emitted gamma-rays from a specific number of atoms E ± ΔE, simply E =n I •But if calculate uncertainty based upon Δn, Δ and. Continuous-FlowReactors Continuous-StirredTankReactor(CSTR) CSTRsareoperatedatsteadystate(accumulation=0)andareassumedtobeperfectlymixed. Professor Dinh research in nuclear reactor thermal hydraulics contributed to severe accident management in LWRs and ALWS plants.

Professor Dinh's current research is focused on data-driven modeling and validation of advanced simulation codes. He is a Fellow of the American Nuclear Society and a recipient of the ANS-THD Technical Achievement Award.The Online Books Page.

Online Books by. General Electric Company. A Wikipedia article about this author is available. General Electric Company: How Electronic Tubes Work (c) (PDF at ) General Electric Company: Television Station Planning () (PDF at ) Help with reading books-- Report a bad link-- .Turinsky, P., CASL: The Consortium for Advanced Simulation of Light Water Reactors: A DOE Energy Innovation Hub for Modeling and Simulation of Nuclear Reactors, Korean Atomic Energy Research Institute, OctoDaejeon, South Korea,