The Experts below are selected from a list of 15 Experts worldwide ranked by ideXlab platform

Joo Hwan Park - One of the best experts on this subject based on the ideXlab platform.

  • Development of a CFD model for the CANDU-6 Moderator analysis using a coupled solver
    Annals of Nuclear Energy, 2008
    Co-Authors: Churl Yoon, Joo Hwan Park
    Abstract:

    Abstract A computational fluid dynamics (CFD) Moderator analysis model by using a coupled solver has been developed for the Moderator analysis of Canada deuterium uranium (CANDU) reactors. For Wolsong Units 2/3/4, a steady-state Moderator circulation under operating conditions and the Local Moderator subcooling during a LOCA transient were evaluated using the CFD tool. When compared to a former study in the Final Safety Analysis Reports, the current analysis provided well-matched trends and reasonable results. This new CFD model based on a coupled solver shows a dramatic increase in the computing speed, when compared to that based on a segregated solver.

Junto Ogawa - One of the best experts on this subject based on the ideXlab platform.

  • Application of 3D Coupled Code SPARKLE-2 to PWR Anticipated Transient Without Scram
    Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle Radioactive Waste Management and Decommissioning; Computational Fluid D, 2014
    Co-Authors: Koji Asano, Hikaru Sakamoto, Satoshi Imura, Junto Ogawa
    Abstract:

    An anticipated transient without scram (ATWS) is an anticipated operational occurrence (AOO) followed by failure of the automatic reactor trip function of the reactor protection system. The failure of the reactor to shut down during the certain AOOs can lead to increase in reactor coolant system (RCS) pressure and decrease in departure from nucleate boiling ratio (DNBR) margin for a pressurized water reactor (PWR). Japanese standard PWRs are equipped with ATWS mitigation system which consists of a diverse mitigation system which is independent from the reactor trip system. The ATWS mitigation system automatically initiates isolation of the main steam line flow and the auxiliary feed water system under condition indicative of an ATWS.Mitsubishi Heavy Industries, Ltd. (MHI) applies 3D coupled code, SPARKLE-2 [1] [2], to the ATWS evaluation. SPARKLE-2 is a 3D coupled code developed by MHI and consist of the PWR system transient analysis code M-RELAP5, the 3D neutron kinetics code COSMO-K [3] and the 3D core thermal-hydraulics code MIDAC [4]. SPARKLE-2 implements the 3D characteristics such as Local Moderator feedback and change in 3D power distribution during transient.Thanks to gain from the 3D calculation, the analysis results show that the plant transients are effectively mitigated by the ATWS mitigation system and the RCS pressure and the minimum DNBR meet the safety criteria. These results also show that operational margins are increased, which enables more flexible design of the reload core.Copyright © 2014 by ASME

Y. Bilodid - One of the best experts on this subject based on the ideXlab platform.

  • Neutron noise observations in German KWU built PWRs and analyses with the reactor dynamics code DYN3D
    Annals of Nuclear Energy, 2018
    Co-Authors: Ulrich Rohde, Marcus Seidl, Sören Kliem, Y. Bilodid
    Abstract:

    Abstract Low-frequency neutronic noise with magnitudes of up to ten percent of the reactor power have been observed in Konvoi-type PWR in Germany in the last years. Several attempts were made to identify the reasons for increased magnitudes of the neutronic fluctuations in comparison with pre-Konvoi reactors, and various hypotheses have been created to explain this effect. In this paper, results of noise simulations performed with use of the reactor dynamics code DYN3D are presented. Fluctuations of the coolant inlet temperature and mass flow rate were considered. Besides of un-correlated fluctuations, correlated temperature fluctuations were simulated. In a preliminary analysis with harmonic temperature oscillations, the importance of existing correlations between temperature fluctuations in the individual fuel assemblies was pointed out. In our analyses, the correlations between fluctuations in the individual fuel assemblies were obtained based on an experimentally validated coolant mixing model. However, the features of the neutronic noise found in the simulations do not correspond to the measurements. Obviously, more complex mechanisms than only temperature and/or mass flow fluctuations have to be considered. Simulated generic fluctuations of the Local Moderator density, that were introduced independently from thermal hydraulics, indicate that potentially deformations or vibrations of the fuel rod lattice or turbulent fluctuations of the coolant temperature and density inside the fuel rod bundles leading to variations of the Local Moderator content might be responsible for the observed neutronic noise. Therefore, advanced models coupling neutronics, thermal hydraulics, turbulence and mechanical modelling have to be developed.

Churl Yoon - One of the best experts on this subject based on the ideXlab platform.

  • Development of a CFD model for the CANDU-6 Moderator analysis using a coupled solver
    Annals of Nuclear Energy, 2008
    Co-Authors: Churl Yoon, Joo Hwan Park
    Abstract:

    Abstract A computational fluid dynamics (CFD) Moderator analysis model by using a coupled solver has been developed for the Moderator analysis of Canada deuterium uranium (CANDU) reactors. For Wolsong Units 2/3/4, a steady-state Moderator circulation under operating conditions and the Local Moderator subcooling during a LOCA transient were evaluated using the CFD tool. When compared to a former study in the Final Safety Analysis Reports, the current analysis provided well-matched trends and reasonable results. This new CFD model based on a coupled solver shows a dramatic increase in the computing speed, when compared to that based on a segregated solver.

Koji Asano - One of the best experts on this subject based on the ideXlab platform.

  • Application of 3D Coupled Code SPARKLE-2 to PWR Anticipated Transient Without Scram
    Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle Radioactive Waste Management and Decommissioning; Computational Fluid D, 2014
    Co-Authors: Koji Asano, Hikaru Sakamoto, Satoshi Imura, Junto Ogawa
    Abstract:

    An anticipated transient without scram (ATWS) is an anticipated operational occurrence (AOO) followed by failure of the automatic reactor trip function of the reactor protection system. The failure of the reactor to shut down during the certain AOOs can lead to increase in reactor coolant system (RCS) pressure and decrease in departure from nucleate boiling ratio (DNBR) margin for a pressurized water reactor (PWR). Japanese standard PWRs are equipped with ATWS mitigation system which consists of a diverse mitigation system which is independent from the reactor trip system. The ATWS mitigation system automatically initiates isolation of the main steam line flow and the auxiliary feed water system under condition indicative of an ATWS.Mitsubishi Heavy Industries, Ltd. (MHI) applies 3D coupled code, SPARKLE-2 [1] [2], to the ATWS evaluation. SPARKLE-2 is a 3D coupled code developed by MHI and consist of the PWR system transient analysis code M-RELAP5, the 3D neutron kinetics code COSMO-K [3] and the 3D core thermal-hydraulics code MIDAC [4]. SPARKLE-2 implements the 3D characteristics such as Local Moderator feedback and change in 3D power distribution during transient.Thanks to gain from the 3D calculation, the analysis results show that the plant transients are effectively mitigated by the ATWS mitigation system and the RCS pressure and the minimum DNBR meet the safety criteria. These results also show that operational margins are increased, which enables more flexible design of the reload core.Copyright © 2014 by ASME