• Title/Summary/Keyword: MATRA-LMR-FB

Search Result 5, Processing Time 0.025 seconds

DEVELOPMENT OF THE MATRA-LMR-FB FOR FLOW BLOCKAGE ANALYSIS IN A LMR

  • Ha, Kwi-Seok;Jeong, Hae-Yong;Chang, Won-Pyo;Kwon, Young-Min;Cho, Chung-Ho;Lee, Yong-Bum
    • Nuclear Engineering and Technology
    • /
    • v.41 no.6
    • /
    • pp.797-806
    • /
    • 2009

  • The Multichannel Analyzer for Transient and steady-state in Rod Array - Liquid Metal Reactor for Flow Blockage analysis (MATRA-LMR-FB) code for the analysis of a subchannel blockage has been developed and evaluated through several experiments. The current version of the code is improved here by the implementation of a distributed resistance model which accurately considers the effect of flow resistance on wire spacers, by the addition of a turbulent mixing model, and by the application of a hybrid scheme for low flow regions. Validation calculations for the MATRA-LMR-FB code were performed for Oak Ridge National Laboratory (ORNL) 19-pin tests with wire spacers and Karlsruhe 169-pin tests with grid spacers. The analysis of the ORNL 19-pin tests conducted using the code reveals that the code has sufficient predictive accuracy, within a range of 5 $^{\circ}C$, for the experimental data with a blockage. As for the results of the analyses, the standard deviation for the Karlsruhe 169-pin tests, 0.316, was larger than the standard deviation for the ORNL 19-pin tests, 0.047.

  • https://doi.org/10.5516/NET.2009.41.6.797 인용 PDF KSCI

A Preliminary Safety Analysis for the Prototype Gen IV Sodium-Cooled Fast Reactor

  • Lee, Kwi Lim;Ha, Kwi-Seok;Jeong, Jae-Ho;Choi, Chi-Woong;Jeong, Taekyeong;Ahn, Sang June;Lee, Seung Won;Chang, Won-Pyo;Kang, Seok Hun;Yoo, Jaewoon
    • Nuclear Engineering and Technology
    • /
    • v.48 no.5
    • /
    • pp.1071-1082
    • /
    • 2016

  • Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  • https://doi.org/10.1016/j.net.2016.08.002 인용 PDF KSCI