Results 21 to 30 of about 308 (180)
CHF prediction in rod bundles using round tube data
Brazilian Journal of Radiation Sciences, 2021 The present work concerns the use of 1995 CHF table for uniformly heated round tubes, developed jointly by Canadian and Russian researchers, for the prediction of critical heat flux in rod bundles geometries.
Maria Auxiliadora Fortini Veloso +1 more
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Nuclear Engineering and Technology, 2023 MARS-KS, a domestic regulatory confirmatory code of Republic of Korea, had been developed by integrating RELAP5/MOD2 and COBRA-TF. The integration of COBRA-TF allowed to extend the capability of MARS-KS, limited to one-dimensional analysis, to multi ...
Yunseok Lee, Taewan Kim
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SERPENT/SUBCHANFLOW COUPLED BURNUP CALCULATIONS FOR VVER FUEL ASSEMBLIES [PDF]
EPJ Web of Conferences, 2021 The continuous improvement in nuclear industry safety standards and reactor designers’ and operators’ commercial goals represent a push for the development of highly accurate methodologies in reactor physics.
Ferraro Diego +6 more
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Mechanism Model of Critical Heat Flux in Bundle Based on High-precision Subchannel Code
Yuanzineng kexue jishu, 2023 Critical heat flux (CHF) is a crucial safety limit for PWR core design. Accurately predicting CHF in the core rod bundle channel can enhance the economy and safety of the reactor. Currently, the prediction of rod bundle CHF in engineering is mainly based
GUO Junliang;JIANG Li;GAO Kui;KONG Huanjun;GUI Miao;SHAN Jianqiang
doaj
MC21/CTF and VERA multiphysics solutions to VERA core physics benchmark progression problems 6 and 7
Nuclear Engineering and Technology, 2017 The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of Consortium for Advanced Simulation of Light Water Reactors (CASL) tools and in-house Python scripts.
Daniel J. Kelly, III +5 more
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Nuclear Engineering and Technology, 2019 The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high ...
Jaejin Lee +2 more
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Nuclear Engineering and Technology, 2021 This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for
Manuel García +10 more
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ANALYSIS OF FUEL LONGTERM EFFECTS USING SUBCHANNEL CODES
Journal of Engineering, 2010 Thermohydraulic computer codes such as COBRA3C are now becoming in practical use as part of power plant licensing process. One of the areas of interest to reactor designer is the effect due to longterm operations and the changes that might occur either due to irradiation (burnup) effects or thermal stresses in the reactor.
E.A. AMIN, M.Y. ABDUL-AHAD
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Assessment of the Subchannel Code CTF for Single- and Two-Phase Flows [PDF]
Nuclear Technology, 2018 As part of the Consortium for Advanced Simulation of Light Water Reactors, the subchannel code CTF is being used for single-phase and two-phase flow analysis under light water reactor operating con...
Xingang Zhao +3 more
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Accuracy and Uncertainty Analysis of PSBT Benchmark Exercises Using a Subchannel Code MATRA
Science and Technology of Nuclear Installations, 2012 In the framework of the OECD/NRC PSBT benchmark, the subchannel grade void distribution data and DNB data were assessed by a subchannel code, MATRA. The prediction accuracy and uncertainty of the zone-averaged void fraction at the central region of the 5
Dae-Hyun Hwang +3 more
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