A review of recent advances in HTGR CFD and thermal fluid analysis

AJ Huning, S Chandrasekaran, S Garimella - Nuclear Engineering and …, 2021 - Elsevier
Abstract The High Temperature Gas-cooled Reactor (HTGR) is an advanced reactor design
being pursued by several different domestic and international organizations due to its high …

Development of core support barrel installation tool for a bypass flow reduction in nuclear reactor core

OC Ondwasi, N Ihn - Nuclear Engineering and Design, 2024 - Elsevier
Reactor internals are installed in the reactor vessel and are subsequently removed during
periodic inspections. To facilitate this process, a gap is maintained between the outlet of the …

Experimental and CFD studies of the bypass flow in a prismatic core of VHTR using a small-scale model

W Kanjanakijkasem, H Wang… - Progress in Nuclear …, 2016 - Elsevier
The bypass flow in a prismatic very high temperature reactor (VHTR) core is an important
parameter in reactor design, which has not been well assessed in publications to date. To …

[PDF][PDF] 温度对狭窄缝隙流动阻力系数影响的试验研究

孟洋, 樊睿辰, 眭曦, 李勇, 郑健涛, 张嘉琪, 王杰… - 核动力 …, 2021 - hdlgc.xml-journal.net
压水反应堆内冷却剂旁漏流流道尺寸多是狭窄缝隙, 狭窄缝隙尺寸很敏感, 很容易受到系统压力,
压差, 温度, 振动等因素的影响. 微小的尺寸变化又会引起阻力系数明显改变 …

Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200

JH Lee, HK Cho, GC Park - Nuclear Engineering and Design, 2016 - Elsevier
The core of the very high temperature reactor (VHTR) PMR200 (a prismatic modular reactor
rated at 200 MW of thermal power) consists of hexagonal prismatic fuel blocks and reflector …

Assessment of Modeling Capabilities and Transient Performance for High-Temperature Reactors

RF Kile - 2023 - trace.tennessee.edu
Modeling and simulation tools play an important role in the design and licensing of reactors,
but models alone are insufficient for licensing reactors. As advanced reactors approach …

Uniform heated scaled-down standard fuel block test to validate core thermofluid analysis code for prismatic gas-cooled reactor

CS Kim, BH Park, ES Kim, MH Kim - Nuclear Technology, 2020 - Taylor & Francis
Abstract The Korea Atomic Energy Research Institute (KAERI) has developed the Core
Reliable Optimization and thermofluid Network Analysis (CORONA) code for core …

Verification and Validation of the Fuel Assembly Bypass Flow Analysis Code PEONY V1. 0

C Chen, J Yan, Y Xi, Y Zhang… - International …, 2022 - asmedigitalcollection.asme.org
Fuel assembly bypass flow is the coolant flow through the guide thimble and the instrument.
The fuel assembly bypass flow should be limited to ensure that there is sufficient coolant in …

Benchmark Validation Experiment of Plenum-to-Plenum Flow Through Heated Parallel Channels

AW Parker, BL Smith - Journal of Verification …, 2022 - asmedigitalcollection.asme.org
This paper documents a computational fluid dynamics (CFD) validation benchmark
experiment for flow through three parallel, heated channels from one plenum to another. The …

[PDF][PDF] Design of Scale-down Standard Fuel Block Test Section to Validate Core Thermo-fluid Analysis Code for Prismatic Gas-Cooled Reactor

CS Kim, BH Park, NI Tak, ES Kim - … of the Korean Nuclear Society Autumn …, 2018 - kns.org
Korea Atomic Energy Research Institute (KAERI) has developed CORONA (Core Reliable
Optimization and thermo-fluid Network Analysis) code for thermofluid analysis of prismatic …