Uniform Heated Scaled-Down Standard Fuel Block Test to Validate Core Thermofluid Analysis Code for Prismatic Gas-Cooled Reactor

Abstract

The Korea Atomic Energy Research Institute (KAERI) has developed the Core Reliable Optimization and thermofluid Network Analysis (CORONA) code for core thermofluid analysis of a prismatic high-temperature gas-cooled reactor (HTGR). KAERI performed scaled-down standard fuel block (SFB) heated tests at a helium experimental loop to validate the CORONA code. The scaled-down SFB was designed based on the core thermofluid design for a 350-MW(thermal) HTGR. The reference test condition was selected to maintain the Reynolds number of the coolant channels and the bypass gaps. The test section had seven coolant holes and 12 fuel holes considering KAERI’s helium loop circulator design. The material of the fuel block was Al₂O₃, selected to simulate the low thermal conductivity of the irradiated graphite at the high-temperature condition. The bypass gap structure was made of stainless steel 304 to minimize gap size deformation at the heated condition. This paper presents a comparison between the test results and the CORONA analysis results. The test parameter was the nitrogen flow velocity (3.6 to 6.0 kg/min) and constant heated condition.

Keywords:

• Prismatic gas-cooled reactor
• standard fuel block
• CORONA
• bypass flow fraction
• heated condition

Acknowledgments

This study was supported by the Nuclear Research and Development Program (2017M2A81014758, 2019M2D1A1058139) of the National Research Foundation of Korea grant funded by the Korean Government.

Nomenclature

$d$ =	=	hydraulic diameter of channel
$f$ =	=	friction factor
Nu =	=	Nusselt number
Pr =	=	Prandtl number
Re =	=	Reynolds number
$T$ =	=	temperature
$x$ =	=	local flow length

Greek

$\mathrm{\Delta }$ =	=	difference
$\eta$ =	=	multiplier for bypass gap friction factor
ρ =	=	density