The High Temperature Test Facility (HTTF) is a one-quarter scale integral test facility model of the Modular High Temperature Gas Reactor. The facility is capable of operating at temperatures like those expected in a loss of forced convection cooling accident. The nominal working fluid is helium although other gases can be used. The facility is configured to simulate a variety of postulated depressurized conduction cooldown (DCC), pressurized conduction cooldown (PCC) and normal operations events.
Purpose
The purpose of the test facility is twofold: to validate the correlations that are currently uses to model relevant heat transfer and fluid flow phenomena required for the integrated simulation of the prismatic core phenomena; and generate results that can be used to validate different simulation methodologies applied in the gas cooled reactors modeling.
Facility Design
The facility is built in modular design to allow core reconfiguration from prismatic to pebble design.
In the current configuration, the graphite prismatic block structure in the MHTGR is simulated by ceramic blocks to capture prototypical core temperature profiles. The HTTF reactor core is built of 10 hexagonal core blocks that are made of a cast ceramic, Greencast 94-F. The reactor core is surrounded by reflectors on each side (2 upper reflectors, 3 bottom reflectors, and side reflectors). The side reflectors are made of a cast ceramic, ShotTech SiC 80 while the top and bottom reflectors are also made of Greencast 94-F. In addition, there are 3 separate structures designed to model the core exit chamber: lower plenum roof, lower plenum (the chamber that houses 163 support posts), and lower plenum floor.
The heating system is a network of electrically heated graphite rodlets/dog bones. The facility is equipped with the network of electrically heated graphite rods that produce approximately 2.2 MWth. There are 210 heater rods arranged in 10 heater banks, with 3 heater legs per heater bank. Each heater leg consists of 7 heater rods.
Detailed facility technical specification and instrumentation plan are available here:
Gutowska, I., Woods, B. G., & Cadell, S. R. (2019). CFD modeling of the OSU High
Temperature Test Facility inlet plenum flow distribution during normal operation. Nuclear
Engineering and Design, 353, 110216. https://doi.org/10.1016/j.nucengdes.2019.110216
Gradecka, M. J., & Woods, B. G. (2016). Development of thermal mixing enhancement
method for lower plenum of the High Temperature Test Facility. Nuclear Engineering
and Design, 305, 81–103. https://doi.org/10.1016/j.nucengdes.2016.02.043
Andre, M. A., Burns, R. A., Danehy, P. M., Cadell S. R., Woods B. G., and Bardet, P.
M., “Velocimetry during depressurized conduction cooldown events in the
HTTF," Nuclear Engineering and Design, 341, pp. 406-414, 2019.
Andre, M. A., Burns, R. A., Danehy, P. M., Cadell S. R., Woods B. G., and Bardet, P.
M., “Development of N2O-MTV for low-speed flow and in-situ deployment to an
integral effect test facility," Exp Fluids (2018) 59:14.
Conference Papers
Gutowska I., Halsted J., Schlamp N., Woods B. G, Manera A., Petrov V., Balestra P
Building Conduction Cooldown Scenarios Experimental Validation Database for HTGRs,
NURETH-19, Brussel, Belgium, March 2022
Gutowska I., Woods B. G, CFD Assessment of LOFA Intra Core Natural Circulation in
the High Temperature Test Facility, NURETH-18, Portland, Oregon, August 2019.
Gutowska I., Cadell S. R., Woods B. G., CFD Air Ingress Analyses on High Temperature
Test Facility, ANS Winter Meeting, Washington DC, US, October/November 2017).
Brumback, K. E., Woods, B. G., & Cadell, S. R. (2017). Overview of Shakedown Testing
for the High Temperature Test Facility. Transactions of the American Nuclear Society,
117, 1619.
Brumback, K. E., Cadell, S. R., and Woods, B. G., “Temperature and Flow
Characteristics of the High Temperature Test Facility during Depressurized Conduction
Cooldown Testing,” Proceedings, Advances in Thermal Hydraulics 2018, Orlando,
FL, November 2018.
Andre, M. A., Bardet, Cadell S.R., and Woods B.G., “Velocimetry/Thermometry
Data Fusion from DCC Tests in the HTTF,” Proceedings, International 18th International
Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH 18), Portland,
Oregon, August, 2019.
Andre, M. A., Bardet, P. M., Danehy, P. M., Burns, R. A., Cadell S.R., and Woods
B.G., “Non-Intrusive Velocity Measurements with MTV During DCC Event In the
HTTF,” Proceedings, International 17th International Topical Meeting on
Nuclear Reactor Thermal Hydraulics (NURETH 17), Xi’an, China, September, 2017.
OREGON STATE UNIVERSITY
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