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Title: DRACS thermal performance evaluation for FHR

Journal Article · · Annals of Nuclear Energy (Oxford)
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  1. The Ohio State University, Columbus, OH (United States)
  2. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
  3. Idaho National Laboratory (INL), Idaho Falls, ID (United States)
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Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS Heat Exchanger and the Natural Draft Heat Exchanger. In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, and to keep the DRACS ready for activation, if needed, during accidents. To help with the design and thermal performance evaluation of the DRACS, a computer code using MATLAB has been developed. This code is based on a one-dimensional formulation and its principle is to solve the energy balance and integral momentum equations. By discretizing the DRACS system in the axial direction, a bulk mean temperature is assumed for each mesh cell. The temperatures of all the cells, as well as the mass flow rates in the DRACS loops, are predicted by solving the governing equations that are obtained by integrating the energy conservation equation over each cell and integrating the momentum conservation equation over each of the DRACS loops. In addition, an intermediate heat transfer loop equipped with a pump has also been modeled in the code. This enables the study of flow reversal phenomenon in the DRACS primary loop, associated with the pump trip process. Experimental data from a High-Temperature DRACS Test Facility (HTDF) are not available yet to benchmark the code. A preliminary code validation is performed by using natural circulation experimental data available in the literature, which are as closely relevant as possible. The code is subsequently applied to the HTDF that is under construction at the Ohio State University.

Research Organization:
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE), Nuclear Energy University Program (NEUP); USDOE
Grant/Contract Number:
AC05-00OR22725; AC07-05ID14517
OSTI ID:
1429226
Alternate ID(s):
OSTI ID: 1247583
Journal Information:
Annals of Nuclear Energy (Oxford), Vol. 77, Issue C; ISSN 0306-4549
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English
Citation Metrics:
Cited by: 13 works
Citation information provided by
Web of Science

References (13)

The advanced high-temperature reactor: High-temperature fuel, liquid salt coolant, liquid-metal-reactor plant journal January 2005
Molten-Salt-Cooled Advanced High-Temperature Reactor for Production of Hydrogen and Electricity journal December 2003
Heat transfer to water from a vertical tube bundle under natural-circulation conditions. [PWR; BWR] report January 1983
Heat transfer from a vertical tube bundle under natural circulation conditions journal December 1985
Dynamics of a Natural Circulation Loop: Analysis and Experiments journal January 1986
An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components report November 2009
Laminar flow in annuli ducts with constant wall temperature journal February 1998
Design of Fluidic Diode for a High-Temperature DRACS Test Facility conference July 2013
Laminar convection in wall sub-channels and transport rates for finite rod-bundle assemblies by superposition journal April 1986
Finite Element Analysis for Laminar flow and heat Transfer in Finite rod Bundles journal March 1987
Turbulent transport in wall subchannels and in finite rod clusters journal June 1991
Assessment of Candidate Molten Salt Coolants for the NGNP/NHI Heat-Transfer Loop report June 2006
Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR) report March 2006

Figures / Tables (24)

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Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
Reactor Cooling Systems
Molten Salt Cooled Reactors
Simulation
NATURAL CONVECTION
HEAT LOSSES
EXPERIMENTAL DATA
HEAT EXCHANGERS
FLUORIDES
GRAPHITE
PUMPS
VALIDATION
COMPUTER CODES
ENERGY BALANCE
Linear Momentum
Conservation Laws
Coated Fuel Particles
ONE-DIMENSIONAL CALCULATIONS
TEST FACILITIES
FHR
DRACS
passive safety
natural circulation
numerical simulation
One-dimensional