DRACS thermal performance evaluation for FHR

Journal Article · Sun Mar 01 00:00:00 EST 2015 · Annals of Nuclear Energy (Oxford)

DOI:https://doi.org/10.1016/j.anucene.2014.10.032· OSTI ID:1177671

Lv, Q. ^[1]; Lin, H. C. ^[1]; Kim, I. H. ^[1]; Sun, X. ^[1]; Christensen, R. N. ^[1]; Blue, T. E. ^[1]; Yoder, G. L. ^[2]; Wilson, D. F. ^[2]; Sabharwall, P. ^[3]

---

1. The Ohio State Univ., Columbus, OH (United States). Dept. of Mechanical and Aerospace Engineering, Nuclear Engineering Program
2. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
3. Idaho National Lab. (INL), Idaho Falls, ID (United States)

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: Idaho National Lab. (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• DOE Contract Number: AC07-05ID14517
• OSTI ID: 1177671
• Report Number(s): INL/JOU-15-34238; TRN: US1500088
• Journal Information: Annals of Nuclear Energy (Oxford), Vol. 77; ISSN 0306-4549
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

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
DRACS
FHR
NATURAL CIRCULATION
NUMERICAL SIMULATION
ONE-DIMENSIONAL
PASSIVE SAFETY