Evaluation of Two 300 MWe Fourth Generation Pb-Bi Reactor System Concepts
- University of Tennessee, Knoxville Tennessee 37996 (United States)
This paper describes the evaluation of two 300 MWe modular Pb-Bi cooled reactor system concepts that can be field assembled from components shipped on standard rail cars or on trucks. Thus, the largest components must be smaller than 12' x 12' x 80' (3.66 m x 3.66 m x 24.4 m) and should weigh no more than 80 tons. One of these systems utilizes a cylindrical two-loop containment vessel for the core and the other is a slab design. The fuel for both designs consists of standard-sized metallic IFR fuel in 17 x 17 square array assemblies with a pitch-to-diameter ratio of 1.15. The coolant outlet temperature is limited by current material technology, which is estimated to be 550 C. The primary coolant inlet temperature is selected to be 350 C. This is well above the melting temperature of Pb-Bi, and it is expected to be sufficiently high to limit transient-induced thermal stresses to acceptable values. Coolant flow rates through the core and external piping are below 1 m/s. The results from neutronics calculations include power distributions, reactivity coefficients, and fuel depletion, and results from heat transfer calculations include temperatures and flow rates at various locations in the primary and secondary systems. The neutronic design calculations are accomplished by using a discrete ordinate transport code and a cross section processing system developed at Oak Ridge National Laboratory. Two-dimensional flux distributions are obtained with the DOORS system, and ORIGEN-S, coupled with KENO, is used for time-dependent depletion calculations. The thermal-hydraulic design of the core consists of heat transfer and fluid flow calculation for an average channel. The inlet and outlet temperatures, along with the fuel centerline temperature, are determined in conjunction with core flow rates, pumping power, and total power output. This is accomplished by using a lumped parameter steady-state model with a spreadsheet and by using a one-dimensional time-dependent model. Results from the thermal-hydraulic calculation obtain a thermal efficiency of 41%, but an efficiency of about 45 % could be obtained. The nominal power density and good thermal conductivity of Pb-Bi will permit decay heat to be handled more effectively than for sodium-cooled design or for light water reactors. The low vapor pressure of Pb-Bi permits the use of a thin walled pressure vessel on the order of centimeters as compared to the 30-40 cm thick PWR vessel, and the high boiling point of the lead bismuth assures that the core will remain covered in the event of a loss of coolant outside the primary vessel. (authors)
- Research Organization:
- The ASME Foundation, Inc., Three Park Avenue, New York, NY 10016-5990 (United States)
- OSTI ID:
- 21062359
- Resource Relation:
- Conference: ICONE 10: 10. international conference on nuclear engineering, Arlington - Virginia (United States), 14-18 Apr 2002; Other Information: Country of input: France
- Country of Publication:
- United States
- Language:
- English
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