Neutronic and depletion analysis of the Pb-AHTR
Conference
·
OSTI ID:20979641
- Department of Nuclear Engineering, University of California, Berkeley, CA 94720-1730 (United States)
The PB-AHTR is a Pebble Bed Advanced High Temperature Reactor that is cooled with the liquid salt flibe (LiF-BeF{sub 2}) rather than helium. This study presents a preliminary neutronic and depletion analysis for the PBAHTR. The attainable burnup is determined as a function of uranium loading per pebble, power density and core dimensions. It is found that the optimal design for a 425 {mu}m UC{sub 0.5}O{sub 1.5} fuel kernel diameter, 10% enriched uranium, features a graphite-to-heavy metal ratio of {approx}360 and its reactivity coefficients are all negative. A comparison with the helium-cooled pebble-bed reactor and with a prismatic-fuel reactor that is cooled with either flibe or helium is also presented. It is found that the PB-AHTR offers similar discharge burnup as the other three designs. As compared to the gas-cooled pebble bed, the PB-AHTR uranium loading and energy generated per pebble are {approx}2.5 times higher. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 20979641
- Country of Publication:
- United States
- Language:
- English
Similar Records
The Pebble Recirculation Experiment (PREX) for the AHTR
Transient Thermal Response of the Pb-AHTR to Loss of Forced Cooling
High Temperature Fluoride Salt Test Loop
Conference
·
Sun Jul 01 00:00:00 EDT 2007
·
OSTI ID:20979639
Transient Thermal Response of the Pb-AHTR to Loss of Forced Cooling
Conference
·
Sun Jul 01 00:00:00 EDT 2007
·
OSTI ID:20979643
High Temperature Fluoride Salt Test Loop
Technical Report
·
Mon Nov 30 23:00:00 EST 2015
·
OSTI ID:1237612