BURNING PLUTONIUM IN THE MULTI-SPECTRUM CANDU-BASED REACTOR (MSCR) USING THE SERPENT CODE
13th International Conference on CANDU Fuel - 2016 Aug. 15-18


Presented at:
13th International Conference on CANDU Fuel
2016 Aug. 15-18
Location:
Kingston, ON Canada
Session Title:
Session 10: Fuel Fabrication and Core Design

Authors:
M. Hussein (Royal Military College of Canada)
H. Bonin (Royal Military College of Canada)
B.J. Lewis (Royal Military College of Canada)
  

Abstract

The actinides present in dismantled nuclear warheads represent a political and security concern. Considering the efficient transmutation of actinides in fast neutron reactors, a novel multi-spectrum reactor has been designed to burn the actinides based on a CANDU6 reactor design with two concentric regions. The inner or the fast neutron spectrum core is fuelled with plutonium oxide mixed with depleted uranium oxide. The plutonium-to-uranium ratio is 14.13%. The MSCR is dedicated to the burning of actinides. Helium is used as a coolant and filling material for the fast core. The outer thermal spectrum core is fuelled with natural uranium with heavy water as both moderator and coolant. In both cores, 37-element CANDU fuel bundles are employed. The lattice pitch for the inner fast neutron core is chosen to maximize the number of fuel channels. The simulations of the multi-spectrum reactor and the burnup calculations were carried out using the Serpent code. The multi-spectrum CANDU reactor produces a uniform power distribution in both the fast and the thermal cores and the form factors of the both core were similar to the safety traditional CANDU value (~1.2 to~1.3). The discharge burnup and the actinide burnup have been significantly improved in comparison to a traditional CANDU 6 reactor at the same power level.

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