Optimum Burnup Distribution in a Continuously Fueled Reactor

Conference · Sun Jun 13 04:00:00 EDT 1976 · Transactions of the American Nuclear Society

DOI:https://doi.org/10.13182/NSE78-A27271· OSTI ID:7154017

Wight, A. L. ^[1]

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1. Atomic Energy of Canada Limited (AECL), Mississauga, ON (Canada)

The Canadian Deuterium-Uranium (CANDU) pressurized heavy water reactor is fueled continuously at power, with alternate channels being fueled in opposite directions (continuous bidirectional fueling). The rate at which channels are refueled in various regions of the core determines the burnup distribution in the core. The burnup distribution in the core determines the power distribution. In present practice, the core is divided radially into two burnup regions having constant average discharge burnup. The limit on maximum neutron flux and the requirement for a critical system determine the size of the inner burnup region and the values of the burnups in the two regions. We can increase the core average exit burnup if we allow the burnup distribution to vary continuously rather than being region wise constant. The purpose of this analysis to derive an optimum burnup distribution that will maximize core average discharge burnup subject to a limit on maximum flux. This equivalent to minimizing the total fuel feed rate. A set of equations describing the optimum distribution of burnup has been derived using calculus of variations techniques. These equations have been solved numerically in one-dimensional cylindrical geometry for homogeneous cores of approximately the size of current generation CANDU reactors.

• Research Organization: Atomic Energy of Canada Limited (AECL), Mississauga, ON (Canada)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• OSTI ID: 7154017
• Conference Information: Journal Name: Transactions of the American Nuclear Society Journal Issue: 1 Journal Volume: 23
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210400* -- Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated
Analysis
BURNUP
CANDU TYPE REACTORS
Canadian Deuterium-Uranium (CANDU)
Continuous Bidirectional Fueling
Equations
FUEL MANAGEMENT
HEAVY WATER MODERATED REACTORS
Homogeneous Cores
MATHEMATICAL MODELS
Nuclear Criticality Safety Program (NCSP)
OPTIMIZATION
One-Dimensional Cylindrical Geometry
PRESSURE TUBE REACTORS
Pressurized Heavy Water Reactor
REACTORS
Solutions
THERMAL REACTORS