scholarly journals A MULTIGROUP, MULTIREGION, ONE-SPACE DIMENSIONAL PROGRAM USING NEUTRON DIFFUSION THEORY

1956 ◽  
Author(s):  
J. Franklin ◽  
E.J. Leshan
Keyword(s):  
Diffusion Theory ◽  
Neutron Diffusion

Linearly Anisotropic Extensions of Asymptotic Neutron Diffusion Theory

1970 ◽  
Vol 41 (2) ◽  
pp. 226-236 ◽  
Author(s):  
R. J. Doyas ◽  
B. L. Koponen
Keyword(s):  
Diffusion Theory ◽  
Neutron Diffusion

Complementary variational principles in neutron diffusion theory

1967 ◽  
Vol 298 (1452) ◽  
pp. 97-101 ◽  
Keyword(s):  
Lower Bounds ◽  
Variational Principles ◽  
Diffusion Theory ◽  
Upper And Lower Bounds ◽  
Neutron Diffusion ◽  
Diffusion In Solids ◽  
Absorption Probability

Complementary variational principles associated with neutron diffusion in solids are presented. The resulting formulae are used to derive new expressions which provide upper and lower bounds for the absorption probability.

Numerical Analysis for a Molten Salt Reactor in the Presence of Fissile Lump

2012 ◽  
Author(s):  
Yao Xiao ◽  
Dalin Zhang ◽  
Zhangpeng Guo ◽  
Suizheng Qiu
Keyword(s):  
Molten Salt ◽  
Diffusion Theory ◽  
Neutron Diffusion ◽  
Delayed Neutron ◽  
Flow And Heat Transfer ◽  
Fuel Salt ◽  
Reactor Types ◽  
New Generation ◽  
Research And Design ◽  
Salt Flow

Molten salt reactors (MSRs) have seen a marked resurgence of interest over the past few decades, highlighted by their inclusion as one of the six Generation IV reactor types. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this paper, the attention is focused on the behaviors of a MSR in the presence of localized perturbations caused by fissile precipitates. A neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors, and the group constants dependent on the temperature are calculated by the code DRAGON. In addition, the k-epsilon turbulent model is adopted to establish the flow and heat transfer. The thermo-hydraulic and neutronic models which are coupled through the temperature, heat source and velocity are coded in a program. The effects of the localized perturbation on the distributions of power, temperature, neutron fluxes and delayed neutron precursors are obtained and discussed in detail. The results provide some valuable information for the research and design of this new generation reactor.

Tensorial formulation of neutron diffusion theory

Meccanica ◽  
1972 ◽  
Vol 7 (3) ◽  
pp. 131-140 ◽  
Author(s):  
Vinico C. Boffi ◽  
Tullio Trombetti
Keyword(s):  
Diffusion Theory ◽  
Neutron Diffusion
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