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Research Papers

Parametric Lattice Study of a Graphite-Moderated Molten Salt Reactor

[+] Author and Article Information
Boris A. Hombourger

Laboratory for Reactor Physics and Systems Behavior, Paul Scherrer Institut (PSI),
CH-5232 Villigen, Switzerland;
Laboratory for Reactor Physics and Systems Behavior,
École Polytechnique Fédérale de Lausanne,
CH-1015 Lausanne, Switzerland e-mail: boris.hombourger@psi.ch

Jiří Křepel

Senior Scientist,
Laboratory for Reactor Physics and Systems Behavior,
Paul Scherrer Institut (PSI),
CH-5232 Villigen, Switzerland

Konstantin Mikityuk

Group Head, Laboratory for Reactor Physics and Systems Behavior,
Paul Scherrer Institut (PSI),
CH-5232 Villigen, Switzerland

Andreas Pautz

Laboratory Head, Professor, Laboratory for Reactor Physics and Systems Behavior, Paul Scherrer Institut (PSI), CH-5232 Villigen, Switzerland;Laboratory for Reactor Physics and Systems Behavior,
École Polytechnique Fédérale de Lausanne,
CH-1015 Lausanne, Switzerland

1Corresponding author.

Manuscript received August 19, 2014; final manuscript received October 20, 2014; published online February 9, 2015. Assoc. Editor: Lin-wen Hu.

ASME J of Nuclear Rad Sci 1(1), 011009 (Feb 09, 2015) (8 pages) Paper No: NERS-14-1038; doi: 10.1115/1.4026401 History: Received August 19, 2014; Accepted December 04, 2014; Online February 09, 2015

Molten salt reactors (MSRs) are promising advanced nuclear reactors for closure of the fuel cycle. This paper discusses the core design of graphite-moderated MSRs, thanks to a parametric study of the fuel and moderator lattice. The study is conducted at equilibrium of the thorium-uranium fuel cycle for several fuel channel radius and moderator block size combinations. The equilibrium composition for each studied configuration is derived with the help of an in-house MATLAB code, EQL0D, which uses the Serpent 2 Monte Carlo neutronics code for the calculation of reaction rates. The results include excess reactivity at equilibrium, mirroring the breeding gain, and the actinide vector composition for each configuration. Moreover, the occurence of an optimum of the excess reactivity per percent uranium-233 was observed. The investigations showed that it is systematically seen at an interchannel distance equal to the neutron slowing-down length in graphite for each configuration and does not depend on the salt channel radius beyond a certain size, which is given by the thermal fission rate reaching the levels of the fast fission rate. In this way, an exotic energy and spatial distribution of the neutrons are attained. The investigations highlight the potential attractiveness, from a neutronics/fuel cycle point of view, of both large fuel channels and moderators with a shorter neutron slowing-down length.

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References

Figures

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Fig. 1

Illustration of heterogeneity for two salt shares

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Fig. 2

Scheme of the EQL0D procedure. The upper-right step (green) is performed by the cell code and the rest of the steps (orange) by the MATLAB-based script

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Fig. 3

Infinite neutron multiplication factor at the beginning of equilibrium cycle for several configurations

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Fig. 4

Uranium-233 weight percentage among actinides in the equilibrium composition

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Fig. 5

Reproduction of the U233 weight percentage at equilibrium using the inverse of the capture rate according to Eq. (2) and normalized to the maximum of the actual value for comparison with Fig. 4

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Fig. 6

Excess reactivity per weight percentage of Uranium-233 at equilibrium as a function of the graphite length between channels

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Fig. 7

Graphite slowing-down length computed according to Eq. (3)

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Fig. 8

Neutron spectra of channels of 1, 10, 50 cm radius and salt share of 30%

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Fig. 9

Neutron spectra of channels of 5 cm and salt shares of 10%, 40%, and 80%

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Fig. 10

Spatial distribution of thermal (above) and fast (below) fission rates (in arbitrary units) in the 30 cm and 40% case

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Fig. 11

Thermal-to-total fission rate ratio according to Eq. (4) as function of the distance between channels

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Fig. 12

Uranium-234-to-Uranium-233 ratio as function of the distance between channels

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Fig. 13

Protactinium-233 capture contribution to the Uranium-234-to-Uranium-233 ratio as function of the distance between channels

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Fig. 14

Uranium-233 capture contribution to the Uranium-234-to-Uranium-233 ratio as function of the distance between channels

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Fig. 15

Calculated bare cylinder critical radius given by Eq. (6) as function of the interchannel distance

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Fig. 16

Calculated migration area given by Eq. (7) as function of the interchannel distance

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