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Research Papers

The Effect of Iron Cross-Section in Thermal Region on Neutron Transport in VVER-1000 Mock-Up in LR-0 Reactor

[+] Author and Article Information
Martin Schulc

Research Center Rez,
Hlavní 130, Husinec Řež 250 68,
Czech Republic
e-mail: martin.schulc@cvrez.cz

Michal Košťál, Davit Harutyunyan, Marie Švadlenková, Vojtěch Rypar, Ján Milčák, Antonín Kolros

Research Center Rez,
Hlavní 130, Husinec Řež 250 68,
Czech Republic

1Corresponding author.

Manuscript received September 30, 2015; final manuscript received August 9, 2016; published online December 20, 2016. Assoc. Editor: Juan-Luis Francois.

ASME J of Nuclear Rad Sci 3(1), 011019 (Dec 20, 2016) (7 pages) Paper No: NERS-15-1198; doi: 10.1115/1.4034568 History: Received September 30, 2015; Accepted August 09, 2016

The iron cross-section in thermal regions influences the thermal neutron flux prediction in steel structural components of reactors and also in regions adjoining them. The thermal neutron flux level is proportional to pin power density in fuel. This quantity is an important criterion reflected in limits and conditions of reactor operation. The new power density evaluation shows notable, well distinguishable discrepancy between calculations realized using the CENDL-3.1 nuclear data library and experimentally determined pin power density in boundary rows of pins. All experiments were carried out in a water–water energetic reactor (VVER-1000) transport mock-up placed in the LR-0 reactor.

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References

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Figures

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Fig. 1

General view on the LR-0 reactor and mock-up core with barrel and baffle simulators

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Fig. 2

Top view of the experimental arrangement showing thermal neutron fluxes measuring points in the RPV simulator

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Fig. 3

Selected pins for power density measurement (red numbering) in the LR-0 reactor

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Fig. 4

C/E-1 comparison for fission density/s normalized per 10−9  A of monitor current in various nuclear data libraries. Gray lines correspond to plus or minus one sigma uncertainty

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Fig. 5

Relative comparison of pin power density in fuel assembly adjacent to the baffle of LR-0 mock-up calculated with ENDF-VII.0 and CENDL-3.1 libraries (baffle is situated behind the top row of pins). Spatial dimensions are given in centimetres

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Fig. 6

(a) Left panel illustrates the relative comparison of pin power density in LR-0 mock-up with the situation in which the baffle was replaced by the moderator; (b) right panel shows the comparison between LR-0 mock-up and case, where the influence of baffle is neglected (neutron importance was set to zero for baffle) in fuel assembly adjacent to the baffle of LR-0 mock-up. All calculations employ ENDF-VII.0 library. Spatial dimensions are given in centimetres

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