Article Sidebar

Published: 2011-12-30
Abstract Views: 74
DOI: https://doi.org/10.53747/nst.v1i4.318
Issue
Vol. 1 No. 4 (2011)
Section
Articles
How to Cite
DO , Q. B., & NGUYEN , H. H. (2011). CALCULATION OF THE MODERATOR TEMPURATURE COEFFICIENT OF REACTIVITY FOR MINIATURE NEUTRON SOURCE REACTORS. Nuclear Science and Technology, 1(4), 42-52. https://doi.org/10.53747/nst.v1i4.318

CALCULATION OF THE MODERATOR TEMPURATURE COEFFICIENT OF REACTIVITY FOR MINIATURE NEUTRON SOURCE REACTORS

DO QUANG BINH1, NGUYEN HOANG HAI2
1 University of Technical Education Ho Chi Minh City,
2 Research and Development Center for Radiation Technology

Main Article Content

Abstract

This paper presents calculated results of the moderator temperature coefficient of reactivity through global reactor calculations and the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (MNSR). In this study, the group constants were calculated with the WIMSD code and the global reactor calculation was accomplished by the CITATION code. This work also presents a method for evaluation of the moderator temperature coefficient of reactivity at different temperatures and its average value in a range of temperatures directly through the values of moderator temperature ​​for MNSRs. This method provides simple analytical representation convenient for the reactor kinetics calculation and nuclear safety assessment.

Article Details

Keywords

MNSR, the moderator temperature coefficient of reactivity, group constants, lattice cell calculation, global reactor calculation

References

[1] A. Hainoun, S. Alissa, Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET, Nuclear Engineering and Design, 235, 33 (2005).
[2] T. B. Fowler, D. R. Vondy, F. B. Kemshell, Nuclear Reactor Core Analysis Code: CITATION. ORNL-TM-2496, RSICC (1971).
[3] WIMSD – A neutronics Code for Standard Lattice Physics Analysis, ANSWERS Software Service, AEA Technology (1997).
[4] G. I. Bell and S. Glasstone, Nuclear Reactor Theory, Van Nostrand Reinhold Company, New York (1970).
[5] James J. Duderstadt, Louis J. Hamilton, Nuclear Reactor Analysis, John Wiley & Sons, New York (1976).
[6] M. J. Halsall, C. J. Taubman, The ‘1986’ WIMS Nuclear Data Library, AEEW-R 2133, AEEE Winfrith, Sep. 1986.
[7] K. Khattab, I. Khamis, Calculation of the top beryllium shim plate worths for the Syrian neutron source reactor, Progress in Nuclear Energy, 44, 33 (2004).
[8] M. Albarhoum, A 3-D neutronics model for the calibration of the control rod of the Syrian MNSR, Progress in Nuclear Energy, 46, 159 (2005).
[9] I. Khamis, I. Sulieman, Monte Carlo simulation of a conceptual thermal column in the Syrian miniature neutron source reactor using MCNP – 4C, Annals of Nuclear Energy, 33, 622 (2006).
[10] A. Hainoun, H. Haj Hassan, N. Ghazi, Determination of major kinetic parameters of the Syrian MNSR for different fuel loading using Monte Carlo technique, Annals of Nuclear Energy, 36, 1663 (2009).