Title: Calculation of fuel burnup and radioactive inventory in the CANDU reactor using the GETERA and MCNP4C codes

Authors: S. Dawahra; K. Khattab; G. Saba

Addresses: Department of Nuclear Engineering, Atomic Energy Commission, P.O. Box 6091, Damascus, Syria ' Department of Nuclear Engineering, Atomic Energy Commission, P.O. Box 6091, Damascus, Syria ' Department of Nuclear Engineering, Atomic Energy Commission, P.O. Box 6091, Damascus, Syria

Abstract: A study of fuel burnup and radioactive inventory for the CANDU bundle (37 element NU) has been carried out to benchmark the GETERA and MCNP4C computer codes for cluster geometry. The infinite multiplication factor was calculated as a function of burnup and compared with available results (IAEA benchmark, KENO VI and MCNPX Monte Carlo codes). Good agreement was observed between the present calculations and the previously published results. The differences between the GETERA code and MCNP literature were −1.85 to −14.0 mk, the GETERA code and transport literature were −4 to −8 mk, the GETERA code and IAEA benchmark were 1 to −8 mk. The GETERA code was then used to calculate the 235U and the 239Pu number densities in the bundle with maximum relative difference less than 7% compared with IAEA benchmark codes. The inventories and the corresponding activities of some important fission products were also calculated. The total radioactivity of the CANDU bundle was calculated (using the GETERA code) summing the radioactivities of all the radionuclides at 3.1 MWd/kg (available in the literature) of fuel burnup and found to be 1.158E15 compared with the WIMSD4 code result (1.122E15 Bq) with relative difference less than 3.2%.

Keywords: fuel burnup; radioactive inventory; CANDU reactors; radioactivity; GETERA code; MCNP4C code; radionuclides; nuclear power plants; NPP; nuclear energy; nuclear reactors.

DOI: 10.1504/IJNEST.2016.078956

International Journal of Nuclear Energy Science and Technology, 2016 Vol.10 No.3, pp.183 - 200

Available online: 04 Sep 2016 *

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