Monte Carlo calculation of the core of the heavy water zero power reactor (HWZPR) using MCNP4c

Moosakhani, A.; Nasrabadi, M.N.; Timuri, B.

doi:10.1016/j.nucengdes.2011.02.005

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Cited by 6 publications

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“…Calculation of neutron flux in the core is very important and necessary for several areas such as power, neutron activation analysis, radiation damage, neutron dosimetry, and thermal neutron physics [10].…”

Section: −→Fission−productsmentioning

confidence: 99%

Neutronic calculations for CANDU thorium systems using Monte Carlo techniques

Saldideh¹,

Shayesteh²,

Eshghi³

2014

Chinese Phys. C

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In this paper, we have investigated the prospects of exploiting the rich world thorium reserves using Canada Deuterium Uranium (CANDU) reactors. The analysis is performed using the Monte Carlo MCNP code in order to understand how much time the reactor is in criticality conduction. Four different fuel compositions have been selected for analysis. We have obtained the infinite multiplication factor, k∞, under full power operation of the reactor over 8 years. The neutronic flux distribution in the full core reactor has already been investigated.

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Section: −→Fission−productsmentioning

confidence: 99%