Parameter Justification to Validate Neutronic Calculation Using MCNP6 for CANDLE Reactor Case Maryam Afifah(a)*, Zaki Su^ud(a), Nuri Trianti(b) , Dwi Irwanto(a)
a) Department of Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia 40132
* 34922002[at]mahasiswa.itb.ac.id
b) Research Centre for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency, PUSPIPTEK, Indonesia 15314
Abstract
This paper presents the survey parameters to justify several factors with the minimum parameters that should be applied to validate neutronic calculation for CANDLE reactor case. In particular, we investigate the effect of neutron number history, inactive-active cycle, the burnup time step length and the recommendation of those factors as a way to improve the computing efficiency using MCNP6 code. This study presented the minimum number of neutron history and the total cycle to minimize the oscillation. However, the keff value with larger length step has a difference of 1% compared to smaller length step application. In this case we explore the methodology that used in MCNP6 to calculate burnup to make sure the numerically stable method was applied. Lastly, this study concluded some recommendations that should be applied for complex core geometry to achieve accuracy and efficiency for neutronic calculation using Monte Carlo method especially using MCNP6 for CANDLE reactor case.
Keywords: CANDLE reactor, Monte Carlo, Depletion, time step length, MCNP6
