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Transport code MCNP
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This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction beff for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo N-Particle transport code MCNP.
7p
christabelhuynh
29-05-2020
13
0
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The methodology uses measured values of intraelement isotope ratios of plutonium and fission product contaminants. MCNP radiation transport codes were used for various reactor core modeling and fuel burnup simulations. A reactor-dependent library of intra-element isotope ratio values as a function of burnup and time since irradiation was created from the simulation results.
10p
minhxaminhyeu3
12-06-2019
9
0
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