A Method to Estimate Fission Product Concentration Uncertainty in a Multi-Time-Step MCNP6 Code Nuclear Fuel Burnup Calculation Academic Article

abstract

• 2019, 2019 American Nuclear Society. The Monte Carlo N-Particle (MCNP6) radiation transport code is widely used to perform material transmutation and depletion calculations using the embedded module CINDER90. CINDER90 is capable of obtaining fission product and transuranic nuclide concentrations with a high level of accuracy in irradiated nuclear fuel. This information is very useful for many nuclear applications including reactor design and analysis, nuclear safeguards, nuclear security, and nuclear forensics, to name a few. However, at present the MCNP6 code does not estimate the overall statistical uncertainty in the nuclide concentrations reported at the end of a depletion calculation. We report our approach using a random sampling method to estimate stochastic uncertainty in fission product nuclide concentration using various parameters reported in MCNP6 output and how these uncertainties are affected by the calculation parameters.

authors

• Chirayath, Sunil

published proceedings

• NUCLEAR TECHNOLOGY

author list (cited authors)

• Minamigawa, Y., Kitcher, E. D., & Chirayath, S. S.

citation count

• 1

complete list of authors

• Minamigawa, Yasuhiro||Kitcher, Evans D||Chirayath, Sunil S

publication date

• January 2020

publisher

• Taylor & Francis  Publisher

published in

• Nuclear Technology  Journal

keywords

• Concentration Uncertainties
• Depletion Calculation
• Monte Carlo Method
• Nuclear Forensics
• Random Sampling

Digital Object Identifier (DOI)

• 10.1080/00295450.2019.1624429

start page

• 73

end page

• 81

volume

• 206

issue

• 1

URL

• http://dx.doi.org/10.1080/00295450.2019.1624429