A Method to Estimate Fission Product Concentration Uncertainty in a Multi-Time-Step MCNP6 Code Nuclear Fuel Burnup Calculation. (2nd January 2020)

• Record Type: Journal Article
• Authors: Minamigawa, Yasuhiro; Kitcher, Evans D.; Chirayath, Sunil S.
• Is Part Of: Nuclear technology. Volume 206:Number 1(2020)
• Page Start: 73
• Page End: 81
• Publication Date: 2020-01-02
• Subjects: Nuclear forensics -- depletion calculation -- Monte Carlo method -- concentration uncertainties -- random sampling; Nuclear engineering -- Periodicals; Nuclear engineering; Nuclear Physics; Periodicals; Periodicals; 621.48
• DOI: 10.1080/00295450.2019.1624429
• Languages: English
• ISSNs: 1943-7471

Abstract

Abstract: 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.