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Meeting Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
Standards Program
The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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Latest News
The U.S. Million Person Study of Low-Dose-Rate Health Effects
There is a critical knowledge gap regarding the health consequences of exposure to radiation received gradually over time. While there is a plethora of studies on the risks of adverse outcomes from both acute and high-dose exposures, including the landmark study of atomic bomb survivors, these are not characteristic of the chronic exposure to low-dose radiation encountered in occupational and public settings. In addition, smaller cohorts have limited numbers leading to reduced statistical power.
Yasuhiro Minamigawa, Evans D. Kitcher, Sunil S. Chirayath
Nuclear Technology | Volume 206 | Number 1 | January 2020 | Pages 73-81
Technical Paper | doi.org/10.1080/00295450.2019.1624429
Articles are hosted by Taylor and Francis Online.
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.