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Conference Spotlight
2025 ANS Winter Conference & Expo
November 9–12, 2025
Washington, DC|Washington Hilton
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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
NRC issues final environmental approval for TerraPower’s SMR
TerraPower announced that the final environmental impact statement for its Natrium small modular reactor has been completed by the Nuclear Regulatory Commission.
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.