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Radiation Protection & Shielding
The Radiation Protection and Shielding Division is developing and promoting radiation protection and shielding aspects of nuclear science and technology — including interaction of nuclear radiation with materials and biological systems, instruments and techniques for the measurement of nuclear radiation fields, and radiation shield design and evaluation.
Meeting Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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
NextGen MURR Working Group established in Missouri
The University of Missouri’s Board of Curators has created the NextGen MURR Working Group to serve as a strategic advisory body for the development of the NextGen MURR (University of Missouri Research Reactor).
Hossein Hashemi-Jozani, Khalil Moshkbar-Bakhshayesh, Soroush Mohtashami, Behzad Rokhbin
Nuclear Technology | Volume 210 | Number 1 | January 2024 | Pages 180-188
Note | doi.org/10.1080/00295450.2023.2224131
Articles are hosted by Taylor and Francis Online.
The computerized simulation of the reactor core is one of the significant steps necessary for designing a nuclear power plant. So far, very suitable Monte Carlo–based codes have been developed (e.g., MCNP, TRIPOLI, KENO, OpenMC, etc.) for the neutronic simulation of the reactor core. In this study, an approach based on Geant4, as an extendable code with the capability to provide a comprehensive reactor core design tool, is developed to calculate the effective multiplication factor (keff) and neutron flux distribution. A combination of the Geant4 code and the NJOY code is applied to calculate the temperature-dependent cross-section library. The C5G7-1D, the Godiva critical facility, and the Jordan subcritical reactor are examined as a benchmarks/case study. The results of the calculation of keff (i.e., relative error < 0.1%) and flux distribution (i.e., relative error <3%) are in very good agreement with the calculation results of the MCNP code and the experimental results. The extensions for the calculation of thermodynamic/thermohydraulic effects as well as the calculation of electron/photon transport and reactor dynamics are under development and will be reported as subsequent results.