ANS is committed to advancing, fostering, and promoting the development and application of nuclear sciences and technologies to benefit society.
Explore the many uses for nuclear science and its impact on energy, the environment, healthcare, food, and more.
Division Spotlight
Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
Meeting Spotlight
2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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!
Latest Magazine Issues
May 2024
Jan 2024
Latest Journal Issues
Nuclear Science and Engineering
June 2024
Nuclear Technology
Fusion Science and Technology
Latest News
Excelsior University student section awarded community education grant
The American Nuclear Society Student Section at Excelsior University in Albany, N.Y., was awarded a $5,000 grant from the ANS Student Section Strategic Fund initiative for its program, Empowering Tomorrow’s Nuclear Innovators: A Collaborative Approach to Nuclear Technology Education and Awareness.
Robert David
Nuclear Technology | Volume 205 | Number 11 | November 2019 | Pages 1488-1494
Technical Paper | doi.org/10.1080/00295450.2019.1597581
Articles are hosted by Taylor and Francis Online.
Finite element analysis is used to study heat transfer from a corium pool at the bottom of the calandria to its surroundings during a severe accident in a CANDU 6 reactor. The shape of the corium crust around the pool and the steady-state heat fluxes exiting the calandria are calculated for representative accident conditions. The sensitivity of the results to several model parameters is examined. Calculated heat fluxes can be compared to measurements of the critical heat flux at different locations on the outside of the calandria in order to assess the possibility of in-vessel retention of the molten core.