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
Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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
Apr 2024
Jan 2024
Latest Journal Issues
Nuclear Science and Engineering
May 2024
Nuclear Technology
Fusion Science and Technology
Latest News
X-energy receives federal tax credit for TRISO fuel facility
Advanced reactor company X-energy has been awarded $148.5 million in tax credits under the Inflation Reduction Act for construction of its TRISO-X fuel fabrication facility in Oak Ridge, Tenn.
Ho Jin Park, Hyung Jin Shim, Han Gyu Joo, Chang Hyo Kim
Nuclear Science and Engineering | Volume 172 | Number 1 | September 2012 | Pages 66-77
Technical Paper | doi.org/10.13182/NSE11-22
Articles are hosted by Taylor and Francis Online.
The purpose of this paper is to present the Monte Carlo (MC) method augmented by the B1 spectrum to generate few-group diffusion theory constants, to assess their qualification in terms of the core depletion analysis, and thus to validate the MC method implemented into the Seoul National University MC code, McCARD, as a few-group diffusion theory constant generator. To do so, two-step core neutronics analyses are conducted for two types of power reactors, pressurized water reactors and very high temperature gas-cooled reactors, by the McCARD/MASTER code system in which McCARD is used as a MC few-group constant generation code and MASTER as a deterministic core analysis code. The two-step calculations for the effective multiplication factors and assembly power distributions of the two types of power reactor cores by McCARD/MASTER are compared with the reference calculations from McCARD, the nuclear design report, or measurements. By showing excellent agreement between McCARD/MASTER and the reference neutronics analyses for the two types of power reactors, it is concluded that the MC method implemented in McCARD can generate few-group diffusion theory constants that are well qualified for high-accuracy two-step core neutronics calculations.