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
May 2024
Jan 2024
Latest Journal Issues
Nuclear Science and Engineering
June 2024
Nuclear Technology
Fusion Science and Technology
Latest News
Developing a new regulatory framework for advanced reactors: Update on Part 53
White
The American Nuclear Society’s Risk-informed, Performance-based Principles and Policy Committee (RP3C) on March 29 held another presentation in its monthly Community of Practice (CoP) series. The presenter, Patrick White with the Nuclear Innovation Alliance (NIA), talked about the current status of efforts to develop a new regulatory framework for advanced reactors—known as 10 CFR Part 53 or simply Part 53. White serves as the research director of the NIA, where he leads their research as well as analysis-based stakeholder and policymaker engagement and education. White’s March 29 presentation is publicly available on YouTube and at ANS’s publication platform Nuclear Science and Technology Open Research (NSTOR).
RP3C chair N. Prasad Kadambi opened the CoP with brief introductory remarks about the RP3C before he welcomed White as the session’s presenter.
White covered three main topics: the history of the existing regulatory frameworks for new reactors, progress to date on the development of the Part 53 rule for advanced reactors, and the current status and next steps for the Part 53 rulemaking process.
Dong-Ho Shin, Su-Jong Yoon, Nam-Il Tak, Goon-Cherl Park, Hyoung-Kyu Cho
Nuclear Technology | Volume 191 | Number 3 | September 2015 | Pages 213-222
Technical Paper | Fission Reactors | doi.org/10.13182/NT14-102
Articles are hosted by Taylor and Francis Online.
In Korea, the Very High Temperature Gas-Cooled Reactor (VHTR) PMR200 is being developed in the Nuclear Hydrogen Development and Demonstration project. Its core consists of hexagonal prism-shaped graphite blocks for the fuel and reflector, and each hexagonal fuel block contains 108 cylindrical coolant holes and 210 fuel compacts. Because of these holes and fuels, the heat transfer in lateral directions in the fuel blocks becomes very complicated. Especially in accident situations when forced convection is lost, the majority of the afterheat flows in the radial direction by conduction across the large number of coolant holes. Moreover, radiation heat transfer is supposed to be added to the radial heat transfer modes owing to the high temperature of the VHTR core. Because of these complexities in radial heat transfer, reliable modeling for effective thermal conductivity (ETC) is required in order to analyze the reactor core thermal behavior using lumped-parameter codes, which are often used to evaluate the integrity of nuclear fuel embedded in the graphite block. In this study, the ETC model adopted in the GAMMA+ code was introduced, and the adequacy of the model was assessed by the commercial computational fluid dynamics (CFD) code CFX-13. The results of the CFD analysis were consistent with the ETC model in general even if a slight disagreement was shown for the case of high temperature. From these analyses, it could be concluded that the ETC model adopted in the GAMMA+ code is an adequate model for the analysis of the PMR200 reactor core. Moreover, it was found that the effect of fuel gap can cause an overprediction of the ETC if the fuel compact thermal conductivity is larger than the applicable range of the model.