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
Isotopes & Radiation
Members are devoted to applying nuclear science and engineering technologies involving isotopes, radiation applications, and associated equipment in scientific research, development, and industrial processes. Their interests lie primarily in education, industrial uses, biology, medicine, and health physics. Division committees include Analytical Applications of Isotopes and Radiation, Biology and Medicine, Radiation Applications, Radiation Sources and Detection, and Thermal Power Sources.
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
Securing the advanced reactor fleet
Physical protection accounts for a significant portion of a nuclear power plant’s operational costs. As the U.S. moves toward smaller and safer advanced reactors, similar protection strategies could prove cost prohibitive. For tomorrow’s small modular reactors and microreactors, security costs must remain appropriate to the size of the reactor for economical operation.
David L. Luxat, Donald A. Kalanich, Joshua T. Hanophy, Randall O. Gauntt, Richard M. Wachowiak
Nuclear Technology | Volume 196 | Number 3 | December 2016 | Pages 684-697
Technical Paper | doi.org/10.13182/NT16-57
Articles are hosted by Taylor and Francis Online.
The Modular Accident Analysis Program (MAAP), Version 5 (MAAP5) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) are widely used integral plant response analysis computer codes. Both programs have been developed over the past 30 years for the purpose of simulating a range of beyond-design-basis accidents. The codes are benchmarked against numerous separate-effects experiments that reflect, to varying degrees, conditions expected to arise in light water reactor accidents. Such separate-effects tests, however, do not completely represent the novel physics that can arise through the interaction of multiple phenomena and physical processes at a reactor scale. Furthermore, aside from the Three Mile Island Unit 2 (TMI-2) core damage event, there is limited information available to evaluate reactor-scale behavior. Both MAAP5 and MELCOR have developed models to capture reactor-scale accident progression that, to a certain extent, extrapolate from separate-effects experiments, with assessment against the TMI-2 event only. Because of the limited information available to assess these extrapolated reactor-scale models, differences in MAAP5 and MELCOR code predictions do exist, most notably in the simulation of in-vessel core-melt progression. While these differences are not necessarily influential for the key metrics evaluated in probabilistic risk assessments, they can have a more pronounced impact on studies assessing the efficacy of accident management measures. This paper reports the first phase of a MAAP-MELCOR crosswalk designed to identify the key core-melt progression modeling differences. The results of this study highlight the impact that assumptions about reactor-scale, in-vessel core debris morphology have on (a) the potential for high temperatures to develop above the reactor core and in the main steam lines and (b) the magnitude and extent of the period for in-vessel hydrogen generation. These examples play critical roles in the evolution of challenges to the reactor pressure vessel pressure boundary and containment and are ultimately central to the evaluation of accident management effectiveness.