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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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!
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February 2024
Latest News
Can hydrogen be the transportation fuel in an otherwise nuclear economy?
Let’s face it: The global economy should be powered primarily by nuclear power. And it probably will by the end of this century, with a still-significant assist from renewables and hydro. Once nuclear systems are dominant, the costs come down to where gas is now; and when carbon emissions are reduced to a small portion of their present state, it will become obvious that most other sources are only good in niche settings. I mean, why use small modular reactors to load-follow when they can just produce that power instead of buffering it?
Randall Gauntt, Donald Kalinich, Jeffrey Cardoni, Jesse Phillips
Nuclear Technology | Volume 186 | Number 2 | May 2014 | Pages 161-178
Technical Paper | Reactor Safety | doi.org/10.13182/NT13-59
Articles are hosted by Taylor and Francis Online.
In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code and developing an understanding of the likely accident progression. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of MELCOR and the Fukushima models against plant data. In this study Sandia National Laboratories developed MELCOR 2.1 models of Fukushima Daiichi Units 1 (1F1), 2, and 3 as well as the Unit 4 spent fuel pool. This paper reports on the analysis of the 1F1 accident. Details are presented on the modeled accident progression, hypothesized mode of failures in the reactor pressure vessel (RPV) and containment pressure boundary, and release of fission products to the environment. The MELCOR-predicted RPV and containment pressure trends compare well with available measured pressures. Conditions leading up to the observed explosion of the reactor building are postulated based on this analysis where drywell head flange leakage is thought to have led to accumulation of flammable gases in the refueling bay. The favorable comparison of the results from the analyses with the data from the plant provides additional confidence in MELCOR to reliably predict real-world accident progression. The modeling effort has also provided insights into future data needs for both model development and validation.