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Division Spotlight
Reactor Physics
The division's objectives are to promote the advancement of knowledge and understanding of the fundamental physical phenomena characterizing nuclear reactors and other nuclear systems. The division encourages research and disseminates information through meetings and publications. Areas of technical interest include nuclear data, particle interactions and transport, reactor and nuclear systems analysis, methods, design, validation and operating experience and standards. The Wigner Award heads the awards program.
Meeting Spotlight
Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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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Nuclear Science and Engineering
February 2025
Nuclear Technology
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Latest News
Reboot: Nuclear needs a success . . . anywhere
The media have gleefully resurrected the language of a past nuclear renaissance. Beyond the hype and PR, many people in the nuclear community are taking a more measured view of conditions that could lead to new construction: data center demand, the proliferation of new reactor designs and start-ups, and the sudden ascendance of nuclear energy as the power source everyone wants—or wants to talk about.
Once built, large nuclear reactors can provide clean power for at least 80 years—outlasting 10 to 20 presidential administrations. Smaller reactors can provide heat and power outputs tailored to an end user’s needs. With all the new attention, are we any closer to getting past persistent supply chain and workforce issues and building these new plants? And what will the election of Donald Trump to a second term as president mean for nuclear?
As usual, there are more questions than answers, and most come down to money. Several developers are engaging with the Nuclear Regulatory Commission or have already applied for a license, certification, or permit. But designs without paying customers won’t get built. So where are the customers, and what will it take for them to commit?
G. P. Airey, A. R. Vaia, R. G. Aspden
Nuclear Technology | Volume 55 | Number 2 | November 1981 | Pages 436-448
Technical Paper | Materials | doi.org/10.13182/NT55-436
Articles are hosted by Taylor and Francis Online.
Inconel 690 is an austenitic nickel base alloy that has been considered for use as steam generator tubing in pressurized water reactors. It has a composition comparable to the currently used Inconel 600 apart from a higher chromium content (30%). Inconel 690 was evaluated for stress corrosion cracking (SCC) resistance in deaerated sodium hydroxide solutions and all-volatile treatment (AVT) environments, and the results are compared with those from thermally treated Inconel 600. The effect of such metallurgical variables as grain size and thermal treatment on SCC resistance was investigated. The caustic SCC resistance of Inconel 690 was significantly improved, compared to the mill-annealed product, when thermally treated over a wide range of temperatures [649 to 871°C (1200 to 1600°F)] and times (1 to 30 h). The SCC resistance of thermally treated Inconel 690 was found to be excellent when exposed to deaerated sodium hydroxide at 343°C (650°F) and 316°C (600°F). There was an increase in caustic SCC susceptibility with decreasing grain size in mill-annealed tubing. However, the corresponding thermally treated tubing showed no caustic SCC susceptibility. No evidence of SCC degradation has been found in either mill-annealed or thermally treated Inconel 690 exposed to high temperature AVT and “pure water” environments.
