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Conference Spotlight
2025 ANS Winter Conference & Expo
November 9–12, 2025
Washington, DC|Washington Hilton
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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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Latest News
Empowering the next generation: ANS’s newest book focuses on careers in nuclear energy
A new career guide for the nuclear energy industry is now available: The Nuclear Empowered Workforce by Earnestine Johnson. Drawing on more than 30 years of experience across 16 nuclear facilities, Johnson offers a practical, insightful look into some of the many career paths available in commercial nuclear power. To mark the release, Johnson sat down with Nuclear News for a wide-ranging conversation about her career, her motivation for writing the book, and her advice for the next generation of nuclear professionals.
When Johnson began her career at engineering services company Stone & Webster, she entered a field still reeling from the effects of the Three Mile Island incident in 1979, nearly 15 years earlier. Her hiring cohort was the first group of new engineering graduates the company had brought on since TMI, a reflection of the industry-wide pause in nuclear construction. Her first long-term assignment—at the Millstone site in Waterford, Conn., helping resolve design issues stemming from TMI—marked the beginning of a long and varied career that spanned positions across the country.
Hideaki Asaka, Yutaka Kukita, Taisuke Yonomoto, Kanji Tasaka
Nuclear Technology | Volume 96 | Number 2 | November 1991 | Pages 202-214
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT91-A34606
Articles are hosted by Taylor and Francis Online.
Three 0.5% hot-leg small-break loss-of-coolant accident experiments are conducted at the ROSA-IV Large-Scale Test Facility (LSTF), a volumetrically scaled full-height model of a pressurized water reactor. Three experiments simulate breaks located at the side, bottom, and top of the horizontal hot-leg piping to investigate the effects of break orientation on system thermal-hydraulic responses. Although the overall system responses in the three experiments are qualitatively the same, the break flow rate is affected significantly by the break orientation for most of the time preceding the initiation of core uncovering: The break flow rate is largest for the bottom break and smallest for the top break. The RELAP5/MOD2 code fails to predict the differences in break flow rate observed in the experiments. However, several modifications, based on separate-effect experiments, made particularly to the break flow calculation models enable this code to simulate the experimental results well.
