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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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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.
Vincent P. Manno
Nuclear Technology | Volume 48 | Number 3 | May 1980 | Pages 281-288
Technical Paper | Fuel | doi.org/10.13182/NT80-A32474
Articles are hosted by Taylor and Francis Online.
The current regulatory requirement that peak cladding temperatures (PCTs) never exceed 1204°C (2200°F) at any time during a loss-of-coolant accident (LOCA) is frequently the most limiting factor in setting core peaking factor limits. Of the many plant specific characteristics involved in predicting a fuel rod’s thermal response to a LOCA, the containment or “back” pressure plays a significant role, especially in pressure suppression pressurized water reactor (PWR) containments. The back pressure effect is studied by comparing the predicted PCT histories at back pressure levels of 138, 155, 172, and 207 kN/m2 (20, 22.5, 25.0, and 30 psia). A typical four-loop PWR with 15 × 15 fuel assemblies is analyzed. The analysis is performed using an in-house LOCA code named HEATUP-R/AEP, which calculates fuel thermal response during core reflood. In addition to temperature, the reflood rates, exit qualities, and cladding oxidation rates are studied. Results show significant increases in PCTs at lower pressure due to enhanced steam binding in the coolant loops.
