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Conference Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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Latest News
Ho Nieh nominated to the NRC
Nieh
President Trump recently nominated Ho Nieh for the role of commissioner in the Nuclear Regulatory Commission through the remainder of a term that will expire June 30, 2029.
Nieh has been the vice president of regulatory affairs at Southern Nuclear since 2021, though he is currently working as a loaned executive at the Institute of Nuclear Power Operations, where he has been for more than a year.
Nieh’s experience: Nieh started his career at the Knolls Atomic Power Laboratory, where he worked primarily as a nuclear plant engineer and contributed as a civilian instructor in the U.S. Navy’s Nuclear Power Program.
From there, he joined the NRC in 1997 as a project engineer. In more than 19 years of service at the organization, he served in a variety of key leadership roles, including division director of Reactor Projects, division director of Inspection and Regional Support, and director of the Office of Nuclear Reactor Regulation.
K. Tasaka, M. Suzuki, Y. Anoda, Y. Koizumi, T. Yonomoto, H. Kumamaru, H. Nakamura, M. Shiba
Nuclear Technology | Volume 68 | Number 1 | January 1985 | Pages 77-93
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT85-A33569
Articles are hosted by Taylor and Francis Online.
The Rig of Safety Assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency-core-cooling-system (ECCS) tests. Experimental results obtained so far confirm that the severest single failure assumption in ECCS is the high-pressure core spray system failure even in a large-break LOCA in a BWR. The measured peak cladding temperature was well below the present safety criterion of 1473 K, even with the single failure assumption in ECCS, and the effectiveness of ECCS for core cooling during a double-ended-break LOCA has been confirmed. The overall agreement between the results calculated by the RELAP4/MOD6/U4/J3 computer code and the experimental results is good. The similarity between the ROSA-III test and a BWR LOCA has been confirmed through the comparison of calculated results for the ROSA-III facility and a BWR system.
