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Division Spotlight
Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
Meeting Spotlight
Utility Working Conference and Vendor Technology Expo (UWC 2024)
August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
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
Virginia utility considers SMRs
Dominion Energy Virginia has issued a request for proposals from leading nuclear companies to study the feasibility of putting a small modular reactor at its North Anna nuclear power plant.
While the utility says it is not a commitment to build an SMR at the site, the RFP is “an important first step in evaluating the technology and the North Anna site to support Dominion Energy customers’ future energy needs consistent with the company’s most recent Integrated Resource Plan.”
Yasuo Koizumi, Hiroshige Kumamaru, Yuichi Mimura+, Yutaka Kukita, Kanji Tasaka†
Nuclear Technology | Volume 96 | Number 3 | December 1991 | Pages 290-301
Technical Paper | Nuclear Reactor Safety | doi.org/10.13182/NT91-A34590
Articles are hosted by Taylor and Francis Online.
Cold-leg small-break loss-of-coolant accident experiments were conducted for break areas ranging from 0.5 to 10% of the scaled cold-leg flow area using the Large-Scale Test Facility (LSTF). The LSTF is a volumetrically scaled simulator of a Westinghouse-type pressurized water reactor. For all the experiments, the core collapsed liquid level was temporarily depressed when liquid in the primary loop U-bend (crossover leg) was being cleared by steam. For scaled break areas <2.5%, the minimum core liquid level was equal to the lowest elevation of the crossover leg. For break areas >5%, the minimum core level was even lower because differential pressures created by the residual liquid holdup in the steam generator (SG) upflow side affected the core liquid level adversely. This influence of SG liquid holdup on the minimum core liquid level was larger for larger break sizes within the range of these experiments; thus, a more severe core level depression was seen for larger break sizes. Also, for the same break size, the core level depression was more severe when higher core power values were used for the simulation of the postscram core power decay. The RELAP5/MOD2 code reasonably well predicted the major phenomena observed in the experiments; however, several shortcomings were found in interfacial drag calculation for the SG U-tube inlet and the hot-leg outlet to the SG inlet plenum and core.