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Division Spotlight
Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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.”
C. D. Fletcher, L. S. Ghan
Nuclear Technology | Volume 95 | Number 2 | August 1991 | Pages 228-246
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT91-A34559
Articles are hosted by Taylor and Francis Online.
Large thermal-hydraulic systems computer codes are most often applied to investigate safety issues in existing nuclear facilities. One such code is applied to aid the design process for a proposed state-of-the-art research reactor. The RELAP5 computer code is used to simulate system response to hypothetical loss-of coolant accidents (LOCAs) in an early design of the Advanced Neutron Source (ANS). Among accident scenarios, a LOCA event is expected to be one of the most challenging to the ANS reactor core; similar analyses for other accident types are in progress. This is the first detailed study of ANS transient system response during accidents, and the outcome of the analysis is used to benefit the design process. The ANS model used is based on an early (preconceptual) cool ing system design layout. This early design has since been superseded by an improved design that is partly based on the results of these studies. The calculated responses of the early design to representative LOCA events are described; the simulations indicate that fuel melting and damage would be experienced for medium and large breaks. The effectiveness of employing a gascharged accumulator on the primary coolant system for preventing fuel damage following medium- and large-break LOCAs is evaluated. As a result of this evaluation, the new ANS design incorporates such accumulators. Analysis uncertainties are addressed, and the findings from this study that were used for the next phase of ANS design are highlighted.