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Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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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Remembering Charles E. Till
Charles E. Till
Charles E. Till, an ANS member since 1963 and Fellow since 1987, passed away on March 22 at the age of 89. He earned bachelor’s and master’s degrees from the University of Saskatchewan and a Ph.D. in nuclear engineering from Imperial College, University of London. Till initially worked for the Civilian Atomic Power Department of the Canadian General Electric Company, where he was the physicist in charge of the startup of the first prototype CANDU reactor in Canada.
Till joined Argonne National Laboratory in 1963 in the Applied Physics Division, where he worked as an experimentalist in the Fast Critical Experiments program. He then moved to additional positions of increasing responsibility, becoming division director in 1973. Under his leadership, the Applied Physics Division established itself as one of the elite reactor physics organizations in the world. Both the experimental (critical experiments and nuclear data measurements) and nuclear analysis methods work were internationally recognized. Till led Argonne’s participation in the International Nuclear Fuel Cycle Evaluation (INFCE), and he was the lead U.S. delegate to INFCE Working Group 5, Fast Breeders.
Tae-Hoon Lee, Young-Soo Kim, Tae-Je Kwon, Hee-Sung Shin, Ho-Dong Kim
Nuclear Technology | Volume 179 | Number 2 | August 2012 | Pages 196-204
Technical Paper | Fuel Cycle and Management | doi.org/10.13182/NT11-77
Articles are hosted by Taylor and Francis Online.
In pyroprocessing it is important to determine the amount of Pu in the various streams of materials involved. This paper presents two approaches to determine the Pu mass of spent fuel assemblies using nondestructive assay and burnup simulation code. Cm balance is adopted and the concept of "Cm ratio," the mass ratio of Pu to Cm, is used for the nuclear material accountancy for the model pyroprocessing facility. The biggest error of the nuclear material accountancy is expected to arise from the determination of Pu mass and Cm ratio in input homogeneously mixed uranium oxide powder, which is assayed nondestructively. One approach to determine the Pu mass and Cm ratio is to apply the average burnup of spent fuel and determine the Pu mass and Cm ratio by using the ORIGEN code. The estimated error in Pu mass determined by this method ranges from 0.94% to 2.33% for a total of 225 spent fuel assemblies of various burnup, initial enrichment, and cooling time. The other approach is to use the functional relationship between the neutron emission rate and Pu mass of spent fuel. The error in Pu mass calculated using this method ranges from -1.68% to 3.86%.