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Division Spotlight
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.
Meeting Spotlight
2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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February 2024
Latest News
Lightbridge announces first U-Zr fuel rod samples extruded at INL
Lightbridge Corporation announced today that it has reached “a critical milestone” in the development of its extruded solid fuel technology. Coupon samples using an alloy of zirconium and depleted uranium—not the high-assay low-enriched uranium (HALEU) that Lightbridge plans to use to manufacture its fuel for the commercial market—were extruded at Idaho National Laboratory’s Materials and Fuels Complex.
Hangbok Choi, Ho Jin Ryu, Gyuhong Roh, Chang Joon Jeong, Chang Je Park, Kee Chan Song, Jung Won Lee, Myung Seung Yang
Nuclear Technology | Volume 157 | Number 1 | January 2007 | Pages 1-17
Technical Paper | Fission Reactors | doi.org/10.13182/NT07-A3798
Articles are hosted by Taylor and Francis Online.
This study describes the mechanical compatibility of the direct use of spent pressurized water reactor fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and fuel handling system in the reactor core by both experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design, which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high-power and high-burnup conditions even though some material properties, such as the thermal conductivity, are a little lower compared to the uranium fuel. However, it is required that the current DUPIC fuel design be changed slightly to accommodate the high internal pressure of the fuel element. It is also strongly recommended that more irradiation tests of the DUPIC fuel be performed to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.