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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.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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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Nuclear News 40 Under 40 discuss the future of nuclear
Seven members of the inaugural Nuclear News 40 Under 40 came together on March 4 to discuss the current state of nuclear energy and what the future might hold for science, industry, and the public in terms of nuclear development.
To hear more insights from this talented group of young professionals, watch the “40 Under 40 Roundtable: Perspectives from Nuclear’s Rising Stars” on the ANS website.
J. E. Houghtaling, J. E. Grund
Nuclear Science and Engineering | Volume 36 | Number 3 | June 1969 | Pages 412-426
Technical Paper | doi.org/10.13182/NSE69-A18738
Articles are hosted by Taylor and Francis Online.
Reactor kinetics investigations have been performed for cold-start-up, hot-start-up, hot-standby, and operating-power reactivity accidents using the UO2-fueled, pressurized-water type SPERT-III reactor. Power excursion behavior was predicted for every SPERT-III experiment by digital computer calculations using the SPERT-developed PARET code. Extrapolations for severe cold-start-up excursion consequences were obtained from severe transient tests on SPERT-III fuel samples in the SPERT-IV capsule driver core. Analyses of the SPERT-III data show that prompt moderator heating was as significant as the Doppler effect in limiting the magnitude of power excursions in the SPERT-III core at operating temperatures. Comparisons of calculations and experimental data demonstrate that PARET is capable of predicting power excursion behavior in SPERT-III within experimental uncertainty for the range of conditions investigated. The SPERT-III integral-core tests also provide a broad base of experimental data for demonstrations of the capabilities of other existing models in predicting non-damaging power excursion behavior in UO2-fueled reactors.