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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
Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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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Reviewers needed for NRC research proposals
The deadline is fast approaching for submitting an application to become a technical reviewer for the Nuclear Regulatory Commission’s fiscal year 2025 research grant proposals.
Kwang J. Jeong, Joon Lim, Il S. Hwang, Hee D. Kim, Martin M. Pilch, Tze Y. Chu
Nuclear Technology | Volume 143 | Number 3 | September 2003 | Pages 347-357
Technical Paper | Materials for Nuclear Systems | doi.org/10.13182/NT03-A3422
Articles are hosted by Taylor and Francis Online.
High-temperature creep tests were performed with an SA533B1 low-alloy steel under both constant load and constant stress conditions. Using the measured minimum creep strain rates as a function of stress and temperature, least-square fittings were made into a Bailey-Norton-type power law equation. Based on the constant stress test results, a constitutive equation was developed for steady-state creep. The constitutive equation was then implemented in elastic-viscoplastic analysis of the lower head of a pressurized water reactor's reactor pressure vessel using a commercial FEM code named ABAQUS 5.8. The FEM model was validated using measured data from the lower head failure experiment conducted at the Sandia National Laboratories. The FEM model using the creep constitutive equation was shown to be capable of accurately predicting the lower head deformation behavior. Additional work, however, is needed to rationalize apparent inconsistency between the constant load data and constant stress data.