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Nuclear Installations Safety
Devoted specifically to the safety of nuclear installations and the health and safety of the public, this division seeks a better understanding of the role of safety in the design, construction and operation of nuclear installation facilities. The division also promotes engineering and scientific technology advancement associated with the safety of such facilities.
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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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Commercial nuclear innovation "new space" age
In early 2006, a start-up company launched a small rocket from a tiny island in the Pacific. It exploded, showering the island with debris. A year later, a second launch attempt sent a rocket to space but failed to make orbit, burning up in the atmosphere. Another year brought a third attempt—and a third failure. The following month, in September 2008, the company used the last of its funds to launch a fourth rocket. It reached orbit, making history as the first privately funded liquid-fueled rocket to do so.
Kun Mo, Hsiao-Ming Tung, Xiang Chen, Yang Zhao, Jon Hansen, James F. Stubbins
Nuclear Technology | Volume 183 | Number 3 | September 2013 | Pages 455-463
Technical Paper | Materials for Nuclear Systems | doi.org/10.13182/NT13-A19433
Articles are hosted by Taylor and Francis Online.
Both Alloy 617 and Alloy 230 have been considered the most promising structural materials for the Very High Temperature Reactor (VHTR). In this study, mechanical properties of both alloys were examined by performing tensile tests at three different strain rates and at temperatures up to 1000°C. This range covers time-dependent (plasticity) to time-independent (creep) deformations. Strain-rate sensitivity analysis for each alloy was conducted in order to approximate the long-term flow stresses. The strain-rate sensitivities for the 0.2% flow stress were found to be temperature independent (m [approximate] 0) at temperatures ranging from room temperature to 700°C due to dynamic strain aging. At elevated temperatures (800°C to 1000°C), the strain-rate sensitivity significantly increased (m > 0.1). Compared to Alloy 617, Alloy 230 displayed higher strain-rate sensitivities at high temperatures. This leads to lower estimated long-term flow stresses. Results of this analysis were used to evaluate the current American Society of Mechanical Engineers (ASME) allowable design limits for each alloy. The study showed that the allowable design stresses in the ASME Boiler and Pressure Vessel Code for either alloy do not provide adequate long-term degradation estimation. Nevertheless, rupture stresses for Alloy 617, developed in the ASME code case N-47-28, can generally satisfy the safety margins at 800°C and 1000°C estimated in the study following the strain-rate sensitivity analysis. Furthermore, additional material development studies might be required, since the design parameters for rupture stresses are constrained such that the current VHTR conceptual designs cannot satisfy the material limits.