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2026 Annual Conference
May 31–June 3, 2026
Denver, CO|Sheraton Denver
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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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Latest News
Seconds Matter: Rethinking Nuclear Facility Security for the Modern Threat Landscape
In today’s rapidly evolving threat environment, nuclear facilities must prioritize speed and precision in their security responses—because in critical moments, every second counts. An early warning system serves as a vital layer of defense, enabling real-time detection of potential intrusions or anomalies before they escalate into full-blown incidents. By providing immediate alerts and actionable intelligence, these systems empower security personnel to respond decisively, minimizing risk to infrastructure, personnel, and the public. The ability to anticipate and intercept threats at the earliest possible stage not only enhances operational resilience but also reinforces public trust in the safety of nuclear operations. Investing in such proactive technologies is no longer optional—it’s essential for modern nuclear security.
Andrew Richard Raymond Telford
Nuclear Technology | Volume 56 | Number 1 | January 1982 | Pages 33-39
Technical Paper | Fission Reactor | doi.org/10.13182/NT82-A32878
Articles are hosted by Taylor and Francis Online.
Tests have been carried out on one of the advanced gas-cooled reactors (AGRs) at Hinkley Point to determine the fuel temperature coefficient of reactivity, an important safety-related parameter. Reactor neutron flux was measured during transients induced by movement of a bank of control rods from one steady position to another. An inverse kinetics analysis was applied to the recorded flux transient to determine the reactivity change as the fuel temperature changed, and the variation of mean fuel temperature was derived from the flux transient by a multiplane thermal-hydraulics code representing an AGR fuel channel The fuel temperature coefficient was then obtained from the slope of a plot of core reactivity against fuel temperature. The uncertainty to be applied to the derived temperature coefficient has been shown to be approximately ±10% at the one standard deviation level The experimental technique has been found to be simple to apply on a commercial reactor and has given consistent results over a range of reactor operating conditions. Calculations of fuel temperature coefficients of reactivity (based on the lattice code, ARGOSY) have been carried out and reactor averaged values deduced for comparison with experiment. The calculated and measured coefficients agree to within one standard deviation over a range of core irradiations and power levels.
