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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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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.
Constantine P. Tzanos, B. Dionne
Nuclear Technology | Volume 179 | Number 3 | September 2012 | Pages 382-391
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT12-A14170
Articles are hosted by Taylor and Francis Online.
To support the safety analysis of the conversion of the BR2 research reactor to low-enriched uranium (LEU) fuel and extend the validation basis of the RELAP code for the analysis of the conversion of research reactors from highly enriched (HEU) fuel to LEU, the simulation of BR2 tests A/400/1, C/600/3, and F/400/1 was undertaken. These tests are characterized by loss of flow initiated at different reactor power levels with or without loss of system pressure, reactor scram, flow reversal, and reactor cooling by natural circulation. This work presents the RELAP analysis of tests C/600/3 and F/400/1 and comparison of code predictions with experimental measurements for peak cladding temperatures during the transient at different axial locations in an instrumented fuel assembly. The simulations show that accurate representation of the power distribution, especially after reactor scram, between the fuel assemblies and the moderator/reflector regions is critical for the correct prediction of the peak cladding temperatures during the transient. Detailed MCNP and ORIGEN simulations were performed to compute the power distribution between the fuel assemblies and the moderator/reflector regions. With these distributions the predicted peak cladding temperatures are in good agreement with experimental measurements.
