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Human Factors, Instrumentation & Controls
Improving task performance, system reliability, system and personnel safety, efficiency, and effectiveness are the division's main objectives. Its major areas of interest include task design, procedures, training, instrument and control layout and placement, stress control, anthropometrics, psychological input, and motivation.
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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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Latest News
College students help develop waste-measuring device at Hanford
A partnership between Washington River Protection Solutions (WRPS) and Washington State University has resulted in the development of a device to measure radioactive and chemical tank waste at the Hanford Site. WRPS is the contractor at Hanford for the Department of Energy’s Office of Environmental Management.
Constantine P. Tzanos, B. Dionne
Nuclear Technology | Volume 176 | Number 1 | October 2011 | Pages 93-105
Thermal Hydraulics | doi.org/10.13182/NT11-A12545
Articles are hosted by Taylor and Francis Online.
The simulation of the BR2 test A/400/1 was undertaken to support the safety analysis of the conversion of the BR2 research reactor to low-enriched uranium (LEU) fuel and to extend the validation basis of the RELAP code for analysis of the conversion of research reactors from highly enriched fuel to LEU. This test was characterized by a steady-state peak heat flux of 400 W/cm2 , total loss of flow without loss of system pressure, reactor scram, flow reversal, and reactor cooling by natural convection. This paper presents the RELAP analysis of test A/400/1 and the comparison of code predictions with experimental measurements of peak cladding temperatures during the transient at different axial locations in an instrumented fuel assembly. The simulations show that accurate representation of the pump coastdown characteristics and of the power distribution, especially after reactor scram, between the fuel assemblies and the moderator/reflector regions are critical for 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 were in a good agreement with experimental measurements.