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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
Lightbridge announces first U-Zr fuel rod samples extruded at INL
Lightbridge Corporation announced today that it has reached “a critical milestone” in the development of its extruded solid fuel technology. Coupon samples using an alloy of zirconium and depleted uranium—not the high-assay low-enriched uranium (HALEU) that Lightbridge plans to use to manufacture its fuel for the commercial market—were extruded at Idaho National Laboratory’s Materials and Fuels Complex.
S. Bourganel, O. Petit, C. M. Diop
Nuclear Technology | Volume 184 | Number 1 | October 2013 | Pages 29-41
Technical Paper | Neutron Transport and Shielding | doi.org/10.13182/NT13-A19866
Articles are hosted by Taylor and Francis Online.
The Électricité de France nuclear park consists of 58 pressurized water nuclear reactors. To ensure their good performance and safety, ex-core neutron shielding studies are regularly performed. For example, neutron flux calculations in ex-core ionization chambers and pressure vessel neutron fluence studies are carried out. In the first case, ex-core ionization chambers are neutron detectors located in the reactor pit, around the reactor vessel. They are dedicated to reactor operation and core protection. In the second case, the calculation of the fast fluence (for energy >1 MeV) in the pressure vessel is used to determine its fracture toughness and integrity. To improve the fluence computations, new efficient parametric methods are under development. For these two problems, Monte Carlo transport codes such as TRIPOLI-4® allow us to perform simulations in realistic complex three-dimensional geometries and to produce reference results.The aim of the present paper is to present together the theoretical background of our approach based on the continuous-energy Green's functions computation and storage to perform both vessel neutron fluence and ex-core chamber responses. The normalized source contribution or importance factor formalism using Green's functions computation is also described, with its associated statistical uncertainty calculation. Application examples to realistic nuclear plant configurations are given.