Neutron Flux Measurement in TRIGA Reactor using Gold Foil Activation Method
Tornike Kimeridze, Tran Canh Hai Nguyen
Abstract
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Tornike Kimeridze, Tran Canh Hai Nguyen
Abstract
Open-access reader
Neutron flux measurement is integral to nuclear physics, with implications across nuclear engineering, medical physics, and material science. This study details neutron flux mapping in the TRIGA reactor in Vienna using the gold foil activation method. Identical gold foils were prepared, subjected to 10-minute irradiation within the reactor core, and subsequently assessed using a 4πβ counter. The neutron flux distribution was theoretically modeled as a Bessel function and cross-referenced against measured activities to yield an experimental thermal neutron flux map. Findings reveal a close alignment between theoretical and experimental data, with a slight flux increase at the core's outermost periphery due to the graphite reflector. This work contributes valuable data on the TRIGA reactor's neutron flux distribution, with broader applicability in the field of reactor experiments and operation.
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Neutron flux measurement is integral to nuclear physics, with implications across nuclear engineering, medical physics, and material science. This study details neutron flux mapping in the TRIGA reactor in Vienna using the gold foil activation method. Identical gold foils were prepared, subjected to 10-minute irradiation within the reactor core, and subsequently assessed using a 4πβ counter. The neutron flux distribution was theoretically modeled as a Bessel function and cross-referenced against measured activities to yield an experimental thermal neutron flux map. Findings reveal a close alignment between theoretical and experimental data, with a slight flux increase at the core's outermost periphery due to the graphite reflector. This work contributes valuable data on the TRIGA reactor's neutron flux distribution, with broader applicability in the field of reactor experiments and operation.
Key concepts: TRIGA, Neutron flux, Research reactor, Neutron cross section, Nuclear physics, Neutron, Neutron temperature, Flux (metallurgy)