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DOE OSTI · 3030618

AGC-4 As-Run Uncertainty Analysis

Holschuh, Tommy [Idaho National Laboratory] (ORCID:0000000215214314)·Quist, Teancum Earl [Idaho National Laboratory]·Windes, William E [Idaho National Laboratory] (ORCID:0000000188030885)
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Abstract

The concept of measuring neutron flux (or fluence rate) with activation foils has existed since the late 1940s and is a well-used methodology. Activation foils are selected for their unique isotopic cross sections that create reaction products with desirable properties: relatively simple radioactive decay scheme, measurable quantities of gamma-rays or betas, and half-lives appropriate for the irradiation and decay durations of the experiment. The activation materials, deemed flux wires, fluence wires, dosimeters, or flux foils, are fabricated to have a mass appropriate for the irradiation as well so the measurement using a radiation detector following the irradiation is sufficient but not overwhelming in terms of total activity. Activation measurement results are expressed in terms of reaction rate, which can then be used to “adjust” the neutron spectrum. The Advanced Graphite Creep (AGC)-4 experiment contained 26 flux wire packages, each containing an iron and a niobium wire inside a vanadium capsule. For all stackup positions, the value of fast to thermal neutrons is higher than expected, but the total neutron fluence is near the predicted value of ~2 x 1022 n/cm2. As compared to the predicted neutron spectrum, there are four takeaways in the results as compared to the predicted ATR irradiation: 1) Additional fast (fission) neutrons present, 2) Spectrum hardening due to temperature, 3) Spectrum hardening due to container attenuation, 4) Total neutron fluence is close to predicted values. Since the neutron energy spectrum is a key parameter when attributing neutron damage in materials, the adjusted energy-dependent spectrum and the total neutron fluence results from the dosimeters can be incorporated for future calculations.

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11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS22 - GENERAL STUDIES OF NUCLEAR REACTORS36 - MATERIALS SCIENCEActivation FoilsAGC-4Flux WiresReactor Dosimetry

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Holschuh, Tommy [Idaho National Laboratory] (ORCID:0000000215214314), Quist, Teancum Earl [Idaho National Laboratory], Windes, William E [Idaho National Laboratory] (ORCID:0000000188030885). 2025-07-08. AGC-4 As-Run Uncertainty Analysis. https://www.osti.gov/biblio/3030618

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