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The Sensitivity of Gamma Portal Monitors to Personnel Intakes of Radioactivity

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Abstract

Gamma radiation portal monitors are commonly used as an element of a radiation protection program at nuclear facilities. While they are are primarily used for personnel monitoring of external contamination, they also have the ability to detect internally deposited radioisotopes.

For nuclear power sites, the radioactive source term is generally composed of multiple nuclides with different activity levels. If the composition and relative fractions of the nuclide mix are known, then a minimum detectable committed effective dose equivalent (MDCEDE) for internal contaminations can be calculated if the gamma efficiency of the monitor is also known. This evaluation shows that the MDCEDE can vary over several decades depending on the radionuclide distributions for three nuclear power sites: an operating PWR, a decayed BWR, and a decayed CANDU.

For these radionuclide distributions, the acute inhalation MDCEDE are approximately .07 rem (0.7 mSv), 0.225 rem (2.3 mSv), and 20 rem (200 mSv), respectively. In the case of the decayed CANDU reactor, the facility had experienced significant fuel leakage during its operating history that caused a substantially higher fraction of transuranic activity to be distributed as compared to the other sites. Understanding the impact of a facility’s radionuclide distribution to the portal monitor’s sensitivity to internally deposited radioactivity can be an important element in understanding and bounding potentially unmonitored internal exposures.

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