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Radiation Protection Dosimetry Advance Access originally published online on March 23, 2005
Radiation Protection Dosimetry 2005 113(4):442-448; doi:10.1093/rpd/nch477
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© The Author 2005. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oupjournals.org

Technical Note

Response of Harshaw neutron thermoluminescence dosemeters in terms of the revised ICRP/ICRU recommendations

K. G. Veinot1,* and N. E. Hertel2

1 Y-12 National Security Complex, P.O. Box 2009, M.S. 8105, Oak Ridge, TN 37831-8105, USA
2 George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332-0405, USA

* Corresponding author: veinotkg{at}y12.doe.gov

Received August 6, 2004, amended February 14, 2005, accepted February 22, 2005

To monitor workers for external neutron radiation dose, the Y-12 National Security Complex utilises the thermoluminescence dosemeters (TLDs) manufactured by Harshaw. At Y-12, the majority of external dose to workers is due to low-energy photon and/or beta particles emitted from uranium and its progeny. However, some neutron dose is expected since neutrons are produced from ({alpha},n) reactions in various compounds found at the plant, including UF4 and UF6. Neutron sources, such as 252Cf, are also used throughout the complex. The Harshaw neutron dosemeter consists of two gamma-sensitive elements (7Li) and two neutron-sensitive elements enriched in 6Li with various shielding/filter materials placed around each of them. In this work, the energy response of the dosemeter to neutrons has been calculated using the Monte Carlo transport code MCNP Version 4-C and, these results are compared with the measured response of the dosemeter to unmoderated and D2O-moderated 252Cf neutrons. The response of the dosemeter has also been determined in terms of the personal absorbed dose and personal dose equivalent as a function of neutron energy based on the recommendations of the ICRP Publication 60 and ICRU Report 49. The energy response of the dosemeter characteristics can be used to generate spectral conversion coefficients for routine neutron absorbed dose and dose equivalent calculations.


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