Institutional members access full text with Ovid®

Share this article on:

Use of Transportable Radiation Detection Instruments to Assess Internal Contamination From Intakes of Radionuclides Part I: Field Tests and Monte Carlo Simulations

Anigstein, Robert; Erdman, Michael C.; Ansari, Armin

doi: 10.1097/HP.0000000000000496
Papers

Abstract: The detonation of a radiological dispersion device or other radiological incidents could result in the dispersion of radioactive materials and intakes of radionuclides by affected individuals. Transportable radiation monitoring instruments could be used to measure photon radiation from radionuclides in the body for triaging individuals and assigning priorities to their bioassay samples for further assessments. Computer simulations and experimental measurements are required for these instruments to be used for assessing intakes of radionuclides. Count rates from calibrated sources of 60Co, 137Cs, and 241Am were measured on three instruments: a survey meter containing a 2.54 × 2.54‐cm NaI(Tl) crystal, a thyroid probe using a 5.08 × 5.08‐cm NaI(Tl) crystal, and a portal monitor incorporating two 3.81 × 7.62 × 182.9‐cm polyvinyltoluene plastic scintillators. Computer models of the instruments and of the calibration sources were constructed, using engineering drawings and other data provided by the manufacturers. Count rates on the instruments were simulated using the Monte Carlo radiation transport code MCNPX. The computer simulations were within 16% of the measured count rates for all 20 measurements without using empirical radionuclide-dependent scaling factors, as reported by others. The weighted root-mean-square deviations (differences between measured and simulated count rates, added in quadrature and weighted by the variance of the difference) were 10.9% for the survey meter, 4.2% for the thyroid probe, and 0.9% for the portal monitor. These results validate earlier MCNPX models of these instruments that were used to develop calibration factors that enable these instruments to be used for assessing intakes and committed doses from several gamma-emitting radionuclides.

*S. Cohen & Associates, 1608 Spring Hill Road, Vienna, VA 2218; †Department of Radiology, Milton S. Hershey Medical Center, Hershey, PA 17033‐2390; ‡Radiation Studies Branch, EHHE, NCEH, Centers for Disease Control and Prevention, Atlanta, GA 30341‐3717.

The authors declare no conflicts of interest.

For correspondence contact: Robert Anigstein, PhD, S. Cohen & Associates, 740 West End Avenue, Apt. 95A, New York, NY 10025, or email at anigstein@cs.com.

(Manuscript accepted 4 January 2016)

© 2016 by the Health Physics Society