Study the shielding for PET/CT imaging installations of Shariati hospital of Tehran with Monte Carlo code

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10.22052/2.2.13

Abstract

Nowadays medical imaging systems like PET/CT are widely used in cancer diagnosis. Due to serious and irreversible harms of ionization radiations, protection of all those who are exposed is the main concern of health issues. The main basis of the calculation of the shielding design in the medical imaging systems is that the absorbed dose should not exceed the allowed limit. Using the Monte Carlo N-Particle Transport Code (MCNP), a potential suite for a positron emission tomography/Computed Tomography (PET/CT) facility was designed. The design considers three different areas as required in a PET/CT facility: scanning room, waiting area, and injection room. MCNP-X calculation code is used to ensure that the controlled and uncontrolled areas (with allowed dose of 0.1 mSv/week and 0.02mSv/week, respectively) does not exceed the limits. Since this center accepts four patients every day with mean total activity of 77 mCi, the dose rates are calculated from 2.6E-06 mSv/Week to 9.35E-3 mSv/week. The simulation results revealed that even if this center increase its service to eight patients per day, the dose rates will be from 4.88E-06 mSv/week to 2.81E-02 mSv/week, which is still below the allowed dose rate

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References

[1] Mark T. Madsen et al, AAPM Task Group 108: PET and PET/CT Shielding Requirements, Med. Phys, 33(1), 2006. [2] PET/CT SHIELDING DESIGN COMPARISIONS A Thesis by Audra Lee Coker. Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of requirements for degree of Master of Science May 2007. [3] Powsner, R.A., Powsner, E.R., Essential Nuclear Medicine Physics, Second edition, Blackwell Publishing, 2006. [4] Sureshbabu, W., Mawlawi, O., PET/CT Imaging Artifacts, Nuclear Medicine Technology, 2005, 33:156-161. [5] Herman Cember, Thomas E. Johnson, Health Physics, Fourth Edition, McGraw-Hill Medical Publishing, 2009. [6] Briesmeister, J. F. (1986). \"MCNP-A general Monte Carlo code for neutron and photon transport.\" LA-7396-M. [7] Shultis, J. K. and R. E. Faw (2011). \"An MCNP primer.\" Kansas State University, Manhattan.

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