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Introduction

THE USE OF MCNP CODE FOR RADIATION TRANSPORT AND DOSIMETRY CALCULATIONS IN TRAINING MEDICAL PHYSICS STUDENTS Abdullah Al Kafi, Nabil Maalej, Akhtar Abbas Naqvi Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261, Saudi Arabia. Introduction.

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Introduction

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  1. THE USE OF MCNP CODE FOR RADIATION TRANSPORT AND DOSIMETRY CALCULATIONS IN TRAINING MEDICAL PHYSICS STUDENTSAbdullah Al Kafi, Nabil Maalej, Akhtar Abbas NaqviPhysics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261, Saudi Arabia

  2. Introduction • Monte Carlo simulation is a stochastic technique that uses random numbers and probability statistics to obtain an answer • First Developed for the Manhattan Project at the Los Alamos National laboratory during World War II. • Used for particle transport and interaction with matter as well as radiation protection and dosimetry • We use Monte Carlo code MCNP 4B/C for our class projects

  3. MCNP Code • MCNP is a general purpose Monte Carlo n-particle transport code used for neutron, photon and electron transport • Neutron energy (l0 MeV to 20 MeV) • Photon and electron energies (1 keV to 1 GeV).

  4. MCNP Setup Input file : • Geometry specification • Materials selection and properties • The location and characteristics of the neutron, photon, or electron source • Output desired (tallies ) • Any variance reduction techniques used to improve efficiency

  5. MCNP Class Project I • Radiological Physics and Dosimetry Course (MEPH 561) • Class Project: Dose distributions in different geometries that simulate human chest, head and leg

  6. Rectangular Geometry to Simulate the Chest

  7. Spherical Geometry to Simulate Head

  8. Cylindrical Geometry to Simulate Thigh

  9. Theoretical Calculations Under charged particle equilibrium, the absorbed dose (D) to a medium can be calculated from the energy fluence  and the mass energy absorption coefficient ; For an ideal broad-beam geometry, the radiant energy of uncharged particles striking the detector through the attenuator at depth x is: If a source emits N photons isotropically, the energy flounce at a distance r from the source and at a depth x in the attenuator; Then the absorbed dose can be approximated by:

  10. Depth-Dose Distribution for Chest Simulation

  11. Depth-Dose Distribution for Head Simulation

  12. Depth-Dose Distribution for Leg Simulation

  13. Percentage Difference Between MCNP and Broad Beam Geometry as a Function of Depth

  14. MCNP Class Project II • Radiotherapy Physics (MEPH 566) • Class Project: Percent Depth Dose (PDD) in a water phantom due to the photons emitted from a cobalt-60 source

  15. Geometry for Calculating Percent Depth Dose(Not True scale) Source 10 cm 5 cm Lead Collimator 80 cm 10 cm 30 cm Water Phantom 30 cm

  16. Percent Depth Dose (PDD): Percent depth dose (PDD): Percentage of the absorbed dose at any depth d to the maximum absorbed dose at reference depth dmax, along the central axis of the beam:

  17. Percent Depth Dose Curve for Co-60 Spectrum.

  18. Summary of Observations • Variations are observed in percent depth dose curve between MCNP and published data at higher depths • These variations are due to the statistical error in calculating the dose from the small number of particles reaching the detector at higher depths

  19. Conclusion • We used Monte Carlo Simulation To simulate Particle Transport and interaction with matter in 2 Medical Physics class projects • Students learned MCNP a very powerful tool for research and development • We are using MCNP to optimize grid design in mammography

  20. Thank you

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