Validation of Monte Carlo Dose Planning Calculations by Epithermal Beam Dose Distribution Measurements in Phantoms
For any routine clinical application of Boron Neutron Capture Therapy fast and accurate dose calculations will be required for treatment planning. Such calculations are also necessary for the planning and interpretation of results from pre-clinical and clinical trials where the speed of calculation is not so critical. A number of treatment planning systems have been developed1,2. A dose calculation system based on the MCNP3 Monte Carlo Neutron transport code has also been developed by Wallace4. This system takes image data from CT scans and constructs a voxel based geometrical model for input into MCNP. To validate the calculations, a number of phantoms were constructed and exposed in the HB11 epithermal neutron beam at the HFR of the CEC Joint Research Centre in Petten5. The doses recorded by arrays of dosimeters in these phantoms were compared with the calculated results from the MCNP dose planning system. The phantoms used included a cylindrical phantom, a human head phantom and a human torso phantom. The measurements used gold, manganese and copper activation foils as well as P-I-N diode neutron dosimeters6 and MOSFET gamma dosimeters. The overall aim was to show that doses to these detectors placed within the phantoms could be accurately predicted by the dose calculation system.
KeywordsNeutron Energy Boron Neutron Capture Therapy Gold Foil Photon Dose Foil Activation
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- 1.F.J. Wheeler and D.W. Nigg, Three dimensional radiation dose distribution analysis for boron neutron capture therapy, Nuclear Science and Engineering, 110: 16–31, 1992.Google Scholar
- 2.R. Zamenhof, J.Brenner, J. Yanch, D. Wazer, H. Madoc-Jones, S. Saris and O. Harling, Treatment planning for neutron capture therapy of glioblastoma multiforme using an epithermal neutron beam from the MITR-I1 research reactor and Monte Carlo simulation in “Progress in Neutron Capture Therapy for Cancer”, B.J. Allen, D.E. Moore, B.V. Harrington, eds., Plenum Press, New York 1992, pp. 173–178.Google Scholar
- 3.J.F. Briesmeister, ed., MCNP-A General Monte Carlo Code for Neutron and Photon Transport, Version 3A, Los Alamos National Laboratory, LA-7396-M, Rev.2, 1986.Google Scholar
- 4.S.A. Wallace, B.J. Allen, J.N. Mathur, Monte Carlo neutron photon transport calculations: Road to modelling from CT scans, these proceedings.Google Scholar
- 5.G. Constantine, L. Dewit, R.L. Moss, B.J. Mijnheer, K. Ravensberg, F. Stecher-Rasmussen, Designing a treatment room for clinical trials at the HFR, Petten, in “Advances in Neutron Capture Therapy”, A.H. Soloway, R.F. Barth, D.E. Carpenter eds., Plenum Press, New York, 1993, pp. 745–748.CrossRefGoogle Scholar
- 6.M.G Carolan, A.B. Rosenfeld, S. Wallace, H. Meriaty, G.J. Storr, V. I. Khivrich, R.L. Moss, B.J. Allen, Silicon Dosimetric Diode for BNCT using epithermal neutron sources, in “Proceedings of the First International Workshop on Accelerator Based Neutron Sources for Boron Neutron Capture Therapy”, Jackson Wyoming, September 1994.Google Scholar
- 7.Tissue Substitutes in Radiation Dosimetry and Measurement, ICRU Report 44, Bethesda Maryland, 1989.Google Scholar
- 8.Standard practice for characterising neutron energy fluence spectra in terms of an equivalent monoenergetic neutron fluence for radiation hardness testing of electronics, ASTM Report, ASTM E 722–93.Google Scholar