Nolan E. Hertel

Calculation of effective dose from measurements of secondary neutron spectra and scattered photon dose from dynamic MLC IMRT for 6 MV, 15 MV, and 18 MV beam energies

Effective doses were calculated from the delivery of 6 MV, 15 MV, and 18 MV conventional and intensity-modulated radiation therapy (IMRT) prostate treatment plans. ICRP-60 tissue weighting factors were used for the calculations. Photon doses were measured in phantom for all beam energies. Neutron spectra were measured for 15 MV and 18 MV and ICRP-74 quality conversion factors used to calculate ambient dose equivalents. The ambient dose equivalents were corrected for each tissue using neutron depth dose data from the literature. The depth corrected neutron doses were then used as a measure of the neutron component of the ICRP protection quantity, organ equivalent dose. IMRT resulted in an increased photon dose to many organs. However, the IMRT treatments resulted in an overall decrease in effective dose compared to conventional radiotherapy. This decrease correlates to the ability of an intensity-modulated field to minimize dose to critical normal structures in close proximity to the treatment volume. In a comparison of the three beam energies used for the IMRT treatments, 6 MV resulted in the lowest effective dose, while 18 MV resulted in the highest effective dose. This is attributed to the large neutron contribution for 18 MV compared to no neutron contribution for 6 MV.

Investigation of secondary neutron dose for 18 MV dynamic MLC IMRT delivery

Secondary neutron doses from the delivery of 18 MV conventional and intensity modulated radiation therapy (IMRT) treatment plans were compared. IMRT was delivered using dynamic multileaf collimation (MLC). Additional measurements were made with static MLC using a primary collimated field size of 10 x 10 cm2 and MLC field sizes of 0 x 0, 5 x 5, and 10 x 10 cm2. Neutron spectra were measured and effective doses calculated. The IMRT treatment resulted in a higher neutron fluence and higher dose equivalent. These increases were approximately the ratio of the monitor units. The static MLC measurements were compared to Monte Carlo calculations. The actual component dimensions and materials for the Varian Clinac 2100/2300C including the MLC were modeled with MCNPX to compute the neutron fluence due to neutron production in and around the treatment head. There is excellent agreement between the calculated and measured neutron fluence for the collimated field size of 10 x 10 cm2 with the 0 x 0 cm2 MLC field. Most of the neutrons at the detector location for this geometry are directly from the accelerator head with a small contribution from room scatter. Future studies are needed to investigate the effect of different beam energies used in IMRT incorporating the effects of scattered photon dose as well as secondary neutron dose.

Secondary neutron spectra from modern Varian, Siemens, and Elekta linacs with multileaf collimators

Neutrons are a by-product of high-energy x-ray radiation therapy (threshold for [gamma,n] reactions in high-Z material -7 MeV). Neutron production varies depending on photon beam energy as well as on the manufacturer of the accelerator. Neutron production from modern linear accelerators (linacs) has not been extensively compared, particularly in terms of the differences in the strategies that various manufacturers have used to implement multileaf collimators (MLCs) into their linac designs. However, such information is necessary to determine neutron dose equivalents for different linacs and to calculate vault shielding requirements. The purpose of the current study, therefore, was to measure the neutron spectra from the most up-to-date linacs from three manufacturers: Varian 21EX operating at 15, 18, and 20 MV, Siemens ONCOR operating at 15 and 18 MV, and Elekta Precise operating at 15 and 18 MV. Neutron production was measured by means of gold foil activation in Bonner spheres. Based on the measurements, the authors determined neutron spectra and calculated the average energy, total neutron fluence, ambient dose equivalent, and neutron source strength. The shapes of the neutron spectra did not change significantly between accelerators or even as a function of treatment energy. However, the neutron fluence, and therefore the ambient dose equivalent, did vary, increasing with increasing treatment energy. For a given nominal treatment energy, these values were always highest for the Varian linac. The current study thus offers medical physicists extensive information about the neutron production of MLC-equipped linacs currently in operation and provides them information vital for accurate comparison and prediction of neutron dose equivalents and calculation of vault shielding requirements.

Gadolinium neutron capture therapy

A computer study of the dose distribution for gadolinium neutron capture therapy is carried out to determine its feasibility. Gadolinium is a potential neutron capture therapy (NCT) agent that produces gamma radiation, conversion electrons, and Auger electrons. The dose distribution from neutrons, neutron-induced gammas, and the reaction products from neutron capture in gadolinium were modeled using the Los Alamos National Laboratory Monte Carlo neutron photon computer code. The results of these calculations are that gadolinium has promise as an NCT agent. Using two parallel opposed epithermal neutron beams for a tumor at an 8.0-cm depth with a gadolinium loading of 100[mu]g/g, the tumor to peak normal tissue dose was determined to be 1.48.

BUMS—Bonner sphere Unfolding Made Simple: an HTML based multisphere neutron spectrometer unfolding package

Abstract A new multisphere neutron spectrometer unfolding package, Bonner sphere Unfolding Made Simple (BUMS) has been developed that uses an HTML interface to simplify data input and code execution for the novice and the advanced user. This new unfolding package combines the unfolding algorithms contained in other popular unfolding codes under one easy to use interface. The interface makes use of web browsing software to provide a graphical user interface to the unfolding algorithms. BUMS integrates the SPUNIT, BON, MAXIET, and SAND-II unfolding algorithms into a single package. This package also includes a library of 14 response matrices, 58 starting spectra, and 24 dose and detector responses. BUMS has several improvements beyond the addition of unfolding algorithms. It has the ability to search for the most appropriate starting spectra. Also, plots of the unfolded neutron spectra are automatically generated. The BUMS package runs via a web server and may be accessed by any computer with access to the Internet at http://nukeisit.gatech.edu/bums.

Personal dose equivalent conversion coefficients for photons to 1 GeV

The personal dose equivalent, H(p)(d), is the quantity recommended by the International Commission on Radiation Units and Measurements (ICRU) to be used as an approximation of the protection quantity effective dose when performing personal dosemeter calibrations. The personal dose equivalent can be defined for any location and depth within the body. Typically, the location of interest is the trunk, where personal dosemeters are usually worn, and in this instance a suitable approximation is a 30 × 30 × 15 cm(3) slab-type phantom. For this condition, the personal dose equivalent is denoted as H(p,slab)(d) and the depths, d, are taken to be 0.007 cm for non-penetrating and 1 cm for penetrating radiation. In operational radiation protection a third depth, 0.3 cm, is used to approximate the dose to the lens of the eye. A number of conversion coefficients for photons are available for incident energies up to several megaelectronvolts, however, data to higher energies are limited. In this work, conversion coefficients up to 1 GeV have been calculated for H(p,slab)(10) and H(p,slab)(3) both by using the kerma approximation and tracking secondary charged particles. For H(p)(0.07), the conversion coefficients were calculated, but only to 10 MeV due to computational limitations. Additionally, conversions from air kerma to H(p,slab)(d) have been determined and are reported. The conversion coefficients were determined for discrete incident energies, but analytical fits of the coefficients over the energy range are provided. Since the inclusion of air can influence the production of secondary charged particles incident on the face of the phantom, conversion coefficients have been determined both in vacuo and with the source and slab immersed within a sphere in air. The conversion coefficients for the personal dose equivalent are compared with the appropriate protection quantity, calculated according to the recommendations of the latest International Commission on Radiological Protection (ICRP) guidance.

Effects of tertiary MLC configuration on secondary neutron spectra from 18 MV x-ray beams for the Varian 21EX linear accelerator

The effect of the jaw configuration and the presence and configuration of the tertiary multileaf collimator (MLC) on the secondary neutron spectra for an 18 MV Varian 21EX linear accelerator (linac) is investigated in detail. The authors report the measured spectra for four collimator (jaw-and-MLC) configurations. These configurations represent the extreme settings of the jaws and MLC and should therefore describe the range of possible fluence and spectra that may be encountered during use of this linac. In addition to measurements, a Monte Carlo model was used to simulate the four collimator configurations and calculate the energy spectra and fluence at the same location as it was measured. The Monte Carlo model was also used to calculate the sources of neutron production in the linac head for each collimator configuration. They found that photoneutron production in the linac treatment head is dominated by the order in which the primary photon beam intercepts the high-Z material. The primary collimator, which has the highest position in the linac head (in a fixed location), is the largest source of secondary neutrons. Thereafter, the collimator configuration plays a role in where the neutrons originate. For instance, if the jaws are closed, they intercept the beam and contribute substantially to the secondary neutron production. Conversely, if the jaws are open, the MLC plays a larger role in neutron production (assuming, of course, that it intercepts the beam). They found that different collimator configurations make up to a factor of 2 difference in the ambient dose equivalent.

Multisphere neutron spectra measurements near a high energy medical accelerator

Photoneutron spectral measurements in the vicinity of a high energy medical accelerator have been performed. A LiI (Eu) scintillator and paired thermoluminescent dosimeters were used in conjunction with moderating multispheres to measure the neutron spectra. The photoneutron fluence rate, fluence lethargy, average energy, and quality factors have been calculated using a recursive unfolding algorithm. The dose rates and dose equivalent rates received by the patient and operators due to these neutrons have been estimated from these unfolded spectra. Doses predicted by a semi-empirical dose-equivalent model were compared to the measured values at each point. The dose equivalent rate to operators was found to be 1.9 x 10(-11) Sv MU(-1).

Transmission of Fast Neutrons Through an Iron Sphere

Integral Experiments have been performed using a homogeneous iron spherical shell to test neutron cross-section data. Neutron leakage spectra from the shell were measured using /sup 252/Cf-fission and (deuterium-tritium) D-T-fusion neutron sources and an Ne-213 spectrometry system. An associated particle detector was used to monitor the absolute D-T neutron source strength as well as any accompanying deuterium-deuterium neutron contamination. The leakage spectra were calculated using the continuous-energy Monte Carlo code VIM and the discrete ordinates S/sub n/ code ANISN employing ENDF/B-IV. For neutron energies between 1 and 5 MeV, the calculations underpredicted the leakage spectrum by factors of 1.4 to 2 for the Californium neutron source and 2 to 3 for the D-T neutron source. The large discrepancies are attributed to inadequate representation of cross-section resonance structure (namely, minima); inadequate representation of the angular and secondary energy distributions for continuum inelastic scattering and (n,2n) reactions also contribute to these discrepancies.

Polyethylene/Boron Composites for Radiation Shielding Applications

Multifunctional composites made with boron are absorbers of low energy nuetrons, and could be used for structural shielding materials. Polyethylene/boron carbide composites were fabricated using conventional polymer processing techniques, and were evaluated for mechanical and radiation shielding properties. Addition of neat boron carbide (powder and nanoparticles) to an injection molding grade HPDE showed superior mechanical properties compared to neat HDPE. Radiation shielding measurements of a 2 wt% boron carbide composite were improved over those of the neat polyethylene.