More realistic Monte Carlo calculations of photon detector response functions

Direct energy spectrum measurement of X-ray from a clinical linac

A realistic X‐ray energy spectrum is essential for accurate dose calculation using the Monte Carlo (MC) algorithm. An energy spectrum for dose calculation in the radiation treatment planning system is modeled using the MC algorithm and adjusted to obtain acceptable agreement with the measured percent depth dose (PDD) and off‐axis ratio. The simulated energy spectrum may not consistently reproduce a realistic energy spectrum. Therefore, direct measurement of the X‐ray energy spectrum from a linac is necessary to obtain a realistic spectrum. Previous studies have measured low photon fluence directly, but the measurement was performed with a nonclinical linac with a thick target and a long target‐to‐detector distance. In this study, an X‐ray energy spectrum from a clinical linac was directly measured using a NaI(Tl) scintillator at an ultralow dose rate achieved by adjusting the gun grid voltage. The measured energy spectrum was unfolded by the Gold algorithm and compared with a simulated spectrum using statistical tests. Furthermore, the PDD was calculated using an unfolded energy spectrum and a simulated energy spectrum was compared with the measured PDD to evaluate the validity of the unfolded energy spectrum. Consequently, there was no significant difference between the unfolded and simulated energy spectra by nonparametric, Wilcoxon's rank‐sum, chi‐square, and two‐sample Kolmogorov–Smirnov tests with a significance level of 0.05. However, the PDD calculated from the unfolded energy spectrum better agreed with the measured compared to the calculated PDD results from the simulated energy spectrum. The adjustment of the incident electron parameters using MC simulation is sensitive and takes time. Therefore, it is desirable to obtain the energy spectrum by direct measurement. Thus, a method to obtain the realistic energy spectrum by direct measurement was proposed in this study.

THE USE OF DECONVOLUTION TECHNIQUE FOR THE ANALYSIS OF GAMMA SPECTROMETRY DATA FROM FIELD MONITORING USING UNMANNED AERIAL VEHICLES

Airborne gamma spectrometry is an effective tool for prompt monitoring and mapping of large areas contaminated after NPP accident, radionuclides leakage cases, an impact of uranium ore mining and processing, etc. Airborne spectrometry data analysis using deconvolution technique enables to calculate air kerma rates and/or radionuclides concentrations as well as identification of radionuclides. Application of this technique on the airborne data (from manned as well as an unmanned survey using drones) is rather specific due to the requirements for short time of one scan data acquisition, a relatively long distance from the source and small detector size, due to the limited payload of the usually used drones. Application of deconvolution techniques for analysis of spectra with very poor statistics, methods and possibilities to improve the processing of such spectra are discussed.

Longitudinal monitoring of whole body counter NaI(Tl) detector efficiency

Assessing accuracy of radiation counting systems over time is critical. We examined long-term WBC performance in detail. Efficiency factors for 54 detectors were updated annually over several years. Newer efficiency values were compared with baseline and with annual values. Overall system efficiency has declined (-1.9% over 3 yrs) and appears to be doing so at an increasing rate. Having more specific performance data on individual components can make the process of system maintenance and repair more straightforward and efficient.

Establishment of 6- to 7-MeV high-energy gamma-ray calibration fields produced using the 4-MV Van de Graaff accelerator at the Facility of Radiation Standards, Japan Atomic Energy Agency

A 6- to 7-MeV high-energy gamma-ray field, produced by the nuclear reaction of (19)F(p, αγ)(16)O, has been established at the Facility of Radiation Standards (FRS) in Japan Atomic Energy Agency for calibration purposes. Basic dosimetric quantities (i.e. averaged gamma-ray energy, air-kerma-to-dose equivalent conversion coefficients and air kerma rates at the point of test) have been precisely determined through a series of measurements using the NaI(Tl) spectrometer and an ionisation chamber coupled with an appropriate build-up material. The measurements obtained comply with values recommended by the International Organization for Standardization for an 'R-F field'. The neutron contamination component for the field has also been measured by means of a conventional neutron dose equivalent meter (the so-called neutron rem-counter) and determined to be ∼ 0.5 % of the total dose equivalent.

Measurement of air kerma rates for 6- to 7-MeV high-energy gamma-ray field by ionisation chamber and build-up plate

The 6- to 7-MeV high-energy gamma-ray calibration field by the (19)F(p, αγ)(16)O reaction is to be served at the Japan Atomic Energy Agency. For the determination of air kerma rates using an ionisation chamber in the 6- to 7-MeV high-energy gamma-ray field, the establishment of the charged particle equilibrium must be achieved during measurement. In addition to measurement of air kerma rates by the ionisation chamber with a thick build-up cap, measurement using the ionisation chamber and a build-up plate (BUP) was attempted, in order to directly determine air kerma rates under the condition of regular calibration for ordinary survey meters and personal dosemeters. Before measurements, Monte Carlo calculations were made to find the optimum arrangement of BUP in front of the ionisation chamber so that the charged particle equilibrium could be well established. Measured results imply that air kerma rates for the 6- to 7-MeV high-energy gamma-ray field could be directly determined under the appropriate condition using an ionisation chamber coupled with build-up materials.

Lithium target performance evaluation for low-energy accelerator-based in vivo measurements using gamma spectroscopy

The operating conditions at McMaster KN Van de Graaf accelerator have been optimized to produce neutrons via the (7)Li(p, n)(7)Be reaction for in vivo neutron activation analysis. In a number of earlier studies (development of an accelerator based system for in vivo neutron activation analysis measurements of manganese in humans, Ph.D. Thesis, McMaster University, Hamilton, ON, Canada; Appl. Radiat. Isot. 53 (2000) 657; in vivo measurement of some trace elements in human Bone, Ph.D. Thesis. McMaster University, Hamilton, ON, Canada), a significant discrepancy between the experimental and the calculated neutron doses has been pointed out. The hypotheses formulated in the above references to explain the deviation of the experimental results from analytical calculations, have been tested experimentally. The performance of the lithium target for neutron production has been evaluated by measuring the (7)Be activity produced as a result of (p, n) interaction with (7)Li. In contradiction to the formulated hypotheses, lithium target performance was found to be mainly affected by inefficient target cooling and the presence of oxides layer on target surface. An appropriate choice of these parameters resulted in neutron yields same as predicated by analytical calculations.

Relevance of accurate Monte Carlo modeling in nuclear medical imaging

Monte Carlo techniques have become popular in different areas of medical physics with advantage of powerful computing systems. In particular, they have been extensively applied to simulate processes involving random behavior and to quantify physical parameters that are difficult or even impossible to calculate by experimental measurements. Recent nuclear medical imaging innovations such as single-photon emission computed tomography (SPECT), positron emission tomography (PET), and multiple emission tomography (MET) are ideal for Monte Carlo modeling techniques because of the stochastic nature of radiation emission, transport and detection processes. Factors which have contributed to the wider use include improved models of radiation transport processes, the practicality of application with the development of acceleration schemes and the improved speed of computers. In this paper we present a derivation and methodological basis for this approach and critically review their areas of application in nuclear imaging. An overview of existing simulation programs is provided and illustrated with examples of some useful features of such sophisticated tools in connection with common computing facilities and more powerful multiple-processor parallel processing systems. Current and future trends in the field are also discussed.

Assessment of response function in two PET scanners with and without interplane septa

The authors have assessed the response function both experimentally and theoretically for two commercial tomographs: CTI 931/08-12 and CTI 953B with and without interplane septa. Monte Carlo simulations were undertaken using the GEANT package from CERN. Spatial resolution (tomographic and axial) was calculated for line sources at various positions in the field of view. Sensitivity and scatter fraction (SF) were calculated for various source geometries as a function of energy discrimination. A very realistic response function in positron emission tomography (PET) is obtained by Monte Carlo methods, using global parameters to account for unsimulated phenomena such as scintillation light transport inside a detector block and its sharing among the various phototubes. Minor discrepancies remain for sensitivity and SF at high energy thresholds and may probably be explained by introducing the observed dispersion in the energy response for the various crystals within a detector block.

Monte Carlo and convolution dosimetry for stereotactic radiosurgery

The dosimetry of small photon beams used for stereotactic radiosurgery was investigated using Monte Carlo simulation, convolution calculations, and measurements. A Monte Carlo code was used to simulate radiation transport through a linear accelerator to produce and score energy spectrum and angular distribution of 6 MV bremsstrahlung photons exiting from the accelerator treatment head. These photons were then transported through a stereotactic collimator system and into a water phantom placed at isocenter. The energy spectrum was also used as input for the convolution method of photon dose calculation. Monte Carlo and convolution results were compared with the measured data obtained using an ionization chamber, a diode, and film.

Generation of photon energy deposition kernels using the EGS Monte Carlo code

The EGS Monte Carlo code was used to generate photon energy deposition kernels which describe the energy deposited by charged particles set in motion by primary, first scattered, second scattered, multiple scattered and bremsstrahlung plus annihilation photons. These were calculated for a water medium irradiated with monoenergetic photons with energies in the range 0.1-50 MeV. In addition to the primary energy deposition kernels, primary charged particle transport was further characterised by computing the effective centre of the voxels, and the effective penetration depth, effective radius and effective lateral distance travelled by these particles. The dose per unit collision kerma for parallel monoenergetic primary photons beta' was calculated. Additional applications of the energy deposition kernels are discussed.