Monte Carlo determination of water concentration effect on gamma-ray detection efficiency in soil samples

Water concentration effect on full energy peak efficiency in a soil sample taken from a soil profile in Erzurum (39°55′ N; 41°16′ E; 200 m above sea level), Turkey was determined using Monte Carlo simulation technique. The dependence of the full energy peak efficiency on the water concentration in the soil was obtained for some particular photon energies ranging from 60 keV to 2 MeV and, as a result, the corresponding correction factors were obtained. It was observed that the correction factor approaches unity with increasing energy and decreases with increasing water concentration.     

Monte Carlo simulations to advance characterisation of landmines by pulsed fast/thermal neutron analysis.

The performance of a detection system based on the pulsed fast/thermal neutron analysis technique was assessed using Monte Carlo simulations. The aim was to develop and implement simulation methods, to support and advance the data analysis techniques of the characteristic gamma-ray spectra, potentially leading to elemental characterisation of innocuous objects using the full spectrum analysis (FSA) approach. The simulations were carried out with a simplified tool, based on a 14MeV DT pulse-neutron source and a bismuth-germanate detector. A MCNP-based method for de-coupling the radiation transport in mixed (n,gamma) fields, to generate separate sets of standard detector gamma-ray responses for individual elements, is outlined. When normalised and experimentally benchmarked in terms of the pulse-neutron source production rate, the standard spectra can be incorporated into algorithms for the FSA of in situ measurements and elemental fingerprinting of the inspected object.     

Glandular breast dose for monoenergetic and high-energy X-ray beams: Monte Carlo assessment.

PURPOSE: To extend the utility of normalized glandular dose (DgN) calculations to higher x-ray energies (up to 120 keV) and to provide the tools for investigators to calculate DgN values for arbitrary mammographic and x-ray spectra. MATERIALS AND METHODS: Validated Monte Carlo methods were used to assess DgN values. One million x-ray photons (1-120 keV, in 1-keV increments) were input to a semicircular breast geometry of thicknesses from 2 to 12 cm and breast compositions from 0% to 100% glandular. DgN values for monoenergetic (1-120 keV) x-ray beams, polyenergetic (40-120 kV, tungsten anode) x-ray spectra, and polyenergetic mammographic spectra were computed. Skin thicknesses of 4-5 mm were used. RESULTS: The calculated DgN values were in agreement within approximately 1%-6% with previously published data, depending on breast composition. DgN tables were constructed for a variety of x-ray tube anode-filter combinations, including molybdenum anode-molybdenum filter, molybdenum anode-rhodium filter, rhodium anode-rhodium filter, tungsten anode-rhodium filter, tungsten anode-palladium filter, and tungsten anode-silver filter. DgN values also were graphed for monoenergetic beams to 120 keV and for general diagnostic x-ray beams to 120 kV. CONCLUSION: The tables and graphs may be useful for optimizing mammographic procedures. The higher energy data may be useful for investigations of the potential of dual-energy mammography or for calculation of dose in general diagnostic or computed tomographic procedures.     

Diffusion cooling of thermal neutrons in basic rock minerals by Monte Carlo simulation of the pulsed neutron experiments

The pulsed neutron experiment (the variable geometric buckling experiment) in spherical geometry has been simulated using the MCNP code. The time decay of the thermal neutron flux has been observed as a function of the sample size. The thermal neutron diffusion cooling coefficient C with the correction F has been determined for three basic rock minerals (quartz, calcite, dolomite) at the given specific densities. The corresponding density-removed parameters have also been obtained.     

Sulfur-33 nanoparticles: A Monte Carlo study of their potential as neutron capturers for enhancing boron neutron capture therapy of cancer

In this paper, the potential effect of enhancing BNCT near the surface of the target volume by means of the addition of the sulfur isotope ³³S is studied. By means of Monte Carlo simulations, it is found a noticeable enhancement effect (local increase of the dose at the isotope site) when it is present at local concentrations that in principle can be reached by means of sulfur nanoparticles. A neutron beam with a high component of 13.5 keV would be required to produce this effect. Some open problems are discussed.     

A feasibility study on gamma-ray tomography by Monte Carlo simulation for development of portable tomographic system

The electron beam X-ray tomographic scanner has been used in industrial and medical field since it was developed two decades ago. However, X-ray electron beam tomography has remained as indoor equipment because of its bulky hardware of X-ray generation devices. By replacing X-ray devices of electron beam CT with a gamma-ray source, a tomographic system can be a portable device. This paper introduces analysis and simulation results on industrial gamma-ray tomographic system with scanning geometry similar to electron beam CT. The gamma-ray tomographic system is introduced through the geometrical layout and analysis on non-uniformly distributed problem. The proposed system adopts clamp-on type device to actualize portable industrial system. MCNPx is used to generate virtual experimental data. Pulse height spectra from F8 tally of MCNPx are obtained for single channel counting data of photo-peak and gross counting. Photo-peak and gross counting data are reconstructed for the cross-sectional image of simulation phantoms by ART, Total Variation algorithm and ML-EM. Image reconstruction results from Monte Carlo simulation show that the proposed tomographic system can provide the image solution for industrial objects. Those results provide the preliminary data for the tomographic scanner, which will be developed in future work.     

Variation of the thermal neutron diffusion cooling properties of wet rock material (Monte Carlo simulations of the pulsed neutron experiments).

The water content in a rock material can significantly change the thermal neutron diffusion parameters with respect to those of the dry medium. The effect has been studied for dolomite, CaMg(CO3)2, by Monte Carlo simulations of the variable buckling experiments for 10 series of samples. The density-removed diffusion cooling coefficient C(M) varies hyperbolically by two orders of magnitude with water content in the range of 0-20%.     

The optimisation of the fast neutron and gamma-ray transmission set-up for moisture measurement of coke.

In the present paper, modelling calculations with the Monte Carlo (MCNP4C) code were performed for the optimisation of the fast neutron and gamma-ray transmission, set-up, used for the humidity measurement of coke. The optimisation focused on maximising the sensitivity of the neutron flux to humidity changes and on lowering neutron-counting error, both leading to higher accuracy of coke moisture determination. Different materials used for the source shielding and neutron collimation, together with different dimensions of the neutron collimators were studied. The results obtained from the Monte Carlo modelling correlate with the real instrument performance.     

Monte Carlo simulation of gamma-ray spectra from natural radionuclides recorded by a NaI detector in the marine environment.

The GEANT4 Monte Carlo code has been used to simulate gamma-ray spectra of natural radionuclides collected by a NaI scintillation detector immersed in seawater. The gamma-rays emitted from the decay of (40)K, and the series of (232)Th and (238)U, were used to describe the radioactive water source around the NaI crystal. The simulated gamma-ray spectra were compared with real data recorded in situ by a newly constructed NaI spectrometer and were found to be in good agreement. The NaI spectrometer was calibrated in the laboratory in a water tank, before its deployment in seawater. Activity concentrations were deduced from the gamma-ray spectra and discussed in comparison with results from the literature.     

Precise Monte Carlo simulation of gamma-ray response functions for an NaI(Tl) detector.

Three widely used methods to calculate the response functions for NaI(TI) detectors were investigated. The methods were Berger-Seltzer's method (Nucl. Instrum. Methods 104 (1972) 317), general Monte Carlo (MC) programs, such as EGS4 (The EGS4 code system, Stanford Linear Accelerator Center, 1985) and MCNP4B (MCNP-a general Monte Carlo N-particle transport code, Los Alamos National Laboratory Report, LA-12625-M, 1997), and special MC programs. The pulse height spectra in a 3" x 3" NaI(Tl) detector due to several gamma-ray sources have been measured to verify the calculated results of these methods. The energies of the sources ranged from 0.4118 to 7.11 MeV. The spectra generated by Berger-Seltzer's method and the general MC programs did not agree well with the experimental data. PETRANS 1.0, the special MC program developed in house, was fairly accurate since it also considered the scintillation efficiency and the single escape peak shift.     

Monte Carlo calculation of epithermal neutron resonance self-shielding factors in wires of different materials.

Epithermal neutron resonance self-shielding factors in wires of materials used as activation detectors or as targets for radionuclide production have been calculated using the MCNP code. The energy dependent self-shielding factor depends on the ratio scattering/capture cross sections. The self-shielding factors for cobalt and gold have been compared with available values. The self-shielding factor depends on various physical and nuclear parameters. However, an adimensional variable can be adopted that describes the self-shielding factors of different materials by a quasi "universal curve".     

A Monte Carlo treatment of particle-size effects in gauging and control by gamma-ray absorptiometry in hydraulic transport systems

The widespread use of γ-ray absorptiometry in non-destructive assay of opaque materials is well established. Analytical treatment of particle-size effects in materials with several corase-grained components requires a theory based on complicated multinomial distributions. As a consequence, no closed expressions can be derived. In this paper, a Monte Carlo model for γ-ray absorptiometry in mixtures with multiple coarse-grained materials is developed. The model calculations are based on an entirely non-analog simulation procedure for γ-ray transport. The weighing factors for forced photon penetration of the different component materials are estimated, and a suitable correction function for particle-size effects is derived. An application on to some specific analysis problems where the model could possibly be used to good advantage has been demonstrated.     

Monte Carlo simulation of neutron irradiation facility developed for accelerator based in vivo neutron activation measurements in human hand bones.

The neutron irradiation facility developed at the McMaster University 3 MV Van de Graaff accelerator was employed to assess in vivo elemental content of aluminum and manganese in human hands. These measurements were carried out to monitor the long-term exposure of these potentially toxic trace elements through hand bone levels. The dose equivalent delivered to a patient during irradiation procedure is the limiting factor for IVNAA measurements. This article describes a method to estimate the average radiation dose equivalent delivered to the patient's hand during irradiation. The computational method described in this work augments the dose measurements carried out earlier [Arnold et al., 2002. Med. Phys. 29(11), 2718-2724]. This method employs the Monte Carlo simulation of hand irradiation facility using MCNP4B. Based on the estimated dose equivalents received by the patient hand, the proposed irradiation procedure for the IVNAA measurement of manganese in human hands [Arnold et al., 2002. Med. Phys. 29(11), 2718-2724] with normal (1 ppm) and elevated manganese content can be carried out with a reasonably low dose of 31 mSv to the hand. Sixty-three percent of the total dose equivalent is delivered by non-useful fast group (> 10 keV); the filtration of this neutron group from the beam will further decrease the dose equivalent to the patient's hand.     

Development of real-time thermal neutron monitor using boron-loaded plastic scintillator with optical fiber for boron neutron capture therapy.

A new thermal neutron monitor for boron neutron capture therapy was developed in this study. We called this monitor equipped boron-loaded plastic scintillator that uses optical fiber for signal transmission as an [scintillator with optical fiber] SOF detector. A water phantom experiment was performed to verify how the SOF detector compared with conventional method of measuring thermal neutron fluence. Measurements with a single SOF detector yielded indistinguishable signals for thermal neutrons and gamma rays. To account for the gamma ray contribution in the signal recorded by the SOF detector, a paired SOF detector system was employed. This was composed of an SOF detector with boron-loaded scintillator and an SOF detector with a boron-free scintillator. The difference between the recorded counts of these paired SOF detectors was used as the measure of the gamma ray contribution in the measured neutron fluence. The paired SOF detectors were ascertained to be effective in measuring thermal neutron flux in the range above 10(6)(n/cm(2)/s). Clinical trials using paired SOF to measure thermal neutron flux during therapy confirmed that paired SOF detectors were effective as a real-time thermal neutron flux monitor.     

Investigations on thermal neutron reflection by activation method.

The prompt neutron detection and the foil activation methods were compared for the determination of the reflection cross-section of thermal neutrons and the hydrogen content of bulk samples. The advantages and limitations of the two methods are discussed.     

Radiation injury of boron neutron capture therapy using mixed epithermal- and thermal neutron beams in patients with malignant glioma.

The purpose of this study was to clarify the radiation injury in acute or delayed stage after boron neutron capture therapy (BNCT) using mixed epithermal- and thermal neutron beams in patients with malignant glioma. Eighteen patients with malignant glioma underwent mixed epithermal- and thermal neutron beam and sodium borocaptate between 1998 and 2004. The radiation dose (i.e. physical dose of boron n-alpha reaction) in the protocol used between 1998 and 2000 (Protocol A, n = 8) prescribed a maximum tumor volume dose of 15 Gy. In 2001, a new dose-escalated protocol was introduced (Protocol B, n = 4); it prescribes a minimum tumor volume dose of 18 Gy or, alternatively, a minimum target volume dose of 15 Gy. Since 2002, the radiation dose was reduced to 80-90% dose of Protocol B because of acute radiation injury. A new Protocol was applied to 6 glioblastoma patients (Protocol C, n = 6). The average values of the maximum vascular dose of brain surface in Protocol A, B and C were 11.4+/-4.2 Gy, 15.7+/-1.2 and 13.9+/-3.6 Gy, respectively. Acute radiation injury such as a generalized convulsion within 1 week after BNCT was recognized in three patients of Protocol B. Delayed radiation injury such as a neurological deterioration appeared 3-6 months after BNCT, and it was recognized in 1 patient in Protocol A, 5 patients in Protocol B. According to acute radiation injury, the maximum vascular dose was 15.8+/-1.3 Gy in positive and was 12.6+/-4.3 Gy in negative. There was no significant difference between them. According to the delayed radiation injury, the maximum vascular dose was 13.8+/-3.8 Gy in positive and was 13.6+/-4.9 Gy in negative. There was no significant difference between them. The dose escalation is limited because most patients in Protocol B suffered from acute radiation injury. We conclude that the maximum vascular dose does not exceed over 12 Gy to avoid the delayed radiation injury, especially, it should be limited under 10 Gy in the case that tumor exists in speech center.     

Optimization of the target of an accelerator-driven neutron source through Monte Carlo numerical simulation of neutron and gamma transport by the PRIZMA code

At Budker Institute of Nuclear Physics, epithermal neutron source for neutron-capture therapy was built and neutron generation was realized. Source is based on tandem accelerator and uses near-threshold neutron generation from the reaction ⁷Li(p,n)⁷Be. The paper describes target optimization through the numerical simulation of proton, neutron and gamma transport by Monte Carlo method (PRIZMA code). It is shown that the near-threshold mode attractive due low activation provides high efficiency of the dose and acceptable therapeutic ratio and advantage depth.     

A Monte Carlo study on neutron and electron contamination of an unflattened 18-MV photon beam

Recent studies on flattening filter (FF) free beams have shown increased dose rate and less out-of-field dose for unflattened photon beams. On the other hand, changes in contamination electrons and neutron spectra produced through photon (E>10 MV) interactions with linac components have not been completely studied for FF free beams. The objective of this study was to investigate the effect of removing FF on contamination electron and neutron spectra for an 18-MV photon beam using Monte Carlo (MC) method. The 18-MV photon beam of Elekta SL-25 linac was simulated using MCNPX MC code. The photon, electron and neutron spectra at a distance of 100 cm from target and on the central axis of beam were scored for 10×10 and 30×30 cm² fields. Our results showed increase in contamination electron fluence (normalized to photon fluence) up to 1.6 times for FF free beam, which causes more skin dose for patients. Neuron fluence reduction of 54% was observed for unflattened beams. Our study confirmed the previous measurement results, which showed neutron dose reduction for unflattened beams. This feature can lead to less neutron dose for patients treated with unflattened high-energy photon beams.     

Extended use of alanine irradiated in experimental reactor for combined gamma- and neutron-dose assessment by ESR spectroscopy and thermal neutron fluence assessment by measurement of 14C by LSC

Gamma- and neutron doses in an experimental reactor were measured using alanine/electron spin resonance (ESR) spectrometry. The absorbed dose in alanine was decomposed into contributions caused by gamma and neutron radiation using neutron kerma factors. To overcome a low sensitivity of the alanine/ESR response to thermal neutrons, a novel method has been proposed for the assessment of a thermal neutron flux using the ¹⁴N(n,p) ¹⁴C reaction on nitrogen present in alanine and subsequent measurement of ¹⁴C by liquid scintillation counting (LSC).     

Long range effects of neutron capture therapy of cancer in mice II. Dose effect

An index is established which compares the outcome of neutron therapy experiments when only a single exposure is given to a transplantable malignant neoplasm in mice if the thermal neutron attenuation in passing through the tumor and tumor boron concentration at the time of exposure are known. A high degree of correlation is shown for the EXIT RAD INDEX in tables and graphs recording and summarizing studies done on some 3500 mice treated for thigh-implanted malignant neoplasms varying from 8–17 mm in dia. at treatment time. The end point was permanent, total regression of the neoplasm occurring without scarring Each animal with a treated neoplasm was followed until death. It is shown that this INDEX makes possible valid comparisons of procedures carried out over several years and in which exposure times vary from 0.1 sec to 23 min in three different reactors at two different boron dosages.