Effect of tissue inhomogeneities on dose distributions from Cf-252 brachytherapy source.

The Monte Carlo method was used to determine the effect of tissue inhomogeneities on dose distribution from a Cf-252 brachytherapy source. Neutron and gamma-ray fluences, energy spectra and dose rate distributions were determined in both homogenous and inhomogeneous phantoms. Simulations were performed using the MCNP5 code. Obtained results were compared with experimentally measured values published in literature. Results showed a significant change in neutron dose rate distributions in presence of heterogeneities. However, their effect on gamma rays dose distribution is minimal.

The evaluation of neutron and gamma ray dose equivalent distributions in patients and the effectiveness of shield materials for high energy photons radiotherapy facilities.

In this study, the MCNP5 code was used to model radiotherapy room of a medical linear accelerator operating at 18 MV and to evaluate the neutron and the secondary gamma ray fluences, the energy spectra and the dose equivalent distributions inside a liquid tissue-equivalent (TE) phantom. The obtained results were compared with measured data published in the literature. Moreover, the shielding effects of various neutron material shields on the radiotherapy room wall were also investigated. Our simulation results showed that paraffin wax containing boron carbide presents enough effectiveness to reduce both neutron and secondary gamma ray doses.

A compact neutron beam generator system designed for prompt gamma nuclear activation analysis.

In this work a compact system was designed for bulk sample analysis using the technique of PGNAA. The system consists of (252)Cf fission neutron source, a moderator/reflector/filter assembly, and a suitable enclosure to delimit the resulting neutron beam. The moderator/reflector/filter arrangement has been optimised to maximise the thermal neutron component useful for samples analysis with a suitably low level of beam contamination. The neutron beam delivered by this compact system is used to irradiate the sample and the prompt gamma rays produced by neutron reactions within the sample elements are detected by appropriate gamma rays detector. Neutron and gamma rays transport calculations have been performed using the Monte Carlo N-Particle transport code (MCNP5).

Detailed dose distribution prediction of Cf-252 brachytherapy source with boron loading dose enhancement.

The purpose of this work is to evaluate the dose rate distribution and to determine the boron effect on dose rate distribution for (252)Cf brachytherapy source. This study was carried out using a Monte Carlo simulation. To validate the Monte Carlo computer code, the dosimetric parameters were determined following the updated TG-43 formalism and compared with current literature data. The validated computer code was then applied to evaluate the neutron and photon dose distribution and to illustrate the boron loading effect.

Spatial and spectral characteristics of a compact system neutron beam designed for BNCT facility.

The development of suitable neutron sources and neutron beam is critical to the success of Boron Neutron Capture Therapy (BNCT). In this work a compact system designed for BNCT is presented. The system consists of (252)Cf fission neutron source and a moderator/reflector/filter/shield assembly. The moderator/reflector/filter arrangement has been optimized to maximize the epithermal neutron component which is useful for BNCT treatment of deep seated tumors with the suitably low level of beam contamination. The MCMP5 code has been used to calculate the different components of neutrons, secondary gamma rays originating from (252)Cf source and the primary gamma rays emitted directly by this source at the exit face of the compact system. The fluence rate distributions of such particles were also computed along the central axis of a human head phantom.

Russian roulette efficiency in Monte Carlo resonant absorption calculations

The resonant absorption calculation in media containing heavy resonant nuclei is one of the most difficult problems treated in reactor physics. Deterministic techniques need many approximations to solve this kind of problem. On the other hand, the Monte Carlo method is a reliable mathematical tool for evaluating the neutron resonance escape probability. But it suffers from large statistical deviations of results and long computation times. In order to overcome this problem, we have used the Splitting and Russian Roulette technique coupled separately to the survival biasing and to the importance sampling for the energy parameter. These techniques have been used to calculate the neutron resonance absorption in infinite homogenous media containing hydrogen and uranium characterized by the dilution (ratio of the concentrations of hydrogen to uranium). The punctual neutron source energy is taken at Es = 2 MeV and Es = 676.45 eV, whereas the energy cut-off is fixed at Ec = 2.768 eV. The results show a large reduction of computation time and statistical deviation, without altering the mean resonance escape probability compared to the usual analog simulation. The Splitting and Russian Roulette coupled to the survival biasing method is found to be the best methods for studying the neutron resonant absorption, particularly for high energies. A comparison is done between the Monte Carlo and deterministic methods based on the numerical solution of the neutron slowing down equations by the iterative method results for several dilutions.

The vacuum geometry effect on neutron transmission and spatial resolution of neutron transmission.

Frequently, shields used against radiation contain some vacuum channels. We have therefore considered an infinite slab with a fixed thickness (thickness 20 lambda with lambda the mean free path of the neutron in the slab) and an infinite plane source of neutrons which arrived on the left side of the slab; transmitted neutrons through the slab to its right side are detected by finite detectors having windows equal to 2 lambda. This slab contains a vacuum channel. This channel has many legs with several horizontal parts. We used the Monte Carlo method for sampling the neutron history in the slab with a spatial biasing technique in order to accelerate the calculation convergence (Levitt, L. B. (1968) Nuclear Science and Engineering 31, 500-504; Jehouani, A., Ghassoun, J. and Aboubker, A. (1994) In Proceedings of 6th International Symposium on Radiation Physics, Rabat, Morocco). We studied the effects of the angle position and the number of horizontal parts of the channel on the neutron transmission. We have studied the effect of the vacuum channel opening (Artigas, R. and Hungerford, H. E. (1969) Nuclear Science and Engineering 36, 295-303) on the neutron transmission; for several values of this opening we have calculated the neutron transmission probability for each detector position. This study allowed us to determine the optimal conditions of vacuum geometries to improve protection against neutrons. In the second part we considered a shield which consists of a slab and a two-legged vacuum channel with two horizontal parts. The spatial distribution of neutrons transmitted through the protection screen was determined. This distribution shows two peaks. The study was made for different distances between the two horizontal parts. We have determined the smallest distance between the two horizontal parts for which the two peaks can be resolved.

The albedo effect on neutron transmission probability.

The aim of this study is to evaluate the albedo effect on the neutron transmission probability through slab shields. For this reason we have considered an infinite homogeneous slab having a fixed thickness equal to 20 lambda (lambda is the mean free path of the neutron in the slab). This slab is characterized by the factor Ps (scattering probability) and contains a vacuum channel which is formed by two horizontal parts and an inclined one (David, M. C. (1962) Duc and Voids in shields. In Reactor Handbook, Vol. III, Part B, p. 166). The thickness of the vacuum channel is taken equal to 2 lambda. An infinite plane source of neutrons is placed on the first of the slab (left face) and detectors, having windows equal to 2 lambda, are placed on the second face of the slab (right face). Neutron histories are sampled by the Monte Carlo method (Booth, T. E. and Hendricks, J. S. (1994) Nuclear Technology 5) using exponential biasing in order to increase the Monte Carlo calculation efficiency (Levitt, L. B. (1968) Nuclear Science and Engineering 31, 500-504; Jehouani, A., Ghassoun, J. and Abouker, A. (1994) In Proceedings of the 6th International Symposium on Radiation Physics, Rabat, Morocco) and we have applied the statistical weight method which supposes that the neutron is born at the source with a unit statistical weight and after each collision this weight is corrected. For different values of the scattering probability and for different slopes of the inclined part of the channel we have calculated the neutron transmission probability for different positions of the detectors versus the albedo at the vacuum channel-medium interface. Some analytical representations are also presented for these transmission probabilities.

The neutron channeling phenomenon.

Shields, used for protection against radiation, are often pierced with vacuum channels for passing cables and other instruments for measurements. The neutron transmission through these shields is an unavoidable phenomenon. In this work we study and discuss the effect of channels on neutron transmission through shields. We consider an infinite homogeneous slab, with a fixed thickness (20 lambda, with lambda the mean free path of the neutron in the slab), which contains a vacuum channel. This slab is irradiated with an infinite source of neutrons on the left side and on the other side (right side) many detectors with windows equal to 2 lambda are placed in order to evaluate the neutron transmission probabilities (Khanouchi, A., Aboubekr, A., Ghassoun, J. and Jehouani, A. (1994) Rencontre Nationale des Jeunes Chercheurs en Physique. Casa Blanca Maroc; Khanouchi, A., Sabir, A., Ghassoun, J. and Jehouani, A. (1995) Premier Congré International des Intéractions Rayonnements Matière. Eljadida Maroc). The neutron history within the slab is simulated by the Monte Carlo method (Booth, T. E. and Hendricks, J. S. (1994) Nuclear Technology 5) and using the exponential biasing technique in order to improve the Monte Carlo calculation (Levitt, L. B. (1968) Nuclear Science and Engineering 31, 500-504; Jehouani, A., Ghassoun, J. and Aboubker, A. (1994) In Proceedings of the 6th International Symposium on Radiation Physics, Rabat, Morocco). Then different geometries of the vacuum channel have been studied. For each geometry we have determined the detector response and calculated the neutron transmission probability for different detector positions. This neutron transmission probability presents a peak for the detectors placed in front of the vacuum channel. This study allowed us to clearly identify the neutron channeling phenomenon. One application of our study is to detect vacuum defects in materials.