Monte Carlo Code MCNP4B Research Articles

Development and characterisation of a head calibration phantom for in vivo measurements of actinides

The investigation of actinides’ internal contamination in human body makes use of a variety of techniques. In large scale screening the technique of “in vivo” evaluation of bone 241Am burden via the determination of the nuclide activity in the skull is often used. For this purpose, adequate calibration procedures and standard phantoms are needed. The present paper summarises the studies and technical procedures followed for the development of a calibration phantom based on a commercial Alderson angiographic head in which a set of 24 241Am point sources were embedded. A theoretical study was first carried out, at the ENEA Institute for Radiation Protection, using the MCNP4-B Monte Carlo code to determine the point source distribution that closely approximates a homogeneous bone contamination. The numerical models were also used to evaluate the resulting degree of approximation. The point sources were prepared at the ENEA National Metrology Institute for ionising radiation quantities and were traceable to the Italian national standard of radionuclide activity. The sources were prepared by quantitatively dispensing a liquid solution onto a plastic disc. The activity of each source was checked by γ-ray spectrometry and the reproducibility of the activity values was determined. Each source was then placed in the optimum position in the skull, given by the Monte Carlo modelling, by a precision mechanical device. The phantom was finally used to calibrate a whole body counter operating at the ENEA Institute for Radiation Protection. The paper reports the main theoretical and experimental aspects of this work, and also discusses the results of the first calibrations.

Absorbed dose calculations for the Ignitor tokamak magnet coils insulator

The Ignitor experiment is an advanced compact high magnetic field tokamak with cryogenically cooled (30 K) normal conductor magnets. Its main purpose is to produce deuterium–tritium plasma regimes where ignition can take place. From the neutronics point of view, the routine operations with 50% of tritium will lead to a very short (4 s) but intense emission of 14 MeV neutrons (up to 3×10 19 n/s). The capability of the boron-free glass reinforced epoxy resin presently adopted for the insulation of the coils conductor to withstand the severe irradiation condition could be a key issue in the performance, reliability and lifespan of Ignitor. In order to better know the radiation environment in which the insulators will operate, calculations of the absorbed dose are performed using the parallelized version of MCNP-4B Monte Carlo code. A detailed 3D geometry description of the tokamak, an accurate representation of the neutron source distribution together with the more recent version of transport cross-section libraries EFF (European Fusion File) are used in this work. The variation of calculated dose depending on the insulator position inside the device, on the neutron energy spectra and on the total number of neutrons produced over the scheduled machine lifetime are discussed in the present paper.

Measurement of Dose-Equivalent Rates around a Cask and Monte Carlo Analysis with Actual Configuration of Fuel Basket

Gamma ray and neutron dose-equivalent rate distributions are measured with an ionization-type gamma-ray survey meter and a moderator-type neutron survey meter respectively around a TN-12A spent fuel transport cask, and the measured dose-equivalent rate distributions are analyzed by the Monte Carlo method. Two models for the aluminum-alloy fuel basket are considered in the Monte Carlo code MCNP 4B. In one case the configuration of the basket with 12 spent fuel assemblies is modeled in detail and the other is the homogenized basket as employed in the Sn code DOT 3.5. In addition, the burn-up distribution is taken into account to generate source neutrons and gamma rays in the z-axis of the spent fuel assemblies in both cases. The essential difference in the dose-equivalent rates is obtained from the Monte Carlo calculations employing the homogenized model and the actual configuration of the basket. Due to employing the actual configuration, the gamma ray and neutron dose-equivalent rates reduce to 67% and 80%, respectively as compared with the homogenized model on the surface of the TN-12A cask. As the results, the good C/Es are obtained: for the neutron it is approximately 1.0 and 1.25 for the gamma ray it is at the center of the cask surface, respectively The effect of the burn-up distribution appears clearly at the off-center cask surface, and in particular, the neutron dose-equivalent rates come close to the measured ones.

Optimization of the BNCT Filter

The Czech Boron Neutron Capture Therapy (BNCT) facility works by an epithermal neutron beam installed at the LVR-15 reactor at Rez near Prague. Several configurations of moderating and shielding materials have been designed in order to ensure appropriate parameters of the beam. The beam filter consists of cylindrical layers of Al, AlF3 and Ti. To decrease the gamma two layers of Pb are implemented. The filter geometry and composition has been optimized with the aim to increase the epithermal neutron fluence rate and decrease the fast neutron dose rate using the MCNP-4B Monte Carlo code with the DLC-189 library.Suitable patterns of the reactor core were also studied especially in regard to a possible installation of fuel units close to the filter input. Results of calculations show that the optimized variant of the reactor core is able to increase the intensity of the fast neutron source incident to the filter by a factor of 2.0.An experimental verification of the beam parameters was performed using different measurement techniques. The neutron energy spectrum was measured with a set of activation foils, by a Bonner spheres spectrometer and a fast neutron spectrometer with a stilbene crystal. The fast neutron kerma rate was calculated from the spectral measurement. Al-P TLD were used to measure the photon absorbed dose.The beam parameters were measured at 10 MW and were found as follows: the epithermal neutron fluence rate (an energy range from 1 eV to 10 keV) of 9.29 × 10l2m−2 s−1, the fast neutron kerma rate in tissue of 2.63 Gy h−1 and the incident gamma dose rate in tissue of 2.9 Gy h−1.

A virtual source model for Monte Carlo modeling of arbitrary intensity distributions.

A photon virtual source model was developed for simulating arbitrary, external beam, intensity distributions using the Monte Carlo method. The source model consists of a photon fluence grid composed of a matrix of square elements, located 25-cm downstream from the linear accelerator target. Each particle originating from the fluence map is characterized by the seven phase space parameters, position (x, y, z), direction (u, v, w), and energy. The map was reconstructed from fluence and energy spectra acquired by modeling components of the linear accelerator treatment head using the Monte Carlo code MCNP4B. The effect of contaminant electrons is accounted for by the use of a sub-source derived from a phase-space simulation of a 25-MV linac treatment head using the code BEAM. The BEAM sub-source was incorporated into the MCNP4B phase-space model and is sampled using a field-size dependent sampling ratio. A Gaussian blurring kernel is convolved with the photon fluence map to account for the finite focal spot size and scattering effects from structures such as the flattening filter and MLC leaves. Depth dose and profile source calculations for 6-MV and 25-MV photon beams, for 5 x 5 cm2, 10 x 10 cm2, and 15 x 15 cm2 field sizes, are in good agreement with measurement and are well within acceptability criteria suggested by the AAPM Task Group Report No. 53. Irregular field calculations compared with film measurement and with a 3-D pencil beam algorithm show that the source model is capable of accurately simulating arbitrary MLC fields.

CT-based dosimetry calculations for 125I prostate implants

Purpose: To evaluate the Monte Carlo code MCNP4B for low-energy brachytherapy calculations, including the effects of interseed attenuation and patient specific heterogeneities, on the calculated dose distribution from transperineal implantation of 125I. Methods and Materials: The Monte Carlo code MCNP4B was used to model and benchmark the absolute dose distribution from two 125I brachytherapy seeds (model 6711 and 6702). Based upon the physical source model, the total photon intensity and differential energy spectrum were evaluated as a function of angle from the transverse bisector of the source. These spectral and intensity data were reformatted to produce probability distributions for sampling from a virtual point source. The virtual source model and a modified version of MCNP4B is then used for simulating arbitrary brachytherapy source configurations within a homogeneous or heterogeneous patient specific computed tomography (CT)-based lattice geometry. Results and Conclusion: Comparison with TG-43 data and the Monte Carlo calculations is excellent with MCNP4B predicting the radial dose function for the 125I 6711 and 6702 sources within 6% for all data points tested. Attenuation effects from neighboring seeds were investigated for pre- and postimplant seed distributions and found to be negligible. Preliminary dosimetry analysis of postimplant seed distributions comparing homogeneous water versus heterogeneous CT simulation geometries indicates an average decrease of approximately 5.6% for the volume of tissue irradiated to a prescription isodose line of 144 Gy.

Calculation of absorbed dose and biological effectiveness from photonuclear reactions in a bremsstrahlung beam of end point 50 MeV

The absorbed dose due to photonuclear reactions in soft tissue, lung, breast, adipose tissue and cortical bone has been evaluated for a scanned bremsstrahlung beam of end point 50 MeV from a racetrack accelerator. The Monte Carlo code MCNP4B was used to determine the photon source spectrum from the bremsstrahlung target and to simulate the transport of photons through the treatment head and the patient. Photonuclear particle production in tissue was calculated numerically using the energy distributions of photons derived from the Monte Carlo simulations. The transport of photoneutrons in the patient and the photoneutron absorbed dose to tissue were determined using MCNP4B; the absorbed dose due to charged photonuclear particles was calculated numerically assuming total energy absorption in tissue voxels of 1 cm3. The photonuclear absorbed dose to soft tissue, lung, breast and adipose tissue is about (0.11-0.12)±0.05% of the maximum photon dose at a depth of 5.5 cm. The absorbed dose to cortical bone is about 45% larger than that to soft tissue. If the contributions from all photoparticles (n, p, 3He and 4He particles and recoils of the residual nuclei) produced in the soft tissue and the accelerator, and from positron radiation and gammas due to induced radioactivity and excited states of the nuclei, are taken into account the total photonuclear absorbed dose delivered to soft tissue is about 0.15±0.08% of the maximum photon dose. It has been estimated that the RBE of the photon beam of 50 MV acceleration potential is approximately 2% higher than that of conventional 60Co radiation.

Studies on the Shielding Properties of Polyboron and Ilmenite-Magnetite Concrete Using a Reactor Neutron Beam

The shielding effectiveness of locally developed polyboron and ilmenite-magnetite (I-M) concrete is investigated using the reactor neutron beam of the 3-MW TRIGA Mark II research reactor at the Atomic Energy Research Establishment, Savar, Dhaka. The effective removal cross sections for the foregoing individual shielding materials as well as their combinations are obtained from transmission data using two-group neutron fluxes defined by a Cd-cutoff value. The experimental transmission factors for I-M concrete and polyboron are compared with those obtained from transport calculations performed with the ANISN deterministic code in the forward mode and the MCNP4B Monte Carlo code. The ANISN code is used for the fast neutron group flux (Cd-cutoff flux), and the MCNP4B code is used for the total neutron flux. The agreement between the experiment and calculation is fairly good at deep penetration, but at initial points, some disagreement is observed. This observation is valid for both polyboron and I-M concrete.

A validation of copper, nickel and iron data in endf/b-vi.2, jef-2.2 and jendl-3.2 through a series of br-1 critical core analysis

This report discusses the lessons learned from a validation of copper, nickel and iron data in ENDF/B-VI.2, JEF-2.2 and JENDL-3.2 through an analysis of 97.6 at. % 239PU metal fueled BR-1 fast reactor benchmarks. The specifications for the benchmarks in this study were taken from the ICSBEP handbook. The Monte Carlo code MCNP4B and continuous-energy cross-section libraries processed from the above three evaluated nuclear data files were used to perform the critical benchmark calculations. There is reasonable agreement between the predicted k eff -values and the benchmarks within the uncertainties of measurements and Monte Carlo calculations, with the exception of the JEF-2.2 and JENDL-3.2 results for the copper-reflected core. The JEF-2.2 and JENDL-3.2 data give higher k eff -values than those of ENDF/B-VI.2 for all the three reflector materials.

Monte Carlo Calculations Using MCNP4B for an Optimal Shielding Design for a 14 MeV Neutron Source

The shielding for 14 MeV neutrons from a D-T generator at SLAC needs to be optimised in terms of shield thickness and cost. Shielding calculations were made using the MCNP4B Monte Carlo code and the ENDF/B-VI cross section set. The following materials were studied: iron, borated polyethylene, three types of normal and heavy concrete, and combinations of iron and borated polyethylene. The effects of shielding geometry (sphere, cube, and the proposed geometry) were also examined. The results of the transmission factor for ambient dose equivalent, as well as the attenuation lengths and the average energies, for neutrons and gammas outside the shielding are presented.