MCNP4C Code Research Articles

Gamma-ray shielding properties for silicate glasses containing Bi2O3, PbO, and BaO

Introduction: With the extensive application of radiation sources and radioactive materials in various fields, the study of attenuation and absorption of X-rays and gamma-rays in several materials has become an essential and interesting field of research. Glass materials are possible alternatives of concretes for radiation-shielding materials. Materials and Methods: The mass attenuation coefficients (µm), effective atomic cross section (σa), effective atomic number (Zeff) and effective electron density (Neff) of silicate glasses containing different concentrations of Bi2O3, PbO, and BaO (30–70% by weight) were obtained using MCNP-4C code, XCOM and XMuDat programs, in the energy range of 10 keV–10 MeV. The interpolation method was employed to extract Zeff and Neff. The Auto-Zeff software was used to determine the Zeff as well. Results: The MCNP-4C Code, XCOM and XMuDat programs and Auto-Zeff results were in good agreement. It was found that the µm and σa values decrease with photon energy. Also the results indicate that σa and Zeff of glasses improve by increasing their Bi2O3, PbO and BaO contents, while increasing these contents to the same fraction has no significant effect on µm in the energy range of 0.3–5 MeV, where the Compton Effect takes over as dominant process and Neff do not follow certain rule. Above 0.1 MeV, lead oxide silicate glass showed the highest values of Zeff and σa. Conclusion: Observed good agreement indicates that the chosen Monte Carlo code and computer programs may be useful to calculate the gamma-ray shielding characteristics of different glass systems, particularly in cases where no experimental data exist.

The plastic tray effect on the spectrum of photoneutrons prudused by 18 MV medical linac using the Monte Carlo method

Introduction: One of the important problems of using high energy linacs is the production of photoneutrons. High energy medical linacs produce secondary neutrons. These phtoneutrons increase patient dose Materials and Methods: In this study, the effect of plastic trays on spectrum of photoneutrons produced by 18MV linac is investigated using the MCNP4C code. To determine the photoneutron spectrum a simplified linac head simulated. In order to contribute the all components of the spectra, an isotropic neutron source was considered with an energy spectrum derived from the relevant equation. Source defined at the center of a thick walled sphere of tungsten. Two kinds of trays, PMMA and Polyethylene, each with 1.5 cm thickness defined just under the aperture of simplified linac head. A typical treatment room simulated with the walls made of standard concrete. Neutron detectors defined as spheres of 10 cm diameter. One of the detectors defined just under the linac head and the other at the adjacent of the bucker door. Three situations were considered, one without tray and two others with trays. Simulation carried out by running the computer to produce 100 millions of particles for each case. F4 tally was used. Results: The results show that spectrum of photoneutrons differs considerably for open field and shielded fields.

The effect of energy on the physical parameters of the electron beam in curtain accelerator: A simulation study

In electron curtain accelerator, the absorbed dose and its uniformity, electron energy loss through beam extract window and its lifetime are key factors for users. To facilitate design and application of curtain electron accelerator in 100–300 keV energy range, with the typical electron beam current (43 mA) and 13-µm-thick extract titanium window, the amount of doses immediately after the extract window, on the target position and at the depth of 1 mm of the target was obtained by MCNP4C code. The results of this simulation showed that with increasing energy, dose increases at indicated positions. It is shown that, by increase the incident beam energy from 100 keV to 300 keV, the temperature on the titanium extract window decreased from 104 to 42 degrees centigrade, respectively. Furthermore, lateral dose distribution and beam energy loss through the typical extract window was obtained based on CASINO numerical simulation package and compared with analytical results.

Calculation of fuel burnup and radionuclide inventory for the HEU and potential LEU fuels in the IRT research reactor

Abstract The study of core conversion of the IRT research reactor from the HEU to LEU fuels has been conducted in this paper. Fuel burnup, buildup of actinides and fission product inventory for the IRT research reactor with UO2-Al dispersion fuel for the various different uranium enrichments and the uranium density such as: (IRT-3M 90% 1.1 g/cm3, IRT-3M 36% 2.51 g/cm3, IRT-4M 19.75% 3.25 g/cm3, IRT-4M 19.75% 3.48 g/cm3, IRT-4M 19.75% 3.71 g/cm3) have been carried out using the GETERA code. The multiplication factor (keff) of the IRT-3M and IRT-4M reactors were calculated using the MCNP4C code. The results showed, after 200 days of the reactor operation time that the total consumptions of 235U for the five cases were 73.85, 73.13, 72.66, 72.80 and 73.21 g respectively. The amounts of 239Pu produced in the core were: 0.43, 3.39, 5.90, 6.13 and 6.35 g respectively. The total radioactivities of fuel assembly were: 2.78E+15, 2.83E+15, 2.86E+15, 2.88E+15 and 2.89E+15 Bq respectively. Finally, from the neutronic view point, the LEU fuel was validated to be a suitable candidate for the fuel conversion of the IRT in the future.

Study on the dose modification factor of strut adjusted volume implant (SAVI) with a 169Yb source using MCNP4C.

The SAVI has gained widespread use for accelerated partial breast irradiation (APBI) brachytherapy. Treatments with SAVI produce inherent heterogeneities including variable backscatter due to proximity to the tissue-air interface and variable cavity contents, causing inaccuracy in the dose calculation. In this study, a model of SAVI with sources of 169Yb developed recently was defined and simulations with MCNP4C code of Monte Carlo were performed through different scenarios to assess the effects of these heterogeneities on the dose distribution. The results showed that the dose delivered to target volume may be lower than the planned dose by up to 9-16%. Therefore, the therapy with169Yb must be viewed with caution and a correction factor must be applied in treatment planning systems. It was also observed that the presence of air cavity changed the relative dose 1% for a symmetric plan compared to a water cavity and up to 5% for an asymmetric plan. It was indicated that the dose modification factor (DMF) did not differ in any significant way to that of the symmetric plan. The effect of the composition on the DMF was negligible because the natures of water and breast tissue were approximately similar. All the obtained results indicated that should 169Yb sources be used with SAVI applications, some amendments of treatment planning systems would be employed.

Benchmark test of JEFF-3.2, FENDL-3.0 and TENDL-2015 with TOF experiments for graphite, silicon carbide, uranium and multiple-slab samples

In order to validate JEFF-3.2, FENDL-3.0 and TENDL-2015 nuclear data libraries, some integral experiments have been conducted with TOF method on the D-T neutron source facility at China Institute of Atomic Energy (CIAE). The measured neutron spectra for Silicon Carbide, Graphite, Uranium and multi-slab sample assemblies were compared with the leakage neutron spectra simulated using the MCNP-4C code with the evaluated nuclear data libraries. The comparison was made by the spectrum shape and the C/E values in five energy regions. From the results we can conclude that all the calculations reproduce well the measurements except at a few neutron energy ranges. It is worth mentioning that the calculated results with FENDL-3.0 are almost the same as those with TENDL-2015for graphite. And compared to JEFF-3.1.1, the nuclear data of carbon were not revised in JEFF-3.2. An interesting phenomenon was significantly observed from the results of multi-slab sample assemblies. The excessive number of elastically-scattered neutrons in the U layer are moderated by the hydrogen in the last layer to lower energies as the C/E ratio satisfactorily becomes consistent with 1.0.

Construction of a Prototype Silver Self-powered Neutron Detector and Simulation of a Developed SPND using MCNP4C Code

Self-powered neutron detectors (SPNDs) are used widely in flux mapping, control, and protection of nuclear reactors. Compared with other neutron detector, SPND has the advantages of small size, low cost, and relatively simple electronics. In this work, a prototype SPND for flux measuring neutron generators has been designed and tested. Using the Monte Carlo simulation code, the output current of the detector was calculated and compared with the experimental results. The agreement between the results suggests that this detector is capable of flux monitoring of the neutron generator. Then, a developed SPND has been introduced with the provision of producing higher output current.

Energy Broadening Model for BGO Array Crystal from 0.059 to 1.332 MeV

The energy resolution and energy broadening parameters are of great importance as far as scintillator detector is concerned. At present, BGO crystal due to its high density, high yield of photoelectric absorption, availability and cost is used in particle physics and medical imaging system especially in positron emission tomography (PET). The purpose of this study is to calculate the Gaussian energy broadening parameters of a 10 × 10 BGO array in order to use it in simulation for analysis of its spectral response in the presence of gamma rays in a wide range of energies. The Gaussian energy broadening parameters were extracted from the experimental spectrum to use them as the input parameters in MCNP4c code. The simulated response of the detector employing these parameters showed a good agreement with the experimental data.

Experiment on uranium slabs of different thicknesses with D-T neutrons and validation of evaluated nuclear data

In order to validate the evaluated nuclear data for 238U, leakage neutron spectra from pure 238U slab samples of different thicknesses bombarded with 14.8MeV D-T neutrons were measured at 60° (for 10cm×10cm×2cm, 10cm×10cm×5cm, 10cm×10cm×11cm samples) and 120° (for 10cm×10cm×5cm, 10cm×10cm×11cm samples) respectively. The measurement was performed at China Institute of Atomic Energy using time-of-flight method. Monte Carlo simulations were carried out by MCNP-4C code combined with various evaluated nuclear data libraries, namely, ENDF/B-VII.1, JENDL-4.0, CENDL-3.1, JEFF-3.2 and TENDL-2015. The experimental results were compared with the corresponding calculated ones, and the calculation-to-experiment ratio (C/E) which indicates the agreement between them was obtained. The results show that the JENDL4.0 fitted the experimental data best.

Determination of gamma-ray shielding properties for silicate glasses containing Bi2O3, PbO, and BaO

The mass attenuation coefficients (μm), effective atomic cross section (σa), effective atomic number (Zeff) and effective electron density (Neff) of silicate glasses containing different concentrations of Bi2O3, PbO, and BaO (30–70% by weight) were obtained using MCNP-4C code, XCOM and XMuDat programs, in the energy range of 10 keV–10 MeV. The interpolation method was employed to extract Zeff and Neff. The Auto-Zeff software was used to determine the Zeff as well. The MCNP-4C Code, XCOM and XMuDat programs and Auto-Zeff results were in good agreement. It was found that the μm and σa values decrease with photon energy. Also the results indicate that σa and Zeff of glasses improve by increasing their Bi2O3, PbO and BaO contents, while increasing these contents to the same fraction has no significant effect on μm in the energy range of 0.3–5 MeV, where the Compton Effect takes over as dominant process and Neff do not follow certain rule. Above 0.1 MeV, lead oxide silicate glass showed the highest values of Zeff and σa. Wherever possible, the simulated and calculated values were compared with experimental data.

Simulation of Gaussian energy broadening in gamma response of a LYSO array detector using a semi-empirical method

The energy broadening parameters for a $10 \times 10$ LYSO:Ce array were determined experimentally in an energy range of 59keV to ∼ 1300 keV for gamma rays. The obtained parameters can be used to improve the Monte Carlo simulation of the spectral behavior of the LYSO-based detectors, especially in PET imaging systems. The detector spectral responses were simulated by the MCNP4c code using the obtained parameters. The obtained results from the simulations were verified using real gamma spectra.

A comparison study for mass attenuation coefficients of some amino acids using MCNP code

Abstract In this study, a novel model of MCNP4C code reported recently was used to determine the photon mass attenuation coefficients of some amino acids at energies, 123, 360, 511, 662, 1170, 1280 and 1330 keV. The simulation results were compared with the XCOM data. It was indicated that the results were highly close to the calculated XCOM values. Obtained results were used to calculate the molar extinction coefficient. All the results showed the convenience and usefulness of the model in calculation of mass attenuation coefficients of amino acids.

A computational study of re-cycling the MTR research reactors with MOX fuel mixed with 241Am, 242Am and 243Am as burnable absorbers actinides using the MCNP4C code

A computational study is performed for re-cycling of the MTR-22MW reactor with MOX (U3O8&PuO2) fuel mixed with (x)Am (241Am, 242Am and 243Am) as Burnable Absorbers Actinides (BAAs) using the MCNP4C code. The MCNP4C code was used to estimate the Criticality Safety Parameters (CSPs) and the Neutronic Parameters (NPs) of the MTR-22MW for the U3O8-Al original fuel and a MOX fuel mixed with (x)Am As BAAs. The obtained results of the CSPs and the NPs showed that re-cycling of the MTR-22MW reactor by MOX fuel mixed with (x)Am As BAAs would not significantly affect the neutronic characteristics of the MTR-22MW as well as its criticality. In addition results showed that by re-cycling of 4% and 3.51% of PuO2 mixed with (x)Am in the MTR-22MW reactor, the fuel enrichment with 235U can be reduced from 19.7% to 6.6% and 6.2%, respectively. In addition, the 235U mass loaded in the reactor core diminishes by about 37.77% and 42.18% for 4% and 3.51% of PuO2 mixed with (x)Am actinides, respectively.

The use of MOX caramel fuel mixed with 241Am, 242mAm and 243Am as burnable absorber actinides for the MTR research reactors

The MOX (UO2&PuO2) caramel fuel mixed with 241Am, 242mAm and 243Am as burnable absorber actinides was proposed as a fuel of the MTR-22MW reactor. The MCNP4C code was used to simulate the MTR-22MW reactor and estimate the criticality and the neutronic parameters, and the power peaking factors before and after replacing its original fuel (U3O8-Al) by the MOX caramel fuel mixed with 241Am, 242mAm and 243Am actinides. The obtained results of the criticality, the neutronic parameters, and the power peaking factors for the MOX caramel fuel mixed with 241Am, 242mAm and 243Am actinides were compared with the same parameters of the U3O8-Al original fuel and a maximum difference is −6.18% was found. Additionally, by recycling 2.65% and 2.71% plutonium and 241Am, 242mAm and 243Am actinides in the MTR-22MW reactor, the level of 235U enrichment is reduced from 4.48% to 3% and 2.8%, respectively. This also results in the reduction of the 235U loading by 32.75% and 37.22% for the 2.65%, the 2.71% plutonium and 241Am, 242mAm and 243Am actinides, respectively.

The effects of simulating a realistic eye model on the eye dose of an adult male undergoing head computed tomography.

In head computed tomography, radiation upon the eye lens (as an organ with high radiosensitivity) may cause lenticular opacity and cataracts. Therefore, quantitative dose assessment due to exposure of the eye lens and surrounding tissue is a matter of concern. For this purpose, an accurate eye model with realistic geometry and shape, in which different eye substructures are considered, is needed. To calculate the absorbed radiation dose of visual organs during head computed tomography scans, in this study, an existing sophisticated eye model was inserted at the related location in the head of the reference adult male phantom recommended by the International Commission on Radiological Protection (ICRP). Then absorbed doses and distributions of energy deposition in different parts of this eye model were calculated and compared with those based on a previous simple eye model. All calculations were done using the Monte Carlo code MCNP4C for tube voltages of 80, 100, 120 and 140 kVp. In spite of the similarity of total dose to the eye lens for both eye models, the dose delivered to the sensitive zone, which plays an important role in the induction of cataracts, was on average 3% higher for the sophisticated model as compared to the simple model. By increasing the tube voltage, differences between the total dose to the eye lens between the two phantoms decrease to 1%. Due to this level of agreement, use of the sophisticated eye model for patient dosimetry is not necessary. However, it still helps for an estimation of doses received by different eye substructures separately.

Benchmarking of JEFF-3.2, FENDL-3.0 and TENDL-2014 evaluated data for tungsten with 14.8 MeV neutrons

Integral experiments on tungsten slab samples were carried out on the D-T neutron source facility at China Institute of Atomic Energy. Leakage neutron spectra from the irradiated tungsten target were measured by the time-of-flight technique. Accuracy of the nuclear data for tungsten was examined by comparing the measured neutron spectra with the leakage neutron spectra simulated using the MCNP-4C code with evaluated nuclear data of the JEFF-3.2, FENDL-3.0 and TENDL-2014 libraries. The results show that the calculations with JEFF-3.2 agree well with the measurements in the whole energy range and all angles, whereas the spectra calculated with FENDL-3.0 and TENDL-2014 have some discrepancies with the experimental data.

Measurement and Monte Carlo simulation of γ-ray dose rate in high-exposure building materials

Natural radioactivity radionuclides in building materials, such as 226Ra, 232Th and 40K, cause indoor exposure due to their gamma-rays. In this research, in a standard dwelling room (5.0 m × 4.0 m × 2.8 m), with the floor covered by various granite stones, was set up to simulate the dose rates from the radionuclides using MCNP4C code. Using samples of granite building products in Iran, activities of the 226Ra, 232Th and 40K were measured at 3.8–94.2, 6.5–172.2 and 556.9–1529.2 Bq kg−1, respectively. The simulated dose rates were 26.31–184.36 nGy h−1, while the measured dose rates were 27.70–204.17 nGy h−1. With the results in good agreement, the simulation is suitable for any kind of dwelling places.

Gamma-ray shielding study of light to heavyweight concretes using MCNP-4C code

In this work, linear and mass attenuation coefficients, half and tenth-value layers, effective atomic number and electron density of different types of concretes were determined at 316.51, 468.07, 511, 662, 1173 and 1332 keV using MCNP-4C code and WinXCom programs. The MCNP-4C and WinXCom results agreed well with each other, with differences of <±1.9%. The results agreed with available experimental data, too, with differences of <±6%. The MCNP-4C results showed better agreement with the experimental data than the WinXCom results. Also, it was found that the effective electron density of studied concretes varies in the range of (2.83–3.2) × 1023 electron/g.

Determining the mass attenuation coefficients for some polymers using MCNP code: A comparison study

In this study, the mass attenuation coefficient for some polymers has been determined using MCNP4C code. The simulation results were compared with available experimental data and predictions from the XCOM program in the gamma ray energies, 59.5, 279.1, 511, 661.6, 662, 1115.5, 1173.2, 1274.5 and 1332.5 keV. Comparison was also made with results obtained by Singh et al. Obtained results were highly in accordance with XCOM data. The results showed that the geometry used here is quite convenient in comparison to Singh's study in the calculation of mass attenuation coefficients of the polymers, especially for low energy photons when compared with the XCOM data. It is also indicated that the MCNP4C model used in this study can be a good candidate to be employed in order to make additional characteristics of different materials. Moreover, this method is useful in cases where no analogous experimental data exist especially for the polymers.

Gamma Ray Shielding Study of Barium–Bismuth–Borosilicate Glasses as Transparent Shielding Materials using MCNP-4C Code, XCOM Program, and Available Experimental Data

Abstract In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10 th value layer values of barium–bismuth–borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium–bismuth–borosilicate glasses have good gamma ray shielding properties from the shielding point of view.