• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.11a
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• pp.114-115
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• 2018

To develop an effective shielding material for spent fuel that emits fast neutrons is necessary. In this study, thermal neutron and fast neutron shielding performance of polyethylene coated boron carbide-incorporated cement paste was quantitatively analyzed by Monte Carlo N-Particle transport code (MCNP) simulations. As the results of the simulations, fast neutrons were effectively shielded through large quantity of hydrogen and boron elements in polyethylene and boron carbide.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4447-4464
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• 2023

This paper focuses on the development of a new computational model of the CNESTEN's TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor k_eff. For that purpose, the T4 model was used to calculate the sensitivity profiles of the k_eff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the k_eff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the k_eff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the k_eff is not negligible and should be taken into account when performing S/U analysis.

• Progress in Medical Physics
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• v.19 no.1
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• pp.21-34
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• 2008

The parallel Monte Carlo electron and photon transport (PMCEPT) code [Kum and Lee, J. Korean Phys. Soc. 47, 716 (2006)] for calculating electron and photon beam doses has been developed based on the three dimensional geometry defined by computed tomography (CT) images and implemented on the Beowulf PC cluster. Understanding the limitations of Monte Carlo codes is useful in order to avoid systematic errors in simulations and to suggest further improvement of the codes. We evaluated the PMCEPT code by comparing its normalized depth doses for electron and photon beams with those of MCNP5, EGS4, DPM, and GEANT4 codes, and with measurements. The PMCEPT results agreed well with others in homogeneous and heterogeneous media within an error of $1{\sim}3%$ of the dose maximum. The computing time benchmark has also been performed for two cases, showing that the PMCEPT code was approximately twenty times faster than the MCNP5 for 20-MeV electron beams irradiated on the water phantom. For the 18-MV photon beams irradiated on the water phantom, the PMCEPT was three times faster than the GEANT4. Thus, the results suggest that the PMCEPT code is indeed appropriate for both fast and accurate simulations.

• Journal of Sensor Science and Technology
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• v.25 no.4
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• pp.264-270
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• 2016

Self-powered neutron detector (SPND) is a sensor to monitor a neutron flux proportional to a reactor power of the nuclear power plants. Since an SPND is usually installed in the reactor core and does not require additional outside power, it generates electrons itself from interaction between neutrons and a neutron-sensitive material called an emitter, such as rhodium and vanadium. This paper presents the simulations of the depletion sensitivity evaluations based on MCNP models of rhodium and vanadium SPNDs and light water reactor fuel assembly. The evaluations include the detail geometries of the detectors and fuel assembly, and the modeling of rhodium and vanadium emitter depletion using MCNP and ORIGEN-S codes, and the realistic energy spectrum of beta rays using BETA-S code. The results of the simulations show that the lifetime of an SPND can be prolonged by using vanadium SPND than rhodium SPND. Also, the methods presented here can be used to analyze a life-time of those SPNDs using various emitter materials.

• Journal of Radiation Protection and Research
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• v.22 no.4
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• pp.289-298
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• 1997

Time-consuming experiments have been required in the development of neutron moisture gauge to induce a relation between the water content in soil and the neutron counts. Applying a monte carlo computer code to simulate the experiments of neutron moisture gauging may contribute to reduce time and efforts for experiments and produce a calibration equation which is more applicable to soil in general. In this study MCNP4A, a monte carlo computer code, was employed to simulate soil experiments and the simulated results were compared with experimental ones. The comparative study showed that MCNP4A is applicable to simulate the experiments and calibration equation can be obtained through simulations. Effects of dry density changes were also studied.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1036-1044
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• 2023

The combination of MCNP6 and the FISPACT codes was used to predict inventories of radioisotopes produced by neutron exposure of a sample in a research reactor. The detailed MCNP6 model of the Budapest Research Reactor and the specific irradiation geometry of the NAA channel was established, while realistic material cards were specified based on concentrations measured by PGAA and NAA, considering the precursor elements of all significant radioisotopes. The energy- and spatial distributions of the neutron field calculated by MCNP6 were transferred to FISPACT, and the resulting activities were validated against those measured using neutron-irradiated small and bulky targets. This approach is general enough to handle different target materials, shapes, and irradiation conditions. A general agreement within 10% has been achieved. Moreover, the method can also be made applicable to predict the activation properties of the near-vessel concrete of existing nuclear installations or assist in the optimal construction of new nuclear power plant units.

• Korean Journal of Environmental Biology
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• v.24 no.4
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• pp.387-391
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• 2006

Computational and experimental dosimetry of Henschke applicator with respect to high dose rate brachytherapy using the MIRD phantom and a remote control afterloader were performed. A comparison of computational dosimetry was made between the simulated Monte Carlo dosimetry and GAMMADOT brachytherapy Planning system's dosimetry. Dose measurements was performed using ion chamber in a water phantom. Dose rates are calculated using Monte Carlo code MCNP4B and the GAMMADOT. Thecomputational models include the detailed geometry of Ir-192 source, tandem tube, and shielded ovoids for accurate estimation. And transit dose delivered during source extension to and retraction from a given dwell position was estimated by Monte Carlo simulations. Point doses at ICRU bladder/rectal pointswhich have been recommened by ICRU 38 was assessed. Calculated and measured dose distribution data agreed within 4% each other. The shielding effect of ovoids leads to 19% and 20% dose reduction at bladder surface and rectal points.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4019-4025
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• 2023

Recently, a new monkey computational phantom, called Visible Monkey, was developed for non-ionizing radiation studies in animal research. In this study, we extended its applications to ionizing radiation studies by implementing the voxel model of the Visible Monkey into three general-purpose Monte Carlo (MC) codes: MCNP6, PHITS, and Geant4. The implementation work for MCNP and PHITS was conducted using the LATTICE, UNIVERSE, and FILL cards. The G4VNestedParameterisation class was used for Geant4. Then, organ dose coefficients (DCs) for idealized photon beams in the antero-posterior direction were calculated using the three codes and compared, showing excellent agreement (differences <3%). Additionally, organ DCs in other directions (postero-anterior, left-lateral, and right-lateral) were calculated and compared with those of the newborn and 1-year-old reference phantoms. Significant differences were observed (e.g., the stomach DC of the monkey was 5-fold greater than that of the 1-year-old phantom at 0.03 MeV) while the differences tended to decrease with increasing energy (mostly <20% at 10 MeV). The results of this study allows conducting MC simulations using the Visible Monkey to estimate organ-level doses, which should be valuable to support/improve monkey experiments involving ionizing radiation exposures.

• Journal of Radiation Protection and Research
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• v.48 no.4
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• pp.184-196
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• 2023

Background: Radiation protection is crucial in various fields due to the harmful effects of radiation. Shielding is used to reduce radiation exposure, but gamma radiation poses challenges due to its high energy and penetration capabilities. Materials and Methods: This work investigates the radiation shielding properties of polyvinylidene fluoride (PVDF) samples containing different weight fraction of tungsten carbide (WC), tungsten trioxide (WO₃), and tungsten disulfide (WS₂). Parameters such as the mass attenuation coefficient (MAC), half-value layer (HVL), mean free path (MFP), effective atomic number (Z_eff), and macroscopic effective removal cross-section for fast neutrons (Σ_R) were calculated using the Phy-X/PSD software. EpiXS simulations were conducted for MAC validation. Results and Discussion: Increasing the weight fraction of the additives resulted in higher MAC values, indicating improved radiation shielding. PVDF-xWC showed the highest percentage increase in MAC values. MFP results indicated that PVDF-0.20WC has the lowest values, suggesting superior shielding properties compared to PVDF-0.20WO₃ and PVDF-0.20WS₂. PVDF-0.20WC also exhibited the highest Z_eff values, while PVDF-0.20WS₂ showed a slightly higher increase in Z_eff at energies of 0.662 and 1.333 MeV. PVDF-0.20WC has demonstrated the highest Σ_R value, indicating effective shielding against fast neutrons, while PVDF-0.20WS₂ had the lowest Σ_R value. The Monte Carlo N-Particle Transport version 5 (MCNP5) simulations showed that PVDF-xWC attenuates gamma radiation more than pure PVDF, significantly decreasing the dose equivalent rate. Conclusion: Overall, this research provides insights into the radiation shielding properties of PVDF mixtures, with PVDF-xWC showing the most promising results.