• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2002.11a
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• pp.379-382
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• 2002

Lately, intensifying screen of the $CaWO_4$ is used to medical treatment and diagnosis of the image. In this paper, we investigated transmission fraction and mass attenuation coefficient of $CaWO_4$ screen about diagnostic x-ray of low energy using MCNP 4C code. Experimentally, for 0.9 mm-$CaWO_4$ screen, the absorbable rate of diagnostic x-ray is more than 95%. according to kVp, the experimental value of mass attenuation coefficient is in a1most agreement with an corrected estimate value of MCNP and the deviation of experimental values is less than ${\pm}7%$. Using the MCNP code through this paper, we can make an estimate of signal and design for construction of the CaWO4/a-Se based digital x-ray image detector and make a good use of the foundation data for development of other materials.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.517-523
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• 2003

The physical properties of polypropylene (PP) membranes under the radiation field were investigated. To calculate radiation flux affecting to PP, it was used MCNP4A Code. The PP membrane and deoxygenation equipment were standardized to bar structure in order to calculate the phonton flux with MCNP4A Code. The change in the properties of the PP membrane to be used in deoxygenation equipment was rarely occurred during the usage work because the radiation level of reactor coolant water was very low level and The doses of radiation workers are very low. From the results, it was found that the Physical properties of PP membranes which used for nuclear power plant reactor coolant water disposal were not rarely changed under the simulated radiation field.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.4
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• pp.351-359
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• 2009

We have developed a computer simulator fur cone-beam computed tomography (CBCT) based on the commercial Monte Carlo code, MCNP. All the functions to generate input files, run MCNP, convert output files to image data, reconstruct tomographs were realized in graphical user-interface form. The performance of the simulator was demonstrated by comparing with the experimental data. Although some discrepancies were observed due to the ignorance of the detailed physics in the simulation, such as scattered X-rays and noise in image sensors, the overall tendency was well agreed between the measured and simulated data. The developed simulator will be very useful for understanding the operation and the better design of CT systems.

• Journal of Radiation Protection and Research
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• v.32 no.1
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• pp.21-26
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• 2007

This study is performed to evaluate the dose rate and to analyze the dose distribution of the gamma irradiation facility (IR-221) by using a Monte Calro simulation, which is helpful of upgrading the radiation processing qualification. Monte Cairo simulation is performed by MCNP4B code. Dose rates were measured at total 369 points with alanine dosimeters to compare the calculation results and the measurements data. The results have shown that the MCNP4B code is very useful to determine the dose distribution of the IR-221 gamma irradiation facility, as the calculation dose rate is within about ${\pm}5%$ of the measurement data. Dosimetry about the gamma irradiation facility usually needs enormous manpower and time. However Monte Cairo calculation method can reduce the tedious dosimetry jobs and improve the irradiation processing qualification, which will probably contribute to obtain the reliability of the irradiation products.

• Journal of Radiation Protection and Research
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• v.19 no.1
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• pp.13-22
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• 1994

The MCNP 4.2 code was used to calculate the thermal neutron flux distributions for $(n,\;{\gamma})$reaction in mainshell, annular plate, and subshell of the calandria of a CANDU 6 plant during operation. The thermal neutron flux distributions in calandria mainshell, annular plate, and subshell were in the range of $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$ which is somewhat higher than the previous estimates calculated by DOT 4.2 code. As an application to shielding analysis, photon dose rates outside the side and bottom shields were calculated. The resulting dose rates at the reactor accessible areas were below design target, $6 {\mu}Sv/h$. The methodology used in this study to evaluate the thermal neutron flux distribution for $(n,\;{\gamma})reaction$ can be applied to radiation shielding analysis of CANDU 6 type plants.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.191-195
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• 2016

Background: The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Materials and Methods: Validation calculations with recent nuclear data evaluations ENDF/BV-II. 0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. Results and Discussion: It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and ${chi}^2$ values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the $k_{eff}$ values. Conclusion: It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1265-1268
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• 2023

One important issue in using radiopharmaceuticals as therapeutic and imaging agents is predicting different organ absorbed dose following their injection. The present study aims at extrapolating dosimetry estimates to a female phantom from the animal data of ⁸⁹Zr radionuclide accumulation using the Sparks-Idogan relationship. The absorbed dose of ⁸⁹Zr radionuclide in different organs of the human body was calculated based on its distribution data in mice using both MIRD method and the MCNP simulation code. In this study, breasts, liver, heart wall, stomach, kidneys, lungs and spleen were considered as source and target organs. The highest and the lowest absorbed doses were respectively delivered to the liver (4.00E-02 and 3.43E-02 mGy/MBq) and the stomach (1.83E-03 and 1.66E-03 mGy/MBq). Moreover, there was a good agreement between the results obtained from both MIRD and MCNP methods. Therefore, according to the dosimetry results, [⁸⁹Zr] DFO-CR011-PET/CT seems to be a suitable for diagnostic imaging of the breast anomalies for CDX-011 targeting gpNMB in patients with TNBC in the future.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.1-11
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• 2004

The hot cell facility for research activities related to the lithium reduction of spent fuel, which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations peformed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}, 2.97{\times}10^{-3} and 1.01{\times}10{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}, 2.99{\times}10^{-3} and 7.88{\times}10^{-2}$ mSv/h, respectively, The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01 mSv/h for the operation area and 0.15 mSv/h for the service (maintenance) area.

• Geophysics and Geophysical Exploration
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• v.15 no.4
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• pp.219-226
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• 2012

This study provides an analysis on results of neutron logging for various borehole environments through numerical simulation based on a Monte Carlo N-Particle (MCNP) code developed and maintained by Los Alamos National Laboratory. MCNP is suitable for the simulation of neutron logging since the algorithm can simulate transport of nuclear particles in three-dimensional geometry. Rather than simulating a specific tool of a particular service company between many commercial neutron tools, we have constructed a generic thermal neutron tool characterizing commercial tools. This study makes calibration chart of the neutron logging tool for materials (e.g., limestone, sandstone and dolomite) with various porosities. Further, we provides correction charts for the generic neutron logging tool to analyze responses of the tool under various borehole conditions by considering brine-filled borehole fluid and void water, and presence of borehole fluid.