• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.403.1-403
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.31-31
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• 1999

• The Journal of Engineering Geology
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• v.27 no.4
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• pp.429-438
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• 2017

Monte Carlo N-Particle (MCNP) modeling algorithm based on the Monte Carlo method was used to perform the simulation of neutron logging in order to increase the reliability and utilization of neutron logs applied in geological and resource engineering fields. To perform the simulation using MCNP, we used a realistic three-dimensional configuration of neutron sonde and formation. Validation of the modeling was confirmed by comparing the calibration curves of sonde manufacture with those calculated by MCNP modeling. After the validation, lithology effects, pore fluid effects, borehole diameter change, casing effect, and effects of borehole water level were investigated through modeling experiments. Numerical tests indicate that changes in neutron count ratio according to the lithology were quantitatively understood. In case of a borehole with a diameter of 3 inches, ratio of counting rates was higher than expected to be interpreted as borehole fluid has small effects on neutron logging. Effect of casing was also small in general, particular when porosity increases. Since modeling results above the groundwater level showed a tendency opposite to those below the groundwater level, neutron logs can be used to detect groundwater level. The modeling results simulated in this study for various borehole environments are expected to be used for data processing and interpretation of neutron log.

• 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.

• 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.

• 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.

• 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.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.1994-1998
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• 2023

This study quantitatively evaluated the source term of a linear accelerator according to target thickness for a 6-20 MeV electron beam using MCNP6. The elements of the target were tungsten and copper, and a composite target and single target were simulated by setting different thickness parameters depending on energy. The accumulation of energy generated through interaction with the collided target was evaluated at 0.1-mm intervals, and F6 tally was used. The results indicated that less than 3% reference error was maintained according to the MCNP recommendations. At 6, 8, 10, 15, 18, and 20 MeV, the energy accumulation peaks identified for each target were 0.3 mm in tungsten, 1.3 mm in copper, 1.5 mm in copper, 0.5 mm in tungsten, 0.5 mm in tungsten, and 0.5 mm in tungsten. For 8 and 10 MeV in a single target consisting only of copper, the movement of electrons was confirmed at the end of the target, and the proportion of escaped electrons was 0.00011% and 0.00181%, respectively.

• 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.