• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.20 no.4
• /
• pp.511-521
• /
• 2022

Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y^-1 were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.

• Nuclear Engineering and Technology
• /
• v.55 no.9
• /
• pp.3506-3513
• /
• 2023

In this paper, we demonstrate the neutron attenuation of dry cask shielding materials using the S670e helium-4 detector manufactured by Arktis Radiation Ltd. In particular, two materials expected to be applied to the TN-32 dry cask manufactured by ORANO Korea and KORAD-21 by the Korea Radioactive Waste Agency (KORAD) were utilized. The measured neutron attenuation was compared with our Monte Carlo N-Particle Transport simulation results, and the difference is given as the root mean square (RMS). For the fast neutron case, a rapid decline in neutron counts was observed as a function of increasing material thickness, exhibiting an exponential relationship. The discrepancy between the experimentally acquired data and simulation results for the fast neutron was maintained within a 2.3% RMS. In contrast, the observed thermal neutron count demonstrated an initial rise, attained a maximum value, and exhibited an exponential decline as a function of increasing thickness. In particular, the discrepancy between the measured and simulated peak locations for thermal neutrons displayed an RMS deviation of approximately 17.3-22.4%. Finally, the results suggest that a minimum thickness of 5 cm for Li-6 is necessary to achieve a sufficiently significant cross-section, effectively capturing incoming thermal neutrons within the dry cask.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4084-4094
• /
• 2022

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. The first cycle of FAST testing was completed and four rodlets were removed for the initial post irradiation examination (PIE). The rodlet design and irradiation conditions were evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % fissions per initial metal atoms [%FIMA]), control rodlets, and a helium-bonded annular rodlet (4.7 %FIMA). Nondestructive PIE has been completed and includes visual inspection, neutron radiography and gamma scanning of the FAST capsules and rodlets. Radiography confirmed the integrity of the experiments, revealed that the annulus in the annular fuel was filled at a modest burnup (4.7 %FIMA), and indicated potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior, except for cesium in the He-bonded annular fuel. Future destructive PIE will be necessary to fully interpret the effects of accelerated irradiation on U-Zr metallic fuel behavior.

• Nuclear Engineering and Technology
• /
• v.54 no.2
• /
• pp.523-531
• /
• 2022

In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U₂₃₅(n, f), $ U_{235}(n\bar{\nu})$ and H₁(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.

• Radiation Oncology Journal
• /
• v.19 no.1
• /
• pp.66-73
• /
• 2001

Purpose : For the research of Boron Neutron Capture Therapy (BNCT), fast neutrons generated from the MC-50 cyclotron with maximum energy of 34.4 MeV in Korea Cancer Center Hospital were moderated by 70 cm paraffin and then the dose characteristics were investigated. Using these results, we hope to establish the protocol about dose measurement of epi-thermal neutron, to make a basis of dose characteristic of epi-thermal neutron emitted from nuclear reactor, and to find feasibility about accelerator-based BNCT. Method and Materials : For measuring the absorbed dose and dose distribution of fast neutron beams, we used Unidos 10005 (PTW, Germany) electrometer and IC-17 (Far West, USA), IC-18, ElC-1 ion chambers manufactured by A-150 plastic and used IC-l7M ion chamber manufactured by magnesium for gamma dose. There chambers were flushed with tissue equivalent gas and argon gas and then the flow rate was S co per minute. Using Monte Carlo N-Particle (MCNP) code, transport program in mixed field with neutron, photon, electron, two dimensional dose and energy fluence distribution was calculated and there results were compared with measured results. Results : The absorbed dose of fast neutron beams was $6.47\times10^{-3}$ cGy per 1 MU at the 4 cm depth of the water phantom, which is assumed to be effective depth for BNCT. The magnitude of gamma contamination intermingled with fast neutron beams was $65.2{\pm}0.9\%$ at the same depth. In the dose distribution according to the depth of water, the neutron dose decreased linearly and the gamma dose decreased exponentially as the depth was deepened. The factor expressed energy level, $D_{20}/D_{10}$, of the total dose was 0.718. Conclusion : Through the direct measurement using the two ion chambers, which is made different wall materials, and computer calculation of isodose distribution using MCNP simulation method, we have found the dose characteristics of low fluence fast neutron beams. If the power supply and the target material, which generate high voltage and current, will be developed and gamma contamination was reduced by lead or bismuth, we think, it may be possible to accelerator-based BNCT.

• Radiation Oncology Journal
• /
• v.28 no.4
• /
• pp.224-230
• /
• 2010

Purpose: We designed the aft-multiple-slit (AMS) system to reduce scatter in cone-beam computed tomography (CBCT). As a preliminary study, we performed a Monte Carlo N-Particle Transport Code (MCNP) simulation to verify the effectiveness of this system. Materials and Methods: The MCNPX code was used to build the AMS geometry. An AMS is an equi-angled arc to consider beam divergence. The scatter-reduced projection images were compared with the primary images only and the primary plus scatter radiation images with and without AMS to evaluate the effectiveness of scatter reduction. To obtain the full 2 dimensional (2D) projection image, the whole AMS system was moved to obtain closed septa of the AMS after the first image acquisition. Results: The primary radiation with and without AMS is identical to all the slit widths, but the profiles of the primary plus scattered radiation varied according to the slit widths in the 2D projection image. The average scatter reduction factors were 29%, 15%, 9%, and 8% when the slit widths were 5 mm, 10 mm, 15 mm, and 20 mm, respectively. Conclusion: We have evaluated the scatter reduction effect of the AMS in CBCT imaging using the Monte Carlo (MC) simulations. A preliminary study based on the MCNP simulations showed a mount of scatter reduction with the proposed system.

• Proceedings of the Korean Vacuum Society Conference
• /
• 2011.08a
• /
• pp.312-312
• /
• 2011

In SiH4/H2 discharge for growth process of hydrogenated amorphous silicon (a-Si:H), silane polymers, produced by SiH2 + Sin-1H2n ${\rightarrow}$ SinH2n+2, have no reactivity on the film-growing surface. However, under the SiH2 rich condition, high silane reactive species (HSRS) can be produced by electron collision to silane polymers. HSRS, having relatively strong reactivity on the surface, can react with dangling bond and form Si-H2 networks which have a close correlation with photo-induced degradation of a-Si:H thin film solar cell [1]. To find contributions of suggested several external plasma conditions (pressure, frequency and ratio of mixture gas) [2,3] to suppressing productions of HSRS, some plasma characteristics are studied by numerical methods. For this study, a zero-dimensional global model for SiH4/H2 discharge and a one-dimensional particle-in-cell Monte-Carlo-collision model (PIC-MCC) for pure SiH4 discharge have been developed. Densities of important reactive species of SiH4/H2 discharge are observed by means of the global model, dealing 30 species and 136 reactions, and electron energy probability functions (EEPFs) of pure SiH4 discharge are obtained from the PIC-MCC model, containing 5 charged species and 15 reactions. Using global model, SiH2/SiH3 values were calculated when pressure and driving frequency vary from 0.1 Torr to 10 Torr, from 13.56 MHz to 60 MHz respectively and when the portion of hydrogen changes. Due to the limitation of global model, frequency effects can be explained by PIC-MCC model. Through PIC-MCC model for pure SiH4, EEPFs are obtained in the specific range responsible for forming SiH2 and SiH3: from 8.75 eV to 9.47 eV [4]. Through densities of reactive species and EEPFs, polymerization reactions and production of HSRS are discussed.

• Journal of the Korean Applied Science and Technology
• /
• v.32 no.2
• /
• pp.330-338
• /
• 2015

The influences of fluorescence, scattering, and flocculation in turbid material by light scattering of N-propyl-N,N-dimethylethanolamine, fluorescence agent and absorption agent were interpreted for the scattered fluorescence intensity and wavelength. They have been studied the molecular properties by the spectroscopy of laser induced fluorescence (LIF) and flocculation. The effects of optical properties in scattering media have been found by the optical parameters(${\mu}_s$, ${\mu}_a$, ${\mu}_t$). Flocculation is an important step in many solid-liquid separation processes and is widely used. When two particles approach each other, interactions of several colloid particles can come into play which may have major effect on the flocculation and LIF process. The values of scattering coefficient ${\mu}_s$ are large by means of the increasing scattering of scatterer, The values have been found that the slope decays exponentially as a function of concentration from laser source to detector by our experimental result. It may also aid in designing the best model for oil chemistry, bio-pharmaceutical, laser medicine and application of medical engineering on LIF and coagulation in particle transport mode.

• Journal of Radiation Protection and Research
• /
• v.41 no.1
• /
• pp.71-79
• /
• 2016

Background: It is necessary to analyze radioactivity of naturally occurring radioactive materials (NORM) in products to ensure radiological safety required by Natural Radiation Safety Management Act. The pretreatments for the existing analysis methods require high technology and time. Such destructive pretreatments including grinding and dissolution of samples make impossible to reuse products. We developed a rapid and simple procedure of radioactivity analysis for thorium series in the products containing NORM. Materials and Methods: The developed method requires non-destructive or minimized pretreatment. Radioactivity of the product without pretreatment is initially measured using gamma spectroscopy and then the measured radioactivity is adjusted by considering material composition, mass density, and geometrical shape of the product. The radioactivity adjustment can be made using scaling factors, which is derived by radiation transport Monte Carlo simulation. Necklace, bracelet, male health care product, and tile for health mat were selected as representative products for this study. The products are commonly used by the public and directly contacted with human body and thus resulting in high radiation exposure to the user. Results and Discussion: The scaling factors were derived using MCNPX code and the values ranged from 0.31 to 0.47. If radioactivity of the products is measured without pretreatment, the thorium series may be overestimated by up to 2.8 times. If scaling factors are applied, the difference in radioactivity estimates are reduced to 3-24%. Conclusion : The developed procedure in this study can be used for other products with various materials and shapes and thus ensuring radiological safety.

• Progress in Medical Physics
• /
• v.28 no.3
• /
• pp.77-82
• /
• 2017

We proposed the objective ROC analysis method based on the setting of threshold value for evaluation of single photon emission computed tomography (SPECT) image. This proposed ROC analysis method uses the quantification computational threshold value to each signal on the SPECT image. The SPECT images for this study were acquired by using Monte Carlo n-particle extended simulation code (MCNPX, Ver. 2.6.0, Los Alamos National Laboratory, USA). The basic SPECT detectors and specific water phantom were realized in the simulation, and we could get the simulation results by the simulation operation. We tried to analyze the reconstructed images using threshold value application based objective ROC method. We can get the accuracy information of reconstructed region in the image. This proposed ROC technique can be helpful when we have to evaluate the weak signal for the NM image. In this study, the proposed threshold value based computational ROC analysis method can provide better objectivity than the conventional ROC analysis method.