• 방사선산업학회지
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• 제17권3호
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• pp.265-273
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• 2023

To effectively and safely manage the radiation exposure to nuclear power plant (NPP) workers in accidents, major overseas NPP operators such as the United States, Germany, and France have developed and applied realistic 3D model radiation dose assessment software for workers. Continuous research and development have recently been conducted, such as performing NPP accident management using 3D-VR based on As Low As Reasonably Achievable (ALARA) planning tool. In line with this global trend, it is also required to secure technology to manage radiation exposure of workers in Korea efficiently. Therefore, in this paper, it is described the application method and assessment results of radiation exposure scenarios for workers in response to accidents assessment technology, which is one of the fundamental technologies for constructing a realistic platform to be utilized for radiation exposure prediction, diagnosis, management, and training simulations following accidents. First, the post-accident sampling after the Loss of Coolant Accident(LOCA) was selected as the accident and response scenario, and the assessment area related to this work was established. Subsequently, the structures within the assessment area were modeled using MCNP, and the radiation source of the equipment was inputted. Based on this, the radiation dose distribution in the assessment area was assessed. Afterward, considering the three principles of external radiation protection (time, distance, and shielding) detailed work scenarios were developed by varying the number of workers, the presence or absence of a shield, and the location of the shield. The radiation exposure doses received by workers were compared and analyzed for each scenario, and based on the results, the optimal accident response scenario was derived. The results of this study plan to be utilized as a fundamental technology to ensure the safety of workers through simulations targeting various reactor types and accident response scenarios in the future. Furthermore, it is expected to secure the possibility of developing a data-based ALARA decision support system for predicting radiation exposure dose at NPP sites.

• 센서학회지
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• 제25권3호
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• pp.229-234
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• 2016

Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the vanadium (V) SPND has been being developed to be used in OPR1000 nuclear power plants. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina insulator with a cylindrical geometry. An MCNP code was used to simulate some factors (neutron self-shielding factor and beta escape probability from the emitter) and space charge effect of an insulator necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND and contribute to the development of TMI (Top-mount In-core Instrumentation) which might be used in the SMART and SMR.

• Nuclear Engineering and Technology
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• 제43권1호
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• pp.13-18
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• 2011

In this study, simulations were done of a 661.6 keV line from a point source of $^{137}Cs$ housed in a lead shield. When increasing the scattering angle from 60 to 120 degrees with a 6061 aluminum alloy target placed at angles of 30 and 45 degrees to the incident beam, the spectra showed that the single scattering component increases and that the multiple scattering component decreases. The investigation of the single and multiple scattering components was carried out using a MCNP5 simulation code. The component of the single Compton scattering photons is proportional to the target electron density at the point where the scattering occurs. The single scattering peak increases according to the thickness of the target and saturates at a certain thickness. The signal-to-noise ratio was found to decrease according to the target thickness. The simulation was experimentally validated by measurements. These results will be used to determine the best conditions under which this method can be applied to testing electron densities or to assess the thickness of samples to locate defects in them.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.277-284
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• 2017

This paper presents the establishment and characterization of a neutron calibration field using a bare $^{252}Cf$ source of low neutron source strength in Vietnam. The characterization of the field in terms of neutron flux spectra and neutron ambient dose equivalent rates were performed by Monte Carlo simulations using the MCNP5 code. The anisotropy effect of the source was also investigated. The neutron ambient dose equivalent rates at three reference distances of 75, 125, and 150 cm from the source were calculated and compared with the measurements using the Aloka TPS-451C neutron survey meters. The discrepancy between the calculated and measured values is found to be about 10%. To separate the scattered and the direct components from the total neutron flux spectra, an in-house shadow cone of 10% borated polyethylene was used. The shielding efficiency of the shadow cone was estimated using the MCNP5 code. The results confirmed that the shielding efficiency of the shadow cone is acceptable.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4678-4684
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• 2023

Using the Monte Carlo method, the impact of the angular distribution of the electron source on the dose distribution for the 2.5 MeV ELV electron accelerator was explored. The experiment measured the 3-D dose distribution in the irradiation chamber for electron energies of 1.0 MeV and 2.5 MeV. The simulation used the MCNP6.2 code to evaluate three angular distribution models of the source: a mono-directional beam, a cone shape, and a triangular shape. Of the three models, the triangular shape with angles θ = 30°, φ = 0° best represents the angle of the scan hood through which the electron beam exits. The MCNP6.2 simulation results demonstrated that the triangular model is the most accurate representation of the angular distribution of the electron source for the 2.5 MeV ELV electron accelerator.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2666-2675
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• 2024

This study explores the potential for tritium production in North Korea's 5MWe graphite-moderated reactor, a facility primarily associated with nuclear weapons material production. While existing research on these reactors has largely centered on plutonium, our focus shifts to tritium, a crucial element in boosted fission bombs. Utilizing the MCNP6 code for simulations, the results estimate that North Korea can possibly produce approximately 7-12 g of tritium annually. This translates to the potential production of 1-3 boosted fission bombs each year. By incorporating tritium production into assessments of North Korea's nuclear capabilities, our methodology provides insights into the dynamics of the country's nuclear force, revealing a more diversified and complex composition than previously assumed. The findings significantly aid policymakers, regulatory bodies, and researchers in comprehending potential proliferation risks associated with graphite-moderated reactors and in developing strategies to address the nuclear threat emanating from North Korea.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1349-1354
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• 2018

Inductively Coupled Plasma Atomic Emission Spectroscopy (ICP-AES) is widely used for the determination of trace elements in geological samples in Iran. In this paper, we have calculated the detection limits of neutron activation analysis (NAA) for some of the common elements in such samples utilizing the ORIGEN and MCNP codes and verified the simulations using the experimental results of three soil standard reference materials, namely, G02.SRM, G18.SRM, and G28.SRM. The results show that while the detection limit of ICP-AES method is usually in the mg/kg range, it is represented to the ${\mu}g/kg$ range for most of the elements of interest using the NAA method, and the simulations can be verified in a tolerance range of 20%.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.3044-3050
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• 2021

This study offers a Monte Carlo alternative for computing mass energy absorption coefficients of any material through calculation of photon energy deposited per mass of the sample and the energy flux obtained inside a sample volume. This approach is applied in this study to evaluate mass energy absorption coefficients of some amino acids found in human body at twenty-eight different photon energies between 10 keV and 20 MeV. The simulations involved a pencil beam source modeled to emit a parallel beam of mono-energetic photons toward a 1 mean free path thick sample of rectangular parallelepiped geometry. All the components in the problem geometry were surrounded by a 100 cm vacuum sphere to avoid any interactions in materials other than the absorber itself. The results computed using the Monte Carlo radiation transport packages MCNP6.2 and GAMOS5.1 were checked against the theoretical values available from the tables of XMUDAT database. These comparisons indicate very good agreement and support the conclusion that Monte Carlo technique utilized in this fashion may be used as a computational tool for determining the mass energy absorption coefficients of any material whose data are not available in the literature.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4244-4252
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• 2022

The accuracy of Monte Carlo (MC) simulations in estimating the computed tomography radiation dose is highly dependent on the accuracy of CT scanner model. A system was developed to observe the 3D model intuitively and to calculate the X-ray energy spectrum and the bowtie (BT) filter model more accurately in Monte Carlo N-particle (MCNP). Labview's built-in Open Graphics Library (OpenGL) was used to display basic surfaces, and constructive solid geometry (CSG) method was used to realize Boolean operations. The energy spectrum was calculated by simulating the process of electronic shooting and the BT filter model was accurately modeled based on the calculated shape curve. Physical data from a study was used as an example to illustrate the accuracy of the constructed model. RMSE between the simulation and the measurement results were 0.97% and 0.74% for two filters of different shapes. It can be seen from the comparison results that to obtain an accurate CT scanner model, physical measurements should be taken as the standard. The energy spectrum library should be established based on Monte Carlo simulations with modifiable input files. It is necessary to use the three-segment splicing modeling method to construct the bowtie filter model.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.384-388
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• 2016

Background: For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. Materials and Methods: In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. Results and Discussion: The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. Conclusion: It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.