• Nuclear Engineering and Technology
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• 제48권3호
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• pp.773-777
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• 2016

A betavoltaic battery was prepared using radioactive $^{63}Ni$ attached to a three-dimensional single trenched P-N absorber. The optimum thickness of a $^{63}Ni$ layer was determined to be approximately $2{\mu}m$, considering the minimum self-shielding effect of beta particles. Electroplating of radioactive $^{63}Ni$ on a nickel (Ni) foil was carried out at a current density of $20mA/cm^2$. The difference of the short-circuit currents ($I_{sc}$) between the pre- and post-deposition of $^{63}Ni$ (16.65 MBq) on the P-N junction was 5.03 nA, as obtained from the I-V characteristics. An improved design with a sandwich structure was provided for enhancing performance.

• 한국표면공학회:학술대회논문집
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• 한국표면공학회 2014년도 추계학술대회 논문집
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• pp.45-45
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• 2014

The importance of magnesium alloys has significantly increased due to their low density, high strength/weight ratio, very good electromagnetic shielding features and good recyclability. However, unfortunately, Mg alloys are very susceptible to corrosion due to their high chemically activities (= -2.356 V vs. NHE at $25^{\circ}C$), hence, most commercial Mg alloys require corrosion protective coatings. Organic coating such as painting, powder coating and electrophoretic deposition of paint (E-paint) is typically used in the final stages of the coating process of Mg alloys. In this study, effect of pre-immersion time on the deposition of E-paint on AZ31 Mg alloy was investigated. It was found that during pre-immersion time, AZ31 Mg alloy rapidly reacts with E-paint solution and paint can be self-deposited on the AZ31 surface without applying of electric current. The pore size on the E-painted AZ31 Mg alloy increased with increasing pre-immersion time from 0 to 5 min. Both adhesion and corrosion resistance of E-painted AZ31 Mg alloy decreased with increasing pre-immersion time. The best E-paint AZ31 Mg alloy, which showed stronger adhesion after water immersion test and good corrosion resistance, was started to deposit after 5 s of pre-immersion time.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.921-928
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• 2013

A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.340-355
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• 2018

The methods and performance of a fast reactor multigroup cross section (XS) generation code EXUS-F are described that is capable of directly processing Evaluated Nuclear Data File format nuclear data files. RECONR of NJOY is used to generate pointwise XS data, and Doppler broadening is incorporated by the Gauss-Hermite quadrature method. The self-shielding effect is incorporated in the ultrafine group XSs in the resolved and unresolved resonance ranges. Functions to generate scattering transfer matrices and fission spectrum matrices are realized. The extended transport approximation is used in zero-dimensional calculations, whereas the collision probability method and the method of characteristics are used for one-dimensional cylindrical geometry and two-dimensional hexagonal geometry problems, respectively. Verification calculations are performed first for various homogeneous mixtures and cylindrical problems. It is confirmed that the spectrum calculations and the corresponding multigroup XS generations are performed adequately in that the reactivity errors are less than 50 pcm with the McCARD Monte Carlo solutions. The nTRACER core calculations are performed with the EXUS-F-generated 47 group XSs for the two-dimensional Advanced Burner Reactor 1000 benchmark problem. The reactivity error of 160 pcm and the root mean square error of the pin powers of 0.7% indicate that EXUF-F generates properly the broad-group XSs.

• 대한방사선기술학회지:방사선기술과학
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• 제41권6호
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• pp.581-586
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• 2018

This study was conducted to analyze factors Influencing Protective Behavior against Radiation Exposure using questionnaires for 231 radiological technologists working in Computed Tomography(CT) examination room with high radiation dose in diagnostic radiology field. Statistical analysis of the collected data revealed that the reasons for partially shielding the examination part in the CT scan were the lack of protective equipment, securing of radiation justification, being annoying and maybe not being harm to adults in order. It was also revealed that the variables influencing the protective behavior were protective behavior against radiation harm, self-efficacy, protective environment, organization culture, protective knowledge and protective instrument in order. The higher the radiological protective environment(${\beta}=0.245$) and the lower the radiological protective knowledge(${\beta}=-0.034$), the more influential the protective behavior against radiation harm was. In this study, it was shown that non examination parts were not shielded in the CT scan. Therefore, it is necessary to improve the level of protective environment, to cultivate knowledge to improve the protective behavior against radiation harm and to have an intervention strategy for concrete action.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2221-2229
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• 2022

In this research, for improving the measurement performance of Prompt Gamma-ray Neutron Activation Analysis (PGNAA) set-up, a new optimization method for set-up design was proposed and investigated. At first, the calculation method for Signal-to-Noise Ratio (SNR) was proposed. Since the SNR could be calculated and quantified accurately, the SNR was chosen as the evaluation parameter in the new optimization method. For discussing the feasibility of the SNR optimization method, two kinds of PGNAA set-ups were designed in the MCNP code, based on the SNR optimization method and the previous signal optimization method, respectively. Meanwhile, the single element spectra analysis method was proposed, and the analysis effect of single element spectra as well as element sensitivity were used for comparing the measurement performance. Since the simulation results showed the better measurement performance of set-up designed by SNR optimization method, the experimental set-ups were built for the further testing, finally demonstrating the feasibility of the SNR optimization method for PGNAA setup design.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2852-2859
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• 2020

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3525-3534
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• 2023

Non-uniformity of Pu and Cm composition in used nuclear fuel was analyzed to determine its effect on Pu accountancy in pyroprocessing, while employing the Pu-to-²⁴⁴Cm ratio method. Burnup simulation of a typical pressurized water reactor fuel assembly, required for the analysis, was carried out using MCNP code. Used fuel nuclide composition, as a function of nine axial and two radial meshes, were evaluated. The axial variation of neutron flux and self-shielding effects were found to affect the uniformity of Pu and Cm compositions and in turn the Pu-to-²⁴⁴Cm ratio. However, the results of the study showed that these non-uniformities do not affect the use of Pu-to-²⁴⁴Cm ratio method for Pu accountancy, if the measurement samples are drawn from the voloxidized powder at the feed step of pyroprocessing. 'Material Unaccounted For' and its uncertainty estimates are also presented for a pyrprocessing facility to verify safeguards monitoring requirements of the IAEA.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4518-4526
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• 2023

NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi'an Jiaotong University. In past few years, improvements have been implemented in TULIP code which is the cross-section generation module of NECP-SARAX, including the treatment of resonance interface, considering the self-shielding effect in non-resonance energy range, hyperfine group method and nuclear library with thermal scattering law. Previous studies show that NECP-SARAX has high performance in both fast and thermal spectrum system analysis. The accuracy of TULIP code in fast and thermal spectrum system analysis is demonstrated preliminarily. However, a systematic verification and validation is still necessary. In order to validate the applicability of TULIP code for full energy range, 147 fast spectrum critical experiment benchmarks and 170 thermal spectrum critical experiment benchmarks were selected from ICSBEP and used for analysis. The keff bias between TULIP code and reference value is less than 300 pcm for all fast spectrum benchmarks. And that bias keeps within 200 pcm for thermal spectrum benchmarks with neutron-moderating materials such as polyethylene, beryllium oxide, etc. The numerical results indicate that TULIP code has good performance for the analysis of fast and thermal spectrum system.

• 대한방사선치료학회지
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• 제26권2호
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• pp.265-272
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• 2014

목 적 : 전뇌 방사선 치료 시 산란선으로 인하여 영향을 받는 갑상선의 피폭선량을 감소시키기 위해 차폐체를 사용하여 갑상선의 차폐 효과를 평가하고자 한다. 대상 및 방법 : 갑상선의 피폭선량을 측정하기 위해 선형가속기(Clinac iX. VARIAN, USA)를 이용하여 6 MV X선, 300 cGy를 인체모형팬텀에 대향 2문 조사하였다. 갑상선의 입사표면선량을 측정하기 위해 인체모형팬텀의 10번째 슬라이스 표면에 유리선량계 다섯 개를 1.5 cm 간격으로 위치시킨 후 차폐체 미사용, bismuth 차폐체 사용, 0.5 mmPb 차폐체 사용, 자체 제작한 1.0 mmPb 차폐체를 사용하여 각각 5회씩 측정하여 평균값을 산출하였다. 또한, 같은 위치에서 갑상선 심부선량을 측정하기 위해서 인체모형팬텀의 10번째 슬라이스 2.5 cm 깊이에서 유리선량계 다섯 개를 1.5 cm 간격으로 위치시킨 후 차폐체 미사용, bismuth 차폐체 사용, 0.5 mmPb 차폐체 사용, 자체 제작한 1.0 mmPb 차폐체를 사용하여 각각 5회씩 측정하여 평균값을 산출하였다. 결 과 : 갑상선의 입사표면선량은 차폐체 미사용 시 44.89 mGy로 측정되었고, bismuth 차폐체는 36.03 mGy, 0.5 mmPb 차폐체는 31.03 mGy, 자체 제작한 1.0 mmPb 차폐체는 23.21 mGy로 측정되었다. 또한, 갑상선의 심부선량은 차폐체 미사용 시 36.10 mGy로 측정되었고, bismuth 차폐체는 34.52 mGy, 0.5 mmPb 차폐체는 32.28 mGy, 자체 제작한 1.0 mmPb 차폐체는 25.50 mGy로 측정되었다. 결 론 : 전뇌 방사선 치료 시 방사선 조사면 밖의 영역에서 발생하는 이차 산란 및 누출 선량에 의해 영향을 받는 갑상선에 대하여 차폐체를 사용했을 때 갑상선 심부는 약 11~30%, 갑상선 표면은 약 20~48% 정도의 피폭선량 감소 효과가 나타났다. 따라서 전뇌 방사선 치료 시 갑상선 차폐체를 사용함으로써 갑상선을 효과적으로 보호하며 치료를 시행할 수 있을 것으로 사료된다.