• Journal of Welding and Joining
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• 제23권2호
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• pp.75-80
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• 2005

The effect of chemical constituents of $(Cr,\;Fe)_7C_3$ carbide phase on its hardness in the chromium-carbide type Cr white iron hardfacing weld deposits has been investigated. In order to examine $(Cr,\;Fe)_7C_3$ carbide phase, a series of filler metals with varying chromium contents was used. The alloys were deposited once or twice on a mild steel plate using the self?shielding flux cored arc welding process. The hardness of $(Cr,\;Fe)_7C_3$ carbide phase was measured by the micro-Vickers hardness test. It was shown that hardness of $(Cr,\;Fe)_7C_3$ carbide phase increased with increasing Cr content in $(Cr,\;Fe)_7C_3$ carbide phase. This behavior of the hardness of $(Cr,\;Fe)_7C_3$ carbide phase was explained by the types of chemical bonds that hold atoms together in $(Cr,\;Fe)_7C_3$ carbide phase.

• 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.

• 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.

• 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.

• 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.

• 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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• 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.

• 대한방사선기술학회지:방사선기술과학
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• 제45권6호
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• pp.523-529
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• 2022

Gamma spectroscopy analysis is widely used for radioactivity analysis, and various factors are required for radioactivity calculations. Among the factors, K3 for each sample significantly influences the results. The previous methods of correcting the self-absorption effect include a computational simulation method and a method that requires making a CRM(certified reference material) identical to the sample medium. However, the above methods have limitations when used in small institutions because they require specialized program utilization skills or high manufacturing costs and large facilities. The aim of this study is to develop a method that can be easily and rapidly applied to radioactivity analysis. After filling the beaker with water, we placed the radiation source in a uniform position and used the measured value as the benchmark. Next, a correction factor was derived based on the difference in the radiation source count of the benchmark and the identically measured sample. For the radiation source, Eu-152, which emits a broad range of energy within the measurement range of gamma rays, and Cs-134 and Cs-137, which are indicator nuclides in environmental radiation analysis, were used. The sample was selected within the density range of 0.26-2.11 g/cm³, and the correction factor was derived by calculating the count difference of each sample compared to the reference value of water. This study presents a faster and more convenient method than the existing research methods for determining the self-absorption effect correction, which has become increasingly necessary.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1204-1212
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• 2024

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.