• 전자공학회논문지 IE
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• 제45권2호
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• pp.6-12
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• 2008

본 연구에서는 아스팔트 함량의 중요성을 인식하여 현재 법적 규제 면제치인 100[${\mu}Ci$]이하의 방사성 동위원소를 이용한 아스팔트 함량측정기의 개발을 목표로 하였다. 이를 위하여 중성자선원의 위치에 따라 아스팔트 함량에 변화를 주는 3가지의 분야로 나누었다. 먼저, 아스팔트 혼합물과 중성자 선원과의 간격을 줄일 경우, 반사체 설치의 경우, 이력수를 변화시켰을 경우로 나누어서 컴퓨터 시뮬레이션을 통하여 살펴보았으며, 만족스러운 오차범위 결과를 얻어 아스팔트 함량측정기기의 개발을 위한 설계 자료로 활용하고자하였다.

• Journal of Radiation Protection and Research
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• 제43권4호
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• pp.131-136
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• 2018

Background: We investigated the current characteristics of a thin-film Ag electrode on a chemical vapor deposition (CVD) diamond. The CVD diamond is widely recognized as a radiation detection material because of its high tolerance against high radiation, stable response to various dose rates, and good sensitivity. Additionally, thin-film Ag has been widely used as an electrode with high electrical conductivity. Materials and Methods: Considering these properties, the thin-film Ag electrode was deposited onto CVD diamonds with varied deposition thicknesses (${\fallingdotseq}50/98/152/257nm$); subsequently, the surface thickness, surface roughness, leakage current, and photo-current were characterized. Results and Discussion: The leakage current was found to be very low, and the photo-current output signal was observed as stable for a deposited film thickness of 98 nm; at this thickness, a uniform and constant surface roughness of the deposited thin-film Ag electrode were obtained. Conclusion: We found that a CVD diamond radiation detector with a thin-film Ag electrode deposition thickness close to 100 nm exhibited minimal leakage current and yielded a highly stable output signal.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1398-1406
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• 2024

The control rod worth is a key safety parameter required to be measured in commercial pressurized water reactors (PWRs). Conventionally, the control rod worth is measured after reaching the critical state, which occupies the considerable time in the zero-power physics test. In this study, an efficient control-rod worth measurement technique has been proposed based on the improved neutron-source multiplication method, which can be implemented with the source-range detector count rates in the subcritical states. Moreover, the noise reduction technique has been adopted to smooth the large fluctuation existing in the original signals. In order to verify the engineering performance of the proposed measurement technique, the measured source-range detector count rates during the rod withdrawal process before reaching critical state in a CNP1000 reactor have been employed. It demonstrated that almost all estimated results of control rod worth satisfy the engineering acceptance criteria, except one control rod with the relative difference over 10 %, which indicates the capability of the proposed method in estimating control rod worth.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1354-1357
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• 2017

Characterization of special nuclear material may be performed using energy spectroscopy of either the neutron or gamma-ray emissions from the sample. Gamma-ray spectroscopy can be performed relatively easily using high-resolution semiconductors such as high-purity germanium. Neutron spectroscopy, by contrast, is a complex inverse problem. Here, results are presented for $^{252}Cf$ and PuBe energy spectra unfolded using a single EJ309 organic scintillator; excellent agreement is observed with the reference spectra. Neutron energy spectroscopy is also possible using a two-plane detector array, whereby time-of-flight kinematics can be used. With this system, energy spectra can also be obtained as a function of position. Spatial-dependent energy spectra are presented for neutron and gamma-ray sources that are in excellent agreement with expectations.

• Nuclear Engineering and Technology
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• 제56권9호
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• pp.3602-3607
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• 2024

The present work is devoted to implementation of the stilbene-based neutron spectrometer energy calibration method. The results of experiments with portable neutron generators and 238PuBe source and scattering materials with known cross sections are used for this method. It is shown that the submitted method makes it possible to carry out fast neutron spectrometry in the energy range from 1 to 15 MeV with the uncertainty of the unfolded neutron energy no more than 200 keV. Neutron spectra unfolding was carried out based on the measured spectra and a Geant4 simulated response matrix. Unfolded spectra were compared with the literature data and reference spectra.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.50-50
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• 1999

• Nuclear Engineering and Technology
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• 제5권4호
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• pp.334-338
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• 1973

$^{232}$ Th 핵분열 물질과 조합된 고체비적검출체를 사용하여 250kw로 정상운전되는 TRIGA Mark-II 원자로의 대차폐수조내에서 열중성자주(thermalizing column)의 중심으로부터 수평방향의 속 중성자 선속밀도 분포를 추정하였다. 속 중성자 스펙트럼이 $^{235}$ U가 열 중성자에 의하여 핵분열이 일어날매 방출되는 중성자 스펙트럼과 같다는 가정을 한 다음, 선속밀도는 고쳬비적검출체로 얻어진 실험 결과로부터 계산되었다. 이와 같은 방법으로 속 중성자 설속밀도 분포의 측정 결과는 도표로서 제시된다.

• 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.

• 방사선산업학회지
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• 제18권3호
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• pp.249-253
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• 2024

For the safe management and disposal of spent fuel, it is essential to accurately determine the quantities of nuclides present within the spent fuel. In this study, a Gamma/Neutron detection system was developed as a part of basic research to measure the burnup profile of spent fuel, and a performance was evaluated using major nuclides. The prototype of the Gamma/Neutron detection system consisted of a CZT sensor and a ³He chamber. For quantitative evaluation, studies were conducted using calibrated ¹³⁷Cs, ¹³⁴Cs, ¹⁵⁴Eu and ²⁵²Cf sources. In the performance evaluation, a field applicability was verified by analyzing the detection characteristics according to the nuclide.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.203-210
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• 2018

As the modernization of the nuclear instrumentation system progresses, research reactors have adopted digital wide-range neutron power measurement (DWRNPM) systems. These systems typically monitor the neutron flux across a range of over 10 decades. Because neutron detectors only measure the local neutron flux at their position, the local neutron flux must be converted to total reactor power through calibration, which involves mapping the local neutron flux level to a reference reactor power. Conventionally, the neutron power range is divided into smaller subranges because the neutron detector signal characteristics and the reference reactor power estimation methods are different for each subrange. Therefore, many factors should be considered when preparing the calibration procedure for DWRNPM channels. The main purpose of this work is to serve as a reference for performing the calibration of DWRNPM systems in research reactors. This work provides a comprehensive overview of the calibration of DWRNPM channels by describing the configuration of the DWRNPM system and by summarizing the theories of operation and the reference power estimation methods with their associated calibration procedure. The calibration procedure was actually performed during the commissioning of an open-pool type research reactor, and the results and experience are documented herein.