• Journal of Radiation Protection and Research
• /
• 제42권1호
• /
• pp.48-55
• /
• 2017

Background: Simultaneous detection of neutrons and gamma rays have become much more practicable, by taking advantage of good gamma-ray discrimination properties using pulse shape discrimination (PSD) technique. Recently, we introduced a commercial CLYC system in Korea, and performed an initial characterization and simulation studies for the CLYC detector system to provide references for the future implementation of the dual-mode scintillator system in various studies and applications. Materials and Methods: We evaluated a CLYC detector with 95% $^6Li$ enrichment using various gamma-ray sources and a $^{252}Cf$ neutron source, with validation of our Monte Carlo simulation results via measurement experiments. Absolute full-energy peak efficiency values were calculated for gamma-ray sources and neutron source using MCNP6 and compared with measurement experiments of the calibration sources. In addition, behavioral characteristics of neutrons were validated by comparing simulations and experiments on neutron moderation with various polyethylene (PE) moderator thicknesses. Results and Discussion: Both results showed good agreements in overall characteristics of the gamma and neutron detection efficiencies, with consistent ~20% discrepancy. Furthermore, moderation of neutrons emitted from $^{252}Cf$ showed similarities between the simulation and the experiment, in terms of their relative ratios depending on the thickness of the PE moderator. Conclusion: A CLYC detector system was characterized for its energy resolution and detection efficiency, and Monte Carlo simulations on the detector system was validated experimentally. Validation of the simulation results in overall trend of the CLYC detector behavior will provide the fundamental basis and validity of follow-up Monte Carlo simulation studies for the development of our dual-particle imager using a rotational modulation collimator.

• Journal of Radiation Protection and Research
• /
• 제42권4호
• /
• pp.212-221
• /
• 2017

Background: This study aims to calculate detector positions as a design of a radioactive source localizing radiation portal monitor (RPM) system using an improved genetic algorithm. Materials and Methods: To calculate of detector positions for a source localizing RPM system optimization problem is defined. To solve the problem, a modified iterative genetic algorithm (MIGA) is developed. In general, a genetic algorithm (GA) finds a globally optimal solution with a high probability, but it is not perfect at all times. To increase the probability to find globally optimal solution rather, a MIGA is designed by supplementing the iteration, competition, and verification with GA. For an optimization problem that is defined to find detector positions that maximizes differences of detector signals, a localization method is derived by modifying the inverse radiation transport model, and realistic parameter information is suggested. Results and Discussion: To compare the MIGA and GA, both algorithms are implemented in a MATLAB environment. The performance of the GA and MIGA and that of the procedures supplemented in the MIGA are analyzed by computer simulations. The results show that the iteration, competition, and verification procedures help to search for globally optimal solutions. Further, the MIGA is more robust against falling into local minima and finds a more reliably optimal result than the GA. Conclusion: The positions of the detectors on an RPM for radioactive source localization are optimized using the MIGA. To increase the contrast of the measurements from each detector, a relationship between the source and the detectors is derived by modifying the inverse transport model. Realistic parameters are utilized for accurate simulations. Furthermore, the MIGA is developed to achieve a reliable solution. By utilizing results of this study, an RPM for radioactive source localization has been designed and will be fabricated soon.

• Nuclear Engineering and Technology
• /
• 제5권3호
• /
• pp.191-201
• /
• 1973

MeV 영역의 속중성자분광을 위해 재래의 radiator system을 개량하여 ringshaped vertical radiator와 cone-shaped horizontal radiator를 공용한 특수한 recoil proton radiator assembly를 사용함으로서 energy 분해능의 저하없이 검출효율을 높이도록 recoil Proton telescope detector를 설계ㆍ제작하였다. 이 검출기에는 입사중성자속에 대한 Si(ti) 검출기의 직접노출을 피함으로서 background를 줄일수 있도록 입사중성자차폐부도 고안 내장되어 있다. 이 개량된 recoil proton telescope detector의 검출효율 및 energy 분해능을 중성자 energy 1-15 MeV에 대하여 radiator system과 Si(Li) 검출기사이의 거리변화에 따라 이론적인 계산치로 도출ㆍ표시하였으며, 실험적검증의 예로서 이 거리를 29cm로 하고 중성자 energy를 14.1 MeV로 하였을 때의 검출기의 제특성측정결과를 얻어 분석하였다. 측정결과의 분석에 의하면 이론에서 추정된것처럼 혼합형 radiator system을 사용하였을 때의 검출 효율은 단일 radiator system을 사용한 재래식 검출기의 검출효율의 2.2배의 증가를 보인데 반하여 energy 분해능의 저하는 불과 30%, background의 증가는 약40% 말만임을 알수가 있었다. 또한 측정에 의한 14.1 MeV 중성자에 대한 energy 분해능은 3.9% FWHM었는데, 이는 이논적인 3.7% FWHM와 거의 완전한 일치를 보이고 있음도 입증되였다.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
• /
• pp.197-201
• /
• 1996

In the detail reaction-rate measurements in a critical assembly using the foil activation method, the measured activations of detector foils have inevitably errors caused by detector foil self-shielding effect. If neutron flux could be approximated to Westcott flux: i.e. well thermalized Maxwellian distribution, these activations of detector foil could be corrected to represent the unperturbated flux at any detected position in the cell with using Westcott option and reaction-rate option of the lattice code, WIMS-AECL. These calculated detector material self-shielding correction factors of the tested fuel, CANFLEX provided much information about neutron spectrum of test lattice cell as well as the correction factors themselves. The results could be verified by another lattice calculations.

• Journal of Radiation Protection and Research
• /
• 제2권1호
• /
• pp.1-8
• /
• 1977

중성자(中性子) 선량(線量) 측정목적(測定目的)으로 사용(使用)할 Posi필름 고체비적검출기(固體飛跡檢出器)의 알파입자(粒子) 비적검출(飛跡檢出) 효율(效率)과 화학부식(化學腐蝕)에 의(依)한 그의 비적형성(飛跡形成) 최적조건(最適條件)을 실험적(實驗的)으로 결정(決定)하였다. $^{10}B$ 및 $^{27}Al$박(箔)과 posi 필름 고체비적검출기(固體飛跡檢出器)로 이루어진 중성자(中性子) 선량계(線量計)를 제작(製作)하고 이것에 의(依)한 중성자(中性子) 선속밀도(線束密度) 및 선량(線量)의 측정범위등(測定範圍等)을 실험결과(實驗結果)와 이론적(理論的)인 근거하(根據下)에서 산출(算出)하였다. 그 결과(結果) posi 필름 고체비적검출기(固體飛跡檢出器)는 핵분열성(核分裂性) 물질(物質)을 입자방출체(粒子放出體)로 병용(倂用)함이 없이 중성자(中性子) 측정(測定)에 효과적(效果的)으로 응용(應用)될 수 있음이 판명(判明)되었다.

• Nuclear Engineering and Technology
• /
• 제48권6호
• /
• pp.1291-1302
• /
• 2016

This paper presents a new and efficient scheme to determine the optimal neutron source position in a model near-equilibrium pressurized water reactor, which is based on the OPR1000 Hanul Unit 3 Cycle 7 configuration. The proposed scheme particularly assigns importance of source positions according to the local adjoint flux distribution. In this research, detailed pin-by-pin reactor adjoint fluxes are determined by using the Monte Carlo KENO-VI code from solutions of the reactor homogeneous critical adjoint transport equations. The adjoint fluxes at each allowable source position are subsequently ranked to yield four candidate positions with the four highest adjoint fluxes. The study next simulates ex-core detector responses using the Monte Carlo MAVRIC code by assuming a neutron source is installed in one of the four candidate positions. The calculation is repeated for all positions. These detector responses are later converted into an inverse count rate ratio curve for each candidate source position. The study confirms that the optimal source position is the one with very high adjoint fluxes and detector responses, which is interestingly the original source position in the OPR1000 core, as it yields an inverse count rate ratio curve closest to the traditional 1/M line. The current work also clearly demonstrates that the proposed adjoint flux-based approach can be used to efficiently determine the optimal geometry for a neutron source and a detector in a modern pressurized water reactor core.

• 한국컴퓨터산업학회논문지
• /
• 제8권1호
• /
• pp.9-16
• /
• 2007

본 연구에서는 아스팔트 함량 변화에 따라서 중성자 계측수가 어떻게 변화되는가를 계산하여 법적 규제 면제치인 $100[{\mu}Ci]$이하의 방사성동위원소를 이용한 아스팔트 함량측정기의 기본 설계 자료로 활용하고자한다. 이를 위하여 1차 년도에서 실시했던 설계자료를 활용하여 아스팔트 함량의 변화에 따라 중성자 계측수가 어떻게 증감이 이루어지고, 또한 감속재인 폴리에틸렌 주변에 흡수체인 카드늄판을 설치했을 때의 계측수의 변화를 MCNP 코드를 이용하여 살펴보았다.

• Journal of Radiation Protection and Research
• /
• 제44권2호
• /
• pp.53-63
• /
• 2019

Background: As neutron fields are always accompanied by gamma rays, it is essential to distinguish neutrons from gamma rays in the detection of neutrons. Neutrons and gamma rays can be separated by pulse shape discrimination (PSD) methods. Recently, we performed characterization of a stilbene scintillator detector and an EJ-301 liquid scintillator detector with a high-speed digitizer DT5730 and investigated optimized PSD variables for both detectors. This study is for providing a basis for developing fast neutron/gamma-ray dual-particle imager. Materials and Methods: We conducted PSD experiments using stilbene scintillator and EJ-301 liquid scintillator and evaluated neutron and gamma ray discriminability of each PSD method with a $^{137}Cs$ gamma source and a $^{252}Cf$ neutron source. We implemented digital signal processing techniques to apply two PSD methods - the charge comparison (CC) method and the constant time discrimination (CTD) method - to distinguish neutrons from gamma rays. We tried to find optimized PSD variables giving the best discriminability in a given experimental condition. Results and Discussion: For the stilbene scintillator detector, the charge comparison method and the constant time discrimination method both delivered the PSD FOM values of 1.7. For the EJ-301 liquid scintillator detector, both PSD methods delivered the PSD FOM values of 1.79. With the same PSD variables, PSD performance was excellent in $300{\pm}100keVee$, $500{\pm}100keVee$, and $700{\pm}100keVee$ energy regions. This result shows that we can achieve an effective discrimination of neutrons from gamma rays using these scintillator detector systems. Conclusion: We applied both PSD methods to a stilbene and a liquid scintillator and optimized the PSD performance represented by FOM values. We observed a good separation performance of both scintillators combined with a high-speed digitizer and digital PSD. These results will provide reference values for the dual-particle imager we are developing, which can image both fast neutrons and gamma rays simultaneously.

• Nuclear Engineering and Technology
• /
• 제55권3호
• /
• pp.1031-1035
• /
• 2023

Neutron is an indirectly ionizing particle without charge, which is normally measured by detecting reaction products. Neutron detection system based on measuring gadolinium-converted gamma-rays is a good way to monitor the neutron because the representative prompt gamma-rays of gadolinium have low energies (79, 89, 182, and 199 keV). Low energy gamma-rays and their high attenuation coefficient on materials allow the simple design of a detector easier to manufacture. Thus, we designed a cadmium zinc telluride detector to investigate feasibility of simultaneous detection of gamma-rays and neutrons by using the Monte-Carlo simulation, which was divided into two parts; first was gamma-detection part and second was gamma- and neutron-simultaneous detection part. Consequently, we confirmed that simultaneous detection of gamma-rays and neutrons could be feasible and valid, although further research is needed for adoption on real detection.

• 방사선산업학회지
• /
• 제17권3호
• /
• pp.293-297
• /
• 2023

The burnup profile of spent fuel should be determined accurately for the safety storage of spent fuel. In this study, a neutron detection system was developed as a part of basic research to analyze the burnup profile of spent fuel, and a performance was evaluated using a radiation source. The prototype of the neutron detection system was based on a ³He proportional chamber. The ³He proportional chamber is often used for neutron measurement and analysis because of its high neutron detection efficiency and simplicity for gamma ray rejection. For quantitative evaluation, tests were conducted using calibrated ²⁵²Cf and ¹³⁷Cs sources. In the performance evaluation, a field applicability was verified by analyzing the detection characteristics according to the nuclide.