http://chineseinput.net/에서 pinyin(병음)방식으로 중국어를 변환할 수 있습니다.
변환된 중국어를 복사하여 사용하시면 됩니다.
- 中文 을 입력하시려면 zhongwen을 입력하시고 space를누르시면됩니다.
- 北京 을 입력하시려면 beijing을 입력하시고 space를 누르시면 됩니다.
RISS 인기검색어
- 검색키워드
- 저자 : Sanggeol Jeong (검색결과 8 건)
- 무료
- 기관 내 무료
- 유료
-
-
Sanggeol Jeong,Giyoon Kim,Jaeyeong Jang,Heejun Chung,Myungsoo Kim 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.2
When the nuclear accident like the Fukushima is occurred, it is required to immediately determine the location of radioactive materials and their activities. Various studies related the unmanned technique to detect and characterize the contaminated area have been conducted. The Korea Institute of Nuclear Nonproliferation and Control (KINAC) has developed a new gamma detection system which consists of nine probes using a silicon photomultiplier (SiPM) and plastic scintillator. The probe is the small gamma detector designed to be carried and dropped near the accident area by the unmanned aerial vehicle. In this paper, we developed the improved design related to the angular dependence of the radioactive contamination detection system with the purpose of increasing the detection efficiency. The detection efficiency, radiation shielding and back-scattering varies depending on the direction of incidence of radiation because the probe has vertical structure of consisting scintillator, photomultiplier, and electric circuits. That is, when the experimental conditions are same except the direction of gamma probe, the result of measurements is different. It causes errors in measuring the radioactivity and location of the radioactive source. Since the direction of the probe is arbitrarily determined during the deployment of the probe through the unmanned aerial vehicle, it is considered changing the design of the scintillator from a conventional 1.0 × 1.0 Φ cylindrical shape to a 1.0 Φ spherical shape. In case of using the spherical scintillator, it is confirmed that angular dependence was reduced through MCNP simulation. The difference in the measurement depending on the direction of the probe could be reduced through additional structure design. Finally, we hope that the developed detection system which has the probes with spherical shape of scintillator can measure the radioactivity and location of the radioactive source in a range of about 100 × 100 m2 by measuring for at least 5 minutes. The field test at Fukushima area will be carried out with JAEA members in order to prove the feasibility of the new system.
-
Criticality Safety Analysis for Concrete Radioactive Waste Drums Using KENO-VI of SCALE
Sanggeol Jeong,Jang Hwan Lim,Seong Ki Lee 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2
Most of the radioactive wastes generated during the nuclear fuel processing activities conducted by KEPCO Nuclear Fuel Co., Ltd. are classified as the categories of intermediate and low-level radioactive waste. These radioactive waste materials are intended for permanent disposal at a designated disposal site, adhering strictly to the waste acceptance criteria. To facilitate the safe transportation of radioactive waste to the disposal site, it is necessary to ensure that the waste drums maintain a level of criticality that complies with the waste acceptance criteria. This necessitates the maintenance of subcritical conditions, under immersion or optimal neutron moderation conditions. This paper presents a criticality safety assessment of concrete radioactive waste under the most conservative conditions of immersion and moderation conditions for waste drums. Specifically, In order to send radioactive waste, which is the subject of criticality analysis, to a disposal facility, pre-processing operations must be performed to ensure compliance with waste accepatance criteria. To meet the physical characteristics required by the accepance criteria, particles below 0.2 mm should not be included. Thus, a 0.3 mm sieve is used to separate particles lager than 0.3 mm, and only those particles are placed in drums. The drums should be filled to achieve a filling ratio of at least 85%. A criticality analysis was conducted using the KENO-VI of SCALE. The Criticality Safety Analysis Results of varying the filling ratio of concrete drums from 85% to 100% presented in an effective multiplication factor of 0.22484. Additionally, the effective multiplication factor presented to be 0.25384 under the optimal moderation conditions. This demonstrates full compliance with the USL and criticality technology standards set as 0.95.
-
Jeong Sanggeol,Kim Dong Yeung,Kim Giyoon,Lee Jun,Chung Heejun,Kim Myungsoo 한국물리학회 2023 THE JOURNAL OF THE KOREAN PHYSICAL SOCIETY Vol.83 No.12
The denuclearization verifcation process involves the localization of nuclear materials in the area of nuclear inspection. Various methodologies based on detector measurement using CsI(Tl), NaI(Tl) scintillators and Geiger–Müller (GM) counters have been studied to localize a nuclear material, but they are not suitable for application to a wide outdoor range. The Korea Institute of Nuclear Nonproliferation and Control (KINAC) has developed a plastic scintillator-based small gamma-ray instrument (probe). In this study, artifcial intelligence-based machine learning was applied to localize radioactive material based on probe measurement values. A localization algorithm model based on a Deep Neural Network (DNN) and Multiple Linear Regression (MLR) which are most used among various machine learning and deep learning algorithms was created. Then, the radioactive material was localized based on the measured value and compared with MCNP6-based simulation data. The performance of the DNN and MLR algorithms was evaluated through a coefcient of determination (R2 ) and Root Mean Square Error (RMSE). The results for the R2 and RMSE of the DNN algorithms model are 0.9488 and 3.5734 m. The R2 and RMSE of the MLR algorithm model are 0.8496 and 7.2452 m.
-
-
Giyoon Kim,Sanggeol Jeong,Jaeyeong Jang,Heejun Chung,Myungsoo Kim 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.2
In emergency situations such as nuclear accidents or terrorism, radioactive and nuclear materials can be released by some environmental reasons such as the atmosphere and underground water. To secure the safety of human beings and to respond appropriately emergency situation, it is required to designate high and low dose rate regions in the early stages by analyzing the location and radioactivity of sources through environmental radiation measurement. This research team has developed a small gamma probe which is featured by its geometrical accessibility and higher radiation sensitivity than other drone detectors. A plastic scintillator and Silicon Photomultiplier (SiPM) were applied to the probe to optimize the wireless measurement condition. SiPM has a higher gain (higher than 106) and lower operating voltage (less than 30 V) compared to a general photodiode. However, the electronic components in the SiPM are sensitively affected by temperature, which causes the performance degradation of the SiPM. As the SiPM temperature increases, the breakdown voltage (VBD) of the SiPM also increases, so the gain must be maintained by applying the appropriate VBD. Therefore, when the SiPM temperature increases while the VBD is fixed, the gain decreases. Thus, the signal does not exceed the threshold voltage (VTH) and the overall count is reduced. In general, the optimal gain is maintained by cooling the SiPM or through a temperature compensation circuit. However, in the developed system, the hardware correction method such as cooling or temperature compensation circuit cannot be applied. In this study, it was confirmed that the count decreased by up to 20% according to the increase in the temperature of the SiPM when the probe was operated at room temperature (26°C). We propose methods to calibrate the total count without cooling device or compensation circuit. After operating the probe at room temperature, the first measured count is set as the reference value, and the correction factor is derived using the tendency of the count to decrease as the temperature increases. In addition, since this probe is used for environmental radiation monitoring, periodic measurements are more suitable than continuous measurements. Therefore, the temperature of the probe can be maintained by adding a power saving interval to the operation sequence of the probe. These two methods use the operation sequence and measurement data, respectively. Thus, it is expected to be the most effective method for the current system where the temperature compensation through hardware is not possible.
-
Jang, Jaerim,Kim, Wonkyeong,Jeong, Sanggeol,Jeong, Eun,Park, Jinsu,Lemaire, Matthieu,Lee, Hyunsuk,Jo, Yongmin,Zhang, Peng,Lee, Deokjung Elsevier 2018 Annals of nuclear energy Vol.114 No.-
<P><B>Abstract</B></P> <P>This paper presents the validation of the continuous-energy Monte Carlo neutron-transport code MCS with the ENDF/B-VII.0 neutron cross-section library for the criticality safety analysis of PWR spent fuel pools and storage casks. The MCS code is developed by the COmputational Reactor physics and Experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for the analysis of Pressurized Water Reactors (PWRs) with high fidelity and high performance. The validation is conducted with 279 selected critical benchmarks from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The 279 validation cases are representative of PWR spent fuel pools and storage casks with <SUP>235</SUP>U enrichment ranging from 2.35 wt% to 4.74 wt%, pin pitch ranging from 1.075 cm to 2.540 cm, moderator to fuel ratio (H/U) ranging from 0.4683 and 11.5398, Energy of the Average Lethargy causing Fission (EALF) ranging from 0.0109 eV to 1.0600 eV, without soluble boron and with soluble boron concentration ranging from 0.015 g/L to 5.030 g/L. The calculation of the effective neutron multiplication factor by MCS is validated by the comparison between experiment and calculation for the selected critical benchmarks. The Upper Safety Limit (USL) of the MCS code is established in accordance to the NUREG/CR-6698 guideline recommended by the NRC (US National Regulatory Commission). The full validation process and determination of USL based on the selected critical benchmarks was also repeated with the MCNP6.1 and SERPENT2.1.27 codes in order to compare the performances of MCS with other reactor analysis codes. This paper demonstrates the capability of the MCS code for the criticality safety analysis of PWR spent fuel pools and storage casks.</P> <P><B>Highlights</B></P> <P> <UL> <LI> Validation of UNIST Monte Carlo code MCS with ENDF/B.VII.0 nuclear data library. </LI> <LI> Criticality safety analysis of PWR spent fuel pool and storage cask. </LI> <LI> Upper safety limits derived with single-sided tolerance band, limit and non-parametric methods. </LI> <LI> Code/code comparison against MCNP6.1 and SERPENT2.1.27 Monte Carlo codes. </LI> </UL> </P>
-
Jaeyeong Jang,Giyoon Kim,Sanggeol Jeong,Heejun Chung,Myungsoo Kim 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.1
Our research team has developed a gamma ray detector which can be distributed over large area through air transport. Multiple detectors (9 devices per 1 set) are distributed to measure environmental radiation, and information such as the activity and location of the radiation source can be inferred using the measured data. Generally, radiation is usually measured by pointing the detector towards the radioactive sources for efficient measurement. However, the detector developed in this study is placed on the ground by dropping from the drone. Thus, it does not always face toward the radiation source. Also, since it is a remote measurement system, the user cannot know the angle information between the source and detector. Without the angle information, it is impossible to correct the measured value. The most problematic feature is when the backside of the detector (opposite of the scintillator) faces the radiation source. It was confirmed that the measurement value decreased by approximately 50% when the backside of the detector was facing towards the radiation source. To calibrate the measured value, we need the information that can indicate which part of the detector (front, side, back) faces the source. Therefore, in this study, we installed a small gamma sensor on the backside of the detector to find the direction of the detector. Since this sensor has different measurement specifications from the main sensor in terms of the area, type, efficiency and measurement method, the measured values between the two sensors are different. Therefore, we only extract approximate direction using the variation in the measured value ratio of the two sensors. In this study, to verify the applicability of the detector structure and measurement method, the ratio of measured values that change according to the direction of the source was investigated through MCNP simulation. The radioactive source was Cs-137, and the simulation was performed while moving in a semicircular shape with 15 degree steps from 0 degree to 180 degrees at a distance of 20 cm from the center point of the main sensor. Since the MCNP result indicates the probability of generating a pulse for one photon, this value was calculated based on 88.6 μCi to obtain an actual count. Through the ratio of the count values of the two sensors, it was determined whether the radioactive source was located in the front, side, or back of the probe.
연관 검색어 추천
이 검색어로 많이 본 자료
활용도 높은 자료
