M. Woo, Jae Hyun Kim, J. Kim, Chewook Yim, Jae Yong Lee, Do-Hyun Kim, Quang Huy Khuat, Bo Kyun Seo, C. Shin, Jong Kyung Kim
Since it is difficult to directly measure neutron energy, the neutron spectrum is estimated from the specific responses of the neutron reaction. Although several algorithms have been proposed to reconstruct the neutron spectrum, no attempt has been made to apply a multi objective optimization technique. This study is based on the idea that reconstructing the spectrum by taking into consideration various prior information simultaneously enables to obtain more reasonable results. The genetic multi-objective optimization scheme was applied to derive the Pareto front of spectrum from activation foil responses. The two objectives of maximizing the Shannon information entropy of the neutron spectrum and minimizing the relative error of the responses were considered. By applying the algorithm, we were able to successfully reduce the solution candidates to Pareto front and improve the validity of the unfolded spectrum.
{"title":"An application of genetic multi-objective optimization algorithm to neutron spectrum unfolding problem","authors":"M. Woo, Jae Hyun Kim, J. Kim, Chewook Yim, Jae Yong Lee, Do-Hyun Kim, Quang Huy Khuat, Bo Kyun Seo, C. Shin, Jong Kyung Kim","doi":"10.15669/PNST.6.230","DOIUrl":"https://doi.org/10.15669/PNST.6.230","url":null,"abstract":"Since it is difficult to directly measure neutron energy, the neutron spectrum is estimated from the specific responses of the neutron reaction. Although several algorithms have been proposed to reconstruct the neutron spectrum, no attempt has been made to apply a multi objective optimization technique. This study is based on the idea that reconstructing the spectrum by taking into consideration various prior information simultaneously enables to obtain more reasonable results. The genetic multi-objective optimization scheme was applied to derive the Pareto front of spectrum from activation foil responses. The two objectives of maximizing the Shannon information entropy of the neutron spectrum and minimizing the relative error of the responses were considered. By applying the algorithm, we were able to successfully reduce the solution candidates to Pareto front and improve the validity of the unfolded spectrum.","PeriodicalId":20706,"journal":{"name":"Progress in Nuclear Science and Technology","volume":"14 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83626452","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
SCALE6.2.1 was released in 2016 and has been used worldwide. It includes new AMPX multigroup format files (AMPX MG libraries) of ENDF/B-VII.0 and ENDF/B-VII.1 and a new nuclear data processing code AMPX-6, which produces AMPX MG libraries. Thus we have produced an AMPX MG library (neutron 200 groups and gamma 47 groups) of JENDL-4.0 in order to disseminate JENDL-4.0. Neutron and gamma spectra inside a sphere of 1 m in radius with a 20 MeV neutron source at the center were calculated with ANISN and the iron experiment at JAEA/FNS was analyzed with DORT, in order to validate the AMPX MG library of JENDL-4.0. As a result, it was verified that the AMPX MG library of JENDL-4.0 had no problems except for the self-shielding correction which is a common problem in AMPX MG libraries.
{"title":"AMPX MG library of JENDL-4.0","authors":"C. Konno","doi":"10.15669/PNST.6.117","DOIUrl":"https://doi.org/10.15669/PNST.6.117","url":null,"abstract":"SCALE6.2.1 was released in 2016 and has been used worldwide. It includes new AMPX multigroup format files (AMPX MG libraries) of ENDF/B-VII.0 and ENDF/B-VII.1 and a new nuclear data processing code AMPX-6, which produces AMPX MG libraries. Thus we have produced an AMPX MG library (neutron 200 groups and gamma 47 groups) of JENDL-4.0 in order to disseminate JENDL-4.0. Neutron and gamma spectra inside a sphere of 1 m in radius with a 20 MeV neutron source at the center were calculated with ANISN and the iron experiment at JAEA/FNS was analyzed with DORT, in order to validate the AMPX MG library of JENDL-4.0. As a result, it was verified that the AMPX MG library of JENDL-4.0 had no problems except for the self-shielding correction which is a common problem in AMPX MG libraries.","PeriodicalId":20706,"journal":{"name":"Progress in Nuclear Science and Technology","volume":"19 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88067145","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
K. Hoshi, N. Tsujimura, Tadayoshi Yoshida, O. Kurihara, Eun-Joo Kim, K. Yajima
When a criticality accident occurs, a screening survey to triage high-dose radiation exposed persons is performed. We established a rapid method for the screening survey by measuring the gamma dose rate mainly from 24Na in the victims’ body with a conventional NaI(Tl) scintillation survey meter. In this study, a water-filled slab phantom containing NaCl was irradiated with neutrons from a 252Cf neutron source. The radioactivity concentration of 24Na produced in the phantom was determined by means of gamma-ray spectrometry and simulations using the Monte Carlo N-Particle Transport Code. The ambient dose equivalent rates at the phantom’s outer surface were simulated by the Monte Carlo N-Particle Transport Code and were directly measured with a NaI(Tl) scintillation survey meter. From the experiments and calculations, we obtained the results that 1 Gy (neutron absorbed dose) corresponds to 18–76 μSv/h (ambient dose equivalent rate) at the surface of the victim’s body, which can be distinguished from normal background levels. Therefore, this method allows us to rapidly screen high-dose radiation-exposed victims.
{"title":"Study of the screening survey using an ambient dose equivalent rate survey meter in criticality accidents","authors":"K. Hoshi, N. Tsujimura, Tadayoshi Yoshida, O. Kurihara, Eun-Joo Kim, K. Yajima","doi":"10.15669/PNST.6.152","DOIUrl":"https://doi.org/10.15669/PNST.6.152","url":null,"abstract":"When a criticality accident occurs, a screening survey to triage high-dose radiation exposed persons is performed. We established a rapid method for the screening survey by measuring the gamma dose rate mainly from 24Na in the victims’ body with a conventional NaI(Tl) scintillation survey meter. In this study, a water-filled slab phantom containing NaCl was irradiated with neutrons from a 252Cf neutron source. The radioactivity concentration of 24Na produced in the phantom was determined by means of gamma-ray spectrometry and simulations using the Monte Carlo N-Particle Transport Code. The ambient dose equivalent rates at the phantom’s outer surface were simulated by the Monte Carlo N-Particle Transport Code and were directly measured with a NaI(Tl) scintillation survey meter. From the experiments and calculations, we obtained the results that 1 Gy (neutron absorbed dose) corresponds to 18–76 μSv/h (ambient dose equivalent rate) at the surface of the victim’s body, which can be distinguished from normal background levels. Therefore, this method allows us to rapidly screen high-dose radiation-exposed victims.","PeriodicalId":20706,"journal":{"name":"Progress in Nuclear Science and Technology","volume":"190 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77269158","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The estimation of neutron duct streaming with a feeder for the power supply system in a JT-60 Tokamak fusion device facility has been performed via neutoronics analysis using a simplified formula for the cylindrical duct streaming model and a suitable model based on the Monte Carlo technique. However, the safety margin is too large for the evaluation using a simplified formula in order to evaluate the neutron streaming of a duct with three feeders (each feeder of size 0.126 m × 0.357 m × 2 m) inside a cylindrical duct (0.8 m diameter × 2 m length in size). Further, a detailed evaluation of neutron streaming of a cylindrical duct with three feeders has not been performed in the facility. In this study, three-dimensional (3D) neutronics analysis using a real-model of the complex structured duct was performed to assess the safety margin of the simplified formula. It was confirmed that the neutron dose attenuation of the penetrating section in the duct using the simplified formula is approximately one order of magnitude larger than the real-model using the 3D calculation code.
利用简化的圆柱形管道流模型公式和基于蒙特卡罗技术的合适模型，通过中子电子学分析对JT-60托卡马克聚变装置供电系统的中子管道流进行了估计。但是，在直径为0.8 m ×长度为2m的圆柱形风管中，采用简化公式评价具有3个给料器(每个给料器尺寸为0.126 m × 0.357 m × 2 m)的风管的中子流时，安全余量太大。此外，该设施还没有对带有三个进料器的圆柱形管道的中子流进行详细评估。在本研究中，使用复杂结构管道的实际模型进行了三维(3D)中子分析，以评估简化公式的安全裕度。用简化公式计算的导管穿透截面的中子剂量衰减比用三维计算程序计算的实际模型约大一个数量级。
{"title":"Estimation of neutron duct streaming with the feeder for the power supply system in the JT-60 tokamak fusion device facility","authors":"A. Sukegawa, Koichi Okuno","doi":"10.15669/PNST.6.95","DOIUrl":"https://doi.org/10.15669/PNST.6.95","url":null,"abstract":"The estimation of neutron duct streaming with a feeder for the power supply system in a JT-60 Tokamak fusion device facility has been performed via neutoronics analysis using a simplified formula for the cylindrical duct streaming model and a suitable model based on the Monte Carlo technique. However, the safety margin is too large for the evaluation using a simplified formula in order to evaluate the neutron streaming of a duct with three feeders (each feeder of size 0.126 m × 0.357 m × 2 m) inside a cylindrical duct (0.8 m diameter × 2 m length in size). Further, a detailed evaluation of neutron streaming of a cylindrical duct with three feeders has not been performed in the facility. In this study, three-dimensional (3D) neutronics analysis using a real-model of the complex structured duct was performed to assess the safety margin of the simplified formula. It was confirmed that the neutron dose attenuation of the penetrating section in the duct using the simplified formula is approximately one order of magnitude larger than the real-model using the 3D calculation code.","PeriodicalId":20706,"journal":{"name":"Progress in Nuclear Science and Technology","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87394547","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
N. Matsuda, S. Kunieda, Tsutomu Okamoto, Kenichi Tada, C. Konno
The Intra-Nuclear Cascade model employed into general-purpose Monte-Carlo simulation codes is not always applicable in the energy region from 20 MeV to several hundreds of MeV. In order to improve accuracy of the Monte-Carlo simulations including this energy region, the fourth version of Japanese Evaluated Nuclear Data Library High Energy File (JENDL-4.0/HE) as a new special purpose file, was released in 2015. It includes evaluated cross-sections for incident neutrons and protons up to 200 MeV for 130 and 133 nuclei, respectively. A Compact version of ENDF (ACE) library of all the neutronand proton-induced reaction data for the MCNP, MCNPX and PHITS codes has been produced with the nuclear data processing code NJOY2016.9, which was modified to keep laboratory angle-energy distribution form (LAW=67) in the ACE library of proton because the original NJOY converts laboratory angle-energy distribution form (LAW=67) to continuum energy-angle distribution form (LAW=61) automatically and PHITS can only treat angle-energy distribution form (LAW=67) for proton. Validations for the ACE library were performed through benchmark calculations for high-energy shielding experiments with PHITS.
{"title":"ACE library of JENDL-4.0/HE","authors":"N. Matsuda, S. Kunieda, Tsutomu Okamoto, Kenichi Tada, C. Konno","doi":"10.15669/PNST.6.225","DOIUrl":"https://doi.org/10.15669/PNST.6.225","url":null,"abstract":"The Intra-Nuclear Cascade model employed into general-purpose Monte-Carlo simulation codes is not always applicable in the energy region from 20 MeV to several hundreds of MeV. In order to improve accuracy of the Monte-Carlo simulations including this energy region, the fourth version of Japanese Evaluated Nuclear Data Library High Energy File (JENDL-4.0/HE) as a new special purpose file, was released in 2015. It includes evaluated cross-sections for incident neutrons and protons up to 200 MeV for 130 and 133 nuclei, respectively. A Compact version of ENDF (ACE) library of all the neutronand proton-induced reaction data for the MCNP, MCNPX and PHITS codes has been produced with the nuclear data processing code NJOY2016.9, which was modified to keep laboratory angle-energy distribution form (LAW=67) in the ACE library of proton because the original NJOY converts laboratory angle-energy distribution form (LAW=67) to continuum energy-angle distribution form (LAW=61) automatically and PHITS can only treat angle-energy distribution form (LAW=67) for proton. Validations for the ACE library were performed through benchmark calculations for high-energy shielding experiments with PHITS.","PeriodicalId":20706,"journal":{"name":"Progress in Nuclear Science and Technology","volume":"76 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86398765","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
T. Fujibuchi, Airi Inoue, Y. Ishigaki, Y. Matsumoto
Medical staff who perform interventional radiology face the risk of exposure to relatively high doses of primary radiation and scattered radiation emitted from patients’ bodies. For monitoring this exposure, we developed a wireless multisensor dosimeter that we named “Pocket Dose,” which provides real-time visualization of dose levels on an Android tablet screen. In this study, we investigate the characteristics of the energy and dose rates of the developed system comprising silicon PIN photodiode detectors, a signal transmitter, and a tablet for data display. The detection unit is equipped with two types of sensors: one for measuring low dose rates and the other for measuring high dose rates. In addition, four detectors are connected to the transmitter with a one-meter-long cable. These detectors are designed to measure the unequal radiation exposures at the chest, neck, and fingers of medical staff. The signals from the detectors transmit count information to the tablet using Bluetooth. The dose-rate linearities of the high-sensitivity and low-sensitivity sensors are confirmed to be less than 14.8 and 82.8 mSv/h, respectively. The system has great potential for energy correction using the energy spectra information. Using this system with wireless communication, real-time management of staff’s exposure becomes easy. Moreover, high expandability is expected since it uses a tablet.
{"title":"Development of a wireless multisensor active personal dosimeter-tablet system","authors":"T. Fujibuchi, Airi Inoue, Y. Ishigaki, Y. Matsumoto","doi":"10.15669/PNST.6.73","DOIUrl":"https://doi.org/10.15669/PNST.6.73","url":null,"abstract":"Medical staff who perform interventional radiology face the risk of exposure to relatively high doses of primary radiation and scattered radiation emitted from patients’ bodies. For monitoring this exposure, we developed a wireless multisensor dosimeter that we named “Pocket Dose,” which provides real-time visualization of dose levels on an Android tablet screen. In this study, we investigate the characteristics of the energy and dose rates of the developed system comprising silicon PIN photodiode detectors, a signal transmitter, and a tablet for data display. The detection unit is equipped with two types of sensors: one for measuring low dose rates and the other for measuring high dose rates. In addition, four detectors are connected to the transmitter with a one-meter-long cable. These detectors are designed to measure the unequal radiation exposures at the chest, neck, and fingers of medical staff. The signals from the detectors transmit count information to the tablet using Bluetooth. The dose-rate linearities of the high-sensitivity and low-sensitivity sensors are confirmed to be less than 14.8 and 82.8 mSv/h, respectively. The system has great potential for energy correction using the energy spectra information. Using this system with wireless communication, real-time management of staff’s exposure becomes easy. Moreover, high expandability is expected since it uses a tablet.","PeriodicalId":20706,"journal":{"name":"Progress in Nuclear Science and Technology","volume":"12 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89415040","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
We measured activation cross sections of aluminum for 0.4, 1.3, 2.2, and 3.0 GeV protons at J-PARC to obtain basic data for nuclear design of beam windows as the test experiment. It was found that data with higher accuracy than current ones will be measured by using precise beam controls and highly accurate beam monitoring. We compared the experimental results, the evaluated data (JENDL-HE/2007), and calculation results with several intra-nuclear cascade models in PHITS. Although the experimental data agreed with JENDL-HE/2007, the calculations underestimated by about 40%, which could come from the evaporation model (GEM) included in PHITS code. It was found that the calculations with original GEM code show good agreement with the experimental data.
{"title":"Measurement of activation cross sections of aluminum for protons with energies between 0.4 GeV and 3.0 GeV at J-PARC","authors":"H. Matsuda, S. Meigo, H. Iwamoto","doi":"10.15669/PNST.6.171","DOIUrl":"https://doi.org/10.15669/PNST.6.171","url":null,"abstract":"We measured activation cross sections of aluminum for 0.4, 1.3, 2.2, and 3.0 GeV protons at J-PARC to obtain basic data for nuclear design of beam windows as the test experiment. It was found that data with higher accuracy than current ones will be measured by using precise beam controls and highly accurate beam monitoring. We compared the experimental results, the evaluated data (JENDL-HE/2007), and calculation results with several intra-nuclear cascade models in PHITS. Although the experimental data agreed with JENDL-HE/2007, the calculations underestimated by about 40%, which could come from the evaporation model (GEM) included in PHITS code. It was found that the calculations with original GEM code show good agreement with the experimental data.","PeriodicalId":20706,"journal":{"name":"Progress in Nuclear Science and Technology","volume":"10 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91478902","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The authors present total amount of released radiocaesium from FNPP1 accident which satisfy mass balance in all components of the environment. The authors also present consensus values of Fukushima Dai-ichi Nuclear Power Plant (FNPP1) derived 137Cs in the environment. Total amount of atmospheric release was 15-20 PBq and direct discharge to the ocean was 3-6 PBq, respectively. 12-15 PBq was deposited in the North Pacific and 3-6 PBq was deposited on land, respectively. Therefore 15-18 PBq was injected as sum of atmospheric deposition and direct discharge in the North Pacific Ocean. The authors also present temporal variation of 137Cs in coastal region close to FNPP1 site. Based on radiocaesium activity concentration and 134Cs/137Cs activity ratios decay corrected to 11 March 2011 in surface water, major source of radiocaesium is still FNPP1 site and contribution from global fallout 137Cs increased at Tomioka and Hasaki depending on the distances from FNPP1 site. A part of radiocaesium in the Sub Tropical Mode Water (STMW) in the Pacific Ocean is already obducted from ocean interior and enter into northern part of East China sea and in Japan Sea.
{"title":"Fukushima radionuclides in the marine environment from coastal region of Japan to the Pacific Ocean through the end of 2016","authors":"M. Aoyama, Y. Inomata, D. Tsumune, Y. Tateda","doi":"10.15669/PNST.6.1","DOIUrl":"https://doi.org/10.15669/PNST.6.1","url":null,"abstract":"The authors present total amount of released radiocaesium from FNPP1 accident which satisfy mass balance in all components of the environment. The authors also present consensus values of Fukushima Dai-ichi Nuclear Power Plant (FNPP1) derived 137Cs in the environment. Total amount of atmospheric release was 15-20 PBq and direct discharge to the ocean was 3-6 PBq, respectively. 12-15 PBq was deposited in the North Pacific and 3-6 PBq was deposited on land, respectively. Therefore 15-18 PBq was injected as sum of atmospheric deposition and direct discharge in the North Pacific Ocean. The authors also present temporal variation of 137Cs in coastal region close to FNPP1 site. Based on radiocaesium activity concentration and 134Cs/137Cs activity ratios decay corrected to 11 March 2011 in surface water, major source of radiocaesium is still FNPP1 site and contribution from global fallout 137Cs increased at Tomioka and Hasaki depending on the distances from FNPP1 site. A part of radiocaesium in the Sub Tropical Mode Water (STMW) in the Pacific Ocean is already obducted from ocean interior and enter into northern part of East China sea and in Japan Sea.","PeriodicalId":20706,"journal":{"name":"Progress in Nuclear Science and Technology","volume":"7 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86071912","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Following the accident at the Fukushima Daiichi Nuclear Power Plant, some kind of unmanned aerial vehicles (UAVs) have been used to collect environmental radiation data. The Japan Atomic Energy Agency has been developed methods for analyzing radiation measurements taken by a small multirotor UAV. In the conventional method, count rate was converted to the air dose rate using a simple parameter. Therefore, such a conversion of the air dose rate using the conventional method is difficult in an area where the distribution of air dose rate is heterogeneous, or the terrain is complex. The goal of this study was to apply the inverse radiation method, which is used in medical applications such as positron emission tomography, for environmental radiation measurement. Basic algorithm was established for conversion from count rate to air dose rate in considering of topographic data. The algorithm was applied to real radiation data obtained by a UAV, and its accuracy was evaluated by comparing that data with ground measurement data. As compared with conventional methods, our algorithm was demonstrated to achieve a more accurate estimation of the air dose rate on the ground.
{"title":"Development of analysis method for airborne radiation monitoring using the inverse problem solutions","authors":"M. Sasaki, Azusa Ishizaki, Y. Sanada","doi":"10.15669/PNST.6.63","DOIUrl":"https://doi.org/10.15669/PNST.6.63","url":null,"abstract":"Following the accident at the Fukushima Daiichi Nuclear Power Plant, some kind of unmanned aerial vehicles (UAVs) have been used to collect environmental radiation data. The Japan Atomic Energy Agency has been developed methods for analyzing radiation measurements taken by a small multirotor UAV. In the conventional method, count rate was converted to the air dose rate using a simple parameter. Therefore, such a conversion of the air dose rate using the conventional method is difficult in an area where the distribution of air dose rate is heterogeneous, or the terrain is complex. The goal of this study was to apply the inverse radiation method, which is used in medical applications such as positron emission tomography, for environmental radiation measurement. Basic algorithm was established for conversion from count rate to air dose rate in considering of topographic data. The algorithm was applied to real radiation data obtained by a UAV, and its accuracy was evaluated by comparing that data with ground measurement data. As compared with conventional methods, our algorithm was demonstrated to achieve a more accurate estimation of the air dose rate on the ground.","PeriodicalId":20706,"journal":{"name":"Progress in Nuclear Science and Technology","volume":"18 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82379564","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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