S. Yoshimura, T. Miyamura, Tomonori Yamada, H. Akiba, H. Kiyoura
This paper presents Part II of the report of our research on three-dimensional finite element seismic response analyses of a full-scale integrated model of the boiling water reactor and reactor building of Unit 1 at Fukushima-Daiichi Nuclear Power Plant subjected to the 2011 off the Pacific coast of Tohoku Earthquake of 9.0 Mw occurring on March 11, 2011. The previous paper （ Part I ） reported the developed analysis method, i.e., the parallel structural analysis solver ADVENTURE_Solid Ver.2 and its pre- / post-modules towards the K computer, model construction, the verification of the analysis model, and computation performance. Part II （ this paper ） reports the results of the eigenanalysis and seismic response analysis of the developed three-dimensional finite element model. These results were precisely compared with those of a lumped mass model. The overall responses of the two different models agreed reasonably well, although the three-dimensional finite element model succeeded in showing local responses more precisely. Finally, we concluded some key roles of petascale simulation for such practical and socially important problems.
{"title":"Seismic Response Analysis of Unit 1 of Fukushima-Daiichi Nuclear Power Plant During the 2011 Off the Pacific Coast of Tohoku Earthquake Using Three-Dimensional Finite Element Method (2nd Report: Results of Eigenanalysis and Seismic Response Analysis)","authors":"S. Yoshimura, T. Miyamura, Tomonori Yamada, H. Akiba, H. Kiyoura","doi":"10.3327/TAESJ.J18.025","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.025","url":null,"abstract":"This paper presents Part II of the report of our research on three-dimensional finite element seismic response analyses of a full-scale integrated model of the boiling water reactor and reactor building of Unit 1 at Fukushima-Daiichi Nuclear Power Plant subjected to the 2011 off the Pacific coast of Tohoku Earthquake of 9.0 Mw occurring on March 11, 2011. The previous paper （ Part I ） reported the developed analysis method, i.e., the parallel structural analysis solver ADVENTURE_Solid Ver.2 and its pre- / post-modules towards the K computer, model construction, the verification of the analysis model, and computation performance. Part II （ this paper ） reports the results of the eigenanalysis and seismic response analysis of the developed three-dimensional finite element model. These results were precisely compared with those of a lumped mass model. The overall responses of the two different models agreed reasonably well, although the three-dimensional finite element model succeeded in showing local responses more precisely. Finally, we concluded some key roles of petascale simulation for such practical and socially important problems.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"8 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437376","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}
J. Kaneda, Yun Wang, Yusaku Maruno, M. Iwanami, S. Ishioka, N. Shigenaka, A. Hasegawa
In this paper, several modified stainless steels （ SSs ） with different contents of Ta （ 0.13 – 0.61 ％） and C （ 0.010 – 0.046 ％） were produced to measure the electrochemical potentiokinetic reactivation （ EPR ） ratio and investigate their crevice corrosion resistance, stress corrosion cracking （ SCC ） susceptibility and crack growth rate （ CGR ） in a simulated boiling water reactor （ BWR ） environment. As a result of the EPR tests, we found that the Ta / C ratio must be ≥ 13 to suppress sensitization by sta-bilization heat treatment. If the Ta / C ratio is ≥ 19, sensitization can be suppressed without stabiliza-tion heat treatment. In the crevice corrosion test under γ -ray irradiation, the maximum corrosion depth in the Ta-modified SSs was smaller than that in type 316L SS. Ta-modified SSs had better crevice corrosion resistance than 316L SS. In the creviced bent beam test, there was no SCC in any of the Ta-modified SSs, whereas cracks were found in four of seven specimens of 316L SS. The CGR test was conducted using 0.5 T-compact tension specimens. Crack growth rates of the Ta-modified SSs were lower than that of 316L SS. The crevice corrosion resistance and SCC resistance were improved by Ta addition. We assumed that Ta addition can improve the repassivation response.
{"title":"SCC and Crevice Corrosion Resistances of Stainless Steel Modified with Tantalum Addition","authors":"J. Kaneda, Yun Wang, Yusaku Maruno, M. Iwanami, S. Ishioka, N. Shigenaka, A. Hasegawa","doi":"10.3327/taesj.j18.036","DOIUrl":"https://doi.org/10.3327/taesj.j18.036","url":null,"abstract":"In this paper, several modified stainless steels （ SSs ） with different contents of Ta （ 0.13 – 0.61 ％） and C （ 0.010 – 0.046 ％） were produced to measure the electrochemical potentiokinetic reactivation （ EPR ） ratio and investigate their crevice corrosion resistance, stress corrosion cracking （ SCC ） susceptibility and crack growth rate （ CGR ） in a simulated boiling water reactor （ BWR ） environment. As a result of the EPR tests, we found that the Ta / C ratio must be ≥ 13 to suppress sensitization by sta-bilization heat treatment. If the Ta / C ratio is ≥ 19, sensitization can be suppressed without stabiliza-tion heat treatment. In the crevice corrosion test under γ -ray irradiation, the maximum corrosion depth in the Ta-modified SSs was smaller than that in type 316L SS. Ta-modified SSs had better crevice corrosion resistance than 316L SS. In the creviced bent beam test, there was no SCC in any of the Ta-modified SSs, whereas cracks were found in four of seven specimens of 316L SS. The CGR test was conducted using 0.5 T-compact tension specimens. Crack growth rates of the Ta-modified SSs were lower than that of 316L SS. The crevice corrosion resistance and SCC resistance were improved by Ta addition. We assumed that Ta addition can improve the repassivation response.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.3327/taesj.j18.036","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437430","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}
hybrid method proposed by Kennedy is referred to （ e.g., in the Atomic Energy Society of Japan standard （ draft ） on the risk assessment method for nuclear fuel facilities ） as one of the seismic probabilistic risk assessment methods for nuclear fuel facilities. Although this method enables us to easily evaluate a plant ʼ s annual probability of failure, the evaluated probability may occasionally be overestimated or underestimated depending on the analysis case. Such a tenden-cy is a major obstacle to judging the appropriateness of the evaluation results of the method. In this paper, the cause of overestimation and underestimation is analyzed, and ways to improve the simplified hybrid method are studied and proposed to enable evaluation results within an acceptable range, which is defined in this paper, to be obtained.
{"title":"Improvement of the Simplified Hybrid Method for Seismic Risk Assessment of Nuclear Fuel Facilities","authors":"K. Mori","doi":"10.3327/taesj.j18.037","DOIUrl":"https://doi.org/10.3327/taesj.j18.037","url":null,"abstract":"hybrid method proposed by Kennedy is referred to （ e.g., in the Atomic Energy Society of Japan standard （ draft ） on the risk assessment method for nuclear fuel facilities ） as one of the seismic probabilistic risk assessment methods for nuclear fuel facilities. Although this method enables us to easily evaluate a plant ʼ s annual probability of failure, the evaluated probability may occasionally be overestimated or underestimated depending on the analysis case. Such a tenden-cy is a major obstacle to judging the appropriateness of the evaluation results of the method. In this paper, the cause of overestimation and underestimation is analyzed, and ways to improve the simplified hybrid method are studied and proposed to enable evaluation results within an acceptable range, which is defined in this paper, to be obtained.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437438","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}
S. Yamaguchi, A. Yusufu, Takuma Shirahama, Y. Murakami, T. Onitsuka, Masayoshi Uno
SiC, which is a promising accident-tolerant fuel cladding, is a non-oxide, and it is known that pas-sive oxidation occurs, where by a protective oxide fi lm of SiO 2 is formed under atmospheric conditions above 900 ℃ . The reaction occurring at this high temperature is important in assessing the soundness of SiC during a severe accident, but the understanding of it is still insuf fi cient. In this study, to evalu-ate the high-temperature oxidation behavior when SiC cladding is exposed to the atmosphere （ 10 5 Pa ） during an accident involving a light-water reactor, an oxidation test was performed for up to 100 h at 1100 to 1500 ℃ . As a result, a SiO 2 oxide fi lm was formed on the surface of SiC, but the formation of bubbles originating from impurities and cracks due to a phase transformation was con fi rmed. In addi-tion, it was observed, for the fi rst time in this research, that a multilayered SiO 2 oxide fi lm was formed at 1500 ℃ . Therefore, it was shown that the oxidation reaction of SiC does not stop depending on the surrounding conditions under high temperature and atmospheric conditions.
{"title":"High-Temperature Stability of SiO2 Oxide Film on Surface of SiC","authors":"S. Yamaguchi, A. Yusufu, Takuma Shirahama, Y. Murakami, T. Onitsuka, Masayoshi Uno","doi":"10.3327/taesj.j18.044","DOIUrl":"https://doi.org/10.3327/taesj.j18.044","url":null,"abstract":"SiC, which is a promising accident-tolerant fuel cladding, is a non-oxide, and it is known that pas-sive oxidation occurs, where by a protective oxide fi lm of SiO 2 is formed under atmospheric conditions above 900 ℃ . The reaction occurring at this high temperature is important in assessing the soundness of SiC during a severe accident, but the understanding of it is still insuf fi cient. In this study, to evalu-ate the high-temperature oxidation behavior when SiC cladding is exposed to the atmosphere （ 10 5 Pa ） during an accident involving a light-water reactor, an oxidation test was performed for up to 100 h at 1100 to 1500 ℃ . As a result, a SiO 2 oxide fi lm was formed on the surface of SiC, but the formation of bubbles originating from impurities and cracks due to a phase transformation was con fi rmed. In addi-tion, it was observed, for the fi rst time in this research, that a multilayered SiO 2 oxide fi lm was formed at 1500 ℃ . Therefore, it was shown that the oxidation reaction of SiC does not stop depending on the surrounding conditions under high temperature and atmospheric conditions.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437460","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}
can be considered that the cooling function during the decommissioning period before starting fuel debris retrieval becomes more important for nuclear safety than in the other periods. This is because it plays a very important role in maintaining the physical state of the fuel and fuel debris to prevent recriticality and thus suppresses the mobility of radioactive materials. Similarly, the confinement function becomes more important during the fuel debris retrieval. Because of the high radiation level, it is very difficult to access the inside of the primary containment vessel and reactor building to inves-tigate and / or repair equipment or components. Thus, it may be necessary to propose a decontamination and decommissioning project with remaining unknowns and to perform extensive research and development to complete the project. Such a project must be carried out strategically with the efficient and effective use of risk information. In this paper, the results of considering the basic concept and methods of risk management during the decommissioning of accident-damaged nuclear power plants from the viewpoint of safety and economic risks are reported.
{"title":"Study on Risk Management during Decommissioning of Accident-Damaged Nuclear Power Plant","authors":"T. Aoki","doi":"10.3327/TAESJ.J18.021","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.021","url":null,"abstract":"can be considered that the cooling function during the decommissioning period before starting fuel debris retrieval becomes more important for nuclear safety than in the other periods. This is because it plays a very important role in maintaining the physical state of the fuel and fuel debris to prevent recriticality and thus suppresses the mobility of radioactive materials. Similarly, the confinement function becomes more important during the fuel debris retrieval. Because of the high radiation level, it is very difficult to access the inside of the primary containment vessel and reactor building to inves-tigate and / or repair equipment or components. Thus, it may be necessary to propose a decontamination and decommissioning project with remaining unknowns and to perform extensive research and development to complete the project. Such a project must be carried out strategically with the efficient and effective use of risk information. In this paper, the results of considering the basic concept and methods of risk management during the decommissioning of accident-damaged nuclear power plants from the viewpoint of safety and economic risks are reported.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437329","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}
J. Aihara, A. Yasuda, S. Ueta, H. Ogawa, Masaki Honda, Koichiro Ohira, Y. Tachibana
Jun AIHARA, Atsushi YASUDA, Shohei UETA, Hiroaki OGAWA, Masaki HONDA, Koichi OHIRA and Yukio TACHIBANA Japan Atomic Energy Agency, Oarai Research and Development Center, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311–1393, Japan Nuclear Fuel Industries, Ltd., 41 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319–1196, Japan Japan Atomic Energy Agency, Tokai Research and Development Center, 2–4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319–1195, Japan (Received May 10, 2019; accepted in revised form June 28, 2019; published online October 25, 2019)
{"title":"Development of Fabrication and Inspection Technologies for Oxidation-Resistant Fuel Element for High-Temperature Gas-Cooled Reactors","authors":"J. Aihara, A. Yasuda, S. Ueta, H. Ogawa, Masaki Honda, Koichiro Ohira, Y. Tachibana","doi":"10.3327/taesj.j19.002","DOIUrl":"https://doi.org/10.3327/taesj.j19.002","url":null,"abstract":"Jun AIHARA, Atsushi YASUDA, Shohei UETA, Hiroaki OGAWA, Masaki HONDA, Koichi OHIRA and Yukio TACHIBANA Japan Atomic Energy Agency, Oarai Research and Development Center, 4002 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki 311–1393, Japan Nuclear Fuel Industries, Ltd., 41 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319–1196, Japan Japan Atomic Energy Agency, Tokai Research and Development Center, 2–4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319–1195, Japan (Received May 10, 2019; accepted in revised form June 28, 2019; published online October 25, 2019)","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437520","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}
Yohei Sato, Tomoya Iwashima, Motohashi Tetsuo, M. Obata
International Research Institute for Nuclear Decommissioning is engaged in a METI-funded project with the title ‘Research and Development of Processing and Disposal of Solid Waste’. This R&D project, which aims to build a waste stream for Fukushima Daiichi Nuclear Power Station, is executed by many organizations. The waste stream should be implemented by organizing several steps from generation to storage / disposal, and mutual feedback between each step is important to obtain the de-sired result. Tools to visualize and share the relations between each R&D theme of this project have been developed to enable the mutual communication of project members who are executing the R&D of different steps.
{"title":"Tools to Visualize and Share Relations between R&D Themes of National Project on Fukushima Daiichi Waste Management","authors":"Yohei Sato, Tomoya Iwashima, Motohashi Tetsuo, M. Obata","doi":"10.3327/TAESJ.J18.028","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.028","url":null,"abstract":"International Research Institute for Nuclear Decommissioning is engaged in a METI-funded project with the title ‘Research and Development of Processing and Disposal of Solid Waste’. This R&D project, which aims to build a waste stream for Fukushima Daiichi Nuclear Power Station, is executed by many organizations. The waste stream should be implemented by organizing several steps from generation to storage / disposal, and mutual feedback between each step is important to obtain the de-sired result. Tools to visualize and share the relations between each R&D theme of this project have been developed to enable the mutual communication of project members who are executing the R&D of different steps.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437386","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}
Kazuo Yoshida, H. Tamaki, N. Yoshida, Ryoichiro Yoshida, Y. Amano, H. Abe
An accident of evaporation to dryness by boiling of high-level liquid waste （ HLLW ） is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium （ Ru ） , are released from the tanks with mixed vapor of water and nitric-acid into the atmosphere. In addition, nitrogen oxides are also released, formed by the thermal decomposition of metal nitrates of fission products （ FP ） in HLLW. It has been observed experimentally that nitrogen oxide strongly affects the transport behavior of Ru under the anticipated atmospheric conditions in cells and / or compartments of the facility building. Chemical reactions of nitrogen oxide with water and nitric acid are also recognized as complex phe-nomena that undergo simultaneously in the vapor and liquid phases. An analysis method has been developed by coupling two types of computer codes to simulate not only thermohydraulic behavior but also chemical reactions in the flow paths of carrier gases for quantitative estimation of the amount of Ru released to the environment. A simulation study has also been carried out with a typical facility building to demonstrate the feasibility of the developed simulation method.
{"title":"Analysis of Chemical Behavior of Nitrogen Oxide Formed by Thermal Decomposition of FP Nitrates in Accident of Evaporation to Dryness by Boiling of Reprocessed High-Level Liquid Waste","authors":"Kazuo Yoshida, H. Tamaki, N. Yoshida, Ryoichiro Yoshida, Y. Amano, H. Abe","doi":"10.3327/TAESJ.J18.019","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.019","url":null,"abstract":"An accident of evaporation to dryness by boiling of high-level liquid waste （ HLLW ） is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium （ Ru ） , are released from the tanks with mixed vapor of water and nitric-acid into the atmosphere. In addition, nitrogen oxides are also released, formed by the thermal decomposition of metal nitrates of fission products （ FP ） in HLLW. It has been observed experimentally that nitrogen oxide strongly affects the transport behavior of Ru under the anticipated atmospheric conditions in cells and / or compartments of the facility building. Chemical reactions of nitrogen oxide with water and nitric acid are also recognized as complex phe-nomena that undergo simultaneously in the vapor and liquid phases. An analysis method has been developed by coupling two types of computer codes to simulate not only thermohydraulic behavior but also chemical reactions in the flow paths of carrier gases for quantitative estimation of the amount of Ru released to the environment. A simulation study has also been carried out with a typical facility building to demonstrate the feasibility of the developed simulation method.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.3327/TAESJ.J18.019","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437277","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}
{"title":"Corrigendum: Demonstration of Rapid Detection System for Radioactive Cesium in the Air toward Releasing the Evacuation Order in the Difficult-To-Return Zone in Fukushima","authors":"T. Maekawa, Y. Oshima","doi":"10.3327/TAESJ.J18.024E","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.024E","url":null,"abstract":"","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437341","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}
M. Machida, S. Yamada, Ayako Iwata, S. Otosaka, Takuya Kobayashi, Masahisa Watababe, H. Funasaka, T. Morita
{"title":"Seven-year Temporal Variation of Cesium-137 Discharge Inventory from the Port of Fukushima Dai-ichi Nuclear Power Plant:","authors":"M. Machida, S. Yamada, Ayako Iwata, S. Otosaka, Takuya Kobayashi, Masahisa Watababe, H. Funasaka, T. Morita","doi":"10.3327/taesj.j18.030","DOIUrl":"https://doi.org/10.3327/taesj.j18.030","url":null,"abstract":"","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.3327/taesj.j18.030","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437396","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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