A. Ohzu, M. Komeda, M. Kureta, Naoki Zaima, Y. Nakatsuka, Shinichi Nakashima
A non-destructive assay system using the fast neutron direct interrogation method has been de-signed and developed to be put into practical use for the determination of the uranium （ 235 U ） mass contained in actual uranium-contaminated waste drums. The method is capable of measuring the fissile mass in a drum by counting the number of fission neutrons resulting from nuclear fission reac-tions between the fissile materials contained in a drum and thermal neutrons generated by 14 MeV fast neutrons irradiated from outside the drum. A performance test employing simulated metal waste drums demonstrated that a natural uranium mass of as low as approximately 10 g could be detected within an error of ± 20 ％ regardless of the distribution of uranium samples in the drum, and the total number of fission neutrons was proportional to the 235 U mass. A demonstration test employing actual waste drums could determine the uranium mass by using a newly developed correction method for de-riving the fissile mass in a drum. It has been proved by the experimental validation tests that the assay system equipped with the correction method is very useful for the accountancy of waste drums.
{"title":"Development of Non-destructive Assay System using Fast Neutron Direct Interrogation Method for Actual Uranium Waste Drums","authors":"A. Ohzu, M. Komeda, M. Kureta, Naoki Zaima, Y. Nakatsuka, Shinichi Nakashima","doi":"10.3327/TAESJ.J15.011","DOIUrl":"https://doi.org/10.3327/TAESJ.J15.011","url":null,"abstract":"A non-destructive assay system using the fast neutron direct interrogation method has been de-signed and developed to be put into practical use for the determination of the uranium （ 235 U ） mass contained in actual uranium-contaminated waste drums. The method is capable of measuring the fissile mass in a drum by counting the number of fission neutrons resulting from nuclear fission reac-tions between the fissile materials contained in a drum and thermal neutrons generated by 14 MeV fast neutrons irradiated from outside the drum. A performance test employing simulated metal waste drums demonstrated that a natural uranium mass of as low as approximately 10 g could be detected within an error of ± 20 ％ regardless of the distribution of uranium samples in the drum, and the total number of fission neutrons was proportional to the 235 U mass. A demonstration test employing actual waste drums could determine the uranium mass by using a newly developed correction method for de-riving the fissile mass in a drum. It has been proved by the experimental validation tests that the assay system equipped with the correction method is very useful for the accountancy of waste drums.","PeriodicalId":8595,"journal":{"name":"Atomic Energy Society of Japan","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2016-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80715919","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. Hamamoto, T. Kawamoto, M. Kondo, Hiroaki Sawahata, S. Takada, Masayuki Shinozaki
A high-temperature engineering test reactor （ HTTR ） has a reactivity control system which is ac-companied with a reserved shutdown system （ RSS ） . The RSS consists of B 4 C / C pellets, a guide tube, an electric plug, a motor which contains a brake and reducer, and so on. In accidents when the control rods cannot be inserted, the electric plug is pulled out by the motor and the B 4 C / C pellets fall into the core by gravity. It was revealed that the motor in the RSS drive mechanism did not work as the result of pre-start-up checks as described in the following: （ 1 ） The oil which separated from the grease of the motor reducer flowed down from the gap of the oil seal, （ 2 ） the separated oil penetrated into the brake, （ 3 ） the penetrated oil mixed with the abrasive powder released from the brake disk, and final-ly, （ 4 ） the adhesive mixture blocked the rotation of the motor. A new evaluation method was proposed to detect signs of the motor sticking. Through the overhaul inspection of all RSS drive mechanisms of an HTTR, it was revealed that the proposed method was effective to evaluate the integrity of the RSS drive mechanism. るとともに，運転時の異常な過渡変化時および事故時に安
{"title":"Establishment of Integrity Evaluation Method for Reserved Shutdown System of High-Temperature Engineering Test Reactor (HTTR)","authors":"S. Hamamoto, T. Kawamoto, M. Kondo, Hiroaki Sawahata, S. Takada, Masayuki Shinozaki","doi":"10.3327/taesj.J15.016","DOIUrl":"https://doi.org/10.3327/taesj.J15.016","url":null,"abstract":"A high-temperature engineering test reactor （ HTTR ） has a reactivity control system which is ac-companied with a reserved shutdown system （ RSS ） . The RSS consists of B 4 C / C pellets, a guide tube, an electric plug, a motor which contains a brake and reducer, and so on. In accidents when the control rods cannot be inserted, the electric plug is pulled out by the motor and the B 4 C / C pellets fall into the core by gravity. It was revealed that the motor in the RSS drive mechanism did not work as the result of pre-start-up checks as described in the following: （ 1 ） The oil which separated from the grease of the motor reducer flowed down from the gap of the oil seal, （ 2 ） the separated oil penetrated into the brake, （ 3 ） the penetrated oil mixed with the abrasive powder released from the brake disk, and final-ly, （ 4 ） the adhesive mixture blocked the rotation of the motor. A new evaluation method was proposed to detect signs of the motor sticking. Through the overhaul inspection of all RSS drive mechanisms of an HTTR, it was revealed that the proposed method was effective to evaluate the integrity of the RSS drive mechanism. るとともに，運転時の異常な過渡変化時および事故時に安","PeriodicalId":8595,"journal":{"name":"Atomic Energy Society of Japan","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2016-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75659078","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}
Shin-ichiro Uesawa, Wei Liu, Lifang Jiao, T. Nagatake, K. Takase, Mitsuhiko Shibata, H. Yoshida
Seawater was injected into the reactors during the accident at TEPCO’s Fukushima Daiichi NPS. However, the effects of the seawater on the cooling performance of the fuel rods and fuel debris are not clear. As possible effects, the change in the physical properties of the coolant and the sea salt de-position on a heat transfer surface and in the coolant are considered. We conducted thermal-hydraulic experiments using an internally heated annulus to determine the effects of seawater under conditions without boiling. The same experiments for water and sodium chloride （ NaCl ） solution were also conducted for the purpose of comparison with the artificial seawater. In these experiments, considering the physical properties of the artificial seawater, the thermal-hydraulic behaviors of the artificial seawater under forced convection （ Re ＞ 2300 ［－］） was estimated from the Dittus-Boelter correlation although sea salt was deposited in the fluid. According to the results of particle image velocimetry （ PIV ） , the velocity distribution in the artificial seawater was NOT different from that in the water and the NaCl solution. For a mixed convection regime, the Nusselt number of the artificial seawater was obtained from the correlation of the Grasholf number, Reynolds number and Prandtl number, as well as those for the water and the NaCl solution. Therefore, considering the physical properties of the artificial seawater, the thermal-hydraulic behavior of the seawater in single-phase flow can be estimated from the conventional thermal-hydraulic correlations for a single-phase flow.
{"title":"Effect of Seawater on Heat Transfer without Boiling in Internally Heated Annulus","authors":"Shin-ichiro Uesawa, Wei Liu, Lifang Jiao, T. Nagatake, K. Takase, Mitsuhiko Shibata, H. Yoshida","doi":"10.3327/TAESJ.J15.024","DOIUrl":"https://doi.org/10.3327/TAESJ.J15.024","url":null,"abstract":"Seawater was injected into the reactors during the accident at TEPCO’s Fukushima Daiichi NPS. However, the effects of the seawater on the cooling performance of the fuel rods and fuel debris are not clear. As possible effects, the change in the physical properties of the coolant and the sea salt de-position on a heat transfer surface and in the coolant are considered. We conducted thermal-hydraulic experiments using an internally heated annulus to determine the effects of seawater under conditions without boiling. The same experiments for water and sodium chloride （ NaCl ） solution were also conducted for the purpose of comparison with the artificial seawater. In these experiments, considering the physical properties of the artificial seawater, the thermal-hydraulic behaviors of the artificial seawater under forced convection （ Re ＞ 2300 ［－］） was estimated from the Dittus-Boelter correlation although sea salt was deposited in the fluid. According to the results of particle image velocimetry （ PIV ） , the velocity distribution in the artificial seawater was NOT different from that in the water and the NaCl solution. For a mixed convection regime, the Nusselt number of the artificial seawater was obtained from the correlation of the Grasholf number, Reynolds number and Prandtl number, as well as those for the water and the NaCl solution. Therefore, considering the physical properties of the artificial seawater, the thermal-hydraulic behavior of the seawater in single-phase flow can be estimated from the conventional thermal-hydraulic correlations for a single-phase flow.","PeriodicalId":8595,"journal":{"name":"Atomic Energy Society of Japan","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2016-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80544470","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}
Since the occurrence probability of multiple steam generator tube rupture (MSGTR) in a PWR is considered to be low, analytical or experimental investigation to prevent such accidents has not been performed explicitly. As new Japanese regulations require continuous effort to enhance the safety of nuclear power plants, low-probability but high-consequence events such as MSGTR should be taken into account to increase safety. In this study, a thermal-hydraulic analysis of multiple tube ruptures in a steam generator (SG) or all the SGs in a station blackout (SBO) situation was performed using the RETRAN-3D code in order to clarify the plant behavior during an MSGTR event and to contribute to risk reduction. The results show that a water supply function to SGs is important to cope with accidents involving MSGTR+SBO to prevent core damage. Furthermore, if the auxiliary feed water system loses its function when 10 tubes rupture in an SG, the time to core exposure is within 1 h and much shorter than that for SGTR, i.e., single tube rupture, in an SG. Therefore, in order to prevent core damage during MSGTR, it is desirable to have an alternative water injection system to quickly replace the auxiliary feed water system.
{"title":"Study of Thermal Hydraulic Behaviors during Multiple Steam Generator Tube Rupture Events in PWR","authors":"Keita Fujiwara, K. Muramatsu, H. Muta","doi":"10.3327/TAESJ.J15.005","DOIUrl":"https://doi.org/10.3327/TAESJ.J15.005","url":null,"abstract":"Since the occurrence probability of multiple steam generator tube rupture (MSGTR) in a PWR is considered to be low, analytical or experimental investigation to prevent such accidents has not been performed explicitly. As new Japanese regulations require continuous effort to enhance the safety of nuclear power plants, low-probability but high-consequence events such as MSGTR should be taken into account to increase safety. In this study, a thermal-hydraulic analysis of multiple tube ruptures in a steam generator (SG) or all the SGs in a station blackout (SBO) situation was performed using the RETRAN-3D code in order to clarify the plant behavior during an MSGTR event and to contribute to risk reduction. The results show that a water supply function to SGs is important to cope with accidents involving MSGTR+SBO to prevent core damage. Furthermore, if the auxiliary feed water system loses its function when 10 tubes rupture in an SG, the time to core exposure is within 1 h and much shorter than that for SGTR, i.e., single tube rupture, in an SG. Therefore, in order to prevent core damage during MSGTR, it is desirable to have an alternative water injection system to quickly replace the auxiliary feed water system.","PeriodicalId":8595,"journal":{"name":"Atomic Energy Society of Japan","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2016-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85159846","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}
One of the most important issues of the current PRA methodology is the precise modeling of dynamic changes such as state transitions among several states including fault (s) or maintenance of the nuclear facility, safety-related systems or components by fault-tree analysis and event-tree analysis. Moreover, though safety-related systems are usually in the stand-by state during normal operating conditions of a nuclear power plant, modeling of the dynamic changes in safety functions, along with changes in component failure rates due to the aging effect in the stand-by state or continuous/intermittent effects originating from external hazards, is also carried out under the same situation. On the basis of the background described above, the authors proposed a reliability analysis methodology of using the Markov state transition model applied to the digital reactor protection system of an ABWR plant, and demonstrated the applicability of the developed methodology using the component failure modes discussed in DIGREL, the task group of WGRisk belonging to OECD/NEA/CSNI. These studies showed that the PRA methodology including the state transition model can consider state transitions of components and time-dependent changes in component failure rates, and the relationship between this methodology and minimal cut sets for calculating the core damage frequency was also clarified.
{"title":"Proposal of PRA Methodology Considering State Transitions and Time-Dependent Failure Rates of Components","authors":"H. Muta, O. Furuya, K. Muramatsu","doi":"10.3327/TAESJ.J15.004","DOIUrl":"https://doi.org/10.3327/TAESJ.J15.004","url":null,"abstract":"One of the most important issues of the current PRA methodology is the precise modeling of dynamic changes such as state transitions among several states including fault (s) or maintenance of the nuclear facility, safety-related systems or components by fault-tree analysis and event-tree analysis. Moreover, though safety-related systems are usually in the stand-by state during normal operating conditions of a nuclear power plant, modeling of the dynamic changes in safety functions, along with changes in component failure rates due to the aging effect in the stand-by state or continuous/intermittent effects originating from external hazards, is also carried out under the same situation. On the basis of the background described above, the authors proposed a reliability analysis methodology of using the Markov state transition model applied to the digital reactor protection system of an ABWR plant, and demonstrated the applicability of the developed methodology using the component failure modes discussed in DIGREL, the task group of WGRisk belonging to OECD/NEA/CSNI. These studies showed that the PRA methodology including the state transition model can consider state transitions of components and time-dependent changes in component failure rates, and the relationship between this methodology and minimal cut sets for calculating the core damage frequency was also clarified.","PeriodicalId":8595,"journal":{"name":"Atomic Energy Society of Japan","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2016-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81890205","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. Ezure, K. Ito, Yuri Kameyama, H. Kamide, T. Kunugi
Toshiki EZURE, Kei ITO, Yuri KAMEYAMA, Hideki KAMIDE and Tomoaki KUNUGI Japan Atomic Energy Agency, 4002 Narita, Oarai-machi, Higashi Ibaraki-gun, Ibaraki 311-1393, Japan NDD Corporation, 1-1-6 Jonan, Mito-shi, Ibaraki 310-0803, Japan Graduate School of Engineering, Kyoto University, Kyoto daigaku-Katsura, Nishikyo-ku, Kyoto 615-8530, Japan (Received November 13, 2015; accepted in revised form March 17, 2016; published online July 27, 2016)
{"title":"Influence of Fluid Viscosity on Vortex Cavitation at a Suction Pipe Inlet","authors":"T. Ezure, K. Ito, Yuri Kameyama, H. Kamide, T. Kunugi","doi":"10.3327/TAESJ.J15.023","DOIUrl":"https://doi.org/10.3327/TAESJ.J15.023","url":null,"abstract":"Toshiki EZURE, Kei ITO, Yuri KAMEYAMA, Hideki KAMIDE and Tomoaki KUNUGI Japan Atomic Energy Agency, 4002 Narita, Oarai-machi, Higashi Ibaraki-gun, Ibaraki 311-1393, Japan NDD Corporation, 1-1-6 Jonan, Mito-shi, Ibaraki 310-0803, Japan Graduate School of Engineering, Kyoto University, Kyoto daigaku-Katsura, Nishikyo-ku, Kyoto 615-8530, Japan (Received November 13, 2015; accepted in revised form March 17, 2016; published online July 27, 2016)","PeriodicalId":8595,"journal":{"name":"Atomic Energy Society of Japan","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2016-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86076592","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}
Fumihiro Nakamori, Y. Ohishi, Masaya Kumagai, H. Muta, K. Kurosaki, K. Fukumoto, S. Yamanaka
{"title":"Mechanical and Thermal Properties of Fe 2 B","authors":"Fumihiro Nakamori, Y. Ohishi, Masaya Kumagai, H. Muta, K. Kurosaki, K. Fukumoto, S. Yamanaka","doi":"10.3327/TAESJ.J16.005","DOIUrl":"https://doi.org/10.3327/TAESJ.J16.005","url":null,"abstract":"","PeriodicalId":8595,"journal":{"name":"Atomic Energy Society of Japan","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2016-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74572195","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}
Takao Nakamura, S. Nakada, Kichisa Iwata, Tsutomu Ono, Fumio Hamasaki
Takao NAKAMURA, Setsuya NAKADA, Kichisa IWATA, Tsutomu ONO and Fumio HAMASAKI Japan Nuclear Safety Institute, 5-36-7 Shiba, Minato-ku, Tokyo 108-0014, Japan Volcano Research Center, Earthquake Research Institute, the University of Tokyo, 1-1-1 Yayoi, Bunkyo-ku, Tokyo 113-0032, Japan Electrical and Mechanical Engineering Office, Nuclear Power Engineering Department, Electric Power Development Co., Ltd., 15-1 Ginza 6-Chome, Chuo-ku, Tokyo 104-8165, Japan (Received February 9, 2016; accepted in revised form June 20, 2016; published online October 4, 2016)
{"title":"Safety Assessment of Nuclear Power Plant under Volcanic Phenomena Part 2: Revision of JEAG4625 on the Safety Assessment of Equipment Used for Measures in Severe Accident and Maintenance Program@@@重大事故等対処施設の影響評価等に関するJEAG4625の改定","authors":"Takao Nakamura, S. Nakada, Kichisa Iwata, Tsutomu Ono, Fumio Hamasaki","doi":"10.3327/TAESJ.J15.027","DOIUrl":"https://doi.org/10.3327/TAESJ.J15.027","url":null,"abstract":"Takao NAKAMURA, Setsuya NAKADA, Kichisa IWATA, Tsutomu ONO and Fumio HAMASAKI Japan Nuclear Safety Institute, 5-36-7 Shiba, Minato-ku, Tokyo 108-0014, Japan Volcano Research Center, Earthquake Research Institute, the University of Tokyo, 1-1-1 Yayoi, Bunkyo-ku, Tokyo 113-0032, Japan Electrical and Mechanical Engineering Office, Nuclear Power Engineering Department, Electric Power Development Co., Ltd., 15-1 Ginza 6-Chome, Chuo-ku, Tokyo 104-8165, Japan (Received February 9, 2016; accepted in revised form June 20, 2016; published online October 4, 2016)","PeriodicalId":8595,"journal":{"name":"Atomic Energy Society of Japan","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2016-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84356284","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}
Masahiko Ariyoshi, T. Takata, Akira Yamaguchi, H. Endo
Masahiko ARIYOSHI, Takashi TAKATA, Akira YAMAGUCHI and Hiroshi ENDO Department of Energy and Environment Engineering, Osaka University, 2-1 Yamada-oka, Suita-shi, Osaka 565-0871, Japan Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656, Japan Central Research Institute of Electric Power Industry, 2-11-1 Iwadokita, Komae-shi, Tokyo 201-8511, Japan (Received July 23, 2015; accepted in revised form January 15, 2016; published online June 15, 2016)
{"title":"Evaluation of the Fuel Melting Character of FBR Core Caused by Seismic Reactivity Insertion","authors":"Masahiko Ariyoshi, T. Takata, Akira Yamaguchi, H. Endo","doi":"10.3327/TAESJ.J15.010","DOIUrl":"https://doi.org/10.3327/TAESJ.J15.010","url":null,"abstract":"Masahiko ARIYOSHI, Takashi TAKATA, Akira YAMAGUCHI and Hiroshi ENDO Department of Energy and Environment Engineering, Osaka University, 2-1 Yamada-oka, Suita-shi, Osaka 565-0871, Japan Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656, Japan Central Research Institute of Electric Power Industry, 2-11-1 Iwadokita, Komae-shi, Tokyo 201-8511, Japan (Received July 23, 2015; accepted in revised form January 15, 2016; published online June 15, 2016)","PeriodicalId":8595,"journal":{"name":"Atomic Energy Society of Japan","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2016-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75398671","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}
Strontium - 90 / Y - 90 are major radionuclides observed in the water samples tested recently at the site of the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company. A simple method of evaluating Sr - 90 concentration in these water samples by measuring β rays from Y - 90 with a GM - detector setup was developed. By applying the precipitation method, Sr - 90 and Y - 90 were separated and quantitatively collected with a filter. β rays from Y - 90 in the filter were measured two times at appropriate intervals by inserting a polyethylene plate of 2 mm thickness as a β -ray absorber. The contribution of γ rays from Cs - 134 and Cs - 137 to the Y - 90 count rates was quantitatively evaluated using a 10-mm-thick acrylic resin plate. From the parent-daughter relationship between Sr - 90 and Y - 90, the Sr - 90 concentration was evaluated using the conversion coefficient of Y - 90 count rate （ cps ） to Sr - 90 concentration （ Bq / cm 3 ） . It was verified that Sr - 90 concentration of below 0.01 Bq / cm 3 in water samples can be correctly measured by this simple method.
{"title":"GM検出器を用いたY-90β線測定による水中のSr-90濃度の高感度・簡便測定法","authors":"近藤 健次郎, 英夫 平山, 雅文 平, 宏樹 松村, 広 岩瀬, 慎一 佐々木","doi":"10.3327/TAESJ.J14.034","DOIUrl":"https://doi.org/10.3327/TAESJ.J14.034","url":null,"abstract":"Strontium - 90 / Y - 90 are major radionuclides observed in the water samples tested recently at the site of the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company. A simple method of evaluating Sr - 90 concentration in these water samples by measuring β rays from Y - 90 with a GM - detector setup was developed. By applying the precipitation method, Sr - 90 and Y - 90 were separated and quantitatively collected with a filter. β rays from Y - 90 in the filter were measured two times at appropriate intervals by inserting a polyethylene plate of 2 mm thickness as a β -ray absorber. The contribution of γ rays from Cs - 134 and Cs - 137 to the Y - 90 count rates was quantitatively evaluated using a 10-mm-thick acrylic resin plate. From the parent-daughter relationship between Sr - 90 and Y - 90, the Sr - 90 concentration was evaluated using the conversion coefficient of Y - 90 count rate （ cps ） to Sr - 90 concentration （ Bq / cm 3 ） . It was verified that Sr - 90 concentration of below 0.01 Bq / cm 3 in water samples can be correctly measured by this simple method.","PeriodicalId":8595,"journal":{"name":"Atomic Energy Society of Japan","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2015-06-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78245290","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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