Atomic Energy Society of Japan最新文献_第9页

Development of Non-destructive Assay System using Fast Neutron Direct Interrogation Method for Actual Uranium Waste Drums 快中子直接探询法铀废桶无损检测系统的研制

Atomic Energy Society of Japan Pub Date : 2016-01-01 DOI: 10.3327/TAESJ.J15.011

A. Ohzu, M. Komeda, M. Kureta, Naoki Zaima, Y. Nakatsuka, Shinichi Nakashima

{"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}

引用次数: 8

Establishment of Integrity Evaluation Method for Reserved Shutdown System of High-Temperature Engineering Test Reactor (HTTR) 高温工程试验堆(HTTR)备用停堆系统完整性评价方法的建立

Atomic Energy Society of Japan Pub Date : 2016-01-01 DOI: 10.3327/taesj.J15.016

S. Hamamoto, T. Kawamoto, M. Kondo, Hiroaki Sawahata, S. Takada, Masayuki Shinozaki

{"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}

引用次数: 0

Effect of Seawater on Heat Transfer without Boiling in Internally Heated Annulus 海水对内加热环空不沸腾换热的影响

Atomic Energy Society of Japan Pub Date : 2016-01-01 DOI: 10.3327/TAESJ.J15.024

Shin-ichiro Uesawa, Wei Liu, Lifang Jiao, T. Nagatake, K. Takase, Mitsuhiko Shibata, H. Yoshida

{"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}

引用次数: 4

Study of Thermal Hydraulic Behaviors during Multiple Steam Generator Tube Rupture Events in PWR 压水堆蒸汽发生器多管破裂过程热水力特性研究

Atomic Energy Society of Japan Pub Date : 2016-01-01 DOI: 10.3327/TAESJ.J15.005

Keita Fujiwara, K. Muramatsu, H. Muta

{"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}

引用次数: 0

Proposal of PRA Methodology Considering State Transitions and Time-Dependent Failure Rates of Components 考虑状态转移和部件时变故障率的PRA方法建议

Atomic Energy Society of Japan Pub Date : 2016-01-01 DOI: 10.3327/TAESJ.J15.004

H. Muta, O. Furuya, K. Muramatsu

{"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}

引用次数: 1

Influence of Fluid Viscosity on Vortex Cavitation at a Suction Pipe Inlet 流体粘度对吸入口涡蚀的影响

Atomic Energy Society of Japan Pub Date : 2016-01-01 DOI: 10.3327/TAESJ.J15.023

T. Ezure, K. Ito, Yuri Kameyama, H. Kamide, T. Kunugi

引用次数: 0

Mechanical and Thermal Properties of Fe 2 B fe2b的力学和热性能

Atomic Energy Society of Japan Pub Date : 2016-01-01 DOI: 10.3327/TAESJ.J16.005

Fumihiro Nakamori, Y. Ohishi, Masaya Kumagai, H. Muta, K. Kurosaki, K. Fukumoto, S. Yamanaka

引用次数: 6

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の改定火山现象下核电厂安全评价第2部分:JEAG4625关于重大事故措施和维修计划设备安全评价的修订JEAG4625

Atomic Energy Society of Japan Pub Date : 2016-01-01 DOI: 10.3327/TAESJ.J15.027

Takao Nakamura, S. Nakada, Kichisa Iwata, Tsutomu Ono, Fumio Hamasaki

引用次数: 0

Evaluation of the Fuel Melting Character of FBR Core Caused by Seismic Reactivity Insertion 地震反应性插入对快堆堆芯燃料熔化特性的影响

Atomic Energy Society of Japan Pub Date : 2016-01-01 DOI: 10.3327/TAESJ.J15.010

Masahiko Ariyoshi, T. Takata, Akira Yamaguchi, H. Endo

引用次数: 0

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・簡便測定法使用GM检测器的Y-90β射线测定水中Sr-90浓度的高灵敏度简便测定法

Atomic Energy Society of Japan Pub Date : 2015-06-30 DOI: 10.3327/TAESJ.J14.034

近藤健次郎, 英夫平山, 雅文平, 宏樹松村, 広岩瀬, 慎一佐々木

{"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}

引用次数: 0