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Cs concentrations in east Japanese riversTsuji, Hideki*; Ishii, Yumiko*; Shin, M.*; Taniguchi, Keisuke*; Arai, Hirotsugu*; Kurihara, Momo*; Yasutaka, Tetsuo*; Kuramoto, Takayuki*; Nakanishi, Takahiro; Lee, S*; et al.
Science of the Total Environment, 697, p.134093_1 - 134093_11, 2019/12
Times Cited Count：17 Percentile：59.38(Environmental Sciences)To investigate the main factors that control the dissolved radiocesium concentration in river water in the area affected by the Fukushima Daiichi Nuclear Power Plant accident, the correlations between the dissolved Cs concentrations at 66 sites normalized to the average Cs inventories for the watersheds with the land use, soil components, topography, and water quality factors were assessed. We found that the topographic wetness index is significantly and positively correlated with the normalized dissolved Cs concentration. Similar positive correlations have been found for European rivers because wetland areas with boggy organic soils that weakly retain Cs are mainly found on plains. However, for small Japanese river watersheds, the building area ratio in the watershed strongly affected the dissolved Cs concentration.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki
Journal of Nuclear Science and Technology, 56(2), p.201 - 209, 2019/02
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was developed to expand application range of an existing computer code. Applicability of the method was demonstrated through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki
JAEA-Research 2017-007, 61 Pages, 2017/07
JAEA-Research-2017-007.pdf:4.3MB
For safety assessment of a steam generator of sodium-cooled fast reactors, it is necessary to evaluate the possibility of occurring tube failure propagation and of water leak rate under sodium-water reaction accident. In the previous studies, a computer code called LEAP-II calculating a wastage-type failure propagation and the water leak rate during long-time event progress was developed. In this study, a numerical method to evaluate the possibility of occurring overheating rupture was introduced into the LEAP-II code to expand application range of this code. The completed code is called LEAP-III. The test analysis on a tube bundle configuration demonstrated that the overheating rupture model could provide conservative prediction.
Uchibori, Akihiro; Kikuchi, Shin; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu, B, 79(808), p.2635 - 2639, 2013/12
Multi-physics analysis system for a heat transfer tube failure event in a steam generator of sodium-cooled fast reactors has been developed. In this study, applicability of the newly constructed numerical models in the analysis system was investigated. The droplet entrainment / transport model which was incorporated into the SERAPHIM code was verified through the analysis of the related experiment. The experimental data about the pressure variation when the droplet entrainment occurs was reproduced by our model successfully. The TACT code is integrated by the numerical models of fluid-structure thermal coupling, stress evaluation and failure judgment of the structure. The fluid-structure thermal coupling model could predict the temperature distribution formed by the flow around the circular cylinder. About the failure judgment model, the predicted time of failure occurrence showed good agreement with the results of the tube rupture simulation experiment.
Uchibori, Akihiro; Kikuchi, Shin; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki
Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 6 Pages, 2013/07
Multiphysics analysis system was newly developed to evaluate possibility of failure propagation occurrence under tube failure accident in a steam generator of sodium cooled fast reactors. The system consists of the computer codes, SERAPHIM, TACT, RELAP5, which are based on the mechanistic numerical models. The SERAPHIM codes calculates multicomponent multiphase flow involving sodium-water chemical reaction. In this study, a numerical model for chemical reaction about production of a sodium monoxide and its transport process were constructed. We also developed the numerical models of the TACT code for evaluation of shell-side flow around an adjacent tube, heat transfer from the fluid to the tube and occurrence of tube failure. In our analysis system, thermal hydraulic behavior of water inside the tube is evaluated by the RELAP5 code. The original heat transfer correlations were corrected for the rapidly heated tube in the present work.
Okumura, Susumu; Ishibori, Ikuo; Kurashima, Satoshi; Yoshida, Kenichi; Yuyama, Takahiro; Ishizaka, Tomohisa; Miyawaki, Nobumasa; Kashiwagi, Hirotsugu; Yuri, Yosuke; Nara, Takayuki; et al.
JAEA-Review 2012-046, JAEA Takasaki Annual Report 2011, P. 172, 2013/01
The AVF cyclotron was operated from May 9th in fiscal 2011 and all the planned experiments on April were cancelled due to scheduled blackouts originated from the Great East Japan Earthquake. The total operation time amounted to 3,038.4 hours. The extended operation from Friday evening to Saturday evening was carried out eight times in order to supply the cancellation on April. The regular yearly overhaul and maintenance were carried out.
Nara, Takayuki; Ishibori, Ikuo; Kurashima, Satoshi; Yoshida, Kenichi; Yuyama, Takahiro; Ishizaka, Tomohisa; Okumura, Susumu; Miyawaki, Nobumasa; Kashiwagi, Hirotsugu; Yuri, Yosuke; et al.
JAEA-Review 2011-043, JAEA Takasaki Annual Report 2010, P. 172, 2012/01
no abstracts in English
Nara, Takayuki; Ishibori, Ikuo; Kurashima, Satoshi; Yoshida, Kenichi; Yuyama, Takahiro; Ishizaka, Tomohisa; Okumura, Susumu; Miyawaki, Nobumasa; Kashiwagi, Hirotsugu; Yuri, Yosuke; et al.
JAEA-Review 2010-065, JAEA Takasaki Annual Report 2009, P. 180, 2011/01
no abstracts in English
Ogawa, Hiroaki; Sugie, Tatsuo; Kasai, Satoshi*; Katsunuma, Atsushi*; Hara, Hirotsugu*; Takeyama, Norihide*; Kusama, Yoshinori
Fusion Engineering and Design, 83(10-12), p.1405 - 1409, 2008/12
Times Cited Count：15 Percentile：66.71(Nuclear Science & Technology)no abstracts in English
Nara, Takayuki; Ishibori, Ikuo; Kurashima, Satoshi; Yoshida, Kenichi; Yuyama, Takahiro; Ishizaka, Tomohisa; Okumura, Susumu; Miyawaki, Nobumasa; Kashiwagi, Hirotsugu; Yuri, Yosuke; et al.
JAEA-Review 2008-055, JAEA Takasaki Annual Report 2007, P. 190, 2008/11
no abstracts in English
Yoshikawa, Ryuji; Ohshima, Hiroyuki; Hamada, Hirotsugu; Kurihara, Akikazu; Uchibori, Akihiro
JAEA-Research 2008-055, 24 Pages, 2008/06
JAEA-Research-2008-055.pdf:3.19MB
In Japan Atomic Energy Agency, thermal hydraulic studies on sodium-water reaction are being performed with the multi-component and multi-phase code SERAPHIM. The interfacial area concentration of sodium droplets in the steam is important for the accurate analysis of sodium-water reaction. In this report, the theoretical analysis and numerical models for gas jets were reviewed to understand the mixing process of sodium and water. As for theoretical analysis, existing critical flow rate, depressurization and entrainment analysis for jet flows were summarized. The applicability of critical flow rate equations for subcooled water at 17MPa were confirmed after investigating its effect of compressibility. Based on the available knowledge on entrained droplet sizes in gas jets, a transport equation of sodium droplet interfacial area concentration was constructed for multiphase flow simulation.
Nara, Takayuki; Agematsu, Takashi; Ishibori, Ikuo; Kurashima, Satoshi; Yoshida, Kenichi; Yuyama, Takahiro; Ishizaka, Tomohisa; Okumura, Susumu; Miyawaki, Nobumasa; Kashiwagi, Hirotsugu; et al.
JAEA-Review 2007-060, JAEA Takasaki Annual Report 2006, P. 204, 2008/03
no abstracts in English
Yuyama, Takahiro; Nara, Takayuki; Ishibori, Ikuo; Kurashima, Satoshi; Yoshida, Kenichi; Ishizaka, Tomohisa; Okumura, Susumu; Miyawaki, Nobumasa; Kashiwagi, Hirotsugu; Yuri, Yosuke; et al.
Proceedings of 5th Annual Meeting of Particle Accelerator Society of Japan and 33rd Linear Accelerator Meeting in Japan (CD-ROM), p.259 - 261, 2008/00
no abstracts in English
Ogawa, Hiroaki; Sugie, Tatsuo; Kasai, Satoshi; Katsunuma, Atsushi*; Hara, Hirotsugu*; Kusama, Yoshinori
Plasma and Fusion Research (Internet), 2, p.S1054_1 - S1054_4, 2007/11
no abstracts in English
Nara, Takayuki; Agematsu, Takashi; Ishibori, Ikuo; Kurashima, Satoshi; Yoshida, Kenichi; Okumura, Susumu; Miyawaki, Nobumasa; Kashiwagi, Hirotsugu; Yuri, Yosuke; Yokota, Wataru; et al.
JAEA-Review 2006-042, JAEA Takasaki Annual Report 2005, P. 196, 2007/02
no abstracts in English
Hishida, Masahiko; Murakami, Tsutomu*; Kisohara, Naoyuki; Fujii, Tadashi; Uchita, Masato*; Hayafune, Hiroki; Chikazawa, Yoshitaka; Usui, Shinichi; Ikeda, Hirotsugu; Uno, Osamu; et al.
JAEA-Research 2006-006, 125 Pages, 2006/03
JAEA-Research-2006-006.pdf:11.55MB
In Phase I of the "Feasibility Studies on Commercialized Fast Reactor Cycle Systems (F/S)", an advanced loop type reactor has been selected as a promising concept of sodium-cooled reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase II, design improvement for further cost reduction and the establishment of the plant concept has been performed. In this study, reactor core design and large-scale plant design have been performed by adopting the modified fuel assembly with inner duct structure and double-wall straight tube steam generator (SG), which concepts were chosen at the interim review of FY 2003. For this SG, safety logics have been studied and the structural concept has been established. And the plant designs improving the in-service inspection (ISI) and repair capability have been performed. Furthermore, elaborate confirmation of the design has been performed reflecting the development of elemental technology, back-up concepts have been proposed. Besides, cost reduction measures have been studied by reducing reactor grade materials, introducing autonomous standardizations, simplifying the design due to deregulation and adopting systemized standards for BOP and NSSS. From now on, reflecting the results of elemental experiments, in-depth design studies and examination of critical issues will be carried out and the plant concept will accomplish in preparation for the final evaluation in Phase II.
Nara, Takayuki; Agematsu, Takashi; Ishibori, Ikuo; Kurashima, Satoshi; Yoshida, Kenichi; Fukuda, Mitsuhiro; Okumura, Susumu; Miyawaki, Nobumasa; Kashiwagi, Hirotsugu; Nakamura, Yoshiteru; et al.
JAEA-Review 2005-001, TIARA Annual Report 2004, P. 370, 2006/01
no abstracts in English
Seino, Hiroshi; Jitsu, Koji*; Kurihara, Akikazu; Ono, Isao*; Hamada, Hirotsugu
Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13), 0 Pages, 2005/05
Blow down analysis code (LEAP-BLOW) and sodium-water reaction jet analysis code (LEAP-JET) have been developed to improve the evaluation accuracy on sodium-water reaction. The validation analyses by these codes were carried out using the data of SWAT-1R test. As the result, though there was a problem in the quantitative evaluation of LEAP-JET, it was possible to obtain the approximately appropriate results.
Toyohara, Daisuke*; Ohno, Shuji; Matsuki, Takuo*; Hamada, Hirotsugu; Miyahara, Shinya
JNC TN9520 2004-004, 151 Pages, 2005/01
JNC-TN9520-2004-004.pdf:139.32MB
A computer code TRACER (Transport Phenomena of Radionuclides for Accident Consequence Evaluation of Reactor) version 2.3 has been developed to evaluate the species and the quantities of fission products (FPs) released into cover gas during fuel pin failure accident in an LMFBR. TRACER version 2.3 includes new and modified models shown below. a) Booth model, a new model for FPs release from fuel. b) Modified model for FPs transfer from fuel to bubble or sodium coolant. c) Modified model for bubbles dynamics in coolant. Computer models, input data and output data of TRACER Version 2.3 are described in this user's manual.
Ohno, Shuji; Nishimura, Masahiro; Hamada, Hirotsugu; Miyahara, Shinya; Sasa, Toshinobu*; Kurata, Yuji*
JNC TN9400 2004-072, 52 Pages, 2005/01
JNC-TN9400-2004-072.pdf:2.88MB
A series of equilibrium evaporation experiment was performed to acquire the essential and the fundamental knowledge about the transfer behavior of lead-bismuth eutectic(LBE) and impurity tellurium in LBE from liquid to gas phase. The experiments were conducted using the transpiration method in which saturated vapor in an isothermal evaporation pot was transported by inert carrier gas and collected outside of the pot. The size of the used evaporation pot is 8cm inner diameter and 15cm length. The weight of the LBE pool in the pot is about 500g. The investigated temperature range was 450deg-C to 750deg-C. From this experiment and discussion using the data in literature, we have obtained several instructive and useful data on the LBE evaporation behavior such as saturated vapor pressure of LBE, vapor concentration of Pb, Bi and Bi2 in LBE saturated gas phase, and activity coefficient of Pb in the LBE. The LBE vapor pressure equation is represented as the sum of Pb, Bi and Bi2 vapor in the temperature range between 550deg-C and 750deg-C as logP[Pa]=10.2-10100/T[K]. The gas-liquid equilibrium partition coefficient of tellurium in LBE is in the range of 10 to 100, with no remarkable temperature dependency between 450deg-C and 750deg-C.
