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Yamamoto, Tomohiko; Matsubara, Shinichiro*; Harada, Hidenori*; Saunier, P.*; Martin, L.*; Gentet, D.*; Dirat, J.-F.*; Collignon, C.*
Nuclear Engineering and Design, 383, p.111406_1 - 111406_14, 2021/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Japan-France collaboration on ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project is launched in 2014. In this project, Japan-France evaluates core assemblies with interferences on seismic event. The object of this study is to verify the seismic evaluation method on core assemblies between Japan and France by comparing the results. The analysis of this benchmark calculation shows a satisfactory agreement between the Japanese and French tools and the figures show a good behavior of the core in horizontal direction under French seismic condition.
Soba, A.*; Prudil, A.*; Zhang, J.*; Dethioux, A.*; Han, Z.*; Dostal, M.*; Matocha, V.*; Marelle, V.*; Lasnel-Payan, J.*; Kulacsy, K.*; et al.
Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10
Yamamoto, Tomohiko; Matsubara, Shinichiro*; Harada, Hidenori*; Saunier, P.*; Martin, L.*; Gentet, D.*; Dirat, J.-F.*; Collignon, C.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05
Japan-France collaboration on ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project is launched in 2014. In this project, Japan-France evaluates core assemblies with interferences on seismic event. The object of this study is to verify the seismic evaluation method on core assemblies between Japan and France by comparing the results. The analysis of this benchmark calculation shows a satisfactory agreement between the Japanese and French tools and the figures show a good behavior of the core in horizontal direction under French seismic condition.
Kim, S. B.*; Zhang, Y.*; Won, S. M.*; Bandodkar, A. J.*; Sekine, Yurina; Xue, Y.*; Koo, J.*; Harshman, S. W.*; Martin, J. A.*; Park, J. M.*; et al.
Small, 14(12), p.1703334_1 - 1703334_11, 2018/03
Times Cited Count：98 Percentile：95.52(Chemistry, Multidisciplinary)Passerini, S.*; Carardi, C.*; Grandy, C.*; Azpitarte, O. E.*; Chocron, M.*; Japas, M. L.*; Bubelis, E.*; Perez-Martin, S.*; Jayaraj, S.*; Roelofs, F.*; et al.
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.780 - 790, 2015/05
Obara, Satoshi; Takaya, Shigeru; Wakai, Takashi; Asayama, Tai; Suzuki, Hiroshi; Saito, Toru; Martin, L.*
Kensa Gijutsu, 16(3), p.24 - 30, 2011/03
For the commercialization of fast breeder reactors (FBR), it is essential to enhance the economic competitiveness by reduction of total cost by elongation of plant service period. In this point of view, it is important to establish the assessment method of integrity of aged weld joint and repair welding for the components of future long life FBR. In the present study, evolution of microstructure is evaluated for the 304SS-304SS weld joint which was used for 88,000h at 526-545
C in the French proto-type fast reactor Phenix (as secondary pipes), and for the repair weld joint made from the 304SS of Phenix and new 316LSS plate. In addition, residual stress of the joints were measured by means of RESA and RESA-II. As the results, the microstructure and the residual stress of the joints had changed in the high temperature-long service environment and by the repair welding, and those of the repair weld joint were correlated with its hardness.
Walsh, M.*; Andrew, P.*; Barnsley, R.*; Bertalot, L.*; Boivin, R.*; Bora, D.*; Bouhamou, R.*; Ciattaglia, S.*; Costley, A. E.*; Counsell, G.*; et al.
Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2011/03
Obara, Satoshi; Takaya, Shigeru; Wakai, Takashi; Asayama, Tai; Suzuki, Hiroshi; Saito, Toru*; Martin, L.*
Hozengaku, 9(1), p.32 - 38, 2010/04
For the commercialization of fast breeder reactors (FBR), it is essential to enhance the economic competitiveness by reduction of total cost by elongation of plant service period. In this point of view, it is important to establish the assessment method of integrity of aged weld joint and repair welding for the components of future long life FBR. In the present study, evolution of microstructure is evaluated for the 304SS-304SS weld joint which was used for 88,000h at 526-545C in the French proto-type fast reactor Phenix (secondary pipes), and for the repair weld joint made from the 304SS of Phenix and new 316LSS plate. In addition, residual stress of the joints were measured by means of RESA and RESA-II. As the results, the microstructure and the residual stress of the joints had changed in the high temperature-long service environment and by the repair welding, and those of the repair weld joint were correlated with its hardness.
Faenov, A. Y.; Magunov, A. I.*; Pikuz, T. A.*; Skobelev, I. Y.*; Giulietti, D.*; Betti, S.*; Galimberti, M.*; Gamucci, A.*; Giulietti, A.*; Gizzi, L. A.*; et al.
JETP Letters, 86(3), p.178 - 183, 2007/08
Times Cited Count：5 Percentile：37.28(Physics, Multidisciplinary)Doyle, E. J.*; Houlberg, W. A.*; Kamada, Yutaka; Mukhovatov, V.*; Osborne, T. H.*; Polevoi, A.*; Bateman, G.*; Connor, J. W.*; Cordey, J. G.*; Fujita, Takaaki; et al.
Nuclear Fusion, 47(6), p.S18 - S127, 2007/06
no abstracts in English
McDonald, D. C.*; Cordey, J. G.*; Thomsen, K.*; Kardaun, O. J. W. F.*; Snipes, J. A.*; Greenwald, M.*; Sugiyama, L.*; Ryter, F.*; Kus, A.*; Stober, J.*; et al.
Nuclear Fusion, 47(3), p.147 - 174, 2007/03
Times Cited Count：51 Percentile：29.82(Physics, Fluids & Plasmas)This paper describes the updates to and analysis of the International Tokamak Physics Activity (ITPA) Global H-node Confinement Database version 3 (DB3) over the period 1994-2004. Global data, for the energy confinement time and its controlling parameters, have now been collected from 18 machines of different sizes and shapes: ASDEX, ASDEX Upgrade, C-Mod CoMPASS-D, DIII-D, JET, JFT-2M, JT-60U, MAST, NSTX, PBX-M, PDX, START, T-10, TCV, TdeV, TFTR and TUMAN-3M. A wide range of physics studies has been performed on DB3 with particular progress made in the separation of core and edge behavior, dimensionless parameter analyses and the comparison of the database with one-dimensional transport code. A key aim of the database has always been to provide a basis for estimating the energy confinement properties of next step machines such as ITER, and so the impact of the database and its analysis on such machines is also discussed.
Be halo nucleus from 
Bi at coulomb barrierMazzocco, M.*; Signorini, C.*; Romoli, M.*; De Francesco, A.*; Di Pietro, M.*; Vardaci, E.*; Yoshida, Koichi*; Yoshida, Atsushi*; Bonetti, R.*; De Rosa, A.*; et al.
European Physical Journal A, 28(3), p.295 - 299, 2006/06
Times Cited Count：46 Percentile：90.03(Physics, Nuclear)The scattering of the radioactive, weakly bound, halo nucleus Be from Bi has been studied at 40 MeV. The measurement performed with a low-intensity and a large-emittance secondary beam could be made using an extremely compact, large solid angle (
2
sr) detecting set-up, based on 8 highly segmented Si telescopes. The 
Be scattering angular distributions, as well as their relative reaction cross-sections, resulted to be rather similar. This may suggest that at Coulomb barrier energies the halo structure and the small weakly binding energy of the Be projectile have no big influence on the reaction dynamics.
and collisionality approaching ITER conditionsCordey, J. G.*; Thomsen, K.*; Chudnovskiy, A.*; Kardaun, O. J. W. F.*; Takizuka, Tomonori; Snipes, J. A.*; Greenwald, M.*; Sugiyama, L.*; Ryter, F.*; Kus, A.*; et al.
Nuclear Fusion, 45(9), p.1078 - 1084, 2005/09
Times Cited Count：51 Percentile：82.4(Physics, Fluids & Plasmas)The condition of the latest version of the ELMy H-mode database has been re-examined. It is shown that there is bias in the ordinary least squares regression for some of the variables. To address these shortcomings three different techniques are employed: (a)principal component regression, (b)an error in variables technique and (c)the selection of a better conditioned dataset with fewer variables. Scalings in terms of the dimensionless physics valiables, as well as the standard set of engineering variables, are derived. The new scalings give a very similar performance for existing scalings for ITER at the standard beta, but a much improvement performance at higher beta.
Wakai, Takashi; Onizawa, Takashi; Ando, Masanori; Aoto, Kazumi; Martin, L.*
Proceedings of Creep and Fracture in High Temperature Components-design and life Assessment Issues (CD-ROM), 11 Pages, 2005/09
For elongation of the plant life, damaged pipes and/or components should be replaced during the plant service period. In this case, some on-site welded joints may take place. In order to assess the structural integrity of such on-site welded joints, metallurgical examination and mechanical tests have been conducted. Comparing the microstructures of aged factory welded joints and those of newly produced on-site welded joints made of an austenitic stainless steel, the influence of heat input during the on-site welding process on the integrity of the joints are discussed.
Ioki, Kimihiro*; Barabaschi, P.*; Barabash, V.*; Chiocchio, S.*; Daenner, W.*; Elio, F.*; Enoeda, Mikio; Gervash, A.*; Ibbott, C.*; Jones, L.*; et al.
Nuclear Fusion, 43(4), p.268 - 273, 2003/04
Times Cited Count：21 Percentile：54.59(Physics, Fluids & Plasmas)Although the basic concept of the vacuum vessel (VV) and in-vessel components of the ITER design has stayed the same, there have been several detailed design improvements resulting from efforts to raise reliability, to improve maintainability and to save money. One of the most important achievements in the VV R&D has been demonstration of the necessary fabrication and assembly tolerances. Recently the deformation due to cutting of the port extension was measured and it was shown that the deformation is small and acceptable. Further development of advanced methods of cutting, welding and NDT on a thick plate have been continued in order to refine manufacturing and improve cost and technical performance. With regard to the related FW/blanket and divertor designs, the R&D has resulted in the development of suitable technologies. Prototypes of the FW panel, the blanket shield block and the divertor components have been successfully fabricated.
Kurisaka, Kenichi; Takamatsu, Misao; Martin, L.*
no journal, ,
This study aims to quantify the probability distribution of the leak flow rate when a sodium leak event takes place, in terms of the effectiveness evaluation of accident management measures using probabilistic safety assessment. For this purpose, sodium leak instances that were experienced in domestic and foreign sodium-cooled fast reactor systems were investigated and analyzed. In most of these leak instances, individual total leak amount is known, but the leak duration time is unknown. Therefore, the previous study needed to assume the leak duration time as a probability distribution to estimate the leak flow rate. In this study, for more realistic evaluation, we investigated both total leak amount and leak duration time of sodium leak instances that were experienced in the Phenix reactor system. For 12 instances where total leak amount was already known, leak duration time became clear. Five leak instances were newly added. The probability distribution of the leak duration time was statistically analyzed by using these data. As a result, it became possible to quantify more realistically the probability distribution of the leak flow rate.
