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Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Nagasawa, Kazuyoshi*; Kurihara, Akikazu; Tanaka, Masaaki
JAEA-Research 2022-009, 125 Pages, 2023/01
JAEA-Research-2022-009.pdf:29.22MB
The design studies of an advanced loop-type sodium-cooled fast reactor (Advanced- SFR) have been carried out by the Japan Atomic Energy Agency (JAEA). At the core outlet, temperature fluctuations occur due to mixing of hot sodium from the fuel assembly with cold sodium from the control rod channels and radial blanket assembly. These temperature fluctuations may cause high cycle thermal fatigue around a bottom of Upper Internal Structure (UIS) located above the core. Therefore, we conducted a water experiment using a 1/3 scale 60 degree sector model that simulated the upper plenum of the advanced loop-type sodium-cooled reactor. And we proposed some countermeasures against large temperature fluctuations that occur at the bottom of the UIS. In this report, we have summarized that the effect of the countermeasure structure to mitigate the temperature fluctuation generated at the bottom of UIS is confirmed, and the Reynolds number dependency of the countermeasure structure and the characteristics of the temperature fluctuation on the control rod surface.
Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki
Hozengaku, 20(3), p.89 - 96, 2021/10
Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS) of Advanced-SFR. Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of UIS. We focused on the temperature fluctuations near the primary and backup control rod channels, and studied the countermeasure structure to mitigate the temperature fluctuation through temperature distribution and flow velocity distribution measurements. As a result, effectiveness of the countermeasure to mitigate the temperature fluctuation intensity was confirmed.
Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki
Hozengaku, 20(3), p.97 - 101, 2021/10
Focusing on the thermal striping phenomena that occurs at a bottom of the internal structure of an advanced sodium-cooled fast reactor (Advanced-SFR) that has been designed by the Japan Atomic Energy Agency, a water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of Upper Internal Structure (UIS). In the previous paper, we reported the effect of measures to mitigate temperature fluctuations around the control rod channels. In this paper, the same test section was used, and a water experiment was conducted to obtain the characteristics of temperature fluctuations around the radial blanket fuel assembly. And the shape of the Core Instrumentation Support Plate (CIP) was modified, and it was confirmed that it was highly effective in alleviating temperature fluctuations around the radial blanket fuel assembly.
Wakai, Takashi; Kobayashi, Sumio; Kato, Shoichi; Ando, Masanori; Takasho, Hideki*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
This paper describes a thermal fatigue test on a structural model with a dissimilar welded joint. In the present design of JSFR, there may be dissimilar welded joints between ferritic and austenitic steels especially in IHX and SG. Creep-fatigue is one of the most important failure modes in JSFR components. However, the creep-fatigue damage evaluation method has not been established for dissimilar welded joint. To investigate the evaluation method, structural test will be needed for verification. Therefore, a thermal fatigue test on a thick-wall cylinder with a circumferential dissimilar welded joint between Mod.9Cr-1Mo steel and 304SS was performed. Since the coefficients of thermal expansion of these steels were significantly different, buttering layer of Ni base alloy was installed between them. After the completion of the test, deep cracks were observed at the HAZ in 304SS, as well as at HAZ in Mod.9Cr-1Mo steel. There were many tiny surface cracks in BM of 304SS. According to the fatigue damage evaluation based on the finite element analysis results, the largest fatigue damage was calculated at HAZ in 304SS. Large fatigue damage was also estimated at BM of 304SS. Fatigue cracks were observed at HAZ and BM of 304SS in the test, so that analytical results are in a good agreement with the observations. However, though relatively small fatigue damage was estimated at HAZ in Mod.9Cr-1Mo steel, deep fatigue cracks were observed in the test. To identify the cause of such a discrepancy between the test and calculations, we performed a series of finite element analyses. Some metallurgical investigations were also performed.
Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 5 Pages, 2016/11
JAEA has been conducting a design study for an advanced large-scale sodium-cooled fast reactor (SFR). Hot sodium from the fuel subassembly can mix with the cold sodium from the control rod (CR) channel at the bottom of Upper Internal Structure (UIS). Temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of UIS. JAEA had performed a water experiment to examine countermeasures for the significant temperature fluctuation generated at the bottom of SFRs UIS. Meanwhile, a self-actuated shutdown system (SASS) is equipped in a backup control rod (BCR) channel to ensure reactor shutdown. The BCR guide tubes have a flow guide structure "flow-collector" to provide reliable operation of SASS. Flow-collector may affect the thermal mixing behavior at the bottom of the UIS. This study has investigated the influence of the flow- collector on characteristics of the temperature fluctuation around the BCR channels.
Kobayashi, Jun; Ezure, Toshiki; Kamide, Hideki; Oyama, Kazuhiro*; Watanabe, Osamu*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
A column type upper internal structure (UIS) is installed in the upper plenum of reactor vessel in JSFR. High cycle thermal fatigue may occur at the bottom plate (CIP) of the UIS where the hot sodium from the fuel subassembly can mix with the cold sodium from the control rod channel and the blanket fuel subassembly. We have been conducted a water experiment using a reactor upper plenum model to grasp the thermal-hydraulic phenomena around control rod (CR) channels, and to obtain countermeasures for significant temperature fluctuation on the CIP. The experimental apparatus has 1/3 scale and 60
sector model of the reactor upper plenum. By the experiment, characteristics of fluid temperature fluctuation between the handling head of the assemblies and the CIP are measured and countermeasure for the significant temperature fluctuation generation will be discussed on the influence of the distance from the handling head outlet to the lower surface of the CIP.
Suzuki, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Hirose, Takanori; Hayashi, Kimio; Tanigawa, Hisashi; Ochiai, Kentaro; Nishitani, Takeo; Tobita, Kenji; Akiba, Masato
Nuclear Fusion, 46(2), p.285 - 290, 2006/02
Times Cited Count：2 Percentile：7.05(Physics, Fluids & Plasmas)This paper presents significant progress in R&D of key technologies on the water-cooled solid breeder blanket for the ITER-TBM in JAERI. By the improvement of heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H, can be obtained by homogenizing it at 1150 C followed by normalizing at 930 C after the HIP process. Moreover, a promising bonding process for a tungsten armor and an F82H structural material was developed by using a uniaxial hot compression without any artificial compliant layer. Also, it has been confirmed that a fatigue lifetime correlation, which was developed for ITER divertor, can be applicable for F82H first wall mock-up. As for R&D on a breeder material, Li
TiO
, the effect of compression loads on thermal conductivity of pebble beds has been clarified. JAERI have extensively developed key technologies for ITER-TBM, and now steps up into an engineering R&D stage, where integrated performance of TBM structures will be demonstrated by scalable mock-ups.
Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi
JAERI-Research 2005-005, 23 Pages, 2005/03
JAERI-Research-2005-005.pdf:1.65MB
Fuel elements used in The Reduced-Moderation Water Reactor (RMWR) have the lamellar structure consisting of MOX pellets and UO blankets in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod. Therefore, it is important to evaluate the local deformation behavior due to the transient temperature distribution. To estimate the thermal deformation behavior, the temperature and stress distribution of the fuel cladding tube assumed in the designed reactor were analyzed. Moreover, basic physical properties and mechanical properties for analyzing the deformation behavior were obtained by experiment using fuel cladding tubes made of candidate alloys. In addition, the appropriate experimental conditions for realizing the practical thermal deformation behavior of the fuel cladding tube was selected by adjusting the testing temperature distribution based on data obtained with thermal analysis.
Hirai, Takeshi*; Ezato, Koichiro; Majerus, P.*
Materials Transactions, 46(3), p.412 - 424, 2005/03
Times Cited Count：111 Percentile：90.2(Materials Science, Multidisciplinary)no abstracts in English
Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi
JAERI-Tech 2004-035, 18 Pages, 2004/03
JAERI-Tech-2004-035.pdf:0.81MB
Fuel elements used in the Reduced-Moderation Water Reactor (RMWR) have the stacking structure consisting of MOX pellets and UO blankets in a fuel rod in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod in comparison with the present LWR fuels. For this reason, it is important to estimate local deformation behavior of the fuel cladding tube with temperature difference caused by MOX pellet and UO blanket. The testing machine was designed to investigate thermal-fatigue behavior under biaxial stress condition. The testing machine consists of the temperature distribution control unit, low cycle fatigue testing unit and internal pressure loading unit, it is also possible to conduct the simulation tests to investigate effects of pressure change with burn-up and longitudinal load change due to operation modes and restriction of fuel rods.
Ishikura, Shuichi*; Kogawa, Hiroyuki; Futakawa, Masatoshi; Kikuchi, Kenji; Haga, Katsuhiro; Kaminaga, Masanori; Hino, Ryutaro
JAERI-Tech 2003-093, 55 Pages, 2004/01
JAERI-Tech-2003-093.pdf:5.41MB
To estimate the structural integrity of the heavy liquid-metal (Hg) target used in a MW-class neutron scattering facility, static and dynamic stress behaviors due to the incident of a 1MW-pulsed proton beam were analyzed. In the analyses, two-type target containers with semi-cylindrical type and flat type window were used as analytical models of the structural analysis codes LS-DYNA. As a result, it is confirmed that the stress generated by dynamic thermal shock becomes the largest at the center of window, and the flat type window is more advantageous from the structural viewpoint than the semi-cylindrical type window. It was confirmed to erosion damage the target container by mercury's becoming negative pressure in the window and generating the cavitation by the experiment. Therefore, it has been understood that the point top of the window was in the compression stress field by the steady state thermal stress because of the evaluation from destroying the dynamic viewpoint for the crack in the generated pit and the pit point, and the crack did not progress.
Sato, Kazuyoshi; Ezato, Koichiro; Taniguchi, Masaki; Suzuki, Satoshi*; Akiba, Masato
Proceedings of the 7th International Conference on Creep and Fatigue at Elevated Temperatures (CREEP7), p.55 - 58, 2001/06
no abstracts in English
Suzuki, Satoshi; Ezato, Koichiro; Sato, Kazuyoshi; Nakamura, Kazuyuki; Akiba, Masato
Fusion Engineering and Design, 49-50, p.343 - 348, 2000/11
Times Cited Count：5 Percentile：37.66(Nuclear Science & Technology)no abstracts in English
Akiba, Masato; Ezato, Koichiro; Sato, Kazuyoshi; Suzuki, Satoshi; Hatano, Toshihisa
Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering (SOFE '99), p.381 - 384, 1999/10
no abstracts in English
Tanaka, Shigeru; R.Matera*; G.Kalinin*; V.Barabash*; Mori, Kensuke*
Journal of Nuclear Materials, 271-272, p.478 - 485, 1999/00
Times Cited Count：4 Percentile：34.81(Materials Science, Multidisciplinary)no abstracts in English
; Sato, Kazuyoshi; *; ; Dairaku, Masayuki; Nakamura, Kazuyuki; Akiba, Masato
Phys. Scr., T81, p.89 - 93, 1999/00
Times Cited Count：4 Percentile：39.16(Physics, Multidisciplinary)no abstracts in English
; *; Nakamura, Kazuyuki; Akiba, Masato
Proceedings of 17th IEEE/NPSS Symposium Fusion Engineering (SOFE'97), 1, p.385 - 388, 1998/00
no abstracts in English
; Sato, Kazuyoshi; Araki, Masanori; Nakamura, Kazuyuki; Dairaku, Masayuki; ; Akiba, Masato
Fusion Technology, 30(3(PT.2A)), p.793 - 797, 1996/12
no abstracts in English
Suzuki, Satoshi; Akiba, Masato; Araki, Masanori; Sato, Kazuyoshi; Yokoyama, Kenji; Dairaku, Masayuki
Journal of Nuclear Materials, 212-215(1), p.1365 - 1369, 1994/09
no abstracts in English
; Araki, Masanori; Akiba, Masato
JAERI-M 93-049, 26 Pages, 1993/03
no abstracts in English
