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Journal Articles

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 1; Proposal of countermeasures to mitigate temperature fluctuations around control rods

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.

Journal Articles

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 2; Proposal of countermeasures to mitigate temperature fluctuations around radial blanket fuel assemblies

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.

Journal Articles

Evaluation of multiaxial low cycle creep-fatigue life for Mod.9Cr-1Mo steel under non-proportional loading

Nakayama, Yuta*; Ogawa, Fumio*; Hiyoshi, Noritake*; Hashidate, Ryuta; Wakai, Takashi; Ito, Takamoto*

ISIJ International, 61(8), p.2299 - 2304, 2021/08

 Times Cited Count:0 Percentile:0(Metallurgy & Metallurgical Engineering)

This study discusses the creep-fatigue strength for Mod.9Cr-1Mo steel at a high temperature under multiaxial loading. A low-cycle fatigue tests in various strain waveforms were performed with a hollow cylindrical specimen. The low cycle fatigue test was conducted under a proportional loading with a fixed axial strain and a non-proportional loading with a 90-degree phase difference between axial and shear strains. The low cycle fatigue tests at different strain rates and the creep-fatigue tests at different holding times were also conducted to discuss the effects of stress relaxation and strain holding on the failure life. In this study, two types of multiaxial creep-fatigue life evaluation methods were proposed: the first method is to calculate the strain range using Manson's universal slope method with considering a non-proportional loading factor and creep damage; the second method is to calculate the fatigue damage by considering the non-proportional loading factor using the linear damage law and to calculate the creep damage from the improved ductility exhaustion law. The accuracy of the evaluation methods is much better than that of the methods used in the evaluation of actual machines such as time fraction rule.

Journal Articles

In situ diffraction characterization on microstructure evolution in austenitic stainless steel during cyclic plastic deformation and its relation to the mechanical response

Kumagai, Masayoshi*; Akita, Koichi*; Kuroda, Masatoshi*; Harjo, S.

Materials Science & Engineering A, 820, p.141582_1 - 141582_9, 2021/07

 Times Cited Count:0 Percentile:0(Nanoscience & Nanotechnology)

Journal Articles

Change in mechanical properties by high-cycle loading up to Gigacycle for 316L stainless steel

Naoe, Takashi; Harjo, S.; Kawasaki, Takuro; Xiong, Z.*; Futakawa, Masatoshi

JPS Conference Proceedings (Internet), 28, p.061009_1 - 061009_6, 2020/02

At the J-PARC, a mercury target vessel made of 316L SS suffers proton and neutron radiation environment. The target vessel also suffers cyclic impact stress caused by the proton beam-induced pressure waves. The vessel suffers higher than 4.5$$times$$10$$^8$$ cyclic loading during the expected service life of 5000 h. We have investigated fatigue strength 316L SS up to gigacycle in the previous studies. The cyclic hardening and softening behavior were observed. In this study, to evaluate the cyclic hardening/softening behavior, the dislocation densities of specimens were measured using the neutron diffraction method at the MLF BL-19. The result showed that the dislocation density of a 316L SS was increased with increasing the number of loading cycles. By contrast, in the case of cold-rolled 316L SS, annihilation and re-accumulation of dislocation by cyclic loading were observed. In the workshop, result of neutron diffraction measurement will be introduced with the progress of fatigue test.

Journal Articles

Failure behavior analyses of piping system under dynamic seismic loading

Udagawa, Makoto; Li, Y.; Nishida, Akemi; Nakamura, Izumi*

International Journal of Pressure Vessels and Piping, 167, p.2 - 10, 2018/11

 Times Cited Count:4 Percentile:50.81(Engineering, Multidisciplinary)

It is important to assure the structural Integrity of piping systems under severe earthquakes because those systems comprise the pressure boundary for coolant with high pressure and temperature. In this study, we examine the seismic safety capacity of piping systems under severe dynamic seismic loading using a series of dynamic-elastic-plastic analyses focusing on dynamic excitation experiments of 3D piping systems which was tested by NIED. Analytical results were consistent with experimental data in terms of natural frequency, natural vibration mode, response accelerations, elbow opening-closing displacements, strain histories, failure position, and low-cycle fatigue failure lives. Based on these results, we concluded that the analytical model used in the study can be applied to failure behavior evaluation for piping systems under severe dynamic seismic loading.

Journal Articles

Thermal fatigue test on dissimilar welded joint between Gr.91 and 304SS

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.

Journal Articles

Remaining fatigue lives of similar surface flaws in accordance with combination rules

Lu, K.; Li, Y.; Hasegawa, Kunio*; Lacroix, V.*

Journal of Pressure Vessel Technology, 139(2), p.021407_1 - 021407_6, 2017/04

 Times Cited Count:1 Percentile:10.91(Engineering, Mechanical)

Journal Articles

Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor; Influence of flow collector of backup CR guide tube

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.

Journal Articles

In situ X-ray diffraction study of the oxide formed on alloy 600 in borated and lithiated high-temperature water

Watanabe, Masashi*; Yonezawa, Toshio*; Shobu, Takahisa; Shiro, Ayumi; Shoji, Tetsuo*

Corrosion, 72(9), p.1155 - 1169, 2016/09

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

Journal Articles

Fatigue lives of multiple flaws in accordance with combination rule

Lu, K.; Li, Y.; Hasegawa, Kunio; Lacroix, V.*

Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 7 Pages, 2016/07

Journal Articles

Creep-fatigue tests of double-end notched bar made of Mod.9Cr-1Mo steel

Shimomura, Kenta; Kato, Shoichi; Wakai, Takashi; Ando, Masanori; Hirose, Yuichi*; Sato, Kenichiro*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

This paper describes experimental and analytical works to confirm that the design standard for SFR components sufficiently covers possible failure mechanisms. Creep-fatigue damage evaluation method in JSME design standard for SFR components has been constructed based on experiments and/or numerical analyses of conventional austenitic stainless steels, such as 304SS. Since the material characteristics of Mod.9Cr-1Mo steel are substantially different from those of austenitic stainless steels, it is required to verify the applicability of the design standards to the SFR components made of Mod.9Cr-1Mo steel. A series of uni-axial creep-fatigue tests were conducted using double-ended notch bar specimens made of Mod.9Cr-1Mo steel under displacement controlled condition with 30 minute holding. The curvature radii of the specimens were 1.6mm, 11.2mm and 40.0mm. The specimen having 1.6mm notch and 11.2mm notch failed from outer surface but the specimen having 40.0mm notch showed obvious internal crack nucleation. In addition, though total duration time of the creep-fatigue test was only 2,000 hours, a lot of creep voids and inter granular crack growth were observed. To clarify the cause of such peculiar failure, some additional experiments were performed, as well as some numerical analyses. We could point out that such a peculiar failure aspect might result from corresponding stress distribution in the cross section. As a result of a series of investigations, possible causes of such peculiar failure could be narrowed down. A future investigation plan was proposed to clarify the most significant cause.

Journal Articles

Water experiments on thermal striping in reactor vessel of Japan Sodium-cooled Fast Reactor; Countermeasures for significant temperature fluctuation generation

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$$^{circ}$$ 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.

Journal Articles

Key achievements in elementary R&D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

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:7.57(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 $$^{circ}$$C followed by normalizing at 930 $$^{circ}$$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$$_{2}$$TiO$$_{3}$$, 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.

Journal Articles

Fatigue assessment of the ITER TF coil case based on JJ1 fatigue tests

Hamada, Kazuya; Nakajima, Hideo; Takano, Katsutoshi*; Kudo, Yusuke; Tsutsumi, Fumiaki*; Okuno, Kiyoshi; Jong, C.*

Fusion Engineering and Design, 75-79, p.87 - 91, 2005/11

 Times Cited Count:8 Percentile:51.68(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Mechanical characteristics evaluation of fuel cladding tube for reduced-moderation water reactor, 1 (Contract research)

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$$_{2}$$ 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.

Journal Articles

ITER relevant high heat flux testing on plasma facing surfaces

Hirai, Takeshi*; Ezato, Koichiro; Majerus, P.*

Materials Transactions, 46(3), p.412 - 424, 2005/03

 Times Cited Count:101 Percentile:90.2(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Tensile and fatigue strength of free-standing CVD diamond

Davies, A. R.*; Field, J. E.*; Takahashi, Koji; Hada, Kazuhiko

Diamond and Related Materials, 14(1), p.6 - 10, 2005/01

 Times Cited Count:19 Percentile:62.24(Materials Science, Multidisciplinary)

A CVD diamond is finding increased application and it is important to study its fatigue properties. The present paper describes research on a batch of di-electric grade CVD material. It was obtained that tensile strength at the nucleation side and the growth were side 690$$pm$$90MPa and 280$$pm$$30MPa, respectively. Some samples survived at least 95% of their critical fracture stress for 10$$^{7}$$ cycles without fatiguing.

Journal Articles

Elastic-plastic FEM analysis on low cycle fatigue behavior for alumina dispersion-strengthened copper/stainless steel joint

Nishi, Hiroshi

Journal of Nuclear Materials, 329-333(Part2), p.1567 - 1570, 2004/08

 Times Cited Count:9 Percentile:54.8(Materials Science, Multidisciplinary)

Elastic-plastic finite element analysis was performed for low cycle fatigue behavior of stainless steel/alumina-dispersion-strengthened copper (DS Cu) joint in order to investigate the fatigue life and the fracture behavior of the joint. As the results, a strain concentration was occurred near the interface of DS Cu for small strain range, however, in the DS Cu for large strain range. The fatigue life and fracture point were evaluated taking account for the strain concentration. The fatigue life and fracture point were consistent with those of the low cycle fatigue test.

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