Onizawa, Takashi; Hashidate, Ryuta
Mechanical Engineering Journal (Internet), 6(1), p.18-00477_1 - 18-00477_15, 2019/02
Aiming at enhancing its economic competitiveness and reducing radioactive waste, JAEA has proposed an attractive plant concept and made great efforts to demonstrate the applicability of some innovative technologies to the plant. One of the most practical means is to extend the design life to 60 years. Accordingly, the material strength standards set by JSME have to be extended from 300,000 to 500,000 hours but this extension requires more precise estimation of creep rupture strength and creep strain of the materials in the long term. This paper describes the development of creep property equations of 316FR stainless steel and Mod.9Cr-1Mo steel considering changes in creep mechanisms at high temperatures in the long term based on evaluations of long-term creep properties of the materials. The creep property equations developed in this study will provide more precise estimation of the creep properties in the long term than the present creep property equations of JSME.
Suwa, Tomone*; Hemmi, Tsutomu*; Saito, Toru*; Takahashi, Yoshikazu*; Koizumi, Norikiyo*; Luzin, V.*; Suzuki, Hiroshi; Harjo, S.
IEEE Transactions on Applied Superconductivity, 28(3), p.6001104_1 - 6001104_4, 2018/04
Harjo, S.; Kawasaki, Takuro; Hemmi, Tsutomu; Ito, Takayoshi*; Nakamoto, Tatsushi*; Aizawa, Kazuya
JPS Conference Proceedings (Internet), 8, p.031001_1 - 031001_5, 2015/09
Kuwabara, Kazumichi; Sato, Toshinori; Sanada, Hiroyuki; Takayama, Yusuke
JAEA-Research 2015-005, 378 Pages, 2015/07
This report presents the results of following rock mechanical investigations conducted at the -500m Stage. (1) Laboratory tests using cores and block samples obtained at the -500m Stage. (2) In-situ stress measurement using Compact Conical-ended Borehole Overcoring (CCBO) method at the -500m Stage. (3) In-situ stress measurements using Differential Strain Curve Analysis(DSCA) method at the -500m Stage. (4) Development of rock mechanical model.
Shinozaki, Takashi; Mihara, Takeshi; Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
JAEA-Research 2014-025, 34 Pages, 2014/12
EDC test is a test method on the mechanical property of fuel cladding tube, and it focuses on the stress condition generated by PCMI under a RIA. We conducted EDC tests which simulate the mechanical conditions during a RIA by using the unirradiated cladding tubes which simulate hydride rim. Circumferential residual strains observed in post-test specimens tended to decrease with increasing the hydrogen concentration in the test cladding tubes and the thickness of the hydride rim. We also prepared RAG tube and performed EDC tests on it. It was observed that circumferential total strains at failure tended to decrease with increasing pre-crack depth on the outer surface of RAG tube specimen. We conducted biaxial stress tests by applying longitudinal tensile load onto RAG tube specimens. It was observed that circumferential total strains at failure under biaxial stress conditions tended to decrease compared to the results under uniaxial tensile condition.
Hirohashi, Masayuki*; Murakami, Haruyuki*; Ishiyama, Atsushi*; Ueda, Hiroshi*; Koizumi, Norikiyo; Okuno, Kiyoshi
IEEE Transactions on Applied Superconductivity, 16(2), p.1721 - 1724, 2006/06
To demonstrate the applicability of NbSn CICCs to ITER, four NbSn model coils have been constructed and tested. The experimental results showed that the measured critical current (Ic) degraded. In addition, the larger is the applied electromagnetic force, the larger the magnitude of the degradation is. The degradation in n-value was also observed. One of the explanations of this degradation is a local strand bending. This consideration has been supported by the test results. However, general dependence of Ic on periodic bending strain has not been clarified in this test since the experiments were carried out at a certain magnetic field, temperature and strain. Therefore, a numerical simulation code was developed to study the general dependence of the Ic and n-value of the NbSn strand on periodic bending strain. A distributed constant circuit model is applied to simulate current transfer among the filaments in the strand. The simulation results show relatively good agreement with the experiment results but some modification in modeling is required for more accurate simulation.
Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi
Nuclear Engineering and Design, 236(2), p.128 - 139, 2006/01
A computer code RANNS was developed to analyze fuel rod behaviors in the RIA conditions. The code performs thermal and FEM-mechanical calculation for a single rod in axis-symmetric geometry to predict temperature profile, PCMI contact pressure, stress-strain distribution and their interactions. An experimental analysis by RANNS begins with pre-test conditions of irradiated rod which are given by FEMAXI-6. Analysis was performed on the simulated RIA experiments in NSRR, FK-10 and FK-12, of high burnup BWR rods in a cold start-up conditions, and PCMI process was discussed extensively. It was revealed that pellet thermal expansion dominates cladding deformation and subjects the cladding to bi-axial stress state, and thermal expansion in the cladding makes the stress in the inner region significantly lower than that in the outer region. Simulation calculations with wider pulses were carried out and the resulted cladding hoop stress was compared with the failure stress estimated in the NSRR experiments.
Ando, Toshinari*; Kizu, Kaname; Miura, Yushi*; Tsuchiya, Katsuhiko; Matsukawa, Makoto; Tamai, Hiroshi; Ishida, Shinichi; Koizumi, Norikiyo; Okuno, Kiyoshi
Fusion Engineering and Design, 75-79, p.99 - 103, 2005/11
no abstracts in English
Hanawa, Satoshi; Sumita, Junya; Shibata, Taiju; Ishihara, Masahiro; Iyoku, Tatsuo; Sawa, Kazuhiro
Transactions of 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT-18), p.600 - 605, 2005/08
no abstracts in English
Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 42(2), p.209 - 218, 2005/02
Regarding high burn-up fuel behavior under LOCA conditions, LOCA-simulated experiments were performed with unirradiated Zircaloy-4 claddings. Claddings containig 100 to 1450 ppm were isothermally oxidized at at 1220 to 1500 K in steam flow, and quenched by flooding water. Axial shrinkage of the rods during the quench was restrained controlling the maximum restraint load at four different levels. Primarily depending on fraction of the cladding thickness oxidized, the claddings fractured into two pieces during the quench, with circumferential cracking. The fracture/non-fracture threshold as for the oxidized fraction decreases as both initial hydrogen concentration and axial restraint load increase. Consequently, it was shown that the threshold is higher than 20% cladding oxidation, e.g. sufficiently higher than the limit in the Japanese ECCS acceptance criteria, irrespective of hydrogen concentration, when the restraint load is below 535 N.
Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 42(1), p.58 - 65, 2005/01
Tube burst tests have been performed with artificially hydrided Zircaloy-4 specimens at room temperature and at 620 K. Pressurization rate was increased to a maximum of 3.4 GPa/s in order to simulate rapid PCMI that occurs in high burnup fuel rods during a pulse-irradiation in the NSRR. Hydrogen content in the specimens ranged from 150 to 1050 ppm. Hydrides were accumulated in the cladding periphery and formed "hydride rim" as observed in high burnup PWR fuel claddings. The hydrided cladding tubes failed with an axial crack at the room temperature tests. Brittle fracture appeared in the hydride rim, and failure morphology was similar to that observed in the NSRR experiments. The hydrides rim obviously reduced burst pressure and residual hoop strain at the tests. The residual hoop strain was very small even at 620 K when thickness of the hydride rim exceeded 18% of cladding thickness. The present result accordingly indicates an important role of the hydrides layer in high burnup fuel rod failure under RIA conditions.
Taguchi, Tomitsugu; Jitsukawa, Shiro; Sato, Michitaka*; Matsukawa, Shingo*; Wakai, Eiichi; Shiba, Kiyoyuki
Journal of Nuclear Materials, 335(3), p.457 - 461, 2004/12
F82H (Fe-8Cr-2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load-displacement curves and the strain distributions obtained from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400C.
Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 41(12), p.1211 - 1217, 2004/12
Hydride precipitation along the radial-axial plane increases in high burn-up BWR fuel claddings. The radial hydrides may have an important role during fuel behavior in a RIA and may reduce ductility of the cladding under PCMI conditions. In order to promote a better understanding of the influence of the radial hydrides on cladding failure behavior under the PCMI conditions, tube burst tests were conducted for unirradiated BWR claddings charged with 200 to 650 ppm of hydrogen. About 20 to 30% of hydrides were re-oriented and precipitated along the radial-axial plane. The claddings exhibited large burst openings with an axial crack at room temperature and 373 K. However, the influence of the radial hydrides on both burst pressure and residual hoop strain was very small. It is accordingly expected that ductility of high burn-up BWR cladding is significantly reduced not only by precipitation of radial hydrides as far as hydrogen concentration and radial hydride fraction range in the present study.
Miwa, Yukio; Tsukada, Takashi
Proceedings of 8th Japan-China Symposium on Materials for Advanced Energy Systems and Fission & Fusion Engineering, p.161 - 168, 2004/10
Irradiation assisted stress corrosion cracking (IASCC) is one of the environmental degradation problems of in-core structural materials for light water reactors. The effects of irradiation and water temperatures on the IASCC were studied using type 316(LN) stainless steels irradiated at 333-673 K to 1.1-16 dpa. IASCC did not occur at 513 K in oxygenated water for specimens irradiated below 573 K to 1.1-16 dpa, but IASCC occurred above 533 K in oxygenated water for all specimens. The irradiation temperature had a strong influence on IASCC susceptibility at 513 K in oxygenated water, so that the irradiation temperature dependence was compared with the temperature dependence of other radiation-induced phenomena.
Journal of Nuclear Materials, 329-333(Part2), p.1567 - 1570, 2004/08
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.
Nakano, Junichi; Miwa, Yukio; Koya, Toshio; Tsukada, Takashi
Journal of Nuclear Materials, 329-333(Part1), p.643 - 647, 2004/08
To study effects of minor elements on the irradiation assisted stress corrosion cracking (IASCC), high purity Type 304 and 316 stainless steels (SSs) were fabricated and added minor elements, Si or C. After neutron irradiation to 3.510n/m (E1MeV), the slow strain rate tests (SSRT) for the irradiated specimens was conducted in oxygeneted high purity water at 561 K. Fracture surface of the specimens was examined using the scanning electron microscope (SEM) after the SSRT. Fraction of intergranular stress corrosion cracking (IGSCC) on the fracture surface after the SSRT increased with netron fluence. Suppression of irradiation hardening and increase of peiod to SCC fracture as benefitical effects of the additional elements, Si or Mo, were not observed obviously. In high purity SS added C, fraction of IGSCC was the smallest in the all SSs, although irraidiation hardening level was the largest in the all SSs. Addition of C suppressed the susceptibility to IGSCC.
Shibata, Taiju; Ishihara, Masahiro; Motohashi, Yoshinobu*; Ito, Tsutomu*; Baba, Shinichi; Kikuchi, Makoto*
Materials Transactions, 45(8), p.2580 - 2583, 2004/08
Fast neutrons (energy 1.610 J) were irradiated to tetragonal zirconia polycrystals containing 3 mol% yttria (3Y-TZP) at the fluence levels of 2.510 (Light irradiation) and 4.310 (Heavy irradiation) m. The irradiation caused no significant swelling in the 3Y-TZP specimens. After the neutron irradiation, superplastic characteristics were examined by tensile tests at a temperature range from 1623 to 1773 K with initial strain rates ranging from 5.010 to 1.6710s. It was found that the elongation to fracture of the irradiated specimens was quite small in comparison with the unirradiated ones. The apparent activation energy for the superplastic flow of the irradiated 3Y-TZP was fairly high, i.e., 781 and 693 kJ・mol for Light and Heavy irradiations, respectively. Atomic displacement damages and defects in the 3Y-TZP caused by the irradiation were thought to be main causes of these property changes.
Ishikura, Shuichi*; Kogawa, Hiroyuki; Futakawa, Masatoshi; Kaminaga, Masanori; Hino, Ryutaro; Saito, Masakatsu*
Nippon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.59 - 66, 2004/03
The development of a MW-class spallation neutron source facility is being carried out under the Japan Proton Accelerator Research Complex (J-PARC) Project promoted by JAERI and KEK. A mercury target working as the spallation neutron source will be subjected to pressure waves generated by rapid thermal expansion of mercury due to a pulsed proton beam injection. The pressure wave will impose dynamic stress on the vessel and deform the vessel, which would cause cavitation in mercury. To evaluate the effect of mercury micro jets, driven by cavitation bubble collapse, on the micro-pit formation, analyses on mercury sphere collision were carried out: single bubble dynamics and collision behavior on interface between liquid and solid, which take the nonlinearity due to shock wave in mercury and the strain rate dependency of yield stress in solid metal into account. Analytical results give a good explanation to understand relationship between the micro-pit formation and material properties: the pit size could decrease with increasing the yield strength of materials.
Nakano, Junichi; Tsukada, Takashi; Tsuji, Hirokazu; Terakado, Shogo; Koya, Toshio; Endo, Shinya
JAERI-Tech 2003-092, 54 Pages, 2004/01
Irradiation assisted stress corrosion cracking (IASCC) is a degradation phenomenon caused by synergy of neutron radiation, aqueous environment and stress on in-core materials, and it is an important issue in accordance with increase of aged light water reactors. Isolating crack initiation stage from crack growth stage is very useful for the evaluation of the IASCC behavior. Hence facility for in-situ observation during slow strain rate test (SSRT) for irradiated material was developed. As performance demonstrations of the facility, tensile test with in-situ observation and SSRT without observation were carried out using unirradiated type 304 stainless steel in 561 K water at 9 MPa. The following were confirmed from the results. (1) Handling, observation and recording of specimen can be operated using manipulators in the hot cell. (2) In-situ observation can be performed in pressurized high temperature water and flat sheet type specimen is suitable for the in-situ observation. (3) Test condition can be kept constantly and data can be obtained automatically for long test period.
Kikuchi, Kenji; Saito, Shigeru; Nishino, Yasuharu; Usami, Koji
Proceedings of 6th International Meeting on Nuclear Applications of Accelerator Technology (AccApp '03), p.874 - 880, 2004/00
Specimens irradiated at SINQ were tested by tensile and fatigue. Speciemns were irradiated by 580MeV proton beams under spallation reaction during two years, transported to JAERI and tested at JAERI Hot Cell. Material is JPCA austenitic stainless steel. Strain-to-necking is over 8% at 250C test temperature and are different from APT handbook database. Fatigue test was conducted at low stress regime of high cycle fatigue. The number of cycles to failure is reduced by factors five to ten. These data will help a design of spallation target in JPARC.