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Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04
Irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steel has been studied as main concern of an aging problem of light water reactor (LWR) materials. It is essential to evaluate corrosion behavior of irradiated materials for mechanistic understanding of IASCC. The aim of this work is to evaluate the corrosion behavior of ion irradiated materials using atomic force microscope (AFM), and evaluate the influence of radiation temperature, radiation damage, H and He implantation.
Onizawa, Kunio; Shibata, Katsuyuki; Kato, Daisuke*; Li, Y.*
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04
Probabilistic Fracture Mechanics (PFM) has been used in the fields of reliability analysis for important structural components. At JAERI, the PFM analysis code PASCAL has been developed. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). Four cases of PTS transients were selected based on the severity for a typical 3-loop PWR. Based on thermal stress analyses, PFM analyses were performed by using PASCAL code focusing on some important variables on the RPV fracture probability. The results showed that non-destructive examination methods had a significant effect on the fracture probability by more than three orders of magnitude. The comparisons of the results using fracture toughness estimation methods between in Japan and USA, and crack geometries between a semi-elliptical surface crack and an infinite surface crack are also made.
Shibamoto, Yasuteru; Iguchi, Tadashi; Nakamura, Hideo; Kukita, Yutaka*
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 11 Pages, 2003/04
The pressure effect on the onset of flow instability in a vertical upflow through a boiling channel is studied both analytically and experimentally. The analytical model is based on the Wallis-Heasley model for linear analysis of one-dimensional homogeneous two-phase flow in thermal equilibrium. The dead-time elements commonly used to represent the time lag in the responses of variables to the inlet velocity perturbation is replaced by first-order lag elements to allow the system characteristic equation to be solved analytically. This approach, although approximate, makes it much easier to identify the main contributor to the instability because the individual components are represented by separate terms in the characteristic equation. The predictions are in reasonable agreement with the data when the system pressure effect on the irreversible pressure loss in the two-phase region is appropriately considered based on calibration experiments.
Kikuchi, Kenji; Saito, Shigeru; Kurata, Yuji; Futakawa, Masatoshi; Sasa, Toshinobu; Oigawa, Hiroyuki; Umeno, Makoto*; Mori, Keijiro*; Takano, Hideki; Wakai, Eiichi
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04
In order to construct ADS Target Test facility in J-PARC project the research and development on Pb-Bi technology have been carried, which cover target design with computer simulation, flowing loop test, stagnant corrosion test, oxygen sensor and cleaning techniques. Obtained results are as follows: Corrosion rate of SUS316 under flowing Pb-Bi at 1m/s at 450
C is 0.1 mm / 3000 hrs. Fe and Cr were melted into lead bismuth from SS316 in the high temperature part and deposited in the low-temperature part according to the difference of solubility. The corrosion thickness decreases with increasing Cr content in the stagnant corrosion test at saturated oxygen concentration. Reliable oxygen sensors are to be developed by using suitable reference electrodes. As a result of cleaning tests, blushing process was needed to remove Pb-Bi effectively after immersion in the silicon oil. The mixed acid easily dissolved Pb-Bi and removed almost perfectly. But specimens themselves were affected by coloring.
Iguchi, Tadashi; Shibamoto, Yasuteru; Asaka, Hideaki; Nakamura, Hideo
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04
Authors investigated the cooling limit under flow instability, by conducting THYNC experiments using a 2
2 bundle test section of electrical rod heaters、whose heated lengths and diameters were 3.71m and 12.3mm. The experimental result indicated periodic rise and rapid drop of the rod temperature under flow oscillation, indicating periodic film boiling. When the heating power increased further, the rod temperature indicated continuous film boiling. The power at the onset of continuous film boiling (cooling limit) under flow oscillation was about 50%-80% of the cooling limit under steady flow condition in THYNC. The ratio of both cooling limits almost agreed with the Umekawa model prediction in cases of P
2MPa and G400kg/m2s. For high pressure and high mass flux conditions, the ratio almost agreed with the empirical model based on the heat balance during one cycle of flow oscillation. TRAC-BF1 code simulated periodic film boiling qualitatively, but the cooling limit under the flow oscillation was not predicted well probably due to inaccurate rewetting prediction.
Ogawa, Toru; Minato, Kazuo; Sawa, Kazuhiro
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 6 Pages, 2003/04
no abstracts in English
Sasa, Toshinobu; Oigawa, Hiroyuki; Tsujimoto, Kazufumi; Nishihara, Kenji; Kikuchi, Kenji; Kurata, Yuji; Saito, Shigeru; Futakawa, Masatoshi; Umeno, Makoto*; Ouchi, Nobuo; et al.
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 9 Pages, 2003/04
JAERI carries out research and development on accelerator-driven system (ADS) to transmute minor actinides and long-lived fission products in high-level radioactive waste. The system is composed of high intensity proton accelerator, lead-bismuth spallation target and lead-bismuth cooled subcritical core with nitride fuel. About 2500 kg of minor actinide is loaded into the subcritical core. Annual transmutation amount using this system is 250 kg with 800MW of thermal output. A superconducting linear accelerator with the beam power of 20 - 30MW is connected to drive the subcritical core. The nitride fuel without uranium, such as (Np, Am, Pu)N, is selected. The fuel irradiated in the ADS is reprocessed by pyrochemical process followed by the re-fabrication of the fuel. Many research and development activities are under way. Especially, to study and evaluate the feasibility of the ADS from physics and engineering aspects, the Transmutation Experimental Facility (TEF) is proposed under a framework of the High-Intensity Proton Accelerator Project.
Kaminaga, Masanori; Haga, Katsuhiro; Kinoshita, Hidetaka; Terada, Atsuhiko*; Koikegami, Hajime*; Hino, Ryutaro
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04
1MW pulsed spallation neutron source using a mercury target will be constructed in order to produce high-intensity neutrons for use in the fields of life and material sciences. The mercury target is proposed using a high-energy (3.0GeV) proton beam with a current of 0.333mA and operating at 25 Hz with a pulse duration less than 1
s. A cross flow type (CFT) mercury target has been designed in order to distribute mercury flow according to an axial heat generation distribution caused by spallation reaction, based on the thermal hydraulic analytical results of 3GeV, 1MW proton beam injection by using the STAR-CD code. This paper presents the final CFD analytical results. In the analysis, an inlet temperature of 50C and an inlet mercury velocity of 1.0m/s were assumed. As results, a maximum velocity of 2.48m/s was observed near the front end of the outlet plenum and a maximum of 125.5C was observed near the beam window where the volumetric heat generation rate was relatively large. The maximum temperature is far below the mercury saturation temperature of 356C under atmospheric pressure. This result satisfied the thermal-hydraulic design criteria of "Maximum mercury temperature in the target shall be less than 300C".
Kogawa, Hiroyuki; Ishikura, Shuichi*; Futakawa, Masatoshi; Kaminaga, Masanori; Hino, Ryutaro
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04
The developments of a MW-class spallation neutron source facility are being carried out under the high-intensity proton accelerator project promoted by JAERI and KEK. A mercury target will be used as a neutron source in the facility. The mercury target vessel made of 316LSS will be subjected to pressure wave generated by rapid thermal expansion of mercury due to a pulsed proton beam injection. The pressure wave will make huge stress on the vessel and will deform the vessel, which would cause cavitation in mercury. To estimate the structural integrity of the mercury target vessel, especially beam window, dynamic stress behaviors due to 1MW-pulsed proton beam injection were analyzed by using FEM code. In the analyses, two types of the target vessels with semi-cylindrical and flat type windows were used as analytical models. As the results, it has been understood that the stress generated in the beam window by the pressure wave could be treated as the secondary stress. Also it was confirmed that the flat type window would be more advantageous from the structural viewpoint than the semi-cylindrical type window.
Haga, Katsuhiro; Kinoshita, Hidetaka; Kaminaga, Masanori; Hino, Ryutaro; Tagawa, Hisato*; Kukita, Yutaka*
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04
The cross-flow type mercury target, in which mercury flows crossing the proton beam path, has been developed as the spallation target of the material and life science facility in the high intensity proton accelerator project. As a part of design optimization, we proposed a mercury target using perforated plates aiming to simplify the inner structure, to make the target assembling easier, and to decrease the assembling cost. Then, the effectiveness of the target structure was investigated by no-heat water experiments and computational analyses. A mockup model of the cross-flow type mercury target using perforated plates was fabricated with plexi-glass and the water flow field was measured using PIV technique and the results were compared with the analytical results. The cross-flow field was realized by perforated plates and the analytical results corresponded well with the experimental results in the proton beam path where the cooling of heat generation is important.
Ide, Hiroshi; Matsui, Yoshinori; Nagao, Yoshiharu; Komori, Yoshihiro; Itabashi, Yukio; Tsuji, Hirokazu; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04
The advanced water chemistry controlled irradiation research device has been developed in JAERI to perform irradiation tests for research on IASCC. The irradiation device consists of the SATCAP (Saturated Temperature Capsule) inserted into the JMTR core and the water control unit installed out-of-core. Regarding the SATCAP, thermohydraulic design of the SATCAP was performed aiming at controlling the specimen temperature with high accuracy and increasing water flow velocity on the specimen surface to improve the controllability of water chemistry. As a result of irradiation test using the new type SATCAP, each specimen temperature and water chemistry were able to be controlled as designed.
Aso, Tomokazu; Kaminaga, Masanori; Hino, Ryutaro; Monde, Masanori*
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04
no abstracts in English
Yonomoto, Taisuke; Okubo, Tsutomu; Iwamura, Takamichi; Ishida, Toshihisa
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04
An innovative concept of the pressure suppression system having functions of water injection and non-condensable gas confinement is developed for the next generation light water reactors (LWRs). The use of the system is advantageous for the mitigation of effects of the loss-of-coolant accidents (LOCAs) in (1) keeping the containment pressure as low as for the conventional LWRs, (2) injecting water to the containment for cooling the reactor pressure vessel (RPV) and/or flooding a break, and (3) confining the non-condensable gas in the drywell. The gas confinement function makes the system considerably suitable for reactor designs with passive cooling systems utilizing heat exchangers, such as the steam generator (SG) secondary side cooling system for an integral reactor, and the passive containment cooling system (PCCS), because it avoids adverse effects of non-condensable gas on the heat transfer performance during LOCAs. The usefulness of the developed concept is confirmed in the RELAP5/MOD3 code calculation.
Iwanaga, Kohei; Yamane, Tsuyoshi; Nishihara, Kenji; Okajima, Shigeaki; Sekimoto, Hiroshi*; Asaoka, Takumi*
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 10 Pages, 2003/04
no abstracts in English
Ouchi, Nobuo; Akaoka, Nobuo*; Asano, Hiroyuki*; Chishiro, Etsuji; Hasegawa, Kazuo; Takeda, Osamu; Yoshikawa, Hiroshi; Matsuoka, Masanori*; Otani, Toshihiro*; Kako, Eiji*; et al.
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 6 Pages, 2003/04
no abstracts in English
Nakajima, Yasuo; Funasaka, Hideyuki; ; Nagaoki, Yoshihiro
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 0 Pages, 2003/00
Ito, Kenji
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 0 Pages, 2003/00
None
; ; Iijima, Takashi
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 0 Pages, 2003/00
None
Iguchi, Yukihiro; ; Kanehia, Yoshiki
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 0 Pages, 2003/00
None
Iguchi, Yukihiro; ; Kanehia, Yoshiki
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 0 Pages, 2003/00
None