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Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
The steam generator (SG) is an important component of a pressurized water reactor. In addition, local wall-thinning has been reported in SG tubes. The burst differential pressure, considering both the internal and external pressures from the primary and secondary coolant systems, should be predicted for the failure probability evaluation or structural integrity assessment of SG tubes. In this study, based on the results of burst tests performed in Japan and the United States, we improved the existing burst pressure estimation method for SG tubes with wall-thinning. In addition, as an example of the utilization of the improved burst pressure estimation method, the conditional failure probabilities for SG tubes with local wall-thinning, which is necessary for probabilistic risk assessment and risk-informed decision making, are calculated considering the dimensions of the wall-thinning.
Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 10 Pages, 2022/07
The steam generator (SG) tube is one of the important components in pressurized water reactors. Flaws such as wall-thinning or stress corrosion cracking have been reported in SG tubes. The burst pressure where both the internal and external pressures from the primary and secondary coolant systems are considered must be predicted to assess the structural integrity of SG tubes. Burst tests were performed by various organizations. On the basis of the test results, failure estimation methods were proposed. In this study, previous burst test data and existing failure estimation methods for SG tubes with wall-thinning or crack were investigated. As a result, the coefficient of the existing estimation method for SG tube with uniform wall-thinning was updated. In addition, failure estimation methods that are suitable for SG tubes with crack or local wall-thinning were proposed by considering the effects of the flaw shape and size on the burst pressure. The applicability of the failure estimation methods was confirmed by comparing the predicted results with the burst test data in actual SG tubes.
Kato, Chiaki
Comprehensive Nuclear Materials, 2nd Edition, Vol.4, p.528 - 563, 2020/08
In spent fuel reprocessing plants, various nitric media are encountered throughout the PUREX process, used in the separation of fission products, uranium, and plutonium. The PUREX process is thus highly corrosive as it takes place at high temperatures under high concentrations of nitric acid solution containing oxidizing metal ions from spent fuel. In this review, the unique chemical properties of nitric acid are first described. Secondly, the process of oxidizing power generation in boiling nitric acid under heat transfer is described using the redox potential and a thermodynamic model of boiling nitric acid. Finally, the corrosion behavior and corrosion acceleration mechanism specific to the reprocessing environments are described from the perspective of solution chemistry.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00353_1 - 19-00353_6, 2020/03
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Wada, Yuki; Le, T. D.; Satou, Akira; Shibamoto, Yasuteru; Yonomoto, Taisuke
Journal of Nuclear Science and Technology, 57(1), p.100 - 113, 2020/01
Times Cited Count：7 Percentile：61.18(Nuclear Science & Technology)Takada, Shoji; Ngarayana, I. W.*; Nakatsuru, Yukihiro*; Terada, Atsuhiko; Murakami, Kenta*; Sawa, Kazuhiko*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 13 Pages, 2019/05
In the loss of core cooling test using HTTR, a technical issue is to improve prediction accuracy of temperature distribution of components in vessel cooling system (VCS). An establishment of reasonable 2D model was started by using numerical code FLUENT, which was validated using the test data by 1/6 scale model of VCS for HTTR. The pressure vessel (PV) temperature was set around 200
C attributed to relatively high ratio of natural convection heat transfer around 20% in total heat removal, which is useful for code to experiment benchmark to improve prediction accuracy. It is necessary to confirm heat transfer flow characteristics around the top of PV which is heated up by natural convection flow which was considered to be affected by separation, re-adhesion and transition flow. The k-
-SST model was selected for turbulent calculation attributed to predict the effects mentioned above adequately. The numerical results using the k--SST model reproduced the temperature distribution of PV especially the top region which is considered to be affected by separation, re-adhesion and transition flow in contract to that using k-
model which does not account the effects.
Uesawa, Shinichiro; Koizumi, Yasuo; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Dai-54-Kai Nihon Dennetsu Shimpojiumu Koen Rombunshu (CD-ROM), 8 Pages, 2017/05
no abstracts in English
Uesawa, Shinichiro; Koizumi, Yasuo; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Thermal Science and Engineering, 25(2), p.17 - 26, 2017/04
no abstracts in English
Umeda, Ryota; Kurihara, Akikazu; Shimoyama, Kazuhito
JAEA-Technology 2016-030, 50 Pages, 2016/12
JAEA-Technology-2016-030.pdf:5.22MB
In case of tube failure of a steam generator in sodium-cooled fast reactors, the reaction jet with high temperature and high velocity under highly alkaline environment is formed by cited exothermic reaction (sodium-water reaction). When the high temperature reaction jet covers the adjacent tubes, the material strength of tube decreases in the high temperature condition, and the adjacent tube may be swollen and failed by inner pressure (overheating tube rupture). For evaluation of the overheating tube rupture, tube failure is judged by comparison the hoop stress loaded by inner pressure with stress strength standard defined as creep strength depending on tube temperature. Thus, it is important to confirm the validation of this failure criterion based on the findings obtained in the simulated experiment of overheating tube rupture. In this report, for consideration on the validation of the failure criteria and elucidation on the failure mode and strength characteristics of failure, the authors carried out the rapid heating rupture experiment for the thin single and double-walled 9Cr steel tubes at high temperature up to 1500 K by using TRUST-2 rig in the Japan Atomic Energy Agency.
Tochio, Daisuke; Nakagawa, Shigeaki
JAERI-Tech 2005-041, 109 Pages, 2005/08
JAERI-Tech-2005-041.pdf:4.48MB
In High Temperature Engineering Test Reactor (HTTR) of 30MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary pressurized water cooler and the intermediate heat exchanger. To remove generated heat at reactor core and to hold reactor inlet coolant temperature to specified temperature, heat exchangers in main cooling system of HTTR should have designed heat exchange performance. In this report, heat exchange performance for ACL in main cooling system is evaluated with previous operation data, and evaluated values are compared with designed value. Moreover, heat exchange performance at full power operation is estimated for the air temperature. As the result, ACL has heat exchange performance removing generated heat at reactor core under the designe ACL inlet air temperature of 33C.
Sugiyama, Tomoyuki; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 41(11), p.1083 - 1090, 2004/11
Times Cited Count：10 Percentile：55.63(Nuclear Science & Technology)The effect of cladding surface pre-oxidation on the rod coolability under reactivity initiated accidents was investigated. NSRR tests on irradiated fuel rods have shown higher rod coolability than that of fresh rods, which arose from suppressed DNB and early quench at the surface. To identify the dominant factor, possible factors such as pellet cracking and so on, were assessed. The most probable factor, the cladding pre-oxidation, was examined by pulse irradiation tests on fresh rods with three cladding surface conditions, no oxide layer, 1
m and 10m-thick oxide layers. Temperature measurements showed increased thresholds for DNB and quench at the pre-oxidized surface, leading to a reduced film boiling duration. The shifts of the critical and minimum heat flux points could be caused by the surface wettability increase. In the present tests, the wettability change was probably dominated by the chemical potential change at the surface due to pre-oxidation. The test results indicate the effects do not depend on the oxide layer thickness, but on the presence of the oxide layer.
Takada, Shoji
Nuclear Engineering and Design, 233(1-3), p.185 - 195, 2004/10
Times Cited Count：6 Percentile：40.63(Nuclear Science & Technology)Experiments are carried out to investigate the effects of the natural convection of superheated gas as well as of the stand pipes on the temperature distributions of the components and the heat removal performance in the water-cooling panel system for the MHTGR for decay heat removal, and to verify the design and evaluation methods. The numerical results of the code THANPACST2 are compared with the experimental data to verify the numerical methods and axi-symmetric model proposed, which can simulate the three-dimensional configuration of the stand pipes on the upper head of the pressure vessel by using porous body cells. The experiments revealed that temperatures increased with elevation on the upper head, because the stand pipes restrict radiation heat transfer to the upper cooling panel and reduce the heat transfer area on the upper head which was superheated by natural convection of helium gas in the pressure vessel. The numerical methods were able to closely duplicate the pattern of the rising temperature profile with elevation around the top of the upper head.
Kato, Chiaki; Yano, Masaya*; Kiuchi, Kiyoshi; Sugimoto, Katsuhisa*
Corrosion Engineering, 52(1), p.53 - 67, 2003/01
The effects of heat-transfer on the corrosion of zirconium was examined in boiling nitric acid solutions with various concentrations. Corrosion mass losses and electrochemical polarization curves were measured on the heat-transfer and isothermal surfaces in the solutions. It was found that the corrosion rate of zirconium was higher on the heat-transfer surface than that on the isothermal surface. The rate increased with increasing nitric acid concentration and solution temperature. The increased oxidization potential on the heat-transfer surface is attributed to the reduction of nitrous acid concentration by the thermal decomposition on the surface and the removal of the decomposition product from solution by boiling bubbles. The redox potential of 12 mol/dm
nitric acid on a boiling heat-transfer surface was very close to the breakdown potential of primary passivity of zirconium. This suggests the initiation of SCC on a boiling heat-transfer surface in a nuclear fuel reprocessing.
Kato, Chiaki; Yano, Masaya*; Kiuchi, Kiyoshi; Sugimoto, Katsuhisa*
Zairyo To Kankyo, 52(1), p.35 - 43, 2003/01
The effects of heat-transfer on the corrosion of zirconium was examined in boiling nitric acid solutions with various concentrations. Corrosion mass losses and electrochemical polarization curves were measured on the heat-transfer and isothermal surfaces in the solutions. It was found that the corrosion rate of zirconium was higher on the heat-transfer surface than that on the isothermal surface. The rate increased with increasing nitric acid concentration and solution temperature. The increased oxidization potential on the heat-transfer surface is attributed to the reduction of nitrous acid concentration by the thermal decomposition on the surface and the removal of the decomposition product from solution by boiling bubbles. The redox potential of 12 mol/dm3 nitric acid on a boiling heat-transfer surface was very close to the breakdown potential of primary passivity of zirconium. This suggests the initiation of SCC on a boiling heat-transfer surface in a nuclear fuel reprocessing.
Uemura, Takuya*; Takeda, Tetsuaki; Ichimiya, Koichi*
Nihon Kikai Gakkai Yamanashi Koenkai Koen Rombunshu (020-4), p.49 - 50, 2002/10
A heat transfer experiment was performed using a horizontal circular tube to obtain the heat transfer and fluid flow characteristics in the tube inserted copper wire. From the results obtained in this experiment, it was found that an amount of heat removal in the tube with copper wire inserts increased about 20% comparing with a tube having a smooth wall. A heat transfer coefficient of the tube inserted copper wire also increased 30% to 50% under the constant pumping power condition.
Takeda, Tetsuaki
Nihon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.199 - 202, 2002/00
There are several methods for enhancement of heat transfer such as attaching various type fins on a channel surface of a heat exchanger tube and a steam reforming tube. The objective of this study is to clarify characteristics of heat transfer and pressure drop in the channel inserted metallic wire with high porosity. A heat transfer experiment was performed using a horizontal circular tube to obtain the heat transfer characteristics of the channel inserted copper wire. From the results obtained in this experiment, it seems to be probable that an enhancement of heat transfer method using metallic wire inserts is further improved under the high temperature conditions in efficiency as compared with the other methods.
Kinoshita, Hidetaka; Terada, Atsuhiko*; Kaminaga, Masanori; Hino, Ryutaro
JAERI-Tech 2001-061, 43 Pages, 2001/10
JAERI-Tech-2001-061.pdf:5.21MB
no abstracts in English
Oba, Toshihiro; Yanagihara, Takao; Kato, Chiaki; Hamada, Shozo
JAERI-Tech 2001-059, 36 Pages, 2001/09
The demonstration test for evaluating reliability of the acid recovery evaporator at Rokkasho Reprocessing Plant has been carried out at JAERI. For the nondestructive mesurement of the thickness of heat transfer tubes in the acid recovery evaporator and short tubes used in corrosion test, we have developed ultra sonic thickness measuring apparatus using immersion method with high resolution. This apparatus can measure and record tube thickness automatically with a personal computer. The results obtained by this apparatus are coincident with the results obtained by a destructive method using an optical microscope.
Saito, Kazuo*; Ishida, Toshihisa
JAERI-Tech 2001-039, 25 Pages, 2001/06
JAERI-Tech-2001-039.pdf:0.94MB
no abstracts in English
Ishiyama, Shintaro; Muto, Yasushi; Tanihira, Masanori*; Ogata, Kan*; Kamito, Yoshimi*; Igari, Toshihide*
Nihon Genshiryoku Gakkai-Shi, 43(6), p.603 - 611, 2001/06
Times Cited Count：4 Percentile：33.39(Nuclear Science & Technology)no abstracts in English
