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JAEA Reports

Report of the erosion-corrosion of metallic materials under solid-liquid two phase flow

Otani, Kyohei; Sato, Tomonori; Kaji, Yoshiyuki; Yamamoto, Masahiro

JAEA-Review 2019-007, 15 Pages, 2019/06


Metallic pipes under solid-liquid two phase flow is damaged by collision of solid particle to the pipe walls, and this phenomenon is named "erosion". In the case of the liquid is corrosive solution, further damage is occurred on the pipe walls chemically, and this named "erosion-corrosion". In the Fukushima Daiichi decommissioning project, the fuel debris will be crushed during removal operation of the debris and micro debris particles would be generated. It is estimated that the pipes of the circulating cooling system would be damaged under the solid-liquid two phase flow containing fuel debris particles. For the reason, the previous study about erosion and erosion-corrosion of metallic materials under solid-liquid two phase flow was surveyed. The survey showed that the damage rate by erosion and erosion-corrosion is influence by a lot of parameter in comparison to the corrosion rate which occurred in no-flow solution. Therefore, it is necessary to pay attention to selecting the experimental method and condition before the investigation about erosion-corrosion of metallic materials under solid-liquid two phase flow is carried out.

Journal Articles

Hardening effect on impact erosion in interface between liquid and solid metals

Futakawa, Masatoshi; Naoe, Takashi*; Kogawa, Hiroyuki; Ishikura, Shuichi*; Date, Hidefumi*

Zairyo, 53(3), p.283 - 288, 2004/03

no abstracts in English

Journal Articles

Design of mercury cirulation system for J-SNS

Kinoshita, Hidetaka; Haga, Katsuhiro; Kogawa, Hiroyuki; Kaminaga, Masanori; Hino, Ryutaro

Proceedings of ICANS-XVI, Volume 3, p.1305 - 1314, 2003/07

The JAERI and the KEK are promoting a plan to construct the spallation neutron source at the Tokai Research Establishment, JAERI, under J-PARC project. A mercury circulation system has been designed so as to supply mercury to the target stably. Conceptual design is almost finished. But, it was necessary to confirm a mercury pump performance, and more, to investigate erosion rate under the mercury flow as well as an amount of mercury remained on the surface after drain. The mercury pump performance was tested under the mercury flow conditions by using an experimental gear pump, which had almost the same structure as a practical mercury pump to be expected, and the erosion rates in a mercury pipeline as were investigated. The discharged flow rates of the gear pump increased linearly with the rotation speed. Erosion rates obtained under the mercury velocity less than 1.6 m/s was found to be so small. For the amount of remaining mercury on the pipeline, radioactivity of this remaining mercury volume was found to be three-order lower than that of the target casing.

Journal Articles

Corrosion-erosion test of SS316 in flowing Pb-Bi

Kikuchi, Kenji; Kurata, Yuji; Saito, Shigeru; Futakawa, Masatoshi; Sasa, Toshinobu; Oigawa, Hiroyuki; Wakai, Eiichi; Miura, Kuniaki*

Journal of Nuclear Materials, 318(1-3), p.348 - 354, 2003/05

 Times Cited Count:25 Percentile:15.15(Materials Science, Multidisciplinary)

Corrosion test of austenitic stainless tube was done under the flowing Pb-Bi condition during 3000 hrs at 450$$^{circ}$$C. Specimen is 316SS produced as a tubing form with 13.8 mm outer diameter, 2 mm thickness and 40 cm length. During the operation, maximum temperature, temperature difference and flow velocity of Pb-Bi at the specimen were kept at 450$$^{circ}$$C, 50$$^{circ}$$C, and 1m/s, respectively. After the test, specimen and components of the loop were cut and examined by optical microscope, SEM, EDX, WDX and X-ray diffraction. Pb-Bi adhered on the surface of the specimen even after Pb-Bi was drained out to the storage tank from the circulating loop. Different results from a stagnant corrosion test were that the specimen surface became rough and the corrosion rate was maximally 0.1mm/3000hrs. And mass transfer from the high temperature to the lower temperature area was observed: crystals of Fe-Cr were found on the tube surface in low-temperature part. The size of crystal was 0.1 $$sim$$ 0.2 mm. The depositing crystal was ferrite grain and the chemical composition ratio (mass%) of Fe to Cr was 9:1.

Journal Articles

Corrosion-erosion of SS316 under flowing Pb-Bi

Kikuchi, Kenji; Kurata, Yuji; Saito, Shigeru; Futakawa, Masatoshi; Sasa, Toshinobu; Oigawa, Hiroyuki; Miura, Kuniaki*

Nippon Kikai Gakkai 2002-Nendo Nenji Taikai Koen Rombunshu, p.273 - 274, 2002/09

Corrosion-erosion properties of type 316 austenitic stainless steel were investigated at 450oC. The study aims at developing ADS, accelerator driven system, for nuclear transmutation of long lived activated nuclei to shorter ones. After 3000 hrs flow of eutectic 45Pb-55Bi loop tubes were cut and analyzed by optical microscope, SEM, EDX, WDX and TEM. It is concluded that corrosion-erosion depth is maximally 0.1mm per 3000 hrs and Cr-Fe crystals were precipitated in the lower temperature of flowing channel. Further more inspection results of Electro Magnetic Pump, Electro Magnetic Flow meter and controlling valve were also reported. Output signal from EMF was stable after certain time duration. In this experiment oxygen content in Pb-Bi was not actively controlled but Pb-Bi was covered by 4N Argon gas.

JAEA Reports

Mercury flow experiments, 4; Measurements of erosion rate caused by mercury flow

Kinoshita, Hidetaka; Kaminaga, Masanori; Haga, Katsuhiro; Hino, Ryutaro

JAERI-Tech 2002-052, 28 Pages, 2002/06


Since the Neutron Scattering Facility will be using mercury as the target material and contain radioactive products, it is necessary to estimate reliability of instruments in a system. The system would be damaged by erosion. An erosion test section and coupons were installed in the mercury loop, and their thickness was measured. As a result, the erosion is about 3$$mu$$m in 1000 hours under 0.7m/s condition. The wall thickness decrease during facility lifetime of 30 years is estimated to be less than 0.5mm. Therfore, the effect of erosion on component strength is extremely small. Moreover, a measurement of residual mercury on the piping surface was carried out. As a result, 19g/m$$^{2}$$ was obtained. Thus, estimation of residual mercury for 150A-sch80 piping is 8.5g/m, and for the mercury target is about 40g. As for the target, radioactivity of the residual mercury is 1.2$$times$$10$$^{12}$$ Bq, which is extremely lower than that in the target casing of 1.0$$times$$10$$^{15}$$ Bq. Then, there is no influence for maintenance and storage of the spent mercury target.

JAEA Reports


; ; *; Yamaguchi, Akira

JNC-TN9400 2000-109, 96 Pages, 2000/11


Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiaied based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1)Ther is no remarkable difference betweeen liquid sodium and luquid Pb-Bi in characteristics of internal flows and free surface charatristics based on Fr number. (2)the AQUA-VOF code has a potentiall enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [thermal Stratification Phenomena] (1)On-set position of thermal entainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. 0n the other hand, the position in the case of C0$$_{2}$$ gas was shifted to upstream side with decreasing of Ri number. (2)Destruction speed of the thermal stratyification interface was dependent on thermal diffusivity as fluid properties. therefor it was concluded that an elimination method is necessary for the interface generated in C0$$_{2}$$ gas. [thermal Striping Phenomena] (1)Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO$$_{2}$$ gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2)To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it isnecessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phenomena] (1)Fundamental behavior of the natural convection in various coolant follows buoyant jet ....

JAEA Reports

Measurements of thermal properties of buffer materials; Measurement of physical properties of buffer materials and improvement of measuring method


JNC-TJ8400 2000-017, 74 Pages, 2000/02


The report concerns the improvement of the method measuring thermal conductivity of buffer materials using a thermistor probe and the measurement of thermal conductivity of compacted bentonites and mixtures of bentonite and silica sand using the proposed method measuring thermophysical properties. The method measuring thermal conductivity is improved in accuracy and the apparatus is improved so as to measure easily with more short time. The calculated values of the conventional correlations predicting thermal conductivity of bentonite and mixture were compared with the exising and present data of thermal conductivity of bentonites and mixtures. The correlation proposed by Sakashita and Kumada can predict the best fitted values with the data of the bentonites and Fricke and Bruggeman correlations are fitted with the data for the mixtures with practical accuracy.

JAEA Reports

Formation and evaluation of functionally gradient material for thermal stress relaxation, 1

; Hirakawa, Yasushi; Kano, Shigeki; Yoshida, Eiichi

PNC-TN9410 98-048, 56 Pages, 1998/03


Planar specimens of functionally gradient material (FGM) for thermal stress relaxation in fast reactor environment were formed and evaluated. FGMs of Al$$_{2}$$O$$_{3}$$-SUS316L system and Y$$_{2}$$O$$_{3}$$-SUS316L system were deposited on SUS316L substrates by low pressure plasma spraying. The deposited coatings with 6 layers in which the ratio of ceramics/SUS316FR changes from 0 to 100% by 20% were successfully formed. Cross-sectional observation of the coatings showed no cracks and the hardness in the coating increased continuously from the substrate to the surface. From the results of X-ray diffraction, there were no changes in the structure of SUS316L and Y$$_{2}$$O$$_{3}$$ between the powder and the coating. On the contrary, in the case of Al$$_{2}$$O$$_{3}$$, $$gamma$$ - Al$$_{2}$$O$$_{3}$$ phase was detected in the coating formed from $$alpha - Al$$_${2}$$$O$$_${3}$$ powder. The specimens were exposed in liquid sodium at 823K or 923K for 3.6Ms(1000h). The coatings were damaged with many cracks in liquid sodium. It was revealed that the bonding strength between the sprayed particles were not sufficient. To improve the stability in liquid sodium, another specimens were formed with changing the chamber pressure during deposition. From the microstructural inspections of the specimens, the coating formed at higher chamber pressure showed less porosity.

JAEA Reports

A study on incubation time for erosion by high temperature sodium

Wakai, Takashi; Aoto, Kazumi

PNC-TN9420 97-003, 11 Pages, 1996/12


Generally, it is required several periods for the damage initiation due to erosion. Such periods are called incubation time. This report describes the calculation of incubation time for erosion of a carbon steel subjected to dripping liquid sodium at elevated temperature, assuming fatigue plays the significant role in the early stage of erosion. It is clalified that the incubation time is remarkably long and it is needless to consider the damage due to erosion if there is no material property degradation. However, taking the material property degradation of carbon steel and liquid sodium into account, the incubation time may be short less than 1 hour in the most pessimistic case. So in this case, erosion must be considered as one of the damage mechanisms of carbon steel.

Journal Articles

Tritium retention in graphite inner wall of JT-60U

Masaki, Kei; ; Ando, Toshiro; Saido, Masahiro; Shimizu, Masatsugu; Hayashi, Takumi; Okuno, Kenji

Fusion Engineering and Design, 31, p.181 - 187, 1996/00

 Times Cited Count:18 Percentile:18.21(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Investigation of plasma facing components in JT-60U operation

*; Ando, Toshiro; ; Arai, Takashi; Neyatani, Yuzuru; Yoshino, Ryuji; Tsuji, Shunji; Yagyu, Junichi; Kaminaga, Atsushi; ; et al.

Journal of Nuclear Materials, 220-222, p.390 - 394, 1995/00

 Times Cited Count:16 Percentile:17.95(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Development and experiences with JT-60U plasma facing components

Ando, Toshiro

NIFS-PROC-12, p.75 - 78, 1993/03

no abstracts in English

Journal Articles

Sputtering characteristics of B$$_{4}$$C-overlaid graphite for keV energy deuterium ion irradiation

Goto, Yoshitaka*; Yamaki, Takahiro*; Ando, Toshiro; Jimbo, Ryutaro*; Ogiwara, Norio; Saido, Masahiro; Teruyama, Kazuhiro*

Journal of Nuclear Materials, 196-198, p.708 - 712, 1992/12

 Times Cited Count:19 Percentile:15.43(Materials Science, Multidisciplinary)

no abstracts in English

Oral presentation

Study on long-term integrity assessment for engineered barriers; Overview of testing plan in under-ground research laboratory

Nakayama, Mariko*; Kobayashi, Masato*; Kawakubo, Masahiro*; Suzuki, Kei*; Eto, Jiro*; Nakayama, Masashi; Ono, Hirokazu; Asano, Hidekazu*

no journal, , 

no abstracts in English

Oral presentation

Research for the erosion of buffer material with the vertical disposal concept, 3; In situ tests on engineering scale at Horonobe URL

Ono, Makoto*; Motoshima, Takayuki*; Shirase, Mitsuyasu*; Yokoyama, Satoshi*; Jo, Mayumi*; Ishii, Tomoko*; Nakayama, Masashi; Ono, Hirokazu

no journal, , 

To understand phenomenon of buffer material erosion in the vertical disposal concept, in situ tests on engineering scale have been carried out at Horonobe URL. As results of water injection with flow rate control, when water flowed in a continuous large flow rate, buffer materials did not swell enough to stop the water flow and the water flow path was confirmed to be formed by aggregated into a single pipe. Further, it was suggested that water pressure of the injection side contributes to the erosion of buffer materials.

Oral presentation

Evaluation of target-wastage in consideration of sodium-water reaction environment formed on the periphery of an adjacent tube in steam generator of sodium-cooled fast reactor, 3

Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Kikuchi, Shin

no journal, , 

Wastage phenomena on adjacent tubes (target-wastage) arise from water/steam leak in steam generators of sodium-cooled fast reactors. Target-wastage is likely to be caused by liquid droplet impingement erosion (LDI) and Na-Fe composite oxidation type corrosion with flow (COCF) in an environment marked by high temperature and high-alkali (reaction jet) due to sodium-water reaction. The authors derived new wastage correlations from COCF and LDI data based on influencing factors which were formed on the periphery of an adjacent tube. In this report, the authors established that the new wastage correlations were applicable to each tube of tube bundle in sodium-water reaction test with straight vertical tube bundle under practical steam generator operation condition, and the time progress of wastage was qualitatively investigated for the two penetrated tubes in the period including the blowdown.

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