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Journal Articles

Characterization and re-activation of oxygen sensors for use in liquid lead-bismuth

Kurata, Yuji; Abe, Yuji*; Futakawa, Masatoshi; Oigawa, Hiroyuki

Journal of Nuclear Materials, 398(1-3), p.165 - 171, 2010/03

 Times Cited Count:17 Percentile:71.00(Materials Science, Multidisciplinary)

Liquid lead-bismuth is a candidate material used for accelerator driven systems and fast reactors. Oxygen control in liquid lead-bismuth is essential in this system and oxygen sensors are the critical instruments for the active oxygen control system. In this study, a re-activation treatment was investigated as well as characterization of oxygen sensors for use in liquid lead-bismuth. The oxygen sensor with a solid electrolyte of yttria-stabilized zirconia (YSZ) and a Pt/gas reference electrode showed almost the same electromotive force values as the theoretical ones at temperatures above 400 or 450 $$^{circ}$$C. After use for longer times than 6000 h, the outputs of the sensor became incorrect in liquid lead-bismuth. The state of the sensor showing incorrect outputs was not recovered by cleaning with nitric acid. However, it was found that the oxygen sensor was able to become a correct sensor indicating theoretical values after a re-activation treatment of the surface of the YSZ.

Journal Articles

Study on low activation decoupler material for MW-class spallation neutron sources

Harada, Masahide; Teshigawara, Makoto; Maekawa, Fujio; Futakawa, Masatoshi

Journal of Nuclear Materials, 398(1-3), p.93 - 99, 2010/03

 Times Cited Count:8 Percentile:48.34(Materials Science, Multidisciplinary)

In JSNS at J-PARC, by means of the Ag-In-Cd (AIC) alloy as the decoupler materials, the high decoupling energy at 1 eV for two decoupled moderators was realized for the first time in MW-class spallation neutron sources. Although the AIC decoupler is superior in the neutronic performance, it has a demerit in high residual radioactivity. To overcome the difficulty, we studied on possibilities of low activation decoupler materials with high decoupling energy. We considered two candidate materials, Au-In-Cd alloy and Boron Carbide based material. Neutronic performance of these materials was investigated by using neutronics calculations. As a result, it was found that both materials could provide pulse characteristics as well as those for the AIC decoupler. As a comparison of the candidate decouplers with the AIC deocupler on viewpoint of not only neutronic performance but also material property and irradiation characteristic, we decide the substitute of the AIC decoupler.

Journal Articles

Oxidation behaviour of P122 and a 9Cr-2W ODS steel at 550$$^{circ}$$C in oxygen-containing flowing lead-bismuth eutectic

Schroer, C.*; Konys, J.*; Furukawa, Tomohiro; Aoto, Kazumi

Journal of Nuclear Materials, 398(1-3), p.109 - 115, 2010/03

 Times Cited Count:58 Percentile:95.51(Materials Science, Multidisciplinary)

The long-term performance of P122 and ODS steel was tested in oxygen-containing flowing lead-bismuth eutectic at 550 $$^{circ}$$C and a flow velocity of 2 m/s by means of exposure experiments in the CORRIDA loop of FZK. The target for the enrichment of dissolved oxygen was a concentration, c$$_{o}$$, of 10$$^{-6}$$mass%. Owing to initial problems in controlling c$$_{o}$$, some of the exposed specimens experienced varying conditions, which allowed for investigating the influence of temporarily low c$$_{o}$$ (down to 10$$^{-9}$$mass%), in addition to constantly high c$$_{o}$$ (averaging 1.6$$times$$10$$^{-6}$$mass%). Maximum exposure times at constantly high c$$_{o}$$ were 10,000 and 20,000 h for P122 and the ODS steel, respectively. Under the testing conditions, both steels were affected by moderate oxidation resulting in the formation of a compact layer of a spinel-type oxide on the surface and a more or less pronounced internal oxidation zone. The thickness of the oxide scale was quantified and used for determining the material loss. For c$$_{o}$$ = 10$$^{-6}$$mass%, logarithmic and power rate laws were fitted to the obtained data as a basis for predicting the metal recession at times exceeding the experimental exposure times.

Journal Articles

Mechanical properties and microstructural stability of 11Cr-ferritic/martensitic steel cladding under irradiation

Yano, Yasuhide; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Shibayama, Tamaki*; Watanabe, Seiichi*; Takahashi, Heishichiro

Journal of Nuclear Materials, 398(1-3), p.59 - 63, 2010/03

 Times Cited Count:12 Percentile:57.83(Materials Science, Multidisciplinary)

The in-reactor creep rupture tests of 11Cr-0.5Mo-2W, V, Nb F/M steel were carried out in the temperature range from 823 to 943 K using Materials Open Test Assembly in the Fast Flux Test Facility and tensile and temperature-transient-to-burst specimens were irradiated in the experimental fast reactor JOYO at temperatures between 693 to 1013 K to fast neutron doses ranging from 3.5 to 102 dpa. The results of post irradiation mechanical tests showed that there was no significant degradation in tensile and transient burst strengths even after neutron irradiation below 873 K, but that there was significant degradation in both strengths at neutron irradiation above 903 K. On the other hand, the in-reactor creep rupture times were equal or greater than those of out-reactor creep even after neutron irradiation at all temperatures. This creep rupture behavior was different from that of tensile and transient burst specimens.

Journal Articles

Effect of tensile stress on cavitation damage formation in mercury

Naoe, Takashi; Kogawa, Hiroyuki; Yamaguchi, Yoshihito; Futakawa, Masatoshi

Journal of Nuclear Materials, 398(1-3), p.199 - 206, 2010/03

 Times Cited Count:5 Percentile:34.76(Materials Science, Multidisciplinary)

Pressure wave induced cavitation is a critical issue to realize a MW-class mercury target because structural integrity of the target vessel is remarkably degraded by the pitting damage. The target vessel suffers tensile stress by welding residual stress and/or thermal stress due to proton beam injection. In this study, in order to examine the effect of tensile stress on pitting damage formation, cavitation erosion test was performed using an ultrasonic homogenizer in mercury. The result showed that the damaged area was increased with increasing in tensile stress. The depth and diameter of pits were larger than that of no-stressed specimen, and the eroded area was increased. Indentation tests under tensile stress were carried out to quasi-statically simulate impact load. Vickers hardness was slightly decreased. Threshold stress of the deformation, i.e. pitting damage formation, was decreased by tensile stress.

Journal Articles

Investigation of beam window buckling with consideration of irradiation effects for conceptual ADS design

Sugawara, Takanori; Kikuchi, Kenji; Nishihara, Kenji; Oigawa, Hiroyuki

Journal of Nuclear Materials, 398(1-3), p.246 - 250, 2010/03

 Times Cited Count:3 Percentile:23.39(Materials Science, Multidisciplinary)

The investigation of the beam window which is a key component in the conceptual design of Accelerator Driven System (ADS) has been performed. In the past study, it was found that the buckling failure was the most severe failure mode for the beam window from the simplified assessment. Detail structural analyses were explored to find a solution for avoiding instantaneous buckling. The results showed that the ellipse shape concepts with the thickness of 2.0-2.4[mm] at the top and the thickness of 2.0-4.0[mm] at the transient part were acceptable under the current ADS design parameters. These investigations, however, did not consider the irradiation effect by neutrons and protons. In this study, investigations based on the latest knowledge for the irradiation effect such as the data obtained in SINQ target irradiation program (STIP) are presented.

Journal Articles

Corrosion resistance of Al-alloying high Cr-ODS steels in stagnant lead-bismuth

Takaya, Shigeru; Furukawa, Tomohiro; Inoue, Masaki; Fujisawa, Toshiharu*; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Kimura, Akihiko*

Journal of Nuclear Materials, 398(1-3), p.132 - 138, 2010/03

 Times Cited Count:64 Percentile:96.44(Materials Science, Multidisciplinary)

Oxide dispersion strengthened (ODS) ferritic steels with excellent high-temperature strength are the candidates for fuel cladding tubes. But, the compatibility with lead bismuth eutectic (LBE) is one of the key issues in accelerator driven system and LBE cooled fast reactors. Addition of Al and increase in Cr may have beneficial influence on the compatibility. Addition of Al, however, causes a decrease in high-temperature strength. A significantly higher Cr concentration results in aging embrittlement. Therefore, we need to find their optimal amount to balance corrosion resistance with high-temperature strength. In this study, the cross sections of the samples after 3,000 h of exposure to LBE with 10$$^{-8}$$ wt% oxygen at 650 $$^{circ}$$C are examined in detail using scanning electron microscope and Auger electron spectroscopy. The observation shows that very thin Al oxide layer is formed continuously between multiple oxide layer/internal oxide zone and matrix, and that such Al oxide layer suppresses further growth of multiple oxide layer/internal oxide zone. The average oxide layer thickness shows a tendency to get thinner by increasing in Al content from about 2 to 4 wt%, although significant dependency on Cr content is not recognized. Furthermore, the additional corrosion test for 5,000 h is conducted. These materials show good corrosion resistance even after 5,000 h of exposure to LBE containing 10$$^{-6}$$ wt% at 650 $$^{circ}$$C. Addition of 3.5 wt% Al is very effective in improving corrosion resistance.

Oral presentation

Microstructural evolution on Ti-modified austenitic stainless steel irradiated by high energy proton

Hamaguchi, Dai; Kikuchi, Kenji; Saito, Shigeru; Endo, Shinya; Yong, D.*

no journal, , 

The microstructural evolutions on austenitic stainless steel JPCA irradiated in SINQ target was investigated. JPCA is Ti-modified type-316 base austenitic stainless steel for reduced swelling, and is one of the candidate materials for the structural material on Japanese ADS. Specimens were irradiated in STIP-I and STIP-II with 580MeV proton. For the sample irradiated to 5.7dpa at 150$$^{circ}$$C on STIP-I, only high density of small loops and black dot defects were observed. The density of the loops did not increase significantly with doses and temperatures, which was around 7$$times$$10$$^{22}$$m$$^{-3}$$, but the size increased from around 15nm at 5.7dpa to 25nm at 19.5dpa on STIP-II irradiated sample. On the other hand, formation of high density small He bubbles with a size of about 2 to 3nm and a density around 1 to 4$$times$$10$$^{23}$$m$$^{-3}$$ were observed for the samples irradiated more than 7.9dpa at above 200$$^{circ}$$C. The densities of the bubbles did not significantly change with doses but the sizes slightly decreased. The helium concentration for the sample with the dose of 7.9dpa was about 600appm, and for the highest dose sample with 19.5dpa was about 1600appm. The formation of bubbles on JPCA was observed at lower temperature compared to EC316LN on STIP-I irradiated samples. This might due to the Ti modification, since Ti in steel is known to be an over-size element and prefer to combine with vacancies, which leads to shorten the incubation period for cavity formation.

Oral presentation

Present status of J-PARC project

Oyama, Yukio

no journal, , 

Oral presentation

Proton irradiation effects on tennsile and bend-fatigue properties of welded F82H specimens

Saito, Shigeru; Kikuchi, Kenji; Hamaguchi, Dai; Usami, Koji; Ishikawa, Akiyoshi; Nishino, Yasuharu; Endo, Shinya; Kawai, Masayoshi*; Yong, D.*

no journal, , 

In several institutes, research and development for an accelerator-driven transmutation system have been progressed. Ferritic/martensitic steels are the candidate material for the beam window. To obtain the irradiation data, the PIE work of the SINQ target irradiation program (STIP) specimens was carried out at JAEA. In this study, the results of PIE on F82H and its welded joint will be reported. The tensile tests were performed for F82H EB and TIG welded specimens. The results indicate that all specimens kept its ductility after 10 dpa irradiation and fractured in ductile manner. The fatigue life of F82H base metal is almost the same as that of unirradiated specimens. Though the number of specimen is limited, the fatigue life of F82H EB (15mm and 3.3mm) welded joints seems to increase after irradiation. The fracture surfaces of the specimens showed transgranular morphology. While F82H TIG welded specimens were not fractured by 10$$^{7}$$ cycles.

Oral presentation

Lesson on LBE control techniques taken from JLBL1 loop

Kikuchi, Kenji; Saito, Shigeru; Hamaguchi, Dai; Tezuka, Masao

no journal, , 

JLBL-1 operation has been done over 18000 hrs to get design data for LBE spallation target in J-PARC facility. Interested materials are austenitic steel because electro-magnetic pump needed non-magnetized materials. The loop was made mainly by SS316 in order to know mass traveling in the closed system. The temperature difference was 50-100 $$^{circ}$$C and flow rate was 1 m/s at the test section. Active control of oxygen concentration in the system was not done but inert gas was used to separate LBE from oxygen environment. Oxygen sensor was set up in the loop later. We experienced blockade of pass in the pump, mass transfer from hot parts to cold parts, precipitations of dissolved components from materials. Lead bismuth was sampled from the loop and replaced by new one. Electro-magnetic flow meter was inspected to take output change into consideration. Most important thing experienced hitherto was a find of erosion-corrosion in the narrow path at the high temperature test section.

Oral presentation

Status of liquid lithium target activities in IFMIF-EVEDA

Nakamura, Hiroo; Agostini, P.*; Groeschel, F.*

no journal, , 

In this presentation, status of the liquid lithium (Li) target activities will be presented. In International Fusion Materials Irradiation Facility (IFMIF), intense neutron flux is produced by D-Li stripping reaction. The liquid Li target system consists of a target assembly, a main loop and a purification loop. Design requirement is to provide a stable Li jet at a maximum speed of 20 m/s. A cold trap and two kinds of hot trap are applied to control impurities below 10 wppm for nitrogen and 1 wppm for tritium. To maintain reliable continuous operation, various diagnostics are attached. The backplate made of RAFMs is located in the most severe region of neutron irradiation (60 dpa/y). Life time of the backplate will be evaluated considering DBTT etc. For replacement of the backplate by remote handling, two design options of replaceable backplate are investigated. In the EVEDA, Li test loop is planned and will start till early 2011.

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