Sako, Hiroyuki; Ahn, J. K.*; Baek, K. H.*; Bassalleck, B.*; Fujioka, H.*; Guo, L.*; Hasegawa, Shoichi; Hicks, K.*; Honda, R.*; Hwang, S. H.*; et al.
Journal of Instrumentation (Internet), 9(4), p.C04009_1 - C04009_10, 2014/04
A TPC has been developed for J-PARC E42 experiment to search for H-dibaryon in (, ) reaction. An event with 2 and 2 protons decaying from H-dibaryon is searched for inside the TPC. The TPC has octagonal prism shape drift volume with about 50 cm diameter with 55 cm drift length filled with Ar-CH (90:10) gas. At the end of the drift volume, 3-layer GEMs are equipped. In order to analyze momenta of produced particles, the TPC is applied with 1 T dipole magnetic field parallel to the drift electric field with a superconducting Helmholz magnet. In order to maximize the acceptance of H-dibaryon events, a diamond target is installed inside the TPC drift volume, in a cylindrical hole opened from the top to the middle of the drift volume. Since extremely high-rate beam is directly injected into the TPC drift volume to the target, a gating grid and GEMs are adopted to suppress positive-ion feedback.
Sugimura, Hitoshi; Imai, Kenichi; Sako, Hiroyuki; Sato, Susumu; Kiuchi, Ryuta; Ichikawa, Yudai; Hwang, S. H.*; Hasegawa, Shoichi; Tanida, Kiyoshi; J-PARC E10 Collaboration*
Physics Letters B, 729, p.39 - 44, 2014/02
We have carried out an experiment to search for a neutron-rich hypernucleus, H, by the Li(,K) reaction at p = 1.2 GeV/c. The obtained missing-mass spectrum with an estimated energy resolution of 3.2 MeV (FWHM) showed no peak structure corresponding to the H hypernucleus neither below nor above the H+2n particle decay threshold. An upper limit of the production cross section for the bound H hypernucleus was estimated to be 1.2 nb/sr at 90% confidence level.
Ioka, Ikuo; Kiuchi, Kiyoshi*; Takizawa, Masayuki*; Ito, Takeshi*
Nihon Kinzoku Gakkai-Shi, 78(1), p.16 - 22, 2014/01
A multi-axis stress field is indispensable to quick and quantitative evaluation of stress corrosion cracking for constructional materials and weld joints of existing industrial plants. The applicability of multi-axis residual stress field into SCC tests was evaluated. The hard sphere ball was stuffed into small flat-plate of type 304SS. Numerical analysis was conducted in order to compare with the experimental results. The numerical analysis was comparatively in agreement with the experimental results. Parameters of the test were selected by numerical analysis to optimize the residual stress of specimen. SCC test in MgCl was performed using the specimen with optical residual stress condition. It is confirmed that the multi-axis residual stress field was useful in quick and quantitative SCC test by comparing the initiation of cracks with the distribution of residual stress obtained by numerical analysis.
Ioka, Ikuo; Suzuki, Jun; Kiuchi, Kiyoshi; Nakayama, Jumpei*
Proceedings of 2nd International Workshop on Structural Materials for Innovative Nuclear Systems (SMINS-2), p.391 - 400, 2012/12
Optimization of composition for high Cr-W-Si Ni base alloy has been studied to apply to a nitric acid with high oxidation-reduction potential of advanced reprocessing plants. The corrosion resistance of the Ni base alloy is superior to that of conventional stainless steels. In addition, The Ni base alloy has an excellent resistance of weld crack and ability of plastic deformation caused by extra high purity (EHP) refining technology. However, the Ni base alloy has a technical limitation in hot working and welding for practical use. Several Ni base EHP alloys with different content of Si and W were manufactured to choose an optimum composition range without losing corrosion resistance. High strain rate tensile tests at high temperature, corrosion tests and weldability tests were carried out to examine the optimum composition range of Ni base EHP alloy.
Kim, G.; Shiba, Kiyoyuki; Sawai, Tomotsugu; Ioka, Ikuo; Kiuchi, Kiyoshi; Nakayama, Jumpei*
Proceedings of 2nd International Workshop on Structural Materials for Innovative Nuclear Systems (SMINS-2), p.273 - 279, 2012/12
Sugimura, Hitoshi; Imai, Kenichi; Sako, Hiroyuki; Sato, Susumu; Adachi, Satoshi*; Tanida, Kiyoshi*; Kiuchi, Ryuta*; Joo, C. W.*
AIP Conference Proceedings 1388, p.602 - 604, 2011/10
no abstracts in English
Ioka, Ikuo; Ishijima, Yasuhiro; Usami, Koji; Sakuraba, Naotoshi; Kato, Yoshiaki; Kiuchi, Kiyoshi
Journal of Nuclear Materials, 417(1-3), p.887 - 891, 2011/10
Fe-25Cr-35Ni EHP alloy was developed with conducting the countermeasure for IASCC. It is composed to adjust major elements, to remove harmful impurities and so on. The specimens were irradiated at 553 K for 25000h using JRR-3. The fluence was estimated to be 1.510n/m. Type 304SS was also irradiated as a comparison material. SSRT test was conducted in oxygenated water at 561 K in 7.7 MPa. The fracture mode of EHP alloy was ductile. IGSCC was not observed in the fracture surface. On the other hand, the fraction of IGSCC on the fracture surface of type 304 was about 70%. Microstructural evolution of EHP and type 304 after irradiation was examined by TEM. The defects induced by irradiation mostly consisted of black dots and frank loops in both specimens. No void was also observed in grain and grain boundary of both specimens. There was a little difference in microstructure after irradiation. It is believed that EHP alloy is superior to type 304 in irradiation.
Ioka, Ikuo; Suzuki, Jun; Kiuchi, Kiyoshi; Nakayama, Jumpei*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 5 Pages, 2011/10
An Extra high purity Nb-W alloy (Nb-W EHP) that is superior to current Zr for cracking in severer environment is developed for reprocessing plants. Nb-W EHP has a technical limitation in weldability for practical use. To choose an optimum composition range of W, Nb-W EHP with different content of W were manufactured. Weld joint specimens were prepared by TIG welding. Tensile, corrosion and Charpy impact tests were carried out to examine the effect of W on chemical and mechanical properties of Nb-W EHP. There was little effect of W on corrosion rate of weld joint of Nb-W EHP. The increase in W improves ultimate tensile strength of Nb-W EHP. A ductile-brittle transition temperature (DBTT) of Nb-W EHP increased with an increase in W. The increase in DBTT is a problem as a structural material. Moreover, DBTT of the weld joint was high about 150K in comparison with the base metal. It is necessary to adjust W content to 8wt% or less in Nb-W EHP from the obtained results.
Nagaoka, Shinichi*; Fukuzawa, Hironobu*; Prmper, G.*; Takemoto, Mai*; Takahashi, Osamu*; Yamaguchi, Takuhiro*; Kakiuchi, Takuhiro*; Tabayashi, Kiyohiko*; Suzuki, Isao*; Harries, J.; et al.
Journal of Physical Chemistry A, 115(32), p.8822 - 8831, 2011/07
Kim, G.; Shiba, Kiyoyuki; Sawai, Tomotsugu; Ioka, Ikuo; Kiuchi, Kiyoshi
Materials Research Society Symposium Proceedings, Vol.1298, p.61 - 66, 2011/04
The irradiation behaviour in two different precipitation hardening types of Ni-base alloys with the ultra high purity grade (EHP), namely, the ' type and G phase type was investigated by multi-ion beam techniques simulated to the irradiation conditions in fuel cladding tubes used in sodium cooled FBRs. Single ion-beam irradiation tests were conducted up to 90 dpa (by Fe or Ni) at 673 K. Triple ion-beam irradiation tests were conducted up to 90 dpa (by Ni, 90 appmHe and 1350 appmH) at 823 K. The irradiation behaviour was examined by nano-indentation tests to irradiation hardening, and the microscopic observation by TEM to the distribution of dislocations, cavities and voids. The behaviour was compared with those of PNC316. The dominating irradiation defects in EHP(') alloy at 673 K by single ion-beam are Frank loops, perfect unfaulted loops and line dislocations. Whereas, those of EHP(WSi) alloy are the irradiation-induced G phase precipitates along planes. Those dominating defect structures at 823 K by triple ion-beam are classified as followings, bimodal distributions in EHP('), bubbles in EHP(WSi) and voids in PNC316. The ratio of void swelling is estimated as nearly 0.01% in EHP(WSi), 0.2% in EHP('), 3.4% in PNC316. From those results, the excellent irradiation properties of EHP(WSi) alloy is clarified as the inhibition effects of secondary irradiation defects.
Ioka, Ikuo; Suzuki, Jun; Kiuchi, Kiyoshi; Nakayama, Jumpei*
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 6 Pages, 2010/05
An Extra High Purity austenitic stainless steel (EHP alloy) was developed with conducting the new multiple refined melting in order to suppress the harmful impurities less than 100 ppm. EHP alloy has great intergranular corrosion resistance. It is considered that intergranular corrosion becomes initiation of SCC. So, we try to apply EHP alloy to weld overlay materials to prevent from SCC. EHP alloy was melted by the new multiple refined method. The conventional weld metals were also prepared as comparisons. Specimens were machined from the welded metal of each material. Intergranular corrosion tests were performed in boiling 8 kmol/m HNO solutions containing 1 kg/m Cr(VI) ions. The intergranular corrosion of conventional weld metals was severer than those of EHP alloys. Crevice Beam bending tests to evaluate susceptibility of SCC were carried out in high temperature water of 561 K with saturated oxygen for 1000 h. Cracks and intergranular corrosion of conventional weld metals were much more than those of EHP alloys. It was confirmed that EHP alloy had excellent SCC resistance in comparison with conventional materials when EHP alloy was used as a weld metal.
Ioka, Ikuo; Kiuchi, Kiyoshi; Nakayama, Jumpei*
Materia, 49(3), p.122 - 124, 2010/03
In reprocessing plants, it is known that the corrosion environment for structural materials is extremely severe to be used boiling nitric acid with oxidized ions related to fission products (FP). Ultra low carbon stainless steels (SUS304ULC) were developed from the experience of corrosion troubles in foreign plants and the operation of domestic plant. In an advanced reprocessing plant, it is predicted that the corrosion environment will be more severe because of the advanced spent fuels which contain much FP than an existing spent fuel. Therefore, it is necessary to develop the new stainless steel with excellent corrosion resistance in comparison with SUS304ULC. We developed an extra high purity stainless steel (EHP alloy) having superior corrosion resistance in the strong acid environment for the advanced reprocessing plant. In the production process of EHP alloy, the practical use was taken into consideration. This report introduces the present conditions of EHP alloy.
Ioka, Ikuo; Suzuki, Jun; Motooka, Takafumi; Kiuchi, Kiyoshi; Nakayama, Jumpei*
Journal of Power and Energy Systems (Internet), 4(1), p.105 - 112, 2010/02
An intergranular corrosion is an important degradation mechanism of austenitic stainless steels for use in a nuclear fuel reprocessing plant. The intergranular corrosion is caused by the segregation of impurities to grain boundaries. An extra high purity austenitic stainless steel (EHP) was developed with conducting the new multiple refined melting to suppress the impurities less than 100ppm. The intergranular corrosion behavior of EHP alloys added various impurities was examined in boiling nitric acid solution with highly oxidizing ions. A multi regression analysis was performed using the obtained data. The degree of influence of the impurities on intergranular corrosion was shown from the analysis. The influence on corrosion rate became small in order of B, P, Si, C, S and Mn. There is little effect of Mn on corrosion rate of EHP-SSs in case of 10000appm or less.
Ioka, Ikuo; Suzuki, Jun; Motooka, Takafumi; Kiuchi, Kiyoshi; Nakayama, Jumpei
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 5 Pages, 2009/06
An intergranular corrosion is an important degradation mechanism of austenitic stainless steels for use in a nuclear fuel reprocessing plant. The intergranular corrosion is caused by the segregation of impurities to grain boundaries. An extra high purity austenitic stainless steel (EHP) was developed with conducting the new multiple refining to suppress the impurities less than 100 ppm. The intergranular corrosion behavior of EHP alloys with various impurities was examined in boiling nitric acid solution. The intergranular corrosion was observed in impurity doped EHP alloys, although no intergranular corrosion was observed in EHP alloy. From the obtained results, an empirical equation between susceptibility of intergranular corrosion and impurities was established by means of the regression analysis. The degree of influence of the impurities on intergranular corrosion was shown.
Ishijima, Yasuhiro; Ioka, Ikuo; Kiuchi, Kiyoshi; Kaneko, Tetsuji*; Okubo, Tsutomu; Yamamoto, Masahiro
Atsuryoku Gijutsu, 47(1), p.12 - 17, 2009/01
We investigate one of these innovative water reactors; Fast Spectrum Light Water Reactor (FLWR). It has unique construction for the reactor core but the fuel cladding material will be exposed in high internal pressure and axial load and complex temperature distribution. Therefore, we conducted a specially designed fatigue-creep test that were simulated several parameters (thermal distribution, temperature variation, internal pressure variation and binding stress) to evaluate an applicability of fuel cladding material for FLWR. Zircalloy-2, which is common cladding material, was used for the test. Test result was confirmed to compare the deformation value between tested and calculated. The result showed the evaluation method could be controlled several parameters simultaneously and the deformation value after the test coincided to the calculated value. This method is sufficient to evaluate thermal deformation characteristics for FLWR.
Ioka, Ikuo; Kato, Chiaki; Kiuchi, Kiyoshi; Nakayama, Jumpei
Journal of Power and Energy Systems (Internet), 3(1), p.31 - 37, 2009/00
Austenitic stainless steels suffer intergranular attack in boiling HNO with oxidants. The intergranular corrosion is mainly caused by impurities at the grain. An extra high purity austenitic stainless steel (EHP alloys) was developed with conducting the new technique in order to suppress harmful impurities less than 100 ppm. The corrosion behavior of type 310 EHP alloy in HNO with highly oxidants was investigated. The straining, aging and recrystallizing (SAR) treated type 310 EHP alloy showed superior corrosion resistance for intergranular attack than solution annealed (ST) type 310 EHP alloy with same impurity level. Boron segregation at the grain boundary was detected in only ST specimen using a FTE method. It is believed that the segregated boron along the grain boundaries in type 310 EHP alloy was one of main factor of intergranular corrosion. The SAR treatment was effective to restrain the intergranular corrosion for type 310 EHP alloy with B up to 7 ppm.
Nakahara, Yukio; Yamamoto, Masahiro; Karasawa, Hidetoshi*; Kiuchi, Kiyoshi; Katsumura, Yosuke*
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
Three types of commercial-grade austenitic stainless steel, Type 304L, Type 316L, and Type 310S, were immersed in deaerated supercritical water (SCW) of 25 MPa, 550 C with -ray irradiation for 1000 hours in total. Absorbed dose rates in SCW are estimated at 5-15 kGy h. High temperature oxidation experiments resulted in the formation of two-layer oxide film in which the outer layer is porous iron oxide and the inner layer is dense iron-chromium-nickel oxide. Rate constants of parabolic rate law in net weight gains of Type 304L SS and Type 310S SS are decreased as the -ray exposure rate is increased. The net weight gain of -ray irradiated Type 316L SS is sharply fluctuating and decreasing, because the flaking of the outer layer occurred. Hematite is formed in the outer layer on all irradiated samples of examined alloys. The concentration of chromium on the surface is increased by -ray irradiation. A chromium-rich part in the inner layer accompanying a nickel-rich part in the metal is formed along to the oxide/metal interface of -ray irradiated Type 304L SS, non-irradiated and -ray irradiated Type 316L SS.
Ioka, Ikuo; Kato, Chiaki; Kiuchi, Kiyoshi; Nakayama, Jumpei
Proceedings of 16th International Conference on Nuclear Engineering (ICONE-16) (CD-ROM), 5 Pages, 2008/05
Austenitic stainless steels suffer intergranular attack in boiling nitric acid with oxidants. The intergranular corrosion is mainly caused by the segregation of impurities to grain. An extra high purity austenitic stainless steel (EHP alloys) was developed with conducting the new multiple refined melting technique in order to suppress the total harmful impurities less than 100ppm. The basically corrosion behavior of type 310 EHP alloy with respect to nitric acid solution with highly oxidizing ions was investigated. The straining, aging and recrystallizing (SAR) treated type 310 EHP alloy showed superior corrosion resistance for intergranular attack. The segregated boron along the grain boundaries was one of main factor of intergranular corrosion from fission track etching results. The SAR treatment was effective to restrain the intergranular attack for type 310 EHP alloy with B less than 7ppm.
Ogawa, Hiroaki; Kiuchi, Kiyoshi
Hyomen Gijutsu, 58(9), p.543 - 549, 2007/09
The oxidation mechanism of nuclear system components accelerated under irradiation was basically examined using a micro-balance device installed an rf low energy plasma source. Six alloys with different Cr contents of Fe-Ni alloy, stainless steels, Ni base alloys and metallic Cr were used in the experiment. The difference in the oxidation behavior between the low energy plasma and thermal equilibrium in oxygen atmosphere of 20 Pa was evaluated at 873 K for 20 hours. The oxidation behavior was analyzed with the in-situ weight change and the surface analysis by X-ray photoelectron spectroscopy (XPS). On the thermal equilibrium, the weight gain is depressed with increasing Cr contents. On the other hand, the weigh loss was observed in the materials with high Cr contents in the low energy oxygen plasma with the high excitation condition. The oxidation behavior was analyzed with the mechanistic model, by considering the evaporation rate of volatile CrO on the surface and the migration rate of O into the substrate.
Motooka, Takafumi; Ishikawa, Akiyoshi; Numata, Masami; Endo, Shinya; Itonaga, Fumio; Kiuchi, Kiyoshi; Kizaki, Minoru
JAEA-Research 2007-031, 20 Pages, 2007/03
An effect of neptunium ions on corrosion of stainless steel in nitric acid solution was investigated by corrosion tests. Type SUS304L stainless steel was used for the tests. The corrosion tests were conducted in 9kmol/m nitric acid solutions containing neptunium ions, where test specimens were immersed or heat-transferred. As a result, we found that neptunium ions promote corrosion of stainless steels in nitric acid solution. This finding would contribute to modifications of the materials for spent fuel reprocessing process.