Kurata, Yuji; Futakawa, Masatoshi; Saito, Shigeru
Journal of Nuclear Materials, 343(1-3), p.333 - 340, 2005/08
In order to study effects of temperature and alloying elements on corrosion behavior in liquid Pb-Bi which will be used for an accelerator driven system (ADS), corrosion tests of various steels were conducted under static liquid Pb-Bi condition. The tests were performed in oxygen-saturated liquid Pb-Bi at 450C and 550C for 3000h. Oxide films were formed during corrosion at 450C and 550C. Corrosion depth of steels decreased at 450C with increasing Cr content of steels. Austenitic stainless steels containing Ni didn't exhibit appreciable dissolution of Ni and Cr at 450C. The thick ferrite layer produced by dissolution of Ni and Cr was found in JPCA and type 316ss at 550C. For this reason the corrosion depth of austenitic stainless steels, JPCA and type 316ss became large. A Si-added austenitic stainless steel showed good corrosion resistance at 550C because a protective oxide film formed on the steel prevented dissolution of Ni and Cr into liquid Pb-Bi.
Saito, Shigeru; Kikuchi, Kenji; Usami, Koji; Ishikawa, Akiyoshi; Nishino, Yasuharu; Kawai, Masayoshi*; Dai, Y.*
Journal of Nuclear Materials, 343(1-3), p.253 - 261, 2005/08
A beam window of a spallation target will be subjected to proton/neutron irradiation, pressure wave and thermal stresses accompanied by high-energy proton beam injection. To obtain the irradiation data, the SINQ target irradiation program (STIP) was initiated in 1996 at Paul Scherrer Institute (PSI) and has been progressing. JAERI takes part in STIP and shares the PIE work. In this study, the results of tensile tests on austenitic stainless steels, JPCA and 316F SS, will be reported. The results indicate that the irradiation causes considerable hardening and degradation of ductility. The YS increases in this study are slightly large in comparison with those irradiated at fission reactor. Strain-to-necking (STN) values show sufficient large ductility of the irradiated JPCA-SA and 316F-SA. The trends of the STN decrease in this study are slightly abrupt in comparison with those irradiated at fission reactor. All specimens, including irradiated at embrittlement temperature for austenitic steels, fractured in ductile manner.
Teshigawara, Makoto; Harada, Masahide; Saito, Shigeru; Kikuchi, Kenji; Kogawa, Hiroyuki; Ikeda, Yujiro; Kawai, Masayoshi*; Kurishita, Hiroaki*; Konashi, Kenji*
Journal of Nuclear Materials, 343(1-3), p.154 - 162, 2005/08
For decoupled and poisoned moderator, a thermal neutron absorber, i.e., decoupler, is located around the moderator to give neutron beam with a short decay time. A B4C decoupler is already utilized, however, it is difficult to use in a MW class source because of He void swelling and local heating by (n,a) reaction. Therefore, a Ag-In-Cd (AIC) alloy which gives energy-dependence of macroscopic neutron cross section like that of B4C was chosen. However, from heat removal and corrosion protection points of view, AIC is needed to bond between an Al alloy (A6061-T6), which is the structural material of a moderator. An AIC plate is divided into a Ag-In (15wt%) and Ag-Cd(35wt%) plate to extend the life time, shorten by burn up of Cd. We performed bonding tests by HIP (Hot Isostatic Pressing). We found out that a better HIP condition was holding at 803 K, 100 MPa for 1 h for small test pieces (f20mm). Though a hardened layer is found in the bonding layer, the rupture strength of the bonding layer is more than 20 MPa, which is less than that of the design stress.
Nemoto, Yoshiyuki; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi
Journal of Nuclear Materials, 343(1-3), p.313 - 317, 2005/08
The aim of this work is to evaluate corrosion behavior of irradiated materials these would be used in spallation neutron sources. Solution annealed high purity Fe-18Cr-12Ni specimens were used in this study. Ni, H+ and He ions were irradiated at 473 to 773K. After corrosion procedure, the specimens were examined with atomic force microscope (AFM) to evaluate corrosion behavior. It was shown that the corroded volume of irradiated area increased with radiation damage. Higher H content at lower temperature accelerated corrosion, but H at higher temperature did not accelerate corrosion. He suppressed corrosion, and corroded volume was larger for the specimens irradiated at higher temperature than these at lower temperature. It is suggested from this study that H and He affect the corrosion behavior of irradiated alloy.
Futakawa, Masatoshi; Naoe, Takashi; Tsai, C.-C.*; Kogawa, Hiroyuki; Ishikura, Shuichi*; Ikeda, Yujiro; Soyama, Hitoshi*; Date, Hidefumi*
Journal of Nuclear Materials, 343(1-3), p.70 - 80, 2005/08
no abstracts in English
Wakai, Eiichi; Jitsukawa, Shiro; Tomita, Hideki*; Furuya, Kazuyuki; Sato, Michitaka*; Oka, Keiichiro*; Tanaka, Teruyuki*; Takada, Fumiki; Yamamoto, Toshio*; Kato, Yoshiaki; et al.
Journal of Nuclear Materials, 343(1-3), p.285 - 296, 2005/08
The dependence of helium production on radiation-hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel doped with B, B and B+B irradiated at 250C to 2.2 dpa. The total amounts of doping boron were about 60 massppm. The range of He concentration produced in the specimens was from about 5 to about 300 appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50 MeV-He irradiation was also performed to implant about 85 appm He atoms at 120C by AVF cyclotron to 0.03 dpa, and small punch testing was performed to obtain DBTT. Radiation-hardening of the neutron-irradiated specimens increased slightly with increasing He production. The 100 MPam DBTT for the F82H+B, F82H+B+B, and F82H+B were 40, 110, and 155C, respectively. The shifts of DBTT due to He production were evaluated as about 70C by 150 appmHe and 115C by 300 appmHe. The DBTT shift in the small punch testing was evaluated as 50C.
Kogawa, Hiroyuki; Ishikura, Shuichi*; Sato, Hiroshi; Harada, Masahide; Takatama, Shunichi*; Futakawa, Masatoshi; Haga, Katsuhiro; Hino, Ryutaro; Meigo, Shinichiro; Maekawa, Fujio; et al.
Journal of Nuclear Materials, 343(1-3), p.178 - 183, 2005/08
A cross-flow type (CFT) mercury target with flow guide blades, which has been developed for JSNS, can suppress the generation of stagnant flow region especially near the beam window where the peak heat density is generated due to spallation reaction. Then, a flat type beam window has been applied to the CFT target from the viewpoint of suppressing dynamic stress caused by a pressure wave, which has been estimated with a mercury model of the linear equation of state. The recent experimental results obtained by using a proton beam incidents to mercury led that a cutoff pressure model in the equation of state of mercury caused a suitable dynamic stress with experimental results. Dynamic stress analyses were carried out with the cutoff pressure model, in which the negative pressure less than 0.15 MPa was not generated. The generated dynamic stress in the flat beam window became much larger than that in a semi-cylindrical type window. However, the generated stress in the semi-cylindrical type beam window was over the allowable stress of SS316L under the peak heat density of 668 W/cc. In order to decrease the dynamic stress in the semi-cylindrical beam window, the incident proton beam was defocused to decrease the peak heat density down to 218 W/cm. As a result, the dynamic stress could be suppressed less than the allowable stress. On the other hand, due to defocus of the proton beam, high heat density was generated on the end of the flow guide blades, which caused high thermal stress exceeding the allowable stress. To decrease the thermal stress, several shapes of the blade ends were studied analytically, which were selected so as not to affect the mercury flow distribution. A simple thin-end blade showed low thermal stress below the allowable stress.
Journal of Nuclear Materials, 343(1-3), p.7 - 13, 2005/08
This paper reports the current status of The Materials and Life Experimental Facility construction under the high intensity proton accelerator projet(J-PARC), which has been conducted by JAERI and KEK collaboratively.Alng with designs and schedule of the neutron source, critical technical issues, e.g., mercury target material and moderator materials, which are still remained to be settled, and activities for development are shown.
Harada, Masahide; Watanabe, Noboru; Konno, Chikara; Meigo, Shinichiro; Ikeda, Yujiro; Niita, Koji*
Journal of Nuclear Materials, 343(1-3), p.197 - 204, 2005/06
For a construction of maintenance and storage scenarios for JSNS, lives of structure material need to be estimated. DPA (Displacement per Atom) was a major index of radiation damage. So we evaluated DPA value of each component. Function of the DPA calculation was equipped to the PHITS code, which was particle and heavy ion transport code. For DPA calculation, displacement cross section was necessary. Displacement cross sections of neutron below 150 MeV were processed by the NJOY code from LA150 library and those of neutron above 150MeV and proton in the all energy region were obtained from energies of fragments calculated in the PHITS. By using the PHITS, we calculated DPA values and DPA mapping. We obtained that the peak DPA values at end of 5000MWh operation were 4.1 for target vessel, 2.8 for reflector and moderator vessels, and 0.4 for proton beam windows, respectively. We estimated the target life at 1 year and the moderator life at 6 year.