Evaluation of source term parameters for spent fuel disposal in foreign countries, 2; Dissolution rates of spent fuel matrices and construction materials for fuel assemblies

Kitamura, Akira; Chikazawa, Takahiro*; Akahori, Kuniaki*; Tachi, Yukio

Genshiryoku Bakkuendo Kenkyu (CD-ROM), 23(1), p.55 - 72, 2016/06

https://doi.org/10.3327/jnuce.23.1_55

The Japanese geological disposal program has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter "direct disposal of SF") as an alternative management option other reprocessing followed by vitrification and geological disposal of high-level radioactive waste. We conducted literature survey of dissolution rate of SF matrix and constructing materials (e.g. zircaloy cladding and control rods) selected in safety assessment reports for direct disposal of SF in Europe and United States. We also investigated basis of release rate determination and assignment of uncertainties in the safety assessment reports. Furthermore, we summarized major conclusions proposed by some European projects governed by European Commission. It was found that determined release rates are fairly similar to each other due to use of similar literature data in all countries of interest. It was also found that the determined release rates were including conservativeness because it was difficult to assign uncertainties quantitatively. It is expected that these findings are useful as fundamental information for determination of the release rates for the safety assessment of Japanese SF disposal system.

Extra radiation hardening and microstructural evolution in F82H by high-dose dual ion irradiation

Ando, Masami; Wakai, Eiichi; Sawai, Tomotsugu; Matsukawa, Shingo; Naito, Akira*; Jitsukawa, Shiro; Oka, Keiichiro*; Tanaka, Teruyuki*; Onuki, Somei*

JAERI-Review 2004-025, TIARA Annual Report 2003, p.159 - 161, 2004/11

http://jolissrch-inter.tokai-sc.jaea.go.jp/pdfdata/JAERI-Review-2004-025-p0001-p0116.pdf

The objectives of this study are to evaluate radiation hardening on ion-irradiated F82H up to 100 dpa and to examine the extra component of radiation hardening due to implanted helium atoms (up to 3000 appmHe) in F82H under ratio of 0, 10, 100 appmHe/dpa.The ion-beam irradiation experiment was carried out at the TIARA facility of JAERI. Specimens were irradiated at 633 K by 10.5 MeV Fe ions with/without 1.05 MeV He ions. Micro-indentation tests were performed at loads to penetrate about 0.40 mm in the irradiated specimens using an UMIS-2000. The results are summarized as follows:1) As a result of the single irradiated F82H, the micro-hardness tended to increase about 30 dpa. 2) The extra radiation hardening was obviously caused by co-implanted helium atoms more than 1000 appm in F82H irradiated at 633 K. 3) In the dual-beam (100 appmHe/dpa) irradiated microstructure, nano-voids and fine defects were observed. It is suggested that the formation of nano-voids causes the extra radiation hardening by helium co-implantation.

Reduced activation martensitic steels as a structural material for ITER test blanket

Shiba, Kiyoyuki; Enoeda, Mikio; Jitsukawa, Shiro

Journal of Nuclear Materials, 329-333(Part1), p.243 - 247, 2004/08

https://doi.org/10.1016/j.jnucmat.2004.04.018

no abstracts in English

Highly thermal conductive sintered SiC fiber-reinforced 3D SiC/SiC composites; Experiments and finite-element analysis of the thermal diffusivity/conductivity

Yamada, Reiji; Igawa, Naoki; Taguchi, Tomitsugu; Jitsukawa, Shiro

Journal of Nuclear Materials, 307-311(Part2), p.1215 - 1220, 2002/12

https://doi.org/10.1016/S0022-3115(02)00957-1

SiC fiber-reinforced SiC composites (SiC/SiC) are considered an advanced structural material for blanket modules of a fusion reactor, which requires high thermal conductivity in order to keep thermal stresses in the material lower than the allowable design stress. The sintered SiC fiber recently developed has obtained high thermal conductivity, so it is highly expected that sintered SiC fiber-reinforced SiC/SiC composites would also show high thermal conductivity. In this study several types of 3D SiC/SiC composites were fabricated by either CVI or PIP method. The results of the thermal conductivity measurements show that the maximum thermal conductivity at room temperature was about 60 W/mK for CVI composites or 25W/mK for PIP ones. These values are considerably higher than those of non-sintered SiC fiber reinforced SiC/SiC composites, which indicates a possibility that the developed materials would be promising. The FEM thremal analysis shows the good agreement between the caluculated and experimental results.

Swelling behavior of TIG-welded F82H IEA heat

Sawai, Tomotsugu; Wakai, Eiichi; Tomita, Takeshi; Naito, Akira; Jitsukawa, Shiro

Journal of Nuclear Materials, 307-311(Part1), p.312 - 316, 2002/12

https://doi.org/10.1016/S0022-3115(02)01200-X

TIG-weld joints of the IEA heat of F82H were irradiated by TIARA. Transmission electron microscope (TEM) specimens were prepared by a focused ion beam (FIB) system. TEM specimens were obtained from the heat affected zone (HAZ) and the weld metal. HAZ specimens had typical bi-modal cavity microstructure after 50 dpa at 450C with He/dpa ratio of 10 appmHe/dpa. Larger voids about 30 nm were observed in the tempered side specimen, while the size of voids in the quenched side specimen was less than 15 nm. Ac1 temperature determined by heat-treated base metal is 820C. Cavity growth in specimens tempered at higher temperature was enhanced, while that in cold worked ones is suppressed.

Effect of simultaneous ion irradiation on microstructural change of SiC/SiC composites at high temperature

Taguchi, Tomitsugu; Wakai, Eiichi; Igawa, Naoki; Nogami, Shuhei*; Snead, L. L.*; Hasegawa, Akira*; Jitsukawa, Shiro

Journal of Nuclear Materials, 307-311(Part2), p.1135 - 1140, 2002/12

https://doi.org/10.1016/S0022-3115(02)01051-6

no abstracts in English

The State and trend of IASCC study

Tsukada, Takashi

Nippon Yosetsu Kyokai "Genshiryoku Kozo Kiki No Zairyo, Sekkei, Seko, Kensa Ni Kansuru Koshukai" Tekisuto, 40 Pages, 2002/00

no abstracts in English

Aging degradation of light water reactor materials; Reactor internal and pressure vessel materials

Tsukada, Takashi; Ebine, Noriya

Nippon AEM Gakkai-Shi, 9(2), p.171 - 177, 2001/06

no abstracts in English

Studies on interaction between cesium iodide and type 316 stainless steel in WIND project

Kudo, Tamotsu; Maruyama, Yu; Shibazaki, Hiroaki*; Hidaka, Akihide; Nakamura, Hideo; Chino, Eiichi; Yuchi, Yoko; Hashimoto, Kazuichiro

JAERI-Conf 2000-015, p.216 - 221, 2000/11

https://doi.org/10.11484/jaeri-conf-2000-015

no abstracts in English

Research on development of high-purity iron-based alloys; Manufacture, analysis of small amount of element and property tests

; *; ; ; Aoto, Kazumi;

JNC-TN9400 2000-059, 43 Pages, 2000/05

JNC-TN9400-2000-059.pdf:2.08MB

The purpose of this study is to understand the material properties of manufacturable high-purity iron and high-purity iron-based alloy in present technology and to get an applicable prospect for the structural and functional material of the frontier fast reactor. Then the about 10kg high-purity iron and iron-based alloy were melted using a cold-crucible induction melting furnace under the ultra-high vacuum. Subsequent to that, the compatibility between the melted material and the high-temperature sodium environment which is a special feature of the fast reactor and tensile property at room and elevated temperatures were investigated using the melted materials. Also, the creep test using the high-purity 50%Cr-Fe alloy at 550C in air in order to understand the high temperature creep property. ln addition, the material properties such as thermal expansion coefficient, specific heat and electrical resistance were measured and to evaluate the outlook for the structural material for the fast reactor. The following results were obtained based on the property test and evaluation. (1)lt was possible to melt the about 10kg high-purity ingot and high-purity 50%Cr-Fe alloy ingot using a cold-crucible induction melting furnace under the ultra-high vacuum. (2)The tensile tests of the high-purity 50%Cr-Fe alloy were performed at room and elevated temperatures in order to understand the deformation behavior. From the experimental results, it was clear that the high-purity 50%Cr-Fe alloy possesses high strength and good ductility at elevated temperatures. (3)The physical properties (the thermal expansion coefficient and specific heat etc.) were measured using the high-purity 50%Cr-Fe alloy. lt was clear that the thermal expansion coefficient of high-purity 50%Cr-Fe alloy was smaller than that of SUS304. (4)From the corrosion test in liquid sodium, the ordinary-purity iron showed the weight loss after corrosion test. However the high-purity iron showed ...