Inverse pole figure mapping of bulk crystalline grains in a polycrystalline steel plate by pulsed neutron Bragg-dip transmission imaging

Sato, Hirotaka*; Shiota, Yoshinori*; Morooka, Satoshi; Todaka, Yoshikazu*; Adachi, Nozomu*; Sadamatsu, Sunao*; Oikawa, Kenichi; Harada, Masahide; Zhang, S.*; Su, Y. H.; et al.

Journal of Applied Crystallography, 50(6), p.1601 - 1610, 2017/12

https://doi.org/10.1107/S1600576717012900

Evaluation of mechanical properties and nano-meso structures of 9-11%Cr ODS steels

Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Oba, Yojiro*; Onuma, Masato*; Koyama, Shinichi; Tanaka, Kenya

Journal of Nuclear Materials, 440(1-3), p.568 - 574, 2013/09

https://doi.org/10.1016/j.jnucmat.2013.04.006

This study carried out mechanical tests and microstructure characterizations of several 9Cr and 11Cr-ODS tempered martensitic steels, and discussed the appropriate chemical composition range of 11Cr-ODS tempered martensitic steel from the viewpoint of high-temperature strength improvement. It was shown that the residual -ferrite fraction in 11Cr-ODS steel was successfully controlled to the same level as the 9Cr-ODS steel by selecting the matrix chemical compositions on the basis of the multi-component phase diagram. The tensile strength decreased with decreasing W content from 2.0 to 1.4 wt%. On the other hand, creep strength at 973 K did not degrade by the decreasing W content. Both tensile strength and creep strength increased with increasing population of the nano-sized oxide particles. Small angle X-ray scattering analysis revealed that titanium and excess oxygen contents were key parameters in order to improve the dispersion condition of nano-sized oxide particles.

Preliminary investigation on capsule for low-temperature irradiation tests

Inaba, Yoshitomo; Ishida, Takuya; Onuma, Yuichi; Saito, Takashi

JAEA-Technology 2009-014, 42 Pages, 2009/05

JAEA-Technology-2009-014.pdf:6.09MB

In order to carry out low-temperature irradiation tests under the high neutron flux in the JMTR core, desirable capsules were investigated from a survey and evaluation of current heat removal techniques. As a result, it was found that the low-temperature irradiation tests can be realized by the development of the capsule with cooling fins or the capsule using a boiling medium. In the case of the irradiation tests at about 100C, the capsule with the fins can be used, and the reactor cooling water cools the capsule including specimens. This technique has few subjects to realize. In the case of the irradiation tests at below 0C, the capsule using the boiling medium can be used, and the cooling of specimens in the capsule by liquid nitrogen is needed. In the present status, it is difficult that the liquid nitrogen is supplied to the capsule, and this technique has to overcome various subjects to realize. The investigation to solve these subjects will be carried out in the near future.

Detaching test on an irradiated mock-up contaminated with tritium from the core of JMTR

Tomita, Kenji; Tsuchiya, Kunihiko; Onuma, Yuichi; Inoue, Shuichi; Watanabe, Hiroyuki; Saito, Takashi; Kikuchi, Taiji; Hayashi, Kimio; Kitajima, Toshio

JAEA-Technology 2008-036, 61 Pages, 2008/06

JAEA-Technology-2008-036.pdf:7.47MB

The second in-situ irradiation experiment using a mock-up (ORIENT-II, JMTR capsule Number: 99M-54J) with a tritium breeder (LiTiO) pebble bed in the Japan Materials Testing Reactor (JMTR) was finished on Aug. 1, 2006. Consideration on the detaching procedure of the irradiated mock-up contaminated with tritium with pebble bed and a detaching test of this mock-up was carried out. The tritium removal properties were examined in the irradiated mock-up, the sweep gas tube, the protective tube and the junction box, for the decreasing of the tritium release to the area of detaching test. Melting/enclosed tests of sealing plug were also carried out for the prevention of tritium leakage from sweep gas lines of pebble bed. Then, the detaching test of the pebble bed was carried out. This report describes the results of tritium removal tests, examination of the detaching procedure, and the detaching test, as well as knowledge obtained from these tests and works.

Fabrication of irradiation capsule for IASCC irradiation tests, 1; Irradiation capsule for crack growth test (Joint research)

Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; et al.

JAEA-Technology 2008-011, 46 Pages, 2008/03

JAEA-Technology-2008-011.pdf:19.39MB

It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, It is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack growth test is reported.

Study on Proper Constitutive Model for Evaluation of Long Term Mechanical Behavior of Buffer Material

Hirai, Takashi; Tanai, Kenji; Kikuchi, Hirohito*; Shigeno, Yoshimasa*; Namikawa, Tsutomu*; Takaji, Kazuhiko*; Onuma, Satoshi*

JNC TN8400 2003-034, 158 Pages, 2004/02

JNC-TN8400-2003-034.pdf:5.26MB

The objective of this report is to make a proposal of the proper constitutive models and parameters for the evaluation of the long term mechanical behavior of the buffer material in the engineered barrier system. In the second progress report by JNC, it was reported that the well designed engineered barrier system is stable and safety on mechanical support of the overpack to ensure stability and stress which acts on the overpack by using analysis which based on the popular constitutive models for the general caly soils. However, the buffer material which has swelling characteristics is considered not to be ordinary clay soils. So it is necessary to select the reliable constitutive models again. Therefor the proper models were selected again systematically in the several models which have been used for the assessment of the behavior of clay soils and the simulation analysis on the laboratory tests were carried out by using these models. From the result of the simulation analysis it appeared that the selected two models were alike to assess the behavior of the buffer material and the parameters which need to simulate the consolidation tests are different from those for the triaxial compression tests. Finally the analysis was conducted to evaluate the effect of the swelling of the overpack by the collosion and the self weight which causes the sedimentation of the overpack. From the analytical result, it was clarified that two kinds of parameter sets are necessary to evaluate the deformation and the stress of the buffer material in the engineered barrier system.

Assessment on mechanical effect of engineering barrier system to fault movement

Hirai, Takashi; Tanai, Kenji; Kikuchi, Hirohito*; Suzuki, Hideaki*; *; Onuma, Satoshi*

JNC TN8400 2003-009, 56 Pages, 2003/03

JNC-TN8400-2003-009.pdf:7.22MB

The objective of this report is to clarify mechanical effect of engineered barrier system to the unavoidable fault movement. From the basic policy of the second progess report by JNC, natural phenomenon which affect strongly to the geological disposal system shoult be avoided. However, small faults as sliprate "C" far from principal fault zone, are difficult to be found out completely. Therefore, it is important to evaluate the influence of these fault movements and to clarify stability and safety of the engineered barrier system. Accordingly, the effect of a rock displacement across a deposition holl was considered and the midium scale test was carried out. Then midium scale test was simulated by Finit Element Method in which the constitutive model of Tresca was adopted to analyze elastoplastic behavior of buffer material. From the result of the midium scale test and the analysis, it was realized that the buffer material diminish shear stress acting on the overpack. Further analytical study was conducted to evaluate the real scale engineered barrier system designed in the second progress report by JNC. From the study, it was apeared that stress in buffer corresponded to the stress calculated for the midium scale test model. Consequently, it was obvious that rock displacement, 80% of buffer thickness, didn't affect overpack if velocity of fault movement was under 10 cm/sec.

Speciation of long-lived radionuclides in soils in the Chernobyl exclusion zone

Amano, Hikaru; Ueno, Takashi; Onuma, Yoshikazu*

Proceedings of the International Workshop on Distribution and Speciation of Radionuclides in the Environment, p.169 - 173, 2000/00

no abstracts in English

Speciation of environmental radionuclides in the Chernobyl 30km zone

Amano, Hikaru; Hanzawa, Yukiko; Watanabe, Miki*; Matsunaga, Takeshi; Ueno, Takashi; Nagao, Seiya; Yanase, Nobuyuki; Onuma, Yoshikazu*

Proceedings of OECD/NEA Workshop on Evaluation of Speciation Technology, p.211 - 218, 1999/00

no abstracts in English

Behavior of long lived radionuclides in surface environment around the Chernobly Nuclear Power Plant

Amano, Hikaru; Matsunaga, Takeshi; Ueno, Takashi; Nagao, Seiya; Watanabe, Miki*; *; Onuma, Yoshikazu*

KURRI-KR-18, p.201 - 212, 1997/00

no abstracts in English