Evaluation of irradiation behavior on oxide dispersion strengthened (ODS) steel claddings irradiated in Joyo/CMIR-6

Yano, Yasuhide; Otsuka, Satoshi; Yamashita, Shinichiro; Ogawa, Ryuichiro; Sekine, Manabu; Endo, Toshiaki; Yamagata, Ichiro; Sekio, Yoshihiro; Tanno, Takashi; Uwaba, Tomoyuki; et al.

JAEA-Research 2013-030, 57 Pages, 2013/11

JAEA-Research-2013-030.pdf:48.2MB

It is necessary to develop the fast reactor core materials, which can achieve high-burnup operation improving safety and economical performance. Ferritic steels are expected to be good candidate core materials to achieve this objective because of their excellent void swelling resistance. Therefore, oxide dispersion strengthened (ODS) ferritic steel and 11Cr-ferritic/martensitic steel (PNC-FMS) have been respectively developed for cladding and wrapper tube materials in Japan Atomic Energy Agency. In this study, the effects of fast neutron irradiation on mechanical properties and microstructure of 9Cr-and 12Cr-ODS steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the CMIR-6 at temperatures between 420 and 835C to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures.

Swelling behaviors in a fuel assembly for the wrapping wire and duct made of modified 316 austenitic stainless steel

Yamagata, Ichiro; Akasaka, Naoaki

Journal of Nuclear Science and Technology, 47(10), p.898 - 907, 2010/10

https://doi.org/10.3327/jnst.47.898

The swelling behaviors in wrapping wire and duct were investigated for a fuel assembly made of modified type 316 austenitic stainless steel irradiated in a fast breeder reactor. The temperature dependence of swelling varied because the peak temperatures of swelling in the wrapping wire and the duct were different. The void distribution in the material was observed by scanning electron microscopy and transmission electron microscopy and it was confirmed that the voids grew within an area of about 100 m from the surface. This phenomenon seemed to be caused by a surface effect on the neutron-irradiated materials.

Effect of high dose/high temperature irradiation on the microstructure of heat resistant 11Cr ferritic/martensitic steels

Yamashita, Shinichiro; Yano, Yasuhide; Tachi, Yoshiaki; Akasaka, Naoaki

Journal of Nuclear Materials, 386-388, p.135 - 139, 2009/04

https://doi.org/10.1016/j.jnucmat.2008.12.082

The heat resistant 11Cr ferritic/martensitic steels were irradiated at 400-670 C up to 100 dpa in FFTF and JOYO. The microstructures of unirradiated 11Cr ferritic/martensitic steels consist of laths, dislocation, and carbide. Almost of the prior austenitic boundaries (PABs) were partially decorated with carbides. It was observed from the results of post irradiation microstructural examinations that the irradiation-induced microstructures were classified into the following three types depending on irradiation temperature; (1) When irradiated at 400-450 C, both dislocation loops and cavities with less than 30 nm in diameter were formed in the ferrite phase. On the other hand, the void swelling was about 0.05%. (2) In the case of irradiation at moderate temperature (500-600 C), the precipitates formation MC carbide was primarily dominated. It was a most noticeable microstructural feature that the carbides; MC and MC grew and covered the PABs at this temperature range. (3) Finally, when irradiation temperature was above 650 C microstructures were drastically-changed. Microstructural observations revealed that formation and growth of equi-axial grain occurred in addition to recovery of laths, growth of carbides simultaneously at high temperature. This remarkable microstructural change might be closely related to a severe degradation in the mechanical properties.

Effects of microstructural evolution on mechanical properties of 11Cr ferritic/martensitic steel after neutron irradiation

Yano, Yasuhide; Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Akasaka, Naoaki; Takahashi, Heishichiro

Materia, 47(12), P. 625, 2008/12

no abstracts in English

Structure of nano-size oxides in ODS steels and its stability under electron irradiation

Oka, Keiichiro*; Onuki, Somei*; Yamashita, Shinichiro; Akasaka, Naoaki; Otsuka, Satoshi; Tanigawa, Hiroyasu

Materials Transactions, 48(10), p.2563 - 2566, 2007/10

https://doi.org/10.2320/matertrans.MD200715

For understanding the microstructural details of nano-size oxide particles, three types of ODS ferritic and austenitic steels were examined by high voltage electron microscopy, EDS and AP-FIM. The oxide included Y, Ti and O and showed a shell-like structure with different composition. The shell-like structure depends on crystal structure of the matrix during fabrication process. To evaluate the irradiation stability of the oxide particles, the electron irradiation was carried out to 47 dpa in the temperature range between room temperature and 923 K. During the irradiation, the oxide particles did not show obvious change in size. The irradiation behavior is discussed comparing with the results recently reported.

Neutron irradiation effects on BC and recovery by annealing

Donomae, Takako; Tachi, Yoshiaki; Sekine, Manabu*; Morohashi, Yuko; Akasaka, Naoaki; Onose, Shoji

Journal of the Ceramic Society of Japan, 115(1345), p.551 - 555, 2007/09

https://doi.org/10.2109/jcersj2.115.551

Use of moderator materials in Fast Breeder Reactor (FBR) is effective for transmutation technology, and BC is one of the candidates. Up to now, the behavior of BC as the Control rod material is well known, but that of BC is hardly investigated. In this paper, the radiation effects of BC pellets, neutron irradiated in the experimental fast reactor JOYO were studied. From the experimental results, it was observed that no macro-cracks were recognized in the irradiated BC pellets. But, bubble nucleation was found in grain and along grain boundaries of BC. And, it was shown that the conductivity of BC was higher than that of BC. During the annealing from room temperature to 1400C, three recovery stages were found on thermal conductivity. It was suggested that, the recovery of BC was related to the dispersion behavior of helium. Judging from these results, as BC was mechanically more stable compared with BC under irradiation, it was shown that BC had high applicability for a moderator.

Microstructural development of a heavily neutron-irradiated ODS ferritic steel (MA957) at elevated temperature

Yamashita, Shinichiro; Akasaka, Naoaki; Ukai, Shigeharu; Onuki, Somei*

Journal of Nuclear Materials, 367-370(1), p.202 - 207, 2007/08

https://doi.org/10.1016/j.jnucmat.2007.03.145

Microstructural observation was done on a neutron-irradiated oxide dispersion strengthened (ODS) ferritic steel, MA957. Since MA957 has been investigated from various viewpoints, special emphases in this study were laid on oxide behaviors including phase stability under irradiation at elevated temperature (973 K). Transmission electron microscopy (TEM) observation of the Y-Ti complex oxide particles showed they were fine (40 nm) whereas the Ti-oxide particles were relatively coarse (300 nm). Dispersion parameters of oxide particles, such as mean size and number density, changed due to irradiation. This fact implies that the recoil resolution of the oxide particles. When irradiated at 973 K, some Y-Ti complex oxides were surviving and interacted with the dislocation structures, which delayed the dislocation recovery and consequently stabilized the elongated grain structures. This is the first evidence showing that oxide particles are effectively functioning as pinning points of dislocations in motion under irradiation to a dose of 100 dpa.

Interaction among dislocation and complex oxide particles in ODS steels heavily-irradiated at high temperature

Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji

Zairyo Kaihatsu No Tameno Kenbikyoho To Oyo Shashinshu, P. 133, 2006/03

Oxide Dispersion Strengthened (ODS) ferritic steel dealt with this study was a MA957 (Fe-0.015C-14Cr-0.3Mo-1.0Ti-0.25Y2O3). The objectives of this study were to understand oxide particle stability of ODS steel during irradiation and interaction among dislocation and oxide particles, reflecting to advanced nuclear reactor design of next generation. Development of some nuclear energy generating systems has been proposed and supported intensively under several international collaboration programs (Generation IV International Forum (GIF), Advanced Fuel Cycle Initiative (AFCI), International Nuclear Energy Research Initiative (I-NERI) etc).Current research issue on ODS ferritic steels is considered to be poverty of experience and understanding on their practical neutron-irradiation behaviors at the temperature higher than 600C.In this research, a MA957, most familiar but primitive 14CrODS ferritic steel contained the highly textured-anisotropic grain structures, was irradiated at 500-700C to fast fluences ranging from 19.8 to 20.8 1026 n/m2 (E 0.1MeV) in the experimental fast reactor JOYO. The dose achieved varied from 99 to 104 dpa. TEM observation and micro-hardness measurement were carried out to clarify the irradiation effects on microstructural evolution of 14CrODS ferritic steel at elevated temperature and high dose. Microstructural examination revealed that all of the highly textured- anisotropic grain structures, following heavy irradiation at the temperature above 600C, have not changed. In addition, large regions in all specimens have retained high dislocation density, contained negligible cavitation.

Mechanical behavior of oxide dispersion strengthened steels irradiated in JOYO

Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Akasaka, Naoaki; Ukai, Shigeharu; Kimura, Akihiko*

Materials Transactions, 46(3), p.493 - 497, 2005/05

https://doi.org/10.2320/matertrans.46.493

Oxide dispersion strengthened (ODS) steels, which are candidate materials for water-cooled solid breeder blankets, were fabricated with several manufacturing parameters, and then irradiated in JOYO to evaluate their irradiation behavior. Engineering stress strain curve of ODS steels irradiated at 673 K exhibited superior material response, i.e., increased tensile strength due to irradiation hardening and no loss of total elongation. Also, their temperature dependence on tensile property indicated that degradation of the tensile property at elevated temperature, which is closely related to phase stability during irradiation, could be avoided due to optimal combination of manufacturing parameters, such as chemical composition, type of inert gas during mechanical alloying, heat-treatment temperature and initial phase of the matrix.

Evaluation of Radiation Embrittlement by Charpy Impact Tests with Miniaturized Specimens

Kurishita, Hiroaki*; Yamamoto, Takuya*; Narui, Minoru*; Yoshitake, Tsunemitsu; Akasaka, Naoaki

JNC TY9400 2004-006, 48 Pages, 2004/04

JNC-TY9400-2004-006.pdf:1.95MB

Radiation embrittlement in high-strength ferritic/martensitic steels of 2WFK and 63WFS and oxide dispersion strengthened (ODS) martensitic steel of H-35 that were irradiated in the experimental fast reactor JOYO is evaluated by instrumented Charpy impact tests for miniaturized (1.5 x 1.5 x 20 mm) and half-sized Charpy V-notch (CVN) specimens. Effects of thermal aging and microstructural evolution during irradiation on radiation embrittlement are described. Next, in order to clarify the specimen size effects on the ductile-to-brittle transition temperature (DBTT) in Charpy impact testing, a method to evaluate the plastic constraint loss for differently sized CVN specimens that may be responsible for the size effects is proposed and applied to 2 WFK. In the method, the constraint factor, a, that is an index of the plastic constraint is defined as a = s*/sy*. Here, s* is the critical cleavage fracture stress which is a material constant and sy* is the uniaxial yield stress at the DBTT at the strain rate generated in the Charpy impact test. The procedures for evaluating each of s* and sy* are described and the result of s* and sy*, thus the value of a, is presented for various types of miniaturized and full-sized CVN specimens of 2 WFK. It is shown that there is the following relationship between a and the specimen size factor, (A*/b2). a=a0-k(A*/b2)0.4 Here, A* is the critical area for cleavage fracture and b is the ligament size. a0 and k are constants depending on a /W (a is the notch depth and W is the specimen width). a increases with increasing a /W.