Synthesis of a Si-Al gel as a starting material of aluminosilicate solids

Sato, Junya; Shiota, Kenji*; Takaoka, Masaki*

Zairyo, 70(5), p.406 - 411, 2021/05

https://doi.org/10.2472/jsms.70.406

An aluminosilicate solid is an inorganic material that has the property of immobilizing heavy metals or radionuclides in the matrix. In this study, aluminosilicates with a Si/Al molar ratio of 0.5 was synthesized from a chemical reagent in order to produce aluminosilicate solids with a low Si/Al molar ratio, which were expected to improve the immobilization of heavy metals and radionuclides contained in the matrix. The synthesized Si-Al gel with a Si/Al molar ratio of 0.5 had little impurity content and was in an amorphous phase. In addition, the compressive strength of the aluminosilicate solid produced by the synthesized Si-Al gel showed a 5 MPa or more, confirming that it can be used as a raw material for aluminosilicate solids. The aluminosilicate solid with a Si/Al molar ratio of 1.25 had a dense surface structure from the result of BSE images and had the highest compressive strength among all samples.

Evaluation of breach characteristics of fast reactor fuel pins during steady state irradiation

Oka, Hiroshi*; Kaito, Takeji; Ikusawa, Yoshihisa; Otsuka, Satoshi

Nuclear Engineering and Design, 370, p.110894_1 - 110894_8, 2020/12

https://doi.org/10.1016/j.nucengdes.2020.110894

The objective of this study is to evaluate the reliability of a cumulative damage fraction (CDF) analysis for the prediction of fuel pin breach in fast rector using experimentally obtained fuel pin breach data for the first time. Six breached fuel pins were obtained from steady state irradiation in the EBR-II. Post irradiation examinations revealed that FP gas pressure was the main cause of creep damage in cladding, and that the stress contribution from FCMI was negligible. CDFs evaluated for these pins using in-reactor creep rupture equation, taking into account the irradiation history of cladding temperature and hoop stress due to FP gas pressure, were in the range of 0.7 to 1.4 at the occurrence of breach. This shows clearly that fuel pin breach occurs when the CDF approaches 1.0. The results indicate that CDF analysis would be a reliable method for the prediction of fuel pin breach when appropriate material strength and environmental effects are adopted.

Technical basis of accident tolerant fuel updated under a Japanese R&D project

Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Nozawa, Takashi; Watanabe, Seiichi*; Kirimura, Kazuki*; Kakiuchi, Kazuo*; Kondo, Takao*; Sakamoto, Kan*; Kusagaya, Kazuyuki*; et al.

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

In Japan, the research and development (R&D) project on accident tolerant fuel and other components (ATFs) of light water reactors (LWRs) has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys (Zircaloy) in LWRs. ATF candidate materials under consideration in the project are FeCrAl steel strengthened by dispersion of fine oxide particles(FeCrAl-ODS) and silicon carbide (SiC) composite, and are expecting to endure severe accident conditions in the reactor core for a longer period of time than the Zircaloy while maintaining or improving fuel performance during normal operations. In this paper, the progresses of the R&D project are reported.

FEMAXI-7 prediction of the behavior of BWR-type accident tolerant fuel rod with FeCrAl-ODS steel cladding in normal condition

Yamaji, Akifumi*; Yamasaki, Daiki*; Okada, Tomoya*; Sakamoto, Kan*; Yamashita, Shinichiro

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

http://www.wrfpm2017.org/upload/draft_file/F-043-PD4.pdf

Features of the accident tolerant fuel performance were evaluated with FEMAXI-7 when the current Zircaloy(Zry) cladding is replaced with FeCrAl-ODS steel cladding (a type of oxide dispersion strengthened steel being developed under the Project on Development of Technical Basis for Safety Improvement at Nuclear Power Plants by Ministry of Economy, Trade and Industry (METI) of Japan) for BWR 99 type fuel rod. In particular, influences of the creep strain rate and thickness of the ODS cladding on the fuel temperature, fission gas release rate (FGR) and pellet-cladding mechanical interaction (PCMI) are investigated.

Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

Kasahara, Shigeki; Kitsunai, Yuji*; Chimi, Yasuhiro; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka

Journal of Nuclear Materials, 480, p.386 - 392, 2016/11

https://doi.org/10.1016/j.jnucmat.2016.06.017

This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. Tensile tests at 290C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Influence of difference in the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. The influence was also found certainly in loss of strain hardening capacity and ductility, although the influence on the yield strength and the Vickers hardness was not clearly observed. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, were considered to contribute to deformation of the austenitic stainless steel.

Developments of high-performance moderator vessel for JRR-3 cold neutron source

Arai, Masaji; Tamura, Itaru; Hazawa, Tomoya

JAEA-Technology 2015-010, 52 Pages, 2015/05

JAEA-Technology-2015-010.pdf:7.11MB

In the Department of Research Reactor and Tandem Accelerator, developments of high-performance CNS moderator vessel that can produce cold neutron intensity about two times higher compared to the existing vessel have been performed in the second medium term plans. We compiled this report about the technological development to solve several problems with the design and manufacture of new vessel. In the present study, design strength evaluation, mockup test, simulation for thermo-fluid dynamics of the liquid hydrogen and strength evaluation of the different-material-bonding were studied. By these evaluation results, we verified that the developed new vessel can be applied to CNS moderator vessel of JRR-3.

Mechanical properties of weldments using irradiated stainless steel welded by the laser method for ITER blanket replacement

Yamada, Hirokazu*; Kawamura, Hiroshi; Nagao, Yoshiharu; Takada, Fumiki; Kono, Wataru*

Journal of Nuclear Materials, 355(1-3), p.119 - 123, 2006/09

https://doi.org/10.1016/j.jnucmat.2006.04.010

In this study, the bending properties of welding joint of irradiated material and un-irradiated material (irradiated/un-irradiated joints) were investigated using SS316LN-IG, which is the candidate material for the cooling pipe of ITER. The results of this study showed that the bending position of joints using un-irradiated material was un-irradiated part and that the bending position of irradiated/irradiated joints was fusion area or HAZ (heat affected zone). Although the bending position of joints was different bor the combination pattern between irradiated and un-irradiated materials, the bending strength of joint was almost same. Additionally, it is confirmed that bending strength did not depend on the combination pattern between the irradiated and un-irradiated materials, nor on the relationship between the heat input direction and the bending load direction.

Anisotropic deformation effect on the fracture of core components made of two-dimensional C/C composite

Sumita, Junya; Shibata, Taiju; Ishihara, Masahiro; Iyoku, Tatsuo; Tsuji, Nobumasa*

Key Engineering Materials, 297-300, p.143 - 147, 2005/11

no abstracts in English

Residual stress evaluation of butt weld sample of high tensile strength steel using neutron diffraction

Suzuki, Hiroshi; Holden, T. M.*; Moriai, Atsushi; Minakawa, Nobuaki*; Morii, Yukio

Zairyo, 54(7), p.685 - 691, 2005/07

https://doi.org/10.2472/jsms.54.685

no abstracts in English

Tensile and fatigue strength of free-standing CVD diamond

Davies, A. R.*; Field, J. E.*; Takahashi, Koji; Hada, Kazuhiko

Diamond and Related Materials, 14(1), p.6 - 10, 2005/01

https://doi.org/10.1016/j.diamond.2004.06.015

A CVD diamond is finding increased application and it is important to study its fatigue properties. The present paper describes research on a batch of di-electric grade CVD material. It was obtained that tensile strength at the nucleation side and the growth were side 69090MPa and 28030MPa, respectively. Some samples survived at least 95% of their critical fracture stress for 10 cycles without fatiguing.