ICMST-Tohoku 2018
Oct. 23 - 26, 2018
Sendai, Japan
ICMST-Shenzhen 2016
Nov 1 - 4, 2016
Shenzhen, China
ICMST-Kobe 2014
Nov 2(Sun) - 5(Wed), 2014
Kobe, Japan
Nuclear Regulation Authority Outline of the New Safety Standards for Light Water Reactors for Electric Power Generation
For Public Comment
Outline of New Safety Standard (Design Basis)
For Public Comment
New Safety Standards (SA) Outline (Draft)
For Public Comment
Outline of New Safety Standard(Earthquake and Tsunami)(DRAFT)

Vol.10 No.2(Aug)
Vol.10 No.1(May)
Vol.9 No.4(Feb)
Vol.9 No.3(Nov)

< Other Issues


Occasional Topics
OTjapan Measures for Tsunami Striking Nuclear Power Station in Japan
Special Article: The Great Tohoku Earthquake (1)
OTjapan The Tragedy of “To Be” Principle in the Japanese Nuclear Industry
EJAMOT_CN3_Figure1_The_outside_view_of_CEFR OTChinaPlanning and Consideration on SFR R&D Activities in China
< All Occasional Topics

Featured Articles
EJAM7-3NT72 A New Mechanical Condition-based Maintenance Technology Using Instrumented Indentation Technique
EJAM7-3NT73 Survey robots for Fukushima Daiichi Nuclear Power Plant

(in English)

Vol.5 No.2previous AA68-SP10(69-70-71)-NT 58 -59
Academic Articles
Vol.5, No.3(2013) p.165 - p.200

Special Issue 10

Materials Degradation and Damage in Nuclear Power Plants

Relevant Field
Ageing Management, Plant Maintenance
Nuclear Power Plant, Material, Materials Degradation, Damage, Alloy 690, Stainless Steel, Duplex Grain Size, Oxidation, Low-Cycle Fatigue, Crack Growth, Stress Corrosion Cracking
  In recent years, aging management and maintenance of nuclear power plant have been one of the most important issues because of the increase in number of highly aged power plants, which have been operated for more than 30 years. We should continue to work on this issue for the future re-operation of nuclear power plant, although almost all the plant has been shut down since the catastrophic earthquake. It is required for this aging management to foresee the potential materials degradation and/or damage exactly and fully understand their mechanisms in advance. Additionally, needless to say, it is also essential to develop an adequate maintenance program based on the foresight, that is, the inspection technology for detecting and monitoring the materials aging degradation and/or damage with high accuracy at the early stage. In this special issue, key papers for the potential aging materials degradation and/or damage such as a duplex grain structure, hydrogen accelerated oxidation and low-cycle fatigue damage have been invited.

Guest Editor,
Shin-ichi Komazaki

Toshiaki HORIUCHI, Naohiro SATOH

An experimental characterization of grain boundaries using field-emission Auger electron spectroscopy has been carried out for Ni2M-stabilized alloy (where M mainly corresponds to Cr) and Alloy 690, which have duplex and non-duplex grain sizes, respectively, in order to determine the relationship between grain-boundary precipitates and the grain structure. Thermodynamic calculations based on the Scheil-Gulliver model with and without back diffusion of the C solute in the solid phase were also performed in order to investigate the solidification process in both alloys. Chromium carbide precipitates, with a predicted composition of M23C6, were observed at grain boundaries in both the Ni2M-stabilized alloy and Alloy 690. The M23C6 precipitates in the Ni2M-stabilized alloy wer considerably coarser than those in the Alloy 690. A small number of coarse titanium carbonitride precipitates were also observed on the fracture surfaces of both alloys at intergranular and intragranular positions. The simulations predicted that the M23C6 precipitates are likely to be formed during the final stages of solidification, and it is thought that this occurs more readily in the Ni2M-stabilized alloy. The results indicate that the duplex grain structure observed in the Ni2M-stabilized alloy is most likely due to the presence of undissolved coarse M23C6 grain-boundary precipitates.

site PDF

Motoki NAKAJIMA, Shin-ichi KOMAZAKI, Tetsuo SHOJI

The influence of hydrogen in steel on the high temperature water oxidation behavior of low carbon austenitic stainless steel Type F316L was investigated in order to clarify the mechanism of SCC initiation in the BWR environment, which is closely associated with the localized oxidation and its acceleration. The steel was charged with hydrogen by means of cathodic electrolysis. And then, the solution-treated (non-charged) and hydrogen-charged steels were subjected to the oxidation test in simulated BWR environment. Experimental results revealed that the size of outer oxide particle increased with increasing hydrogen content, resulting in the hydrogen accelerated oxidation (HAO). Additionally, the oxide of the hydrogen-charged steel was mainly NiFe2O4, whereas Fe3O4was predominantly formed on the non-charged one. From the result of the small punch test in the BWR environment, it was also indicated that the effect of hydrogen on the oxidation might be almost equivalent to that of applied stress.

site PDF

Masayuki KAMAYA

Quantifying the low-cycle fatigue damage accumulated in nuclear power plant components is one of the important issues for aged plants. In this study, detailed observations of crack initiation and growth were made using scanning electron microscopy in order to correlate the crack size and the magnitude of the fatigue damage. Type 316 stainless steel specimens were subjected to the strain-controlled axial fatigue test (strain range: 1.2%) in air at room temperature. The test was interrupted several times in order to observe the specimen surface. The spatial distribution of inhomogeneously accumulated damage by cyclic loading was identified by crystal orientation measurements using the electron backscatter diffraction technique. Cracks were initiated at grain boundaries and slip steps, where relatively large damage accumulated. The changes in the number of cracks and their length were quantified. The crack growth rates were well correlated with the strain intensity factor. The change in crack size during the fatigue test was predicted using the obtained growth rate and assumed initial crack size. The fatigue lives estimated by the crack growth prediction agreed well with those obtained experimentally. It was concluded that the fatigue damage could be estimated from the crack size measured in plant components.

EJAM Vol.5 p.165-200 Academic Articles Special Issue on "Materials Degradation and Damage in Nuclear Power Plant "