The need to reliably perform a fatigue life assessment is obvious, if we have the ageing infrastructure in our power plants in view. Concerning components manufactured in austenitic stainless steel, in Germany, in PWR, we do not have in the primary circuit, as main cooling line (MCL), an austenitic component. The German design is based on pipes, made from the same steel grade as the reactor pressure vessel (RPV) – but cladded on the inner side by austenitic stainless steel. Therefore, Germany, for instance, has not the NDT inspection problems, other countries have, with the dissimilar metal weld between RPV nozzle and MCL. However, the surge line and the pressurizer spray line are fully made from austenitic stainless steel and they are exposed to thermomechanical fatigue. Concerning their fatigue behavior over lifetime of a NPP, many open questions still exist. The paper describes fatigue experiments, monitoring the material in the servo-hydraulic fatigue machine and the development of the PHYBAL procedure for physically based life evaluation. This new technology, in contrast to the standardized SN-Curve evaluation, only asks for a small number of specimens. In case of a homogeneous material, only three tests, 1 incremental step test and 2 constant amplitude tests, are required for the estimation of a complete S-N-fatigue curve by calculation.

In March 2011, the Fukushima Nuclear Power Plant suffered a severe accident because of the tsunami waves when the Great East Japan Earthquake occurred and because of this accident the containment vessel was exposed to high temperatures. However, studies related to concrete with extremely high temperature exposure are very few. For this reason, the aim of this study is to understand the structural soundness of the concrete structures exposed to high temperature by conducting heating experiments of concrete specimens which were made to reproduce the concrete structure of the Nuclear Plant. From the test results, cracks were observed at the specimen surface and they grew with the increase of temperature. Additionally, it was observed that concrete crack volume grew to temperatures around 800°C, but this growth saturated more than this temperature to 1000°C. It was observed that the amount of water content inside concrete decreased with temperature. Moreover, the chemical composition change of concrete was accompanied with the water content decrease. Results on other material properties are also to be presented.

In order to ensure safe and stable Nuclear Power Plant (NPP) operation, effective ageing management (AM) program and implementation of the program is necessary. For this purpose, the “Code on Implementation and Review of Nuclear Power Plant Ageing Management Programs”, which was prepared by the Atomic Energy Society of Japan (AESJ), is used in Japan. The Code has been endorsed and used in the regulation by Japanese regulatory body.  Since the first edition was issued in 2007, continuous effort for improvement was made, and the second revision was issued quite recently in 2015. 
The Code has the following features;
- Specify ageing management actions according to plant operational years from the early stage of plant operation.
- Utilize “Summary Sheet of Ageing Phenomena”, which has been made based on the results of aging management technical evaluation and the actual operation experience of Nuclear Power Plants (NPPs) in Japan.
-  8 ageing phenomena which should be selected and subject to AM technical evaluation are listed, and the evaluation procedures including the seismic safety evaluation are enacted in the Attachment as a part of the Code.
In the revision, not only new information and data are added reflecting the further operating experience and the latest knowledge including IAEA IGALL, but also the evaluation procedures and “Summary Sheet of Ageing Phenomena” are enriched in terms of seismic effects. 
The approach to effective AM program and their implementation are explained and discussed.

The main coolant line (MCL) is a part of the primary coolant boundary and it is the key factor for the reactor safety. In order to guarantee the integrity of the primary circuit, full volumetric ultrasonic examination is required for the MCL welds of the European Pressurized Reactor (EPR) according to the RSE-M code 2010 Version. The difficulties for application of the ultrasonic inspection on the MCL welds are mainly caused by limited access space and coarse austenitic grain structure. In this study, inspection system including automated scanner and transmitter/receiver longitudinal wave technique has been developed in consideration of the multiple constraints. The qualification process of the ultrasonic examination on the test mock-ups including flaw sizing results is also presented and commented.

With the reserve dwindling of coal, oil and other fossil energy, nuclear power becomes more and more important in recent years. As one of the important part in the nuclear power plant (NPP), the transit pipeline of cooling water plays a key role in the safe operation of NPP. The pipeline used in NPP not only requires excellent mechanical properties, corrosion resistance and wear resistance, but also needs great anti-aging performance. In this paper, the pipeline used in NPP is briefly introduced. Some newly emerged composite pipes, such as plastic lined metal composite pipe, ceramic-lined metal composite pipe, are discussed on the possibility applied in NPP. Furthermore, based on the tape casting process and centrifugal-thermite process, a simple process for fabricating the glass coating steel pipe, which is called as Slurry-Centrifugal-Casting, is proposed.