Aging management is one of the main issues to keep the safety of currently operating light-water reactors. In many cases plant aging is caused by material’s degradation due to energetic particle irradiation, high temperature, high pressure, etc. In order to keep the safety of light-water reactors, material’s degradation is accurately inspected, and then, based on the results thus obtained the appropriate plant design and timely maintenance should be made. To do so, we should develop the methodology to predict the behavior of materials under severe environments. In the present study, the theoretical and computational investigation was done to understand the mechanism of oxidation process of fuel cladding, which is one of the main degradation processes. Our calculation results have indicated that a difference between the oxidation rates of Zr at normal operating temperature and high temperature may be explained by the stress effect on the diffusion process of an oxygen atom in the oxide film.

A recently developed nondestructive method, called Magnetic Adaptive Testing, which is based on systematic measurement and evaluation of minor magnetic hysteresis loops was applied for detection of local material thinning in a thick, L-shaped ferromagnetic carbon steel, which simulates T-tubes with reinforcing plates. Artificially made slots were reliably detected with a good signal/noise ratio from the other side of the specimen, even through a covering ferromagnetic plate.

The contribution reports to R&D performed under the umbrella of the German Nuclear Safety Research Program in the last decade by the Fraunhofer-IZFP within the context of nondestructive materials characterization of materials ageing phenomena which can occur during the lifetime of a nuclear power plant. The Cu-rich low-alloy heat-resistant steel 15 NiCuMoNb 5 (WB 36, material number 1.6368) which is used as piping and vessel material in boiling water reactor (BWR) and pressurized water reactor (PWR) nuclear power plants in Germany was under special investigations. In all damage situations observed in plants, the operating temperature was between 320℃ and 350℃ and the damage was combined with a shift in the ductile-to-brittle-transition-temperature (ΔDBTT) of the notched-bar impact test to higher temperatures. Micromagnetic NDT techniques were applied to characterize the material degradation which is enhanced when low cycle fatigue loads are superimposed to the precipitation process of Cu-rich particles. As the essential results correlations were found between micromagnetic properties on one hand and microhardness and the Cu-content on the other hand.
As Cu-rich-precipitates play also an essential role in neutron embrittlement of pressure vessel material, the same NDT techniques were applied to predict ΔDBTT at irradiated Charpy specimen. Also here high correlations were found between the micromagnetic properties and the ΔDBTT with small residual standard error. Austenitic stainless steel AISI 321 is used in German plants as surge- and spray-pipeline material which under service is exposed to cyclic thermo-mechanical loads. The contribution documents the ability of ultrasonic time-of-flight measurements to sensitively characterize the whole fatigue life from the beginning to the end of life. As the experiments were performed at 300°C too, EMAT-transducers have been applied.

Electromagnetic acoustic resonance (EMAR) provides accurate and stable evaluation. Its capability has been demonstrated through online monitoring using a large-scale corrosion test loop operating at high temperature. This study uses EMAR to evaluate the thickness of pipes in a nuclear power plant during its shutdown through signal processing based on superposition of nth compression. Sections of piping evaluated with EMAR include those in long-term service, where thinning may produce scale-like surfaces, and those having complicated geometry. Moreover, we compare measurement results obtained with EMAR and with ultrasonic testing (UT). The accuracy of EMAR depends on the pipe geometry, such as the pipe diameter and whether the pipe is straight or an elbow, the presence of welding, and complicated wall thinning. We consider the causes of the difference in thickness values between EMAR measurements and UT. Finally, we discuss how to implement EMAR in pipe inspection.

Since it was reported that Flow-accelerated corrosion was largely influenced by a swirling flow, it is important to clarify the characteristics of swirling flows formed in pipings. In this study, a flow experiment with 2D-PIV measurement was conducted in order to evaluate the swirling flow generated downstream of a three-dimensionally-connected dual elbow which comprises two 90-degree elbows directly connecting to each other. The ratio of the elbow curvature radius to a pipe inner diameter,Υ , was varied as an important parameter, and three types of elbow, Υ = 1.0, 1.5, 2.0, were tested. The Reynolds number was set at 300,000 and the inlet flow condition was fully developed turbulent flow. Experimental results showed that the swirl number of the swirling flow generated in the dual elbow became high and fluctuated largely as the cross-section where the swirl number was evaluated moved downstream of the elbows when Υ was small. In addition, high velocity secondary flows were likely to appear locally downstream of the elbows when Υ was small. In the parallel cross-section to the pipe axis, the high velocity area formed not in an island-like shape but in a belt-like one. The locally high velocity component in the circumferential and the axial direction implied formation of a spiral flow downstream of the dual elbow. And the spiral flow became strong when Υ = 1.0. Moreover the swirl numbers at the outlet of the second elbow were approximately same as that 8D downstream of the dual elbow in all of the cases, and the swirling flow was less likely to decay compare to the secondary flow generated in the single elbow and the dual elbow with a 2D-straight part between the elbows.

A simplified technique of obtaining the shape of inclined defect at welded zone of high chromium steel was proposed. Magnetic flux leakage measurement and the inverse analysis using singular value decomposition were used. From the magnetic flux density distribution obtained under the high magnetic field, adequate defect shape was obtained without the influence of welded zone. Under the low magnetic field, adequate defect shape was obtained by subtracting the magnetic field of welded zone from measured magnetic flux distribution.

Guided wave (GW) inspection systems offer the capability for online health monitoring of structures and hence the potential for transitioning from schedule-based maintenance to conditioned-based maintenance of nuclear plants. However, a major problem in guided wave NDE is the presence of multiple propagation modes at any actuation frequency that travel with different velocities and interfere with each other which makes interpretation of GW signals complicated. Decomposition of the measured signal into its constituent components is a critical requirement for accurate analysis in terms of detection, location and characterization of structural defects. This paper addresses the above problem using a new two-stage Mode Decomposition (MD) algorithm. The first stage employs fast high-resolution Time-Frequency Representation (TFR) of the signal based on Reassigned Spectrogram with Chirp Transform kernel (RSCT), which is used for preprocessing. The second stage includes efficiently implemented Matching Pursuit with dispersion-based dictionary. The performance of the algorithm is demonstrated on 2-mm thick aluminum plates with surface-bonded piezoelectric wafer sensors. The results show that the algorithm identifies the modes and separates closely spaced or overlapped S0 and A0 wave packets. Automated analysis of the MD output allows for detection of flaw indications corresponding to different modes using standard diagnostic imaging algorithm.

After Fukushima Accident (March 11, 2011), the Atomic Energy Society of Japan (AESJ) started to develop the standard of Tsunami Probabilistic Risk Assessment (PRA) for nuclear power plants in May 2011. As Japan is one of the countries with frequent earthquakes, a great deal of efforts has been made in the field of seismic research since the early stage. To our regret, the PRA procedures guide for tsunami has not yet been developed although the importance is held in mind of the PRA community. Accordingly, AESJ established a standard to specify the standardized procedure for tsunami PRA considering the results of investigation into the concept, the requirements that should have and the concrete methods regarding tsunami PRA referring the opinions of experts in the associated fields in December 2011 (AESJ-SC-RK004:2011).

Current regulatory testing and inspection requirements are reviewed and a summary of degradation experience is presented. Techniques commonly used to inspect NPP concrete structures to assess and quantify age-related degradation are summarized. An approach for conduct of condition assessments of structures in NPPs is presented. Criteria, based primarily on visual indications, are provided for use in classification and assessment of concrete degradation. Materials and techniques for repair of degraded structures are generally discussed.

In this paper, method on the durability evaluation in nuclear power plant concrete structures was investigated. In view of the importance of evaluating the degree of deterioration of reinforced concrete structures, relationships should be formulated among the number of years elapsed, t, the amount of action of a deteriorative factor, F, the degree of material deterioration, D, and the performance of the structure, P. Evaluation by PDFt diagrams combining these relationships may be effective. A detailed procedure of durability evaluation for a reinforced concrete structure using PDFt concept is presented for the deterioration of rebar corrosion caused by neutralization and penetration of salinity by referring to the recent papers.