With the higher safety requirements and the rising public concern of reactor safety, ageing management of nuclear power plants are facing new challenges in recent years. Multiscale modeling is the tool for ageing prediction during and beyond reactor design lifetime. It is important to update the latest predictions to existing licensed reactors continuously, however, experimental data at microscopic scale is lacking for the validation and calibration of modeling results. Atom probe tomography (APT) is a new and powerful analyzing tool with near-atomic resolution. It is good at studying issues like second-phase precipitation and grain boundary cracking, and its 3D atom map is highly suitable for the comparison with modeling. APT is promising for future reactor ageing management and maintainology, but only a small fraction of researchers can utilize the new technique currently. In this work, training for potential APT users is proposed and designed, so that more researchers could benefit from APT technique in the future. The advantages and the basic mechanisms of APT are introduced. The possible errors that are most important to APT users and modeling researchers are explained. A practice section is also designed, which enables the participants to experience the visualization of atom maps and the tuning of precipitate definition. The concept, structure and experience of this training could be a good starting point for the design and organization of better APT training in the future.

Atom probe tomography is a promising tool for future ageing management of nuclear power plants. For a wider application of atom probe technique in the future, a training program for university students and young researchers has been proposed, and a practice section has been designed to provide chances to experience the data analysis of atom probe. However, the available commercial software for the data analysis is expensive and lacks enough flexibility. In this work, a data analysis solution for atom probe with free codes and free software has been set up. The requirement of atom map visualization has been discussed, and the atom map data has been successfully visualized with the open-source software ParaView. A precipitate searching code has been developed based on the maximum separation method, which is the most popular method currently. The precipitate searching code consists of three steps: maximum separation step, envelope step and erosion step. The detailed algorithm of the code has been discussed, and the effects of tuning the precipitate searching parameters in the code have been demonstrated. With this set of tools, the training participants could get a vivid picture of the advantages and limitations of the atom probe technique.

The aim of this work presented here is to investigate the applicability of ultrasonic guided wave for evaluation of laser beam butt-welding quality. First, ten in total test pipes having welding seam are prepared. Two piece of pipe are jointed and continuous laser beam is irradiated on the edges, varying laser irradiation power, welding side and surface profile of the adjacent edges of the pipe. Second, ultrasonic guided wave testing experiment is performed on the pipes. Torsional mode guided wave is excited by Electromagnetic Acoustic Transducers. Finally, the experimental results are analyzed and issues are discussed. The reflection wave bullet from the poor interface of the welding seam is clearly observed, whereas no reflection from fine welded line. From the aspect of laser irradiation power, welding side and surface profile of the adjacent edges, the relation between the interface condition and detection wave bullet are analyzed. It is found that the ultrasonic guided wave technologies have the potential for evaluating laser beam butt-welding seam.

Fiber Bragg Gratings (FBG) were fabricated by use of point by point method using precisely focused picosecond laser pulses, The FBGs can be the best candidate for strain gauge for high temperature industrial plants. For the best use of heat resistant characteristic, we demonstrated to embed the FBG sensors in metal mold to an elbow in the sodium loop. Metal adhesive is made of fine silver particles. This adhesive is applicable at very high temperature, having a high thermal conductivity and being activated in strength at 200℃. A sodium loop was constructed to develop maintenance technologies for liquid sodium handling. The main tank was filled with liquid sodium of 6 ton. Sodium was circulated with temperature of ~500℃, the flow speed of 5 m/s. In daily operation, the FBGs successfully detect periodic distortion at the elbow due to the thermal expansion. During emergency cooling, a sudden shrinking of the loop was recorded. We propose to apply FBGs to an advanced remote monitoring system for high temperature industrial plants.

The adherence of sodium on the external surface of steam generator tubes of Fast Breeder Reactors adds an additional level of complexity in the in-service inspection of SGs using eddy current testing. The paper focusses on research and development of novel multi-frequency algorithms, based on eddy current signals, special developed and tuned to enhance the signal from defects located on external tube surface and to suppress electromagnetic noise from the conductive sodium and tube support plates. The application of the multi-frequency algorithm is investigated using both three-dimensional finite element simulations and experimental measurements from SG tubes soaked and then drained of sodium, in a sodium tank mock-up.

After Fukushima Daiichi nuclear power plant accident, the importance of nuclear security is increased, especially as a threat to nuclear power plants, sabotage by insider is significant. In response to the increasing threats to Nuclear Power Plant, human malicious behavior detection is necessary for nuclear security. Hand motion is an important part of human activity and has a high contribution for high-accuracy detection of insider malicious behaviors. Hand motions can be distinguished by the position of each fingertip, both stretched and bend fingers of both left and right hands can be classified as different parts by using depth data and body index frame of Microsoft Kinect v2. Fingers were identified by using K-means clustering algorithm. Finally, it was built a hand motion time series data by using the developed real-time hand motion detection system. However, as malicious behaviors detection isn’t enough for nuclear security, future malicious behaviors prediction should be taken into consideration.
In this research, the real-time hand motion detection system was developed by using Kinect v2. In addition, we explored the possibility of malicious behavior detection and prediction by using Stacked Auto-Encoder.

The neutron irradiation embrittlement of reactor pressure vessel (RPV) steels needs to be properly predicted and managed for the safe long-term operation of light water reactors. To investigate the effects of the solute elements Ni, Mn and Si on the irradiation hardening at high dose levels, pertinent to long-term operation, in RPV materials without Cu, ion irradiation for Fe-1.0Ni-0.2Si, Fe-1.0Ni-1.4Mn and Fe-1.0Ni-1.4Mn-0.2Si alloys was carried out at 290 oC up to 5.0 dpa by Fe-ions with the energy of 2.8 MeV. Nano-indentation techniques were applied to evaluate the ion irradiation hardening. The results show that Mn significantly accelerates the irradiation hardening and addition of Si reduces the irradiation hardening. The depth dependence of irradiation hardening was discussed, and the irradiation hardening was the most significant at the indentation depth of about 120 nm in all alloys. The hardening behavior in deeper region was slightly different from that observed at about 120 nm depth, which may be caused by the interactions between solutes and defects.

Stress corrosion cracking (SCC) occurs in several key structural components of nuclear power plants over time. A sufficiently precise quantitative technique is necessary to guarantee the safety and efficient operation of the key structural components when SCC occurs. Quantitative evaluation of SCC profiles using eddy current testing (ECT) signals often underestimates the crack depth due to the local conductive property and the ill-posedness of the inverse problem. In this paper, a quantitative evaluation method based on features of ECT signals is proposed to improve the sizing accuracy of SCC profiles. The proposed strategy includes a numerical SCC model based on the local conductivity distribution, a hybrid inverse strategy combined with the conjugate gradient method and the particle swarm optimization algorithm, and a quantitative evaluation method using features of ECT signals. Reconstruction is conducted with simulated ECT signals from conductive cracks, and the results show that the proposed strategy is effective in improving the sizing precision of SCC profiles.

Reactor Pressure Vessel (RPV) during Pressurized Thermal Shock (PTS) loading is a critical issue in assessing the safety of nuclear power plants. The most severe situation takes places during Emergency Core Cooling Systems (ECCS) cold water injection in the cold legs due to a Loss-Of-Coolant Accident (LOCA). Conventionally, one-dimensional thermal hydraulic analysis has been performed as well as the simple model fracture mechanics analysis. In the present study, the three-dimensional Computational Fluid Dynamics (CFD) simulation and a comprehensive fracture mechanics analysis are performed. A reference design of a four-loop RPV is applied, and three different cases of the mass flow rates are considered in the analysis. Based on temperature distribution obtained by CFD, the fracture mechanics analysis were carried out to investigate the structural integrity, where submodeling technique is employed. Our results indicate that the worst crack location is identified and the dependence of Stress Intensity Factor (SIF) on the positon of RPV is clarified. It is useful information to inspect and maintain the RPV integrity.

The resistance to environment-assisted cracking (EAC) like SCC (stress corrosion cracking) and aging embrittlement are the most important problems on the materials performance of type 316L steel used in the advanced nuclear power plants. Extra high purity, grade type 25Cr-35Ni EHP, austenitic stainless alloy was developed by means of minimizing impurity using the special melting technology. It has the excellent corrosion resistance in LWR environments and high temperature steam under gamma ray irradiation.
In this research, cladding technology of type 25Cr-35Ni EHP alloy on the base metal of type 316L steel is developed for reactor core materials by the diffusion bonding method using hot rolling. The corrosion resistance was tested by aging up to 600 degrees in Ar under gamma-ray irradiation. The effect of cladding and aging embrittlement of type 316L steel due to sigma phase precipitation is evaluated by Charpy impact tests.
From these results, it was clarified that the resistance to corrosion and aging embrittlement of type 316L steel are possible to improve by including cladding of type 25Cr-35Ni EHP alloy, of a thickness of more than 2 mm . The cladding of EHP alloy is considered to be one of the most widely used materialtechnology for preventing the aging degradation of austenitic stainless steels like type 316L used in current nuclear power plants.

For safety and certain removal of the melted down nuclear fuel debris, the localization of radiation sources should be required. In this paper, we propose an approach to estimate the accurate localization of radiation sources by utilizing a gamma-ray CT method with a detector mounted on a mobile robot that has a pose uncertainty. Gamma-ray CT method aggregates measurement data obtained from the detector to localize radiation sources. For applying gamma-ray CT method, detector’s position that the measurement data obtained is important. In a simulation experiment, we confirmed the detector’s position errors and correlated them with the accuracy of the localization of radiation sources.

Buried pipes in nuclear power plants (NPPs) service in complex environments. Failure in these pipes may result in severe consequences, so that non-destructive testing (NDT) of buried pipes in NPPs is of great importance. Alternating current field measurement (ACFM) method is a potential approach to detect failure in buried pipes due to its advantages such as non-contacting measurement and possible to be applied to long pipes. To evaluate the performance of ACFM for buried pipes, a numerical method and a simulation code are proposed and developed in this paper to calculate the current distribution in a conductive object due to both alternative conduction current injection as well as those induced by the alternative magnetic field due to non-steady current. Comparisons between the current density and related magnetic field results in an elongated plate with cracks obtained by using the developed numerical code and the ANSYS software demonstrated the validity of the proposed numerical method. By using the developed numerical code, dependences of ACFM signals on the lift-off, crack depth, and excitation frequency are simulated and analyzed aiming to improve the ACFM testing conditions for the buried pipe inspection.

A huge tsunami induced by the earthquake which occurred on March 11, 2011 in the Pacific Ocean off the coast of the Tohoku region in Japan led the TEPCO Fukushima Dai-ichi (Number-1) nuclear power station into the station blackout due to a loss of all external power, and the units 1 to 3 (out of 6 units) have led to the core meltdown by losing all core cooling functions. On the other hand, three BWR units in operation at the Onagawa nuclear power station of Tohoku Electric Power Company, which is located most closely to the epicenter, could keep safe from any serious damage despite the fact that the tsunami run-up height had reached to 13.8m. The tough decision was made at the time when the nuclear power plant at the Onagawa site was newly constructed with the land elevation of the ground height by 14.8m. Considering such eventual sequences and as lessons learnt, discussions are made on what should be taken into account in the design and construction, and the operation as the standard and regulation to reduce the risk on ISO12100 of the safety design principle as the world’s global standard and the confidence-rebuilding measures for future nuclear power plants.

This study is aimed at proposing a new maintenance concept; performance based maintenance (PBM). In a flaw tolerance assessment, time duration until the next inspection is determined by crack growth prediction assuming material degradation such as stress corrosion cracking and fatigue. In most cases, the crack growth prediction is excessively conservative and no cracking is found by inspections. The fact of no indication by inspections implies that cracks have not grown to more than a detectable size for the inspection technique applied. In the PBM concept, this inspection result is considered in determining the next inspection schedule. In this study, applicability of the PBM concept was shown for fatigue degradation of stainless steel components in a pressurized water reactor primary water environment. The relationship between crack depth and number of cycles obtained in the author’s previous study was used for the crack growth prediction. Then, crack growth for a postulated fatigue crack was predicted considering strain gradient in the depth direction caused by thermal loading. It was shown that the duration until the next inspection could be optimized based on the inspection result together with the crack growth curve. A longer operation time before the inspection resulted in a longer duration until the next inspection.

In Fukushima Daiichi nuclear power plant (NPP), the environment with high dose rate causes a serious problem. The dose is especially high inside of the reactor building, so it is difficult for people to access there. Therefore, many kind of work in this severe environment by the remote devices is required strongly. These works include decontamination to reduce the environmental dose rate. MHI has implemented the development of remote-operated robot, and we continue to struggle for settling the accident of Fukushima Daiichi NPP. This article describes the actual performance of “MHI-MEISTeRⅡ” which was developed in 2015 to decontaminate for an upper floor at Fukushima Daiichi NPP.

Reactivity measurement is crucial in monitoring nuclear reactors. Proper estimation of neutron flux derivative is one of the key to improve the reactivity estimation based on inverse kinetics (IK) method. Signal differentiation is widely used in system monitoring and control. The key problem in differentiator design is how to obtain a satisfactory tradeoff between differentiation exactness and robustness to uncertainties and noises. By proposing a new finite-time stabilizer for 2-order integrator chain, a new finite-time-convergent 2-order differentiator with bounded estimation error is developed. This differentiator is then applied to reactivity measurement through solving the IPK equation. Numerical simulation results not only verify the theoretical results but also show its satisfactory performance.

Graphite is an important moderator material in nuclear reactors, which has a complex structure, where complexity refers to two main respects: defects and disorder within the crystallites, and porosity including cracks and pores with lengths varying over six orders of magnitude, from nanometers to millimeters. A comprehensive characterization of the structure, needed e.g. to predict structural changes for long-term safe operation of nuclear reactors, requires the complementary input from several techniques to cover the relevant length scale range. In this work X-ray/neutron diffraction, small angle neutron scattering (SANS), spin echo SANS (SESANS) and neutron imaging have been applied to investigate the structure of nuclear graphite from atomic to macroscopic length scales: from nm to mm. The results provide a key to model and quantify disorder at the atomic level. At the meso- and macro-scopic levels a fractal structure is found that spans over an extraordinary large scale of lengths of 6 orders of magnitude and has fractal dimensions close to 2.5, a case where surface and mass fractal dimensions coincide.

Fenton and Fenton-like oxidation processes have been developed to effectively disintegrate and mineralize spent radioactive ion exchange resins from nuclear power plants. In this paper, the degradation of spent cationic resins by Fenton-like process for chemical oxygen demand (COD) removal and weight reduction was investigated. The effects of initial pH, Cu2+ concentration and H2O2 dosage on resin degradation by the Cu2+/H2O2 oxidation process was studied. The results showed that lower initial pH value brought higher COD removal rate. With the increasing amount of Cu2+ concentration and H2O2 dosage, the COD removal rate of resins first increased and then decreased. The degradation efficiency (in term of COD removal rate) and weight reduction percentages were 99% and 39%, respectively, at pH 0.75 and T 95 C with 0.2 M Cu2+, and 35 mL 30% H2O2. The scanning electron microscopy images provided the information of morphological changes of resin during the dissolution process, which clearly revealed the degradation of resins.