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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
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Outline of New Safety Standard (Design Basis)
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EJAM6-3NT66 Development of the Air dose Rate Evaluation System (ARES)

(in English)


Vol.7 No.1previous AASP14(87-88-89-90-91-92-93-94-95-96-97-

Academic Articles
Vol.7, No.1(2015) p.1 - p.137

Special Issue 14

The 2nd International Conference on Maintenance Science and Technology (ICMST-Kobe 2014)

This special issue includes selected papers that were submitted at the 2nd International Conference on Maintenance Science and Technology (ICMST-Kobe 2014). ICMST-Kobe 2014 was held on November 2–5, 2014 at Kobe University Centennial Hall and Takikawa Memorial Hall, and was chaired by Prof. Fumio Kojima of Kobe University. There were 102 presentations and 200 participants from eight countries.
The conference was an international forum for presentations and discussions on important issues relating to maintenance and for the exchange of up-to-date information on advanced maintenance technologies relating to nuclear power plants. Maintenance science and technology is an engineering discipline that provides a basis for scientifically rational maintenance activities. The conference provided an opportunity to consider the contribution of maintenance activities to the safety management of the operation of nuclear power plants through thoughtful discussions.
The scope of the conference covered the question, “What is Resilience for Nuclear Power Plants, for Industries, and for Society?” The concept of “resilience” was originally taken to be the recovery capacity of an ecosystem in the face of environmental challenges, but is now attracting interest from engineers as a concept that can be used to develop existing frameworks of system safety, reliability, and risk engineering. The decommissioning of the Fukushima Dai-ichi plant was a crucial issue discussed at the conference.
The papers presented at the conference are being published in two steps: first, short papers have been published in the proceedings booklet, and then, full-length papers are being published in two special issues of E-Journal of Advanced Maintenance (EJAM) after peer review following the conference. Thirty-three full-length papers were submitted for publication in the special issues, with 19 papers accepted for this, the first, issue. Other accepted papers will be published in the second issue.
Finally, I would like to thank the associate editors of the special issues, Prof. Takayuki Aoki, Prof. Naoto Kasahara, Dr. Ichiro Komura, Prof. Kazunori Morishita, Prof. Takao Nakamura, Prof. Hiroshi Shimoda, Dr. Naoki Soneda, Dr. Shigeru Takaya, and Prof. Noritaka Yusa, for their hard work in reviewing the full-length papers and also acknowledge all reviewers for their careful reviews.

Guest Editor,
Toshiyuki Takagi


The oscillation of a thermal stratification layer can induce thermal fatigue damage on structures with nuclear components. To evaluate the thermal stress induced by thermal stratification oscillation, a frequency response function was developed in our previous research. However, this function does not consider the thickness of the stratified layer. Thus, it is difficult to evaluate the stress generated by actual thermal stratified layers having finite thicknesses with sufficient accuracy. To clarify the effects of layer thickness on induced thermal stress, finite element simulations were conducted under various fluid conditions. As a result, it was clarified that the non-dimensional layer thickness Ht*, which is the ratio of layer thickness to layer oscillation length, can explain the thermal stress response mechanism with layer thickness. Based on the clarified mechanisms, the frequency response function was improved. Applicability of the proposed function to a closed branch pipe of a Light Water Reactor (LWR) and the upper plenum of a pressure vessel of a Fast Breeder Reactor (FBR) was validated through comparison with finite element simulations.

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The progress of the nuclear industry within the United States (U.S.) in responding to regulatory changes following the events at Fukushima Dai-ichi is significant and rapidly evolving. The U.S. Nuclear Regulatory Commission (USNRC) has issued questions related to submitted utility approaches in satisfying Order EA-12-049. In addition to responding to these questions, the U.S. nuclear industry has continued to perform plant modifications, procure equipment, establish national response centers, establish new plant procedures, and address other Regulatory Orders and Directives. In addition to these required activities, the U.S. nuclear industry has also identified means to utilize these new capabilities to provide other benefits.

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Daisuke KANAMARU, Shigetaka OKANO, Masahito MOCHIZUKI

In this study, semi-nondestructive measurement of non-equibiaxial stress using the indentation technique without a reference load under the non-stress state has been applied to low-carbon austenitic stainless steel welds. In order to clarify the material dependency of the applied procedure, two constants (the load ratio and the conversion factor) for stress determination were experimentally quantified and compared with those of high-strength structural steel. The accuracy of stress measurement by the procedure applied to low-carbon austenitic stainless steel welds was verified through a comparison with the X-ray diffraction and stress relief methods. Based on the results, the applied procedure had less dependency on the material properties of these steels and was in good agreement with the conventional techniques in the overall distribution of residual stress.

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The application of the software PVrisk is discussed which was compiled in 2008 to solve probabilistic fracture mechanical problems in case of pressure vessels and pipes by utilizing approaches based on the Failure Assessment Diagram (FAD). The probabilistic feature is embedded due to Monte Carlo simulation (MCS). Compared to other tools available the advantage of PVrisk is to take the combination of fracture mechanics and NDT results based on Probability of Detection (POD) data into account. This data can be the result of an experimental POD trial or the result of computed POD-curves due to numerical modeling of the inspection process. This paper uses POD curves by evaluating experimental data obtained in a project in the German nuclear safety research program. The NDT task was the inspection of welds in austenitic stainless steel plates by using ultrasonic testing (UT) and the application of phased array transducers (PAUT – Phased Array UT) as inspection technique. Taking into account a realistic statistical population of half-elliptical, axial oriented, subsurface and surface breaking cracks in a larger diameter austenitic pipe as well as the material properties such as elastic limit Rp0.2, tensile strength Rm, and fracture toughness KIC numerical statistical modeling is used to calculate the probability of failure (POF). The advantage of using NDT with optimized POD is documented.

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Axel HILL, Dr. Cristoph STIEPANI, Michael DRECHSLER

Severe accidents might cause the release of airborne radioactive substances to the environment of the NPP either due to containment leakages or due to intentional filtered containment venting. In the latter case aerosols and iodine are retained, however noble gases are not retainable by the FCVS or by conventional air filtration systems like HEPA filters and iodine absorbers. Radioactive noble gases nevertheless dominate the activity release depending on the venting procedure and the weather conditions. To prevent unacceptable contamination of the control room atmosphere by noble gases, AREVA GmbH has developed a noble gas control room accident filtration system (CRAFT) which can supply purified fresh air to the control room without time limitation. The retention process is based on dynamic adsorption of noble gases on activated carbon. The system consists of delay lines (carbon columns) which are operated by a continuous and simultaneous adsorption and desorption process. CRAFT allows minimization of the dose rate inside the control room and ensures low radiation exposure to the staff by maintaining the control room environment suitable for prolonged occupancy throughout the duration of the accident. CRAFT consists of a proven modular design either transportable or permanently installed.

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The postulated fatigue crack growth curve (P-curve), which represents the relationship between crack size and fatigue damage, has been developed for Type 316 stainless steel. In this study, further improvement of the P-curve was made. Particularly, the incubation period before initiation of a 0.1 mm depth crack was newly included in the P-curve. Then, the P-curve was extended for thermal fatigue loading, which is the main cause of fatigue damage in nuclear power plant components. The stress and strain gradient in the depth direction derived from thermal convection analysis was reasonably considered in the crack growth prediction. It was shown that the fatigue damage (DF) of a component could be quantified using the P-curve and crack size detected by inspection, or even if no crack is detected, possible DF could be estimated from the detectable crack size of the inspection technique. By incorporating the probability of detection (POD) of an inspection technique into the P-curve, the PD-curve, which is the relationship between the DF and POD, was developed. It was discussed that the PD-curve can be used to show the specifications of inspection techniques necessary for ensuring component integrity.

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Jonghoon IM, Hiromitsu FUJII, Atsushi YAMASHITA, and Hajime ASAMA

When images are acquired in bright condition, it can cause a loss of highlight details (over exposure) in bright area and a loss of shadow details (under exposure) in dark area. Over and under exposure causes a big problem when people investigate dangerous place like Fukushima nuclear power plant through the camera attached remote control robot. In this paper, we propose a method to correct the over and under exposure image to solve the problem. The image processing consists of four steps. Firstly, multiple images are acquired by alternately turning on and off each illumination which set in different positions. Then the image obtained first is defined as input image 1, the image obtained second is defined as input image 2 and the image obtained N-th is defined as input image N. Secondly, luminance of the images is corrected. Thirdly, over and under exposure regions in the image are extracted from the input image 1. Finally, the over and under exposure regions in the input image 1 are compensated by other images. The results show that the over and under exposure regions in the input image are recovered by our proposed method.

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Kouki MARUYAMA, Iliana MARINOVA, Yoshifuru SAITO

Previously we have succeeded in developing a new ECT sensor called ∞ coil. This new ECT sensor is a relatively high sensibility and has the high liftoff characteristics compared with that of conventional ECT sensor.
However, the ∞ coil confronts to a serious difficulty to apply the curved surface specimens. To overcome this difficulty, this paper has worked out a flat ∞ coil. This flat ∞ coil exhibits a high sensitivity not only to the curved surface but also to the flat surface specimens because of its highly shape flexibility to fit the curved target surface and widely spreader-able exciting coils.
Intensive numerical simulations employing 3D FEM have been carried to show the usefulness of the flat ∞ coil. The experimented results have verified the validity of the numerical simulations. Thus, we have confirmed the versatile capabilities of the flat ∞ coil.

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Takao NAKAMURA, Ryosuke FUJIKAWA, Mikiya MATSUSHITA, Shigeki ABE, Masayuki KAMAYA

After Fukushima Daiichi NPP accident, the improvement of nuclear safety is highly requested in order to prevent re-occurrence of severe accident. New scheme of fatigue evaluation to confirm system safety in NPPs is enhanced to be established based on Defense in Depth concept. This study focuses on the direction to reconstruct Grand Design of fatigue evaluation to ensure system safety based on fatigue failure analysis of Japanese NPPs. The issues of fatigue evaluation we have now is investigated in order to review Grand Design on the basis of new role of fatigue management in plant long term safe operating scheme that is required after Fukushima daiichi NPP accident.

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Li Liu, Zheng Cao Li, Naoto Sekimura

Currently China has 22 units of nuclear power reactors (NPPs) in operation, 27 units under construction. The first NPP Qinshan-I has been operated for 23 years, while the design life time is 30 years. In order to maintain safety operation of NPPs, it is very important to detect ageing effects, to understand related reduction in safety margins and to take actions before loss of integrity and functions. Ageing management program is set up for systems, structures and components important to safety to understand, monitor, and manage ageing effects in China. In addition, ageing, as one of safety factors, is reviewed during Periodical Safety Review. The regulation on licence renewal or long term operation should be decided by Chinese regulatory body in the near future.

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Hideaki NEMORI, Iliana MARINOVA, and Yoshifuru SAITO

To evaluate a stress effect to the ferromagnetic properties, this paper proposes a methodology utilizing a frequency response of the domain based magnetization model. A key idea of this approach is based on the following facts that the parameters identification of a domain based magnetization model has been successfully developed by a harmonic balance approach. Since the parameters of the model are extremely sensitive to the various measurement conditions such as temperature, mechanical stress and so on, investigation to the model parameter deflection while controlling a particular measurement condition reveals its controlled parameter effect to the tested ferromagnetic materials. As a result, it is found that the stress effect to a parameter expressing hysteretic property has been clearly deflected its values depending on the externally applied stresses.

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Yu KAMIJI, Masashi TANIGUCHI, Yasuo NISHIHATA, Ryuji NAGAISHI, Hirohisa TANAKA, Shingo HIRATA, Mikiya HARA, Ryutaro HINO

For hydrogen mitigation, a new type passive autocatalytic recombiner has been developed, as one of safety accident managements at a nuclear power plant. This new recombiner has an automotive monolithic substrate catalyst in order to reduce weight and to improve hydrogen treating capacity, environmental resistance, and product quality. The activation energy has been experimentally estimated at 5.75 kJ/mol with stoichiometric composition gas flow. The fed hydrogen and oxygen to the monolithic catalyst were immediately recombined in both the dry and the humid atmospheres. Moreover, it has been found that gamma-ray irradiation up to 1.0 MGy could promote catalytic activity, because the specific area of the catalyst and the surface area of the precious metals increased. In addition, no significant difference was observed in the compositional distribution in the monolithic catalyst by EPMA. As the result, an automotive monolithic catalyst is applicable to practical installation of PAR into NPPs.

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Naoyuki ISHIDA, Akinori TAMURA, Toshinori KAWAMURA, Kazuaki KITOU, Mamoru KAMOSHIDA

The Fukushima Daiichi Nuclear Power Plant accident and its consequences have led to extensive rethinking about the safety technologies used in boiling water reactors (BWRs). As one of the options of the safety technologies, we have been developing passive cooling systems consisting of a water-cooling system and an infinite-time air-cooling system. These systems achieve core cooling without electricity and are intended to cope with a long-term station blackout (SBO). Both these cooling systems remove relatively high decay heat for the initial 10 days after an accident, and then the infinite-time air-cooling system continues to remove attenuated decay heat after this period. To obtain heat transfer data for the design of the water-cooling system, we conducted heat transfer tests using a full-scale U-shaped single tube. The data were obtained at a system pressure of 0.2 to 3.0MPa (absolute) and inlet steam velocity of 5 to 56m/s. To enhance heat transfer of the air-cooling system, we successfully implemented some air-cooling enhancing technologies. The performance was evaluated by heat transfer data obtained from the element heat transfer tests. The heat transfer performance increased at least 100% with the enhancement technologies compared with a bare tube. From these test results, we confirmed good feasibility for application of the cooling systems.

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Shigetaka OKANO, Misa MIYABE, Masahito MOCHIZUKI

In this study, a new methodology is proposed for semi-nondestructive determination of anisotropic stress-strain curves for steel by using asymmetric indenters with different vertical angles. The functional relation between the indentation load-depth curve and the representative stress for steel was derived from the π theorem, and then the relation was expressed as a function of the work-hardening coefficient and constant depending on the geometry of the indenter. With the proper selection of representative strain level varies according to the geometry of indenter, this relation is directly dependent on the geometry of indenter. On the basis of π theorem and FE solutions, the relation between the representative strain level and the geometry of the indenter was quantified. Subsequently, specific function formulae indicating the relation between the indentation load-depth curve and the representative stress for anisotropic steel were constructed with a focus on the effect of the degree of anisotropy and the vertical angle of indenter. By using these formulae, the representative stresses for anisotropic steel can be quantified at different representative strain levels and thus the anisotropic stress-strain curve can also be obtained. The usefulness of developed procedure was evaluated through numerical simulation.

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This study presents electrical condition monitoring technique aiming at measuring volume resistivity (ρ) precisely for the tubular-shaped cables insulators. Tubular electrodes were designed to detect ρ for nuclear grade flame retardant ethylene propylene rubber (FR-EPR) insulated cables that underwent thermal accelerated ageing. FR-EPR insulated cables were heated at 125 °C for 5040 hours by two methods: with jacket and without jacket, while only FR-EPR cable with jacket was heated at 150 °C for 336 hours. ρ in insulators heated with jacket sharply decreased at ageing times 800-3480 hours and 300-336 hours during ageing at 125 °C and 150 °C respectively, hence obeying “induction-time” behavior as confirmed from elongation at break degradation curves at similar ageing conditions. Reduction of ρ in insulators heated at 125 °C without jacket is less significant compared to ageing with jacket.

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Japanese regulations for thermal fatigue issues are mainly based on the guideline of the Japan Society of Mechanical Engineers (JSME). However, the methods of the guideline include unclear safety margins, owing to existing uncertainties on loading and strength parameters. As a first step toward improving the JSME guideline to quantify safety margins with probabilistic approaches, a reliability evaluation procedure for thermal fatigue of mixing tees suitable for codes and standards has been discussed in this paper. While Monte Carlo simulations would be unsuitable for engineering use due to its cumbersome process and requirement for professional knowledge of probabilistic theory, simplified reliability analysis methods, such as the advanced first-order second-moment method, are more applicable in code designs. In addition, partial safety factors are easier to create a simpler measure of safety margins for code users who are unfamiliar with probabilistic reliability analysis methods. Thus, this paper has proposed a practical reliability evaluation procedure, which is based on the load and resistance factor design concept, for the thermal fatigue induced by random fluid temperature fluctuations in a mixing tee.

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Li Liu, Kenta Murakami, Kenji Dohi, Kenji Nishida, Akiyoshi Nomoto, Naoki Soneda, Zhengcao Li, Naoto Sekimura

The accurate understanding and prediction of neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is a very important issue to ensure the safety and continued operation of nuclear power plant. High-density nano-scale Cu-rich precipitates, which are typically alloyed with Si, Ni or Mn, have been considered as the main contributor to the hardening embrittlement of RPV steels. In our work, four series of RPV model alloys, Fe-Cu, Fe-Cu-Si, Fe-Cu-Ni and Fe-Cu-Ni-Mn, are thermally aged at 450˚C for up to 380 hours. The effects of Si, Ni and Mn on hardening are discussed. Furthermore, Atom Probe Tomography (APT) was used to study Cu-rich precipitates.

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Shingo HIRATA, Tomoaki MOURI, Minoru IGARASHI, Manabu SATOH, Yu KAMIJI, Yasuo NISHIHATA, Masashi TANIGUCHI, Hirohisa TANAKA, and Ryutaro HINO

A new type passive autocatalytic recombiner to mitigate released hydrogen gas at the accident of nuclear power plants and related facilities has been developed. In order to realize easy and anywhere installation, this new recombiner has advanced automotive catalysts with features such as light and small, high durability, and high performance. Downsizing and light-weighting of recombiner has been conducted with thermal hydraulic and structural analysis incorporating an automotive catalyst model and its characterization results obtained through experiments. Thermal hydraulic and structural analysis show PAR concept proposed is feasible for hydrogen mitigation in nuclear power plants and related facilities.

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Shigeki ABE, Takao NAKAMURA, Masayuki KAMAYA

In order to establish sophisticated management of aging degradation and to achieve high reliability of components in nuclear power plants, it is required to reveal the mechanism of aging degradation and to quantify its deterioration. In low-cycle fatigue regime, it was shown that the number of cycles to specimen failure (fatigue life) can be estimated by predicting crack growth. The application of crack length for representing fatigue damage will make it possible to measure fatigue damage by inspection. In previous study, crack growth with inhomogeneous rate and fatigue life in a typical condition were predicted with statistical model of micro crack growth. In this study, fatigue test in expanded condition was conducted in air at room temperature in order to define the relationship between damage factor (DF = number of cycles/fatigue life) and crack length. Crack initiation and crack propagation during fatigue test were measured by replica investigation periodically. Statistical model of micro crack growth was applied to the observation result and its applicability for various strain ranges was demonstrated. The relationship between the damage factor and crack length was shown and its use for maintenance was discussed.

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EJAM Vol.7 p.1-137 Academic Articles Special Issue on "The 2nd International Conference on Maintenance Science and Technology for Nuclear Power Plants (ICMST-Kobe 2014)"