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Journal Articles

Proposal of inspection rationalization method and application for sodium cooled fast reactor

Yada, Hiroki; Takaya, Shigeru; Enuma, Yasuhiro

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet),  ( ), 7 Pages, 2020/08

Journal Articles

A Simplified method for evaluating sloshing impact pressure on a flat roof based on Wagner's theory

Takaya, Shigeru; Fujisaki, Tatsuya*

Mechanical Engineering Journal (Internet), 7(3), p.19-00526_1 - 19-00526_10, 2020/06

In severe seismic conditions, sloshing waves are considered to even reach a roof slab of a reactor vessel. The structural integrity of roof slabs is required to be evaluated against sloshing impacts. However, there is no widely recognized evaluation method for sloshing impact pressure on flat roofs yet. Therefore, in this paper, a simplified evaluation method is proposed based on Wagner's theory, which is a well-known classic theory for evaluating impact pressures on rigid wedges dropping on water surfaces. In the proposed method, we assume an equivalent wedge on a flat roof. The impact pressure on the equivalent wedge is evaluated by applying Wagner's theory. Computational fluid dynamics analysis is conducted to confirm that a key assumption of Wagner's theory is applicable to the evaluation of sloshing impact on a flat roof. In addition, the predictability of the proposed method is investigated by comparing literature data of sloshing experiments with the estimated values.

Journal Articles

Influence of dead weight and internal pressure to seismic buckling probability of fast reactor vessels

Takaya, Shigeru; Sasaki, Naoto*

Mechanical Engineering Journal (Internet), 7(3), p.19-00549_1 - 19-00549_9, 2020/06

Seismic buckling of vessels is one of main concerns for the design of fast reactor plants in Japan. In previous studies, we discussed evaluation methods of seismic buckling probability of vessels by taking account of seismic hazards in order to rationalize seismic buckling evaluation, and proposed a rule for seismic buckling of vessels based on the load and resistant factor design method. The proposed method deals with only seismic load, but in actuality, dead weight and internal pressure also exist. In this study, the rule was expanded so that dead weight and internal pressure can be taken into account. Furthermore, the influences of dead weight and internal pressure to seismic buckling evaluation were discussed. As result, it was shown that approximately 10 to 20% of further rationalization of allowable seismic load could be achieved by considering dead weight and internal pressure in the evaluation.

Journal Articles

Development of prototype reactor maintenance, 3; Application to valves of sodium-cooled reactor prototype

Chikazawa, Yoshitaka; Takaya, Shigeru; Tagawa, Akihiro; Kubo, Shigenobu

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 6 Pages, 2019/05

A maintenance management required to prototype nuclear power reactors has been developed. One of important mission of a prototype reactor is to develop maintenance program for commercial reactors step by step securing safety. Since operating experience at the early stage is limited, the maintenance program for the prototype reactor should be a progressive one. It has to be modified and improved frequently taking into account R&D insight and operation experiences. Additionally, the maintenance program has to consider features of the prototype reactor even at the early stage. To select maintenance grades on particular components/systems, risk informed and graded approaches are effective. And maintenance programs have to take into account degradation mechanism originally due to reactor features. In this paper, applications for maintenance program on sodium valves of prototype fast breeder reactor Monju are studied as an example of prototype sodium-cooled reactors (SFR).

Journal Articles

Proposal of a simple evaluation method for sloshing impact pressure on flat roofs

Takaya, Shigeru; Fujisaki, Tatsuya*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07

Sloshing is one of important issues for both loop-type and tank-type fast reactors with free liquid surface. Periods of seismic vibration are lengthened by base isolation systems installed to prevent damages to facilities during earthquakes, and get close to the natural periods of sloshing. As a result, sloshing is promoted. Sloshing waves are assumed to even reach a roof slab of a reactor vessel in severe seismic conditions. It is important to evaluate structural integrity for sloshing impacts on roofs. However, there are not any established evaluation methods for impact pressure on flat roofs yet. Therefore, in this study, a simple evaluation method is proposed based on Wagner's theory. The effectiveness of the proposed method is illustrated using computational fluid dynamics analysis and literature data of sloshing experiments.

Journal Articles

Proposal on LBB evaluation conditions for sodium cooled fast reactor pipes and effects of pipe parameters

Yada, Hiroki; Takaya, Shigeru; Wakai, Takashi; Nakai, Satoru; Machida, Hideo*

Nippon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00389_1 - 17-00389_15, 2018/03

no abstracts in English

Journal Articles

Creep-fatigue evaluation method for weld joints of Mod.9Cr-1Mo steel, 1; Proposal of the evaluation method based on finite element analysis and uniaxial testing

Ando, Masanori; Takaya, Shigeru

Nuclear Engineering and Design, 323, p.463 - 473, 2017/11

AA2016-0317.pdf:0.77MB

 Times Cited Count:1 Percentile:80.48(Nuclear Science & Technology)

In the present study, a method for creep-fatigue life evaluation of Mod.9Cr-1Mo steel weld joint was proposed based on finite element analysis (FEA). Since the point of the creep-fatigue life evaluation in the weld joint is a consideration of the metallurgical discontinuities, FEA was performed using a model with three material properties, a base metal (BM), weld metal (WM) and a heat-affected zone (HAZ) formed in the base metal due to the welding heat input, to consider the mutual relationships among them. The material properties of these three materials were collected and utilized in FEA for considering such metallurgical discontinuities. The creep-fatigue life estimated using the proposed evaluation method based on the FEA results were compared with available creep-fatigue test data, and the proposed method was found to predict the number of cycles to failure within a factor of 3.

Journal Articles

Numerical analysis of flow-induced vibration of large diameter pipe with short elbow

Takaya, Shigeru; Fujisaki, Tatsuya*; Tanaka, Masaaki

Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07

Flow-induced vibration (FIV) of a hot-leg piping is one of main concerns in the design of an advanced loop-type sodium cooled fast reactor. We have been developing numerical analysis models to deal with this issue. In this study, computational fluid dynamics (CFD) simulation of a 1/3 scaled-model of the hot-leg piping was conducted. The results such as velocity profiles and power spectral densities (PSD) of pressure fluctuations were compared with experiment ones. The simulated PSD of pressure fluctuation at the recirculation region agreed well with the experiment. Then, stress induced by FIV was evaluated using pressure fluctuation data calculated by the CFD simulation. The calculated stress generally agrees well the measurement values, which indicates the importance of precise evaluation of the PSD of pressure fluctuation at the recirculation region for evaluation of FIV of the hot-leg piping with a short elbow.

Journal Articles

Load and resistance factor design approach for seismic buckling of fast reactor vessels

Takaya, Shigeru; Sasaki, Naoto*; Asayama, Tai; Kamishima, Yoshio*

Mechanical Engineering Journal (Internet), 4(3), p.16-00558_1 - 16-00558_12, 2017/06

In this study, we developed a new design rule for the prevention of seismic buckling of vessels using the load and resistance factor design method to enable more rational vessel designs. The effectiveness of the new design rule was illustrated in comparison with the current provision.

Journal Articles

Development of prototype reactor maintenance, 1; Application to piping system of sodium-cooled reactor prototype

Kotake, Shoji*; Chikazawa, Yoshitaka; Takaya, Shigeru; Otaka, Masahiko; Kubo, Shigenobu; Arai, Masanobu; Kunogi, Kosuke; Ito, Takaya*; Yamaguchi, Akira*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

A maintenance management required to prototype nuclear power reactors is proposed. Monitoring and control of sodium impurity and thermal transient are extremely important for sodium boundary maintenance for sodium-cooled fast reactors. At the fast stage of the prototype reactor Monju operation, degradation mechanism on the piping should be demonstrated based on operation experiences. Therefore inspection on a representative position for crack indication and pipe thickness is proposed. Due to less experience of SFR plants, early detection of boundary failure is considered. For a matured operation stage, when degradation mechanism is well demonstrated based on inspection data, inspection cycle could be extended. And for commercial reactors, maintenance without inspection will be established based on accumulated operation experiences including those of the prototype reactor Monju.

Journal Articles

Development of prototype reactor maintenance, 2; Application to piping support of sodium-cooled reactor prototype

Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji*; Ito, Takaya*; Yamaguchi, Akira*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

Applications for maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of the piping support could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports.

Journal Articles

Proposal of maintenance management of nuclear power plants at R&D stage by taking account of their features

Takaya, Shigeru; Chikazawa, Yoshitaka; Hayashida, Kiichi; Tagawa, Akihiro; Kubo, Shigenobu; Yamashita, Atsushi

Hozengaku, 15(4), p.71 - 78, 2017/01

A maintenance management suitable to nuclear power plants (NPP) at R&D stage was discussed. Objectives of maintenance management of NPP at R&D stage was first clarified. Next, applicability of codes for maintenance management of commercial NPP to NPP at R&D stage was discussed. Then, requirements and consideration for maintenance management of NPP at R&D stage was proposed. Finally, an example that the proposal was applied to setting maintenance program of sodium-cooled fast reactor was presented.

Journal Articles

Development of simple estimation method for the influence of parameter uncertainty of probability distributions against evaluation result of probabilistic fracture mechanics

Okajima, Satoshi; Takaya, Shigeru; Asayama, Tai

Nippon Kikai Gakkai Rombunshu (Internet), 83(845), p.16-00434_1 - 16-00434_13, 2017/01

no abstracts in English

Journal Articles

Creep-fatigue evaluation method for weld joint of Mod.9Cr-1Mo steel, 2; Plate bending test and proposal of a simplified evaluation method

Ando, Masanori; Takaya, Shigeru

Nuclear Engineering and Design, 310, p.217 - 230, 2016/12

AA2016-0318.pdf:2.62MB

 Times Cited Count:1 Percentile:83.2(Nuclear Science & Technology)

In the present study, to develop an evaluation procedure and design rules for Mod.9Cr-1Mo steel weld joints, a method for creep-fatigue life evaluation of Mod.9Cr-1Mo steel weld joints was proposed based on the finite element analysis (FEA) results with a series of cyclic plate bending tests of the longitudinal and horizontal seamed plates. The strain concentration and redistribution behaviors were evaluated and failure cycles were estimated using FEA by considering the test conditions and metallurgical discontinuities in the weld joints. The elastic follow-up factors calculated from a comparison of the elastic and inelastic FEA results were determined to be less than 1.5. Based on the estimated elastic follow-up factors obtained via inelastic FEA, a simplified technique using elastic FEA was proposed for evaluating the creep-fatigue life in Mod.9Cr-1Mo steel weld joints.

Journal Articles

Determination of in-service inspection requirements for fast reactor components using System Based Code concept

Takaya, Shigeru; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Asayama, Tai

Nuclear Engineering and Design, 305, p.270 - 276, 2016/08

AA2016-0006.pdf:0.51MB

 Times Cited Count:2 Percentile:69.72(Nuclear Science & Technology)

In our previous study, we proposed a new process for determining the in-service inspection (ISI) requirements using the System Based Code concept. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other on plant safety. In this study, the ISI requirements for a reactor guard vessel (RGV) and core support structure (CSS) of a prototype sodium-cooled fast breeder reactor in Japan (Monju) were investigated using the proposed process. It was shown that both components had sufficient reliability even assuming unrealistic severe conditions. The failure occurrences of these components were practically eliminated. Hence, it was concluded that no ISI requirements were needed for these components. The proposed process is expected to contribute to the realization of effective and rational ISI by properly taking into account plant-specific features.

JAEA Reports

Maintenance management of nuclear power reactors at the stage of research and development

Takaya, Shigeru; Chikazawa, Yoshitaka; Hayashida, Kiichi; Tagawa, Akihiro; Kubo, Shigenobu; Yamashita, Atsushi

JAEA-Research 2016-006, 66 Pages, 2016/07

JAEA-Research-2016-006.pdf:3.4MB

A maintenance management required to nuclear power reactors at the R&D stage was discussed. It is the most important to ensure safety of nuclear power plants by taking account of characteristics of nuclear power reactors at the R&D stage. In addition, it is needed to establish a system of maintenance management technologies suitable for reactor types. In this report, objectives of maintenance management of nuclear power reactors at the R&D stage was clarified. Next, requirements and consideration for maintenance management was discussed according to the objectives. "Codes for maintenance management of nuclear power plants" and "Guides for maintenance management of nuclear power plants" were refereed in the discussion. Then, a draft of codes for maintenance management of nuclear power plants at the R&D stage were newly proposed. Finally, an example that the draft codes were applied to components containing sodium, typical components of sodium-cooled fast reactor, was presented.

JAEA Reports

Determination methodologies for input data including loads considered for reliability evaluation of fast reactor components

Yokoi, Shinobu*; Kamishima, Yoshio*; Sadahiro, Daisuke*; Takaya, Shigeru

JAEA-Data/Code 2016-002, 38 Pages, 2016/07

JAEA-Data-Code-2016-002.pdf:1.51MB

Many efforts have been made to implement the System Based Code concept aiming at optimizing margins dispersed in existing codes and standards. Failure probability calculated based on statistical information such as a type of probability distribution, mean (or median) and variance (or standard deviation) for random variables is expected to be a promising quantitative index for margin optimization. Statistical information on material strength, which is also required to calculate the failure probability, has been already reported in JAEA-Data/Code 2015-002 "Structural Properties of Material Strength for Reliability Evaluation of Components of Fast Reactors -Austenitic Stainless Steels-" whereas others have not been identified yet. This report provides methodologies and basic ideas to determine statistical parameters of other random variables, especially variable loads, necessary for reliability evaluation.

Journal Articles

Numerical analysis of unsteady phenomena in upper plenum and hot-leg piping system of advanced loop-type sodium cooled fast reactor

Takaya, Shigeru; Fujisaki, Tatsuya*; Tanaka, Masaaki

Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 5 Pages, 2016/07

JAEA is conducting R&D of an advanced loop-type sodium cooled fast reactor. The cooling system is planned to be simplified by employing a two-loop configuration and shortened piping with less elbows than Monju in order to reduce construction costs. The design increases flow velocity in the hot-leg piping and induces large flow turbulence around elbows. Therefore, flow-induced vibration is one of main concerns. The flow field in the hot-leg piping is affected by flow disturbance at the inlet, so it is important to evaluate flow field including the upper plenum. In this study, we analyzed unsteady fluid flow by using an integrated model of the upper plenum and the hot-leg piping system. Unsteady Reynolds averaged Navier-Stokes simulation with Reynolds stress model was used. In general, the simulation results obtained by using the integrated model show a similar tendency with the experiment results of 1/3 scaled-model of hot-leg piping with deflect flows. The coupling effect of swirling and deflected flows seems to be not significant although further investigation is needed.

Journal Articles

Evaluation of fatigue strength of similar and dissimilar welded joints of modified 9Cr-1Mo steel

Takaya, Shigeru

Journal of Pressure Vessel Technology, 138(1), p.011402_1 - 011402_9, 2016/02

 Times Cited Count:0 Percentile:100(Engineering, Mechanical)

This paper presents an evaluation method for the fatigue strength of similar and dissimilar welded joints of modified 9Cr-1Mo steel, which is a candidate structural material for a demonstration fast breeder reactor being developed in Japan. The discontinuity of the mechanical properties across a welded joint causes a non-homogeneous strain distribution, and this effect should be considered in the evaluation of the fatigue strength of welded joints. In this study, a "2-element model," which comprises base metal and welded metal, was employed. First, the strain ranges of each element are calculated, and second, the fatigue lives of each element are evaluated. Finally, the shorter fatigue life is selected as the fatigue life of the welded joint. The failure position can be also estimated by this model. The evaluation results were compared with experimental data obtained at elevated temperature, and the results were in good agreement.

Journal Articles

Development of high temperature magnetic sensor

Takaya, Shigeru; Arakawa, Hisashi*; Keyakida, Satoshi*

Hozengaku, 14(3), p.81 - 87, 2015/10

A magnetic sensor which can be applied to measurement at elevated temperature was newly developed. It is a kind of flux gate magnetic sensor. Permendur was employed for a magnetic core instead of Permalloy which is commonly used because Permendur has much higher Curie point, about 1000$$^{circ}$$C. Heat resistant ceramic coating Cu wires were used for coils. External magnetic field is detected by shift of peak position of differential permeability during induction of triangular excitation current. A magnetic core has race track shape with a fine part to make peak position more clear and increase detectability. The output of the developed sensor showed good linearity with external magnetic field even at 500$$^{circ}$$C. Furthermore, the durability of the sensor was discussed, and it was shown that decrease in coil performance after some operation at elevated temperature seems to be a critical issue.

165 (Records 1-20 displayed on this page)