Shigeru Takaya

Development of 2012 Edition of JSME Code for Design and Construction of Fast Reactors: (3) Development of the Material Strength Standard of Modified 9Cr-1Mo Steel

This paper describes the material strength standard of Modified 9Cr-1Mo (ASME Gr.91) steel in the design code for fast reactors of 2012 edition published by the Japan Society of Mechanical Engineers. Modified 9Cr-1Mo is to be used for primary and secondary coolant circuits, including intermediate heat exchangers and steam generators for the Japan Sodium Cooled Fast Reactor (JSFR). Modified 9Cr-1Mo steel was developed in Oak Ridge National Laboratory in the United States. Application of Modified 9Cr-1Mo to JSFR needs the material strength standard. Therefore, the authors developed the material strength standard. The material strength standard involved allowable limits such as S0, Sm, Su, Sy, SR and St and so on, environment effects such as sodium effects. In addition, material characteristic equations (Creep rupture equation, creep strain equation and equation of best fit curve for low-cycle fatigue life and so on) necessary for the allowable limits were involved. This paper describes the contents of the material strength standard.Copyright

Development of 2012 Edition of JSME Code for Design and Construction of Fast Reactors: (2) Development of the Material Strength Standard of 316FR Stainless Steel

This paper describes the material strength standard of 316FR stainless steel in the design code for fast reactors of 2012 edition published by the Japan Society of Mechanical Engineers. 316FR stainless steel is to be used for a reactor vessel and internals for the Japan Sodium Cooled Fast Reactor (JSFR). 316FR was developed in Japan by optimizing chemical composition within the specifications of SUS316 in the Japanese Industrial Standard which is equivalent to Type 316 stainless steel. The optimization was performed from the viewpoint of maximizing the creep resistance under fast breeder conditions. Application of 316FR stainless steel to JSFR needs the material strength standard. Therefore, the authors developed the material strength standard. The material strength standard involved allowable limits such as S0, Sm, Su, Sy, SR and St and so on, environment effects such as irradiation effects and sodium effects. In addition, material characteristic equations (Creep rupture equation, creep strain equation and equation of best fit curve for low-cycle fatigue life and so on) necessary for the allowable limits were involved. This paper describes the contents of the material strength standard.Copyright

Study on Minimum Wall Thickness Requirement for Seismic Buckling of Reactor Vessel Based on System Based Code Concept

The minimum wall thickness required to prevent seismic buckling of a reactor vessel (RV) in a fast reactor is derived using the system based code (SBC) concept. One of the key features of SBC concept is margin optimization; to implement this concept, the reliability design method is employed, and the target reliability for seismic buckling of the RV is derived from nuclear plant safety goals. Input data for reliability evaluation, such as distribution type, mean value, and standard deviation of random variables, are also prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Minimum wall thickness required to achieve the target reliability is evaluated, and is found to be less than that determined from a conventional deterministic design method. Furthermore, the influence of each random variable on the evaluation is investigated, and it is found that the seismic load has a significant impact.

Elaboration of the System Based Code Concept — Activities in JSME and ASME: (4) Joint Efforts of JSME and ASME

This paper describes the ongoing activities at the Joint Task Group for System Based Code established in 2012 by the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) in the ASME Boiler and Pressure Vessel Code Committee. The Joint Task Group aims at developing alternative rules for ASME Boiler and Pressure Vessel Code Section XI Division 3, inservice inspection requirements for liquid metal reactors. The alternative rules will be developed based on the System Based Code concept which was originally proposed in Japan and is being elaborated both in JSME and ASME. The alternative rules are for sodium-cooled fast reactors where some of the components could have difficulties in conforming to the current requirements primarily due to accessibility. The alternative requirements would consist of a set of relieved requirements and a logic flow through which the applicability of them is judged. The logic flow considers both component structural integrity and the plant safety goals. The issuance of a Code Case is envisioned around 2016. Further efforts to integrate the process into a new framework being developed in Section XI would cover various types of reactors.Copyright

Development of Limit State Design for Fast Reactor by System Based Code

Limit state design was newly developed in order to apply the margin exchange which is one of the innovative concepts of the System Based Code (SBC). It was shown that limit state design method is applicable to plant design instead of current design criteria. In this report, working example of a reactor vessel of a Fast Reactor subject to thermal load is conducted to demonstrate this concept. As the result allowable stress was increased by changing the acceptance criteria from current design criteria to limit state design criteria.Copyright

Development of a magnetic sensor system for predictive IASCC diagnosis on stainless steels in a nuclear reactor

The authors previously reported that magnetic flux leakage d ata, measured using a flux gate (FG) sensor, showed a correlation with the irradiation-assisted stress corros ion cracking (IASCC) susceptibility of neutron-irradiate d austenitic stainless alloys. This paper presents a study conducted to develop a diagnostic system that can detect IASCC precursors in stainless steels by measuring the magnetic properties of the material. The eddy current method and alternating current (AC) magnetization method were used, as these will be more practical for use in actual reactors. Probes were developed for these measurement methods, providing sufficient tolerance for en vironments in nuclear reactors. An attempt was also made to improve spatial resolution by manufacturing a smaller probe. A sensor system was designed for remote control, performance tests were conducted by measuring neutron-irradiated specimens and mock specimens, and magnetic data were evaluated by comparing the IASCC susceptibility of the specimens. It was proved that the sensor system developed in this study is capable of detecting IASCC precursors. Further developments necessary for application in actual nuclear reactors and the mechanism of correlation between magnetic properties and IASCC susceptibility were also discussed. Many of the problems experienced in existing light water reactors (LWR) are caused by damage to structural materials. Stress corrosion cracking (SCC) and irradiation-assisted SCC (IASCC) of structural materials have been especially serious problems, and it has not been possible to predict these in the reactor design process. These problems will become more significant when existing reactors start to age from now on, and when advanced reactors are operated and structural materials are used in more severe irradiation environments in future. It is therefore import ant to develop diagnostic techniques that can

Development of Creep–Fatigue Evaluation Method for 316FR Stainless Steel

In the design of fast reactor plants, the most important failure mode to be prevented is creep–fatigue damage at elevated temperatures. 316FR stainless steel is a candidate material for the reactor vessel and internal structures of such plants. The development of a procedure for evaluating creep–fatigue life is essential. The method for evaluating creep–fatigue life implemented in the Japan Society of Mechanical Engineers code is based on the time fraction rule for evaluating creep damage. Equations such as the fatigue curve, dynamic stress–strain curve, creep rupture curve, and creep strain curve are necessary for calculating creep–fatigue life. These equations are provided in this paper and the predicted creep–fatigue life for 316FR stainless steel is compared with experimental data. For the evaluation of creep–fatigue life, the longest time to failure is about 100,000 h. The creep–fatigue life is predicted to an accuracy that is within a factor of 2 even in the case with the longest time to failure. Furthermore, the proposed method is compared with the ductility exhaustion method to investigate whether the proposed method gives conservative predictions. Finally, a procedure based on the time fraction rule for the evaluation of creep–fatigue life is proposed for 316FR stainless steel.

Development of Integrated Numerical Analysis Model for Unsteady Phenomena in Upper Plenum and Hot-Leg Piping System of Japan Sodium-Cooled Fast Reactor

Flow-induced vibration of hot-leg pipings is one of concerns for the design of Japan Sodium-cooled Fast Reactor (JSFR) which is now being developed. The flow field in the hot-leg pipings is supposed to be affected by flow disturbances at the entrance, so it is important to evaluate flow fields including the upper plenum. In this study, a simulation model of the upper plenum and the hot-leg piping system of JSFR was developed. Unsteady fluid flow analyses were then conducted by unsteady Reynolds averaged Navier-Stokes simulation (URANS) with Reynolds stress model. The appropriateness of the calculated results was discussed by comparing available scale model test results. Furthermore, a prototype model for vibration analysis of the hot-leg piping was developed. In the model, the transient pressure data predicted by the URANS were used as input data for the vibration analysis. The number of element was significantly reduced from that of CFD model by considering the correlation length of stress fluctuation. In addition, a stress mapping tool from a CFD model to a model for vibration analysis was created.Copyright

Evaluation of Fatigue Strength of Similar and Dissimilar Welded Joints of Modified 9Cr-1Mo Steel

This paper presents evaluation methods of 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 mechanical properties across welded joint causes a non-homogeneous strain distribution, and this effect should be taken into account for evaluation of fatigue strength of weld joints. In this study, ‘2-element model’, which is consisted of base metal and weld metal, was employed. Firstly, strain ranges of each element are calculated, and secondly fatigue lives of each element are evaluated. Finally, shorter fatigue life is chosen as fatigue life of the weld joint. Failure position can be also estimated by this model. Evaluation results were compared with experimental data at elevated temperature, and it was shown that they agree well.Copyright

Elaboration of the System Based Code Concept — Activities in JSME and ASME: (3) Guidelines on Structural Reliability Evaluation for FBR

This paper describes the outline of the guidelines on structural reliability evaluation for the passive components of the fast breeder reactor (FBR). The guidelines are now being prepared by the task force for the system based code in the Japan society of mechanical engineers in order to contribute to reducing differences in evaluated structural reliability by evaluators. They consist of five chapters, which are “General rules”, “Reliability evaluation”, “Failure scenario setting”, “Modeling”, and “Failure probability calculation”, respectively. In the chapter of “Reliability evaluation”, the general procedures of reliability evaluation are explained. Detailed procedures at each step are explained in the following chapters in the guidelines. Evaluation procedures for accumulation damage and crack propagation due to creep-fatigue interaction which is a typical degradation phenomenon in FBR are also provided in the appendix of the guidelines. In addition, some examples of random variables are prepared as standard input data for structural reliability evaluation.© 2014 ASME