Kazuyuki Tsukimori

A Study on Fatigue and Creep-Fatigue Life Assessment Using Cyclic Thermal Tests With Mod.9Cr-1Mo Steel Structures

In a component design at elevated temperature, fatigue and creep-fatigue is one of the most important failure modes, and fatigue and creep-fatigue life assessment in structural discontinuities is important issue to evaluate structural integrity of the components. Therefore, to assess the failure estimation methods, cyclic thermal loading tests with two kinds of cylindrical models with thick part were performed by using an induction heating coil and pressurized cooling air. In the tests, crack initiation and propagation processes at stress concentration area were observed by replica method. Besides those, finite element analysis (FEA) was carried out to estimate the number of cycles to failure. In the first test, a shorter life than predicted based on axisymmetric analysis. Through the 3 dimensional FEA, Vickers hardness test and deformation measurements after the test, it was suggested that inhomogeneous temperature distribution in hoop direction resulted in such precocious failure. Then, the second test was performed after improvement of temperature distribution. As a result, the crack initiation life was in a good agreement with the FEA result by considering the short term compressive holding. Through these test and FEA results, fatigue and creep-fatigue life assessment methods of Mod.9Cr-1Mo steel including evaluation of cyclic thermal loading, short term compressive holding and failure criterion, were discussed. In addition it was pointed out that the temperature condition should be carefully controlled and measured in the structural test with Mod.9Cr-1Mo steel structure.Copyright

Proposals of Guidelines for High Temperature Structural Design of Fast Reactor Vessels

Main loadings of reactor vessels in fast reactor plants are thermal stresses induced by fluid temperature change at transient operation. Structures respond to them with elastic plastic creep deformation under high temperature conditions. It can induce incremental deformation and creep fatigue crack at critical portions around the sodium surface, thermal stratification layer and core support structures. Those phenomena are so complex that design evaluation becomes sometimes too conservative. In order to achieve precise high temperature design for realizing compact reactor vessels of fast reactor plants, such guidelines are proposed as for thermal load modeling, structural analysis and strength evaluation. This paper gives the brief summary of these guidelines. GUIDELINES FOR THERMAL LOAD MODELING: One of main difficulties of thermal load modeling is their inducement mechanism by interaction between thermal hydraulic and structural mechanics. Design evaluation requires envelope load conditions with considering scatter of design parameters. Proposed guidelines enable precise load modeling by grasping sensitivities of thermal stress to design parameters including thermal hydraulic ones. GUIDELINES FOR INELASTIC DESIGN ANALYSIS: Guidelines are proposed to apply inelastic analysis methods for design of reactor vessels. There are so many influence parameters in inelastic analysis that conservative and unique solutions are hardly found. To overcome such difficulties, mechanism and main parameters of inelastic behaviors of reactor vessels were clarified. Guidelines give conservative results within the same mechanism as expected reactor vessels. HIGH TEMPERATURE STRENGTH EVALUATION METHOD: Incremental deformation and creep fatigue strength evaluation methods were proposed. Accumulated strain is limited within no influence of fatigue and creep-fatigue strength. Taking design conditions of reactor vessels into account, creep fatigue evaluation considers strain concentration and an intermediate stress hold effect on creep-fatigue strength. Influences of thermal aging were also confirmed.Copyright

Development of 2012 Edition of JSME Code for Design and Construction of Fast Reactors: (6) Design Margin Assessment for the New Materials to the Rules

New 2012 edition of JSME code for design and construction of fast reactors (FRs code) was published by Japan society of mechanical engineers (JSME). Main topic of the current JSME FRs code 2012 edition is registration of the two new materials, 316FR and Mod.9Cr-1Mo steel. Besides the allowable strength values and material properties were standardized for the registration, the design margins for the new materials to the rules for the components and piping serviced at elevated temperature described in the JSME FRs code were assessed. To confirm the design margins, a series of the assessment program for the new materials to the conventional design rules was performed using the evaluation of the experimental data and finite element analysis. Namely, the design margin including the evaluation procedure of creep-fatigue damage, strain range and the others were assessed based on the background concept of the conventional JSME FRs code. Since a number of the evaluation procedures described in the JSME FRs code were investigated, a several topical assessments of these are reported in this paper. Besides the assessed results of the evaluation of the accumulated creep-fatigue damage and enhanced creep strain are reported, the assessments results of the design margin including the concept of the elastic follow-up originally applied in the JSME FRs code were covered in this paper. Through these assessments, the enough design margins for new materials to the rules were confirmed.Copyright

Development of LBB Assessment Method for Japan Sodium Cooled Fast Reactor (JSFR) Pipes (5): Crack Growth Assessment Method for Pipes Made of Mod.9Cr-1Mo Steel

For sodium pipes of Japan Sodium cooled Fast Reactor (JSFR), the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy Leak-Before-Break (LBB). The vessels of JSFR are connected by thin wall pipes with a large diameter made of modified 9Cr-1Mo steel and the internal pressure of the pipes is very low. Modified 9Cr-1Mo steel has relatively large yield stress and small work hardening coefficient compared to the austenitic stainless steels which are currently used in the conventional plants. Therefore, these material characteristics of modified 9Cr-1Mo steel must be taken into account in LBB assessment, as well as geometrical and structural features of JSFR pipes. In order to demonstrate LBB aspects of the JSFR pipes, the authors have proposed a LBB assessment flowchart and developed assessment methods of unstable fracture and crack opening displacement (COD) for the thin wall pipes with large diameter made of modified 9Cr-1Mo steel. This paper studies the master curve to estimate the crack length when a postulated initial crack unexpectedly grows and penetrates the pipe thickness. In order to obtain the fatigue crack and creep crack growth characteristics of modified 9Cr-1Mo steel pipes, fatigue crack and creep crack growth tests were conducted using compact tension (CT) specimens and crack growth rates for both fatigue and creep at elevated temperature were obtained. Based on the obtained material characteristics and the results of a series of crack growth calculations, a relationship between the penetrated crack length and the ratio of membrane to total stress, so called as master curve, was proposed. In this study, master curves were proposed for pipes made of modified 9Cr-1Mo steel as a function of pipe geometry, i.e. the ratio of radius to thickness.Copyright

3D RF‐ECT SIMULATIONS OF ISI OF U‐BEND SG TUBES IN FBR

The paper focus on 3‐dimensional finite element simulations of the steam generator inspection using remote field eddy current testing (RF‐ECT) in order to asses the performance of the in‐service inspection in SG tubes. It is evaluated the influences of the SG tubes U‐bend curvature on the eddy current sensor signal by taking into account a 3D model in which also multiple SG tubes are located close to each other, as in the nuclear reactor.

Theoretical Modeling of Creep Behavior of Bellows and Some Applications

The use of bellows expansion joints is an effective method to rationalize various piping systems in industry. In the structural design, the requirements for preventing failures such as ratcheting, fatigue, and buckling should be satisfied. The mechanisms of some failure modes of bellows are different from those of vessels and piping components, which makes it difficult to estimate the behaviors. In the case of high-temperature operation, creep behavior of bellows should be considered. In this paper, a simplified theoretical modeling of creep behavior of bellows is presented. The formulation is developed by using Nortons law for creep property of bellows material and assuming meridional bending stress is dominant. According to this modeling, bellows convolution dimensions are considered directly. The excessive creep deformation problem of bellows under internal pressure and the elastic follow-up behavior problem of a piping system with bellows expansion joints are examined as the applications of this modeling. The results are compared with detailed analysis results by FEM, and the applicability and the validity of this modeling is discussed.

Ultimate Strength of a Thin Wall Elbow for Sodium Cooled Fast Reactors Under Seismic Loads

With a purpose of identifying the failure mode and associating the ultimate strength of piping components against seismic integrity, many kinds of failure tests have been conducted for thick wall piping for light water reactors (LWRs). However, there are little failure test data on thin wall piping for sodium cooled fast reactors (SFRs). In this paper, a series of failure tests on thin wall elbows for SFRs is presented. Based on the tests, the failure mode of a thin wall piping component under seismic loads was identified to be fatigue. The safety margin included in the current design methodology was clarified quantitatively.

Structural Materials and Code Development for Japanese Sodium-Cooled Fast Reactors

This paper describes the latest status on the development of elevated temperature materials and structural codes for Japanese sodium-cooled fast reactors (SFRs). Based on the extensive research and development activities in the last decades in Japan, two materials, 316FR and Modified 9Cr-1Mo steels were recently incorporated into the 2012 Edition of Fast Reactor Design and Construction Code of the Japan Society of Mechanical Engineers (JSME). Structural design methodologies are continuously being improved towards the next major revision planed in 2016 Edition where methodologies for a 60-year design of Japanese demonstration fast reactor will be provided. Codes and guidelines for fitness-for-service, leak-before-break evaluation and reliability assessment are concurrently being developed utilizing the System Based Code concept aiming at establishing an integrated code system that encompasses a life cycle of SFRs.Paper published with permission.Compilation Copyright

Development of Structural Codes for JSFR Based on the System Based Code Concept

This paper overviews the current status of the ongoing research and development as well as activities for codification of structural codes for the Japan Sodium Cooled Fast Reactor (JSFR), the demonstration fast reactor which is in the phase of conceptual study. Not only the design and construction code which has been published and updated on a regular basis, codes on welding, fitness-for-service, leak-before-break evaluation as well as the guidelines for structural reliability evaluation are being developed. The basic strategy for the development is to fully take advantage of the favorable technical characteristics associated with sodium-cooled fast reactors; the codes will be developed based on the System Based Code concept, a concept that materializes code rules that are most suitable to the reactor types they are applied to. The above mentioned set of codes are planned to be published from the Japan Society of Mechanical Engineers in 2016.Copyright

Study on Piping Response Under Multiple Excitation Validation for Elastic-Plastic Analysis of Piping

Piping in a nuclear power plant is usually laid across several floors of a single building or adjacent buildings, and is supported at many points. As the piping is excited by a large earthquake through multiple supporting points, seismic response analysis by multiple excitations within the range of plastic deformation of piping material is necessary to obtain the precise seismic response of the piping. The verification of the dynamic analysis method of piping under an elastic domain, which is excited by multiple seismic inputs, was performed in our study last year and the correspondence of a piping response between an analysis and an experiment have been confirmed [17][18]. However, few experiments under plastic deformation conditions have been performed to verify the validity of multiple excitation analysis under a plastic deformation range. To obtain better understanding of the behavior of piping under a large seismic input, it is important to investigate the seismic response by multiple excitations and to verify the validity of the analytical method by multiple excitation experiments.This paper reports the validation results of the seismic elastic-plastic time history analysis of piping compared with the results of the shaking test of a 3-dimensional piping model under a plastic deformation range using triple uni-axial shake table. Three directional strains from the analysis and the experiments were compared in order to validate the analysis method. As a result, it is confirmed that the elastic-plastic analysis by time history excitation shows good agreement with the test results.Copyright