Takashi Wakai

Creep Strength Evaluation of Welded Joint Made of Modified 9Cr-1Mo Steel for Japanese Sodium Cooled Fast Reactor (JSFR)

This paper describes a proposal of provisional allowable stress for the welded joints made of modified 9Cr-1Mo steel (ASME Gr.91) applicable to the structural design of Japanese Sodium cooled Fast Reactor (JSFR). For the early commercialization of the SFRs, economic competitiveness is one of the most essential requirements. One of the most practical means to reduce the construction costs is to diminish the total amount of structural materials. To meet the requirements, modified 9Cr-1Mo steel has attractive characteristics as a main structural material of SFRs, because the steel has both excellent thermal properties and high temperature strength. Employing the steel to the main pipe material, remarkable compact plant design can be achieved. There is only one elbow in the hot leg pipe of the primary circuit. However, in such a compact piping, it is difficult to keep enough distance between welded joint and high stress portion. In the welded joints of creep strength enhanced ferritic steels including ASME Gr.91 (modified 9Cr-1Mo) steel, creep strength may obviously degrade especially in long-term region. This phenomenon is known as “Type-IV” damage. Though obvious strength degradation has not observed at 550°C yet for the welded joint made of modified 9Cr-1Mo steel, it is proper to suppose strength degradation must take place in very long-term creep. Therefore, taking strength degradation due to “Type-IV” damage into account, the allowable stress applicable to JSFR pipe design was proposed based on creep rupture test data acquired in temperature accelerated conditions. Available creep rupture test data of welded joints made of modified 9Cr-1Mo steel provided by Japanese steel vender were collected. The database was analyzed by region partition method. The creep rupture data were divided into two regions of short-term and long-term and those were individually evaluated by regression analyses with Larson Miller Parameter (LMP). Boundary condition between short-term and long-term was half of 0.2% proof stress of base metal at corresponding temperature. First order equation of logarithm stress was applied. For conservativeness, allowable stress was proposed provisionally considering design factor for each region. Present design of JSFR hot leg pipe of primary circuit was evaluated using the proposed allowable stress. As a result, it was successfully demonstrated that the compact pipe design was assured. For validation of the provisional allowable stress, a series of long-term creep tests were started. In future, the provisional allowable stress will be properly reexamined when longer creep rupture data are obtained. In addition, some techniques to improve the performance of welded joints were surveyed and introduced.Copyright

Development of LBB Assessment Method for Japanese Sodium Cooled Fast Reactor (JSFR) Pipes: 1—Study on the Premise for the Standardization of Assessment Procedure

This paper describes the premise for the standardization of Leak Before Break (LBB) assessment procedure applicable to Japanese Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr-1Mo steel. For the early commercialization of the SFRs, economic competitiveness is one of the most essential requirements. Japan Atomic Energy Agency (JAEA) proposes an attractive plant concept and studies the applicability of some innovative technologies to the plant. One of the most practical means to reduce the construction costs is to reduce the total amount of structural materials. To meet the requirements, modified 9Cr-1Mo steel has attractive characteristics as a main structural material of SFRs, because the steel has both excellent thermal properties and high temperature strength. By employing the steel as the main structural material, remarkably compact plant design can be achieved. Since the ductility and toughness of the steel is relatively inferior to those of conventional austenitic stainless steels, a LBB assessment technique suitable for the pipes made of modified 9Cr-1Mo steel may be required. In addition, since the SFR pipes are mainly subjected to displacement controlled thermal loads, it is expected that fast unstable fracture is unlikely. Taking both material and structural features into account, the framework to establish a precise LBB assessment procedure for SFR pipes must be organized. For the standardization of the LBB procedure, the main investigation items were defined as follows: (1) Approval of the assessment flowchart eliminating uncertainty due to small scale leakage, e.g. self plugging phenomenon and influence of crack surface roughness on leak rate. (2) Proper selection of LBB assessment objects in JSFR. (3) Distinguishment between the matters covered by a design code and LBB, i.e. assumption of initial flaw(s). (4) Development of creep and/or fatigue crack extension assessment technique, including collection of necessary material data. (5) Development of unstable fracture assessment technique. (6) Development of leak rate evaluation technique. (7) Characterization of loads for LBB assessment. (8) Standardization of the procedure as the Japan Society of Mechanical Engineers (JSME) code.Copyright

A Study for Proposal of Welded Joint Strength Reduction Factors of Modified 9Cr-1Mo Steel for Japan Sodium Cooled Fast Reactor (JSFR)

This paper proposes provisional welded joint strength reduction factors (WJSRF) of modified 9Cr-1Mo steel (ASME Gr.91) applicable to the structural designing of “Japan sodium cooled fast reactor (JSFR)”. In the welded joints of creep strength enhanced ferritic steels including modified 9Cr-1Mo steel, creep strength may obviously degrade especially in long-term region. This phenomenon is known as “Type-IV” damage. The authors had proposed provisional allowable stress for the welded joints made of the steel in PVP 2010 conference, taking creep strength degradation due to “Type-IV” damage into account. Available creep rupture data of the welded joints made of the steel provided by Japanese steel venders were collected. The temperature range was from 500 to 650°C. The database was analyzed by stress range partitioning method. The creep rupture data were divided into two regions of short-term and long-term and those were individually evaluated by regression analyses with Larson Miller Parameter (LMP). The difference in the creep failure mechanisms between short-term and long-term regions is taken into account in this method. Boundary between these regions was half of 0.2% proof stress of the base metal at corresponding temperature. First order polynomial equation of logarithm stress was applied. For conservativeness, allowable stress was proposed provisionally considering design factor for each region. JSME (Japan Society of Mechanical Engineers) published a revised version of the elevated temperature design code in last year. Modified 9Cr-1Mo steel was officially registered in the code as a new structural material for sodium cooled fast reactors. The creep rupture curve for the base metal of the steel was standardized by employing stress range partitioning method, same as for the welded joint. However, second order polynomial equation of logarithm stress was applied in the analysis for the base metal. In addition, the creep rupture data obtained at 700°C were included in the database and data ruptured in very short term, i.e. smaller than 100 hours, were excluded from the analysis. Thus, there are some differences between the procedures to determine the creep rupture curves for base metal and welded joint made of modified 9Cr-1Mo steel. This paper discusses the most feasible procedure to determine the creep rupture curve of the welded joint of the steel by performing some case studies to focus on physical adequacy and harmonization with the determination procedure of the creep rupture curve for the base metal. Then, the WJSRF are provisionally proposed based on the design creep rupture stress intensities. In addition, the design of JSFR pipes was reviewed taking WJSRF into account.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

A Screening Method for Prevention of Ratcheting Strain Derived From Movement of Temperature Distribution

In this paper, we simplify the existing method and propose a screening method to prevent thermal ratcheting strain in the design of practical components. The proposed method consists of two steps to prevent the continuous accumulation of ratcheting strain. The first step is to determine whether all points through the wall thickness are in the plastic state. This is based on an equivalent membrane stress, which comprises the primary stress and the secondary membrane stress. When the equivalent stress exceeds the yield strength in some regions of the cylinder, the axial lengths of these regions are measured for the second step. The second step is to determine whether the accumulation of the plastic strain saturates. For this purpose, we define the screening criteria for the length of the area with full section yield state. When this length is sufficiently small, residual stress is generated in the direction opposite to the plastic deformation direction. As a result of residual stress, further accumulation of the plastic deformation is suppressed, and finally shakedown occurs. To validate the proposed method, we performed a set of elastoplastic finite element method (FEM) analyses, with the assumption of elastic perfectly plastic material. Not only did we investigate about the effect of the axial length of the area with full section yield state but also we investigated about effects of spatial distribution of temperature, existence of primary stress, and radius thickness ratio.

Observation and Evaluation of Plastic Collapse for Double-Notch Pipe

The plastic collapse behavior and strength were investigated for an austenitic stainless steel pipe with two 90° through-wall notches perpendicular to the pipe axis direction. Double-notch specimens with various notch separation distances were coated with photo-plastic film. Arbitrary combined axial tensile and bending loads were applied to the specimens. Changes in the photoplastic fringe pattern were observed during the tests to investigate the plastic collapse behavior. The plastic collapse strength was evaluated using a model based on an elastic-perfectly plastic body. The photo-plastic fringe patterns at the experimental plastic collapse point differed based on the loading history. Thus, the plastic collapse behavior depends on the loading history. In addition, the plastic collapse strength differed based on the loading history and hardly depended on the notch separation distance. The experimental plastic collapse occurred before reaching the theoretical plastic point for only some pure tension loading tests. Thus, the model analysis based on an elastic-perfectly plastic body used in this study might give an unconservative estimate for the plastic collapse of a stainless steel pipe subjected to a pure tension load.

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

Proposal of the Screening Method for Prevention of the Accumulation of the Ratcheting Strain Derived From the Movement of the Temperature Distribution

The prevention of excessive deformation by thermal ratcheting is important in the design of high-temperature components of fast breeder reactors (FBR). This includes evaluation methods for a new type of thermal ratcheting caused by a traveling temperature distribution. Igari et al. [1] proposed a mechanism-based evaluation method to evaluate thermal ratcheting caused by temperature distributions traveling long and short distances.In this paper, we simplify the existing method and propose a screening method to prevent thermal ratcheting strain in the design of practical components. The proposed method consists of two steps to prevent the continuous accumulation of ratcheting strain.The first step is to determine whether all points through the wall thickness are in the plastic state. This is based on an equivalent stress, which comprises the primary stress, the thermal membrane stress, and the thermal bending stress. When the equivalent stress is less than the yield strength of the cylinder material, overall plastic deformation through the wall thickness does not occur. When the equivalent stress exceeds the yield strength in some regions of the cylinder, the ranges of these regions are measured for the second step. To prevent the acceleration of the plastic deformation due to creep, we define the upper limit of the equivalent stress based on the relaxation strength, Sr.The second step is to determine whether the accumulation of the plastic strain saturates (i.e. if shakedown occurs). For this purpose, we define the screening criteria for the range of the plastic region. When the range of the plastic region is sufficiently small, residual stress is generated in the direction opposite to the plastic deformation direction. As a result of residual stress, further accumulation of the plastic deformation is suppressed, and finally shakedown occurs. If the range of the plastic region exceeds the defined criteria, a more detailed evaluation method (e.g. inelastic finite element analysis) may be used for the component design.To validate the proposed method, we performed a set of elasto-plastic finite element method (FEM) analyses, with the assumption of elastic perfectly plastic material.Copyright

Effect of Rotational Stiffness Evolution at Crack Part on Critical Crack Size in SFR

This paper describes the efficiency of the deployment of rotational stiffness evolution model in the critical crack size evaluation for Leak Before Break (LBB) assessment of Sodium cooled Fast Reactor (SFR) pipes. The authors have developed a critical crack size evaluation method for the thin-walled large diameter pipe made of modified 9Cr-1Mo steel. In this method, since the SFR pipe is mainly subjected to displacement controlled load caused by thermal expansion, the stress at the crack part is estimated taking stiffness evolution due to crack into account. The stiffness evolution is evaluated by using the rotational spring model. In this study, critical crack sizes for several pipes having some elbows were evaluated and discuss about the effect of the deployment of the stiffness evolution model at the crack part on critical crack size. If there were few elbows in pipe, thermal stress at the crack part was remarkably reduced by considering the stiffness evolution. In contrast, in the case where the compliance of the piping system was small, the critical crack size could be estimated under displacement controlled condition. As a result, the critical crack size increases by employing the model and LBB range may be expected to be enlarged.Copyright