Yuji Nagae

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 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

Strength of 316FR Joints Welded by Type 316FR/16-8-2 Filler Metals

Type 316 stainless steel with low-carbon and medium-nitrogen contents called 316FR stainless steel is a candidate structural material for reactor vessels and internals of future-generation fast breeder reactors (FBRs). The reactor vessel cannot be manufactured from rolled or forged steel, but can be built at reasonable cost by welding rolled steel plates. In this manufacture approach, the reliability of the welded joint must be indicated. Two types of filler metals are candidates for 316FR welded joints: types 316FR and 16-8-2 filler metals. The chemical composition of type 316FR filler metal is close to that of the stainless steel; type 16-8-2 filler metal contains lower amounts of Ni, Cr, and Mo than that of the stainless steel. This study evaluated the need to consider the welded joint strength reduction factors in 316FR welded joints under design of future-generation FBRs. To this end, the tensile and creep strengths of types 316FR and 16-8-2 weld metals were measured, and the effect of δ-ferrite in weld metals was evaluated in creep strength tests of 316FR welded joints. In tensile and creep strengths of 316FR welded joints welded by both metal types, the welded joint strength reduction factors were immaterial. The creep strength of 316FR welded joints was negligibly affected by δ-ferrite levels from 4.1 to 7.0 ferrite number (FN) in the Welding Research Council-1992 diagram. Furthermore, the tensile and creep strengths of 316FR welded joints by two methods (gas tungsten arc welding (GTAW) and shielded metal arc welding (SMAW)) were the same. Therefore, the tensile and creep strengths of 316FR welded joints in above condition are ensured the reliability of similar to 316FR stainless steels.

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.

Evaluation Method of Creep-Fatigue Life for 316FR Weldment

The most important failure mode to be prevented is creep-fatigue at elevated temperatures in fast reactors. 316FR stainless steel is a candidate material for the reactor vessel and internal structures. A method to evaluate creep-fatigue life, based on the time fraction rule, has been already developed in base metal of 316FR stainless steel. Development of procedure in evaluating creep-fatigue life is also necessary for the weldment of 316FR stainless steel by similar fillers or 16-8-2 fillers. Compared between mechanical properties of weldment and those of base metal, strength-reduction factors for weldment have been evaluated. Strength-reduction factor for fatigue has been proposed. It is considered that strength-reduction factor for creep strength is not necessary. Creepfatigue life could be evaluated in the same way for weldments of similar fillers and 16-8-2 fillers, because a difference in mechanical properties between both filler metals is negligible. Creep-fatigue life by the time fraction rule using analytical relaxation curve for weldments were compared with experimental data, and a method to evaluate creep-fatigue life for the weldments of 316FR stainless steel has been proposed.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

Development of 2012 Edition of JSME Code for Design and Construction of Fast Reactors: (4) Creep–Fatigue Evaluation Method for Modified 9Cr–1Mo Steel

This paper describes a creep–fatigue evaluation method for modified 9Cr–1Mo steel, which has been newly included in the 2012 edition of the JSME code for design and construction of fast reactors. In this method, creep and fatigue damages are evaluated on the basis of Miner’s rule and the time fraction rule, respectively, and the linear summation rule is employed as the failure criterion. Investigations using material test results are conducted, which show that the time fraction approach can conservatively predict failure life if margins on the initial stress of relaxation and the stress relaxation rate are embedded. In addition, the conservatism of prediction tends to increase with time to failure. Comparison with the modified ductility exhaustion method, which is known to have good failure life predictability in material test results, shows that the time fraction approach predicts failure lives to be shorter in long-term strain hold conditions, where material test data is hardly obtained. These results confirm that the creep–fatigue evaluation method in the code has implicit conservatism.Copyright

Extrapolation of Creep Strength by Fracture Energy for 316FR Stainless Steel at 823K

316FR stainless steel is a candidate material to be used for a reactor vessel and internals for fast reactors with a design life of 60 years at an operating temperature of 823K. This paper describes an extrapolation approach based on fracture energy for calculating creep strength. A change in fracture energy is assumed to be expressed as a power-law function of time to failure and energy density rate. The energy density rate is calculated using initial stress, rupture elongation, and time to rupture. It is important to evaluate a change in rupture elongation for the extrapolation of creep strength at 823K. The time to rupture at 823K is estimated and extrapolated on the basis of the fracture energy approach. This paper shows the validity of extending the design life to 60 years by using the Larson–Miller parameter compared with the estimation by the fracture energy approach.Copyright

Nondestructive Evaluation of Neutron Irradiation Damage on Austenitic Stainless Steels by Measurement of Magnetic Flux Density

Magnetic flux densities for neutron irradiated specimens of austenitic stainless steels, SUS304 and Fast Breeder Reactor grade type 316 (316FR), were measured by using a flux gate (FG) sensor to investigate the nondestructive evaluation method of irradiation damage parameters, dose and He content. Specimens were irradiated in each one of the experimental fast reactor JOYO, the Japan Materials Testing Reactor, and the Japan Research Reacter-3M (JRR-3M), or in both of JRR-3M and JOYO (coupling irradiation). Irradiation in various reactors and the coupling irradiation provided irradiation conditions which could be hardly obtained by irradiation in a single reactor. The range of dose, He content and irradiation temperature of the neutron irradiated samples studied in this paper were 0.01–30 displacement per atom (dpa), 1.0–17 appm and 470–560 °C, respectively. Magnetic flux density increased with dose although there may be a threshold dose for magnetic property to change between 2 and 5 dpa for 316FR. This result shows the possibility of nondestructive evaluation of dose by measuring magnetic flux density by an FG sensor. On the other hand, magnetic flux density did not depend on He content.Copyright