Tai Asayama

Evaluation Procedures for Irradiation Effects and Sodium Environmental Effects for the Structural Design of Japanese Fast Breeder Reactors

In the structural design of fast breeder reactors, irradiation effects and sodium environmental effects on structural materials have to be taken into account. In this paper, firstly, an evaluation procedure for irradiation effects on the mechanical properties of 316FR (FBR Grade 316 stainless steel), which is a newly developed stainless steel for the Japanese demonstration fast breeder reactor, is proposed. The procedure gives a limit of accumulated fast neutron fluence (E>0.1 MeV) as a function of temperature, so that the minimum tensile fracture elongation of 10 percent, which is the threshold for material to stay ductile, is maintained. Furthermore, the procedure determined a creep life reduction factor and a creep rate increase factor as a function of accumulated thermal neutron fluence (E<0.4 eV), within the limitation of the accumulated fast neutron fluence, to account for the creep life reduction and the increase of creep rate due to irradiation. Secondly, an evaluation procedure for sodium environmental effects on the integrity of 316FR and modified 9Cr-1Mo steel was proposed. It gave a corrosion allowance as a function of temperature, oxygen content, and service time, based on corrosion tests. It determined that no correction factors that correspond to sodium environment on design allowable stresses, etc., are needed, because no adverse effects of sodium on the mechanical properties of 316FR and modified 9Cr-1Mo steel were to be expected in the service conditions of FBRs. Both the procedures have been incorporated into the Japanese Elevated Temperature Structural Design Guide for Demonstration Fast Breeder Reactor.

An Overview of Creep-Fatigue Damage Evaluation Methods and an Alternative Approach

Renewed interest in elevated temperature nuclear reactors has occasioned a reassessment of creep-fatigue damage evaluation methods. Points to be improved in the current methods employed in Subsection NH of the ASME BP either directly through inelastic analysis or indirectly through manipulation of elastic analyses. Second, because the test specimen itself incorporates the hardening, softening and aging effects of the structure it represents, it is not necessary do rely on theoretical modeling of these effects in an artificial separate accounting of creep and fatigue damage.Copyright

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: (1) Overview

This paper presents the main features of 2012 Edition of JSME (Japan Society of Mechanical Engineers) Code for Design and Construction of Fast Reactors (JSME FR Code). The first edition of the JSME FR Code was published in 2005 based on the requirements that had been applied to the Japanese Prototype Fast Breeder Reactor Monju. The latest 2012 Edition incorporated 316FR stainless steel and Mod.9Cr-1Mo steel of which application to the Japanese demonstration fast reactors is expected. 316FR is a low-carbon nitrogen added fast reactor grade 316SS which has been developed in Japan. Allowable stresses up to 300,000 hours along with various material properties equations from which they have been derived have been codified. The applicability of the code rules on elevated temperature design to the new materials has been extensively investigated. Focus was on the creep-fatigue interaction evaluation methods, and predictability in long-term regions was carefully demonstrated based on material tests. Various structural tests were also conducted to verify that sufficient margins are maintained for the newly implemented materials. The JSME FR Code will be further upgraded in the 2016 and codes such as fitness-for-service code and leak-before-break evaluation code and a guideline for reliability evaluation of static components will be developed utilizing the System Based Code concept.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: (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 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

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

Probabilistic Prediction of Crack Depth Distributions Observed in Structures Subjected to Thermal Fatigue

The application of risk-based technologies not only to in-service inspections but also to the design of components and systems, encompassing a plant life-cycle, is the way to be pursued for the improvement of design of new reactors such as fast breeder reactors. When doing so, it is necessary to develop an analytical method that is capable of estimating failure probabilities without a failure database that can only be established on the long-time accumulation of operational experiences. The prediction method should estimate failure probabilities based on actual mechanisms that cause failure. For this purpose, this study developed a structural reliability evaluation method using probabilistic prediction of crack depth distributions for thermal fatigue, which is one of representative failure modes to be prevented in components of nuclear plants. This method is an extension of probabilistic fracture mechanics approach but is capable of modeling crack initiation, crack propagation, and crack depth density distribution at a given cycle. To verify the methodology, crack depth distribution observed in thermal fatigue test specimens was evaluated, and it was shown that the method could reproduce the observed crack depth distributions fairly well. This is considered to explore the possibility that probabilistic fracture mechanics approach can be verified by experiments, which was deemed impossible so far. Further improvement such as explicit implementation of interaction mechanisms between adjacent cracks will allow this methodology to be applied to the procedure of optimization of in-service inspection planning, as well as to the optimization of safety factors in component design of nuclear plants.

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.