Takashi Onizawa

Bio: Takashi Onizawa is an academic researcher from Japan Atomic Energy Agency. The author has contributed to research in topics: Creep & Sodium-cooled fast reactor. The author has an hindex of 4, co-authored 13 publications receiving 65 citations.

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

Abstract: 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 © 2010 by ASME

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

Abstract: 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 © 2013 by ASME

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

Abstract: 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 © 2013 by ASME

9%Cr heat resistant steels: Alloy design, microstructure evolution and creep response at 650 °C

Abstract: In this work 9%Cr alloys were designed supported by computational thermodynamic methods. Two sets of alloys were produced: 9%Cr alloys with 0.1%C and 0.05%C and 9%Cr alloys containing ∼0.03% Ti with 0.1%C and 0.05%C (always wt%). Microstructure investigations showed good agreement with the predicted phases of the thermodynamic modeling. The volume fraction of precipitated M 23 C 6 carbides is directly related to the carbon content of the alloys. For Ti-containing alloys the precipitation of nano-sized Ti-rich MX carbonitrides was observed. The microstructure evolution (sub-grain and particle size) during creep at 650 °C/100 MPa was investigated by STEM-HAADF. The sub-grain size evolution and the coarsening of precipitates (MX carbonitrides, M 23 C 6 and Laves phase) were more pronounced for Ti-containing alloys. 9Cr alloys without Ti and with low carbon content presented the highest creep strength of all investigated alloys.

Design Study and R&D Progress on Japan Sodium-Cooled Fast Reactor

Abstract: This paper describes the progress of the design study and research and development (R&D) for the Japan Sodium-cooled Fast Reactor (JSFR) implemented in the “Fast Reactor Cycle Technology Development (FaCT)” project. A sodium-cooled fast reactor with an electric power of 1,500MWe is targeted for commercialization at around 2050, and a demonstration reactor assuming a power output from 500 to 750MWe is planned to start operation at around 2025. R&D on innovative technologies to achieve economic competitiveness and enhance reliability and safety is carried out for the commercialization. A compact reactor vessel without a vessel wall cooling system is pursued in consideration of the wall thickness enough to resist the severest seismic condition. A two-loop cooling system with shortened highchromium steel piping is a crucial feature, and studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being carried out. A double-walled straight tube steam generator is investigated to...

Comparative study of autogenous tungsten inert gas welding and tungsten arc welding with filler wire for dissimilar P91 and P92 steel weld joint

Abstract: Creep strength enhanced ferritic/martensitic 9Cr-1Mo-V-Nb (P91) steel is also designated as ASTM A335 used for out-of-core and in-core (piping, cladding, ducts, wrappers, and pressure vessel) of Gen IV reactors. In present investigation, the dissimilar weld joint of P91 and P92 steel were made using the autogenous tungsten inert gas (TIG) welding with single pass, double side pass and multi-pass gas tungsten arc (GTA) welding with filler wire. Microstructure evolution in sub-zones and mechanical properties of dissimilar welded joints were studied in as-welded and post weld heat treatment (PWHT) condition. Formation of δ-ferrite patches in weld fusion zone and heat affected zones (HAZs) and their influence on the mechanical behaviour of the welded joints were also studied. Presence of higher content of ferrite stabilizer in P92 steel have resulted the formation of δ-ferrite patches in weld fusion zone as well as HAZs. The δ-ferrite was observed in autogenous TIG welds joints. The δ-ferrite patches were formed in as-welded condition and remained in the microstructure after the PWHT. The δ-ferrite patches leads to reduction in Charpy toughness of autogenous TIG welds joint and also lower down the average hardness of weld fusion zone. Peak hardness and poor impact toughness were observed for autogenous TIG welds joint as compared to GTA welds. For microstructure characterization, field-emission scanning electron microscope (FESEM) with energy dispersive spectroscopy (EDS) and optical microscope were utilized.