Masayuki Sukekawa

Exploratory research on creep and fatigue properties of 9Cr-steels for the steam generator of an FBR

Abstract Research on the applicability of 9Cr-steels to the steam generator of the demonstration fast breeder reactor was performed by the Subcommittee of the Japan Welding Engineering Society as a four-year program from 1985. In this program, exploratory tests, which included tensile, creep rupture and low-cycle fatigue tests, were conducted on three kinds of 9Cr-steels (Mod.9Cr-1Mo, 9Cr-1Mo-V-Nb, and 9Cr-2Mo) and their weldments. This paper describes the summary of results obtained in this program. Among the tested 9Cr-steels, Mod.9Cr-1Mo steel shows the best creep rupture strength and its weldment indicates almost the same level of creep rupture strength and the base metal at 500 and 550°C. The low-cycle fatigue properties of Mod.9Cr-1Mo steel is also discussed from its relation to the tensile properties.

Experimental Investigation of Strain Concentration Evaluation Based on the Stress Redistribution Locus Method

Evaluating the local strain in structural discontinuities is an important technology in high temperature design of fast reactor components because the failure mode in high-temperature fatigue or creep fatigue damage usually results from the crack initiation and growth from such locally high strained areas. A rationalized method of evaluating strain concentration that was experimentally studied is discussed. The stress redistribution locus (SRL) method had been proposed to improve the accuracy with which local stress and strain can be evaluated in the structural discontinuities. This method is based on the concept that the locus of stress redistribution from an elastic to an inelastic state, or that during relaxation, strongly depends on the structure, and the locus almost coincides with the locus obtained through elastic-creep analysis. High-temperature fatigue tests of circumferentially notched specimens were conducted accompanied by the measurement of local strain carried out with a capacitance-type strain gauge. The measured strain was compared with the predictions made with SRL, and the methods accuracy was evaluated. SRL improved the accuracy of inelastic strain estimation while remaining reasonably conservative in comparison with Neubers rule, which is used in high temperature design codes.

A Comparative Study of Negligible Creep Curves for Rational Elevated Temperature Design

In Japanese elevated temperature standard, creep considering design is required for all ferrite steels applied over 375°C and all austenitic stainless steels applied over 425°C regardless of the operating time. On the other hands, ASME Sec.III Subsection NH, RCC-MR and R5 provide the additional rules to determine the negligible creep range. In those standards, each material is evaluated as non-creep considering design region, although there are varieties of applicable materials and the rules to settle the negligible creep range in each standard. 316FR and Mod.9Cr-1Mo are candidate materials of Japan sodium-cooled fast reactor (JSFR), and those high creep resistant properties extend the negligible creep damage area over the conventional temperature limits. Extension of non-creep design area widens design windows and simplifies the creep analysis procedure. To reply those requirements, authors already proposed original negligible criterion and discussed about it. In this paper, we recall the backgrounds of the negligible creep criterion which have already been proposed. Then the negligible creep criterion and relating property in each standard were compared. For estimating the evaluation procedure of each criterion, the common material properties used in “Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)” were applied to each standard’s criteria. All standards have the negligible creep curves/regions for type 18Cr-8Ni steels and type 18Cr-12Ni-2.5Mo steels, although ASME Sec.III Subsection NH defines just the criteria of negligible creep for the rule of inelastic strain limits. On the diagram of temperature-negligible creep time, the negligible creep curves of 316L(N)(1S) in RCC-MR and R5 exist between those of SUS316 and 316FR in FDS. The negligible creep regions defined in all standards are similar for austenitic stainless steels, although those criteria are different. Comparison of the negligible creep curves by each criterion with FDS’s material properties indicated that the criterion in FDS provides the most conservative curve. In case of Mod.9Cr-1Mo steel, FDS and R5 provide relationship between temperatures and time for estimating the negligible creep time. ASME Sec. III Subsection NH provides only procedures and has no practical allowable values, and RCC-MR doesn’t have the negligible creep curve. Comparison of the negligible creep curves in each criterion with FDS’s material properties indicated that FDS’s criterion allows the longest negligible creep. The negligible creep criteria in ASME Sec.III Subsection NH, RCC-MR and R5 are not practicable for Mod.9Cr-1Mo. On the other hands, FDS criterion raises the temperature limits from conventional 375°C to about 425°C even when the components designed lifetime is 60years. Sensitivities to the difference of criteria and material properties were discussed and concluded that negligible creep curve is strongly dependent on the combination of criteria and material properties. Some evaluations proved that the negligible creep curves in FDS are moderately conservative and practicable.Copyright

Main Features of JSME Design and Construction Code for Fast Reactors

Since its foundation in 1997, the Main Committee on Power Generation Facility Codes, MC-PGFC, of the Japan Society of Mechanical Engineers, JSME, has issued a number of nuclear codes including the rules on design and construction and the rules on fitness-for-service for nuclear power plants. Some of these JSME nuclear codes have been endorsed by the regulatory body, and are now utilized in the regulatory processes of the actual plants. Among these nuclear codes recently published is the “rules on design and construction for fast reactors”. It includes as its main body design rules on class 1 components for elevated temperature services. This paper overview the main features of the code.Copyright

A Rational Identification of Creep Design Area Using Negligible Creep Curves

For extension of non-creep design area and simplification of design procedures, a rational identification method of creep design area by negligible creep (NC) curves was studied. NC curves of six kinds of austenite stainless and ferrite steels for fast reactors were determined based on domestic material data. NC curves provide the relation between temperature and time that does not induce damageable creep strain under the constant stress 1.5Sm (Sm: design stress intensity). In existing Japanese design guides, non-creep design area is severely restricted by constant upper temperature limit for austenite stainless steel and ferrite steel. In the case of 316FR steel and SUS410J3, which are candidate materials of Japanese commercialized fast reactors and have excellent material property, this limit can be extended by NC curve concept considering the duration of high temperature operation. NC curves under secondary stress considering stress relaxation were also studied. However, rationalization effect was insufficient whereas evaluation process was too complex. Therefore, at the present stage, NC curves at constant stress level 1.5Sm were adopted to identify creep design area. The concept of NC curve was introduced into the interim structural design guide for commercialized fast reactors in Japan to simplify the creep design of fast reactor systems. Utilizing these curves, non-creep design becomes possible for components operated at comparatively lower temperature in normal condition.Copyright