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.

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.

In a component design at elevated temperature, fatigue and creep-fatigue is one of the most important failure modes, and fatigue and creep-fatigue life assessment in structural discontinuities is important issue to evaluate structural integrity of the components. Therefore, to assess the failure estimation methods, cyclic thermal loading tests with two kinds of cylindrical models with thick part were performed by using an induction heating coil and pressurized cooling air. In the tests, crack initiation and propagation processes at stress concentration area were observed by replica method. Besides those, finite element analysis (FEA) was carried out to estimate the number of cycles to failure. In the first test, a shorter life than predicted based on axisymmetric analysis. Through the 3 dimensional FEA, Vickers hardness test and deformation measurements after the test, it was suggested that inhomogeneous temperature distribution in hoop direction resulted in such precocious failure. Then, the second test was performed after improvement of temperature distribution. As a result, the crack initiation life was in a good agreement with the FEA result by considering the short term compressive holding. Through these test and FEA results, fatigue and creep-fatigue life assessment methods of Mod.9Cr-1Mo steel including evaluation of cyclic thermal loading, short term compressive holding and failure criterion, were discussed. In addition it was pointed out that the temperature condition should be carefully controlled and measured in the structural test with Mod.9Cr-1Mo steel structure.

Main loadings of reactor vessels in fast reactor plants are thermal stresses induced by fluid temperature change at transient operation. Structures respond to them with elastic plastic creep deformation under high temperature conditions. It can induce incremental deformation and creep fatigue crack at critical portions around the sodium surface, thermal stratification layer and core support structures. Those phenomena are so complex that design evaluation becomes sometimes too conservative. In order to achieve precise high temperature design for realizing compact reactor vessels of fast reactor plants, such guidelines are proposed as for thermal load modeling, structural analysis and strength evaluation. This paper gives the brief summary of these guidelines. GUIDELINES FOR THERMAL LOAD MODELING: One of main difficulties of thermal load modeling is their inducement mechanism by interaction between thermal hydraulic and structural mechanics. Design evaluation requires envelope load conditions with considering scatter of design parameters. Proposed guidelines enable precise load modeling by grasping sensitivities of thermal stress to design parameters including thermal hydraulic ones. GUIDELINES FOR INELASTIC DESIGN ANALYSIS: Guidelines are proposed to apply inelastic analysis methods for design of reactor vessels. There are so many influence parameters in inelastic analysis that conservative and unique solutions are hardly found. To overcome such difficulties, mechanism and main parameters of inelastic behaviors of reactor vessels were clarified. Guidelines give conservative results within the same mechanism as expected reactor vessels. HIGH TEMPERATURE STRENGTH EVALUATION METHOD: Incremental deformation and creep fatigue strength evaluation methods were proposed. Accumulated strain is limited within no influence of fatigue and creep-fatigue strength. Taking design conditions of reactor vessels into account, creep fatigue evaluation considers strain concentration and an intermediate stress hold effect on creep-fatigue strength. Influences of thermal aging were also confirmed.

Spontaneous vapor explosion can occur when a layer of the high temperature molten material lies on the water pool or on the moisture floor. This is so-called base-triggered vapor explosion. The base-triggered vapor explosion is supposed to occur in the case of a severe accident in a nuclear reactor and in other industrial facilities. It is very important to clarify the occurrence condition and possibility of the base-triggered vapor explosion from the viewpoints of the prediction and the prevention of the vapor explosion. In order to evaluate the occurrence conditions and to clarify the micro-mechanism of the base-triggered vapor explosion, the experimental apparatus to observe the base-triggered vapor explosion from the bottom of the floor to above is designed and constructed. The experiments using U-Alloy95 as a simulant material are conducted. Consequently, the microscopic behavior at the interface between the molten material and water can be observed in detail with this experimental apparatus. The PIV analysis is conducted to the visual observation data obtained on the experiments in order to evaluate the overall behavior of molten material. The digital auto-correlation method is also applied to the visual observation data in order to evaluate the interfacial shape and the velocity distribution at the interface between the molten material and water. And blowout velocity of the molten material at vapor explosion is evaluated from the visual data obtained on the experiment. The generated pressure at the vapor explosion is estimated by using the blowout velocity. In addition, the experimental results are compared with the thermal interaction zone (TIZ) theory.