Japanese Reactor Pressure Vessel Steels Research Articles

G0310404 国内原子炉圧力容器鋼lこ対する破壊靭性カーブの開発(材料力学部門一般セッション:圧力容器・ねじ)

G0310404 国内原子炉圧力容器鋼lこ対する破壊靭性カーブの開発(材料力学部門一般セッション:圧力容器・ねじ)

Evaluation of Fracture Toughness by Master Curve Approach Using Miniature C(T) Specimens

The fracture toughness master curve shows the relationship between the median of fracture toughness and temperature in the ductile–brittle transition temperature region of ferritic steels such as reactor pressure vessel (RPV) steels. The master curve approach specified in the ASTM standard theoretically provides the confidence levels of fracture toughness in consideration with the inherent scatter of fracture toughness. The authors have conducted several fracture toughness tests for typical Japanese RPV steels with various specimen sizes and shapes and ascertained that the master curve can be accurately applied to the specimens with a thickness of 0.4-in. or larger. With respect to using the master curve method with the current surveillance program for operating RPVs, the utilization of miniature specimens is important. Miniature specimens, which can be taken from the broken halves of surveillance specimens, are necessary for the efficient determination of the master curve from the limited volume of the available materials. In this study, fracture toughness tests were conducted for typical Japanese RPV steels, particularly SFVQ1A forged and SQV2A plate materials, using the miniature C(T) specimens with a thickness of 4 mm, following the procedure in the ASTM standard. The results show that the differences in the test temperature, evaluation method, and specimen size did not affect the master curves, and the fracture toughness indexed by the reference temperature, To, obtained from miniature C(T) specimens were consistent with those obtained from the standard and larger C(T) specimens. It was also found that valid reference temperatures can be determined with a realistic number of miniature C(T) specimens, i.e., less than ten, if the test temperature was appropriately selected. Thus, the master curve method using miniature C(T) specimens could be a practical method to determine the fracture toughness of actual RPV steels.

Evaluation of Fracture Toughness by Master Curve Approach Using Miniature Specimens

The fracture toughness Master Curve gives a universal relationship between the median of fracture toughness and temperature in the ductile-brittle transition temperature region of ferritic steels such as reactor pressure vessel (RPV) steels. Considering the possible application of the Master Curve method coexistent with the present surveillance program for operating RPVs, the utilization of miniature specimens is quite important for efficient determination of the Master Curve from limited volume of the materials of concern. In this study, fracture toughness tests were conducted for typical Japanese RPV steels using the miniature C(T) specimens with the thickness of 4 mm following the procedure of the ASTM standard. The results showed that the differences in test temperature, evaluation method, and specimen size did not affect the Master Curves, and the fracture toughness indexed by the reference temperature, To, obtained from miniature C(T) specimens were consistent with those obtained from standard and larger C(T) specimens.

Characteristics of the New Embrittlement Correlation Method for the Japanese Reactor Pressure Vessel Steels

Neutron irradiation embrittlement of reactor pressure vessel steels is an important aging issue for the long-term operation of light water reactors. A new embrittlement correlation method was developed by Central Research Institute of Electric Power Industry and the Japanese electric utilities in 2007. This method is primarily based on the fundamental understandings on the embrittlement mechanisms, i.e., microstructural changes were modeled by the mathematical form of rate equations, and the predicted microstructural changes were further correlated with the mechanical property changes in transition temperature region. The coefficients of the rate equations were optimized using the Japanese surveillance data of RPV embrittlement. This method was adopted as the revision of the Japanese code, JEAC4201-2007, in 2007. In this paper, after a brief explanation on the new correlation method, the predictions of the new method will be investigated through comparisons with the previous correlation, JEAC4201-2004, and the U.S. surveillance data in order to identify the characteristics of the new method.

Embrittlement Correlation Method for the Japanese Reactor Pressure Vessel Materials

Abstract A new embrittlement correlation method developed for the Japanese reactor pressure vessel (RPV) steels is presented. The Central Research Institute of Electric Power Industry and the Japanese electric utilities conducted a project to develop a new embrittlement correlation method for the Japanese RPV steels based on the understandings on the mechanisms of the RPV embrittlement. In addition to the information from the literatures, we generated new information by characterizing the microstructural changes in the surveillance materials of the Japanese commercial reactors. We found that in low Cu materials, solute atom clusters containing little or no Cu atoms are formed at relatively low fluence of 3×1019 n/cm2, E>1 MeV. The volume fraction of the solute atom clusters has a good correlation with the Charpy transition temperature shift regardless of the Cu content. We also found that the microstructure of the boiling water reactor surveillance material is different from that of the archive material irradiated in material testing reactor. The understandings on the RPV embrittlement mechanisms were formulated using a set of rate equations, and the coefficients of the equations were optimized using the ΔRTNDT values of the Japanese surveillance database. This method considers the effect of neutron flux. Only one set of coefficients was developed, and they are independent of the product form. Predictions of the new embrittlement correlation method were compared with those of the recent U.S. correlation method as well as the U.S. surveillance data. The comparison shows the characteristics of the present method.

1008 超小型試験片を用いたマスターカーブ破壊靭性評価(OS10-2 供用エネルギー機器の経年変化と健全性評価-破壊評価-)

The fracture toughness Master Curve gives a universal relationship between the median of fracture toughness and temperature in the ductile-brittle transition temperature region of ferritic steels such as reactor pressure vessel (RPV) steels. Considering the possible application of the Master Curve method coexistent with the present surveillance program for operating RPVs, the utilization of miniature specimens is quite important for efficient determination of the Master Curve from limited volume of the materials of concern. In this study, fracture toughness tests were conducted for typical Japanese RPV steels using the miniature C(I') specimens with the thickness of 4 mm following the procedure of the ASTM standard. The results showed that the differences in test temperature, evaluation method, and specimen size did not affect the Master Curves, and the fracture toughness indexed by the reference temperature, T_o, obtained from miniature C(T) specimens were consistent with those obtained from standard and larger C(f) specimens.

マスターカーブ法による国産原子炉圧力容器鋼の破壊靭性評価(保全・設備診断技術,<特集>第13回動力・エネルギー技術シンポジウム)

The fracture toughness Master Curve is a universal relationship between the median of fracture toughness and temperature in the ductile-brittle transition temperature region of ferritic steels. The Master Curve approach specified in the ASTM standard theoretically provides the confidence levels of fracture toughness in consideration of the inherent scatter of fracture toughness. In this study, fracture toughness tests for Japanese reactor pressure vessel steels, with various specimen sizes and shapes were conducted. The differences in test temperature, evaluation method, and specimen size and shape did not affect the Master Curves, ascertaining that the Master Curve approach could be well applied to these classs of steels. The adequacy of the lower bounds based on the ASTM standard was also verified by comparing with experimental fracture toughness data.

The effect of chemical composition on irradiation embrittlement

The Japanese Pressurized Thermal Shock (PTS) integrity study is now being conducted by the Japan Power Engineering and Inspection Corporation (JAPEIC) under contract with the Ministry of International Trade and Industry (MITI). This paper introduces the study plan and the interim results on irradiation embrittlement of reactor pressure vessel steels. Five kinds of base metal plates and two kinds of weld metals were irradiated in the Japan Material Testing Reactor (JMTR). After irradiation, Charpy test and fracture toughness ( K IC , J IC , K Id , J Id ) test were conducted. Principal results are as follows. 1. (1) Change of material properties of some Japanese reactor pressure vessel steels caused by neutron irradiation was obtained. 2. (2) Predictive equations of irradiation embrittlement ΔRT NDT for Japanese PWR reactor pressure vessel steels were developed.