The effect of the neutron flux at high fluence on the microstructural and hardness changes of reactor pressure vessel (RPV) steels was investigated in succession to the previous study [1]. An accelerated test reactor irradiation of copper containing RPV materials, previously irradiated in commercial reactors, was carried out at the lowest possible neutron fluxes in order to obtain neutron fluences up to approximately 1×10 20 n/cm 2 (E>1MeV). State-of-the-art experimental techniques such as three-dimensional atom probe were applied to carry out advanced quantitative characterization of defect features in the materials. Results for the same materials irradiated in both high and low flux conditions are compared. For neutron fluences above 6×10 19 n/cm 2 (E>1MeV) the difference in the neutron fluence dependence of the increase in hardness is not seen for any neutron flux condition. The number densities and the diameters of solute atom clusters for the low flux irradiation materials tend to be lower and larger, respectively, than that for the high flux irradiation materials, while the volume fraction of solute atom clusters increases linearly with increasing neutron fluence, and the effect of neutron flux is not significant. The component elements and the chemical composition of the solute atom clusters formed by irradiation for the same material do not change regardless of the neutron fluence and flux. The square root of the volume fraction of the solute atom clusters provides a good correlation with the increase in hardness.

The effect of the neutron flux at high fluence on the microstructural and hardness changes of a reactor pressure vessel (RPV) steel was investigated. An accelerated test reactor irradiation of a RPV material, previously irradiated in commercial reactors, was carried out at the lowest possible neutron fluxes in order to obtain neutron fluences up to approximately 1×10 20 n/cm 2 (E>1MeV). State-of-the-art experimental techniques such as three-dimensional atom probe were applied to carry out advanced quantitative characterization of defect features in the materials. Results for the same material irradiated in both high and low flux conditions are compared. For neutron fluences above 6×10 19 n/cm 2 (E>1MeV) the difference in the neutron fluence dependence of the increase in hardness is not seen for any neutron flux condition. The volume fraction of solute atom clusters increases linearly with neutron fluence, and the influence of neutron flux is not significant. The component elements and the chemical composition of the solute atom clusters formed by the irradiation do not change regardless of the neutron fluence and flux. The square root of the volume fraction of the solute atom clusters is a good correlation with the increase in hardness.

Neutron radiation embrittlement of reactor pressure vessel (RPV) steels is one of critical issues for the structural integrity assessment of the RPVs. Especially, the embrittlement at high fluences is of great interest for the long term operation of light water reactors because information on the mechanical property changes as well as embrittlement mechanisms is limited at high fluences. In this study, microstructural analyses were conducted on the RPV steels irradiated to high fluences in order to confirm the applicability of the trend curve at high fluence region. Steels investigated are five base metals and a weld metal with their copper content ranging from 0.02 to 0.25 wt. %. These steels were irradiated in the material test reactors to fluence up to 1.3 × 10 20 n/cm 2 , E > 1MeV, at temperature of about 290 °C. After irradiation, transmission electron microscope (TEM) observations were performed to characterize the nano-meter scale microstructural changes due to irradiation. Formation of dislocation loops was observed. Number density and diameter of dislocation loop was investigated. Effects of chemical composition of steel and fluence on dislocation loop formation are discussed.

The Japan Electric Association Code, JEAC 4201, “Method of Surveillance Tests for Structural Materials of Nuclear Reactors” was revised in 2007. Major revised points of JEAC4201-2007 are: 1) A new irradiation embrittlement correlation method, 2) Reconstitution of surveillance specimen, 3) Long-term surveillance program. An overview of the revised JEAC 4201-2007 is presented in this paper.

The Master Curve Approach for assessing fracture toughness of reactor pressure vessel (RPV) steels has been accepted throughout the world. The Master Curve Approach using fracture toughness data obtained from RPV steels in Japan has been investigated in order to incorporate this approach into the Japanese Electric Association (JEA) Code 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components”. This paper presents the applicability of the Master Curve Approach for Japanese RPV steels.

A new irradiation embrittlement correlation method for reactor pressure vessel (RPV) materials was developed by the Central Research Institute of Electric Power Industry (CRIEPI). In order to incorporate this new correlation method into the Japan Electric Association Code, JEAC 4201, “Method of Surveillance Tests for Structural Materials of Nuclear Reactors”, sensitivity studies for PWR and BWR reactor vessel materials were performed and the features of the new correlation method was investigated. Based on those studies, the prescription of the new correlation method in JEAC 4201 was formulated. The results of the studies and the overview of the revised JEAC 4201 are presented in this paper.

It is well known that as the embrittlement due to neutron irradiation on reactor pressure vessel (RPV) steels, there is the tendency of the reduction of Charpy absorbed energy at upper shelf region (USE), in addition to the shift of ductile-brittle transition temperature. Concerning to the regulation of the upper shelf region, no method is provided to evaluate integrity for RPV steels with USE of less than 68J in Japanese codes. Under the circumstance, the reduction tendency of USE using simulated Japanese RPV steels, irradiated by fast neutron up to 10 24 n/m 2 in OECD Halden test reactor, was investigated to establish the basis of the USE prediction after 60 years plant operation for the integrity assessment of the RPVs. The USE prediction equations have been developed through the regression analyses of the test reactor data combined with Japanese surveillance test data. This research was entrusted by the Ministry of Economy, Trade and Industry in Japan.

The Japan Electric Association Code, JEAC 4201, “Method of Surveillance Tests for Structural Materials of Nuclear Reactors” was originally published in 1970 so as to design a surveillance program for monitoring radiation induced changes in mechanical properties of beltline materials in light-water moderated nuclear power reactor vessels and to evaluate the test results. Recently in 2004 the code was revised, and a new method for predicting the decrease in upper shelf Charpy impact energy (USE) of beltline materials was incorporated based on the results of the research performed as a national project by the Japan Power Engineering and Inspection Corporation. Japanese surveillance tests program, including materials selections, type of specimens, number of capsules, withdrawal schedule, and evaluation of the results, is overviewed. Furthermore, methods for predicting the decrease in USE and reference temperature shift for beltline materials are also presented.

The Japan Electric Association Code, JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” was originally published in 1973 in order to prescribe test methods, fracture toughness requirements and acceptance standards for materials used in nuclear power plant components. The code was recently revised so as to incorporate a new method to evaluate the structural integrity of reactor vessels with upper shelf Charpy impact energy (USE) less than 68J, based on the results of the researches performed as national projects by the Japan Power Engineering and Inspection Corporation and so on. In this paper, some contents of the code, which are applicable for reactor vessels, such as J integral based integrity evaluation method for reactor vessels with low USE including predicting J resistance curves (J-R curves) by using USE and temperature, and methods to evaluate integrity against pressurized thermal shock events and to determine pressure temperature limits, are overviewed.