Takatoshi Hirota

Applicability of Miniature C(T) Specimen to Evaluation of Fracture Toughness for Reactor Pressure Vessel Steel

Irradiation embrittlement of Japanese reactor pressure vessels (RPV) is usually monitored by conducting tests on irradiated RPV material according to surveillance test program. Although fracture toughness specimens are contained in Japanese PWR surveillance capsule, the number of specimens is limited due to capacity of capsule. In order to evaluate lower bound of fracture toughness considering its scatter with higher reliability, it is expected to obtain additional fracture toughness data using remaining broken specimens of irradiated materials.One of solutions to this problem is specimen reconstitution technique. However, it is difficult to make numbers of specimens by reconstitution because of need for specific equipments and time-consuming machining operations. As an alternative method, fracture toughness test using miniature C(T) specimens with dimension of 4×10×10mm, which can be taken from broken halves of Charpy specimen, is proposed and the studies to verify the reliability and robustness of evaluation method have been conducted in the Japanese round robin program since 2010.In this study, fracture toughness tests were performed on Japanese SA 533 Gr.B Cl.1 steel using miniature C(T) specimens and the effect of specimen size on reference temperature T0 was studied by the Master Curve approach. In addition, the issues related to application to irradiated materials were discussed.Copyright

Evaluation on Constraint Effect of Reactor Pressure Vessel Under Pressurized Thermal Shock

In recent years, the integrity of reactor pressure vessel (RPV) under pressurized thermal shock (PTS) accident has become controversial issue since the larger shift of RTNDT in some higher fluence surveillance data raised a concern on RPV integrity. Under PTS condition, the combination of thermal stress due to a temperature gradient and mechanical stress due to internal pressure causes considerable tensile stress inside the wall of RPV. Currently, RPV integrity is assessed by comparing stress intensity factor on a crack tip under PTS condition and a reference toughness curve based on the fracture toughness data of irradiated compact specimens. Since PTS loading is large enough to cause plastic deformation, a crack tip behavior on the inner surface of RPV can be explained by elastic-plastic fracture mechanics using the J-integral.In this study, 3D elastic plastic finite element analyses were performed to assess the crack tip behavior on surface of a RPV under Loss of coolant Accident, which causes one of the most severe PTS condition. In order to quantify the constraint effect on a surface crack, J-Q approach was applied. The constraint effect of a surface crack was compared with a compact specimen and its influence on the fracture toughness was assessed.As a result, the difference of constraint effect was clearly obtained. And it is recommended to consider constraint effects in the evaluation of structural integrity of RPV under PTS.Copyright

Simplified J-Integral Evaluation Method for Evaluation of Reactor Pressure Vessel in the Upper-Shelf Region

It is well-known that the Upper-Shelf Energy of reactor pressure vessel (RPV) steels is reduced due to neutron irradiation and J-integral (J-applied) is used for the structural integrity evaluation in upper-shelf region. It is very time consuming to calculate the J-integral in detail by using 3-dimensional elastic-plastic FE analysis. U.S. Regulatory Guide 1.161[1] applies a simplified J-applied calculation method based on elastic calculation of the stress intensity factor, KI . However, this method for thermal stress can be applied only for constant cooling rates. From this point of view, a simplified J-applied evaluation method which can be applied to design transients and has high accuracy has been studied. In order to develop the new evaluation method, KI based on 1-dimensional elastic analysis, KI based on 3-dimensional elastic FE analysis and J-applied based on 3-dimensional elastic-plastic FE analysis are calculated and compared with each other. An accurate J estimation method for design transients from 1-dimensional elastic analysis results is proposed and the severest transient in each condition of each RPV group is evaluated in this paper.Copyright

Applicability of Fracture Toughness Curves Developed for Japanese Pressure Vessel Steels to Structural Integrity Evaluation

Surveillance tests have been conducted on Japanese Pressurized Water Reactor (PWR) plants for more than 40 years to monitor irradiation embrittlement of reactor pressure vessel (RPV) beltline materials. Fracture toughness specimens are contained as well as tensile and Charpy impact specimens in a surveillance capsule and utilized for structural integrity evaluation. Therefore, a lot of fracture toughness data have been obtained by fracture toughness tests using such as Compact Tension (CT) and Wedge Opening Loading (WOL) specimens. More than one thousand data have been accumulated for both unirradiated and irradiated materials until 2013. Additionally, in terms of fracture toughness, Master Curve (MC) concept has been widely used for fracture toughness transition curve expression of ferritic steels. Considering such a situation, the new fracture toughness curves using Tr30, which denotes Charpy V-notch 30ft-lb transition temperature, as an indexing parameter were developed based on MC concept depending on product form for Japanese RPV steels in 2014.In this study, applicability of the newly developed curves of Japanese RPV steels to structural integrity evaluation is investigated. Especially, this paper focused on conservatism of the curves and the adequate margin to be added in evaluation of RPV integrity employing statistical methodology.Copyright

Proposal for Update on Evaluation Procedure for Reactor Pressure Vessels Against Pressurized Thermal Shock Events in Japan

The evaluation procedure for the reactor pressure vessel integrity of Japanese PWR plants against Pressurized Thermal Shock (PTS) events is prescribed in the Japan Electric Association Code, JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” since 1991. The current procedure was developed based on the PTS verification test program, which was conducted as Japanese national project and the related studies in 1980’s.Since much progress has been made on fracture mechanics, fracture toughness, in-service inspection techniques/results and so on, it is preferred to advance the current procedure for more credible evaluation by reflecting the latest knowledge.This paper describes the outline of the studies to update the current procedure.© 2014 ASME

Alternative Reference Temperature Based on Master Curve Approach in Japanese Reactor Pressure Vessel Steel

Master Curve approach is the effective method to evaluate the fracture toughness of the ferritic steels accurately and statistically. The Japan Electric Association Code JEAC 4216-2011, “Test Method for Determination of Reference Temperature, To, of Ferritic Steels” was published based on the related standard ASTM E 1921-08 and the results of the investigation of the applicability of the Master Curve approach to Japanese reactor pressure vessel (RPV) steels. The reference temperature, To can be determined in accordance with this code in Japan.In this study, using the existing fracture toughness data of Japanese RPV steels including base metals and weld metals, the method for determination of the alternative reference temperature RTTo based on Master Curve reference temperature To was statistically examined, so that RTTo has an equivalent safety margin to the conventional RTNDT. Through the statistical treatment, the alternative reference temperature RTTo was proposed as the following equation; RTTo = To + CMC + 2σTo. This method is applicable to the Japan Electric Association Code JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” as an option item.Copyright

Development of Prediction Equations on Charpy Upper Shelf Energy for Japanese RPV Steels

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 1024 n/m2 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.Copyright