Seiji Asada

Verification of Alternative Criteria for Shakedown Evaluation Using Flat Head Vessel

Alternative stress evaluation criteria suitable for Finite Element Analysis (FEA) proposed by Okamoto et al. [1] have been studied by the Committee on Three Dimensional Finite Element Stress Evaluation (C-TDF) in Japan. Thermal stress ratchet criteria in plastic FEA are now under consideration. Two criteria are proposed: evaluating variations in plastic strain increments and evaluating variations in the elastic core region. To verify the validity of these criteria, calculations were performed for several typical models in C-TDF [2]. This paper shows calculations and evaluation results of a Flat Head Vessel for shakedown. To study shakedown criteria for gross structural discontinuity, a flat head vessel is surveyed. The flat head vessel consists of a stiff flat head and a shell and is subjected internal pressure and thermal cycle. The elastic shakedown area and the plastic area are compared and plastic strain increments are surveyed. A shakedown evaluation method based on distribution of elastic-plastic strain range is proposed.© 2002 ASME

Round robin calculations of collapse loads : A torispherical pressure vessel head with a conical transition

Round robin calculations of collapse loads for a pressure vessel were made by sixteen teams in Japan. The model is composed of a cylinder and a trispherical head with a conical transition. The structure is an example in which the stress classifications specified in the ASME Code are not strictly applicable. The calculations were performed to clarify the issue of the evaluation procedure using the limit analysis method specified in the ASME Code, Sec. 3, and to check the sensitivity of calculation models and programs. It is found that the stress in the knuckle region has certain characteristics of secondary stress, yet still dominates the collapse of the vessel. Using the limit analysis to prove the validity of stress classifications is recommendable. The sensitivity of the calculation methods is not so significant. Therefore, it is concluded that the limit analysis can be used as a standard procedure in regulations.

PWSCC of Nickel Base Alloys in Vapor Phase Environment of Pressurizer

PWSCC incidents of Alloy 600 in vapor phase environment of pressurizer have been confirmed at several PWR plants. Vapor phase of pressurizer is filled with vapor from primary water, and the inner surface is covered with liquid film. Chemistry of the liquid film may be different from primary water, and this may cause different PWSCC susceptibility. Therefore the chemistry of liquid film of vapor phase has been investigated using simulated mock-up tests, and PWSCC susceptibility of 152 weld metal and TT600 (SG tube) has been investigated under the chemistry of the liquid film of vapor phase and primary water. According to the result of the chemistry investigation tests using mock-up of pressurizer, the liquid film environment was evaluated as follows: DH2 concentration: 300cc/kg·H2 O, B:150ppm, Li<0.1ppb, pH320°C :5.6 under the primary water chemistry condition is DH2 concentration:30cc/kg·H2 O, B:1950ppm, Li:3.7ppm, pH340°C :6.9. DH2 concentration of the liquid film is ten times higher and pH is lower than that of primary water. PWSCC susceptibility tests have been performed under the environment of the liquid film and primary water. No PWSCC crack propagation of 152 weld metal is confirmed in vapor phase environment. Crack growth rate of TT600 in vapor phase environment of pressurizer is not particularly high compared with that in primary water environment. It is confirmed that Alloy 690 (152 weld metal) has no PWSCC susceptibility under vapor phase environment of pressurizer. The difference of PWSCC susceptibility for Alloy 600 between vapor phase of pressurizer and primary water environment is not significant.Copyright

Optimization of Environmental Fatigue Evaluation: Step 1

In order to introduce environmental effects into fatigue evaluation of design and construction codes, the environmental fatigue evaluation method should not only be established, but the conservativeness of the codes, such as safety factors of design fatigue curve and simplified elastic-plastic analysis method (Ke factor), etc. should also be reviewed. Then plant designers should optimize total system of fatigue evaluation according to the objective of the codes by properly selecting design transient conditions and stress analysis methods used in fatigue evaluations as necessary. In addition, investigation of measures for reducing fatigue should be performed to mitigate possible fatigue initiators and alternative evaluation methods in case that the evaluation result should exceed the criteria specified in the design and construction codes. This paper discusses the present status in the review of these items for the Japanese PWR plants and future prospects to tackle on the application of environmental fatigue evaluation in design stage of plant construction.© 2007 ASME

Proposal of Fatigue Life Equations for Carbon and Low-Alloy Steels and Austenitic Stainless Steels as a Function of Tensile Strength

Fatigue life equations for carbon & low-alloy steels and also austenitic stainless steels are proposed as a function of their tensile strength based on large number of fatigue data tested in air at RT to high temperature. The proposed equations give a very good estimation of fatigue life for the steels of varying tensile strength. These results indicate that the current design fatigue curves may be overly conservative at the tensile strength level of 550 MPa for carbon & low-alloy steels. As for austenitic stainless steels, the proposed fatigue life equation is applicable at room temperature to 430 °C and gives more accurate prediction compared to the previously proposed equation which is not function of temperature and tensile strength.Copyright

Development of Load Multipliers (Z-Factors) in Elastic-Plastic Fracture Mechanics Evaluation in Rules on Fitness-for-Service for Nuclear Power Plants

Not only nonmandatory Appendix C, “Evaluation of Flaws in Piping,” in ASME Boiler & Pressure Vessel Code Section XI but also Appendix E-9, “Elastic-Plastic Fracture Mechanics Evaluation,” in the JSME Rules on Fitness-for-Service for Nuclear Power Plants use the load multiplier Z-factor that is applied to elastic-plastic fracture mechanics evaluation for a circumferential flaw of austenitic stainless steel piping and ferritic steel piping. The Z-factor is defined as the ratio of the limit load to the load at fracture load. Basically, the Z-factor equations were conservatively formulated by using the Z-factors for circumferential through-wall flaws. However, the Codes require flaw evaluation for circumferential surface flaws. Accordingly, Z-factors for circumferential surface flaws should be developed to have the consistency. Therefore Z-factor equations of austenitic stainless steel piping and ferritic steel piping have been developed for circumferential surface flaws.© 2012 ASME

Master Curve Approach for Some Japanese Reactor Pressure Vessel Steels

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.Copyright

Overview of the Japanese Code of Surveillance Test Program for Reactor Vessels

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.Copyright

Structural Evaluation for Repaired J-Weld Portion of Reactor Vessel Head Penetration

Leakage was found in a Reactor Vessel (RV) Head Penetration of Ohi unit 3 of the Kansai Electric Power Co., Inc. in May 4, 2004. Non-destructive examinations identified flaws in a J-weld portion of the Head Penetration. The J-weld portion was repaired by using Embedded Flaw Repair Technique [1] that performs welding of 52 weld metal on the J-weld surface remaining the flaws. In order to show the structural integrity of the J-weld portion, a fracture mechanics evaluation was performed in accordance with the Rules on Fitness-for-Service for Nuclear Power Plants of the JSME Codes for Nuclear Power Generation Facilities, JSME S NA1-2002 [2] (hereafter, the JSME Fitness-for-Service Rules) and literatures related. The flaw was characterized as both case of an embedded flaw and a surface flaw and KI for each flaw was directly calculated by using FE analysis. Fatigue crack growth analysis using KI calculated showed the amount of the crack growth was quite small. The fracture mechanics evaluation followed confirmed that the result satisfied the criteria. This paper explains the method and results for evaluation of the structural integrity of the J-weld portion.Copyright

KI Evaluation Method on Crack Growth of Bottom Mounted Instrumentation Penetrations in a Reactor Pressure Vessel

A method to evaluate a Mode I stress intensity factor KI for bottom mounted instrumentation (BMI) penetrations (nozzles) in a reactor pressure vessel (RPV) has been developed in order to perform the crack growth analysis simply and accurately. As regards the KI evaluation method, the several studies has been reported for a simple pipe with inside surface crack [1][2][3]. However, the study on the structure like the BMI nozzle joined with the vessel head by J-welding, has never been reported. Therefore, KI for BMI nozzles should be evaluated considering the load and the displacement by this interaction. The KI is evaluated simply by multiplying three correction factors to the one for a two-dimensional crack in a flat plate. These factors are to correct the flat plate’s KI to pipe’s one, two-dimensional crack shape to three-dimensional one, and to consider bottom head influence by the interaction. As regards a stress distribution in the wall thickness to be used, it is divided into the load control type and the displacement control one because the excessive conservativeness is excluded. Finally, the KI is calculated by superposing the one evaluated for each stress type. The evaluated KI was verified by comparing with the detailed three-dimensional FEA results. A simplified and accurate KI evaluation method for BMI nozzles in a RPV is proposed in this paper.Copyright