Flaw Evaluation Procedures Research Articles

Application of fracture evaluation method for aged elbow pipes

This paper evaluated a closed formed solutions of stress intensity factor K and J-integral for a flawed elbow of French Code, RSE-M, in order to introduce a flaw evaluation for aged elbows into the JSME Rules on Fitness-for-Service (JSME Rules), which have flaw evaluation procedures using K and J for only straight pipes and cylinder-shaped vessels in the current version. The stress distribution of an elbow is more complex than straight pipe and depends on not only flaw shape, flaw direction and radius/thickness but also bending radius, which may result in smaller allowable flaw size than that of a straight pipe. The authors performed K and J-integral calculation using FE analysis (FEA) focusing on an aged primary coolant pipe of PWR plant in Japan and compared with the solutions of the RSE-M code. As a result, it was confirmed that K and J solutions by the RSE-M code gave generally good result with conservativeness. Some cases of K showed large unconservative predictions by the RSE-M code and it was found that those were for the elbows with geometries out of the application range of the RSE-M code. The authors will develop a flaw evaluation for elbows with larger application range than that of the RSE-M code.

Suitability of fatigue crack growth thresholds at negative stress ratios for ferritic steels and aluminum alloys in flaw evaluation procedures

Fatigue crack growth thresholds for ferritic steels and aluminum alloys in flaw evaluation documents are reviewed. The WRC (Welding Research Council) Bulletin, ASME (American Society of Mechanical Engineers) Code Section VIII, IIW (International Institute of Welding) and BS (British Standards) 7910 give constant fatigue crack growth threshold values at negative stress ratios. However, the definitions of the thresholds at negative stress ratios are different between these flaw evaluation documents. From fatigue crack growth tests and collection of fatigue crack growth threshold data, the thresholds are affected by compressive stresses. It can be said that the thresholds at negative stress ratios are not constant. Therefore, the thresholds given by the WRC Bulletin, ASME Section VIII and IIW are slightly un-conservative. The threshold given by BS 7910 is considerably conservative. A suitable definition of the threshold at negative stress ratios for application in flaw evaluation procedures is proposed as the variable threshold expressed by full range of stress intensity factors.

Flaw Evaluation Procedure for Cast Austenitic Stainless-Steel Materials Using a Newly Developed Statistical Thermal Aging Model

Abstract Thermal embrittlement of some cast austenitic stainless steels (CASS) occurs at reactor operating temperatures leading to very low fracture toughness. Because of their low aged toughness with high variability, flaw evaluations of CASS material need to be established with an understanding of the materials aged condition, especially since most U.S. pressurized water reactor (PWR) nuclear plants have been given plant life extensions for 60-year operation. A flaw evaluation procedure for CASS materials is presented here using a new statistical model developed to predict the toughness of fully aged CASS using the materials' chemical compositions. In this procedure, the dimensionless-plastic-zone-parameter (DPZP) analysis is used to determine when limit-load is applicable and also approximate the elastic-plastic correction factor (Z-factor) to predict the failure stress for CASS pipe/fittings with a circumferential surface crack. The procedure was validated against several CF8m pipe test results which include various pipe diameters, crack sizes, ferrite contents, failure modes. The as-developed flaw evaluation procedure was also used to determine the Z-factors for four different pipe diameters for a database of 274 pipe/elbows in U.S. PWR plants –solving 1096 sample problems to understand what range of Z-factors in U.S. PWR plants (for CF8m CASS materials). Finally, the applicability of the CF8m-based statistical model for use with CF3 and CF8 CASS materials was also verified with available test results. The outcome of this work has been implemented in an ASME boiler and pressure vessel (BPV) code case and has been approved in 2020 as Code Case N-906.

Effects of Crack Closure on the Fatigue Crack Growth Rates of Ferritic Steels Subjected to Severe Reversing Loads

Abstract Low-alloy steels are extensively used in pressure boundary components of nuclear power plants. The structural integrity of the components made of low-alloy steels can be evaluated by the flaw evaluation procedure provided by Section XI of the ASME Boiler and Pressure Vessel Code. According to the Code, the stress intensity factor range ΔK can be used to determine the fatigue crack growth rates of the material. However, it has been reported that the fatigue crack growth rate under severe reversing loads is also strongly influenced by crack closure behavior. This paper discusses the relation between applied stresses and the fatigue crack growth rate for cracks in low-alloy steels exposed to air. Compressive-tensile cyclic loadings are applied to center-notched plates to obtain the fatigue crack growth curves. The test data demonstrate that effective stress intensity factor range predicted by our closure model described the crack growth property more accurately. A comparison among crack closure models indicates that our crack closure model is suitable to predict the crack growth rates when low constraint conditions are assumed at the crack tip due to severe reversing loads.

Technical Basis for Flaw Acceptance Criteria for Cast Austenitic Stainless Steel Piping

Abstract According to the current ASME Code Section XI, IWB-3640 and Appendix C flaw evaluation procedure, cast austenitic stainless steel (CASS) piping with ferrite content (FC) less than 20% is treated as wrought stainless steel. For CASS piping with FC equal or greater than 20%, there was no flaw evaluation procedure in the ASME Code prior to the 2019 Edition. In this paper, the technical basis for the recently approved Code change containing flaw acceptance criteria for CASS piping is presented. The procedure utilizes the current rules in ASME Code Section XI, IWB 3640/Appendix C and the existing elastic-plastic correction factors (i.e., Z-factors) for other materials in the Code. The appropriate Z-factor to use for the CASS piping is determined based on the FC (using Hull's equivalent factor). Experimentally measured fully saturated fracture toughness and tensile data of the three most common grades of CASS material in the U.S. (CF3, CF8, and CF8M) were used to determine the flaw acceptance criteria in the Appendix C Code method. As described here, the method is conservative since it utilizes the fully saturated condition of CASS materials. In addition, it is simple and consistent with the current regulatory guidance on aging management of CASS piping.

Collapse loads for circumferentially through-wall cracked pipes subjected to combined torsion and bending moments

Pressurized piping items in power plants may experience combined torsion and bending moments during operation. Currently, there is a lack of guidance in flaw evaluation procedures for combined loading modes of pressure, torsion and bending loads. Recently, collapse bending moments for pipes under torsion moments were analyzed by finite element modelling. Equivalent moments defined as the root of the sum of the squares of the torsion and bending moments are shown to be equal to pure bending moments for various diameter pipes containing circumferentially part through cracks. This paper focuses on behaviour of plastic collapse moments for pipes with circumferential through-wall cracks using finite element analysis, and describes the behaviour of the equivalent bending moments for flaw evaluation procedures, referring the results of part through cracked pipes.

Effects of Torsion on Equivalent Bending Moment for Limit Load and EPFM Circumferential Pipe Flaw Evaluations

The circumferential flaw evaluation procedures in ASME Boiler and Pressure Vessel Code Section XI nonmandatory Appendix C are currently limited to straight pipes under pressure and bending loads without consideration of torsion loading. The Working Group on Pipe Flaw Evaluation of the ASME Boiler and Pressure Vessel Code is developing guidance for considering the effects of torsion by a mean of an equivalent bending moment, which is a square root of sum square combination of bending moment and torsion load with a weighted factor for torsion moment. A torsion weighted factor, Ce, is established in this paper using large strain finite element limit load analysis with elastic perfectly plastic materials. Planar flaws and nonplanar flaws in a 10.75 in. (273 mm) OD pipe are investigated. Additionally, a finite element J-integral calculation is performed for a planar through wall circumferential flaw with elastic plastic materials subjected to bending and torsion load combinations. The proposed Ce factor for planar flaws is intended for use with the ASME B&PV Code Section XI, Appendix C for limit load and Elastic Plastic Fracture Mechanics (EPFM) circumferential planar flaw evaluations.

Universal Weight Function Consistent Method to Fit Polynomial Stress Distribution for Calculation of Stress Intensity Factor

Procedures for the analytical evaluation of flaws in nuclear pressure boundary components are provided in Section XI of the ASME B&PV Code. The flaw evaluation procedure requires calculation of the stress intensity factor. Engineering procedures to calculate the stress intensity factor are typically based on a polynomial equation to represent the stress distribution through the wall thickness, where the polynomial equation is fitted using the least squares method to discrete data point of stress through the wall thickness. However, the resultant polynomial equation is not always an optimum fit to stress distributions with large gradients or discontinuities. Application of the weight function method enables a more accurate representation of the stress distribution for the calculation of the stress intensity factor. Since engineering procedures and engineering software for flaw evaluation are typically based on the polynomial equation to represent the stress distribution, it would be desirable to incorporate the advantages of the weight function method while still retaining the framework of the polynomial equation to represent the stress distribution when calculating the stress intensity factor. A method to calculate the stress intensity factor using a polynomial equation to represent the stress distribution through the wall thickness, but which provides the same value of the stress intensity factor as is obtained using the Universal Weight Function Method, is provided in this paper.

Inclusion of Torsion Loads in Section XI Flaw Evaluation Procedures for Pipes Containing Circumferential Planar Surface Crack-Like Flaws on the Basis of Limit Load Analysis

Piping systems in power plant may experience combined bending-torsion loads in the presence of planar crack-like flaws. ASME Boiler and Pressure Vessel Code Section XI nonmandatory Appendix C provides flaw evaluation procedures for pipes with flaws. These are currently limited to straight pipes under pressure and bending loads and no provision is made for torsion loading. The working group on pipe flaw evaluation is developing guidance for including the torsion load within the existing solutions provided in the Appendix C for bending loading on a straight pipe under fully plastic fracture regime. This paper reports on the finite element limit load analyses performed on the straight pipe containing a circumferential planar crack-like flaw. Pipe diameters were ranging from 4 in. (100 mm) to 24 in. (600 mm) nominal diameter (OD) and R/t ranging from 6 to 40. For the purpose of nonmandatory Appendix C flaw evaluation, it is concluded that the torsion loads can be combined with bending loads using the root of the sum of the squares (RSS) method of Section III of the ASME Boiler Code, without any additional weighting on torsion.

JSME FFS Rule Making of Flaw Evaluation for Nickel Based Alloy Components

Nickel based alloy is used for light water reactor power plant components such as penetration nozzles and butt weld nozzles of reactor vessels. In recent years, stress corrosion cracking (SCC) occurred at the nickel based alloy weld joints of the power plants in Japan and foreign countries. Integrity of cracked components is evaluated according to the flaw evaluation procedure prescribed in the maintenance rule issued by The Japan Society of Mechanical Engineers (JSME). However, the latest version (2008 edition) of Rules on Fitness-for-Service for Nuclear Power Plants of JSME maintenance rule does not provide the flaw evaluation procedure for nickel based alloy components. The working group of flaw evaluation of JSME code and standard committee have been involved in establishing the flaw evaluation procedure from 2008 and three code cases were published by JSME. This paper introduces the brief contents of the code cases for flaw evaluation for nickel based alloy components.

Determination of the elastic-plastic fracture mechanics Z-factor for alloy 182 weld metal flaws

One of the ways that the ASME Section XI code incorporates elastic-plastic fracture mechanics (EPFM) in the Section XI Appendix C flaw evaluation procedures for circumferential cracks is through a parameter called Z-factor. This parameter allows the simpler limit-load (or Net-Section-Collapse) solutions to be used with a multiplier from EPFM analyses. This paper shows how 3-D finite element (FE) analyses were employed to investigate the sensitivity of the crack-driving force as a function of crack location (i.e., crack in the center of weld, or closer to the stainless or low alloy steel sides) in an Alloy 182 dissimilar metal weld (DMW), and how an appropriate (or equivalent) stress-strain curve was determined for use in the J-estimation schemes. The J-estimation schemes are then used to cover a wider range of variables, i.e., pipe diameters, cracks lengths, and also incorporate crack growth by ductile tearing. The Z-factor equations as a function of pipe diameter were calculated using the LBB.ENG2 J-estimation scheme along with the most conservative equivalent stress-strain curve from the FE analyses. The proposed Z-factor approach was then validated against an Alloy 182 DMW full-scale pipe test that had a circumferential through-wall crack in the fusion line. The predicted EPFM maximum load showed excellent agreement with the experimental result. Furthermore, it was shown that the proposed Z-factor equation is not sensitive to the location of the crack.

Defect assessment procedures for plants

There are a significant number of procedures available for use in plants mostly used in a company or of a national character. A number of fracture mechanics based low temperature assessment procedures have been developed during the last decade within Europe and world wide. Guidelines and methodologies for high temperature assessment also exist within several European countries. Furthermore, new procedures are being worked on in various EC projects. The present paper gives a review and comparison of the available defect assessment procedures, presented based on mainly the available published documents by the project partners of EC TNW project PLAN, Cluster Task C2. The survey on usage of DAPs preceeds the reviews of low and high temperature procedures, followed by the unified procedures of SINTAP and HIDA, respectively. The French codified flaw evaluation procedure that was not included in the above mentioned reviews is also presented.

Japanese fitness-for-service code for nuclear power plants—summary of flaw evaluation procedures

A Japanese fitness-for-service code for nuclear power plant components has been developed at the Japan Society of Mechanical Engineers. The code prescribes methods for the evaluation of flaws, which are detected during inservice inspection for pressure vessels and pipes in nuclear power plants. This paper describes the summary of flaw evaluation procedures, including comparisons with the ASME Code Section XI.

RSE-M. A general presentation of the French codified flaw evaluation procedure

This paper describes the flaw evaluation procedures in the French RSE-M code, which describes the rules for in-service inspection of nuclear power plant components. Some specific novel features of the flaw evaluation procedures are described including the method for estimating J for combined thermal and mechanical loads, treatment of small-scale yielding and stress intensity factor solutions. Finally, the paper summarises ongoing work areas.

Development of a flaw evaluation handbook of the high pressure institute of Japan

This paper introduces a handbook, which is edited by the High Pressure Institute of Japan to support engineers who evaluate flaws detected in nuclear power components according to the Japanese fitness-for-service code. The handbook is written in Japanese and contains basic information on fracture mechanics as well as the specific flaw evaluation procedures and material properties data stipulated in the code. The main features of the handbook are summarized in the paper.

Overview of a draft of new maintenance standards

Abstract One of the topics on structural standards in Japan is the completion of a draft of new maintenance standards which covers inservice inspection (ISI), flaw evaluation and repair procedures. This paper describes the background of Japan's current structural codes and standards, the necessity of introducing flaw evaluation, and the outline of the activities of developing the Draft of Maintenance Standards in the Japan Power Engineering Inspection Corporation. The essential features of the Draft of Maintenance Standards for ISI, flaw evaluation, and repair technologies, respectively, and future activities for updating the Draft of Maintenance Standards are presented.

Flaw evaluation of theramally aged cast stainless steel in light-water reactor applications

Cast stainless steel may be used in the fabrication of the primary loop piping, fittings, valve bodies, and pump casings in light-water reactors. However, this material is subject to embrittlement due to thermal aging at the reactor temperature, that is, 290°C (550°F). The Argonne National Laboratory (ANL) recently completed a research program and the results indicate that the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of submerged arc welds (SAWs). Thus, the US Nuclear Regulatory Commission (NRC) staff has accepted the use of SAW flaw evaluation procedures in IWB-3640 of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to evaluate flaws in thermally aged cast stainless steel for a license renewal evaluation. Alternatively, utilities may estimate component-specific fracture toughness of thermally aged cast stainless steel using procedures developed at ANL for a case-by-case flaw evaluation.

Low-cycle fatigue crack growth considerations in pipe fracture analyses

This paper presents some results from an ORNL cyclic pipe experiment that demonstrated the magnitude of low-cycle fatigue crack growth from simulated seismic loading. The concern with this result is that current pipe flaw evaluation procedures do not consider the cyclic crack growth during a seismic event, i.e. the peak dynamic load is used in a static analysis. This was the main incentive for the analyses developed in this paper. Additional applications of the methodology developed in this paper would be high stress-low cycle fatigue transients that can occur in service. In this paper, we show that the low-cycle fatigue crack growth data from the ORNL pipe test agreed well with the extrapolation of high-cycle fatigue crack growth data from small specimens. The next step was to evaluate the accuracy of several different methods to predict the low-cycle fatigue crack growth. In this evaluation it was found that for low-cycle fatigue LEFM analysis would significantly underpredict the crack growth and EPFM analyses were needed. Three different EPFM analyses were used. The GE/EPRI method was found to be overly conservative, which is consistent with comparisons with past pipe fracture data. The LBB.NRC and LBB.ENG2 J-estimation schemes provided good predictions of the experimental crack growth data when the crack-closure load was taken into account.

Threshold stress criterion for delayed hydride crack initiation at a blunt notch in zirconium alloys

The flaw evaluation procedures, currently used for CANDU (Canada Deuterium Uranium) nuclear reactor zirconium alloy pressure tubes, are based on the prediction of the initiation of delayed hydride cracking (DHC) at a blunt notch by assuming that failure of hydrided material occurs when the peak tensile stress in the vicinity of a notch attains a critical value pc. However, since failure involves plastic straining of the hydrided material, the value of pc cannot be unique. The paper quantifies this non-uniqueness in terms of the ratio pcp∗ where p∗ is the threshold stress appropriate to the base-line case where a hydrided region forms at the planar surface of a semi-infinite solid, that is subjected to a uniform tensile stress, and where no allowance is made for the compressive stress induced by hydriding. The extent to which pc exceeds p∗ depends upon the ductility of the hydrided material and the stress gradient in the vicinity of a notch. In the latter context, it is shown how the ratio pcp∗ increases as the hydrided region length decreases (smaller number of thermal cycles), and as the notch root radius decreases.

Evaluation of flaws in nuclear piping

Abstract This paper summarizes development of flaw evaluation procedures and allowable flaw sizes for piping used in light water reactors, and focuses on the research conducted by the Electric Power Research Institute in support of Section XI of the ASME Code. The paper also presents development of elastic-plastic fracture analysis methods needed in the development of the code evaluation procedures.