Analysis of HDPE Failure Data in Support of Development of Flaw Acceptance Criteria for Butt Fusion Joints

There is a need for ASME B&PV Code procedures and acceptance criteria for evaluation of flaws detected by inspection of high density polyethylene (HDPE) piping items in safety Class 3 systems. To support the development of flaw acceptance criteria for butt fusion joints in HDPE pipes, a series of coupon tests have been completed for specimens cut from butt fused HDPE pipes with surface or subsurface flaws placed in the joints prior to fusion process. Specimens containing known flaw sizes were tested under axial load at accelerated stresses and temperatures until failure; or until a prescribed number of test hours was reached. The failure time from the tests has been correlated to the net section stress and the stress intensity factor, and the results showed that the failure time can be better represented by the stress intensity factor. The test results were then used to fit the Brown and Lu formula that predicts the time to failure due to the slow crack growth of flaws as a function of stress intensity factor and temperature. With the developed Brown and Lu equation, the allowable stress intensity factors for a piping lifetime of 50 years at the maximum code allowable temperature of 60°C have been proposed for both surface and subsurface flaws in HDPE butt fusion joints. Examples of what might be corresponding allowable flaw sizes in the butt fusion joints of piping are also provided.

An Assessment of Uncertainty in Design Factors and Strain-Rate Inputs for Environmentally-Assisted Fatigue and Related Margins

In the case of ASME Class 1 pressure vessels and piping code, as in other similar codes, the design adequacy for fatigue is based on the cumulative usage factor (CUF), with recent augmentation to account for possible environmental effects. This deterministic quantification utilizes several engineering parameters (inputs) and (multiplicative) empirical factors. Although the fixed values of some of these design factors and S–N curves are based on underlying experimental data, the associated uncertainties are not explicit in the resulting fatigue assessment that is effectively based on the singular, calculated quantities of CUF and Fen, projected for a specified service. As such, the resulting fatigue margin and associated conservatism remain implicit or inconsistent and unquantifiable. At the same time, there is an increased demand for either extending the life of existing systems or for new systems with economically viable or better optimized fatigue designs. One approach to address this is to use a more realistic evaluation offered by probabilistic techniques that take into account the various uncertainties. Such an approach to supplement the deterministic analysis was recently proposed by the author keeping the existing and familiar framework of CUF based assessment, while satisfying acceptable component reliability to meet the fatigue design adequacy. The CUF formulation includes an explicit consideration of the k-factors (for material, loading history, surface and size effects) as adjustments to the S–N data. The objective of this paper is to assess the impact of k-factors and their uncertainty on the failure probability and on the number of load-cycles for specified target reliability. Also, similar assessment is made for the impact of strain-rate variable and its uncertainty on the allowable load-cycles. This is illustrated with a typical application of the CUF analysis of a safety injection nozzle safe-end. The approach taken consists of parametric analysis of the CUF-based probability of failure by individually removing the factors and/or their uncertainty, and comparing the results with the base case where all factors and associated uncertainties are maintained at their original values. Results of this analysis and their implications are discussed, along with a generally applicable relation between the deterministic CUF and the probability of failure.

Application of ASME Code Section XI Appendix L Using 3-D Finite Element Analyses

As part of a fatigue management program for subsequent license renewal, a flaw tolerance evaluation based on ASME Code, Section XI, Appendix L may be performed. The current ASME Code, Section XI, Appendix L flaw tolerance methodology requires determination of the flaw aspect ratio for initial flaw size calculation. The flaw aspect ratios listed in ASME Section XI, Appendix L, Table L-3210-2, for austenitic piping for example, are listed as a function of the membrane-to-gradient cyclic stress ratio. The Code does not explicitly describe how to determine the ratio, especially when utilizing complex finite element analyses (FEA), involving different loading conditions (i.e. thermal transients, piping loads, pressure, etc.). The intent of the paper is to describe the methods being employed to determine the membrane-to-gradient cyclic stress ratios, and the corresponding flaw aspect ratios (a/l) listed in Table L-3210-2, when using finite element analysis methodology. Included will be a sample Appendix L evaluation, using finite element analysis of a pressurized water reactor (PWR) pressurizer surge line, including crack growth calculations for circumferential flaws in stainless steel piping. Based on this example, it has been demonstrated that, unless correctly separated, the membrane-to-gradient cyclic stress ratios can result in extremely long initial flaw lengths, and correspondingly short crack growth durations.

Proposal for a Total Fatigue Life Assessment Methodology That Predicts Fatigue Life From Defect Initiation to Through Wall Leak

Even with improvements to remove excessive conservatisms, current fatigue assessment approaches can result in high Cumulative Usage Factors (CUFs) for some analyses. In order to improve plant availability from these assessments and mitigate future changes to design codes, an improvement in understanding in this area is desirable. Hence the proposal for a Life Assessment Methodology (LAM) was created. The LAM is a concept for an approach based on modelling each stage of fatigue life to predict total fatigue life, as a means of minimising conservatism in an assessment, where necessary. It should also be capable of incorporating statistical methods to assign reliability figures to calculated plant lives. This paper describes the proposed definition of the LAM and how a proof of concept version of the LAM was developed to assess the Bettis Bechtel Stepped Pipe (BBSP) test. The results were presented with two seeded cases (fixed inputs) and a range of lives corresponding to associated Target Reliabilities (TRs). The Best Estimate (BE) and TR associated lives produced were based on using the latest methods available for calculating Fatigue Initiation (FI) and Fatigue Crack Growth (FCG), whereas the seeded Effective Strain Range (ESR) comparison case used current deterministic assessment methods. The results for the case study concluded that there is a benefit to pursuing the development of the LAM when compared to traditional assessment methods. It highlighted and quantified the conservatism present in traditional assessment methods for these cases as well as the need to understand the required TR for a specific component as this can have a large effect on the predicted life. With further refinements to the method, a more realistic and robust output of the total fatigue life distribution (for specific cases) would be obtained, which in turn would allow us to better quantify the conservatism associated with a TR.

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

Recent Argonne National Laboratory Estimated Thermal Expansion Coefficients for Reactor Pressure Boundary Base Metal, Similar and Dissimilar Metal Welds

The paper presents estimated thermal expansion coefficients of 316SS and 508LAS base, 316SS filler or similar metal weld (SMW), and 316SS-508LAS dissimilar metal weld (DMW) filler and butter weld metals. These base and weld metals are typically used in nuclear reactor pressure boundary components. Accurate estimation of the expansion coefficients of these metals is essential for accurate estimation of thermal-mechanical stress in reactor pressure boundary components. In this paper we present the expansion coefficients of 316SS and 508LAS base, 316SS-SMW filler, and 316SS-508LAS DMW filler and butter weld metals. The coefficients were estimated based on our own experimental data. The corresponding expansion coefficient results and the FE validation results are presented in this paper. We anticipate that these types of results can be used as guidelines for choosing appropriate expansion coefficients for thermal-mechanical stress analysis of safety critical nuclear reactor components.

Codes and Standards for Managing Degradation of Boilers in Service

Degradation of pressure equipment is becoming an important issue due to increasing asset service time in process and power plants across Europe. For this reason it is important to assess life consumption of these assets to avoid catastrophic failures. Therefore it is necessary to refer to national/international normative on this subject. At present time the Italian thermotechnical committee (CTI) has drawn up a comprehensive set of norms which help the user to set up an inspection plan to investigate and assess degradation of pressure vessels and boilers. In the first part of this paper creep damage of Steam Generators is analyzed. For this purpose results of INAIL (Istituto Nazionale per l’ Assicurazione contro gli Infortuni sul Lavoro) database of steam boilers with 100’000 service hours or more is illustrated. Critical components are identified with reference to materials, geometry and operating parameters (pressure, temperature and time). At the end of the design life cycle, components of pressure equipment operated in creep regime must subjected to specific checks to estimate their residual life and the suitability for further use in safety conditions. The procedure allows to define reinspection intervals keeping acceptable the risk associated with the further use of the component related to creep even in evidence of defects in progress. The first check must be performed after 100,000 hours of effective use. Then, residual life evaluations must be repeated according to period of time that are defined as function of the results of all the checks carried out. In the second part of this paper boiler degradation is discussed with reference to NDT results and in-field inspection campaigns which are carried out traditionally after 45 years of service time, to minimize the risk of pressure components failures. In this paper results of different case studies are discussed with reference to degradation mechanisms and applicable standards.

Fracture Toughness Characterization in the Transition and Ductile Regime of an A508 Type Weld Metal With the Mini-CT Geometry Before and After Irradiation

The mini-CT geometry is being adopted worldwide for application to measure the fracture toughness properties. This geometry is particularly adapted for irradiated materials given the limited space available for irradiation in high flux material test reactors such as the BR2 reactor. A series of twenty (20) mini-CT specimens taken from an A508-type weld were irradiated in the BR2 reactor at 290 °C to a fluence level of ∼5 × 1019 n/cm2, E > 1MeV. They were precracked and 20 % side grooved before irradiation. In parallel, twenty four (24) unirradiated mini-CT specimens were available for testing. The main objective of this paper is to measure fracture toughness in the transition and ductile regime before and after irradiation and to compare the results with low flux surveillance data.

INCEFA-PLUS Project: Lessons Learned From the Project Data and Impact on Existing Fatigue Assessment Procedures

The INCEFA+ project (INcreasing Safety in nuclear power plants by Covering gaps in Environmental Fatigue Assessment) is a five year project supported by the European Commission HORIZON2020 programme, which will conclude in June 2020. This project aims to generate and analyse Environmental Assisted Fatigue (EAF) experimental data (approximately 230 fatigue data points generated on austenitic stainless steel), and focuses on the effect of several key parameters such as mean strain, hold times and surface finish, and how they interact with environmental effects (air or PWR environment). This work focuses on the analysis of the data obtained during the INCEFA+ project. More specifically, this paper discusses how the outcome of this analysis can be used to evaluate existing fatigue assessment procedures that incorporate environmental effects in a similar way to NUREG/CR-6909. A key difference between these approaches and the NUREG/CR-6909 is the reduction of conservatisms resulting from the joint implementation of the adjustment sub-factor related to surface finish effect (as quantified in the design air curve derivation) and a Fen penalization factor for fatigue assessment of a location subjected to a PWR primary environment. The analysis presented in this paper indicates that the adjustment (sub-)factor on life associated with the effect of surface finish in air (as described in the derivation of the design air curve in NUREG/CR-6909) leads to substantial conservatisms when it is used to predict fatigue lifetimes in PWR environments for rough specimens. The corresponding margins can be explicitly quantified against the design air curve used for EAF assessment, but may also depend on the environmental correction Fen factor expression that is used to take environmental effects into account.

Study on Acoustic Emission Detection Technology of Fiber Reinforced Plastic Pressure Vessel

Fiber reinforced plastics are used in pressure vessel manufacturing because of their high strength and corrosion resistance.Defects may occur in the manufacture and use of the pressure vessel. To ensure safe operation of the pressure vessel, it is necessary to conduct periodic safety assessment of the pressure vessel put into operation. It is difficult to evaluate the safety status of fiber-reinforced plastic pressure vessels by conventional nondestructive testing.Acoustic emission detection technology is a dynamic detection method, which has obvious advantages for the performance and fracture process of fiber reinforced plastic materials. ASME section V or ASTM section on acoustic emission detection of FRP pressure vessels, in which the localization of defects is mainly based on acoustic emission instruments. Due to the anisotropy of FRP material, the instrument can only give the area of the defect, and then use other non-destructive testing methods supplementary detection, so the author proposes a regional positioning method, which can locate defects more accurately. In this paper, acoustic emission detection method and lead breaking method were used to simulate the deficiency, and acoustic velocity attenuation and variation of fiber reinforced plastics were studied, and confirmative tests were carried out to obtain the positioning accuracy of the deficiency in different areas.In order to achieve the acoustic emission (AE) response behavior of stretching damage of glass fiber composites with fiber pre-broken and weak bonding, stretching tests and real-time AE monitoring of glass fiber composites were conducted.Experimental results showed that damage model such as matrix cracking and fiber fracture and bending could be occurred in the process of damage and failure. The composition and content of signal frequency of AE is also different because of difference of preset defect.