ASME Conference Presenter Attendance Policy and Archival Proceedings

2018;():V01AT00A001. doi:10.1115/PVP2018-NS1A.

This online compilation of papers from the ASME 2018 Pressure Vessels and Piping Conference (PVP2018) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Codes and Standards: API 579/ASME Code Fitness-for-Service Activities

2018;():V01AT01A001. doi:10.1115/PVP2018-84669.

Leak-Before-Break (LBB) has been applied in various industries for decades, and this paper explores using it for a new application. In the refining industry, various process units contain hydrogen at elevated temperatures where high temperature hydrogen attack (HTHA) can occur. This mechanism involves the reaction between hydrogen and carbides to form methane, but also the diffusion of hydrogen occurs in the steel. Under certain temperature and hydrogen partial pressures, the methane formation can cause grain boundary cavitation which leads to fissuring and eventually macroscopic cracking. Generally one designs to avoid such cracking from occurring following the so-called “Nelson Curves” contained in API RP 941; however, in recent years it has been found that non-stress relieved carbon steels are susceptible to HTHA below the original API 941 curve. As a result, the refining industry has experienced a number of leaks in piping and vessels. This paper presents some developmental efforts to apply LBB to non-stress relieved seamless carbon steel piping girth welds susceptible to (HTHA) cracking in refinery applications. Much of this approach builds on analyses, results, and experience from the commercial nuclear industry LBB efforts over the last 30 years. This paper will discuss the results of both mechanical testing as well as detailed modelling efforts to evaluate LBB technology to this new application for circumferential cracks, which to date implies that LBB may be applicable to seamless pipe girth welds. Cracks in tees or other components were not addressed in this work. Axial cracks in seam welds are not addressed in this work.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A002. doi:10.1115/PVP2018-84847.

Commonly used piping vibration screening limits are typically justified by experience and lack a well-documented technical basis. This paper presents technical background for future Level 1 Fitness-for-Service (FFS) vibration screening criteria. The criteria assess the risk of fatigue in process piping due to bending mode type vibrations. Finite element analysis (FEA) of 20,000 randomly generated candidate-piping models and high-cycle welded joint fatigue curves for both constant amplitude and variable amplitude loading form the stress limits and basis for the proposed criteria. Most importantly, the proposed criteria aligns with historically used allowable vibration limits rooted in substantial experience. The allowable stress basis implemented in this paper considers periodic and random vibrations making it applicable to situations of mechanically induced, two-phase flow induced, turbulent-induced vibration of single-phase process fluid, or wind-induced, which may be manifested as either periodic or random. To reduce conservatism, limits are set for butt-welded and non-butt welded mainline piping to prevent use of a single blanket limit that may lead to unnecessary piping support alterations/additions, or costly piping configuration changes and unit downtime. Furthermore, the proposed Level 1-type criteria are consistent with previously proposed FFS Level 2 and 3 piping vibration fatigue evaluations [1] intended for inclusion in the ASME FFS-1/API 579 (API 579) Fitness for Service Standard [2].

Commentary by Dr. Valentin Fuster
2018;():V01AT01A003. doi:10.1115/PVP2018-84858.

A recent comprehensive investigation into residual stress distributions in narrow gap welds in pressure vessels and pipe components are presented in this paper, covering component wall thickness from 1” (25.4mm) to 10” (254mm), component radius to wall thickness ratio from 2 to 100, and linear welding heating input from low (300 J/mm) to high (18000 J/mm). By means of a residual stress decomposition technique, two key parameters that govern through-thickness residual stress distributions in terms of their membrane and bending content have been identified. One is component radius to wall thickness ratio (r/t) and the other is a characteristic heat input density () having a unit of J/mm3. With these two parameters, a unified functional form for representing through-thickness residual stress profile in narrow gap welds is proposed for supporting fitness for service assessment, e.g., using f API 579-RP. Its validity is further confirmed by full-blown thermomechanical finite element residual stress analyses for a number of selected narrow gap weld cases.

Topics: Stress , Welded joints
Commentary by Dr. Valentin Fuster
2018;():V01AT01A004. doi:10.1115/PVP2018-84860.

As a further extension to the structural strain method first introduced by Dong et al [1], this paper presents an enhanced structural strain method which incorporates material nonlinearity and for two typical weld structures, i.e. weldment with plate sections (e.g. gusset weld or cruciform weld etc.) and weldment with beam sections. (e.g. pipe structures). A modified Ramberg-Osgood is introduced to capture nonlinear stress strain behavior of the material. A set of numerical algorithms is used to deal with complex stress state induced by structural effect such as beam section and plane strain condition. The proposed structural strain method is then applied to analysis of fatigue data of weldment made from different materials including steel, aluminum and titanium. It is shown that the enhanced structural strain method provides a unified way to correlate fatigue life of weldment in both high cycle and low cycle fatigue regime. The method is also used to study ratcheting problem raised up by Bree. A modified Bree diagram is given by considering material nonlinearity.

Commentary by Dr. Valentin Fuster

Codes and Standards: ASME Section XI Code Activities

2018;():V01AT01A005. doi:10.1115/PVP2018-84381.

Linear elastic fracture mechanics based flaw evaluation procedures in Section XI of the ASME Boiler and Pressure Vessel Code require calculation of the stress intensity factor (KI). The 2015 Edition of ASME Section XI [1] implemented a number of significant improvements in Article A-3000, including closed-form equations for calculating stress intensity factor influence coefficients (Gi) for circumferential flaws on the inside surface of cylinders. In the 2017 Edition [2], closed-form equations for axial flaws on the inside and outside surfaces of cylinders have been implemented.

In this paper, closed-form equations are developed for circumferential cracks on the OD surface of cylinders, based on tabular data from API 579 (2007 Edition) [3]. The equations presented, represent a complete set of Ri/t, a/t, and a/ ratios. The closed-form equations provide G0 and G1 coefficients while G2 through G4 are obtained using a weight function representation for the KI solutions for a surface crack. These equations permit the calculation of the Gi coefficients without the need to perform tabular interpolation. The equations are complete up to a fourth order polynomial representation of the applied stress. The fourth-order representation for stress will allow for more accurate fitting of highly non-linear stress distributions, such as those depicting high thermal gradients and weld residual stress fields.

Topics: Stress , Cylinders
Commentary by Dr. Valentin Fuster
2018;():V01AT01A006. doi:10.1115/PVP2018-84467.

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 procedure of flaw evaluation provided by Section XI of the ASME Boiler and Pressure Vessel Code. According to the Code, the range of stress intensity factor ΔK can be used to determine the fatigue crack growth rates of the material. However, it has been reported that crack closure behavior also strongly influence the fatigue crack growth rate under strong compressive load cycles.

This paper discusses the relation between ΔK and the fatigue crack growth rate for cracks in low alloy steels exposed to air. Compressive-tensile cyclic loadings were applied to center-notched plates to obtain the fatigue crack growth curves. The test data demonstrated that effective SIF range ΔKeff more accurately described the crack growth property due to plasticity induced crack closure. Comparing the test results with the reference crack growth curves in the ASME Code Section XI, it may seem that the crack growth prediction based on the Code underestimates the crack growth rates for compressive-tensile cyclic loadings under high stress level.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A007. doi:10.1115/PVP2018-84771.

The ASME Boiler & Pressure Vessel Code Section XI Committee is currently developing a new Division 2 nuclear code entitled the “Reliability and Integrity Management (RIM) program,” with which one is able to arrive at a risk-informed, NDE-based engineering maintenance decision by estimating and managing all uncertainties for the entire life cycle including design, material selection, degradation processes, operation and non-destructive examination (NDE). This paper focuses on the uncertainty of the NDE methods employed for preservice and inservice inspections due to a large number of factors such as the NDE equipment type and age, the operator’s level and years of experience, the angle of probe, the flaw type, etc. In this paper, we describe three approaches with which uncertainty in NDE-risk-informed decision making can be quantified: (1) A regression model approach in analyzing round-robin experimental data such as the 1981–82 Piping Inspection Round Robin (PIRR), the 1986 Mini-Round Robin (MRR) on intergranular stress corrosion cracking (IGSCC) detection and sizing, and the 1989–90 international Programme for the Inspection of Steel Components III-Austenitic Steel Testing (PISC-AST). (2) A statistical design of experiments approach. (3) An expert knowledge elicitation approach. Based on a 2003 Pacific Northwest National Laboratory (PNNL) report by Heasler and Doctor (NUREG/CR-6795), we observe that the first approach utilized round robin studies that gave NDE uncertainty information on the state of the art of the NDE technology employed from the early 1980s to the early 1990s. This approach is very time-consuming and expensive to implement. The second approach is based on a design-of-experiments (DEX) of eight field inspection exercises for finding the length of a subsurface crack in a pressure vessel head using ultrasonic testing (UT), where five factors (operator’s service experience, UT machine age, cable length, probe angle, and plastic shim thickness), were chosen to quantify the sizing uncertainty of the UT method. The DEX approach is also time-consuming and costly, but has the advantage that it can be tailored to a specific defect-detection and defect-sizing problem. The third approach using an expert panel is the most efficient and least costly approach. Using the crack length results of the second approach, we introduce in this paper how the expert panel approach can be implemented with the application of a software package named the Sheffield Elicitation Framework (SHELF). The crack length estimation with uncertainty results of the three approaches are compared and discussed. Significance and limitations of the three uncertainty quantification approaches to risk assessment of NDE-based engineering decisions are presented and discussed.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A008. doi:10.1115/PVP2018-84809.

Limit load solutions have been applied to estimate the collapse load of a component made of ductile material. Worldwide maintenance codes for power plants, such as ASME Boiler and Pressure Vessels Code, Section XI, and JSME fitness-for-service code, describe limit load solutions under the assumption of a single flaw. Detected flaws are, however, not always a single flaw, and adjacent flaws due to stress corrosion cracking have been detected in power plants. Thus, development of a limit load solution to estimate the collapse load in the case of multiple flaws remains an issue of structural integrity evaluation. Under the aim of developing a method for evaluating the effect of multiple flaws on collapse load as a part of a limit load solution, fracture tests of flat plates and pipes with multiple flaws were conducted. Although experimental approaches have been attempted to establish the evaluation method, further efforts are required to incorporate the evaluation procedure into a code rule.

Effective parameters for considering reduction of collapse load on the basis of test results for specimens with multiple flaws were identified. Test results clearly show a correlation between collapse load and ratios of net-section areas. This correlation leads to the conclusion that distance parameters and flaw length of a smaller flaw determine the existence of an effect on the collapse load by multiple flaws. To investigate the physical sense of the correlation, finite element analysis (FEA) was performed. The FEA results show that strain distributions at the flaw tip under several conditions correspond at the time of maximum load of the fracture tests regardless of the effect of multiple flaws. Also according to the FEA results, the extent of the strain field is linearly proportional to flaw length. These FEA results are consistent with the correlation obtained by the test results.

Topics: Stress
Commentary by Dr. Valentin Fuster
2018;():V01AT01A009. doi:10.1115/PVP2018-84940.

Fatigue crack growth rates are expressed as a function of the stress intensity factor ranges. The fatigue crack growth thresholds are important characteristics of fatigue crack growth assessment for the integrity of structural components. Almost all materials used in these fatigue tests are ferritic steels. As a result, the reference fatigue crack growth rates and the fatigue crack growth thresholds for ferritic steels were established as rules and they were provided by many fitness-for-service (FFS) codes. However, the thresholds are not well defined in the range of negative stress ratio. There are two types of thresholds under the negative stress ratio. That is, constant thresholds and increment of thresholds with decreasing stress ratios. The objective of this paper is to introduce the thresholds provided by FFS codes and to analyze the thresholds using crack closure. In addition, based on the experimental data, definition of the threshold is discussed to apply to FFS codes. Finally, threshold for ferritic steels under the entirely condition of stress ratio is proposed to the ASME Code Section XI.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A010. doi:10.1115/PVP2018-84961.

Crack closure during fatigue crack growth is an important phenomenon for predicting fatigue crack growth amount. Many experimental data show that fatigue cracks close at not only negative loads but also positive loads during constant amplitude loading cycles, depending on applied stress levels.

The Appendix A-4300 in the ASME Code Section XI provides two equations of fatigue crack growth rates expressed by stress intensity factor range for ferritic steels under negative stress ratio. The boundary of two fatigue crack growth rates is classified by the magnitude of applied stress intensity factor range with the consideration of crack closure.

The objective of this paper is to investigate the influence of the magnitude of the stress intensity factor range on crack closure. Fatigue tests have been performed on ferritic steel specimens in air environment at room and high temperatures. Crack closures were obtained as a parameter of stress ratio. It was found that crack closure occurs at a smaller applied stress intensity factor range than the definition given by the Appendix A-4300.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A011. doi:10.1115/PVP2018-85051.

In ASME Section XI Appendix C for analytical evaluation of flaws in piping, a screening procedure is prescribed to determine the failure mode and analysis method for the flawed pipe. The end-of-evaluation period flaw dimensions, temperature, material properties, and pipe loadings are considered in the screening procedure. Equations necessary to calculate components of the screening criteria (SC) include stress intensity factor (K) equations. The K-equation for a pipe with a circumferential inside surface flaw in the 2017 Edition Section XI Appendix C-4000 is for a fan-shaped flaw. Real surface flaws are closer to semi-elliptical shape. As part of Section XI Working Group on Pipe Flaw Evaluation (WGPFE) activities, revision to stress intensity factor equations for circumferential surface flaws in Appendix C-4000 has been proposed. The proposed equations include closed-form equations for stress intensity influence coefficients G0 for membrane stress and Ggb for global bending stress for circumferential inside surface flaws. The rationale for the Code changes and technical basis for the proposed stress intensity factor equations are provided in this paper.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A012. doi:10.1115/PVP2018-85093.

Article A-3000 of Appendix A in ASME Section XI provides methods to calculate stress intensity factors that are used in Section XI linear elastic fracture mechanics based flaw evaluation procedures. The ASME Section XI Working Group on Flaw Evaluation has been in the process of rewriting Article A-3000 of Appendix A. The rewrite of Article A-3000 includes implementation of closed-form equations for stress intensity factor influence coefficients for cylinder geometries.

Closed-form relations for stress influence coefficients G0 and G1 for axial flaws on the outside surface in cylinders were recently developed and implemented into the 2017 Edition of ASME Section XI Appendix A. The closed-form equations were implemented with one restriction on the application related to very long flaws. This restriction was taken as an interim approach to addressing a technical concern from the US NRC staff. NRC staff had technical concern on the large percentage fitting errors for the G1 influence coefficients at surface point for some very long flaws. An action was assigned within the ASME Section XI Working Group on Flaw Evaluation to investigate the accuracy of surface point G values for very long flaws. The intent of the investigation is to provide technical justification for using the closed-form equations with no restriction and to identify any issues in the source data or during the fitting process. This paper describes current results from this ongoing investigation.

Commentary by Dr. Valentin Fuster

Codes and Standards: Environmental Fatigue Issues

2018;():V01AT01A013. doi:10.1115/PVP2018-84034.

INCEFA-PLUS is a five-year project supported by the European Commission HORIZON 2020 programme. The project commenced in mid-2015. Sixteen organisations from across Europe have combined forces to deliver new experimental data which will support the development of improved guidelines for assessment of environmental fatigue damage to ensure safe operation of nuclear power plants. Within INCEFA-PLUS, the effects of mean strain and stress, hold time, strain amplitude and surface finish on fatigue endurance of austenitic stainless steels in light water reactor environments are being studied experimentally, these being issues of common interest to all participants. The data obtained is being collected and standardised in an online environmental fatigue database, implemented with the assistance of an INCEFA-PLUS led CEN (European Committee for Standardization) workshop on this aspect. Later in the project it is planned that INCEFA-PLUS will develop and disseminate methods for including the new data into assessment approaches for environmental fatigue degradation.

This paper provides an overall update to project developments since it was last presented at PVP2017 [[1]]. As well as being a standalone paper, the paper will also serve as an introduction to more detailed papers also being submitted covering 4 specific aspects of the project. In particular, this paper summarises:

• The results for Phase 1 testing.

• The agreed plans for Phase 2 testing

• A summary of emerging sensitivities to, and inter-dependencies between:

○ Mean strain and stress, surface finish, strain amplitude and hold time.

○ Environment

○ Material

○ Laboratory

• Latest thinking on direction for the project in its last two years.

• The latest thoughts on how the project results will be used to advance development of improved assessment guidelines.

• A summary of dissemination achieved and planned for the forthcoming year.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A014. doi:10.1115/PVP2018-84081.

INCEFA-PLUS stands for INcreasing safety in nuclear power plants by Covering gaps in Environmental Fatigue Assessment. It is a five year project supported by the European Commission HORIZON2020 program that commenced in mid-2015 and in which sixteen organizations from across Europe participate. Specifically, the effects of mean strain/stress, hold time, strain amplitude and surface finish on fatigue life of austenitic stainless steels in light water reactor environments are being studied, these being issues of common interest to all participants.

The project will develop proposals for improvements to methods for environmental fatigue assessment of nuclear plant components. Therefore, extensive testing capacity is being solicited in various laboratories across Europe in order to add to the existing amount of published data on environmentally assisted fatigue.

Since there currently is no standard on environmental fatigue testing, it was imperative to come up with and agree upon a testing procedure within the consortium to minimize lab-to-lab variations in test results. This was done prior to the first phase of testing, but an update of the procedure was required after review of initial results, when additional potential lab-to-lab differences were identified. The current status of the so-called test protocol, and the key areas of difference found between different testing facilities, will be discussed.

Due to the large test matrix within INCEFA-PLUS, distributed amongst various test laboratories, it has been necessary to develop a method to assign a data quality level to each test result, and a minimum data quality requirement for results that will be included in the project’s datasets used for analysis. Furthermore, the project has triggered international interest in facilitating mutual data access, and this requires data is gathered in a common database with data quality ratings applied. Ways to address the evaluation of data quality will be discussed.

In a way, both activities, on a test protocol and on data review, jointly contribute to data quality by, respectively, ensuring a pre-test, common test procedure and a post-test, harmonized data evaluation.

The large number of participants in the INCEFA-PLUS project presents a unique opportunity to gain consensus on light water reactor environment fatigue testing procedures and data quality assessment from experts working in a range of different organizations. The test protocol and data quality ratings developed within the INCEFA-PLUS project could be adopted by other organizations, or possibly used as the basis for future testing standards documents to harmonize approaches across the nuclear industry.

Topics: Fatigue , Testing
Commentary by Dr. Valentin Fuster
2018;():V01AT01A015. doi:10.1115/PVP2018-84197.

The update of the ASME III design fatigue curve for stainless steel in conjunction with the Fen model described in the NUREG/CR-6909 report has been criticized since publication. Data used to develop curves and models raises more questions than it answers.

Material testing in a simulated light water reactor environment is difficult due to the temperature and pressure involved. The experimental challenge makes it tempting to take shortcuts where they should least be taken. Facing and overcoming the challenges, direct strain-controlled fatigue testing has been performed at VTT using a unique tailored-for-purpose EAF facility. The applicable ASTM standards E 606 and E1012 are followed to provide results that are directly compatible with ASME Code Section III.

Several earlier PVP papers (PVP2016-63291, PVP2017-65374) report lower than calculated experimental Fen factors for stabilized stainless steels. In this paper new results, in line with the previous years’ conclusions, are presented for nonstabilized AISI 304L tested with dual strain rate waveforms.

To model environmental effects more accurately, an approach accounting for the damaging effect of plastic strain is proposed. A draft Fen model, similar in structure to the NUREG model but with additional parameters, is shown to significantly improve the accuracy of Fen prediction.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A016. doi:10.1115/PVP2018-84240.

Additional fatigue rules within the ASME Boiler and Pressure Vessel Code have been developed over the past decade or so, such as those in Code Case N-792-1 [1], which provides an acceptable method to describe the effects of BWR and PWR environments on the fatigue life of components. The incorporation of environmental effects into fatigue calculations is performed via an environmental factor, Fen, and depends on factors such as the temperature, dissolved oxygen and strain rate. In the case of strain rate, lower strain rates (i.e., from slow transients) aggravate the Fen factor which counters the long-held notion that step (fast) transients cause the highest fatigue usage.

A wide range of other factors, such as surface finish, can have a deleterious impact on fatigue life, but their impact on fatigue life is typically considered by including transition sub-factors to construct the fatigue design curve from the mean behavior air curve rather than in an explicit way, such as the Fen factor. An extensive amount of testing and evaluation has been conducted and reported in References [2] [3] [4] [5] [6] [7] and [8] that were used to both revise the transition factors and devise the Fen equations contained in Code Case N-792-1.

The testing supporting the definition of Fen was performed on small-scale laboratory specimens with a polished surface finish on the basis that the Fen factor is applicable to the design curve without any impact on the transition factors.

The work initiated by AREVA in 2005 [4] [5] [6] suggested, in testing of austenitic stainless steels, an interaction between the two aggravating effects of surface finish and PWR environment on fatigue damage.

These results have been supported by testing carried out independently in the UK by Rolls-Royce and AMEC Foster Wheeler (now Wood Group) [7], also on austenitic stainless steels. The key finding from these investigations is that the combined detrimental effects of a PWR environment and a rough surface finish are substantially less than the sum of the two individual effects. These results are all the more relevant as most nuclear power plant (NPP) components do not have a polished surface finish. Most NPP component surfaces are either industrially ground or installed as-manufactured.

The previous studies concluded that explicit consideration of the combined effects of environment and surface finish could potentially be applicable to a wide range of NPP components and would therefore be of interest to a wider community: EDF has therefore authored a draft Code Case introducing a factor, Fen-threshold, which explicitly quantifies the interaction between PWR environment and surface finish, as well as taking some credit for other conservatisms in the sub-factors that comprise the life transition sub-factor used to build the design fatigue curve .

The contents of the draft Code Case were presented last year [9]. Since then, other international organizations have also made progress on these topics and developed their own views. The work performed is applicable to Austenitic Stainless Steels only for the time being.

This paper aims therefore to present an update of the draft Code Case based on comments received to-date, and introduces some of the research and discussions which have been ongoing on this topic as part of an international EPRI collaborative group on environmental fatigue issues. It is intended to work towards an international consensus for a final version of the ASME Code Case for Fen-threshold.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A017. doi:10.1115/PVP2018-84251.

High temperature water environments typical of LWR operation are known to significantly reduce the fatigue life of reactor plant materials relative to air environments in laboratory studies. This environmental impact on fatigue life has led to the issue of US-NRC Regulatory Guide 1.207 [1] and supporting document NUREG/CR-6909 [2] which predicts significant environmental reduction in fatigue life (characterised by an environmental correction factor, Fen) for a range of actual and design basis transients. In the same report, a revision of the fatigue design curve for austenitic stainless steels and Ni-Cr-Fe alloys was proposed [2]. This was based on a revised mean curve fit to laboratory air data and revised design factors to account for effects not present in the test database, including the effect of rough surface finish. This revised fatigue design curve was endorsed by the NRC for new plant through Regulatory Guide 1.207 [1] and subsequently adopted by the ASME Boiler and Pressure Vessel (BPV) Code [3]. Additional rules for accounting for the effect of environment, such as the Fen approach, have been included in the ASME BPV Code as code cases such as Code Case N-792-1 [4].

However, there is a growing body of evidence [5] [6] [7] and [8] that a rough surface condition does not have the same impact in a high temperature water environment as in air.

Therefore, application of Fen factors with this design curve may be unduly conservative as it implies a simple combination of the effects of rough surface and environment rather than an interaction. Explicit quantification of the interaction between surface finish and environment is the aim of a number of recent proposals for improvement to fatigue assessment methods, including a Rule in Probationary Phase in the RCC-M Code and a draft Code Case submitted to the ASME BPV Code as described in References [9] and [10]. These approaches aim to quantify the excessive conservatism in current methods due to this unrecognised interaction, describing this as an allowance for Fen effectively built into the design curve. A number of approaches in various stages of development and application are discussed further in a separate paper at this conference [11].

This paper reports the results of an extensive programme of strain-controlled fatigue testing, conducted on two heats of well-characterised 304-type material in a high-temperature simulated PWR environment by Wood plc. The baseline behaviour in environment of standard polished specimens is compared to that of specimens with a rough surface finish bounding normal plant component applications. The results reported here substantially add to the pool of data supporting the conclusion that surface finish effects in a high-temperature water environment are significantly lower than the factor of 2.0 to 3.5 assumed in construction of the current ASME III fatigue design curve. This supports the claim made in the methods discussed in [9] [10] and [11] that the fatigue design curve already incorporates additional conservatism for a high-temperature water environment that can be used to offset the Fen derived by the NUREG/CR-6909 methodology. At present, this observation is limited to austenitic stainless steels.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A018. doi:10.1115/PVP2018-84252.

Light water reactor coolant environments are known to significantly enhance the fatigue crack growth rate of austenitic stainless steels. However, most available data in these high temperature pressurized water environments have been derived using specimens tested at positive load ratios, whilst most plant transients involve significant compressive as well as tensile stresses. The extent to which the compressive loading impacts on the environmental enhancement of fatigue crack growth, and, more importantly, on the processes leading to retardation of those enhanced rates is therefore unclear, potentially leading to excessive conservatism in current assessment methodologies.

A test methodology using corner cracked tensile specimens, and based on finite element analysis of the specimens to generate effective stress intensity factors, Keff, for specimens loaded in fully reverse loading has been previously presented. The current paper further develops this approach, enabling it to be utilized to study a range of positive and negative load ratios from R = −2 to R = 0.5 loading, and provides a greater understanding of the development of stress intensity factor within a loading cycle.

Test data has been generated in both air and high temperature water environments over a range of loading ratios. Comparison of these data to material specific crack growth data from conventional compact tension specimens and environmental crack growth laws (such as Code Case N-809) enables the impact of crack closure on the effective stress intensity factor to be assessed in both air and water environments. The significance of indicated differences in the apparent level of closure between air and water environments is discussed in the light of accepted growth laws and material specific data.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A019. doi:10.1115/PVP2018-84301.

Environmentally Assisted Fatigue (EAF) is receiving nowadays an increased level of attention for existing Nuclear Power Plants (NPPs) as utilities are now working to extend their life. In the wake of numerous experimental fatigue tests carried out in air and also in a PWR environment, the French RCC-M code [1] has recently been amended (in its 2016 edition) with two Rules in Probatory Phase (RPP), equivalent to ASME code-cases, “RPP-2” and “RPP-3” [2] [3]. RPP-2 consists of an update of the design fatigue curve in air for stainless steels (SSs) and nickel-based alloys, and is also associated with RPP-3 which provides guidelines for incorporating the environmental penalty “Fen” factor in fatigue usage factor calculations.

Alongside this codification effort, an EAF screening has recently been carried out within EDF DT [4] on various areas of the primary circuit of the 900 MWe plants of the EDF fleet. This screening led to the identification of a list of 35 “sentinel locations” which are defined as areas most prone to EAF degradation process. These locations will be subjected to detailed EAF analysis in the stress report calculations (according to the above-mentioned RCC-M code cases) for the fourth decennial inspection of the 900 MWe (VD4 900 MWe) power plants.

The potential impact of EAF on the secondary circuit components is another question to address in anticipation of the VD4 900 MWe, as they may be considered as class 1 or class 2 equipment for RCC-M application according to the equipment specification. This paper presents the approach proposed by EDF towards an exemption of environmental effects consideration for secondary circuit components. The argument is first based on a review of experimental campaigns led in Japan and France (respectively on fatigue test specimens and at the component scale) which indicate a Dissolved Oxygen (DO) content threshold below which environmental effects are almost inexistent. The (conservative) value of 40 ppb has been selected consistently with NUREG/CR-6909 revision 0 [5]. The second part of the argument is built, on the one hand, on the analysis of the EDF Technical Specifications for Operation (STE) which narrows the scope of the study only to unit outages, and, on the other hand, on the analysis of 5 years of operations of all 900 MWe plants of the EDF fleet (equivalent to 170 reactor-years). It has been shown that the DO content rarely exceeded the 40 ppb threshold in the secondary coolant, and that in this case, the considered locations were not submitted to any fatigue loading.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A020. doi:10.1115/PVP2018-84346.

One of the goals of ASME Section XI is to ensure that systems and components remain in safe operation throughout the service life, which can include plant license extensions and renewals. This goal is maintained through requirements on periodic inspections and operating plant criteria as contained in Section XI IWB-2500 and IWB-3700, respectively. Operating plant fatigue concerns can be caused from operating conditions or specific transients not considered in the original design transients. ASME Section XI IWB-3740, Operating Plant Fatigue Assessments, provides guidance on analytical evaluation procedures that can be used when the calculated fatigue usage exceeds the fatigue usage limit defined in the original Construction Code. One of the options provided in Section XI Appendix L is through the use of a flaw tolerance analysis.

The flaw tolerance evaluation involves postulation of a flaw and predicting its future growth, and thereby establishing the period of service for which it would remain acceptable to the structural integrity requirements of Section XI. The flaw tolerance approach has the advantage of not requiring knowledge of the cyclic service history, tracking future cycles, or installing systems to monitor transients and cycles. Furthermore, the flaw tolerance can also justify an inservice inspection period of 10 years, which would match a plant’s typical Section XI in-service inspection interval. The goal of this paper is to demonstrate a flaw tolerance evaluation based on ASME Section XI Appendix L for several auxiliary piping systems for a typical PWR (Pressurized Water Reactor) nuclear power plant. The flaw tolerance evaluation considers the applicable piping geometry, materials, loadings, crack growth mechanism, such as fatigue crack growth, and the inspection detection capabilities. The purpose of the Section XI Appendix L evaluation is to demonstrate that a reactor coolant piping system continues to maintain its structural integrity and ensures safe operation of the plant.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A021. doi:10.1115/PVP2018-84461.

Change in the fatigue life due to application of the mean strain was investigated for Type 316 stainless steel in simulated pressurized water rector (PWR) primary water environment. The tests were conducted by controlling the strain range to 1.2% for different strain rates of 0.4, 0.004, or 0.001%/s. The applied mean strain was 15% in nominal strain. In addition, cold worked specimens were also subjected to the tests without applying the mean strain. The tests using the cold worked specimens were regarded as the tests with the mean strain without increase in surface roughness due to application of plastic deformation. By inducing the cold working at low temperature, the effect of martensitic phase on the fatigue life was also examined. It was shown that the fatigue life of the stainless steel was reduced in the PWR water environment and the degree of the fatigue life reduction was consistent with the prediction model prescribed in the code issued by the Japan Society of Mechanical Engineers (JSME) and NUREG/CR-6909. Increases in peak stress and stress range due to cold working did not cause any apparent influence on the fatigue life. It was also shown that the 10.5 wt% martensitic phase induced by the low temperature cold working and the increase in the surface roughness caused by application of 15% mean strain did not bring about further fatigue life reduction. It was concluded that the effects of the mean strain, cold working, and martensitic phase were minor on the fatigue life in the PWR water environment. The current JSME and NUREG/CR-6909 models were applicable for predicting the reduction in fatigue due to the PWR water environment even if the mean strain or cold working was applied.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A022. doi:10.1115/PVP2018-84490.

In order to conduct effective and rational maintenance activity of components in nuclear power plants, it is proposed to manage fatigue degradation based on crack size corresponding to an extent of cumulative fatigue damage. The purpose of this study focuses on the influence of strain rate in simulated reactor coolant environment for fatigue crack initiation and growth. 3-dimensional replica observations were conducted for environmental fatigue test specimens in different strain rates. Crack initiation and growth were observed in the experiments. It is clarified that low strain rate influences crack propagation and coalescence and increases crack growth rate that finally decrease fatigue life.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A023. doi:10.1115/PVP2018-84561.

The Sodium Fast-cooled Reactor (SFR), are generation IV nuclear power plants, have a target operating temperature of 550°C which makes creep-fatigue behavior more critical than a generation III nuclear power plants. So it is important to understand the nature of creep-fatigue behavior of the piping material, Grade 91 steel. The creep-fatigue damage diagram of Grade 91 steel used in ASME-NH was derived using a conventional time-fraction testing method which was originally developed for type 300 stainless steels. Multiple studies indicate that the creep-fatigue damage diagram of Grade 91 steel developed using this testing method has excessive conservatism in it. Therefore, an alternative testing method was suggested by separating creep and fatigue using interrupted creep tests. The suggested method makes it possible to control creep life consumption freely which was difficult with the previous method. It also makes it easier to observe the interaction between creep and fatigue mechanisms and microstructural evolution. In conclusion, an alternative creep-fatigue damage diagram for Grade 91 steel at 550°C was developed using an interrupt creep fatigue testing method and FE model simulation.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A024. doi:10.1115/PVP2018-84610.

Fatigue testing campaigns are a common feature in the design and operation of advanced engineering systems in the aerospace and power generation sectors. The resulting data are typically of a high inherent technical and financial value. Presently, these data are typically transferred between departments and companies by way of ad-hoc solutions reliant on obsolete or proprietary technologies, including CSV files, MS Excel® files, and PDFs. In these circumstances there is significant potential for data loss, inconsistency, and error. To address these shortcomings, there is a need for a systematic means of transferring data between different digital systems. With this in mind, a series of CEN Workshops on engineering materials data have taken place with a view to developing technologies for representing and exchanging engineering materials data. Most recently, a CEN Workshop on the topic of fatigue test data has delivered data formats derived from the ISO 12106 standard for axial strain-controlled fatigue testing. This paper describes the methodology for developing the data formats and demonstrates their use in the scope of the INCEFA-PLUS project on increasing safety in nuclear power plants by covering gaps in environmental fatigue assessment.

Topics: Fatigue testing
Commentary by Dr. Valentin Fuster
2018;():V01AT01A025. doi:10.1115/PVP2018-84698.

In the past 10 years, different laboratory test results lead the International Standard Development Organizations (SDO) to review their fatigue design rules in different directions, in particular to consider consequences of environmental effects on existing design rules.

The key document that ask different questions to Code developers is the USNRC NUREG 6909 report: “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials” that confirms some environmental effects on S-N fatigue tests on small specimen. The open question is: how to transfer these results to Fatigue Design Rules of plant components?

This paper will review existing codified rules in major nuclear Codes; in particular USA ASME Boiler and Pressure Vessel Code Section III and French AFCEN RCC-M Code.

The paper will make a first historical background of these Codes and analyze background of these rules by consideration of field experience and existing tests results.

To conclude, the paper will summarize to-day “fatigue road maps” to evaluate margins and screening criteria to assure reliable and safe codified design fatigue life evaluation.

Topics: Fatigue , Design
Commentary by Dr. Valentin Fuster
2018;():V01AT01A026. doi:10.1115/PVP2018-84879.

Environmentally assisted fatigue of nuclear plant materials in the Pressurised Water Reactor (PWR) coolant environment is a phenomenon that has been extensively studied over the past 30 years. Methods for accounting for the PWR environment in an ASME Section III fatigue assessment are presented in NUREG/CR-6909. The deleterious effect of environment is described through a Fen factor dependent upon strain rate, temperature and the dissolved oxygen content of the water. The formulae which describe the Fen are based upon correlations observed in test data, predominantly from tests conducted with constant temperature and strain rate (triangular or sawtooth loading). Actual loading histories encountered during service are far more complex, with both strain rate and temperature, and therefore Fen, varying through the cycle. NUREG/CR-6909 Draft Rev 1 recommends the Modified Rate Approach (MRA) to account for this type of loading.

There is a substantial and growing body of data for conditions in which the strain rate and/or temperature change within the load cycle, for which MRA does not generally perform well in describing the deleterious effect of environment in these complex waveform conditions. In particular, MRA does not predict the observed difference in life when the temperature is varied in-phase or out-of-phase within the strain waveform, or when the slow portion of the strain rate is moved from the top to the bottom of the waveform.

An alternative approach called the Strain-Life Weighted (SNW) Fen method was presented in PVP2017-66030 and additional validation testing was proposed.

This paper develops the SNW method further into a general approach for all stainless steels and presents additional new validation data, including a range of isothermal and non-isothermal plant realistic waveforms and a more extensive review of open literature data.

It is concluded that the SNW method offers a significant improvement in fatigue life prediction capability for plant realistic complex waveforms compared to MRA and provides residuals similar to that of standard waveform data. It is thus considered to be suitably validated to propose a code case for use in ASME Section III fatigue assessments.

Topics: Temperature
Commentary by Dr. Valentin Fuster
2018;():V01AT01A027. doi:10.1115/PVP2018-84923.

Pressurized water reactor environments are known to reduce the fatigue life of austenitic stainless steel components when compared to air environments. Laboratory testing has provided a means of quantifying this, allowing conservative plant assessments to be made. The majority of this testing has been isothermal and carried out on membrane loaded hollow or solid specimens. The geometry and loading of laboratory test specimens is significantly different to that experienced on plant, where complex strain waveforms are generally out of phase with temperature changes, and significant through wall strain gradients may be present. To address the issue of realistic loading, a test facility has been developed which can simulate thermal shock loading on a tubular specimen.

The capability of the test facility was presented at the PVP2016 conference [PVP2016-63161]. Since then the facility has evolved, with modifications made to the rig configuration and specimen geometry in order to maximize the strain amplitude from the thermal shock, including the adoption of an annular flow geometry. These modifications were designed to optimize both the heat transfer coefficient and the speed of cycling between hot and cold water in order to induce a thermal strain that can cause mechanical failure within practicable test durations. In order to calculate the magnitude of the thermal strain, detailed calculations were required both in terms of thermal hydraulics as well as stress analyses. The latest stress analysis has been combined with state of the art life prediction models to estimate the time for crack initiation.

This paper presents the results of the latest stress analysis and life prediction, including the derivation of the heat transfer coefficient for an annular flow region. The life prediction method uses best estimate strain-temperature histories from elastic-plastic finite element analysis (FEA). Heat-specific material properties have been developed during accompanying tests within the same experimental programme, and have been applied to enable cyclic hardening to be taken into account. The comparison of the prediction to an on-going test is also discussed.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A028. doi:10.1115/PVP2018-84935.

Strain controlled LCF testing extended to 10 million cycles revealed an abrupt endurance limit enforced by secondary hardening. In elevated temperatures the ε-N curve is rotated and endurance limit is lowered, but not vanished. When very low strain rates are applied at 325°C in simulated PWR environment, fatigue life is reduced, but far less than predicted according to NUREG/CR-6909. It is possible, but not probable that the difference is due to different stainless grades studied. We assume that the test method plays a more important role.

We have repeatedly demonstrated in different tests campaigns that interruptions of straining with holds aiming to simulate steady state normal operation between fatigue relevant cycles can notably extend the fatigue endurance. Further proof is again presented in this paper. The suspected explanation is prevention of strain localization within the material microstructure and also in geometric strain concentrations. This actually suggests, that hold effects should be even more pronounced in real components.

Cyclic behavior of austenitic steels is very complex. Transferability of laboratory data to NPP operational conditions depends on test environment, temperature, strain rate and holds in many ways not considered in current fatigue assessment procedures. In addition to penalty factors, also bonus factors are needed to improve transferability. Furthermore, it seems that the load carrying capacity of fatigued stainless steel is not compromised before the crack growth phase. Tensile tests performed after fatigue tests interrupted shortly before end-of-life condition in 325°C (N ≈ 0.85 × N25) showed strength and ductility almost identical to virgin material.

This paper provides new experimental results and discusses previous observations aiming to sum up a state of the art in fatigue performance of German NPP primary loop materials.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A029. doi:10.1115/PVP2018-84936.

Unexpected findings on time and temperature dependent behavior have been recorded during our research on fatigue performance of niobium stabilized stainless steel.

Cyclic straining at 325°C and low strain rates resulted in higher stress responses than in higher rate tests. This effect is particular strong in PWR water environment. Subsurface bulk effect in environment is in contrast to the assumption on similar responses in air and environment, which is the foundation of the ‘companion specimen’ method where the strain in environment is measured from a parallel specimen similarly tested in air.

Our data shows that environmental effects caused by PWR water cannot be isolated as a separate issue. Environment, temperature and strain rate are factors, which interactively affect the cyclic response and fatigue performance of stainless steel in relevant temperatures and loading conditions.

The current ASME Code Section III design curve is based on different translation of the laboratory data than that made by Langer et al. The resulting effect is not as radical as caused by replacement of the original test data to a new highly scattered data base. But also the procedural changes have detectable effects and open a door for a provocative question:

“is the design curve still compatible with the code?”

Topics: Temperature , Pipes
Commentary by Dr. Valentin Fuster

Codes and Standards: Fatigue and Ratcheting Issues in Pressure Vessel and Piping Design

2018;():V01AT01A030. doi:10.1115/PVP2018-84233.

This paper describes an experimental validation of the enhanced reference stress method to calculate fatigue J-integral ranges, which are effective in predicting the fatigue crack propagation rate under low–cycle fatigue loadings. Although J-integral type fracture mechanics parameters can be calculated via elastic–plastic finite element analysis (FEA) of the crack geometry, performing such an analysis is costly and requires a high–end computer. A simplified method for estimating the elastic–plastic J-integral is therefore desired. Herein, several representative simplified methods for estimating the elastic–plastic J-integral were applied to crack propagation prediction and compared with each other. The experiments referred to was a previously performed cyclic bending tests using wide–plate specimens containing a semielliptical surface crack. Limit load correction factors to improve the accuracy of the reference stress method were estimated by performing an elastic–plastic FEA. The predicted crack propagation behaviors were compared against the test results.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A031. doi:10.1115/PVP2018-84952.

The loads on the equipment nozzles are generally generated by the piping stress engineer by doing the stress analysis of entire closed loop systems. Subsequently the nozzle loads are passed on to the engineers of the pressure vessel equipment. The value of the loads which have been worked out for the nozzle mostly depends upon the methods/concept by which the piping stress engineer has evaluated the piping loop. Nozzle flexibility/stiffness is the important parameter in evaluation of various components of nozzle loads. The objective of this paper is to explain the effect/influence of flexibility/stiffness generated from three different methods (Anchor, WRC and Finite element method) on nozzle load evaluation and shell/nozzle junction stresses.

WRC297 bulletin [6] gives the reference to nozzle flexibility in the appendix A, example no.3.

The work presented in this paper is an attempt to compare the nozzle loads calculated by evaluating the flexibilities/stiffness in various methods. Further an attempt has been made to consolidate the results of junction local stresses obtained by the various methods of stiffness/flexibilities which would result in realistic results and overall code acceptable stresses without the results being either overly conservative or unconservative.

Topics: Stress , Nozzles , Stiffness
Commentary by Dr. Valentin Fuster
2018;():V01AT01A032. doi:10.1115/PVP2018-85050.

The ASME Boiler and Pressure Vessel Code (Section III and Section VIII) provides requirements to avoid a ratcheting (accumulating permanent strain) condition under cyclic thermal load application. The ratchet check in this code is based on the solutions presented by Miller in 1959. One important focus in Miller’s work was to estimate the accumulated plastic strain under cyclic loading. The existing pressure vessels and piping codes have been adopting Miller’s ratchet boundary solution where there is no cyclic plastic accumulation of strain. However, some of these codes also provide limit on accumulated plastic strain under ratcheting conditions. Since the cyclic loading also causes fatigue damage in thee component, the question how to account for the interaction of ratchet deformation, which may contribute to damage in the material, and fatigue damage arises, since the fatigue curves are obtained from tests in the absence of ratcheting. This paper investigates the solutions to calculate growth strain (incremental plastic strain) and their application in design including taking into account the interaction with fatigue. Finite element analysis is presented to validate the analytical solutions.

Commentary by Dr. Valentin Fuster

Codes and Standards: Fatigue Monitoring and Related Assessment Methods

2018;():V01AT01A033. doi:10.1115/PVP2018-84007.

During design of Class I components in Nuclear Power Plants, cumulative usage factors (CUF) are conservatively calculated to estimate fatigue damage, and results must be below the limits of the applicable codes. Nevertheless, when these results are used to evaluate the possibility of using these components for an extended life, the results are frequently above code limits. Many Nuclear Power Plants have installed commercial fatigue monitoring systems at critical components in order to assess transient severity and cycle count for life extension fatigue calculations among other reasons. Since the commissioning of the system, unexpected operation modes and thermal stratification was discovered and evaluations needed to be done. Findings, interpretations and solving are presented in this paper.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A034. doi:10.1115/PVP2018-84597.

The present paper combines the fatigue properties rapid assessment approach using uniaxial test specimens proposed by Risitano and co-workers with the nondestructive testing (NDT) inspection approach proposed by Sakagami and co-workers to monitor the onset of fatigue in a reduced scale pipeline test specimen that was previously dented and subsequently subjected to cyclic pressure loading. In addition to the use of the conventional infrared (IR) thermographic method, the present paper uses a self-reference lock-in IR thermography method based on Thermoelastic Stress Analysis (TSA) and its deviation from traditional applications due to the presence of fatigue damage and plastic strains. The paper concludes showing that is possible to predict and monitor and detect fatigue initiation and damage using IR and TSA techniques applied to the thin wall pipe loaded under cyclic hydrostatic pressure.

Topics: Fatigue , Thermography , Pipes
Commentary by Dr. Valentin Fuster
2018;():V01AT01A035. doi:10.1115/PVP2018-84979.

This work analyzes the applicability of the ASME Boiler and Pressure Vessel Code procedures to calculate fatigue crack initiation under multiaxial stresses and/or strains, in particular when caused by non-proportional loads that lead the principal directions at the critical point to vary with time, e.g. due to out-of-phase bending and torsion loads induced by independent sources. Classic uniaxial fatigue damage models are usually inappropriate for analyzing multiaxial loads, since they can generate highly inaccurate predictions. Moreover, it is shown that the ASME procedures can lead to non-conservative results for non-proportional load histories.

Commentary by Dr. Valentin Fuster

Codes and Standards: Hydrogen Effects on Material Behavior for Structural Integrity Assessment

2018;():V01AT01A036. doi:10.1115/PVP2018-84099.

As is well known, low alloy steels are widely used as materials for high pressure vessels because of their high tensile strength and reasonable price, but also show severe hydrogen embrittlement. Therefore, in 2016, the authors introduced a scenario for the safe use of low alloy steels in highly pressurized hydrogen gas as a “Guideline” at ASME PVP 2016 [1]. Following discussions with stakeholders and experts in recent years, we published Technical Document (TD) as an industrial standard prior to regulation, on the safe use of ground storage vessels made of low alloy steels in Hydrogen Refueling Stations (HRSs) based on performance requirements. This article presents an outline of the TD describing the required types of testing as performance requirements for confirming the good hydrogen compatibility of low alloy steels, such as controlling tensile strength in an appropriate range, confirming leak-before-break, determining the life of ground storage vessels by fatigue testing and determining the inspection term by fatigue crack growth analysis using the fatigue crack growth rate in highly pressurized hydrogen.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A037. doi:10.1115/PVP2018-84178.

Hydrogen induced cracking occurs at the welded position of the structure due to concentration of hydrogen during cooling process of welding. In order to prevent the hydrogen induced cracking, Pre-Heat Treatment (PHT) is conducted. However, since PHT takes high cost, it is important to find out the suitable PHT condition based on computational mechanics. One of authors has been proposed α multiplication method which magnifies the hydrogen driving term in the diffusion equation to find out detailed behaviors of hydrogen concentration around a local stress field.

In this study, in order to clarify the effect of PHT on hydrogen diffusion and concentration behaviors, a coupled analysis of heat transfer – thermal stress – hydrogen diffusion combining with α multiplication method was conducted for the model of y-grooved weld joint under various PHT conditions.

This analytical method is as follows. At first, heat transfer analysis was conducted by finite difference method (FDM). And, temperature at each grid obtained by heat transfer analysis was interpolated to each node for thermal stress analysis by the finite element method (FEM). Then, thermal stress was calculated for each node using the interpolated temperature. After that, thermal stress obtained by this analysis was interpolated to each grid point for analysis of hydrogen diffusion by FDM. Using the interpolated thermal stress, stress driven hydrogen diffusion analysis was performed. By conducting sequentially these calculations mentioned above, hydrogen diffusion and concentration behaviors during cooling process were analyzed. The temperature of weld metal was 1500°C. And at initial state, hydrogen was introduced in weld metal. Thermal stress analysis was conducted under plane strain condition.

As a result, hydrogen diffusion and concentration behaviour at weld joint during cooling process was found to be typical at the site of maximum hydrostatic stress and to be affected not only the gradient of hydrostatic stress but also the gradient of diffusion coefficient induced by temperature distribution.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A038. doi:10.1115/PVP2018-84192.

Hydrogen induced cracking occurs at the welded position of the structure due to concentration of hydrogen during cooling process of welding. A square groove weld joint is one of typical one in engineering field. Hydrogen embrittlement cracking is sometimes caused during cooling process of a weld joint. For such case, hydrogen diffusion and concentration behaviour is a significant factor. One of authors has been proposed α multiplication method which magnifies the hydrogen driving term in the diffusion equation to find out detailed behaviours of hydrogen concentration around a local stress field. In this paper, to clarify hydrogen diffusion behaviour in the square groove weld joint, a coupled analysis of heat transfer – thermal stress – hydrogen diffusion combining with α multiplication method was conducted. From these results, it was found out that for the case of a square groove weld joint, since thermal stress was not highly localized for the case of using usual value of thermal expansion coefficient of steel, hydrogen concentration behaviour is not typical. However, if thermal stress is highly localized, hydrogen was found to be localized in the side of HAZ (heat affected zone) at the interface of WM (weld metal) and HAZ and is much more typical near the outer surface side of weld joint. Hydrogen diffusion and concentration behaviours were also found to be dominated not only by local thermal stress gradient, ∇σ but also by diffusion coefficient gradient, ∇D caused by temperature difference during cooling process. In this paper, effects of these factors on hydrogen concentration were investigated based on a coupled analysis of heat transfer – thermal stress – hydrogen diffusion combining with α multiplication method.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A039. doi:10.1115/PVP2018-84486.

2.25Cr-1Mo-0.25V steel, which is a high strength low alloy (HSLA) steel, has been widely used for structural material of hydrogenation reactor due to its excellent combination of mechanical properties and resistance to hydrogen embrittlement (HE). However, it still suffers serious hydrogen damage during the aggressive service environment. When sufficient hydrogen concentrates in the reactor steel, the ductility and strength of the steel will be greatly decreased. Such a phenomenon of reduction of toughness is known as HE, and it can significantly weaken the safety and reliability of equipment. Therefore, the aim of this investigation is to focus on the effect of hydrogen on fracture toughness behavior of 2.25Cr-1Mo-0.25V steel. The fracture mechanics specimens in geometry of single edge notch bending were used. The immersion charging method was used to pre-charge hydrogen inside the specimens. Moreover, the fracture toughness of specimens with and without hydrogen pre-charging were measured following the ASTM E1820 standard. Finally, the fracture morphology was observed by scanning electron microscopy (SEM) to identify the HE mechanisms. The results of the present investigation showed that the pre-charged hydrogen resulted in significant reduction of fracture toughness of 2.25Cr-1Mo-0.25V steel, indicating a reduced crack growth resistance of specimens in the presence of hydrogen. Furthermore, the uncharged specimens failed in a ductile manner, whereas the fracture of pre-charged specimens is a mixed ductile and brittle fracture mode. It was believed that the hydrogen-induced decohesion (HEDE) mechanism contributed to the HE in hydrogen pre-charged specimens.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A040. doi:10.1115/PVP2018-84723.

Archival materials test data on austenitic stainless steels for service in high pressure hydrogen gas has been reviewed. The bulk of the data were from tests conducted prior to 1983 at the Savannah River Laboratory, the predecessor to the Savannah River National Laboratory, for pressures up to 69 MPa (10,000 psi) and at temperatures within the range from 78 to 400 K (−195 to 127 °C). The data showed several prominent effects and correlations with test conditions:

• There was a significant reduction in tensile ductility as measured by reduction of area or by the total elongation with hydrogen. Hydrogen effects were observed when the specimens were tested in the hydrogen environment, or the specimens were precharged in high pressure hydrogen and tested in air or helium.

• There was a significant reduction in fracture toughness with hydrogen (and sometimes in tearing modulus which is proportional to the slope of the crack resistance curve).

• The effects of hydrogen on ductility can be correlated to the nickel content of the iron-chromium-nickel steels. The optimal nickel content to retain the high tensile ductility in these alloys was 10 to at least 20 wt. %.

• The effects of hydrogen can be correlated to the grain size. Large grain sizes exhibited a greater loss of ductility compared to small grain sizes.

The Savannah River Laboratory test data, especially those not readily available in the open literature, along with the sources of the data, are documented in this paper.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A041. doi:10.1115/PVP2018-84726.

Cr-Mo and Ni-Cr-Mo high-strength low-alloy steels are candidate materials for the storage of high-pressure hydrogen gas. Forging materials of these steels have been used for such an environment, while there has been a strong demand for a higher performance material with high resistance to hydrogen embrittlement at lower cost. Thus, mechanical properties of Cr-Mo and Ni-Cr-Mo steels made of quenched and tempered seamless pipes in high-pressure hydrogen gas up to 105 MPa were examined in this study. The mechanical properties were deteriorated in the presence of hydrogen that appeared in reduction in local elongation, decrease in fracture toughness and accelerated fatigue-crack growth rate, although the presence of hydrogen did not affect yield and ultimate tensile strengths and made little difference to the fatigue endurance limit. It is proposed that pressure vessels for the storage of gaseous hydrogen made of these seamless line pipe steels can be designed.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A042. doi:10.1115/PVP2018-85116.

Flaws found during in-service inspection of Zr-2.5Nb pressure tubes in CANDU(1) reactors include fuel bundle scratches, debris fretting flaws, fuel bundle bearing pad fretting flaws and crevice corrosion flaws. These flaws are volumetric and blunt in nature. A key structural integrity concern with in-service blunt flaws is their susceptibility to delayed hydride cracking (DHC) initiation, particularly for debris fretting flaws under flaw-tip hydride ratcheting conditions. Hydride ratcheting conditions refer to situations when flaw-tip hydrides do not completely dissolve at normal operating temperature, and accumulation of flaw-tip hydrides occurs with each reactor heat-up/cool-down cycle. A significant number of in-service flaws are expected to be under hydride ratcheting conditions at late life of pressure tubes.

DHC initiation evaluation procedures based on process-zone methodology for flaws under hydride ratcheting conditions are provided in CSA (Canadian Standards Association) N285.8-15. The process-zone model in CSA N285.8-15 predicts whether DHC initiation occurs or not for given flaw geometry and operating conditions, regardless of the number of reactor heat-up and cool-down cycles. There has been recent new development. Specifically, a cycle-wise process-zone model has been developed as an extension to the process-zone model in CSA N285.8-15. The cycle-wise process-zone model is able to predict whether DHC initiation occurs or not during a specific reactor heat-up and cool-down cycle under applied load. The development of the cycle-wise process-zone model was driven by the need to include flaw-tip stress relaxation due to creep in evaluation of DHC initiation. The technical basis for the development of the cycle-wise process-zone model for prediction of DHC initiation under flaw-tip hydride ratcheting conditions is described in this paper.

Commentary by Dr. Valentin Fuster

Codes and Standards: Hydrogen Flakes Assessment in the RPVs

2018;():V01AT01A043. doi:10.1115/PVP2018-84087.

Following the flaw indications found in summer 2012 in two Belgian Reactors Pressure Vessels (RPV), WENRA recommended [1] the nuclear safety authorities in Europe to verify the material quality and integrity of the RPV in a 2-step approach: 1) a comprehensive review of the manufacturing and inspection records of the forgings of the RPV, 2) an additional UT examination of the base material of the vessels if needed.

In this context, and to consolidate scientific basis on this issue, IRSN, the French technical safety organization, conducted, with CEA support, a test program aiming at studying the consequences of hydrogen flakes in large forgings of primary equipment (RPV, steam generator, pressurizer).

Framatome provided the material to be investigated, namely two blocks of a steam generator vessel shell in 18MND5 steel: a block without flake — the reference block — and a block including a high density of hydrogen flakes. This shell — so called VB395 — was rejected because of an incident which occurred during the degassing heat treatment.

Fracture toughness has been evaluated from 85 tests in the ductile range and the ductile-to-brittle transition range of the material. The test results on usual 0.5T-CT specimens were compared to those on specimens containing a hydrogen flake replacing the fatigue precrack. The latter were interpreted using 3D elastic-plastic X-FEM simulations allowing the modelling of the irregular flake geometry.

Furthermore, large scale bending specimens with multiple flakes have been tested at −100°C. These tests were interpreted thanks to 3D X-FEM simulations allowing the analysis of the hydrogen flake interaction in terms of KJ.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A044. doi:10.1115/PVP2018-84155.

During the 2012 outage of the Belgian nuclear power plants (NPP) Doel 3 and Tihange 2 non-destructive testing (NDT) measurements revealed a high quantity of indications in the upper and lower core shells of the reactor pressure vessels (RPV). A root cause analysis leads to the most likely hypothesis that the indications are hydrogen flakes in segregated zones of the RPV ferritic base material. The laminar and quasi-laminar orientation (0° – 15° inclination to the pressure retaining surface) of the hydrogen flakes, the interaction of several adjacent flakes and the mechanical loading conditions lead to a mixed-mode behavior at the crack tips.

In the framework of an ongoing research project, experimental and numerical investigations are conducted with the aim to describe the failure behavior of such complex crack configurations. The experiments are carried out using two ferritic materials. One is a non-irradiated representative RPV steel (SA 508 Class 2) and the second material is a special lower bound melt of a modified 22NiMoCr3-7 steel (FKS test melt KS 07 C) containing hydrogen flakes. A material characterization is done for both materials including tensile specimens, notched round bars, shear-, torsion- and compact-tension-shear (CTS) - specimens to investigate different stress states. Furthermore, flat tensile specimens with eroded artificial crack fields are used to investigate the interaction between the cracks in different arranged crack fields. Numerical simulations are carried out with extended micromechanical based damage mechanics models. For the description of ductile failure an enhanced Rousselier model is used and an enhanced Beremin model to calculate the probability of cleavage fracture. To account the sensitivity for low stress triaxiality damage by shear loading, the Rousselier model was enhanced with a term to account for damage evolution by shear. The Beremin model will be enhanced with a term to account for different levels of triaxiality. For the numerical simulations in the transition region of ductile-to-brittle failure a coupled damage mechanics model (enhanced Rousselier and Beremin) will be used. In this paper, the current status of the ongoing research project and first results are presented.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A045. doi:10.1115/PVP2018-84575.

In fracture mechanics, a flaw behavior in pressure vessels is assessed with respect to the material fracture toughness.

Fracture toughness which most Fitness-for-Service (FFS) codes relies on, only considers mode-I crack opening. However, in presence of tilted flaws, like quasi-laminar hydrogen flakes, this mode-I toughness may be too severe, and a mixed mode I+II fracture toughness seems to be more appropriate.

In order to address the assessment of the fracture toughness curve, mixed mode I+II tests were performed by the authors on ferritic steel samples by adjusting the standard mode I CT specimen geometry to a geometry subjected to mixed mode I+II. Then, XFEM simulations of the mixed mode tests were performed in order to calculate the J-integral along the crack front.

Based on tests and calculations results, the paper explains how the authors work towards proposing a method to measure the material fracture toughness in case of flaws subjected to mixed mode (I+II) loading.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A046. doi:10.1115/PVP2018-84688.

Non-destructive testing measurements in the Belgian nuclear power plants Doel 3 and Tihange 2 revealed a high quantity of indications in the upper and lower core shells of the reactor pressure vessels. The most likely explanation is that the indications are hydrogen flakes positioned in segregated zones of the base material of the pressure vessel. These hydrogen flakes have a laminar and quasi-laminar orientation to the pressure retaining surface. Under mechanical loading the crack tips undergo predominantly mixed mode loading conditions, where the induced stress and strain fields of the single crack tips influence each other. Due to these specific loading conditions, the assumptions for classical standardized fracture mechanical methods are not met. Currently, there is no verified concept for the evaluation of such kind of crack fields.

Therefore the mechanical behavior of components with laminar crack fields and the interaction of cracks in such crack fields are investigated in an ongoing research project. Relevant parameters to describe crack fields in terms of crack size, crack location and crack orientation are derived from literature and own nondestructive measurements. Damage mechanical approaches are used in finite element calculations to investigate the interaction of cracks. Advanced damage mechanical models will be used to investigate crack initiation, crack growth and coalescence of cracks in crack fields. According to the results, representative parameters for crack fields will be derived and critical crack formations determined. The results will be evaluated and compared with state of the art approaches and standards.

Commentary by Dr. Valentin Fuster

Codes and Standards: Improvement of Flaw Characterization Rules in Fitness-for-Service Codes

2018;():V01AT01A047. doi:10.1115/PVP2018-84019.

When discrete multiple flaws are in the same plane, and they are close to each other, it can be determined whether they are combined or standalone in accordance with combination rules provided by fitness-for-service (FFS) codes, such as ASME, JSME, BS7910, FKM, WES2805, etc. However, specific criteria of the rules are different amongst these FFS codes.

On the other hand, plastic collapse bending stresses for stainless steel pipes with circumferential twin flaws were obtained by experiments and the prediction procedure for collapse stresses for pipes with twin flaws were developed analytically. Using the experimental data and the analytical procedure, plastic collapse stresses for pipes with twin flaws are compared with the stresses in compliance with the combination criteria. It is shown that the calculated plastic collapse stresses based on the combination criteria are significantly different from the experimental and analytical stresses.

Topics: Stress , Pipes , Collapse
Commentary by Dr. Valentin Fuster
2018;():V01AT01A048. doi:10.1115/PVP2018-84120.

When multiple flaws are detected in pressure retaining components during inspection, the first step of evaluation consists of determining whether the flaws shall be combined into a single flaw or evaluated separately. This combination process is carried out in compliance with proximity rules given in the Fitness-for-Service (FFS) Codes. However, the specific criteria for the rules on combining multiple flaws into a single flaw are different among the FFS Codes.

In this context, revised and improved criteria have been developed, to more accurately characterize the interaction between multiple subsurface flaws in operating pressure vessels. This improved approach removes some of the conservatism in the existing ASME Code approach, which was developed in the 1970s based on two flaws interacting with each other.

This paper explains in detail the methodology used to derive improved flaw proximity rules through three-dimensional FEM and XFEM analyses. After the presentation of the calculations results and the improved criteria, the paper also highlights the multiple conservatisms of the methodology using several sensitivity analyses.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A049. doi:10.1115/PVP2018-84703.

After a short review of the 3 Codes in term of flaw evaluation, this paper will consider the Failure Assessment Diagrams (FAD) proposed in each of them.

The cracked components are evaluated by a dedicated diagram for margin evaluation of ductile tearing resistance of the components: the elastic stress intensity factor of the crack normalized by the toughness of the material on one axis and the applied stresses normalized by a Reference Stress in the other axis.

The 2017 Edition of RSE-M Appendix 5.4 and 5.6, the 2017 Edition of ASME XI Appendix H and the 2016 Edition of API 579 Part 9 will be used in this first comparison.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A050. doi:10.1115/PVP2018-84822.

In this paper, the combination rule for circumferential multiple-cracked pipe assessment is investigated using finite element damage analysis. The FE damage analysis based on the stress-modified fracture strain model is validated against limited fracture test data of two circumferential surface cracked pipes. Then systematic parametric study is performed using FE damage analysis for symmetrical surface cracked pipes. Failure bending stresses are calculated using the combination rule and the net-section collapse load approach for single crack provided in ASME BPV Code. It is found that predicted failure bending stress using the combination rule might be non-conservative when the distance between two cracks is short. To overcome the problem, a new combination criterion based on crack dimensions is proposed and compared with numerical data.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A051. doi:10.1115/PVP2018-84838.

In this study, the effect of longitudinal distance H between non-aligned twin cracks is investigated using finite element damage analysis. The FE damage analysis based on the stress-modified fracture strain model is used to calculate the failure stress of non-aligned twin cracked pipe. Parametric study on the axial distance H between non-aligned twin cracks with various crack depths and lengths were conducted and compared with predictions using the alignment rules and the net-section collapse load approach for single crack provided in ASME Code. It is shown that the trend of the predicted collapse bending stresses for the non-aligned twin cracked pipes using FE damage analysis are different from the ones using the alignment rule.

Topics: Stress , Pipes , Failure
Commentary by Dr. Valentin Fuster
2018;():V01AT01A052. doi:10.1115/PVP2018-84951.

Bending stresses at incipient plastic collapse for pipes with circumferential surface flaws are predicted by net-section stress approach. Appendix C-5320 of ASME B&PV Code Section XI provides a formula of bending stress at the plastic collapse, where the formula is applicable for both inner and outer surface flaws. That is, the collapse stresses for pipes with inner and outer surface flaws are the same, because of the pipe mean radius at the flawed section being entirely the same.

Authors considered the separated pipe mean radii at the flawed ligament and at the un-flawed ligament. Based on the balances of axial force and bending moment, formulas of plastic collapse stresses for each inner and outer flawed pipe were obtained. It is found that, when the flaw angle and depth are the same, the collapse stress for inner flawed pipe is slightly higher than that calculated by Appendix C-5320 formula, and the collapse stress for outer flawed pipe is slightly lower than that by Appendix C-5320 formula, as can be expected. The collapse stresses derived from the three formulas are almost the same in most instances. For less common case where the flaw angle and depth are very large for thick wall pipes, the differences amongst the three collapse stresses become large.

Topics: Stress , Pipes , Collapse
Commentary by Dr. Valentin Fuster
2018;():V01AT01A053. doi:10.1115/PVP2018-84960.

When multiple surface flaws are detected in pressure components, their potential interaction is to be assessed to determine whether they must be combined or evaluated independently of each other. This assessment is performed through the flaw characterization rules of Fitness-For-Service (FFS) Codes. However, the specific combination criteria of surface flaws are different among the FFS Codes. Most of the time, they consist of simple criteria based on distance between flaws and flaw depth. This paper aims at proposing alternative characterization rules reflecting the actual level of interaction between surface planar flaws. This interaction depends on several parameters such as the relative position of flaws, the flaw sizes and their aspect ratio. Thanks to numerous three-dimensional XFEM simulations, best suited combination criteria for surface planar flaws are derived by considering the combined influence of these parameters.

Commentary by Dr. Valentin Fuster

Codes and Standards: Integrity of Cast Stainless Steel Pipe

2018;():V01AT01A054. doi:10.1115/PVP2018-84601.

In the framework of a pressurized water reactor primary loop replacement, elbows of different types were produced in cast austenitic stainless steel grade Z3CN 20-09 M. For that type of component, acceptance tests to check the sufficient mechanical properties include room and hot temperature tensile tests, following the RCC-M CMS – 1040 and EN 10002 specifications. A large test campaign on standard 10mm diameter specimens was performed and exhibited a high scattering in yield stress and ultimate tensile strength values. As a consequence, some acceptance tensile tests failed to meet the required minimal values, especially the ultimate tensile strength. Optical and electronic microscopy revealed that the low values were due to the presence of very large grain compared to the specimen gage diameter. However, tensile tests strongly rely on the hypothesis that the specimen gage part can be considered as a representative volume element containing a number of grains large enough so that their variation in size and orientation gives a homogeneous response. To confirm the origin of the scattering, a huge experimental tensile test campaign with specimens of different diameters was conducted. In parallel, FE calculations were also performed. From all those results, it was concluded that it was necessary to improve the RCC-M code for that type of test for cast stainless steel: to do so, a modification sheet was sent and is being investigated by AFCEN.

Topics: Stainless steel
Commentary by Dr. Valentin Fuster
2018;():V01AT01A055. doi:10.1115/PVP2018-85015.

In order to determine the target flaw depths for performance demonstration (PD) of nondestructive testing of cast austenitic stainless steel (CASS) pipes, probabilistic fracture mechanics (PFM) code, “PREFACE”, was developed to evaluate ductile fracture and plastic collapse considering change in mechanical properties due to thermal aging and uncertainties of the mechanical properties. In the previous study[1], the tables of the target flaw depths for the highest ferrite content material of the Japanese CASS pipes at the fully saturated thermal aging condition were calculated for several flaw lengths and several stress levels by the PREFACE code. However, the sensitivity study of the ferrite content on the target flaw depth revealed that the fully saturated thermal aging condition may not be the most severe condition for the target flaw depth. In this study, the PREFACE code was modified to improve the dependency of ferrite content on true stress-strain curve and to apply limit load method directly at the failure mode of plastic collapse. To confirm of the correct relation between ferrite content and aging time, deterministic sensitivity analyses were performed. For validation of the PFM function, the results of the deterministic analysis and PFM analysis were compared.

Topics: Pipes
Commentary by Dr. Valentin Fuster

Codes and Standards: Integrity of Reactor Pressure Vessels and Internals for Codes

2018;():V01AT01A056. doi:10.1115/PVP2018-84140.

Reactor internals are components that are no typical pressure boundary but they are nevertheless very important as they hold fuel elements and all reactor control system elements and thus must ensure their safe and reliable operation during the whole reactor life under all operating and even beyond bases regimes.

In principle, reactor internals can be replaced but their weight, quantity of very high activated material and cost such possibility practically excluded.

Thus, evaluation of the reactor internals condition and prediction of their behavior during the whole or even extended lifetime is of high importance. Reactor internals are subjected to very high neutron irradiation that could initiated not only stress corrosion (irradiation assisted) cracking but also large embrittlement and changes in dimensions (swelling and creep).

VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was initiated and co-ordinated by the Czech and was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for Nuclear Power Plants (NPPs) with WWER (Water-Water-Energetic-Reactor = PWR type) type reactors, as those codes were developed only for design and manufacture and were not changed since their second edition in 1989.

VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes.

Within the last upgrading and principal extending of this VERLIFE Procedure was developed within the 3-years IAEA project (in close co-operation with another project of the 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) that started in 2009 with final approval and editing in 2013”) a part dealing with the evaluation of reactor internals lifetime was elaborated..

This IAEA VERLIFE procedure for internals has been implemented into the existing Normative Technical Documentation (NTD) ASI (Czech Association of Mechanical Engineers), Section IV – Evaluation of Residual Lifetime of Components and Piping in WWER type NPPs.

Main damaging mechanisms that should be taken into account in reactor internals and the procedure are described in detail with necessary formulae for materials of internals:

- Radiation hardening

- Radiation embrittlement

- Radiation swelling

- Radiation creep

- Swelling under stress effect

- Swelling inducing embrittlement

- Irradiated assisted stress corrosion cracking

- Transformation austenite-ferrite

and also the method for evaluation of the resistance against non-ductile failure of postulated defect.

The paper will describe these main principles and also more detailed information on the procedure for evaluation of reactor internals will be given.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A057. doi:10.1115/PVP2018-84385.

Fracture toughness from a CT specimen is used as a material constant for fracture evaluation, but it has a large constraint, which provides too conservative evaluation results. In ductile to brittle transition temperature (DBTT) region ferritic steel which is material of RPV has a large scatter and it becomes important to know the accurate scatter of an irradiated material because of less margin of RPV’s integrity after a long term operation. In this paper to establish a more precise fracture evaluation method in DBTT region for an irradiated RPV with a postulated surface flaw, fracture analysis procedures considering constraint effect, the Beremin model and damage mechanics model and a coupled model of these models were applied to the specimens with different constraints, which were 1/2TCT specimens and flat plate specimens with a semicircular flaw under tensile load. For evaluation of pure cleavage fracture of flat plate specimens, a Beremin model with plastic strain effect was applied with incorporation of plastic strain effect. Further, for ductile fracture, the local strain criterion of ASME Section VIII was applied to the specimens with different geometries and its applicability was discussed.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A058. doi:10.1115/PVP2018-84722.

For several decades the nuclear industry has used structural weld overlays (SWOL) to repair and mitigate cracking within pressurized water reactor (PWR) components such as nozzles, pipes and elbows. There are two known primary mechanisms that have led to cracking within PWR components. One source of cracking has been primary water stress corrosion cracking (PWSCC). Numerous SWOL repairs and mitigations were installed in the early 2000s to address PWSCC in components such as pressurizer nozzles. However, nearly all of the likely candidate components for SWOL repairs have now been addressed in the industry. The other cause for cracking has been by fatigue, which usually results from thermal cycling events such as leakage caused by a faulty valve close to the component. The PWR components of most concern for fatigue cracking are mainly stainless steel. Thus, ASME Section XI Code Case N-504-4 would be a likely basis for SWOL repairs of these components, although this Code Case was originally drafted to address stress corrosion cracking (SCC) in boiling water reactors (BWR). N-504-4 includes the requirements for the SWOL design and subsequent analyses to establish the design life for the overlay based on predicted crack growth after the repair.

This paper presents analysis work performed using Code Case N-504-4 to establish the design life of a SWOL repair applied to a boron injection tank (BIT) line nozzle attached to the cold leg of an operating PWR. The overlay was applied to the nozzle to address flaws found within the stainless steel base metal during inservice examination. Analyses were performed to calculate the residual stresses resulting from the original fabrication and the subsequent SWOL repair. In addition, post-SWOL operating stresses were calculated to demonstrate that the overlay does not invalidate the ASME Section III design basis for the nozzle and attached pipe. The operating and residual stresses were also used for input to a fatigue crack growth (FCG) analysis in order to establish the design life of the overlay. Lastly, the weld shrinkage from the application of overlay was evaluated for potential impact on the attached piping, restraints and valves within the BIT line. The combined analyses of the installed SWOL provide a basis for continued operation for the remaining life of the plant.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A059. doi:10.1115/PVP2018-85130.

The American Society of Mechanical Engineers (ASME) published Section XI Code Case N-648-1 [1] in order to provide alternative examinations of reactor vessel nozzle inner radii. The Code Case was created because ultrasonic examination of the inner radius regions of reactor vessels nozzles is not practical within the operating fleet and the likelihood of flaws developing within these locations is extremely low. Justification for using alternative visual examinations was provided in a paper published at the 2001 Pressure Vessel and Piping (PVP) Technology Conference [2]. This 2001 PVP paper used linear elastic fracture mechanics (LEFM) to demonstrate tolerance for flaws significantly larger than would be detected using nondestructive examination techniques.

However, the Code Case [1] and PVP paper [2] were only applicable to operating plants in the United States. Thus, there was a need to provide a similar fracture analysis considering the AP1000® design to support elimination of volumetric examinations of the nozzle inner radius regions. It was also important to consider improvements in facture mechanics techniques that have been recently published in the ASME Code. The ductile behavior of the material at operating temperatures allow for the use of elastic plastic fracture mechanics (EPFM) methods which provides significantly improved flaw tolerance results. This paper compares results from analyses using LEFM and the EPFM methods for the AP1000 reactor vessel nozzle inner radii region and demonstrates tolerance for large flaws within these regions in order to support a basis for elimination of volumetric inspection during in-service and pre-service examination for the AP1000 design.

Commentary by Dr. Valentin Fuster

Codes and Standards: International Session for Fast Reactor Design and Construction

2018;():V01AT01A060. doi:10.1115/PVP2018-84161.

RCC-MRx Code is the result of the merger of the RCC-MX 2008, developed in the context of the research reactor Jules Horowitz Reactor project, and the RCC-MR 2007, which established rules applicable to the design of components operating at high temperature and to the Vacuum Vessel of ITER.

This code has been issued in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires) in 2012, 2015, and a new edition will be published by the end of 2018.

This new edition integrates a significant evolution for the rules dedicated to the ratchetting evaluation through the so-called “efficiency diagram rule”. This rule was initially developed to present a less penalizing rule than the classical “3 Sm” rule and to analyze the interaction between creep and ratchetting in the high temperature domain.

Since the first edition of RCC-MR code, several modifications have been made to this rule to improve its representativeness for this kind of damage. These modifications were motivated and justified by numerous experimental tests.

The last significant evolution is thus the occasion to present through this article, the background of this rule, the major evolutions already incorporated in the code and also the on-going developments.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A061. doi:10.1115/PVP2018-84242.

A program for a high-temperature design analysis and defect assessment has been developed for an elevated temperature evaluation according to the RCC-MRx for Generation IV and fusion reactor systems. The program, called ‘HITEP_RCC-MRx,’ consists of three modules: ‘HITEP_RCC-DBA,’ which computerizes the design-by-analysis (DBA) for class 1 components such as the pressure vessel and heat exchangers according to RB-3200 procedures, ‘HITEP_RCC-PIPE,’ which computerizes the design-by-rule (DBR) analysis for class 1 piping according to RB-3600 procedures and ‘HITEP_RCC-A16,’ which computerizes high-temperature defect assessment according to the A16 procedures. It is a web-based program, and thus can operate on a smartphone as well as on a personal computer once it is connected to the URL. The program has been verified with a number of relevant example problems on DBA, Pipe, and A16. It was shown from the verification works that HITEP_RCC-MRx with the three modules conducts a design evaluation and a defect assessment in an efficient and reliable way.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A062. doi:10.1115/PVP2018-84706.

The 2018 edition of the RCC-MRx Code (1) will be issued, by the end of the 2018, in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires). This Code set up rules applicable to research reactor components (coming from the RCC-MX 2008 developed in the context of the Jules Horowitz Reactor project), to reactor components operating at high temperature (coming from the RCC-MR 2007) and to the Vacuum Vessel of ITER (also coming from the RCC-MR 2007).

This edition takes the benefits of an important feedback of the users, such as Jules Horowitz Reactor or ASTRID project, but also from ITER and MYRRHA projects.

In parallel, in compliance with the EC’s objectives and its own policy of openness, AFCEN proposes to make its codes evolve, taking into account the needs and expectations of European stakeholders (operators, designers, constructors, suppliers ...) through a workshop called CEN Workshop 64 phase 2. The end of the workshop, planned for 2018, will allow to integrate recommendations issued from this work in the code.

This paper gives an overview of the performed work and also identifies the work to be done for a development of a standard such as RCC-MRx code.

Commentary by Dr. Valentin Fuster

Codes and Standards: Master Curve Fracture Toughness and Other Small Specimen Mechanical Properties (Joint With MF-12)

2018;():V01AT01A063. doi:10.1115/PVP2018-84066.

In this paper, different techniques to test notched Small Punch (SPT) samples for the estimation of the fracture properties in aggressive environments are studied, based on the comparison of the micromechanisms at different rates.

Pre-embrittled samples subsequently tested in air at conventional rates (0.01 and 0.002 mm/s) are compared to embrittled ones tested in environment at the same rates (0.01 and 0.002 mm/s) and at a very slow rate (5E−5 mm/s); a set of samples tested in environment under static loads that produce very slow rates complete the experimental results. To close the study, numerical simulations based on obtaining a punch rate that produces an equivalent CTOD growing rate in the edge of the notch to the one at the crack tip of a C(T) specimen for a given solicitation rate is carried out.

As a conclusion, is recommended to test SPT notched specimens in environment at very slow rates, of arround E−6 mm/s, when characterizing in Hydrogen Embrittlement (HE) scenarios, in order to allow the interaction material-environment to govern the process.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A064. doi:10.1115/PVP2018-84142.

Small punch test specimens are widely used for a long time as they are simple to produce and requires only a small volume of material. This fact is advantageous especially for high activity materials but also for assessment of operational damage in components materials when component integrity and strength may not be affected. In the same time, no test standard exists and several different specimen types and test procedures have been developed in different place.

Thus, to unify this activity, considerable attention has been paid since 2012 to the standardization of small punch test technique within the American Society of Testing and Materials (ASTM). In 2016 a large InterLaboratory Study has been launched within the ASTM subcommittee E10.02 - Behavior and Use of Nuclear Structural Materials, involving 12 laboratories and 6 evaluated structural materials from the nuclear and non-nuclear power plant components.

Paper describes the current status of ASTM standardization, results of the InterLaboratory Study, first analysis of the results with respect to some important test parameters, lessons learned and open questions remaining to be solved for the successful completion of the standardization process.

Topics: Testing
Commentary by Dr. Valentin Fuster
2018;():V01AT01A065. doi:10.1115/PVP2018-84250.

The extension of the operation period of nuclear plants requires an accurate characterisation of the vessel materials, in order to monitor their embrittlement due to neutron irradiation. This need poses a challenge, since the availability of specimens inside the vessels to characterise their evolution is rather scarce. Therefore, innovative techniques have to be applied, in order to reduce the number of tests and the volume of the specimens. In this paper, the Master Curve approach has been employed, combined with the use of small punch notched specimens. The Master Curve methodology allows to evaluate the embrittlement of the material using a single parameter: the reference temperature, T0. This parameter has been estimated for several steels by means of modified small punch specimens, which are characterised by their reduced dimensions: only 10 × 10 × 0.5 mm. The obtained results have been compared with those obtained by means of conventional testing and a methodology to estimate T0 by means of small punch tests together with the Master Curve has been proposed.

Topics: Temperature , Testing
Commentary by Dr. Valentin Fuster
2018;():V01AT01A066. doi:10.1115/PVP2018-84297.

The standard Master Curve (MC) deals only with materials assumed to be homogeneous, but MC analysis methods for inhomogeneous materials have also been developed. Especially the bi-modal and multi-modal analysis methods are becoming more and more standard. Their drawback is that these methods are generally reliable only with sufficiently large data sets (number of valid tests, r ≥ 15–20). Here, the possibility of using the multi-modal analysis method with smaller data sets is assessed, and a new procedure to conservatively account for possible inhomogeneities is proposed.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A067. doi:10.1115/PVP2018-84535.

An evaluation of the fracture toughness of the heat-affected zone (HAZ), which is located under the weld overlay cladding of a reactor pressure vessel (RPV), was performed. Considering inhomogeneous microstructures of the HAZ, 0.4T-C(T) specimens were manufactured from the cladding strips locations, and Mini-C(T) specimens were fabricated from the distanced location as well as under the cladding. The reference temperature (To) of specimens that were aligned with the middle section of a cladding strip (HAZMCS) was ∼12°C higher than that of specimens that were aligned with cladding strips at the overlap (HAZOCS). To values of partial area in the HAZ were obtained using Mini-C(T) specimen. The To values obtained near the side of the cladding were ∼13°C higher than those away from the cladding. To values of HAZ for both 0.4T-C(T) and Mini-C(T) specimens were significantly lower than that of the base metal at a quarter thickness by 40°C–60°C. Compared to the literature data that indicated fracture toughness at the surface without overlay cladding and base metal of a quarter thickness in a pressure vessel plate, this study concluded that the welding thermal history showed no significant effect on the fracture toughness of the inner surface of RPV steel.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A068. doi:10.1115/PVP2018-84690.

The use of miniature compact tension (mini-CT) specimens for fracture mechanics was experimentally demonstrated to allow the characterization of ferritic steels in the transition regime. In particular, the master curve transition temperature T0 can confidently be determined according to the ASTM E1921 standard using mini-CT specimens. This means that specimen size effect is well taken into account if loss of constraint is limited by restricting the test temperature range to remain below the allowed maximum loading level. In the upper shelf ductile regime, where stable crack growth occurs, a number of challenges should be overcome to use such a geometry to derive the crack resistance curve, or JR-curve, transferrable to a structure. Indeed, despite a large scatter, the experimental data on several materials suggest a size effect that underestimates the crack resistance when reducing specimen size.

The crack resistance behavior of several reactor pressure vessel materials was investigated with square-sized ligament compact tension specimens of various size ranging from 1 inch-thickness (B = 25 mm) to the smallest thickness (B = 4.2 mm) of the mini-CT. The main objective of this paper is to estimate the crack resistance behavior of RPV steels that would be obtained with a standard 1T-CT specimen by using mini-CT with the appropriate specimen size correction. After a series of scaling attempts that were not successful, based on a simple dimensional analysis, a simple analytical formulation based on specimen thickness and ligament is suggested to account for specimen size effect for the CT geometry. Reasonable agreement could generally be found on a number of RPV materials between crack resistance measured with mini-CT and standard 1T-CT specimens.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A069. doi:10.1115/PVP2018-84749.

The original WWER-440 surveillance had 6 sets of specimens and each set had 12 Charpy, 12 COD (crack opening displacement) and 6 tensile specimens made from base material, weldment and HAZ (heat affected zone). The Charpy size precrack TPB (three point bend) COD specimens were located at the end of the chains, where the flux is rapidly decreasing.

During the period of 1970–90, when the WWER-440-V213 units were designed, built and started to operate, the Charpy impact transition curve measurement was the accepted method to evaluate the radiation embrittlement. The technology and the standards to use small size fracture mechanical specimens in surveillance capsules were not developed at the time period when most of the second generation reactors — including the WWER-440 V 213 type — were designed, therefore the fracture toughness specimens were considered less interesting for the utilities and the safety authorities. Fracture toughness curves were elaborated in the laboratories on large size unirradiated specimens and radiation embrittlement adjustments were made according to the Charpy shift. However, during the past 30 years fracture mechanics has rapidly developed, and the testing moved to the direction of using small and mini sized specimens. The development of the Master Curve evaluation method [4,5] allowed the use of small specimens for fracture toughness testing in surveillance programs, and the results obtained on irradiated specimens may be used directly in the lifetime evaluation. The purpose of this work was to develop a specimen production technology and testing procedure to measure these data using the remnants of irradiated surveillance Charpy specimens, and the comparison of the data calculated from CMOD and LLD on irradiated CrMoV type RPV material and weldment.

Topics: Testing , Surveillance
Commentary by Dr. Valentin Fuster
2018;():V01AT01A070. doi:10.1115/PVP2018-84804.

Mini-CT specimens are becoming a highly popular geometry for use in reactor pressure vessel (RPV) community for direct measurement of fracture toughness in the transition region using the Master Curve methodology. In the present study, Mini-CT specimens were machined from previously tested Charpy specimens of the Midland low upper-shelf Linde 80 weld in both, unirradiated and irradiated conditions. The irradiated specimens have been characterized as part of a joint ORNL-EPRI-CRIEPI collaborative program. The Linde 80 weld was selected because it has been extensively characterized in the irradiated condition by conventional specimens, and because of the need to validate application of Mini-CT specimens for low upper-shelf materials — a more likely case for some irradiated materials of older generation RPVs. It is shown that the fracture toughness reference temperatures, To, derived from these Mini-CT specimens are in good agreement with To values previously recorded for this material in the unirradiated and irradiated conditions. However, this study indicates that in real practice it is highly advisable to use a much larger number of specimens than the minimum number prescribed in ASTM E1921.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A071. doi:10.1115/PVP2018-84889.

The load and temperature history during pressurized thermal shock (PTS) event is highly depending on the crack edge location in wall thickness direction of a reactor pressure vessel (RPV) beltline region. Therefore, the consideration of plant specific through-wall fracture toughness distribution, which is not considered in the current codes and regulations [1,2], may improve the structural integrity assessment for PTS event.

The Master Curve (MC) method [3,4] is one of the methods, which can directory evaluate the fracture toughness of ferritic materials with relatively low number of any size of specimens. CRIEPI has proposed the use of very small C(T) (Mini-C(T)) specimens for the MC method. The appropriateness of Mini-C(T) technology has been demonstrated through a series of researches and round robin activities [5, 6, 7, 8, 9].

The present study evaluated the through-wall fracture toughness distribution of irradiated IAEA reference material (JRQ) by means of combination of MC method and Mini-C(T) specimens. Four thickness locations between inner surface to 1/4-T was selected. Those four layers were separately subjected to the Mini-C(T) MC evaluation in two different laboratories. Both laboratories could separately obtain valid and consistent reference temperature, To, from all the tested layers. Inner most layer exhibits 80 °C lower To compared to the 1/4-T location even though the layer has the highest fluence of 5.38 × 1019 n/cm2, while that in 1/4-T location is 2.54 × 1019 n/cm2. The results demonstrate that initial toughness distribution is dominant in the general trend of fracture toughness distribution even after the material was highly irradiated.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A072. doi:10.1115/PVP2018-84906.

The Master Curve (MC) method can be used to directly determine fracture toughness of ferritic reactor pressure vessel (RPV) materials. CRIEPI has been working on the development of a testing technique to apply very small C(T) (called Mini-C(T)) specimens for the MC method. The appropriateness of using Mini-C(T) specimens for several materials including un-irradiated plate, forging, weld metal and irradiated plate has been demonstrated. Through a series of investigations, it was determined that more invalid data, due to ductile crack growth (DCG), can occur when using small size specimens.

Linde-80 weld metal, used in the fabrication of some RPVs, is known as low upper shelf material, which tends to exhibit more DCG than high upper shelf materials. In the present study, two sets of 15 Mini-C(T) specimens were machined and pre-cracked from irradiated Linde-80 weld metal. Each set of specimens were provided to two different laboratories (A and B). The laboratories separately conducted the MC tests. DCG occurred even in the lower test temperature condition. About half of specimens for lab A showed excessive DCG and were subjected to the censoring. Some of specimens were rejected since the test temperature is outside of the specified range (T-To < −50°C). As a result, lab A could not obtain valid To with 15 specimens. Lab B also experienced DCG, however were able to obtain a sufficient number of valid KJc data points to determine a valid To. The obtained ToQ (lab A) and To (lab B) are sufficiently close to each other and suggests that Mini-C(T) can be used even for the low upper shelf material if the number of available specimens are sufficiently large. The combined dataset from labs A and B estimated To = 31.5°C, which is in the scatter band of To obtained by pre-cracked Chapry (PCCv), 0.5TC(T) or 1TC(T) specimens in a past Heavy-Section Steel Irradiation (HSSI) project. The overall result suggests that To can be estimated using Mini-C(T) specimens for the lower upper shelf weld material, but 15 is a marginal number of specimens for a valid estimation.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A073. doi:10.1115/PVP2018-84950.

An irradiated low-upper-shelf Linde 80 weld metal has been tested by four laboratories as part of an inter-laboratory assessment of use of the miniature compact tension [mini-C(T)] test specimen for Master Curve fracture toughness evaluation following ASTM E1921. The preliminary results from each of the laboratories have been compiled and evaluated together to assess the validity and use of the mini-C(T) specimen for an irradiated reactor pressure vessel material which can exhibit ductile crack growth at low temperatures relative to cleavage initiation fracture toughness. The preliminary results from this mini-C(T) testing can also be compared to extensive specimen test results from larger C(T) specimens of the same irradiated material. Comparisons of the results from each of the laboratories and some inter-laboratory differences in the fracture testing are assessed. The evaluations indicate reasonable agreement between the mini-C(T) and larger specimen results, but the selection of test temperature and the number of test specimens needed to obtain reliable results are more difficult when testing a low-upper-shelf toughness material.

Topics: Metals , Testing
Commentary by Dr. Valentin Fuster
2018;():V01AT01A074. doi:10.1115/PVP2018-84967.

The Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals .

In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor.

The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.

Commentary by Dr. Valentin Fuster
2018;():V01AT01A075. doi:10.1115/PVP2018-84994.

Miniature compact tension (Mini-C(T)) specimen can be an effective tool by utilizing together with Master Curve (MC) methodology for fracture toughness evaluation of irradiated reactor pressure vessel (RPV) steels. Recently, Mini-C(T) specimen has been incorporated into the Japanese standard test method related to MC methodology, JEAC4216-2015 and several studies were found focusing on applicability of Mini-C(T) specimen to irradiated RPV materials. However, there exist some other issues to be resolved considering application to irradiated materials. One of them is violation against the limitation criteria for ductile crack growth (DCG) specified in the standards. In general, upper shelf energy (USE) of RPV materials tends to decrease as well as shift in Charpy transition temperature due to neutron irradiation embrittlement. It may cause reduction in resistance of material against DCG and this leads to the problem peculiar to low USE materials such that the limitation for DCG might be dominant rather than that for KJclimit. Therefore, it is possible to fail to obtain valid KJc data even within valid temperature range of MC methodology, i.e. −50°C ≤ T-To ≤ 50°C, for low USE materials using Mini-C(T) specimens.

In this study, the RPV steel with USE lower than 68J was made simulating reduction of USE due to neutron irradiation. Fracture toughness tests were performed using Mini-C(T) specimens as well as the standard 1T-C(T) specimens. Based on the test results, the validity for DCG limitation was also evaluated for each datum by post-test observation of fracture surface. Additionally, effectiveness of added side grooves and double thickness of specimen was examined as a countermeasure for Mini-C(T) specimen.

Topics: Reactor vessels
Commentary by Dr. Valentin Fuster
2018;():V01AT01A076. doi:10.1115/PVP2018-85065.

Eurofer97 is one of leading candidates of reduced activation ferritic martensitic (RAFM) steels for first wall structural materials of early demonstration fusion power plants. During fusion plant operation, high neutron irradiation damage on first wall materials can cause irradiation embrittlement and reduce the fracture toughness of RAFM steels. Therefore, it is critical to select proper testing techniques to characterize the fracture toughness of RAFM steels with high fidelity. In this manuscript, we present the feasibility study of using pre-cracked miniature multi-notch bend bar specimens (M4CVN) with a dimension of 45mm (length) × 3.3mm (width) × 1.65mm (thickness) to characterize the transition fracture toughness of Eurofer97 steel based on the ASTM E1921 Master Curve method. The testing yielded a provisional Master Curve reference temperature ToQ of −89°C of unirradiated Eurofer97 steel heat J362A in the normalized and tempered condition. The results are within the normal scatter range of Master Curve reference temperature T0 for Eurofer97 steel, indicating suitability of applying M4CVN specimens for characterizing the transition fracture toughness of Eurofer97 steel.

Commentary by Dr. Valentin Fuster

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