ASME Conference Presenter Attendance Policy and Archival Proceedings

2016;():V01AT00A001. doi:10.1115/PVP2016-NS1A.

This online compilation of papers from the ASME 2016 Pressure Vessels and Piping Conference (PVP2016) 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presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.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

2016;():V01AT01A001. doi:10.1115/PVP2016-63598.

The continuous pipe bend behavior is well elaborated in literature. It is characterized by local ovalization of each cross section during bending which results in enhanced flexibility of it as compared to straight pipe. When pipe bend approaches some other structural elements of a piping system the end effect take place which can be described by so called long shell solution. This long solution is, in fact, a semi-membrane Vlasov’s solution when the derivative of any geometrical or force function in axial direction is much smaller than in the circumferential one [1].

Mitred bend is formed by conjunction by welding of two oblique sections of initially straight pipes. Its behavior during loading by pressure or bending moment is not evident and poorly described in standards. The goal of this paper is to give a set of general functions within a thin cylindrical shell theory which will give the opportunity to consider the mitred bend as an element of a piping system. Here we additionally introduce the so called short solution when the derivative of any parameter in axial direction is much bigger than that in circumferential one. Its main goal is to give the local behavior of stress in the vicinity of the oblique weld. Each of these two solutions satisfy by differential equations of forth order.

The complete theoretical solution for a particular mitred bend is compared with

a) existing analytical solutions and formulas;

b) numerical results obtained by FEM with distinction of the zones of influence of a long as well as short shell solution;

c) experimental data on real mitred bends given in the literature.

Topics: Stress , Pipes
Commentary by Dr. Valentin Fuster
2016;():V01AT01A002. doi:10.1115/PVP2016-63678.

The third edition of API 579-1/ASME FFS-1 Fitness-For-Service includes a new Part 14 covering fatigue assessment procedures for in-service components. This new part provides methods for estimating the time to crack initiation using strain-life approaches, which are important for low-cycle fatigue, and is written as a multi-tiered approach covering screening, current design code methods, and advanced methods that take into account the latest in technology. Cycle counting methods for both welded joint and smooth-bar fatigue methods are included for uniaxial and multiaxial loading histories. In addition, the multiaxial incremental plasticity correction procedure presented in the new Level 3 Assessment represents a significant update to the plasticity correction used previously. Part 14 of API 579-1/ASME FFS-1 is a procedural document that is written for accurate and practical implementation. WRC Bulletin 550 is the basis for Part 14 and provides the background and supporting documentation for its development. This paper summarizes the work that went into developing Part 14, and WRC 550, and provides a specific example of its implementation for a practical problem of interest using software developed by The Equity Engineering Group, Inc. In-depth comparisons, key findings, and limitations of the presented methods are included.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A003. doi:10.1115/PVP2016-63695.

A methodology is presented for Level 1 and 2 Fitness-for-Service (FFS) assessments of process and power piping subject to random vibration loading. The intent is to provide a basis for random vibration assessment based on concepts from spectral fatigue, which is simplified to the degree that a non-specialist can conduct the FFS assessment with little prior knowledge of the subject matter. The proposed Level 1 FFS assessment is based on extension and generalization of established industry screening curves for piping vibration. The measured (overall) RMS vibration level is compared against allowable values given by the curves evaluated at the average crossing frequency of the measured vibration. The proposed Level 2 FFS assessment utilizes an allowable cyclic stress. The allowable stress is a function of the target design life as well as the average crossing frequency and kurtosis of the measured vibration. Both approaches are illustrated by example.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A004. doi:10.1115/PVP2016-63756.

API 579-1/ASME FFS-1[1] Part 5 (2007 edition) contains procedures for assessing local metal loss based on failure by plastic collapse. Equation (5.10) defines “acceptable” tip radii for grooves to ensure a plastic collapse failure mode. Grooves failing the radius check must be treated as a crack-like flaws. The validity of Equation (5.10) is questionable, and it may be excessively conservative. This paper presents new rules for groove assessment based on brittle and ductile failure models. Computation of the Weibull stress using finite element analysis (FEA) was employed to determine the minimum groove radius required to eliminate the possibility of cleavage fracture. The Bao-Wierzbicki ductile failure model was used with FEA to evaluate burst pressure and to determine a new groove radius criterion defining the plastic collapse regime, allowing categorization as metal loss. Groove-like flaws categorized as neither sharp cracks nor as metal loss are evaluated using an effective toughness concept. This concept quantifies the difference in fracture response between a sharp crack and a notch with a finite tip radius. The upcoming 2016 API 579-1/ASME FFS-1 rules remove the excessive conservatism found in Part 5 of the 2007 edition and avoid the abrupt transition between crack and metal loss assessment types based on groove radius.

Topics: Brittleness , Failure
Commentary by Dr. Valentin Fuster

Codes and Standards: ASME Section XI Code Activities

2016;():V01AT01A005. doi:10.1115/PVP2016-63051.

The fatigue threshold behavior of stainless steel was assessed in high temperature air and hydrogenated deaerated water environments as a function of stress ratio (R). Fatigue threshold experiments were conducted on four different heats of type 304, 304/304L, and 308L austenitic stainless steels in 250°C air and water environments at stress ratios ranging from 0.1 to 0.8. Air and water experiments showed that operational threshold ΔK (ΔKTH) values ranged from 4.3–6.0 and 3.9–5.3 MPa√m, respectively. ΔKTH values were observed to generally decrease with increasing R which is attributable to crack closure effects. The water ΔKTH measurements were either consistent with or lower than air threshold measurements, and the potential roles of the competing effects of crack closure and hydrogen enhanced planar slip will be discussed in the context of these results. Load history effects in the form of overloads and underloads were shown to significantly impact ΔKTH measurements and these results motivated testing aimed at characterizing material property based intrinsic ΔK threshold (ΔKTH*) values. The ΔKTH* values for stainless steel fatigue crack growth in 250–288°C air and water environments are estimated to be 3 and 2 MPa√m, respectively.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A006. doi:10.1115/PVP2016-63827.

This paper will provide the bases for the requirements in the Beyond Design Basis Events (BDBE) evaluation performed in accordance with the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (BPVC), Section XI, Code Case (CC). The CC provides rules to facilitate the affected components Return to Service (RTS) after a BDBE magnitude Earthquake. The paper describes the bases for the examination of Reactor Coolant Pressure Boundary (RCPB) Structures, Systems and Components (SSCs) for determining the actual seismic loading and magnitude. The implications of the examination data based on design allowable stress values. The pipe loads determined shall be used to calculate stresses on the pumps and valves.

The paper also describes the methodology for seismic events lasting longer than 100 cycles.

The Cumulative Usage Factor (U) due to the event is calculated from individual cycles as; U = U1 + U2 + U3 + ... + Un. Aftershocks are accounted for in the methodology. Fatigue usage from the event that increases the total U to a value greater than 0.8 shall be included as high risk location(s) in the ASME BPVC, Section XI, Inservice Inspection (ISI) Program.

Topics: Design , Earthquakes
Commentary by Dr. Valentin Fuster
2016;():V01AT01A007. doi:10.1115/PVP2016-63872.

Reference curves of fatigue crack growth rates for ferritic steels in air environment are provided by the ASME Code Section XI Appendix A. The fatigue crack growth rates under negative R ratio are given as da/dN vs. Kmax, It is generally well known that the growth rates decreases with decreasing R ratios. However, the da/dN as a function of Kmax are the same curves under R = 0, −1 and −2. In addition, the da/dN increases with decreasing R ratio for R < −2.

This paper converts from da/dN vs. Kmax to da/dN vs. ΔKI, using crack closure U. It can be obtained that the growth rates da/dN as a function of ΔKI decrease with decreasing R ratio for −2 ≤ R < 0. It can be seen that the growth rate da/dN vs. ΔKI is better equation than da/dN vs. Kmax from the view point of stress ratio R. Furthermore, extending crack closure U to R = −5, it can be explained that the da/dN decreases with decreasing R ratio in the range of −5 ≤ R < 0. This tendency is consistent with the experimental data.

Topics: Steel , Fatigue cracks
Commentary by Dr. Valentin Fuster
2016;():V01AT01A008. doi:10.1115/PVP2016-63946.

ASME Code Case N-513-4 allows evaluation of flaws in elbows to be treated as straight pipe if the flaw is located within a distance √Rt from the weld centerline, provided the elbow is welded to straight pipe. In addition, N-513-4 provides guidance for evaluation of flaws in elbows when the flaw is remote from the weld, but this approach may be increasingly conservative for flaw locations closer to the weld. For the instance where a flaw evaluation is necessary at a weld adjoining a 90° elbow to another 90° elbow, it is not clear if the stress distribution would be significantly different from a weld connecting a 90° elbow to straight pipe. This paper will employ finite element models to examine the stress distribution near elbow-to-elbow welds under a moment load and compares the results to straight pipe. The results of this paper may be used as a supporting basis to modify the Code Case guidance in a future revision.

Topics: Pipes
Commentary by Dr. Valentin Fuster
2016;():V01AT01A009. doi:10.1115/PVP2016-64023.

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. Article A-3000 of Appendix A in ASME Section XI prescribes a method to calculate the stress intensity factor for a surface or subsurface flaw by making use of the flaw location stress distribution obtained in the absence of the flaw. The 2015 Edition of ASME Section XI implemented a number of significant improvements in Article A-3000, including closed-form equations for calculating stress intensity factor influence coefficients for circumferential flaws on the inside surface of cylinders. Closed-form equations for stress intensity factor influence coefficients for axial flaws on the inside surface of cylinders have also been developed. Ongoing improvement efforts for Article A-3000 include development of closed-form relations for the stress intensity factor coefficients for flaws on the outside surface of cylinders. The development of closed-form relations for stress intensity factor coefficients for axial flaws on the outside surface of cylinders is described in this paper.

Topics: Stress , Cylinders
Commentary by Dr. Valentin Fuster

Codes and Standards: Development of Stress Intensity Factors for Codes (Joint With M&F)

2016;():V01AT01A010. doi:10.1115/PVP2016-63262.

Stress intensity factor (SIF) is one of the key parameters in structural integrity assessment. Weight function method has been used in flaw acceptance assessment codes and standards, such as R6 and BS7910, to calculate SIF of a semi-elliptical (part) surface flaw. In this method, stress distribution across the section thickness is described by a polynomial equation, and SIF is estimated using geometry functions fi and stress components σi. The SIF solutions are available for both the deepest and the surface points of part surface flaw in R6 and BS7910. However, a case study from this work shows that the SIF estimation using the current methods are not always conservative when a flaw is at stress concentration, such as weld toe. This results in an optimistic limiting defect sizes and jeopardizes the safety. To account for the effect of stress concentration on SIF, one solution is to use SIF magnification factor and stress concentration factor, but this approach could be overly conservative. Although the original research used power law stress distribution in calculation of SIF, it is not clear whether the developed geometry function factors are suitable for a flaw at steep gradient stress concentration zone. The same question is for the similar SIF solutions of French RCC-MR code, as the model used to derive the SIF does not include stress concentration.

This paper briefly reviews the weight function SIF solutions and compares them with the 3D FEA results of surface flaws in plate and pipe with various dimensions and flaw sizes. The guidance is provided on how to use weight function SIF solutions for surface flaws at stress concentration region for structural integrity analysis.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A011. doi:10.1115/PVP2016-63362.

In this paper, the stress intensity factor KI for the crack front line a − ε(1 + cosmθ), which is slightly perturbed from a complete circular line with a radius of a, is determined. The method used in this study is based upon the perturbation technique developed by Rice for solving the elastic field of a crack whose front slightly deviates from some reference geometry. It is finally shown that the solution for the stress intensity factor matches the results of a three-dimensional finite element analysis.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A012. doi:10.1115/PVP2016-63424.

In some cracks attributed to primary water stress corrosion cracking, the crack depth a was greater than half-length of the crack 0.5ℓ. This paper presents details of stress intensity factor solutions for circumferential surface cracks with large aspect ratios a/ℓ in piping system subjected to global bending. The stress intensity factor solutions for semi-elliptical surface cracks were obtained by finite element analyses with quadratic hexahedron elements. Solutions at the deepest and the surface points of the cracks with various aspect ratio (0.5 ≤ a/ℓ ≤ 4.0), crack depth ratio (0.01 ≤ a/t ≤ 0.8) and pipe sizes ( 1/80 ≤ t/Ri ≤ 1/2) were investigated, where t and Ri are wall thickness and inner radius of pipe, respectively. Proposed stress intensity factor solutions for cracks with a/ℓ = 0.5 are consistent with the values reported in the previous study. The solutions developed in this study are widely applicable to various engineering problems related to crack evaluation in piping systems.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A013. doi:10.1115/PVP2016-63479.

Stress intensity factor (SIF) solutions for subsurface flaws near the free surface in plates were numerically investigated based on the finite element analyses. The flaws with aspect ratios a/ℓ = 0, 0.1, 0.2, 0.3, 0.4 and 0.5, the normalized ratios a/d = 0, 0.1, 0.2, 0.4, 0.6 and 0.8 and d/t = 0.01 and 0.1 were taken into account, where a is the half flaw depth, ℓ is the flaw length, d is the distance from the center of the subsurface flaw to the nearest free surface and t is the wall thickness. Fourth-order polynomial stress distributions in the thickness direction were considered. Based on the results, it can be concluded that the numerical SIF solutions obtained in this study are useful in engineering applications.

Commentary by Dr. Valentin Fuster

Codes and Standards: Environmental Fatigue Issues (Joint With M&F)

2016;():V01AT01A014. doi:10.1115/PVP2016-63115.

Advanced Fatigue Methodologies (AdFaM), a joint project of European research laboratories, vendors and plant operators, was launched in 2014 to build on the results from recent laboratory studies of fatigue behavior of austenitic stainless steels under NPP-relevant conditions that showed improved lifetimes compared to the best fit to test data presented in NUREG/CR-6909, and to further investigate transferability between specimen test results and the fatigue behavior of NPP components during plant operation.

In particular, AdFaM has focused on an empirical and mechanistic investigation of the effects of hold times on fatigue life. A small number of previous test results suggest an increase in fatigue life for stabilized grades of austenitic stainless steel when hold times (ranging from several hours to days) are introduced into a test between periods of strain-controlled cyclic loading. Tests incorporating hold times may be more representative of material behavior in NPPs, where temperature transients due to start-ups, shutdowns and major power changes may be separated by long periods of steady state operation.

Under AdFaM, fatigue endurance tests incorporating hold times have been completed on stabilized and non-stabilized stainless steel grades (Types 304L and 347) and the mechanisms responsible for the observed variations in fatigue life have been investigated using a range of microscopy techniques.

Results confirm a significant extension of fatigue life due to hold times in both stabilized and non-stabilized grades. Life extension appears to be linked to hold hardening observed in cyclic behavior, and this link has been investigated through microstructural characterization of fatigue specimens examined before and after holding at elevated temperature.

This project helps to improve the understanding of transferability of results from small specimen tests (without hold times) to analysis of NPP components and provides insights that will contribute towards continuing development of fatigue design curves and analysis methods in Design Codes such as ASME Code Section III and KTA 3201/3211.

The AdFaM project is now complete. The valuable results and insights gained from this work demonstrate the significant benefits of collaborative research between various industrial and academic partners in the area of fatigue of NPP materials.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A015. doi:10.1115/PVP2016-63125.

In the wake of numerous experimental tests carried out in air and also in a PWR environment, both abroad and in France, an update of the current thermal fatigue codification is underway in France. Proposals are currently being integrated in the RCC-M code [1].

In parallel, it is necessary to evaluate the impact of codification evolution on the RCS components. In the USA, such evaluations have already been implemented for license renewal to operate power plants beyond their initial 40 years of operation. In order to reduce the scope of the calculations to perform, a preliminary screening was carried out on the various areas of the primary system components: this screening is detailed in an EPRI report [2]. The output of this screening process is a list of locations that are most prone to EAF degradation process and it is on these zones only that detailed EAF calculations are carried out.

In France, a similar approach was defined in the perspective of the fourth ten-year visit of the 900 MWe plants (VD4 900 MWe) so as to map out all the locations that are most impacted by EAF and hence concentrate the calculation effort on these specific areas for the VD4 900 MWe.

In that respect, a specific methodology to evaluate the factor to account for environmental effects or Fen [3] based on correlations [4] for hot and cold shocks was established. These correlations use data that is readily accessible in transient description documents and stress reports such as temperature change, heat transfer coefficients, ramp duration and geometry. The need for these correlations is specific to the French context due to a need for a preliminary and yet precise idea of the overall impact of the modifications brought to the RCC-M code in fatigue before the VD4 900 MWe.

This paper presents the results of the screening method that was applied to the whole RCS of the 900 MWe NPP fleet.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A016. doi:10.1115/PVP2016-63127.

The French nuclear industry has to face nowadays a series of challenges it did not have to face a decade ago. The most significant one is to ensure a reliable and safe operation of Nuclear Power Plants (NPP) in a context of both an ageing reactor fleet and new builds. The new constructions need rules that integrate a strong operation feedback while the older NPPs need rules that will guarantee the life extension beyond 40 years of operation.

In this context, a new edition of the French RCC-M Code is planned for 2016. This new edition integrates the modifications made to the Code as a result of Requests for Modification (RM), which can be submitted by anyone and which help to continuously improve the quality and robustness of the Code.

Concerning fatigue analyses, the RCC-M Code steering committee has acknowledged end of 2014 the reception of two RM to modify the fatigue design curve for austenitic stainless steels and Nickel base alloys, as well as to integrate environmental effects in the fatigue evaluation for austenitic stainless steel components. The contents of these two RM were based on the proposals presented in Reference [1].

AFCEN required a technical review of these two RM and this task was performed by a working group composed by French and international experts. This process concluded to the approval of these two RM to be integrated to the 2016 edition of the RCC-M Code. This paper offers a presentation of these two new Rules in Probation Phase (RPP), this format being quite similar to Code Cases proposed by ASME Code.

Topics: Fatigue
Commentary by Dr. Valentin Fuster
2016;():V01AT01A017. doi:10.1115/PVP2016-63134.

Fatigue crack growth of austenitic stainless steels can be enhanced significantly in high temperature light water reactor coolant environments and an ASME Code Case, N-809, has recently been developed to provide fatigue crack growth rate curves for these alloys in pressurized water reactor environments. However, under some conditions, the enhanced rates can decrease to rates close to those in air at long rise times, a process referred to as retardation which is not taken account of in the Code Case. An improved understanding of the mechanisms of both enhancement and retardation would be beneficial to determining whether advantage could be taken of these retarded rates in plant assessment. A number of studies have been undertaken to evaluate fatigue crack growth behavior in both air and water environments in order to provide mechanistic insight. Progress on this work will be described. The data from air and inert environments support the proposed mechanism of environmentally enhanced fatigue by environmental enhancement of planar slip, although it is not yet possible to differentiate between the impact of oxidation and corrosion hydrogen on the level of enhancement in aqueous environments. Testing in high temperature water environments suggests that both corrosive blunting and/or oxide-induced closure mechanisms may contribute to crack growth rate retardation under specific circumstances.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A018. doi:10.1115/PVP2016-63148.

High temperature water environments, typical of light water reactor primary coolant, are known to lead to significant environmental enhancement of fatigue crack growth of austenitic stainless steels. For PWR environments. these effects have recently been codified in ASME Code Case N-809. However, just as for the detrimental effect of these environments on fatigue endurance, plant experience indicates that crack growth rates must be significantly lower than predictions based on laboratory data using simple sawtooth waveforms. In order to explain this discrepancy, a significant amount of research has been conducted to quantify factors leading to crack growth rate retardation with sulfur content having been identified as significant in promoting crack growth rate retardation. However, the inherent conservatisms in current analysis techniques may be just as significant in generating the perceived over-conservatism of environmental fatigue crack growth laws such as Code Case N-809.

The current work looks at the impact of waveform shape and spectrum loading on the level of environmental enhancement for a given stress intensity factor range and total rise time by considering simplified transients and loading spectra. The observations suggest that simplified definitions of total rise time used in fatigue assessments can lead to large over-estimation of actual fatigue damage. These data form the basis of an analytical methodology being developed by RollsRoyce (presented in a separate paper at this conference) aimed at partitioning damage across the loading cycle in order to remove over-conservatisms in current analytical methodologies.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A019. doi:10.1115/PVP2016-63149.

INCEFA-PLUS is a major new five year project supported by the European Commission HORIZON2020 program. The project commenced in mid 2015. 16 organizations 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.

Prior to the start of INCEFA-PLUS, an in-kind study was undertaken by several European organizations with the aim of developing the current state of the art for this technical area. In addition to stress/strain amplitude, this study identified three additional experimental variables which required further study in order to support improved assessment methodology for environmental fatigue, namely the effects of mean stress/strain, hold time and surface finish. Within INCEFA-PLUS, the effects of these three variables on fatigue endurance of austenitic stainless steels in light water reactor environments are therefore being studied experimentally. The data obtained will be collected and standardized in an online environmental fatigue database. A dedicated CEN workshop will deliver a harmonized data format facilitating the exchange of data within the project but also beyond.

Based on the data generated and the resulting improvement in understanding, it is planned that INCEFA-PLUS will develop and disseminate methods for including the new data into assessment procedures for environmental fatigue degradation. This will take better account of the effects of mean stress/strain, hold time and surface finish.

This paper will describe the background to the project and will explain the expectations for it.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A020. doi:10.1115/PVP2016-63161.

Light water reactor coolant environments are known to significantly reduce the fatigue life of austenitic stainless steels. However, most available data are derived from isothermal testing of membrane loaded tensile specimens, whereas the majority of plant loading transients result from thermal transients and involve significant through-wall strain gradients. This paper describes the development of a high temperature water facility to enable both thin and thick wall hollow fatigue endurance specimens to be subjected to thermal and mechanical loading for a wide range of thermal cycles including rapid shock loading.

Thermal shock loading from 300°C to between 40 and 150°C has been achieved and Finite Element Analysis, FEA, has been used to calculate the thermally induced strain profiles through a 12mm thick-wall specimen. This indicates peak surface thermal strain ranges of up to 0.8% for a transient between 300 and 40°C. Testing is underway to investigate the impact of the strain gradient and thermal waveform on the fatigue life of this specimen where significantly longer lives may be expected compared to membrane loaded specimens.

The ability within the same facility to apply simulated thermal shock profiles to both thick-wall specimens and mechanically loaded thin wall specimens provides a powerful tool to assess the impact of thermal fatigue loading and thermal strain gradients on component life.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A021. doi:10.1115/PVP2016-63291.

Direct strain controlled LCF data for solid specimens is still very rare. In PVP2013-97500 and PVP2014-28465 we reported results for niobium stabilized X6CrNiNb1810mod steel (type 347) fatigued in 325°C and 200°C PWR water according to VGB water chemistry specification. New data in this paper further confirms the conclusions: we are unable to repeat as high Fen factors or short lives as predicted according to NUREG/CR-6909.

The slowest strain rate used 4·10−6 in 325°C water would predict Fen > 12, i.e. laboratory specimen data below the current ASME design curve, but our results are superior for this steel generally used in German NPP’s. However, the difference is not necessarily grade specific. Use of 100% relevant fabricated material batch and standard LCF methodology are regarded to play an important role.

Notable hardening can be measured, when long duration holds in elevated temperatures are introduced between blocks of cyclic strains at lower temperatures. This is the case for thermal gradient loaded primary circuit components, e.g. the PWR pressurizer spray lines or surge line, which connects the pressurizer to primary coolant line. In PVP2011-57942 we reported improved endurances in fatigue tests aiming to roughly simulate steady state operation between fatigue transients in such NPP components. New test types have been introduced to generalize the results.

Mechanisms of time and temperature dependent relaxation of fatigue damage and/or improvement of material fatigue performance during holds are not yet fully revealed, but the rate controlling thermal activation energy is below shown to be near that for vacancy and interstitial atom diffusion. This allows us to draft a thermodynamic prediction model.

Improved accuracy of fatigue assessment helps in focusing optimally scheduled nondestructive testing to the most relevant locations and maintaining high level of reliability without excessive cost and radiation doses for inspection personnel.

This paper provides previously unpublished experimental results and proposes methods to improve transferability of laboratory test data to fatigue assessment of NPP components. The effects of material, water environment, temperature and service loading patterns are discussed.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A022. doi:10.1115/PVP2016-63294.

A tailored-for-purpose environmental fatigue testing facility was previously developed to perform direct strain-controlled tests on stainless steel in simulated PWR water. Strain in specimen mid-section is generated by the use of pneumatic bellows, and eddy current measurement is used as a feedback signal. The procedure conforms with the ASTM E 606 practice for low cycle fatigue, giving results which are directly compatible with the major NPP design codes.

Past studies were compiled in the NUREG/CR-6909 report and environmental reduction factors Fen were proposed to account for fatigue life reduction in hot water as compared to a reference value in air. This database exclusively contained non-stabilized stainless steels, mainly tested under stroke control. The applicability of the stainless steel Fen factor for stabilized alloys was already challenged in past papers (PVP2013-97500, PVP2014-28465). The results presented in this paper follow the same overall trend of lower experimental values (4.12–11.46) compared to those expected according to the NUREG report (9.49–10.37).

In this paper results of a dual strain rate test programme on niobium stabilized AISI 347 type stainless steel are presented and discussed in the context of the NUREG/CR-6909 Fen methodology. Special attention is paid to the effect of strain signal on fatigue life, which according to current prediction methods does not affect the value of Fen.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A023. doi:10.1115/PVP2016-63478.

The environmental fatigue evaluation for piping is performed based on conservative formulas and environmental correction factors (Fen) calculation because of the difficulty of finding transient strain history and to reduce the cost of the analysis. Therefore, more sophisticated analyses using a finite element model of piping would be required to meet Code requirement (cumulative fatigue usage factor including the environmental effects, CUFen < 1).

The environmental fatigue evaluation was performed considering the transient stress history for pressurizer surge and spray piping. The stress histories due to the thermal moment and thermal gradient through the wall thickness were evaluated using a three-dimensional piping system model. The strain rate for Fen calculation was calculated in accordance with draft Code Case 10-293 and 14-1177.

For the purpose of comparison, environmental fatigue usage factors and Fen are calculated based on the ASME Code NB-3600 rules. The differences of Fen, CUF, and CUFen are reviewed between the methods using NB-3200 rule and NB-3600 rule.

Finally, this paper shows that environmental fatigue evaluation considering transient history is an effective method in lowering the environmental CUF value of the piping.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A024. doi:10.1115/PVP2016-63497.

The reference Fatigue Crack Growth (FCG) rate behaviour for austenitic stainless steel in a deaerated Pressurised Water Reactor (PWR) environment is provided in ASME Section XI Code Case N-809. This FCG law is dependent on temperature, R ratio and environment, which is defined through a load rise time parameter. The basis for this law is contained within a large dataset of testing carried out by various industry sources.

Code Case N-809 defines rise time as the time for which the stress is increasing during a stress cycle, based on specimen testing in which sawtooth loading (typically 85% rise/15% fall) was primarily used. However it was found from testing of more complex waveforms carried out by Amec Foster Wheeler, which is reported in a separate paper at this conference, that the FCG rate is not always well characterised by this definition of rise time and was often found to be overly conservative. A number of different waveforms were considered, including simplified two-stage linear waveforms where the loading rate is different in the top and bottom half of the cycle, and those more representative of plant transients.

This paper presents a method to account for the environmental enhancement of complex waveforms with variable loading rate by weighting the loading rate according to its position within the loading cycle. In this way an effective rise time can be obtained for use in Code Case N-809. This method was found to give good agreement with the experimental data generated using complex waveforms.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A025. doi:10.1115/PVP2016-63584.

Small specimen fatigue testing is challenging in simulated LWR coolant environments at elevated temperatures and pressures. Two approaches to isothermal uniaxial testing in such environments have been developed: use of an autoclave to contain the environment around the specimen, which is conventionally of a solid design (e.g. circular cross-section, parallel sided gauge length); and use of a thin-walled hollow or tubular specimen, where the coolant environment passes through the bore of the specimen.

It is often assumed that fatigue lives measured using these two specimen designs are equivalent. However, recent isothermal strain-controlled fatigue endurance tests on a single heat of Type 304L stainless steel at Amec Foster Wheeler — on behalf of Rolls-Royce — have indicated a significant difference in life from testing of these different specimen designs in high temperature PWR coolant, with hollow specimens consistently giving shorter lives.

This paper presents those test results, and identifies a range of possible reasons for the differences in fatigue life through consideration of relevant literature and laboratory examination of failed specimens.

These new test results have potentially significant implications for test programmes in which solid specimen test results in air are compared to hollow specimen results in LWR environments, and for fatigue databases that include results from testing of both specimen types.

The use of a conversion factor, to be applied to fatigue lives from hollow specimens tests to allow comparison to solid specimen test results, is discussed. Further work to investigate the relevance of findings to further heats of material and to a wider range of loading conditions is identified.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A026. doi:10.1115/PVP2016-63640.

This paper demonstrates the feasibility of using theoretically-motivated crack tip strain rate (CTSR) models to estimate environmentally-assisted fatigue (EAF) crack growth rate (CGR) in light water reactor (LWR) environments. Four models, each combining one of two CTSR expressions with one of two theoretical derivations from Faraday’s Law, were fitted to measured CGR data under dynamic loadings. The four models were compared with each other and with experimental crack growth data from examples where various austenitic stainless steel base metals and nickel-alloy welds were tested in LWR environments under EAF, periodic partial unloading (PPU) with various hold times, loading gradients with increasing and decreasing stress intensity factor K (±dK/da), and constant K loading. All four models produced good fits to the data on some examples, and the models using the newer derivation performed well on all examples. Default model parameters and an equation for the distance from the crack tip at which strain rate is estimated were successful in simplifying the application of the models. Both variable effects and measured CGR were well modeled. Advantages and issues of theoretically-based CTSR models are presented.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A027. doi:10.1115/PVP2016-63796.

In order to develop new design fatigue curves for austenitic stainless steels, carbon steels and low alloy steels and a new design fatigue evaluation method that are rational and have a clear design basis, the Design Fatigue Curve (DFC) subcommittee was established in the Atomic Energy Research Committee in the Japan Welding Engineering Society. Mean stress effects for design fatigue curves are to be considered in the development of design fatigue curves. The Modified Goodman approach for mean stress effects is used in the design fatigue curves of the ASME B&PV Code. Tentative design fatigue curves were developed and studies on the effect of mean stress and design factors are on-going. Development of design fatigue curves, effect of mean stress and design factors is needed to establish a new fatigue design evaluation method. The DFC subcommittee has studied correction approaches for mean stress effects and the approaches of modified Goodman, Gerber, Peterson and Smith-Watson-Topper were compared using test data in literature. An appropriate approach for mean stress effects are discussed in this paper.

Topics: Fatigue , Stress , Design
Commentary by Dr. Valentin Fuster
2016;():V01AT01A028. doi:10.1115/PVP2016-63798.

To understand the fatigue behavior of austenitic stainless steels in a simulated PWR primary water environment, the patterns were studied. Austenitic stainless steel Type 316 plate was used as the test material.

Regarding non-isothermal testing: isothermal and non-isothermal fatigue tests were carried out for several patterns of temperature change and strain rate change. Typically, fatigue lives for non-isothermal tests with an out-of-phase strain change pattern were longer than those for isothermal tests.

Regarding strain holding testing: multiple groups of strain range cycles were separated by a long hold time and several test cases were carried out. Testing shows there is little difference in fatigue life for strain holding tests with high strain amplitude.

Commentary by Dr. Valentin Fuster

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

2016;():V01AT01A029. doi:10.1115/PVP2016-63027.

This paper examines conservatisms associated with two aspects of piping fatigue life as applied in NB 3600 fatigue calculations. First, piping component allowable cyclic life is defined by stress-life (S-N) curves based on constant displacement push-pull test results of small diameter (≤ 0.375 inch (≤ 9.5 mm)) smooth test specimens in air at room temperature. These life estimates are subsequently applied to all Class 1 piping components regardless of their actual size or thickness. Secondly, all component cyclic stresses are treated as uniform through-thickness membrane stresses when most transients induce stresses that vary across the thickness of the pipe.

The application of fatigue usage correction factors that account for: 1) increased cyclic life associated with the growth of engineering size fatigue cracks in thicker components and 2) the presence of actual through-thickness stress gradients is considered. An example case study indicates that, depending on the nature of the applied cyclic loading and pipe section thickness, considerable reductions in design fatigue usage can be obtained.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A030. doi:10.1115/PVP2016-63287.

The most common method used to determine the crack initiation life of a component containing a stress raiser in the low cycle fatigue regime is to calculate the maximum strain and then to use a strain-life curve. General practice is to base fatigue life estimates on the stabilized strain amplitude and to neglect the effects of transient behavior due to cyclic hardening or softening and ratcheting. For certain structures in which the accumulation of plastic strains may be significant, a separate check may be performed to ensure that these strains remain below a specified level. An objective of this research is to understand the notch tip local strain ratcheting and shakedown through finite element analyses and physical experiments. Towards planning a set of notched flat coupon experiments, this study performed analyses of various notched coupons under force-controlled cyclic loading. A question that will be addressed, what is the notch tip failure mechanism under a force-controlled load cycle with a non-zero mean force? Smooth specimens under such a force-controlled load cycle normally results in strain ratcheting. It is investigated whether notch tip strain responds in a similar manner under a force controlled loading cycle. The analysis results show that the strain ratcheting rate at the notch tip depends on the sharpness of the notch. In case of semi-circular and blunt elliptical notches shakedown of strain ratcheting within 25 cycles is observed, whereas for the sharp elliptical notch strain ratcheting doesn’t shakedown after 300 cycles. A novel observation made from the analysis results is that the mean stress at the notch tip gradually decreases with inelastic cycle while the stress amplitude remains unchanged. These result and future experimental plan on notch specimens are presented in this article.

Topics: Stress , Geometry
Commentary by Dr. Valentin Fuster
2016;():V01AT01A031. doi:10.1115/PVP2016-63410.

The stress states of elbow and tee pipes are complex and different from those of straight pipes. Several researchers have reported the low-cycle fatigue lives of elbows and tees under cyclic bending with internal pressure conditions. In this work, finite element analyses were carried out to simulate the reported experimental results of elbows and tees. The crack initiation area and the crack growth direction were successfully predicted by the analyses. The analytical results showed that the revised universal slope method can accurately predict the low-cycle fatigue lives of elbow and tee pipes under internal pressure conditions regardless of differences in shape and dimensions.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A032. doi:10.1115/PVP2016-63434.

Fatigue life can be divided into cycles of crack initiation and those in which the initiated crack grows to macroscopic size. In crack growth analysis, it is possible to consider the effect of the strain or stress gradient in the depth direction on the fatigue life. Therefore, flaw tolerance assessments allow reasonable fatigue life prediction. The fatigue life is reduced in the primary water environment of pressurized water reactor (PWR) nuclear power plants, and the correction factor Fen is used for considering the fatigue life reduction in fatigue damage assessments. To apply the flaw tolerance concept to a PWR water environment, the correction factor must be applied not to the fatigue life but to the number of cycles for crack growth. In this study, the fatigue life reduction in the PWR environment was correlated to the crack growth acceleration for a flaw tolerance assessment. The crack growth rates were obtained from fatigue life tests and crack growth tests performed in the PWR environment using Type 316 stainless steel. Then, the fatigue life was estimated by predicting the crack growth from an initial depth of 20 μm. It was concluded that a reasonable flaw tolerance assessment can be performed by using the strain intensity factor. The fatigue life reduction was successfully replaced with the crack growth acceleration.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A033. doi:10.1115/PVP2016-63635.

As early as the 1950’s, practitioners observed progressive distortion in thin-walled pressure vessels subjected to a constant axial stress and a cyclic thermal stress. In the late 1960’s, Bree [1] developed a theory and corresponding diagram plotting the primary membrane stress versus the cyclic thermal stress which delineates the various zones of plastic behavior. The zones include elastic cycling, plastic cycling, elastic cycling after initial plasticity, and ratcheting leading to incremental growth. This paper revisits the original Bree problem and investigates many of the recent advancements made to alleviate several of the simplifying assumptions Bree made in developing the diagram, to bring it in line with more modern operating conditions. In particular, a novel modification to the Bree diagram to account for the ratio of the yield stress at the operating extremums is proposed. This paper also reviews some advancements made to incorporate creep into the problem and discusses the operating conditions wherein creep effects may be significant. The outcomes of this paper will help expand the applicability of the Bree diagram, broadening its scope to encompass operating conditions more representative of modern applications.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A034. doi:10.1115/PVP2016-63773.

The ASME B&PV Code provides design by analysis rules that address failure mechanisms under cyclic loading. One of these potential failure mechanisms is incremental plastic collapse, or ratcheting. Miller presented the technical basis for the present Code requirements in a technical paper in 1959. Miller’s equations for the ratchet boundary address a beam under a cyclic through-thickness thermal gradient acting together with a steady axial mechanical load. This ratchet boundary applies approximately to a pressurized cylinder with through-thickness thermal bending stress. Conditions arise sometimes in practice where cooling or heating is applied simultaneously to the inner and outer surface of pressure boundary. The extreme case of such a scenario arises when both surfaces experience the same thermal condition such that there is a cyclic thermal stress but both zero membrane thermal stress and zero thermal bending stress The question is, could ratcheting occur in this case?

This paper derives the ratchet boundary for cases when the maximum temperature occurs mid-way through the thickness. The linearized stress due to thermal loading is zero. The solution is obtained using FE analysis and the Non-Cyclic Method (NCM) that has been proposed previously by the authors. The NCM is a generalization of the static shakedown theorem and allows the ratchet boundary to be calculated for both elastic and elastic-plastic cyclic stress states.

Topics: Stress , Membranes
Commentary by Dr. Valentin Fuster
2016;():V01AT01A035. doi:10.1115/PVP2016-63861.

Traditional design fatigue analyses of pressure vessels and piping equipment have typically used linear-elastic stress analyses, where the stresses caused by various loads, such as thermal, pressure, bending moments, etc. are combined using the principle of linear superposition. Based on high stress locations, geometric and material discontinuities, and other engineering judgements, stress classification lines (SCLs) were defined for where fatigue usage factors would be calculated. It was then necessary to apply simplified elastic-plastic penalty factors, based on the through-wall linearized stresses, to the peak stress amplitudes, in order to account for the nonlinear behavior of materials. Nonlinear finite element analysis that directly calculates strains were not typically used, because of computing and material modeling limitations. However, such analyses, even for complex three-dimensional structures, have become much more practical today with advancements in computing speed and storage capacity. ASME Section III Subarticle NB-3200 includes a provision for performing nonlinear (or “plastic”) analysis (NB-3228.4(c)), but little to no guidance is provided for how to perform the analysis itself. In addition, the procedure for computing the strain range, as currently written in the Code, has been identified as being limited to a uniaxial stress condition and is fundamentally inconsistent with the traditional elastic methodology. This paper provides a proposal for an improved approach for computing fatigue usage and strain rates using nonlinear plastic analysis. Additional guidance for performing these analyses is provided, as they are expected to be used more frequently into the future.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A036. doi:10.1115/PVP2016-64028.

Previous stress analyses have shown that the requirement of exclusively using the branch pipe cross-section properties for the NB-3600 Class 1 branch connection stress analyses can be very severe (conservative), when compared with using the branch nozzle cross-section properties. In the analyses performed for a 2015 PVP technical paper (Ref. 2), one of the questions raised was the fact that only two specified bending moments were used in those previous stress analyses: 1,000,000 in-lbf as the run bending moment and 20,000 in-lbf as the branch bending moment. In fact, that was the main question that was raised. Therefore, the purpose of this technical paper is to perform additional stress analyses, based on various bending moments on the run side and on the branch side. The impact of using these various bending moments will be evaluated.

The 2001 Edition of the ASME-Code removed a note at the bottom of the branch connection NB-3600 Figure. This note is required to be able to perform more accurate NB-3600 piping stress analyses of the branch connection. There is at this time a suggestion to restore that note.

Commentary by Dr. Valentin Fuster

Codes and Standards: Fatigue Assessment and Management — A Probabilistic Perspective

2016;():V01AT01A037. doi:10.1115/PVP2016-63670.

The cycle time, and, in particular, the duration of the quenching phase, is known to play an important role in the rate of damage accumulation in a coke drum. A shorter cycle is desirable because it allows for increased production but this comes at the cost of more frequent repairs and a shorter overall life. Therefore, a trade-off decision needs to be made in order to balance these two effects and maximize profitability.

This paper provides guidance for making these sorts of decisions by looking at several different coke drum operating strategies from the perspective of a high level financial analysis, taking into account many cost/benefit factors in order to determine the optimal strategy that maximizes profitability with various levels of confidence. It is shown that the most critical factor is predicting the rate of repairs as the drum ages. This depends on many factors and can only be determined probabilistically due to the many uncertainties involved. One of the most important factors driving the repair rate is the frequency at which fatigue cracks initiate and grow. A probabilistic fatigue model is used to describe this, taking into account the large surface area of the coke drum, which provides more potential nucleation sites for fatigue crack initiation.

It is shown that there is, indeed, an optimal, intermediate cycle time for certain conditions. For other situations running with the shortest possible cycle time is shown to be the best choice.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A038. doi:10.1115/PVP2016-63676.

A coke drum experiences operating conditions that are severe and unrelenting, and their lifespans reflect this. In this paper, a coupled thermal-mechanical model of an operating coke drum is presented. The impacts of important operating parameters, such as the cycle time and quench rate, on an ideal operating cycle and a cycle that produces hotspots are discussed. The model is constructed in three stages: (i) the through-wall temperature distribution is probabilistically determined using a Bayesian inverse approach that calculates the heat flux profile on the inside of the drum wall from a temperature reading on the outside of the wall; (ii) a thermal-mechanical finite element analysis of the drum wall is conducted using the heat flux profile as a spatially and temporally varying boundary condition — to obtain the stress and strain behavior at a critical location (weld seam) on the drum; and (iii) hotspots are introduced in the vicinity of the weld through a novel approach, and their impact on the plastic behavior of the drum wall is examined. The influence of the operating conditions on the behavior of the coke drum is to be incorporated into a financial analysis of the entire coke drum’s life cycle — in a companion paper — to determine an optimal operating strategy.

Topics: Coke
Commentary by Dr. Valentin Fuster

Codes and Standards: Fatigue Monitoring and Related Assessment Methods

2016;():V01AT01A039. doi:10.1115/PVP2016-63374.

In accordance with the recommendation of USNRC and the U.S. license renewal experiences, the effect of reactor coolant environment on the fatigue life has to be considered for the continued operation of operating nuclear power plants as well as for the design of new plants in Korea. The reason is that it is very important to maintain the structural integrity and reliability of the nuclear power plants against the fatigue failure during operation. Fatigue monitoring system has been considered as a practical way to ensure safe operation of the nuclear power plants in terms of the fatigue. The fatigue monitoring system evaluates various plant conditions and their effects on the monitored location to give quantified value that indicates accumulated fatigue damage up to date. From this, the authors have developed a fatigue monitoring system, named NuFMS (Nuclear Fatigue Monitoring System) in web environment and has been being applied widely to Korean nuclear plants. In this paper, overall configuration and characteristics of the NuFMS are described in detail.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A040. doi:10.1115/PVP2016-63747.

This paper presents the transference of a well-established nuclear power industry technique to perform environmentally-assisted fatigue monitoring for application to offshore, subsea High Pressure, High Temperature (HPHT) components using existing instrumentation. Applying fatigue monitoring is an effective way to manage these assets against the threat of fatigue failures and meet regulatory expectations during the service life. Our objective is to provide the background and overview of the approach so that design engineers better understand and evaluate its potential in contrast to traditional design changes that may actually produce an unintended consequence that increases the fatigue potential associated with heavy section thicknesses. This paper describes the overall analysis approach that may be used to perform fatigue monitoring of an HPHT component. At the conclusion, recommendations are provided for further research and development to close existing gaps.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A041. doi:10.1115/PVP2016-63917.

A comparison of currently available codes for assessment of fatigue crack growth, including ASME (America Society of Mechanical Engineers) SEC. XI, FKM (Forchungskuratorium Maschinenbau) guideline, WES (Japan Welding Engineering Society) 2805, BS7910 and JSME (The Japan Society of Mechanical Engineers), was carried out by paying attention to the suitability of application and the easiness to obtain the parameters, based on fatigue crack growth data of Cr-Ni-Mo-V steel welded joints. Results showed that fatigue crack growth curves provided by the FKM or WES were good choice when few inputs were at hand while the curves in the BS7910, JSME and ASME were recommended for precise estimation. It was indicated that the assessment of welded joints solely by fatigue crack growth behavior at base metal part and the assessment of fatigue crack growth for the aged condition by as-received one both resulted in non-conservativeness, albeit dependent on the range of stress ratios, R. A new bilinear form of fatigue crack growth model independent of R was developed based on transition point occurred in the near-threshold regime. This constituted the bilinear approach to fatigue assessment, and thus contributed to the optimization of fatigue assessment in the near-threshold regime.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A042. doi:10.1115/PVP2016-63931.

This paper points out some relevant aspects of the simplified elasto-plastic fatigue analysis as addressed in the ASME Code Section III Subsection NB and its application to two structural components that are subjected to a slow or to a fast thermal transient. The structural components considered are a thick-walled pipe and a nozzle-to-vessel junction. For the case of the thick-walled pipe, a closed form analytical solution proposed by Albrecht for pipes subjected transient temperature loading was implemented and its results were compared to coupled thermal and mechanical finite element analyses using a commercial finite element software. The application of the analytical solution allows for an optimization of the time consumed to obtain the stresses that occur across the thickness of the pipe as a function of time, i.e. the membrane plus bending plus peak stress range, Sp. The analytical solution equally allows for the linearization of the stress components actuating along the pipe thickness for all time steps considered within the thermal stress solution. This yields the membrane plus bending stress range, Sn, and allows for a design code conforming plasticity correction by means of Ke factors. In the considered case of the nozzle-to-vessel junction, a finite element solution was used. It was one aim of the study to point out, that under fast transients loading situations the relevant stresses Sp and Sn do not necessarily coincide with each other.

In the ASME Code the alternating stress Sa is a function of the factor Ke and of the range of Sp, with Ke being a function of the range of Sn and of the material properties. Consequently, a non-conservative fatigue analysis may result in the case of performing cycle counting only based on the time history of the critical Sp values and simply assigning the corresponding Sn and Ke values. This paper exemplifies one of those cases and proposes a method to overcome this problem.

Commentary by Dr. Valentin Fuster

Codes and Standards: High Pressure Hydrogen Storage and Transportation

2016;():V01AT01A043. doi:10.1115/PVP2016-63315.

The storage of hydrogen in a compressed gaseous form offers the simplest solution in terms of infrastructure requirements and has become the most highly developed hydrogen storage method. Low cost and large vessels for bulk hydrogen storage are needed at central production plants, geologic storage sites, terminals and refueling stations. A multifunctional steel layered vessel (MSLV) for stationary hydrogen storage with maximum design pressure of 98 MPa has been developed. First of all, the basic structure and characteristics of the vessel were introduced. Secondly, the stress in the cylindrical shell of the MSLV was studied based on the ribbon-width-direction effective normal stress and shear stress sub-models. Besides, the stresses in the hemispherical head and reinforcing ring were obtained by combining finite element analysis with experiments in the meantime. Finally, safety of the vessel was evaluated mainly by hydrogen compatibility tests of the weld joints of austenitic stainless steel S31603 under 98MPa gaseous hydrogen according to ANSI/CSA CHMC 1-2014, as well as MSLV’s feature of burst resistant and easy for online safety monitoring. Research shows that hydrogen embrittlement of MSLV was mitigated, because the stress in the inner shell of MSLV is low, and austenitic stainless steel and its weld are well compatible with high pressure hydrogen.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A044. doi:10.1115/PVP2016-63371.

A novel Steel Concrete Composite Vessel (SCCV) was designed and engineered for stationary high-pressure gaseous hydrogen storage applications. SCCV comprises four major innovations: (1) flexible modular design for storage stations for scalability to meet different storage pressure and capacity needs, flexibility for cost optimization, and system reliability and safety, (2) composite storage vessel design and construction with an inner steel vessel encased in a pre-stressed and reinforced outer concrete shellshell, (3) layered steel vessel wall and vent holes to address the hydrogen embrittlement (HE) problem by design, and (4) integrated sensor system to monitor the structural integrity and operation status of the storage system. Together, these innovations form an integrated approach to make the SCCV cost competitive and inherently safe for stationary high-pressure hydrogen storage services.

A demonstration SCCV has been designed and fabricated to demonstrate its technical feasibility. Capable of storing approximately 89 kg of gaseous hydrogen at 6250 psi (430 bar), the demonstration vessel was designed to include all major features of SCCV design and fabricated with today’s manufacturing technologies and code/standard requirements. Two crucial tests have been performed on this demonstration vessel. A hydro-test was successfully carried out to 8950 psi per ASME VIII-2 requirements. The cyclic hydrogen pressure test between 2000 psi and 6000 psi is currently being performed to validate its use for high-pressure hydrogen storage. Multiple sensors, such as pressure sensors and strain gages, were incorporated in the demonstration SCCV to collect information to validate the design and operation of SCCV. Key design parameters and test data on its performance are summarized in this paper.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A045. doi:10.1115/PVP2016-63389.

Hydrogen desorption from 6061 aluminum alloy during deformation was investigated by a new type of ultrahigh vacuum apparatus. The new apparatus which combined tension and outgassing experiments can be used to measure hydrogen desorption from the specimen during deformation. Outgassing corresponded to the deformation process, which initially released surface-adsorbed hydrogen. After elastic deformation, the as-received specimen released hydrogen because of the rupturing of the surface oxide layer. The hydrogen-charged specimen showed lower hydrogen concentration and hydrogen release rate than the as-received specimen. The second hydrogen release peak was due to the hydrogen accumulation at the interface between the oxide layer and matrix. The irregular sharp hydrogen release peaks resulted from the hydrogen bubble exposed to the surface during the rupture of the surface oxide layer.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A046. doi:10.1115/PVP2016-63568.

Current standards governing the design, qualification and in-service inspection of carbon fibre composite cylinders do not facilitate to optimise cylinder design. The requirements have been adapted from standards for metallic cylinders and cannot easily quantify the degradation processes in composite materials. In this article, the results of hydraulic and hydrogen pressure cycle life tests performed on composite reinforced tanks with a metal liner (type 3) and with a high density polymer liner (type 4) are shown. Moreover, the degradation measured by means of residual strength of the tanks after the cycling tests have been compared. It has been found that the most critical aging for metal based composite cylinder is the gaseous cycling while type 4 designs seem to be more sensitive to hydraulic cycling at high temperature.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A047. doi:10.1115/PVP2016-63597.

To prevent the on-board storage tank from burst at vehicle fire scenario, pressure relief device (PRD) is required to be installed to the tank and timely activated to release internal high-pressure hydrogen. Actually, there are two types of PRDs (i.e. thermally-activated and pressure-activated PRDs), and four types of tanks such as all-metal, hoop/fully-wrapped with metal liner and fully-wrapped with plastic liner. Great importance should be attached to the using of PRDs for all types of tanks in consideration of the risk of tank burst caused by fire. However, there are great differences in the requirements for the using of PRDs in hydrogen storage tank standards such as GTR-HFCV, ISO/TS 15869, JARI S 001 and TSG R006. Compared with compressed natural gas tank standards, PRD requirements in hydrogen storage tank standards are discussed in this paper. Moreover, key influencing factors on the activation of thermally-activated and pressure-activated PRDs are analyzed in detail based on fire test data. Finally, some advices for the using of PRDs of hydrogen storage tanks are proposed.

Commentary by Dr. Valentin Fuster

Codes and Standards: High Temperature Codes and Standards

2016;():V01AT01A048. doi:10.1115/PVP2016-63252.

Present work includes creep analysis of ASTM A 335-Grade P22 welds on the high-temperature-section (superheater/reheater) lower headers of bottom-supported heat-recovery steam generators (HRSG): they may be critical because of the long continued service (175000 hours or twenty years), internal reactions, stress intensification at the end-plug and finned-tube joint to the cylinder respectively. Weld life results from the Italian creep code compare to those predicted by the American standard API 579-1. Classical methods applied on the circumferential weld show consistent Von-Mises equivalent stress with both ASME and Italian pressure formulae’s (for the cylinder wall). Numerical-model stress analysis for the finned-tube tee weld shows results consistent with those on both sides of the intersection (pressure-load condition, linear); it shows instead unacceptable values for thermal-load condition, linear and nonlinear. Creep strain values are generally small for both joints, with rates in the weld metal lower than base’s for the circumferential weld. Analysis considers a strength reduction factor for the welds, disregarding possible residual stresses; it assumes a Norton-Bailey power-law for the creep, experimental data from short duration tests providing the basis for extracting coefficients. API 579-1 (Level 1 through 3) assessment results are consistent with the Italian creep code. Creep lives on the weld are consistent with those on both sides of the intersection, the circumferential weld showing lives longer than 300000 hours.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A049. doi:10.1115/PVP2016-63647.

Cr-Mo steel reactors being purchased for the oil refineries, gasification, and other industries have excellent standards for achieving fit for service designs, material specifications, and fabrication requirements. These standards include the ASME Pressure Vessel Code and API Recommended Practice (RP) 934-A for the 2¼Cr-1Mo alloy family and RP 934-C for 1¼Cr-½Mo. The RPs’ scopes indicate that they were developed for the most severe, high pressure refinery services, which typically require thick wall reactors. However, these standards are also widely used by industry as guidance for thinner Cr-Mo equipment. This paper aims to identify and present the appropriate requirements derived from the RPs that can be justified for less severe services and/or thin wall equipment falling outside the intended scope of API RP 934-A and 934-C.

Commentary by Dr. Valentin Fuster

Codes and Standards: Hydrogen Effects on Material Behavior for Structural Integrity Assessment (Joint With MF-3)

2016;():V01AT01A050. doi:10.1115/PVP2016-63073.

Zr-2.5Nb pressure tubes in CANDU 1 reactors are susceptible to hydride formation when the solubility of hydrogen in the pressure tube material is exceeded. As temperature decreases, the propensity to hydride formation increases due to the decreasing solubility of hydrogen in the Zr-2.5Nb matrix. Experiments have shown that the presence of hydrides is associated with reduction in the fracture toughness of Zr-2.5Nb pressure tubes below normal operating temperatures. Cohesive-zone approach has recently been used to address this effect. Using this approach, the reduction in fracture toughness due to hydrides was modeled by a decrease in the cohesive-zone restraining stress caused by the hydride fracture and subsequent failure of matrix ligaments between the fractured hydrides. As part of the cohesive-zone model development, the ligament thickness, as represented by the radial spacing between adjacent fractured circumferential hydrides, was characterized quantitatively. Optical micrographs were prepared from post-tested fracture toughness specimens, and quantitative metallography was performed to characterize the hydride morphology in the radial-circumferential plane of the pressure tube. In the material with a relatively low fraction of radial hydrides, further analysis was performed to characterize the radial spacing between adjacent fractured circumferential hydrides. The discrete empirical distributions were established and parameterized using continuous probability density functions. The resultant parametric distributions of radial hydride spacing were then used to infer the proportion of matrix ligaments, whose thickness would not exceed the threshold value for low-energy failure. This paper describes the methodology used in this assessment and discusses its results.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A051. doi:10.1115/PVP2016-63156.

Zirconium alloys, as used in water-cooled nuclear reactors, are susceptible to a time-dependent failure mechanism known as Delayed Hydride Cracking, or DHC. Corrosion of zirconium alloy in the presence of water generates hydrogen that subsequently diffuses through the metallic structure in response to concentration, temperature and hydrostatic stress gradients. As such, regions of increased hydrogen concentration develop at stress concentrating features, leading to zirconium hydride precipitation. Regions containing zirconium hydride are brittle and prone to failure if plant transient loads are sufficient.

This paper demonstrates the application of the Extended Finite Element Method, or XFEM, to the assessment of the DHC susceptibility of stress concentrating features, typical of those considered in the structural integrity assessment of heavy water pressure tube reactors. The method enables the calculation of a DHC threshold load. This paper builds on the process-zone approach that is currently used to provide the industry-standard DHC assessment of zirconium alloy pressure tubes and also recent developments that have extended the application of the process-zone approach to arbitrary geometries by the use of finite element cohesive-zone analysis. In the standard cohesive-zone approach, regions of cohesive elements are situated in discrete locations where the formation of zirconium hydride is anticipated. In contrast, the use of XFEM based cohesive formulations removes the requirement to define cohesive zones a priori, thereby allowing the assessment of geometries in which the location of hydride material is not known.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A052. doi:10.1115/PVP2016-63729.

Structural components of nuclear reactors made from zirconium alloys are subject to degradation mechanisms associated with hydrogen pickup during their operating life. Even small amounts of hydrogen isotopes (∼tens of wt. ppm) can significantly reduce the material’s fracture toughness and make it susceptible to a sub-critical crack growth mechanism known as Delayed Hydride Cracking (DHC). The mitigation of potentially costly failures of these components requires assessments based upon not only the bulk concentration of hydrogen, but the concentration of hydrogen precipitated as hydrides, and in solution. The behaviour of the zirconium-hydrogen system near ∼100 ppm hydrogen concentration continues to be the subject of research, primarily due to its otherwise complex behaviour and critical role in nuclear plant operations. There is an anisotropic volumetric expansion associated with the precipitation of zirconium hydride, and this precipitation event results in plasticity in the surrounding matrix and compressive strain in the hydride phase. This leads to both a pronounced solubility hysteresis upon dissolution and precipitation, as well as variances in solubility behaviour depending on the prior thermal and mechanical history.

In the present study, high-energy synchrotron x-ray diffraction is used to study the evolution of hydrogen solubility in Zr-2.5 wt% Nb pressure tube material with different hydrogen concentrations in situ during thermal cycling between 100 and 400°C. This technique provides the ability to directly measure the amount of hydride in a given sample at different temperatures, and to evaluate zirconium-hydrogen precipitation and dissolution kinetics.

Topics: X-rays , Precipitation
Commentary by Dr. Valentin Fuster
2016;():V01AT01A053. doi:10.1115/PVP2016-64009.

Flaws found during in-service inspection of CANDU(1) Zr-2.5Nb pressure tubes 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. Crack initiation from in-service flaws can be caused by the presence of hydrogen in operating pressure tubes and resultant formation of hydrided regions at the flaw tips during reactor heat-up and cool-down cycles. Zr-2.5Nb pressure tubes in the as-manufactured condition contain hydrogen as an impurity element. During operation, the pressure tube absorbs deuterium, which is a hydrogen isotope, from the corrosion reaction of the zirconium with the heavy water coolant. In addition, deuterium ingresses into the pressure tube in the rolled joint region. The level of hydrogen isotope in pressure tubes increases with operating time.

Over the years, Canadian CANDU industry has carried out extensive experimental and analytical programs to develop evaluation procedures for crack initiation from in-service flaws in Zr-2.5Nb pressure tubes. Crack initiation experiments were performed on pressure tube specimens with machined notches to quantify resistance to crack initiation under various simulated flaw geometries and operating conditions such as operating load and hydrogen concentration. Predictive engineering models for crack initiation have been developed based on understandings of crack initiation and experimental data. A set of technical requirements, including engineering procedures and acceptance criteria, for evaluation of crack initiation from in-service flaws in operating pressure tubes has been developed and implemented in the CSA Standard N285.8. A high level review of the development of these flaw evaluation procedures is described in this paper. Operating experience with the application of the developed flaw evaluation procedure is also provided.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A054. doi:10.1115/PVP2016-64024.

In-service flaws in cold-worked Zr-2.5 Nb pressure tubes in CANDU(1) reactors are susceptible to a phenomenon known as delayed hydride cracking (DHC). The material is susceptible to DHC when there is diffusion of hydrogen atoms to a service-induced flaw, precipitation of hydrides on appropriately oriented crystallographic planes in the zirconium alloy matrix material, and development of a hydrided region at the flaw tip. The hydrided region could then fracture to the extent that a crack forms and DHC is said to have initiated. Examples of in-service flaws are fuel bundle scratches, crevice corrosion marks, fuel bundle bearing pad fretting flaws, and debris fretting flaws. These flaws are volumetric in nature. Evaluation of DHC initiation from the flaw is a requirement of Canadian Standards Association (CSA) Standard N285.8. This paper describes the validation of the weight function based process-zone model for evaluation of pressure tube flaws for DHC initiation. Validation was performed by comparing the predicted threshold load levels for DHC initiation with the results from DHC initiation experiments on small notched specimens. The notches in the specimens simulate axial in-service flaws in the pressure tube. The validation was performed for both un-irradiated and pre-irradiated pressure tube material.

Commentary by Dr. Valentin Fuster

Codes and Standards: Hydrogen Flakes Assessment in RPVs

2016;():V01AT01A055. doi:10.1115/PVP2016-63231.

In the summer of 2012, the detection of a large number of quasi-laminar flaw indications in the reactor vessel beltline ring forgings of two Belgian pressurized water reactors (Doel Unit 3 and Tihange Unit 2) posed a significant safety threat that led the licensee to shutdown both plants. Those indications were identified by the licensee as hydrogen flakes that developed during the fabrication of the forgings. As a prerequisite for a potential restart of the units, the Belgian Nuclear Safety Regulator, the Federal Agency for Nuclear Control (FANC), requested the licensee to provide, for each unit, a safety case demonstrating the acceptability of the reactor pressure vessel for continued operation. As the technical subsidiary of the FANC, Bel V performed a safety evaluation of the condition of the reactor pressure vessels. The paper documents the approach Bel V used in his safety evaluation and the criteria he defined to evaluate the acceptability of the hydrogen flaking damage in the reactor pressure vessels.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A056. doi:10.1115/PVP2016-63593.

In this paper, work done for defining the lowest temperature for ductile fracture initiation in piping and fittings was extended to a sample reactor pressure vessel (RPV). The methodology used is called “Master Curve of Fracture Transition Temperatures” and was developed from correlations of thousands of laboratory tests and hundreds of pipe tests originally presented in 2005. Since then it has been extended to; low and high strength line-pipe steel base metals (oil/gas industry), blunt flaws as well as sharp cracks, girth welds for the oil/gas industry, fracture of pipe fittings/valves, and is the technical basis for the lowest temperature for ductile crack initiation in ferritic piping in Appendix C of ASME Section XI. Since the methodology is quite different than traditional approaches for nuclear component applications, the general methodology will be reviewed, as well as analysis results showing how surveillance capsule Charpy data could be used to predict the lowest temperature where ductile crack initiation would occur in RPVs. Once this temperature is established, then the upper-shelf toughness values can be used to determine if the failure is EPFM or limit-load, and the associated failure stresses. This temperature could be used for defining the pressure-temperature limits to assure that the RPV material has a high flaw tolerance. This methodology was proposed for the Doel 3 and Tihange 2 RPVs in Belgium.

One key concern for operation of RPVs is determining the lowest operating temperature where ductile crack initiation is still expected which can be based on studies of irradiation effects on the measured toughness from Charpy tests of material in surveillance capsules. With ductile crack initiation the flaw tolerance is quite high even with long-term irradiation damage to the material. In fact, if the toughness is closer to limit-load than LEFM in a FAD analysis, then as long as there is ductile initiation, the irradiation effects increase the strength which could increase the flaw tolerance. There are some ongoing efforts within the ASME Section XI activities to define the minimum temperature where ductile initiation occurs in fracture toughness testing (typically based on a mixture of Charpy and 1T CT specimen data); however, there are still thickness and constraint effects on the fracture toughness for more precise application to a thick-walled vessel with a postulated axial surface crack.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A057. doi:10.1115/PVP2016-63632.

The 2012 in-service vessel inspections performed at Doel-III and Tihange-II Nuclear Power Plant revealed a large number of ultrasonic indications in the forged core shells suggesting nearly-laminar defects attributed to hydrogen flakes induced during the component manufacturing process. These observations have initiated a very large experimental test program not only for reliably and accurately characterizing the ultrasonic indications but also to characterize the mechanical properties in presence of hydrogen flakes. A key point of the Safety Case was the justification of the material properties to be used in the structural integrity analysis, in particular the local fracture toughness of the material ahead of hydrogen flakes. An innovative methodology was used to manufacture fracture toughness specimens, including both compact tension and Charpy size specimens, with a hydrogen flake substituting the standard fatigue pre-crack. As a result, it is possible to measure the fracture toughness directly at the tip of real hydrogen flakes. This paper describes the procedure and the results that were obtained in comparison to standard fatigue precracked specimens.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A058. doi:10.1115/PVP2016-63765.

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected in the reactor pressure vessels, mainly in the lower and upper core shells. The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process.

As a consequence, both units remained core unloaded pending the elaboration of an extensive Safety Case demonstrating that they can be safely operated.

The Structural Integrity Assessment of the RPVs, through the Flaw Acceptability Analysis, aimed at demonstrating that the identified indications do not jeopardize the integrity of the reactor vessel in all operating modes, transients and accident conditions.

This demonstration, presented in this paper, has been done on the basis of a specific innovative methodology inspired by the ASME XI procedure but adapted to the nature and number of indications found in the Doel 3 and Tihange 2 RPVs.

Topics: Hydrogen
Commentary by Dr. Valentin Fuster
2016;():V01AT01A059. doi:10.1115/PVP2016-63766.

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, specific ultrasonic in-service inspections revealed a large number of quasi-laminar indications in the base metal of the reactor pressure vessels, mainly in the lower and upper core shells. The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process.

As a consequence, a Flaw Acceptability Assessment had to be performed as a part of the Safety Case demonstrating the fitness-for-service of these units.

In that framework, detailed analyses using eXtended Finite Element Method were conducted to model the specific character of hydrogen flakes. Their quasi-laminar orientation as well as their high density required setting up 3D multi-flaws model accounting for flaw interaction.

These calculations highlighted that even the most penalizing flaw configurations are harmless in terms of structural integrity despite the consideration of higher degradation of irradiated material toughness.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A060. doi:10.1115/PVP2016-63767.

In the framework of the hydrogen flakes issue concerning the reactor pressure vessels of the two Belgian NPP’s Doel 3 and Tihange 2, the Federal Agency of Nuclear Control required to perform tests on large scale specimens taken from a block representative of the pressure vessels with the double objective of validating the structural integrity approach and of verifying the load capacity of the specimens affected by flakes.

The large scale tests were led on many kinds of specimens: 4 points bending specimens, CT specimens and tensile specimens containing hydrogen flakes or flawed with EDM notches. All of these tests have been simulated using extend finite element method (XFEM).

The paper describes the linear elastic and elastic-plastic fracture mechanics calculations performed in the frame of these large scale tests using XFEM and presents the comparison between simulations and experiments.

A focus is done on the XFEM capabilities to model 3D complex shaped flaws like hydrogen flakes.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A061. doi:10.1115/PVP2016-63860.

During 2012 RPVs inspections, nearly-laminar indications were detected in the lower and upper core shells of Doel 3 and Tihange 2 RPVs. As a consequence, the Doel 3 and Tihange 2 NPPs have to stay in cold shutdown until it was proved that they can be safely operated. The performance and capability of the applied UT technique to detect and characterize hydrogen flakes had to be confirmed. To achieve this objective, extensive investigations were launched on blocks extracted from a steam generator shell known for containing hydrogen flakes.

The paper describes the UT validation performed and the results of the destructive tests which enable to demonstrate the capability of a straight ultrasonic beam to detect and size nearly laminar indications and to correctly identify ligaments between two adjacent indications. It addresses the formal extension of qualification performed by correlating UT measurements acquired on a large numbers of real hydrogen flakes, simulations of UT responses and destructive examinations.

Lessons learned from the experiments are finally described.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A062. doi:10.1115/PVP2016-63878.

During the summer outages of 2012, large numbers of indications were found inside the shell material of the Doel 3 and Tihange 2 reactor pressure vessels (RPV). Therefore both plants remained in cold shutdown with their core unloaded. A series of examinations, tests and inspections were performed leading to the conclusion that the indications are hydrogen flakes and that they do not affect the structural integrity of the RPVs, regardless the operating mode, transient or accident condition. All this was documented in the Safety Case Reports issued in December 2012 and the Safety Case Addenda issued in April 2013.

Based on those reports, the Belgian Federal Agency for Nuclear Control (FANC) authorized the restart of both units that went back on-line in June 2013. In parallel, a number of requirements from the FANC were addressed such as the qualification of the applied Ultrasonic Testing (UT) procedure, and mechanical testing of irradiated specimens containing hydrogen flakes. The preliminary tests showed unexpected results regarding the shift in RTNDT (Reference Temperature for Nil Ductility Transition) under irradiation that could not confirm the hypotheses considered in the initial Safety Case Reports. Therefore, the Licensee Electrabel decided to shut down both plants immediately.

In order to fully address this concern, the material test programme was extended including several RPV materials and covering additional irradiation campaigns. This led to a modification of the irradiation embrittlement trend curves considered in the structural integrity analysis. In addition, the qualification of the UT procedure led to an updated cartography of the flakes. The structural integrity assessments of both RPVs were revised accordingly. The final Safety Case Reports, confirming the fitness-for-continued operation of both RPVs, were submitted to the FANC in October 2015. FANC allowed restart of both units on November 17th, 2015.

The paper gives a historical overview of the Doel 3 and Tihange 2 RPV Safety Cases and explains how the roadmap was built in order to demonstrate the RPV’s structural integrity in the presence of hydrogen flakes.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A063. doi:10.1115/PVP2016-63881.

The formation of hydrogen flakes in Reactor Pressure Vessels (RPV’s) forgings is a well known phenomenon that can affect forging components during manufacturing. Following recent data from Belgium, it has been the object of important assessments during the last 3 years. We will present in this paper elements pertaining to the RPV’s of the French nuclear fleet.

A thorough review of fabrication processes and specifically of ultrasonic examination of forgings, including the oldest ones has demonstrated that the risk of leaving undetected hydrogen flakes in RPV’s forgings is not a problem in French RPVs. The French Regulator also clearly shared this position point in 1985. Parts with defects were once again observed at the beginning of 2012 but the defects were attributed to errors in hydrogen measurement at the steel maker shop. Those hydrogen induced defects were detected at an early manufacturing stage (before quality heat treatment) by ultrasonic examination, and the concerned parts were consequently rejected through the normal quality control process.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A064. doi:10.1115/PVP2016-63882.

During the 2012 outage at Doel 3 (D3) and Tihange 2 (T2) Nuclear Power Plants (NPP), a large number of nearly-laminar indications were detected mainly in the lower and upper core shells. The D3/T2 shells are made from solid casts that were pierced and forged.

Restart authorization in 2013 was accompanied by a number of “mid-term” requirements, to be completed during the first operating cycle after the restart. One of these requirements was the mechanical testing of irradiated specimens containing hydrogen flakes. These tests showed unexpected results regarding the shift in the Reference Temperature for Nil Ductility Transition (RTNDT) of the flaked material VB395 (Steam Generator shell rejected because of flakes) after irradiation.

This paper presents the root cause analysis of this unexpected behaviour and its transferability (or not) to the D3/T2 Reactor Pressure Vessels (RPVs).

A mechanistic and a manufacturing based approach were used, aiming at identifying the microstructural mechanisms responsible for the atypical embrittlement of VB395 and evaluating the plausibility of these mechanisms in the D3/T2 RPVs.

This work was based on expert’s opinions, literature data and test results. Both flaked and unflaked samples have been investigated in irradiated and non-irradiated condition.

All hydrogen-related mechanisms were excluded as root cause of the unexpected behaviour of VB395. Two possible mechanisms at the basis of the atypical embrittlement of VB395 were identified, but are still open to discussion. These mechanisms could be linked to the specific manufacturing history of the rejected VB395 shell.

Since the larger than predicted shift in transition temperature after irradiation of VB395 is not linked with the hydrogen flaking and since none of the specific manufacturing history features that are possible root causes are reported for the D3/T2 RPVs, the D3/T2 shells should not show the unexpected behaviour observed in VB395.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A065. doi:10.1115/PVP2016-63883.

This paper presents the root cause analysis and characterisation of the indications discovered in two Belgian pressurized water reactor (PWR) pressure vessels (Doel 3 and Tihange 2), during in service inspections performed in 2012.

The only plausible mechanism at the origin of the detected indications is hydrogen flaking during fabrication. Flaking could occur because of the local combination of high hydrogen concentration, stresses and susceptible microstructure.

The phenomenology study performed on several materials including flaked materials provided precious information for the Safety Case: it confirmed the cause of the indications and evidenced that flaking exclusively occurs in segregated zones and that the ligament between flakes is sound.

The paper also briefly addresses the possibility of evolution of the indications during operation: low cycle fatigue during transients is identified as the only mechanism likely to induce flaw growth in operation. A particular attention has been paid to the potential hydrogen effects.

Commentary by Dr. Valentin Fuster
2016;():V01AT01A066. doi:10.1115/PVP2016-63901.

During the summer outages of 2012, large numbers of nearly-laminar indications were found in the core shells of the Doel 3 and Tihange 2 reactor pressure vessels (RPV). As a consequence, both units remained in cold shutdown with their core unloaded. A series of examinations, tests and inspections were performed leading to the conclusion that the indications are hydrogen flakes and that they do not affect the structural integrity of the RPV, regardless of the operating mode, transient or accident condition. All this was documented in the Safety Case reports issued in December 2012 and in the Safety Case Addenda issued in April 2013 [1]. Based on those reports, the Belgian Federal Agency for Nuclear Control (FANC) authorized the restart of both units which went back on-line in June 2013.

A key input required for this Safety Case was the definition of the appropriate material properties, in particular fracture toughness, for the RPV shells affected by hydrogen flakes. A material testing program on non-irradiated materials evaluated aspects like the possible effects of macro-segregations and local segregations (ghost lines) and of specimen orientation on the fracture toughness. The irradiation embrittlement sensitivity of the zone of macro-segregation in which the flakes are located was evaluated on the basis of the maximum enrichment in Cu, P and Ni in macro-segregations based on literature data. This was the basis of the trend curve of RTNDT evolution vs. fluence used in the Safety Cases submitted in 2012–2013.

The restart authorization in 2013 was accompanied by a number of “mid-term” requirements, to be completed during the first operating cycle after the restart. One of these requirements was the mechanical testing of irradiated specimens containing hydrogen flakes, in order to confirm the conservativeness of the RTNDT trend curve used for the structural integrity analyses.

After a first irradiation campaign of a material containing hydrogen flakes in the BR2 reactor of the Belgian Nuclear Research Center SCK.CEN, atypical results were obtained and the utility decided to shut down the units in March 2014. Detailed investigations involving three additional irradiation campaigns in BR2 including other reference materials, among which another material affected by hydrogen flakes, were performed in order to characterize this atypical behaviour and to derive a new conservative RTNDT trend curve. The resulting trend curve was accepted by the FANC and was used in the 2015 Safety Cases [1]. An overview of the Doel 3 and Tihange 2 safety cases is given in [6].

The paper summarizes the results of the material investigations on non-irradiated and irradiated materials and the process leading to the definition of this conservative RTNDT trend curve.

Commentary by Dr. Valentin Fuster

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