bluuebunny/crossref_metadata_2025 huggingface dataset:
Relaxation of Exemption Requirement of PWHT for SA-508 Grade 1A, by Consideration Surface Welding Residual Stress As Evaluated by Instrumented Indentation Testing Method
Authors: Kim Jong-hyoung , Lee Jun Sang , Choi Sungki , Kim Jong-sung , Kwon Dongil
Published: October 2017

Generally, post-weld heat treatment is applied to decrease welding residual stress and improve the mechanical properties and microstructure of weldment, and its performance has been recommended for many years [1, 2]. However, current steel-making technology has improved significantly and, steel toughness levels have generally improved substantially [1]. Additionally for several quenched and tempered steels, it is reported that in some cases, mechanical properties such as tensile strength and impact toughness are degraded after post-weld heat treatment [3]. In addition, for large steel assemblies, post-weld heat treatment can be expensive, so that there is an economic incentive to avoid post-weld heat treatment [2]. The research presented here suggests a way to exempt post-weld heat treatment for SA-508 Grade 1A material, which is used for pressure vessels in nuclear power plants, by considering both mechanical properties and residual stress to simplify the welding procedure. Weldments made of 120 mm thick SA-508 Grade 1A should be post-weld heat treated, according to current ASME BPV Code. In order to increase the PWHT exemption thickness to 120 mm, we performed mechanical tests using welding coupons without PWHT; the test results satisfied current mechanical property criteria. We present a residual stress acceptance criterion based on brittle fracture criteria in this research.
Numerical Study of an Impusively Loaded Vessel Containing Double Versus Single Closure Bolt Patterns
Authors: Hess Joseph E. D.
Published: October 2017

Impulsively loaded pressure vessels are often closed using a bolted joint configured in a double staggered row pattern. The bolted joint design must maintain the placement of the vessel opening covers to support the structural integrity of the shell and also provide the necessary preload of sealing surfaces for leak prevention. Good design practice suggests configuring tensile loaded bolted joints with a double rows pattern in order to minimize prying against the bolt head induced by localized moments. Double bolt row patterns allow moments induced by load offsets to be reacted through contact of the faying surfaces of the bolted members and if separation occurs by differential axial loading of the two bolt rows. This acts to reduce direct prying of the mated members against the bolt heads. Material cost and operational time savings could be realized if a single bolt row design with acceptable performance was implemented. In this paper a detailed finite element model is described and calculation results are presented for two vessel configurations subjected to an impulsive load; a double staggered 64 bolt pattern and a single row 32 bolt pattern. Finite element results are compared to each other and to the rules of ASME Code Case 2564 in Section VIII, Division 3. Special attention is given to the loading induced in the bolts and to the relative deflection of faying surfaces containing seals. It will be shown that reducing the bolt count per opening from 64 to 32 results in increased peak response of the bolts, seal opening gaps, and shell. Nonetheless a single row bolt pattern does appear feasible and within the bounds of the Code Case.
Notched Fatigue of Austentic Alloys in Gaseous Hydrogen
Authors: Nibur Kevin A. , Gibbs Paul J. , Foulk James W. , San Marchi Chris
Published: October 2017

Circumferentially notched specimens of several austenitic stainless steel alloys subjected to positive load ratio, load-controlled fatigue have been cycled to failure in high pressure hydrogen gas. The number of cycles to failure for a given applied stress amplitude varies among the alloys tested indicating that control of fatigue life in hydrogen environments may be attained through informed alloy selection. The number of cycles to initiate a crack does not vary significantly among the alloys tested, however the total life to failure varied by over an order of magnitude. This difference in life is attributed to variations of the stress and strain fields ahead of the blunt notch. These fields are influenced by the strength and strain hardening characteristics of each alloy and they dictate the driving force for fatigue crack growth while the crack is small.
Evaluation of ASME Class 3 HDPE Flanged Joints
Authors: Musto Thomas M. , Frazee Glenn R. , Marohl Michael P. H.
Published: October 2017

In the design of piping systems, there are many options for transitioning between HDPE and metallic piping. One common option is the use of flanged joints. As a result of the visco-elastic nature of HDPE, the use of HDPE-to-metallic flanged joints requires special design considerations. When HDPE-to-metallic flanged joints are used in ASME Class 3 systems, the design is further complicated by the requirements provided in the ASME B&PV Code, Section III for flanged joint analysis.

This paper examines the differences between HDPE piping flanged joints and metallic piping flanged joints, including consideration of industry guidance and available industry testing results. The paper provides a proposed methodology for evaluating ASME Class 3 HDPE-to-metallic flanged joints and HDPE-to-HDPE flanged joints, including the determination of required bolt torque values and the determination of the maximum internal pressure that the joint can resist without experiencing leakage.
Minimizing Visual Effects and Optimizing Routes and Locations for Geothermal Steam Gathering System
Authors: Jonsson Magnus Thor , Magnusdottir Lilja
Published: October 2017

In geothermal power plants, the steam system is a costly component of the infrastructure in terms of construction requirements. The steam system consists of geothermal wells, well head equipment, pipelines and separators connecting wells to the separators and the separators to the power plant. This paper describes an approach for optimum design of a steam system, with focus on topology and route selection for pipelines transporting two-phase fluid from geothermal wells to a separation station and single-phase pipelines from the separator stations to the power plant.

This article proposes a new approach for selecting the locations of the separators and power plant and routes for both two phase-flow and single-phase flow pipelines in a geothermal steam gathering system. Multiple weighted distance maps calculated by uniform cost Dijkstra’s algorithm are used to find the optimum location of a site for a steam separator based on the flow capacity of geothermal wells. The two-phase flow routes are monotonic and the incline is slight in order to minimize the pressure drop and the slug flow conditions in the pipeline. Once the location of separators has been optimized, the optimum location of the power plant can be determined based on the route selection for both the pipelines transporting steam to the power plant and the pipelines transporting the used geothermal fluid from the power plant to the injection wells. The objective is to minimize both cost and visual effects.

A comparison of this method to traditional methods shows that when the method described in this paper is used, shorter routes and cheaper systems can be designed with less environmental impact.
An Analytical-Based Structural Strain Method for Low Cycle Fatigue Evaluation of Girth-Welded Pipes
Authors: Pei Xianjun , Wang Wei , Dong Pingsha
Published: October 2017

As a further extension to the structural stress based master S-N curve method adopted by ASME Div 2 since 2007, this paper presents an analytical-based structural strain method for girth-welded piping components. Here, structural strain is defined as outer and inner fiber strains calculated corresponding to a deformation condition in which a pipe section plane before deformation remains as a plane after deformation. The analytical formation takes into account all possible plastic deformation conditions a pipe section subjected to a combined remote cyclic bending and axial tension. A simple numerical procedure is used for solving both outer fiber and inner fiber strains, as well as the corresponding elastic core size. For fatigue evaluation purpose, the outer fiber strain can be used to calculate the corresponding pseudo elastic structural stress range so that the structural stress based master S-N curve can be directly used. Under linear elastic deformation conditions, the structural strain definition becomes exactly the same as that calculated by the structural stress method which is the basis on the ASME Div 2 master S-N was developed.

A set of a recent full scale girth-welded pipe component test data in low-cycle regime was analyzed using the structural strain method. The results showed that all these new test data fall well within the ASME Div 2 master S-N curve scatter band defined by mean+-standard deviations. In addition to its demonstrated effectiveness, the key advantage of this structural strain method is its simplicity for dealing with girth-welded pipe sections, since finite element stress analysis is no longer needed.
Adoption of the Composite Reinforced Pressure Vessels (CRPV) Into the ASME BPV Code
Authors: Biel Richard C. , Cano Gregory
Published: October 2017

Adoption of composite reinforced pressure vessels (CRPV) into the ASME Boiler and Pressure Vessel Code represented advancement in the technology of pressure vessels. The advantage of this construction technique is that the weight of a CRPV for compressed gas service built may be reduced to about one-half conventional pressure vessel of the same capacity. The concept of hoop wrapping fibers in a plastic composite (>90% fiber fill) makes full utilization of the fiber strength as the fibers share the hoop load with a metal cylinder. With reduced hoop stresses in the metal, a substantial reduction in wall thickness is attainable.

The process of adoption of this technology presented several challenges and some robust administrative hurdles. These included coordination with ASME BPV Code Section X for the composite application and Section VIII for the steel design and overall acceptance of the Case.

The most vexing technical challenge was the inspection of an unfinished weld on the inside of the shell from the outside of the shell. The next challenge was to gain consensus on the testing criteria for the acceptance of finished vessels.

Case 2390 was drafted in the winter of 2000 and spring of 2001 and approved for publication after nine revisions with an approval date of October 9, 2002. The Case was subsequently adopted into the body of ASME BPV Code Section VIII, Division 3 [1] (VIII-3) in the 2010 edition.
Effect of Variable Temperature on the Fatigue Life and Crack Growth Rates of Austenitic Stainless Steels in PWR Coolant Environments
Authors: Platts N. , Gill P. , Cruchley S. , Grieveson E. , Twite M.
Published: October 2017

The Pressurized Water Reactor (PWR) primary coolant environment is known both to significantly reduce the fatigue life of austenitic stainless steels and to lead to enhanced fatigue crack propagation rates. Relationships for the impact of the PWR coolant environment on fatigue life have been presented in NUREG/CR-6909 using an environmental fatigue correction factor (Fen), which is a function of temperature. Fatigue crack growth behavior has been codified in ASME Code Case N-809 in terms of parameters such as rise time, stress intensity factor, load ratio and temperature. However, plant performance suggests that the application of these predicted environmental effects using current assessment procedures for fatigue for plant transient loading may be unduly pessimistic. One potential reason for this over-conservatism is thought to be that, although the majority of plant design transients result from variations in thermal loading, most available data are derived from isothermal testing.

For the calculation of fatigue initiation life, NUREG/CR-6909 gives guidance on the effective temperature to be used in assessments of thermal transients. Recent results from Thermo-Mechanical Fatigue (TMF) testing on stainless steels in PWR coolant show that this guidance is conservative for out-of-phase cycling of temperature and loading, and potentially non-conservative for in-phase thermal loading. In contrast, code case N-809 gives no guidance on the effective temperature for fatigue crack growth assessments, resulting in maximum temperatures frequently being adopted for assessments of thermal transients. There is therefore a need for a clearer understanding of the impact of variable temperatures during transients on the predicted levels of environmental fatigue.

This paper describes test facilities developed to permit measurements of both thermo-mechanical fatigue life and fatigue crack growth rates in pressurized water reactor environments. Initial test results obtained using these facilities are presented. The fatigue life data have been generated for a range of applied strain amplitudes, 0.45% to 1%, using temperature cycling between 100°C and 300°C. These data, for both in- and out-of-phase temperature and loading, are compared to the predictions of the “weighted Fen” model which is detailed in a separate paper, PVP2017-66030. Similarly, crack growth rate data generated for cycles between 140°C and 280°C are presented and comparisons made against the predictions of the “weighted K rate” (WKR) method detailed in paper PVP2017-65645. In both cases, the test results suggest that the weighted models are able to provide good predictions of an effective temperature to be used in fatigue assessment methods, which offer a significant improvement in the treatment of variable temperatures compared to current assessment practice.
Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs: Fatigue Crack Growth Analysis of Hydrogen Flakes
Authors: Lacroix Valéry , Dulieu Pierre
Published: October 2017

During the 2012 outages at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected, mainly in the lower and upper core shells of the Reactor Pressure Vessels (RPVs). 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 Assessment, 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 has been done on the basis of a specific methodology inspired by the ASME B&PV Code Section XI procedure but adapted to the nature and the number of indications found in the Doel 3 and Tihange 2 RPVs.

As requested by Article IWB-3610(a) of ASME B&PV Code Section XI, one of the parts that have to be addressed through the Flaw Acceptability Assessment is the Fatigue Crack Growth (FCG) Analysis of the flaws in the core shells until the end-of-service lifetime of the RPVs. Due to the large number of flaws in the core shells, a specific methodology has been developed in order not to perform the FCG Analysis of each flaw separately.

The paper describes this simplified approach aiming at distributing the flaws according to their inclination and at defining envelope flaws covering the actual flaws to carry out FCG Analysis. Furthermore, the paper highlights and quantifies the conservatisms of this analysis leading finally to demonstrate that the FCG of hydrogen flakes is not a concern in Doel 3 and Tihange 2 RPVs.
Proposal of a New Subsurface-to-Surface Flaw Transformation Rule for Fatigue Crack Growth Analyses
Authors: Lacroix Valéry , Bouydo Afaf , Katsumata Genshichiro , Li Yinsheng , Hasegawa Kunio
Published: October 2017

When a subsurface flaw is located near the component free surface, the first step consists of characterizing the flaw as surface or subsurface in compliance with subsurface-to-surface flaw proximity rules.

The re-characterization process from subsurface to surface flaw is addressed in all Fitness-for-Service (FFS) Codes. However, the specific criteria for the rules on transforming subsurface flaws to surface flaws are different among the FFS Codes.

This re-characterization concept is essential and important for subsurface flaws in the flaw assessment procedures. It is applied for three stages of the flaw assessment: at service inspection for flaw characterization, at subcritical crack growth calculation, such as fatigue crack growth, and at ductile/brittle fracture assessment. In this frame, fatigue crack growth experiments were recently conducted by the authors and it was highlighted that the subsurface-to-surface transformation is highly sensitive to the aspect ratio a/l of the flaw whereas the proximity factors in the rules are defined by constant values i.e., regardless of the flaw aspect ratios a/l.

The authors have therefore proposed a new subsurface-to-surface flaw proximity rule based on experimental data and equivalent fatigue crack growth rates.

Then, the authors demonstrated through numerous Fatigue Crack Growth (FCG) calculations that the current ASME B&PV Code Section XI surface proximity factor should be updated according to the type of component i.e., piping or vessel.

The paper summarizes all the steps leading to the improvement of the ASME Code Section XI subsurface-to-surface proximity rule, from the fatigue crack growth experiments to the studies of the suitability of the current flaw-to-surface proximity factor.

Furthermore, based on additional fatigue crack growth calculations and more refined investigations, the paper proposes finally a new limit value for the surface proximity factor. As a result, a proposal for modification of the ASME Code Section XI, Appendix C is provided. The paper is used for the technical basis of this proposal.
Development of Japanese High Pressure Vessel Standard HPIS C106 With ASME Section VIII Division 3
Authors: Terada Susumu
Published: October 2017

Many high pressure vessels are used in isostatic pressing, polyethylene process and crystal growth application. The design condition of these high pressure vessels becomes more severe in pressure, temperature and cyclic operation. It was desired that design code for such high pressure vessels be issued enabling more reasonable design than ASME Section VIII Div.1 and Div.2. Against above request, ASME Sec. VIII Div.3 was issued in 1997. While in Japan the subcommittee for high pressure vessels in HPI was started in October 1997 in order to issue the Japanese code for high pressure vessels. At first the background of ASME Div.3 was investigated and then “Rules for Construction of High Pressure Vessels: HPIS C 106” was issued in 2005. That was some differences from ASME Div.3, because we considered that ASME Div.3 should be modified. The author has also been appointed as a member of ASME SG-HPV Committee since 2003. The author has proposed some modification and addition of rules for ASME Div.3 since 2000 and most of them already have been approved and incorporated in ASME Div.3. The background of these modification and addition of rules are shown in this paper.
The Master Curve Fracture Toughness Evaluation of Irradiated Plate Material JRQ Using Miniature-C(T) Specimens
Authors: Yamamoto Masato
Published: October 2017

The Master Curve approach is a powerful tool to evaluate material-specific fracture toughness of ferritic steels, such as RPV steels, using a limited number of specimens. However, preparing sufficient volume of material for the generally used 25.4mm-thickness fracture toughness specimens is difficult for irradiated reactor pressure vessel (RPV) steels of existing surveillance programs. Utilization of miniature specimens, which can be machined from broken halves of standard Charpy specimens, is a possible solution to address this issue. CRIEPI has been working on the development of test technique utilizing a miniature C(T) (Mini-C(T)) specimens, whose dimensions are 4×10×9.6 mm (0.16 inch thickness specimen). The basic applicability of the Mini-C(T) Master Curve approach on un-irradiated materials had been confirmed for the base metals of typical Japanese RPV steels [1]. International round robin tests [2–4] confirmed the reproducibility of fracture toughness data obtained by Mini-C(T) specimens.

Applicability of the Mini-C(T) specimen on neutron irradiated materials is of another important subject to be ensured. In the present study, the European standard plate material JRQ, which is in irradiated state up to the fluence of 1.85×1019 n/cm2 was subjected to the Master Curve evaluation with Mini-C(T) specimens. Two laboratories, A and B, as well as CRIEPI were involved in this study. Both of the laboratories separately and successfully carried out machining, pre-cracking and fracture toughness testing without excessive technical difficulty even though the material was highly irradiated. Consistent reference temperature, To was estimated as 40°C (Lab. A) or 32°C (Lab. B). The difference between the two laboratories was reasonably small. To was also consistent with that by pre-cracked Charpy V-notch specimens (PCCv), or 1-inch thickness C(T) specimens examined in the former IAEA Coordinated Research Program [5]. As a results of the investigation, the relevance of using Mini-C(T) specimens for irradiated JRQ material was demonstrated.
Prediction of Ductile Material Failure by Using Innovative Damage Mechanics Concepts
Authors: Schaffrath Simon , Feldmann Markus , Brinnel Victoria , Münstermann Sebastian , Novokshanov Denis
Published: October 2017

Material failure and plastic instability are currently in ASME and EN pressure vessel and piping standards prevented by applying experience-based safety factors. Due to the few existing experimental test results and the inaccuracy of the incorporated design equations, these safety factors are quite high. In particular, modern high strength steels are penalized because of the small yield-to-tensile ratio. Apart from monotonic loading conditions, an economical design of pressurized components to prevent material failure due to ultra-low-cycle fatigue (ULCF) is a big topic.

Experimental tests are in general very expensive and not suitable for the common design. Therefore, a numerical based prediction of the actual burst pressure and the resistance against strong cyclic loading conditions would be favourable. During the last decades innovative damage mechanics concepts have been developed and successfully validated in the scope of structural and plant engineering. Depending on the size of the component and the applied loading condition, several different damage mechanics models are available.

During the studies presented in this paper, a phenomenological damage mechanics model proposed by Bai and Wierzbicki has been utilized. It has been developed to describe the material failure in the upper shelf of the toughness temperature transition curve in case of monotonic loading conditions. Considering an extension based on the effective strain concept proposed by Ohata and Toyoda, the combined model is also able to predict failure in the range of ULCF.

The presented work aims at validating the accurate numerical prediction of failure by using this damage mechanics model. Validation is performed by comparing numerical and experimental results of a burst test of a pressure vessel and cyclic deformation tests of almost constant pressurized bended pipes. The pressure vessel made of HSLA steel P690Q had a length of 3.0 m, a diameter of 1.2 m and a wall thickness of 50 mm. Tests on bended pipes can be separated in two series. A first series made of steel X60 with a diameter of 16 “ and a wall thickness of 9.5 mm as well as a second series made of steel X65 with a diameter of 8.625 “ and a wall thickness of 5.6 mm.

The comparison of experimental and numerical results shows an acutely satisfying prediction of failure. Both, time point and location of failure coincide well. Mesh size dependency must be considered but the approach seems to be promising for further applications.
Experimental Research on Cyclic Response, Hold Effects and Fatigue of Stainless Steel
Authors: Solin Jussi , Alhainen Jouni , Seppänen Tommi , Karabaki H. Ertugrul , Mayinger Wolfgang
Published: October 2017

Our research on fatigue performance of stainless steel and transferability of laboratory data to nuclear power plant operational conditions continues. The focus is in quantification of time and temperature dependent damage relaxation during holds introduced within strain controlled LCF fatigue tests with niobium stabilized X6CrNiNb1810mod steel. These holds aim to simulate steady state normal operation between fatigue relevant cycles at start-up, shut-down or power changes in PWR primary circuit components, e.g. the pressurizer spray lines and surge line.

Amplified cyclic hardening was observed at strain rates approaching zero at normal operation temperatures (≤325°C). Even more pronounced static hardening is consistently measured during holds in elevated temperatures (≥200°C). Beneficial effects of holds in material endurance were shown five years ago. The latest results suggest another beneficial change in component fatigue performance. In addition to improved material response, de-localization of strain is demonstrated in this paper. Our target is a thermodynamic prediction model for improved assessment of fatigue with normal operation periods. The model should quantify the life extension due to long periods in normal operation at operational temperatures.
Flaw Evaluation Procedure for Cast Austenitic Stainless Steel Materials Using Thermal Aging Models
Authors: Uddin M. F. , Wilkowski G. M. , Pothana S. , Brust F. W.
Published: October 2017

Thermal embrittlement of cast austenitic stainless steels (CASS) can occur at reactor operating temperatures potentially leading to a reduction in their fracture toughness. Some aged CASS materials have the potential to have exceedingly low toughness and also show high toughness variability due to the nature of their microstructure. The experimentally measured JIc values for CASS materials showed a large scatter when plotted against ferrite number (FN) or chrome equivalent number (Creq). Because of their low aged toughness with such a large variability, flaw evaluations of CASS material needs to be done carefully, especially since most US PWR nuclear plants have been given plant-life extensions for 60-year operation, and consideration of further extension to 80 years is underway. However, the ASME Section XI Appendix C flaw acceptance criterion currently does not have a recommended procedure for flaw evaluation for CASS materials with FN ≥ 20, and the Working Group recognizes that the changes might also be needed for CASS with FN less than 20.

In this paper, a flaw evaluation procedure for fully aged CASS materials is presented using JIc values at LWR operating temperatures predicted from several existing thermal-aging toughness degradation models. All available thermal aging models for CASS materials were evaluated which predict fully aged (lower saturated toughness condition) fracture toughness of CASS based on their chemical compositions. A set of 20 experimental test data was analyzed by using all models to find the most accurate thermal aging models. Using the most accurate models, correlations between predicted JIc values and French Creq-Fr and ASTM A800 FN were developed from a database of 274 pipe/elbows in US PWR plants whose chemical compositions were known. Finally, the correlation was used to determine the elastic-plastic fracture correction factor (Z factor) for CASS pipe and fittings as a function of pipe diameter and their chemical compositions from material certification sheet using the Dimensionless-Plastic-Zone-Parameter (DPZP) analysis. The DPZP analysis is a relatively simple curve-fitting procedure through full-scale circumferential surface-cracked pipe tests developed in pipe fracture projects funded by the USNRC, and was checked against a full-scale aged CF8m pipe fracture test. After determining the chemical composition specific Z factor for CASS materials, the flaw evaluation can be performed according to the ASME Section XI Appendix C procedures.
ASME Pressure Design of Class 1 Pipe Bends Using Elastic-Plastic FEA
Authors: Abdelsalam Usama , Glover Andrew , Vijay Dk
Published: October 2017

In a previous paper, it was demonstrated that a tight radius bend with uniform wall thickness equal to the pressure based thickness of the corresponding straight pipe does not meet the linear elastic criteria of NB-3221 Design Loadings. It was also demonstrated that the allowable wall thickness circumferential distribution of the ASME B&PV Code SEC XI Code Case N-597-2 achieves a uniform linear elastic stress intensity over the entire bend that meets the linear elastic criteria of NB-3221. As stated by the ASME B&PV Code under NB-3221, the provisions of NB-3228 may provide relief from certain of the linear elastic stress limits if plastic analysis techniques are applied.

In this paper, the plastic collapse analysis approach of the COG FFSG is adopted to explore the relative safety margin when analyzing various Class 1 tight radius pipe bends’ configurations made of Carbon Steel SA-106 Grade B. In addition to the power law material model, three elastic plastic material models are constructed based on ASME B&PV Code Section II Part D material properties. The pressure-based thickness for the same size straight pipe is uniformly used in modelling the tight radius bends (no increased thickness on the intrados). It is demonstrated that the results from the ASME B&PV based flow stress model are marginal and that the ASME B&PV based bi-linear material model and the limit analysis results are not meeting the collapse pressure criterion. It is also demonstrated that the uniform pressure-based thickness along with the power law material model meets the requirements for the adopted plastic collapse pressure providing a considerable life extension.
Investigation of Burst Pressure in T-Joints With Wall-Thinning by Using FEA
Authors: Yamaguchi Atsushi
Published: October 2017

Boilers and pressure vessels are heavily used in numerous industrial plants, and damaged equipment in the plants is often detected by visual inspection or non-destructive inspection techniques. The most common type of damage is wall thinning due to corrosion under insulation (CUI) or flow-accelerated corrosion (FAC), or both. Any damaged equipment must be repaired or replaced as necessary as soon as possible after damage has been detected. Moreover, optimization of the time required to replace damaged equipment by evaluating the load carrying capacity of boilers and pressure vessels with wall thinning is expected by engineers in the chemical industrial field. In the present study, finite element analysis (FEA) is used to evaluate the load carrying capacity in T-joints with wall thinning. Burst pressure is a measure of the load carrying capacity in T-joints with wall thinning. The T-joints subjected to burst testing are carbon steel tubes for pressure service STPG370 (JIS G3454). The burst pressure is investigated by comparing the results of burst testing with the results of FEA. Moreover, the maximum allowable working pressure (MAWP) of T-joints with wall thinning is calculated, and the safety margin for the burst pressure is investigated. The burst pressure in T-joints with wall thinning can be estimated the safety side using FEA regardless of whether the model is a shell model or a solid model. The MAWP is 2.6 MPa and has a safety margin 7.5 for burst pressure. Moreover, the MAWP is assessed the as a safety side, although the evaluation is too conservative for the burst pressure.
Recapturing Net Positive Suction Head Margins in Boiling Water Reactor Emergency Core Cooling Systems
Authors: Bilanin Alan J. , Kaufman Andrew E. , Bilanin Warren J.
Published: October 2017

Boiling Water Reactor pressure suppression pools have stringent housekeeping requirements, as well as restrictions on amounts and types of insulation and debris that can be present in the containment, to guarantee that suction strainers that allow cooling water to be supplied to the reactor during a Loss of Coolant Accident remain operational. By introducing “good debris” into the cooling water, many of these requirements/restrictions can be relaxed without sacrificing operational readiness of the cooling system.
Solution of a Sample Problem Related to Revision 1 of Code Case N-830
Authors: Kirk Mark , Xu Steven , Lui Cheng , Erickson Marjorie , Kim Yil , Lee Darrell , Shim Do Jun , Stevens Gary
Published: October 2017

Within the American Society of Mechanical Engineers (ASME) the Section XI Working Group on Flaw Evaluation (WGFE) is currently working to develop a revision to Code Case (CC) N-830. CC N-830 permits the direct use of fracture toughness in flaw evaluations as an alternative to the indirect/correlative approaches (RTNDT-based) traditionally used in the ASME Code. The current version of N-830 estimates allowable fracture toughness values in the transition regime as the 5th percentile Master Curve (MC) indexed to the transition temperature T0. The proposed CC N-830 revision expands on this capability by incorporating a complete and self-consistent suite of models that describe completely the temperature dependence, scatter, and interdependencies between all fracture metrics (i.e., KJc, KIa, JIc, J0.1, and J–R) used currently, or useful in, a flaw evaluation for conditions ranging from the lower shelf through the upper shelf. Papers presented in previous ASME Pressure Vessel and Piping (PVP) Conferences since 2014 provide the technical basis for these various toughness models.

This paper contributes to this overall CC N-830 documentation suite by presenting the results of a sample problem run to assess the proposed revision of the CC. The objective of the sample problem was (1) to determine if the revised CC was written with adequate clarity to permit different engineers to accurately and consistently calculate the various allowable toughness values described by the equations in the CC, (2) to assess how these allowable toughness values would be used to calculate allowable flaw depths using standard ASME SC-XI approaches, and (3) to compare allowable flaw depths calculated using established Code practices (RTNDT-based) to those calculated using proposed CC practices (T0-based). The sample problem demonstrated that (1) the CC was written with sufficient clarity to allow different engineers to arrive at the same estimated value of allowable toughness, (2) the latitude associated with the provisions of the ASME Code pertinent to estimation of allowable flaw depth are responsible for some differences in the allowable flaw depth values reported by different participants, and (3) current Code estimates of allowable flaw depth are far more conservative (that is: smaller) than values estimated by the candidate CC methods based on the MC, this mostly due to the generally-conservative bias of the Code’s RTNDT & KIc approach. The candidate CC methods provide much more consistent conservatism than current Code approaches for all conditions in the operating nuclear reactor fleet via their use of an index temperature (T0) defined by actual fracture toughness data and a temperature dependence defined by those data.

The WGFE is continuing to evaluate candidate approaches to estimate allowable toughness values for CC N-830 using a T0-indexed Master Curve. Associated work is addressed by two companion papers presented at this conference.
Surface Stress Analysis of Internally Corroded Pipes Under External Pressure
Authors: Chen Zhanfeng , Shen Xiaoli , Ye Hao , Yan Sunting , Jin Zhijiang
Published: October 2017

Corrosion often leads to the failure of transporting pipelines. The surface stresses on the corroded pipes are related to the failure pressure. In this paper, a double circular arc (DCA) model is developed to calculate the surface stress of the internal corroded pipes under external pressure. In addition, a critical corrosion ratio and a critical thickness-to-diameter ratio are presented to determine the location of the maximum stress. Based on the stress function method and bipolar coordinates, an analytical solution of the DCA model was obtained. And then the stress distributions on the internal and external surfaces of the corroded pipes were determined. Next, the equivalent and hoop stresses at several locations in the cross section of the corroded pipes were discussed. The calculated results were validated using finite element method (FEM). Results show that the maximum stresses vary from the internal surface to the external surface with the increase of the corrosion ratio or the thickness-to-diameter ratio. Our research provides a benchmark for approximate solutions to predict the failure pressure and assess the integrity of the corroded pipelines.
An Experimental Study About Two-Phase Damping Ratio on a Tube Bundle Subjected to Two-Phase Flow
Authors: Kim Key Sun , Sim Woo Gun , Dagdan Banzragch
Published: October 2017

An analytical model was developed by Sim to calculate the two-phase damping ratio for upward two-phase flow perpendicular to horizontal tube bundles. To verify the model, the present experiment is performed with a typical normal square array of cylinders subjected to the two-phase flow of air-water in the tube bundles. The diameter of cylinder is 18mm and the pitch ratio to diameter is 1.35. Using a pressure transducer and data acquisition system, pressure loss along the flow direction in the tube bundles is measured to evaluate the two-phase Euler number and the two-phase friction multiplier. The drag force along the flow direction on a tube is measured to calculate the drag coefficient and the two-phase damping ratio. The experimental results of the two-phase damping ratios are compared with the analytical results given by Sim’s model for homogeneous two-phase flow. It was found that, as increasing the mass flux, the drag force and the drag coefficients given by experimental test are close to the results calculated by the homogeneous model. As a result, the damping ratio can be evaluated by the homogeneous model for bubbly flow of sufficiently large mass flux.
Hydrogen Effect on the Deformation Behavior of Austenitic Stainless Steels Investigated by Nanoindentation
Authors: Zhang Lin , Hong Yuanjian , Zheng Jinyang , An Bai , Zhou Chengshuang
Published: October 2017

A full understanding of hydrogen effect on deformation is important to reveal the mechanism of hydrogen embrittlement. The effects of thermal gaseous hydrogen charging on 304 and 310S austenitic stainless steels have been examined by using nanoindentation. It is first found by using nanoindentation continuous stiffness measurement that hydrogen decreases the elastic modulus and increases the hardness in the initial stage of plastic deformation, while the hydrogen effect becomes weaker and remains nearly constant with further plastic deformation. Hydrogen increases the creep displacement in 310S steel, which indicates that hydrogen facilitates ambient creep. α′ martensite restrains the nanoindentation creep and hydrogen has little effect on the creep induced by α′ martensite.
Initial Thermal Testing of the Pipe Over-Pack Container With a Combustible Waste Matrix for Los Alamos National Laboratory
Authors: Oka Jude M. , Stone Tim A. , Smith Paul H. , Riglin Jacob D. , Caviness Mike L. , Carson Peter H. , Croce Mark P.
Published: October 2017

Los Alamos National Laboratory (LANL) Technical Area 55 (TA 55) utilizes several different container types to store and transfer special nuclear material and waste for numerous programs. The Pipe Over-pack Container (POC) is a vented carbon steel container with a removable lid designed for storing or transferring nuclear waste. Additionally, TA 55 has been tasked to manage and store all of its Transuranic (TRU) waste for the foreseeable future. Being able to place more material into a single container conserves the physical storage space available at TA 55. Conducting thermal testing with a combustible waste matrix in the POC can benefit stakeholders on increasing the material limits of the container. The results of these measurements will establish new wattage limits to meet LANL’s transportation requirements. The current LANL Transportation Safety Document (TSD) limits the amount of heat source plutonium to 10 grams per POC, corresponding to approximately 5 Watts of heat. Combustible loading consisted of an arrangement of cellulosics, plastics, rubber and other contents such as tape. An initial arbitrary limit was set at 60 °C reflecting previous research on nitrate contaminated cheesecloth within a combustible loaded matrix. The limit will ensure the sensitivity of nitrate contaminated contents is undisturbed. Two tests were conducted on the waste matrix at 9.3 watts and 15 watts payloads. Each test was conducted within an 8 hour work day while observing the transient response to ensure the testing does not exceed the temperature limit. Results show that with a 15 watt payload the temperature at the source reaches the 60 °C limit within 5 hours of testing. Other areas such as the cheesecloth and plywood are under the arbitrary limit. Observations and pictures showed no signs of material degradation from heat loading which allow us to move past the limit and investigate higher payloads. ANSYS numerical modeling has complemented the efforts by producing predictions within 5% error of experimental results. These results along with previous POC thermal testing will be presented to the Packaging and Transportation (OS-PT) group at LANL to assist in raising the heat source plutonium loading limit.
Fitness-for-Service Assessment of Steam Generator Tubing Subject to Dealloying Degradation
Authors: Mousavi Rosita , Duan Xinjian , Kozluk Michael , Wang Min , Shi Yihai
Published: October 2017

A new degradation mechanism has been observed in Monel 400 Steam Generator tubing material, a nickel-copper alloy (63Ni-28Cu-2½Fe) with the ASME material designation SB-163/N04400. The location is above the top preheater support plate of the two re-circulating steam generator in one of the units of the Pickering Nuclear Generating Station.

This paper provides a brief description of the regulatory environment, OPG’s steam generator life cycle management plans, the Canadian Industry’s fitness-for-service guidelines for steam generator tubes, and the afflicted steam generators. The paper then goes on to discuss the following activities that were conducted to support the technical basis to justify that the steam generators fit to be returned to service:

• Inspection scope expansion, methods, and results.

• Examination of removed tubes.

• Condition monitoring assessment.

• Operational assessment.

• Burst-pressure tests of removed tubes and of fabricated test specimens.

• Degradation specific flaw model and acceptance standards.

• Flaw growth rate predictions.

• Plugging limit adopted.
Applicability of Miniature C(T) Specimen to Fracture Toughness Evaluation for the Irradiated Japanese Reactor Pressure Vessel Steel
Authors: Sugihara Takuji , Hirota Takatoshi , Sakamoto Hiroyuki , Yoshimoto Kentaro , Tsutsumi Kazuya , Murakami Takeshi
Published: October 2017

Fracture toughness evaluations for irradiated reactor pressure vessel (RPV) steels are essential in the structural integrity assessment of RPVs. In Japanese pressurized water reactor (PWR) plants, fracture toughness tests are conducted for irradiated RPV steels through the surveillance tests and fracture toughness data are obtained. Lately, the Master Curve (MC) approach has become the main stream in fracture toughness evaluation. However, there can be the case that the number of fracture toughness data is not enough for the MC method in some Japanese PWR plants because of limited numbers of fracture toughness specimens contained in the surveillance capsules. On the other hand, for the Japanese PWR plants, a surveillance capsule generally contains a lot of Charpy impact specimens which miniature C(T) (Mini-C(T)) specimens with a size of 4×10×9.6mm can be taken from. Therefore, it is planned that additional fracture toughness tests are performed using Mini-C(T) specimens after the Charpy impact tests to obtain sufficient fracture toughness data for the MC method.

Applicability of the Mini-C(T) specimen to the MC evaluation has been studied in a series of international round robin test programs coordinated by Central Research Institute of Electric Power Industry (CRIEPI). In these programs and the related studies, it was demonstrated that the reference temperature (To) can be determined by the Mini-C(T) specimens without any specific difficulties for the unirradiated RPV base metals. In addition, CRIEPI has recently reported on the basis of their studies that the fracture toughness tests could be successfully performed on the typical unirradiated RPV weld metal and the valid To can be determined with the data obtained from the weld metal as well as base metals. However, few papers reported applicability of the Mini-C(T) specimen to the MC evaluation for irradiated RPV steels.

In this study, fracture toughness tests using Mini-C(T) specimens were conducted on the irradiated Japanese Industrial Standards (JIS) SFVQ1A steel (equivalent to ASME A508 Cl.3 steel). The Mini-C(T) specimens were machined out from some broken halves of Charpy impact specimens used in a surveillance test of an actual Japanese PWR plant by a wire cut electric spark machine followed by fatigue precracking. After the fracture toughness tests, the evaluation was performed on the obtained fracture toughness data according to the MC method. The effect of specimen size on To was studied and applicability of the Mini-C(T) specimen was discussed by comparing the existing results of fracture toughness tests using the 1/2T-C(T) specimens conducted in the surveillance test. In addition, the issues to obtain valid To for irradiated materials were discussed.
Elastic-Plastic Fracture Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shock
Authors: Guo Huajing , Wang Zhongxian , Lam Poh-Sang
Published: October 2017

Three-dimensional finite element models are used to analyze a reactor pressure vessel with an axial semi-elliptical surface crack subjected to pressurized thermal shock. During the thermal shock event, the J-A2 two-parameter fracture theory is used to investigate the temperature-dependent constraint effect at the deepest point and the surface point of the crack. Using the R6 methodology, a series of constraint-based crack failure assessment curves during the thermal shock can be obtained. It was found that the crack tip constraint should be considered for developing a more realistic failure criterion.