bluuebunny/crossref_metadata_2025_split huggingface dataset:
Use of Average Temperature in Fen Calculations
Authors: Weitze W. F. 
Published: December 2020

Abstract
The effect of the light reactor water environment on fatigue damage is referred to as environmentally assisted fatigue (EAF). This effect is accounted for by applying an environmental fatigue correction factor, Fen, to calculated fatigue usage. In providing guidelines for calculation of Fen, Revision 0 of NUREG/CR-6909 [1] permits temperature averaging for the case of a constant strain rate and linear temperature response, and permits it in other cases as well, but only if the average temperature used produces results that are consistent with the modified rate approach [1, p. A.5]. Revision 1 of NUREG/CR-6909 [2] modifies this slightly, requiring that the threshold temperature be used in averaging instead of the minimum if the minimum is below the threshold [2, p. A-6]. In both cases, the benchmark for accuracy is the modified rate approach [2, Section 4.4].
In this paper, we use real world examples to compare Fen values based on the modified rate approach with those using average strain rate and temperature. We also examine how to select the rise time used to calculate average strain rate in those cases where it is not obvious. We find that temperature averaging is conservative if rise time and other parameters are correctly selected.
Assessment of the Continued Need for Independent Requirements on Transition Temperature and Upper Shelf Charpy Impact Toughness Metrics in Regulations Pertaining to Nuclear Reactor Pressure Vessels
Authors: Kirk Mark , Yamamoto Masato , Erickson Marjorie 
Published: December 2020

Abstract
The toughness requirements for the ferritic steels used to construct the primary pressure boundary of a nuclear power plant include both transition temperature metrics as well as upper-shelf metrics. These separate specifications for transition and upper shelf toughness find their origins in decisions made during the 1970s and 1980s, a time when there was much less empirical and theoretical knowledge concerning the relationship between these quantities. Currently, significant evidence exists to demonstrate a systematic relationship between transition and upper shelf toughness metrics for RPV-grade steels and weldments (e.g., the equations in draft Code Case N-830-1, empirical correlation between Charpy transition temperature and upper shelf metrics, etc.). This paper explores these relationships and demonstrates that, in many cases, the joint specification of transition temperature and upper shelf toughness values is redundant and, therefore, unnecessary.
Methods to Appropriately Account for Uncertainties in Structural Integrity Assessment Models
Authors: Erickson Marjorie , Kirk Mark 
Published: December 2020

Abstract
To ensure an appropriate and/or conservative assessment of structural integrity it is essential to account for the uncertainties inherent to the various inputs and models that, collectively, contribute to a structural integrity assessment. While the methods used to account for uncertainties will differ, this applies equally to assessments performed using either deterministic or probabilistic approaches. Oftentimes the overall model used for a structural integrity assessment is itself comprised of multiple inputs and models, which themselves may be inter-related and/or correlated. In these circumstances the quest to ensure that all uncertainties are addressed can result in the same uncertainty — or uncertainty source — being accounted for multiple times. Such “double-counting” of uncertainties introduces un-needed conservatism to the assessment and should be avoided.
In this paper we use the linked fracture toughness models contained in the recently proposed Revision 1 to ASME Section XI Code Case N-830 to provide examples of uncertainty treatment in analyses using multiple models. Identification of sources of uncertainty in each model used in a multi-model analysis can help to ensure that each source is accounted for appropriately and not multiple times. The CC N-830-1 models are used to demonstrate the effects of various uncertainty treatment strategies and the pitfalls that arise from treating sources of uncertainty twice.
Evaluation of Nozzle P-T Limit Curves for Korea Optimized Power Reactor
Authors: Lee Hyunchul , Yoo Choonsung , Maeng Youngjae , Yoo Sunghoon 
Published: December 2020

Abstract
The purpose of the Pressure – Temperature (P-T) limit curve is to prevent a failure of reactor pressure vessel during operation of reactor coolant system. In Korea, P-T limit curves have to meet 10CFR50 Appendix G [1] according to Nuclear Safety and Security Commission Notification 2017-20. The P-T limit curves have been traditionally evaluated based on the beltline region which is most affected by neutron irradiation. However due to the geometric discontinuity, the inside corner regions of the vessel nozzles are the most highly stressed regions of reactor vessel. These higher stresses can potentially result in more restrictive P-T limits, even if the reference nil ductility transition temperatures (RTNDT) for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries. In 2014, the NRC issued Regulatory Issue Summary (RIS) 2014-11 [2], which require the consideration of reactor vessel nozzles in P-T limits curve generation.
In this paper, P-T limit curves for Korea optimized power reactor (OPR-1000) nozzles at the end of license were evaluated. And then these curves were compared to the traditional beltline region P-T limit curves in order to verify which curve is more limiting.
Treatment of the Interaction With the Free Surface of the Component for Combined Subsurface Flaws: Technical Basis for Revision of IWA-3300 and Table IWB/IWC-3510-1
Authors: Lacroix Valéry , Dulieu Pierre , Hasegawa Kunio , Mares Vratislav 
Published: December 2020

Abstract
When flaws are detected in pressure retaining components, assessments have to be done in order to demonstrate the fitness-for-service (FFS) of the component for continued operation. This FFS demonstration is performed in accordance with FFS Codes providing flaw assessment procedures and acceptance standards. Before performing analyses, a flaw characterization has to be carried out in order to determine unequivocally the flaw geometry. This flaw characterization is done according to rules provided in the FFS Codes and hence appears as crucial for the rest of the flaw assessment.
The first step of the flaw characterization addresses the interaction of the flaw and the free surface of the component: if a subsurface flaw is located near the free surface, this step consists of characterizing the flaw as surface or subsurface according to subsurface-to-surface flaw proximity rules. The recharacterization process from subsurface to surface flaw is addressed in all fitness-for-service (FFS) Codes. The second step of the flaw characterization addresses the interaction of the flaw with adjacent flaws: if a flaw is located near another flaw, this step consists of combining the flaws between them according to flaw proximity rules.
However, in some FFS Codes and in the ASME B&PV Section XI Code particularly, there is a lack on how to treat the interaction of a combined flaw and the free surface of the component. The ASME B&PV Section XI Code flaw characterization is not clear on this topic which could lead to misinterpretations and unreliable flaw assessment results.
Some typical examples of unrealistic flaw assessment results due to these misinterpretations of the ASME B&PV Code Section XI flaw characterization rules are depicted in this paper. After analyzing more in-depth the origin of the inconsistencies based on 3D Extended Finite Element Method (XFEM) calculations, the paper is used as Technical Basis for the improvement of the ASME B&PV Code Section XI in order to clarify the treatment of combined flaw in the flaw characterization (IWA-3300) and in the flaw acceptability assessment as well (IWB/IWC-3510-1).
A Comparison of Stress Intensity Factor Solutions for Surface Cracked Plates
Authors: Aird C. J. 
Published: December 2020

Abstract
The present paper compares the stress intensity factor (SIF) solutions for surface-cracked plates presented in R6, BS 7910, API 579-1/ASME FFS-1 and RSE-M. Both extended surface cracks and semi-elliptical surface cracks are considered. In general, the agreement between the various solutions is found to be good at within 5%. However, particular cases are highlighted which are associated with larger differences between the SIFs. For extended surface cracks, the various solutions agree closely at within 5% in all but one case. For semi-elliptical surface cracks, the agreement between the various solutions is generally better at the deepest point than at the so-called ‘surface’ points. In particular, the surface point SIFs corresponding to the Vainshtok and Varfolomeyev weight function solution which appears in API-579-1/ASME FFS-1 differ significantly from the other solutions in some cases. It is concluded that further work is required in order to establish an appropriate definition of the surface point SIF for use in structural integrity calculations.
Development of Design Method for High Temperature Nuclear Reactor Cladded Components
Authors: Barua B. , Messner M. C. , Jetter R. I. , Sham T.-L. 
Published: December 2020

Abstract
High temperature nuclear reactors plan to use highly corrosive coolant such as molten salts, molten lead, and lead-bismuth eutectic mixtures. The existing Class A metallic materials qualified in the ASME Section III, Division 5 rules for high temperature nuclear reactors are not ideal for resisting corrosion when exposed to these coolants. One option to overcome this limitation would be to Code-qualify new corrosion-resistant materials for Class A service, however this process is long and expensive and requires long-term creep test data. A near-term alternative would be to allow designers to clad the existing Class A base materials with non-qualified corrosion-resistant materials. However, there are currently no ASME design rules for cladded components to guard against creepfatigue failure and ratcheting strain accumulation in elevated temperature nuclear service. This work addresses this deficiency by proposing a design strategy for cladded components that does not require long-term testing of clad materials. The proposed approach relies on approximate design analysis methods for two types of clad materials — soft clad that creeps faster than the base material and hard clad that creeps slower and has higher yield stress than the base material. The proposed approach treats a soft clad material as perfectly compliant and a hard clad material as linear elastic. Sample finite element analyses of representative high temperature reactor components are performed to verify the approach. At the end, a complete set of design rules is provided for each of the two types of cladded components.
Application of Environmentally Assisted Fatigue Transient Ramp Times Optimization in Design Evaluations and Fatigue Monitoring Algorithms
Authors: Matthews Eric , Gray Mark 
Published: December 2020

Abstract
As part of the effort for a nuclear plant to undergo license renewal, the effect of reactor water environment on fatigue life must be addressed for limiting component locations. One method to incorporate the effects of reactor water environment into the fatigue evaluations of metal components is to apply an environmental fatigue penalty factor (Fen) to the partial usage factor obtained from the design fatigue curve for each stress cycle. Fatigue evaluations have historically been performed by assuming that temperature transient loads occur at conservatively high rates to maximize the stress response and corresponding fatigue usage values. However, with consideration of reactor water environmental effects on fatigue, transients with slower rates generally produce higher Fen values that could potentially result in higher environmental fatigue usage values than transients with identical temperature changes but faster rates. A generic parametric study was performed in MRP-218 to characterize limiting transient ramp rates with respect to environmental fatigue usage for a range of piping geometry and material configurations. This paper describes the application of the parametric study results to optimize thermal hydraulic and stress response modeling assumptions with respect to transient rates and downstream effects on environmental fatigue results in both design and monitoring fatigue evaluations.
Local Failure Analysis of Bolted Flanges
Authors: Li Qi , Sulwinski Rafal , Boellstorff Charles 
Published: December 2020

Abstract
Protection against local failure is one of the integral components in the design-by-analysis requirements in ASME BPVC Section VIII, Division 2. Of the methods offered by the ASME, the Local Strain Limit procedure outlined in 5.3.3.1 is the typical calculation method. However, it has been found that relying on this procedure alone can lead to untenable utilization results if used on certain analyses with varied load paths.
The flange described in this study was calculated using “design by analysis” according to Part 5 of ASME BPVC Section VIII, Division 2. The elastic-plastic stress analysis method was used. The flange was loaded with an initial bolt pre-tension and then with internal pressure. During the local failure calculation, an abnormal condition was encountered in the form of a large spike in the history curve of the ratio between plastic strain and limiting triaxial strain.
An investigation found that despite being in a stress state below yield stress, some nodes had a non-zero plastic strain and high triaxiality factor. This was caused by the load sequence: first, the bolt pre-tension and then internal pressure. The flange was first bent due to the pre-tension load, and later experienced bending in the opposite direction after the internal pressure load was applied. This resulted in a relatively low stress state with a high triaxiality factor and non-zero plastic strain in certain areas, which then showed high utilization under the local failure strain limit criterion.
This paper will discuss how this issue can be avoided by using the strain limit damage calculation procedure 5.3.3.2 outlined in ASME BPVC Section VIII, Division 2.
Codes and Standards for Managing Degradation of Boilers in Service
Authors: Delle Site Corrado , Artenio Emanuele , Sepede Gennaro , Chini Matteo , Giacobbe Francesco 
Published: December 2020

Abstract
Degradation of pressure equipment is becoming an important issue due to increasing asset service time in process and power plants across Europe. For this reason it is important to assess life consumption of these assets to avoid catastrophic failures. Therefore it is necessary to refer to national/international normative on this subject. At present time the Italian thermotechnical committee (CTI) has drawn up a comprehensive set of norms which help the user to set up an inspection plan to investigate and assess degradation of pressure vessels and boilers. In the first part of this paper creep damage of Steam Generators is analyzed. For this purpose results of INAIL (Istituto Nazionale per l’ Assicurazione contro gli Infortuni sul Lavoro) database of steam boilers with 100’000 service hours or more is illustrated. Critical components are identified with reference to materials, geometry and operating parameters (pressure, temperature and time). At the end of the design life cycle, components of pressure equipment operated in creep regime must subjected to specific checks to estimate their residual life and the suitability for further use in safety conditions. The procedure allows to define reinspection intervals keeping acceptable the risk associated with the further use of the component related to creep even in evidence of defects in progress. The first check must be performed after 100,000 hours of effective use. Then, residual life evaluations must be repeated according to period of time that are defined as function of the results of all the checks carried out. In the second part of this paper boiler degradation is discussed with reference to NDT results and in-field inspection campaigns which are carried out traditionally after 45 years of service time, to minimize the risk of pressure components failures. In this paper results of different case studies are discussed with reference to degradation mechanisms and applicable standards.
Overview of EASICS Validation Experiments and Code Comparison of R5, RCC-MRX and ASME III Division V
Authors: James Peter , Coon David , Austin Colin , Underwood Nicholas , Chevalier Marc , Dean David 
Published: December 2020

Abstract
The “Establishing AMR Structural Integrity Codes and Standards for UK GDA” (EASICS) project has recently been established to help support the acceptance of Advanced Modular Reactors, or AMRs, which are typically based on high temperature Generation IV reactors. The EASICS project is looking to help define the requirements for codes and standards for the design of AMRs, to ensure that state-of-the art knowledge will be brought to bear on developing the required design and assessment methodologies. The EASICS project is to run from July 2019 to March 2021.
To support this, the work presented in this paper provides an overview of two interacting aspects of this programme. The first is to perform validation tests for high temperature creep-fatigue assessments of a plant relevant component. The second aspect is to use these results, along with those of other identified tests, to provide a comparison of internationally recognised approaches for the assessment of high temperature (creep regime) components. This paper provides an initial overview of the works and progress; the final results will be presented in subsequent paper(s).
Specialist testing of a plant relevant component (thin tubes butt welded together) under high temperature loading conditions will be performed. The validation testing is to include two fatigue tests and four creep-fatigue tests on 316H welded tubes. The tubes will be subject to strain controlled tension/compression (R-ratio of −1), with the possibility to apply a displacement controlled dwell. The tests will be conducted at 525 C.
To help enhance interaction with the code bodies, and to understand the impact of differences in the approaches, comparative assessments will be performed when adopting R5, ASME Section III Div 5 and RCC-MRx. One comparison will be based around the tests detailed here. The second assessment comparison will consider the Evasion mock-up tests provided by CEA (sodium based thermal shock tests). The third assessment case currently still needs to be defined in the programme at the time of writing. The subsequent discussions over results, differences and potential impact to the codes will help inform a guidance document to best assess AMRs in the UK.
Probabilistic Leak Before Break Using the R6 Methodology
Authors: Gill Peter 
Published: December 2020

Abstract
The integrity of a component in a safety critical industry is determined by carrying out Engineering Critical Assessments (ECA). These are designed to provide a conservative estimate of the life of a component based on conservative inputs/methodology. It is becoming increasingly apparent that for many applications these methods are overly conservative. The only physical way to really assess the reliability of a component is by producing many thousands, if not millions of a specific component and calculating a failure probability based on testing/OPEX. This is simply not feasible for the components in, for example, a nuclear reactor, and probabilistic techniques are becoming increasingly important as a means to understand the reliability of a component. This information can then be used to assess risk and inform inspection programmes.
Typically a probabilistic method relies on assigning distributions to various input parameters and evaluating a probability integral, usually by Monte-Carlo analysis. A previous PVP paper developed Monte Carlo methods using the R6 fracture mechanics procedure. Although providing good insight into the likelihood of failure, these analyses were simplified and not readily applied to realistic plant situations. Further development would enable much more of the technology contained within R6 to be applied within probabilistic software. The following new features of the software are presented in this paper:
• the latest K and limit load solutions from R6 for through wall circumferential defects
• Simplified V factor approach to account for secondary stresses
• two phase flow (water) based on the latest SQUIRT methodology
• global bending, through wall bending, weld residual stress
This enables a full probabilistic leak detection calculation for circumferential through wall cracks in pipes. Examples of probabilistic Leak-before-Break calculations for PWR pipework are presented in the paper.
How congenital Zika virus impacted my child’s functioning and disability: a Brazilian qualitative study guided by the ICF
Authors: Campos Taynah Neri Correia , Schiariti Veronica , Gladstone Melissa , Melo Adriana , Tavares Jousilene Sales , Magalhães Adriana Gomes , Longo Egmar 
Published: December 2020

IntroductionThe Zika virus outbreak in Brazil has had devasting social, medical and financial consequences for families. Both researchers and clinicians are measuring longer-term outcomes to understand the impact of the Zika on child development, functioning and disability. Outcomes and tools used to measure them are very varied and we are unclear how meaningful they are to families and children. This study aimed to identify the parents’ perspectives on relevant areas of functioning and disability that should be included as outcome measures for children with congenital Zika syndrome (CZS), as guided by the International Classification of Functioning, Disability and Health (ICF).MethodsThis qualitative study included parents or caregivers of children aged 0–5 years with confirmed CZS from two states in northeastern Brazil. Interviews were conducted using focus groups. Content mapping followed the WHO’s ICF linking rules. Three raters analysed the content using NVivo V.11.ResultsThirty-two caregivers participated in six focus groups, 88% were mothers with an average age of 30 years. Most children were male (59%) and all were level V (severe) to on the Gross Motor Function Classification System (GMFCS). Overall, 825 themes were mapped to 36 ICF categories. Although parents mentioned areas across all ICF domains, they reported that areas of mobility, eating and recreation were most relevant for them. In addition, environmental factors were highly identified as barriers, specifically services, policies and access to assistive devices. The most predominant facilitators within the environment were; immediate family support, kind relationships with therapists and support from the extended family.ConclusionsAlthough parents emphasised issues related to mobility, their greatest concerns involved environmental factors, such as access and quality of health and social services, systems and policies. These results reinforce the importance of including parents’ perspectives when selecting or developing outcome measures for CZS.