bluuebunny/crossref_metadata_2025_split huggingface dataset:
Essential Elements of an Asset Integrity Management Program for Ammonia and Methanol Plants
Authors: Jaske Carl E. , Weichel Steven J. , Brongers Michiel P. H. 
Published: October 2017

World-class ammonia and methanol plants typically produce more than 500,000 metric tons of ammonia or methanol per year. These plants utilize pressure vessels, piping, and tanks that operate over a wide range of temperatures and pressures. The materials of construction range from carbon steel to corrosion-resistant and heat-resistant alloys. Ensuring the safe and reliable operation of these facilities requires an effective asset integrity management program. This paper reviews the essential elements of an asset integrity management program and provides recommendations for judging the effectiveness of the program. The essential elements of an asset integrity management program include leadership, risk management, personnel and contractor competence, management of change, learning from events, emergency response, and implementation of quality assurance, maintenance, inspection, fitness-for-service assessment, repair, and replacement.
Management commitment to the program is a key factor in leadership. Risk is managed by mitigating the consequences of an incident as well as minimizing its likelihood; a robust risk-based inspection (RBI) program is typically part of the risk management. In-service degradation mechanisms of the materials that are used in pressure vessels, piping, and tanks include corrosion, fatigue, creep, and metallurgical embrittlement. If defects are identified by inspection, fitness for service assessment is performed to determine what action is to be taken. Training and certification of personnel and contractors is required to make sure that this work is properly performed. Incidents and near misses that occur in the plant and in the industry need to be reviewed to identify areas for potential program improvements. Timely and appropriate emergency response can minimize the consequences of an incident.
Sensitivity Analysis for XLPR Acceptance Testing
Authors: Sallaberry Cédric J. , Kurth Robert E. , Brust Frederick W. , Kurth Elizabeth A. 
Published: October 2017

The US Nuclear Regulatory Commission (US NRC) and the Electric Power Research Institute (EPRI) have jointly funded the development of the probabilistic analysis tool xLPR (extremely Low Probability of Rupture) version 2.0 for calculating the probability of a leak before break for welded pipes [1]. The acceptance testing for the program was completed in July of 2016. This allowed the release of the code to the development team. The released version of xLPR v 2.0 was used to run probabilistic simulations on selected scenarios and then perform uncertainty and sensitivity analysis on selected results. Uncertainty Analysis (UA) informs on how much the uncertainty affect the outputs while Sensitivity Analysis (SA) ranks the uncertain parameters in term of importance (i.e., influence over uncertainty) for selected outputs.
In this paper, we present the methodology used to perform Sensitivity Analysis using three regression techniques as well as scatterplots for graphical study. The results generated are consistent both with the experts’ expectations and our understanding of the equations used in the sub-models.
A Study for Evaluating Local Damage to Reinforced Concrete Panels Subjected to Oblique Impact of Deformable Projectile
Authors: Nishida Akemi , Ohta Yoshimi , Tsubota Haruji , Li Yinsheng 
Published: October 2017

Many empirical formulas have been proposed for evaluating local damage to reinforced concrete structures caused by impacts of rigid projectiles. Most of these formulas have been derived based on impact tests perpendicular to the target structures. Up to now, few impact tests oblique to the target structures have been carried out. This study has been conducted with the purpose of proposing a new formula for evaluating the local damage caused by oblique impacts based on previous experimental and simulation results.
In this paper, the results of simulation analyses for evaluating the local damage to reinforced concrete panels subjected to normal and oblique impacts by deformable projectile, by using the simulation method that the validity was confirmed through simulation analysis of past impact experimental results. Based on the results of these simulation analyses, the quantitative evaluation of reduction effects in the local damage caused by the difference of the impact angle is investigated.
Fatigue With Hold Times Simulating NPP Normal Operation Results for Stainless Steel Grades 304L and 347
Authors: Karabaki H. Ertugrul , Solin Jussi , Twite Marius , Herbst Matthias , Mann Jonathan , Burke Grace 
Published: October 2017

The cyclic behavior and endurance of austenitic stainless steels tested under NPP-relevant laboratory conditions has been studied. It had been earlier shown that long intervals between fatigue transients can affect the fatigue performance in stainless steels that are generally used in NPP primary piping. If this can be confirmed, then the transferability between laboratory results, design curves and the fatigue behavior of NPP components during plant operation shall be addressed. In addition to coolant water environmental effects, the material response during steady state normal operation should also be accounted for.
Advanced Fatigue Methodologies (AdFaM), a joint project of European research laboratories, vendors and plant operators was focused on empirical and mechanistic investigations to confirm the claimed effects of hold times on fatigue life.
Strain-controlled fatigue tests incorporating accelerated hold times at temperatures between 290°C and 420°C were performed on stabilized and non-stabilized stainless steel grades, which are used in Germany and the UK. Two material batches of both alloy types (304L and 347) were studied. The mechanisms responsible for the observed variations in stress response and fatigue life have been investigated using a range of microscopy techniques.
The results confirmed the extension of fatigue life due to hold times in both stabilized and non-stabilized grades. This life extension appears to be linked to hold hardening observed in the cyclic behavior of both alloys.
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. This gives reason to consider the transferability of standard laboratory test data to fatigue assessments of NPP components, and to propose a new factor ( Fhold ) as part of an advanced fatigue methodology and realistic transferability factor: Freal = Fen × Fhold.
Technical Basis for Flaw Acceptance Criteria for Cast Austenitic Stainless Steel Piping
Authors: Shim D. J. , Cofie N. G. , Dedhia D. , Harris D. O. , Griesbach T. J. , Amberge K. 
Published: October 2017

According to the current ASME Code Section XI, IWB-3640 and Appendix C flaw evaluation procedure, cast austenitic stainless steel (CASS) piping with ferrite content less than 20% is treated as wrought stainless steel. For CASS piping with ferrite content equal or greater than 20%, there is currently no flaw evaluation procedure in the ASME Code.
In this paper, the technical basis for a proposed flaw acceptance criteria for CASS piping is presented. The procedure utilizes the current rules in ASME Code Section XI, IWB 3640/Appendix C and the existing elastic-plastic correction factors (i.e., Z-factors) for other materials in the Code. The appropriate Z-factor to use for the CASS piping is determined based on the ferrite content (using Hull’s equivalent factor). Experimentally measured fully saturated fracture toughness and tensile data of the three most common grades of CASS material in the US (CF3, CF8 and CF8M) were used to determine the flaw acceptance criteria in the proposed method. The proposed method is conservative since it utilizes the fully saturated condition of CASS materials. In addition, it is simple and consistent with current regulatory guidance on aging management of CASS piping.
Probabilistic Assessment of CANDU Reactor Core for Risk of Pressure Tube Failure due to Presence of In-Service Flaws
Authors: Cho David , Mok Danny H. B. , Xu Steven X. , Scarth Douglas A. 
Published: October 2017

Technical requirements for analytical evaluation of in-service Zr-2.5Nb pressure tubes in CANDU(1) reactors are provided in the Canadian Standards Associate (CSA) N285.8. The evaluation must address all in-service degradation mechanisms including the presence of in-service flaws. Flaws found during in-service inspection of CANDU Zr-2.5Nb pressure tubes, including fuel bundle scratches, debris fretting flaws, fuel bundle bearing pad fretting flaws, dummy bundle bearing pad fretting flaws, erosion-shot flaws and crevice corrosion flaws, are volumetric and blunt in nature. These in-service flaws can become crack initiation sites during pressure tube operation and potentially lead to pressure tube failure. Any detected flaws that do not satisfy the criteria of acceptance as per CSA N285.4 must be analytically evaluated to justify continued operation of the pressure tube. Moreover, the risk of pressure tube failure due to presence of in-service flaws in the entire reactor core must be assessed.
A review of assessment of the risk of pressure tube failure due to presence of in-service flaws in CANDU reactor core is provided in this paper. The review covers the technical requirements in the CSA N285.8 for evaluating degradation mechanisms related to flaws in the reactor core. Current Canadian industry practice of probabilistic assessment of reactor core for pressure tube failure due to presence of in-service flaws is described, including evaluation of flaws for crack initiation, subsequent crack growth to through-wall penetration, and pressure tube rupture due to unstable crack growth prior to safe shutdown of the reactor. Operating experience with the application of probabilistic assessment of reactor core for the risk of pressure tube failure due to presence of in-service flaws is also provided.
Comparison of Methods for Structural Stress Determination According to EN 13445-3 Annex NA
Authors: Trieglaff Ralf , Beckert Martin , Rudolph Jürgen , Hauser Fabian 
Published: October 2017

The European Pressure Vessel Standard EN 13445 provides in its part 3 (Design) a simplified method (Clause 17) and a detailed method for fatigue assessment (Clause 18). Clause 18 “Detailed Assessment of Fatigue Life” is under revision within the framework of the European working group CEN/TC 54/WG 53 - Design methods to reach a significant increase in user-friendliness and a clear guideline for the application.
This paper is focused on the new informative annex NA “Instructions for structural stress oriented finite elements analyses using brick and shell elements”. As an essential amendment for the practical user the determination of structural stress ranges for fatigue assessment of welds is further specified in this new annex. Different application methods for the determination of structural stresses are explained in connection with the requirements for finite element models and analyses.
This paper will give an overview of the proposed approaches of structural stress determination in annex NA of the revised EN 13445-3. It will present the results of fatigue analyses according to EN 13445-3 Clause 18 from a round robin test with participants of the German working group for different application examples. The estimated structural stress ranges for the proposed approaches of structural stress determination will be compared. This comparison leads to general remarks and open points for the different approaches.
Furthermore it is concludes that further comparisons, e.g. with results of other pressure vessel design codes are desirable.
Decay Length in Pressure Vessels
Authors: Adibi-Asl R. , Seshadri R. 
Published: October 2017

In theory of shell, decay length is defined as the distance affected by localized external (applied loads) or internal (edge effect, discontinuity) forces and moments, beyond which the effect of these loads becomes negligible. The concept of decay length becomes relevant when assessing the interaction effects caused by adjacent discontinuities.
In this paper, the decay lengths for several shells geometries, specifically cylindrical, spherical and conical shells, are reviewed. The available expressions for decay lengths in the literature are listed and are compared with finite element analysis to demonstrate the accuracy of some of the solutions.
Determination of the Gasket Stress Distribution
Authors: Schaaf Manfred 
Published: October 2017

The European Standard EN 1591-1 is used for the calculation of bolted flanged joints, stress analysis as well as for tightness proofs. In this calculation procedure gasket characteristics according to EN 13555 are used to describe the mechanical and the tightening behavior of gasket materials.
With further developments in the calculation algorithm and the use of the realistic gasket behavior in the calculation more detailed results can be obtained, which are comparable to results obtained from Finite Element Analysis. The flange rotation and the resulting uneven gasket stress distribution in the radial direction during the assembly of the flanged joint is the fundamental principle in this development. The effective compressed gasket width has influence on the required gasket forces for the tightness proof as well as on the mechanical behavior of the flanged joint, and thus also on the stress analysis.
In this paper, the determination of the effective gasket width using a newly developed approach [1] is optimized and the verification of this approach with Finite Element Analysis for several different gasket materials and flange geometries is shown.
Dealloying of As-Welded Microstructures in Aluminum Bronzes in Essential Cooling Water Service
Authors: Kirchhofer Rita , Vaillancourt Henry , Garner Michael , Cipolla Russell C. 
Published: October 2017

Dealloying of aluminum-bronze (Al-Brz) is a potential environmentally-induced degradation mechanism in piping systems and components exposed to marine, brackish, or raw water service. Examples of this form of degradation have been observed in ASME Code Class 3 Essential Cooling Water systems. The degradation mechanism is loss of material due to selective leaching (i.e., dealuminification) due to localized corrosion. The dealloying occurs in the eutectoid phase in binary alloys; the eutectoid consists of α + β or α + β/γ2, and is dependent on aluminum content of the alloy, cooling rate during fabrication/welding, and the presence of a local crevice for the corrosion process to concentrate. Cast components are susceptible to general dealloying but unlike them, weldments of wrought pipe have not exhibited the same extent of degradation. The dealloying process leaves micro-voids in the microstructure within the eutectoid. Leakage (weepage) will occur once dealloying has extended through the wall thickness via a network of porous transformed phase. The dealloying process also causes a reduction in mechanical strength and fracture toughness properties.
A detailed metallurgical evaluation has been completed to establish a fundamental understanding of the resistance of the pipe welds to dealloying that explains the better performance of the welds versus the castings. Factors affecting the dealloying susceptibility, such as aluminum and iron content, cooling rates during the solidification process, and the resulting microstructural differences were studied and evaluated. It was found that dealloying will not penetrate into the material unless a continuous network of susceptible microstructure is present.
The difference in susceptibility between castings and weldments to dealloying was also determined. The findings of this comparative evaluation were significant and are being used in the aging management program to justify continued use of aluminum bronze piping welds for extended service for both above ground and below ground (buried) piping systems.
Nuclear Power Plant Fires and Explosions: Part II — Hydrogen Ignition Overview
Authors: Leishear Robert A. 
Published: October 2017

Major accidents that were affected by hydrogen fires and explosions included Chernobyl, Three Mile Island, and Fukushima Daiichi. Smaller piping explosions have occurred at Hamaoka and Brunsbüttel Nuclear Power Plants. An overview of pertinent topics is presented here to compare similarities and differences between these accidents.
In particular, a hydrogen ignition mechanism is presented here, where fluid transients, or water hammer, may cause pressures to compress flammable hydrogen gas in reactor systems. As the gas compresses, it heats to temperatures sufficient to cause autoignition, or dieseling. Autoignition then leads to fires or explosions in nuclear power plant systems.
To explain this evolving theory on hydrogen ignition during fires and explosions, various nuclear power plant hydrogen accidents require discussion. For example, Chernobyl explosions were unaffected by water hammer, while a Three Mile Island hydrogen fire was a direct result of water hammer following a reactor meltdown, and explosions that followed a meltdown at Fukushima Daiichi occurred during a water hammer event. Other piping damages also occurred during water hammer events. The primary purpose of this paper is to serve as a literature review of past accidents and to provide new insights into those accidents.
In short, what is known versus what is unknown is discussed here with respect to the ignition sources of nuclear power plant fires and explosions. How can nuclear power plant safety be assured unless previous fire and explosion causes are understood? Prior to this work, they were not understood.
Nuclear Power Plant Fires and Explosions: Part I — Plant Designs and Hydrogen Ignition
Authors: Leishear Robert A. 
Published: October 2017

Requiring further investigation, hydrogen explosions and fires have occurred in several operating nuclear reactor power plants. Major accidents that were affected by hydrogen fires and explosions included Chernobyl, Three Mile Island, and Fukushima Daiichi. Smaller piping explosions have occurred at Hamaoka and Brunsbüttel Nuclear Power Plants. This paper is the first paper in a series of publications to discuss this issue. In particular, the different types of reactors that have a history of fires and explosions are discussed here, along with a discussion of hydrogen generation in commercial reactors, which provides the fuel for fires and explosions in nuclear power plants.
Overall, this paper is a review of pertinent information on reactor designs that is of particular importance to this multi-part discussion of hydrogen fires and explosions. Without a review of reactor designs and hydrogen generation, the ensuing technical discussions are inadequately backgrounded. Consequently, the basic designs of pressurized water reactors (PWR’s), boiling water reactors (BWR’s), and pressure-tube graphite reactors (RBMK) are discussed in adequate detail. Of particular interest, the Three Mile Island design for a PWR is presented in some detail.
Sleep disturbance in family caregivers of children who depend on medical technology
Authors: Keilty Krista , Cohen Eyal , Spalding Karen , Pullenayegum Eleanor , Stremler Robyn 
Published: October 2017

ObjectivesFamily caregivers of children who depend on medical technology (CMT) provide highly skilled care up to 24 hours per day. Sleep disruption places family caregivers at risk for poor health and related outcomes that threaten their long-term caregiving capacity. Few studies exist that have measured sleep in family caregivers, and most have relied entirely on subjective measures.MethodsIn a prospective cohort study, family caregivers of CMT (n=42) and caregivers of healthy children (n=43) were recruited. Actigraphy data and a concurrent sleep diary were collected for 6 days/7 nights. Measures of sleep quality, depression, sleepiness, fatigue and quality of life were also administered.ResultsFamily caregivers of CMT averaged fewer hours of sleep per night (mean (SD)) (6.56 ± 1.4 vs 7.21 ± 0.6, p=0.02) of poorer quality (7.75 ± 2.9 vs 5.45 ± 2.8, p<0.01) than the control group. Three times as many family caregivers of CMT scored in the range for significant depressive symptomatology (12(33%) vs 4(10%), p=0.01) and experienced excessive daytime sleepiness (16(38%) vs 5(12%), p<0.01). Fatigue was also more problematic among family caregivers of CMT (22.12 ± 9.1 vs 17.44 ± 9.0, p=0.02).ConclusionsFamily caregivers of CMT are at risk of acute and chronic sleep deprivation, psychological distress and impaired daytime function that may threaten their capacity for sustained caregiving. Family caregivers of CMT may be important targets for screening for sleep disorders and the development of novel sleep-promoting interventions.
Compliance with nutrition screening in a children’s hospital
Authors: Marino Luise V , Beattie Robert Mark 
Published: October 2017

IntroductionThere has been an increased emphasis on the nutritional care of children in hospital with the recognition that those admitted to hospital, particularly with chronic conditions, are at significant nutrition risk. However, although nutrition risk screening tools (NRST) are widely used in acute hospital settings compliance with their use is poorly reported.MethodsOne-day cross-sectional audit of anthropometry/NRST records, Southampton Children’s Hospital. Inclusion criteria were all in-patients present on the ward between 8 am and 4 pm. Comparison was made with previous data.ResultsOne hundred and thirteen records were reviewed. Anthropometric measures; weight recorded in 96.4% length/height recorded in 64% . This reflects a significant improvement on previous length/height data of 19%. Compliance with NRST was poor; only 17% of records were completed.ConclusionCompliance with basic anthropometry was high and better than previously reported although use of nutritional screening tools was poor. This raises questions about the usefulness of current tools in clinical practice and whether measurement of height, weight and assessment of intake and nutrition risk may be more appropriate.
Comparative analysis of autofluorescence and OCT angiography in Stargardt disease
Authors: Guduru Abhilash , Lupidi Marco , Gupta Arushi , Jalali Subhadra , Chhablani Jay 
Published: October 2017

AimsTo characterise the vasculature of the retina in patients with Stargardt disease (STGD) using optical coherence tomography angiography (OCTA) and to compare these functional findings with fundus autofluorescence (FAF) imaging.MethodsThis observational study included consecutive patients with STGD. The choriocapillaris (CC) layer was analysed on OCTA and retinal pigment epithelium (RPE) changes were assessed on FAF. Areas of CC and RPE impairment were quantified and correlated.ResultsTwenty-two patients suffering from different stages of STGD were enrolled. OCTA revealed a vascular rarefaction with vascular prominence in 15 (35%) eyes while the remaining cases had pure vascular rarefaction. On FAF imaging, 25 (58%) eyes had a peripapillary halo, a hypofluorescent lesion in foveal area and the presence of both hypofluorescent and hyperfluorescent changes. On FAF imaging, the average area of RPE impairment was 6.7±4.4 mm2 while on OCTA a hypointense area of 4.2±3.6 mm2 was shown at the level of the CC layer. Mann-Whitney U test showed a statistically significant difference in terms of lesion extension between these two findings (p=0.004).ConclusionRPE damage on FAF appears to be significantly larger than CC layer vessel loss on OCTA, which suggests that RPE damage might precede that of CC.