bluuebunny/crossref_metadata_2025_split huggingface dataset:
Interobserver reliability of computed tomographic contouring of canine tonsils in radiation therapy treatment planning
Authors: Murakami Keiko , Rancilio Nicholas J. , Plantenga Jeannie Poulson , Moore George E. , Heng Hock Gan , Lim Chee Kin 
Published: December 2017

AbstractIn radiation therapy (RT) treatment planning for canine head and neck cancer, the tonsils may be included as part of the treated volume. Delineation of tonsils on computed tomography (CT) scans is difficult. Error or uncertainty in the volume and location of contoured structures may result in treatment failure. The purpose of this prospective, observer agreement study was to assess the interobserver agreement of tonsillar contouring by two groups of trained observers. Thirty dogs undergoing pre‐ and post‐contrast CT studies of the head were included. After the pre‐ and postcontrast CT scans, the tonsils were identified via direct visualization, barium paste was applied bilaterally to the visible tonsils, and a third CT scan was acquired. Data from each of the three CT scans were registered in an RT treatment planning system. Two groups of observers (one veterinary radiologist and one veterinary radiation oncologist in each group) contoured bilateral tonsils by consensus, obtaining three sets of contours. Tonsil volume and location data were obtained from both groups. The contour volumes and locations were compared between groups using mixed (fixed and random effect) linear models. There was no significant difference between each group's contours in terms of three‐dimensional coordinates. However there was a significant difference between each group's contours in terms of the tonsillar volume (P < 0.0001). Pre‐ and postcontrast CT can be used to identify the location of canine tonsils with reasonable agreement between trained observers. Discrepancy in tonsillar volume between groups of trained observers may affect RT treatment outcome.
A quilting subcutaneous suture pattern to reduce seroma formation and pain 24 hours after midline celiotomy in dogs: A randomized controlled trial
Authors: Travis Blake M. , Hayes Galina M. , Vissio K. , Harvey H. J. , Flanders James A. , Sumner Julia P. 
Published: December 2017

AbstractObjectiveTo determine the influence of a quilting suture pattern tacking the subcutaneous tissues to the deep fascia on complications after midline celiotomy in dogs.Study designSingle‐center, randomized, blinded, controlled trial conducted in a veterinary teaching hospital.AnimalsFour hundred thirty‐two dogs undergoing midline celiotomy for ovariohysterectomy (n = 249) or other abdominal procedures (n = 183).MethodsDogs were randomly assigned to (1) a quilting group, subcutaneous tissue with apposition plus tacking to the rectus fascia or (2) a nonquilting group, with apposition of subcutaneous tissue without tacking. Randomization was stratified on hospital admitting service. Primary outcome was the incidence of incisional seroma. Secondary outcomes included postoperative pain the day after surgery and surgical site infection (SSI). Outcomes were assessed during the first 30 postoperative days.ResultsNo differences were detected between the quilting group (n = 183) and the nonquilting group (n = 175) in terms of illness severity, surgical procedure performed, surgeon's experience, duration of surgery, intraoperative complications, or methods of surgical closure other than the intervention under study. In an intent‐to‐treat analysis, the incidence of incisional seroma was lower in the quilting group (odds ratio = 0.30, 95% CI = 0.13‐0.67, P = .004). Pain assessed 24 hours postoperatively was lower in the quilting group (P = .03). The incidence of SSI did not differ between groups.ConclusionTacking the subcutaneous tissues to the deep fascia is indicated to reduce seroma during celiotomy closure.
Three‐dimensional morphology and volume of the free Achilles tendon at rest and under load in people with unilateral mid‐portion Achilles tendinopathy
Authors: Nuri Leila , Obst Steven J. , Newsham‐West Richard , Barrett Rod S. 
Published: December 2017

New Findings
What is the central question of this study?
The aim was to determine the effect of mid‐portion Achilles tendinopathy (MAT) on free Achilles tendon three‐dimensional morphology and volume at rest and under load in people with unilateral MAT.

What is the main finding and its importance?
Tendinopathic tendon had a larger resting tendon cross‐sectional area and anteroposterior diameter relative to healthy tendon. When loaded, tendinopathic tendon experienced a reduction in transverse morphology (i.e. cross‐sectional area, anteroposterior and mediolateral diameters) and overall volume reduction. In contrast, the healthy tendon remained isovolumetric and bulged along the anteroposterior axis. These findings suggest a fundamental reorganization of tendinopathic tendon matrix components and altered tendon fluid content when under load.

AbstractMid‐portion Achilles tendinopathy (MAT) adversely affects free Achilles tendon (AT) structure and composition. However, it is not known how these pathological alterations associated with MAT change the normal three‐dimensional (3‐D) morphology of free AT at rest and under load throughout the entire free tendon length. Here, we used 3‐D ultrasound to examine the effect of unilateral MAT on free tendon 3‐D morphology [length, cross‐sectional area (CSA), anteroposterior (AP) diameter and mediolateral (ML) diameter] and volume at rest and during a submaximal (50%) voluntary isometric plantarflexion contraction bilaterally in individuals with unilateral MAT (n = 10) compared with a matched healthy control group (n = 10). The tendinopathic free AT had a greater CSA relative to the control tendons along the entire tendon length, which was mainly driven by a greater tendon AP diameter. Under load, the tendinopathic tendon experienced greater longitudinal and transverse strains than the control tendons. In contrast to the control tendons, which experienced a reduction in tendon CSA and ML diameter, bulged along the AP axis and behaved isovolumetrically under load, the tendinopathic tendon experienced a reduction in tendon CSA, AP diameter and ML diameter and an overall volume reduction. Overall, these findings suggest that the magnitude of longitudinal strain and volume change and the corresponding magnitude and direction of transverse strain under load are altered in MAT compared with normal tendon. These findings are indicative of a fundamental reorganization of the tendon matrix and alterations in tendon fluid content and distribution under load in tendinopathic tendon.
Implementation of ASME OM Requirements in IST Programs in Spain: Special Focus on ISTA General Requirements
Authors: Fiol Martínez Francisco , García Rocasolano Jesús 
Published: December 2017

Throughout the world, IST Programs are usually required by the Regulatory Body that holds authority over the site. IST programs have several sources. Typically, these include IST Codes and Standards, Plant Technical Specifications, Final Safety Analysis Report and, should the plant have developed it, the Probability Risk Assessment. Rulemaking clarifications, modifications and requirements play a key role connecting all applicable documentation. In Spain, the Spanish Regulatory Body, CSN (Consejo de Seguridad Nuclear) requires all NPPs to develop and implement an IST Program according to the Codes and Standards of the country of design origin. As a result, all Spanish NPPs that have been designed in the US follow a 10CFR50.55a and ASME OM IST-based approach. In order to be able to operate, Spanish NPPs must have an official document called “MISI”, which stands for Manual of In-Service Inspection. The scope of this manual is wide: at the very least MISIs include in their scope ASME OM, ASME Section XI and Appendix J requirements. In this presentation, we explain how we intertwine NRC Regulations with our Regulatory Body’s Regulations, and applicable Codes and Standards specially focusing on ASME OM ISTA requirements.
Paper published with permission.
Smooth Running Pumps
Authors: Moenkedick Betsy , McGaha Mel 
Published: December 2017

Smooth Running Pumps have been an Industry issue since 1988. This caused many nuclear plants to obtain a Relief Request to use alternative requirements than those specified in the tables in subsections ISTB or ISTF of the ASME OM Code. Code Case OMN-22, Smooth Running Pumps, was approved in January 2017 for pumps with very low reference value vibration levels. This Code Case specifies the alternative requirements that may be used in lieu of the applicable Code subsections.
Paper published with permission.
Terry Turbopump Expanded Operating Band
Authors: Osborn Douglas M. , Ross Kyle , Cardoni Jeff , Solom Matt , Zhao Haihua , O’Brien James , Vierow-Kirkland Karen , Bergman Mark , Bunt Randy 
Published: December 2017

The Terry turbine is a small, single-stage, compound-velocity impulse turbine originally designed and manufactured by the Terry Steam Turbine Company purchased by Ingersoll-Rand in 1974. Terry turbines are currently manufactured and marketed by Dresser-Rand. Terry turbines were principally designed for waste-steam applications. Terry turbopumps are ubiquitous to the US nuclear fleet as a steam driven turbopump in either the reactor core isolation cooling system (RCIC) and high pressure coolant injection systems for boiling water reactors (BWRs) or in the auxiliary feedwater system (AFW) system for pressurized water reactors (PWRs).
Prior to the accidents at Fukushima Daiichi, assumptions and modeling of the performance of Terry turbopumps were based mostly on generic vendor use of NEMA SM23 Steam Turbine for Mechanical Drive Service guidance [1]. However, the RCIC/AFW system performance (i.e., the Terry turbopump) under beyond design basis event (BDBE) conditions is poorly known and largely based on conservative assumptions used in probabilistic risk assessment (PRA) applications. For example, common PRA practice holds that battery power (DC) is required for RCIC operation to control the vessel water level, and that loss of DC power results in RCIC flooding of the steam lines and an assumed subsequent failure of the RCIC turbopump system. This assumption for PRA implies that RCIC operation should terminate on battery depletion which can range from 4 to 12 hours. In contrast, real-world observation from Fukushima Daiichi Unit 2 shows that RCIC function was not terminated by uncontrolled steam line flooding or loss of control power, and in fact provided coolant injection for nearly three days [2].
There is a current effort being undertaken by the US industry, the US Department of Energy (DOE), and the Government of Japan to investigate the true operating band of the Terry turbopump for BDBE conditions. This paper provides a summary of the experimental and modeling efforts to date.
Paper published with permission.
What Is Your Actual Pump Flow Rate?
Authors: Perkins Norman F. , Stacy Philip S. 
Published: December 2017

What appears to be a simple question is often quite difficult to answer depending on the quantity of flow; and size, type, and location of piping. Even the reason for asking the question can be varied and complex — ranging from environmental regulation, investment decisions, aging infrastructure improvement planning, and new equipment evaluation. Absolute field performance testing of power plant equipment yields valuable data that can be used in a variety of ways.
National and International codes list several methods to measure water flow in a performance application and provide realistic uncertainty estimates. Codes and standards exist for equipment evaluation and contractual performance tests. These Codes, though, are sometimes viewed as costly or perceived to impose additional risk on suppliers. Herein, we will present how to obtain performance test data and how that data can be used.
In many rehabilitation or regulation driven projects, an accurate representation of the state of the existing power plant is desired. Pump curves typically do not represent an accurate depiction of flow due to equipment degradation, changes in system components/geometry, and/or bio-fouling. While the testing may be considered costly, it can often be justified as part of a rehabilitation project. Absolute testing provides a lower uncertainty that can yield more definitive estimates of return on investment to justify projects that might be otherwise considered marginal.
Case studies will be discussed that illustrate these points, including:
• Flow measurement feasibility and site testing at a nuclear thermal plant
• In-situ flow testing to calibrate existing ultrasonic flow meters at a biomass thermal plant
• Condenser performance testing at a nuclear thermal plant
Paper published with permission.
New Validated Methodology for the Required Force to Operate Balanced Disk Globe Valves
Authors: Leutwyler Zachary , Beasley Kenneth , Leutwyler Emil , Mistry Mital 
Published: December 2017

Electric Power Research Institute (EPRI) contracted Kalsi Engineering, Inc (KEI) to perform flow loop testing, computational fluid dynamic (CFD) analyses, and methodology development to more accurately predict flow-induced forces in balanced globe valves. The flow loop test conditions included single-phase and two-phase water flow, straight pipe and upstream-flow-disturbance pipe configurations, and two 4-inch balanced disk globe valves test specimens with a combination of quick opening and linear trim. CFD predictions were performed with a commercial grade dedicated version of ANSYS CFX 16.0 software.
The methodology was developed to utilize key dimensional characteristics of the disk and cage to determine the effective area through the stroke. The methodology accounts for trim characteristics, flow orientation, disk style, maximum valve DP and maximum flow rate. The model is validated for fluid temperatures between 70 °F and 160 °F, flow velocities up to 45 ft/sec. The methodology was validated against flow loop test data over a range of flow conditions, disk styles, and trim characteristics.
Paper published with permission.
Delivering the Nuclear Promise Through Effective ISE and IST Programs for Dynamic Restraints
Authors: Palmer Matt 
Published: December 2017

Delivering cost reductions via the Nuclear Promise can appear to be at odds with the safe operation and maintenance of nuclear facilities. However, In-Service Examination (ISE) and In-Service Testing (IST) programs can deliver significant gains in efficiency and effectiveness with proper application of the ASME O&M code. Along with scheduled maintenance prescribed by the manufacturer, Dynamic Restraints (snubbers) require periodic visual inspection and testing to ensure the installed population will perform its safety function during seismic events or dynamic operational transients. Methods prescribed in the ASME O&M Code Subsection ISTD are effective in identifying bad actors and verifying the operational readiness of the population, but can come at a significant cost when not properly utilized, especially when the penalty for a failed test or inspection is applied to the ISE or IST campaign. The Nuclear Promise can be realized in a snubber ISE or IST program with a thorough understanding of the intent of the prescribed testing and the mechanics of the safety functions to be verified. With this understanding, legacy requirements that were grandfathered into a program can be examined as to their relevance, and procurement specifications and testing procedures can be written that are pertinent and current to industry best practices.
This paper, through the lens of a snubber manufacturer and ASME certificate holder, examines some common and uncommon examples found in industry that add significant cost, time, or dose to a snubber ISE/IST program, and the basis for eliminating them. The methodology used to evaluate an ISE/IST program requirement and determine its effectiveness in verifying a snubber’s safety function while satisfying the O&M code could be used for other components under the jurisdiction of the O&M code. In this manner, the Nuclear Promise can be safely delivered in an ISE/IST program that does not compromise the intent or integrity of Code requirements.
Paper published with permission.
Actuator Fatigue Cycles: Delivering the Nuclear Promise
Authors: Kropp Ronald A. 
Published: December 2017

Motor-Operated Valves (MOVs) in nuclear plants have an equipment qualification (EQ) life of 2000 cycles. Historically, nuclear generating stations have not counted actuator cycles. As nuclear plants age beyond 30 years and licenses are extended from 40 to 60 years, the accumulated actuator operational cycles have become an issue. Without hard data, plants are not able to “prove” that actuator fatigue cycle count is less than 2000 cycles. The only viable option to address this issue is actuator replacement. Actuator replacement is expensive and frequently require long lead times for procurement. Nuclear plants generally have fifty to one hundred qualified actuators and replacing all qualified actuators is prohibitively expensive.
The questions is simply how to deliver the nuclear promise without wholesale replacement of actuators and expending significant capital or O&M dollars.
Paper published with permission.
TVA OM Code Mandatory Appendix III Readiness Assessment
Authors: Thacker Bob H. , Mills Jerry V. , Estep Neal E. 
Published: December 2017

To prepare for implementation of ASME OM Code [1] Mandatory Appendix III (Appendix III) for inservice testing of motor-operated valves (MOVs), Tennessee Valley Authority (TVA) performed a comprehensive assessment at all three of their nuclear sites to identify gaps between their legacy IST and MOV programs and an IST program that meets the requirements of Appendix III. This assessment reviewed each paragraph of Appendix III and T VA governing documents to determine how the requirements are already being met or are missing in the legacy MOV program(s). Secondly, the assessment performed a high level overview of TVA’s MOV programs in response to NRC Generic Letters 89-10 [6] and 96-05 [7] and identifies areas for improvement for TVA consideration. This paper presents the assessment purpose and objectives, scope, approach and methods, references, summary of significant gaps, and proposed actions to resolve these gaps prior to Appendix III implementation.
Paper published with permission.
Component Cooling Water (CCW) Pump Assessment Report
Authors: Tomlinson Watson , Kochaniec Steven 
Published: December 2017

Three (3) Component Cooling Water (CCW) pumps, IR 8X18SE, appeared to have been operated beyond their original manufacturer pump curves without proper justification and analysis. The testing was performed to determine exactly where the pumps were operating relative to the original head-capacity curve, at different system demands. Additionally, the results would be compared to the customers flow and pressure measuring capabilities.
Paper published with permission.
Realignment of ASME Operations Maintenance Committee: Improving Responsiveness and Efficiency
Authors: Ruggiero T. 
Published: December 2017

The O&M Code was developed when it was decided to move Pump and Valve Inservice Testing (IST) Requirements from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI to a standalone Code. The Code review process structure at the time was quite small and generally consisted of changing Section XI Subsections IWP and IWV into OM language. At the same time, new testing techniques were being developed that included check valve condition monitoring and current trace testing of motor actuated valves. This necessitated adding groups that were specific to these new initiatives.
Although that was several decades ago, these groups remained and, over the years, it was identified that actions, such as Inquiries, were taking much too long to process. This became abundantly clear with the development of the newly published Mandatory Appendix IV for Air Operated Valve Testing.
This paper discusses how the Code Committee became the organization that it is and how a new realignment will streamline the Code process and make it more efficient and responsive to the industry/regulatory needs.
Paper published with permission.
Pump and Valve Inservice Testing: How Pump and Valve Testing Evolved
Authors: Ruggiero T. 
Published: December 2017

The O&M Code was developed when it was decided to move Pump and Valve In-Service Testing Requirements from ASME Boiler and Pressure Vessel Code Section XI to a standalone Code. IST for Pumps was originally is ASME Section XI IWP and for Valves IWV. Safety and Relief Valves were a Power Test Code and not in the scope of the ASME Boiler and Pressure Vessel Code. IWP and IWV were developed after plants had been designed and built. The desire was that no back-fits were to be required to comply with IST requirements. After the 1986 Edition, IWP and IWV requirements were moved into O&M. Appendix 1 of OM was what used to be the Power Test Code.
While this was going on, the NRC issued what has been called “the Richardson Letter”. Among other things, that letter required that IST for pumps better asses the condition of the pumps by putting higher accuracy instrumentation on the test pipe. For many plants, this was the minimum recirculation pipe. Over the course of time, the committee was able to get agreement that if a centrifugal pump were tested “back on its curve” increased instrument accuracy would be meaningless. This was the genesis of what we now call comprehensive Pump testing. Additionally, there were several alternative methods for valve testing that had been developed. It became clear, that simple periodic stroke timing of a power operated valve was simply not adequate for detecting degrading performance.
This presentation will discuss how Pump and Valve In Service Testing evolved to what it is today and discuss what might be alternatives in the future.
I want to thank Robert Parry, who provided some insights into this presentation specifically where my memory needed a bit of jogging.
Paper published with permission.
Plant Performance History of an Innovative Gate Valve in Critical Service Applications
Authors: Kalsi M. S. , Alvarez Patricio , White Thomas , Green Micheal 
Published: December 2017

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper.
Paper published with permission.