bluuebunny/crossref_metadata_2025_split huggingface dataset:
A Critical Understanding of “Low-Stress” Spiral Wound Gaskets
Authors: Jamison Alton 
Published: November 2022

Abstract
Low-stress spiral wound gaskets are marketed as an alternative to standard spiral wound gaskets, requiring less torque to seat the gasket. Spiral wound gaskets are common commodities used in piping reliability staff are constantly looking for different alternatives. Multiple manufacturers offer a low-stress version of spiral wound gaskets. Do these spiral wound gaskets offer a low-stress solution, and do they perform better than a standard spiral wound gasket? This paper will go beyond the marketing of “low-stress” spiral wound gaskets and examine the construction and engineering behind these gasket designs. Manufacturers of spiral wound gaskets have made subtle changes and the “low-stress” technology has become a common theme throughout the spiral wound gasket market. Multiple chemical and petrochemical plants use these designs in their piping systems and sometimes as replacements for ASME recommended spiral wound gaskets with inner rings. Low-stress spiral wound gaskets have multiple designs from additional graphite in the filler to an anti-buckling design which are marketed as requiring less initial preload to seat. This paper will examine the validity of these gaskets and determine if they are low-stress and if they provide a credible seal in a bolted flanged joint. [1]
Large-Scale Parametric Modeling of Spent Nuclear Fuel Dynamics in the 30 cm Package Drop Scenario
Authors: Kadooka Kevin , Fitzpatrick James , Klymyshyn Nicholas 
Published: November 2022

Abstract
Packages used to transport spent nuclear fuel (SNF) are required by the U.S. Code of Federal Regulations 10 CFR 71.71 to demonstrate satisfactory performance during a drop scenario. While the CFR is meant to ensure safe package function, it does not evaluate survival of the SNF within. The U.S. Department of Energy Spent Fuel and Waste Science and Technology program is working on closing the knowledge gap related to the response of SNF to external mechanical loads, including the 30 cm package drop scenario in the CFR. In support of this effort, LS-DYNA finite element simulations were developed by Pacific Northwest National Laboratory (PNNL) to model generic drop scenarios at both the package and fuel assembly level. The models were validated against one-third scale package and full scale fuel assembly drop test data and were exercised to predict fuel cladding strains in a narrow range of model configurations.
This work describes a large-scale parametric study conducted by PNNL using the previously developed and validated PWR finite element model, with the addition of a new generic BWR assembly model. The motivation for the parametric study was to characterize the broad range of SNF responses in the 30 cm package drop scenario. This was accomplished by varying the drop orientation, fuel assembly type (17 × 17 PWR and 10 × 10 BWR), burnup, cladding temperature, spacer grid buckling load, package mass, impact limiter stiffness, and mechanical gap conditions within the basket. A MATLAB framework was developed to automate LS-DYNA model generation and execution on PNNL institutional computing resources. In total, over 2000 simulations were performed. For each simulation, the SNF response was quantified in terms of permanent grid deformation, fuel rod contact pressure, and strains within the fuel rods, guide tubes, and water rods. The results provide valuable insight into the range of responses that could be reasonably expected from SNF in the 30 cm drop scenario, as well as the sensitivity to each input parameter. The results of this parametric study are a key component of the testing and modeling strategy the Spent Fuel and Waste Science and Technology program is using to close the external loads knowledge gap.
Electron Beam Welds in 316L Part 1: Weld Production, Residual Stress Measurements and Predictions
Authors: Horne Graeme , Moffat Andrew 
Published: November 2022

Abstract
There is enthusiasm for new civil nuclear plants in the UK to adopt power beam welding technologies, which could offer several advantages over conventional techniques. In particular, reduction in the time taken to produce and inspect a weld, and thus the cost of manufacturing components. However, a blocker to the adoption of these technologies is a shortage of accepted methodologies for demonstrating the integrity of these joints, which forms part of the requirements of the generic design assessment within the UK regulatory environment.
Residual stresses can contribute towards crack driving force and thus should be accounted for when assessing the integrity of a component or its tolerance to damage. Whilst bounding residual stress fields are often used, it is often desirable to have more realistic estimations that capture the through-wall residual stress distribution, which also allows them to be decomposed into membrane, bending and self-equilibrated components to aid stress classification.
Material-specific weld residual stress profiles already exist, for example Level 3 profiles in the UK’s R6 procedure. However, they are for arc welding techniques. This work seeks to provide a framework for the generation of weld residual stress profiles for power beam welds and is split over two papers:
1. Weld Production, Residual Stress Measurements and Predictions;
2. A Methodology and Example for Parameterised Residual Stress Profiles.
In this paper an experimentally-validated programme of weld modelling is presented to be used as an input to the framework for producing parametric through-thickness weld residual stress profiles presented in the second paper.
An Experimental and Analytical Study on the Development of Extrapolation Method for the Creep-Fatigue Life of Alloy 617 to Low Strain Ranges and Long Hold Times at 950°C
Authors: Hou Peijun , Sham Ting-Leung , Wang Yanli 
Published: November 2022

Abstract
Creep-fatigue is the most severe damage mode for high temperate components under cyclic loading. Experimentally, standard creep-fatigue testing is mostly performed at strain ranges of 0.3% and above. There is a major data gap in creep-fatigue failure information for strain ranges at 0.3%, or lower, which are more prototypical for advanced reactor plant operations. To develop design curves for creep-fatigue evaluations, it will require a method for extrapolating the creep-fatigue cyclic lives to lower strain ranges and longer hold times. In this study, methods for creep-fatigue life prediction of Alloy 617 at lower strain ranges and longer hold times at 950°C were assessed. Two types of experimental data were generated to support the development of the creep-fatigue design curves at low strain ranges and long hold times. First, standard strain-controlled creep-fatigue testing using the concept of block-strain range was conducted to generate the information needed for extrapolating the creep-fatigue design curves to lower strain ranges and longer hold times. Second, creep-fatigue experiments with elastic follow-up were performed to failure to verify the proposed design curves. Based on this new testing approach and analysis method, a preliminary creep-fatigue design curve was developed for Alloy 617 at 950°C. This study shows the potential of generating a set of creep-fatigue design curves with different hold times within a reasonable amount of test times and testing effort at the low strain range region.
Electron Beam Welds in 316L Part 2: A Methodology and Example for Parameterised Residual Stress Profiles
Authors: Horne Graeme , Elliott Ben , Moffat Andrew 
Published: November 2022

Abstract
There is enthusiasm for new civil nuclear plants in the UK to adopt power beam welding technologies, which could offer several advantages over conventional techniques. In particular, reduction in the time taken to produce and inspect a weld, and thus the cost of manufacturing components. However, a blocker to the adoption of these technologies is a shortage of accepted methodologies for demonstrating the integrity of these joints, which forms part of the requirements of the generic design assessment within the UK regulatory environment.
Residual stresses can contribute towards crack driving force and thus should be accounted for when assessing the integrity of a component or its tolerance to damage. Whilst bounding residual stress fields are often used, it is often desirable to have more realistic estimations that capture the through-wall residual stress distribution, which also allows them to be decomposed into membrane, bending and self-equilibrated components to aid stress classification.
Material-specific weld residual stress profiles already exist, for example Level 3 profiles in the UK’s R6 procedure. However, they are for arc welding techniques. This work seeks to provide a framework for the generation of weld residual stress profiles for power beam welds and is split over two papers:
1. Weld Production, Residual Stress Measurements and Predictions;
2. A Methodology and Example for Parameterised Residual Stress Profiles.
In this paper, the second in this series, a framework is presented for producing parametric through-thickness weld residual stress profiles using the validated modelling from the first paper in the series.
A Viscoplastic Model for Alloy 800H for Use With the Section III, Division 5 Design by Inelastic Analysis Methods for Class A Components
Authors: Messner M. C. 
Published: November 2022

Abstract
Recently, the ASME approved a new non mandatory appendix to Section III, Division 5, Subsection HB, Subpart B of the ASME Boiler & Pressure Vessel Code providing guidance on the Division 5 design by inelastic analysis methods for Class A components. This appendix provides guidance on constructing and validating nonlinear constitutive models for use with the Code rules as well as approved material models the designer can use without additional validation. An eventual goal for the new appendix is to provide reference models for all the HBB Class A materials. This paper describes an inelastic constitutive model for Alloy 800H, suitable for use with the design by inelastic analysis methods. The model uses a unified viscoplastic form and a detailed kinematic hardening model to capture the key features of high temperature cyclic behavior in Alloy 800H. At lower temperatures the model switches to a rate independent reponse to better capture the available experimental data. The paper describes the model form, calibration database, and details how the model was calibrated and validated against experiment data. The goal for this model is to incorporate it into the ASME Code, via the newly approved appendix to Subsection HB, Subpart B, to continue to expand the range of prebuilt constitutive models available to designers.
Mechanical Properties and the Effect of Hydrogen on Base Metal and Welds of 9Cr-1Mo-V Steel
Authors: Zuazo Rodriguez Ian , Alberini Stefano , Bouillot Claude , Cornut Dany , Fusari Fausto 
Published: November 2022

Abstract
9Cr-1Mo-V steel or Grade 91 has been typically used up to now in thermal power plants operating at high temperatures, e.g. 600°C. In another potential application, such as refinery equipment like hydrocracker and hydrotreater reactors, the choice is still the 2¼Cr-1Mo-¼V. The main reason is that the average design temperature of hydrocracker and hydrotreater reactors, approximately 450°C, has not been increased in the last 20 years making the use of 2¼Cr-1Mo-¼V steel more suitable and cheaper in comparison to Grade 91.
However, some recent interest has been observed for its application in the petrochemical and chemical sector, including for new process technologies related to green economy market sectors. In these applications the influence of hydrogen could be detrimental for the steel and in the past, some investigations concluded positively on the use of Grade 91. In recent years, an improved control of the chemistry intended to improve creep properties has led to the industrial production of the SA-387 Grade 91 Type 2.
Therefore, due to a renewed interest and an improved chemistry, tests were planned and carried out with the aim of assessing the hydrogen effect on the Grade 91 Type 2. This was done using welded test samples covering both full plate butt joints and weld overlays using two stainless steels, 308L and 347. In the butt joints, the impact transition curves of the weld metal, heat affected zone and base materials were compared to those after a heat treatment under high hydrogen pressure. The resistance to hydrogen induced disbonding was evaluated on the weld overlays. Finally, the influence of step cooling treatment on impact toughness was also investigated for both base and weld metal.
Analysis to Relate Data From Radial Compression Tests on Helical Springs to Tensile Material Properties
Authors: Gagnon Andre , Metzger Donald 
Published: November 2022

Abstract
In CANDU nuclear reactors, pressure tubes reside within a calandria tube with separation maintained by helical springs installed in the annular space. Evaluation of material degradation due to the unique operating environment requires testing of ex-service spring material by compressing short spring segments by two diametrically opposed forces. The load vs. displacement results combined with the geometry allows for the stress-strain behavior to be derived. The test specimens are effectively unmodified from the as-received condition so accurate characterization of the geometry is required. Since the test response is mainly bending, error in the radial section dimension is augmented by powers of 2 and 3 when calculating bending stress and specimen stiffness respectively. Additional complications come from nonuniform loading of the coils due to end effects.
Detailed analysis that accounts for end effects is applied to the linear elastic portion of the load curve to accurately quantify the specimen dimensions. With geometry determined, the nonlinear portion of the tensile curve is adjusted to reproduce the entire load curve up to the failure point. Examples are presented to demonstrate how the load corresponding to the yield point and outer fiber stress at the failure point can be determined.
Rocking Response Analysis of Flat-Bottom Cylindrical Tanks Considering Rotational Inertia of Content Liquid
Authors: Yoshida Yuichi , Taniguchi Tomoyo , Nakashima Teruhiro 
Published: November 2022

Abstract
This paper proposes an analytical method for calculating rocking responses of unanchored tanks during seismic ground motion. Some researchers have proposed the analytical methods to calculate seismic responses of unanchored tanks without considering the rotational inertia of content liquid. In other words, most of method for a time-history analysis of tank rocking responses are assumed that an equilibrium equation in direction of rotation can be formulated as a static equilibrium equation in which the overturning moment and the restoring moment are equal. Even a model that considers rotational inertia of content liquid was proposed in the past, the definition of moment of inertia is ambiguous. On the other hand, there a study that derives moment of inertia from a physical background and alleges the importance of the moment of inertia in the tank rocking response.
In this study, an analytical method of a time history analysis of tank rocking motion considering the rotational inertia of content liquid is proposed. Furthermore, results of the proposed method are compared with results of the FEM dynamic analysis to validate the accuracy of the proposed method. Finally, comparisons with results of the conventional calculation method are also conducted.
Inverse Conduction Method for Complex Thermal Loading
Authors: Gilman Timothy 
Published: November 2022

Abstract
For U.S. nuclear power plants, transfer functions have typically been utilized in fatigue monitoring applications to characterize a pressure vessel or piping component’s local, inside fluid temperatures, which are the primary forcing functions for stress cycling and the accumulation of fatigue damage, crack nucleation, and crack growth. The transfer function approach involves using simplified, conservative thermal-hydraulic engineering principles to develop algorithms that calculate the local temperatures, using existing plant instruments, such as temperatures, flow rates, valve positions, etc.
With the increased ease of adding instrumentation using modern thermocouple and wireless data technology, numerous practical uses are available to plants now, such as:
• More accurately and easily characterizing local thermal loading of components for analysis or fatigue monitoring, including flaw tolerance methods.
• Determining the inside temperature loading in normally-stagnant branch lines subject to thermal cycling. Accurate knowledge of inside temperatures in these lines can potentially eliminate the need for costly, frequent inspections when a significant temperature range threshold is not exceeded [1] [2].
• Inside temperate loading, based on measured data, can also provide the basis for detailed evaluations in emergent situations when cracks or new loadings are discovered.
With thermocouple data, temperatures on the outside surface of components are measured, but inside surface temperatures are needed to compute accurate stresses. There are challenges with attempting to deterministically calculate inside temperatures, because of numerical instability issues caused by temperature fluctuations, measurement inaccuracies, or both. The problem becomes increasingly complex when multiple instruments are required to characterize the loading, such as with thermal stratification or with thermal cycling caused by in-leakage interacting with swirl flow at a branch nozzle.
Because of stability challenges and the multivariate nature of the problem, a robust, generalized solution for computing inside temperatures at multiple different thermal zones is nontrivial.
Drawing on previous work, this paper discusses the challenge along with a proposed solution.
Predicting Pressure Relief Valve Generated Acoustic Induced Pipe Wall Vibration
Authors: Swindell Rob , Izuchi Hisao , Karczub Denis , Eilers Daniel , Mann J. Adin 
Published: November 2022

Abstract
The calculation of acoustic induced vibration (AIV) for piping downstream of a valve is a critical step in predicting the damage from extreme levels of noise generated by pressure relief valves in flare systems. Three noise prediction schemes are considered for this purpose: International Electrotechnical Commission (IEC) 60534-8-3, the Carucci-Mueller (C-M) formulation for sound power, and an industry valve noise prediction methodology published in the 1980’s. The application of these prediction methods is reviewed utilizing data from a full-scale test system consisting of an NPS6x8 pressure relief valve flowing into a NPS12 tailpipe that is connected through a tee to an NPS20 header. The results show good correlation between the IEC-based predictions and measured internal sound pressure and pipe wall vibration in the AIV frequency region. The industry method provides useful predictions without requiring the level of detailed information needed for the IEC method, whilst the C-M sound power model has limitations when applied to discrete predictions of vibration and strain levels. Observations are also made regarding the relative importance of the FIV contribution to the overall dynamic stresses and associated fatigue life.
A Novel Method to Predict the Concentricity of Aero-Engine Rotor Considering the Assembly Process of Bolted Flange Joints
Authors: Zhu Linbo , Yang Yilong , Huang Hongwei , Bouzid Abdel-Hakim , Hong Jun 
Published: November 2022

Abstract
The concentricity of the aero-engine rotor is an important parameter to evaluate the quality of final assembly and directly affects the vibration characteristics, especially for high-speed rotation. The use of a bolted flange joint is the basic type of connection in aero-engine rotors. During the initial tightening of the bolts, large deformation usually occurs in the flange because its thickness is small. The deformation in the flange has a major impact on the concentricity and consequently the dynamic behavior of the aero-engine rotor. This paper proposes a novel stack-build assembly method to predict the concentricity of multi-stage rotors considering the deformation of the flange. The Small Displacement Torsor (SDT) method is employed to construct the mathematical model of part errors. The homogeneous coordinate transformation method is used to analyze the deviation propagation in the bolted flange joint of each stage part. A finite element (FE) model is built to obtain the deformation of the bolted flange joints by simulating the assembly process. The deformation of the flange is involved in the stack-build assembly model as an error matrix. Furthermore, the influence of the assembly process such as interference and preload, tightening sequence on the concentricity is investigated. The results show that bolted flange joints have a significant effect on concentricity, especially for the complex geometry at the flange interface. The developed approach is validated by experimental tests conducted on a multi-stage rotor. This study can provide guidance and enhance the dynamic performance of bolted joints for aero-engine rotors.
Determination of Irradiated Stainless Steel Properties and Its Effects on Reactor Vessel Internals
Authors: Kwak Dong-Hyeon , Sim Jae Min , Chang Yoon-Suk , Kong Byeong Seo , Jang Changheui 
Published: November 2022

Abstract
Reactor vessel internals (RVIs) consist of austenitic stainless steels (ASSs) which have excellent material properties. Meanwhile, the high radiation environment of a reactor can cause the degradation of components. Since changed material properties are important for long-term operation, experimental researches related to tensile and fracture properties had been conducted. However, it is limited to investigate these researches due to their high radioactivity and small quantity. Thus recent researches have been dedicated to small specimens such as nanoindentation and micropillar compression tests and so on with ion-irradiation. In this study, micropillar compression tests were carried out for virgin and irradiated 304 ASSs to obtain microscopic mechanical behaviors. The trial sets of finite element (FE) analyses were performed to derive dislocation density based material constitutive equations for austenite phase by comparing with test results. Subsequently, representative volume elements analyses with periodic boundary conditions were adopted to estimate overall tensile stress-strain curves as well as 0.2% offset yield strengths (YSs) under the virgin and irradiated states. Finally, the effect of irradiated properties on typical RVIs were investigated. As typical results, optimized material parameters related to dislocation density based formulations were revealed, and microscopic stress-strain curves were reasonably comparable with test results. The estimated YS values were compared with the experimental results and corresponded within 9.09%. The overall deformation, stress and strain behaviors of typical RVIs were examined considering estimated properties, of which details and key findings will be discussed.