bluuebunny/crossref_metadata_2025_split huggingface dataset:
Application of One-Dimensional Hydrogen Event Pressure Pulse Propagation and Diminishment Model to Analysis and Design of Hanford WTP Piping
Authors: Pellman Abby M. , Collin Jean E. , Minichiello John C. 
Published: October 2017

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will process waste slurries capable of generating a dispersed non-condensable gas phase. Under postulated conditions, the generated bubbles may coalesce to form flammable gas pockets. Ignition of such a gas pocket may result in a deflagration or detonation that can transmit a pressure pulse into the surrounding gas-liquid slurry. The transmitted pressure pulses will impose structural loads on the piping and associated supports. These loads, and their evolution as the pressure pulse propagates throughout the system, must be considered in the design of WTP piping. Recent work has demonstrated good agreement between experimental data and predictions of the pressure diminishment associated with the propagation of a pressure pulse through a bubbly medium using a simplified onedimensional pursuit model (i.e., model of rarefaction wave overtaking shock front).
This paper describes the application of the pursuit model to flammable gas pocket scenarios postulated in the design of WTP piping. The model is adapted to consider the effect of the time-evolution of the ignited gas pocket and parametric simulations are used to demonstrate the effect of key variables on the propagation and diminishment of the pressure pulse in piping. The practical applicability of the pursuit model for WTP piping design / analysis is illustrated through the development of a correlation that appropriately represents the results of the pursuit model for a wide range of conditions and that could be readily incorporated into existing structural analysis tools. Arguments are presented supporting the applicability of the simplified model even for scenarios in which the ignited gas pocket does not satisfy the model’s onedimensional assumption.
Study on the Stream-Wise Fluidelastic Instability of Rotated Square Arrays of Circular Cylinders Subjected on Cross-Flow
Authors: Nakamura Tomomichi , Tsujita Tomoki 
Published: October 2017

Stream-wise fluidelastic instability has recently been in the spotlight due to a practical event in steam generators at a nuclear power plant [1]. The instability has been reported to occur easily in rotated triangular arrays, but at all hardly in normal triangular arrays [2][3]. In square arrays, no instability has been reported in the authors’ test facility [4].
This paper presents the test results on rotated square arrays of the pitch-diameter ratios from 1.2 to 1.5. The tests have been conducted with a wind tunnel, where cylinders have been constrained to move in only one specific direction, either stream-wise or transverse to the flow. The results indicate that fluidelastic vibrations occur only in the stream-wise direction. The critical factor corresponds to the previous data reported in some guidelines [5], but here the factor is related mainly to stream-wise instability.
Performance Evaluation of a Respirator Vortex Cooling Device
Authors: Elizondo Ashley D. , Iacovone Robert K. 
Published: October 2017

The United States Department of Energy’s Savannah River Site (SRS) in Aiken, South Carolina, is dedicated to promoting site-level risk-based inspection practices in order to maintain a safe and productive work environment. Protective suits are worn by personnel working in contaminated environments. These suits require that cooling be applied to keep the interior temperature within safe and comfortable limits. A vortex tube, also known as the Ranque-Hilsch vortex tube, can provide the necessary cooling. As mechanical devices void of moving components, vortex tubes separate a compressed gas into hot and cold streams; the air emerging from the “hot” end reaching a temperature of 320 degrees F, and the air emerging from the “cold” end reaching a temperature of negative 25 degrees F [1]. Routing the cold stream of the vortex tube to the user’s protective suit facilitates the required cooling.
Vortex tubes currently in use at SRS are pre-set, through sole modification by and within the SRS Respiratory Equipment Facility (REF), to provide a temperature reduction between 40 and 45 deg F. When a new model of vortex tube capable of user adjustment during operation recently became available, prototype testing was conducted for product comparison. Similar cooling performance between the old and new models is achievable. Implementing the use of the new model of vortex tube at SRS will result in significant cost savings because the product could be shipped directly to the end user, circumventing adjustment by the REF. Production units were acquired to be subjected to complete product analysis at SRS utilizing a statistical test plan. The statistical test plan, data, thermodynamic calculations, and conclusions were reviewed.
Residual Stress Assessment for Pressure Vessel Subject to Thermal Shock
Authors: Hsiao Jushin , Qian Haiyang , Brunner Christopher , Bober Thomas R. , D’Amico Lynn , Loomis Robert W. 
Published: October 2017

API 579-1/ASME FFS-1 Part 9 provides assessment procedures for evaluating crack-like flaws in components to determine if it is fit for continued service. Although residual stress distribution is required as an input to perform a fatigue life assessment, no procedure or guideline is available for evaluating this crack driving force resulting from thermal shocks.
Through a systematic analysis, a conservative residual stress distribution can be obtained for pressure vessels subject to thermal shocks. For the two thick-walled vessels considered, the maximum residual stress occurs when the vessel is half filled with water. The conservative residual stress provides the needed input when using API 579-1/ASME FFS-1 for evaluating crack-like flaws in components. Dependence of the residual stress on film coefficient, temperature difference between water and metal surface, and water level inside the vessel is also presented so that refinement can be made on life assessment when additional field data becomes available.
Assessment of a Stress-Free Temperature Model for Residual Stresses in Surface Cladding of a Reactor Pressure Vessel
Authors: Bass B. Richard , Williams Paul T. , Dickson Terry L. , Klasky Hilda B. 
Published: October 2017

This paper describes further results from an ongoing study of a simplified engineering model that is intended to account for effects of clad residual stresses on the propensity for initiation of preexisting inner-surface flaws in a commercial nuclear reactor pressure vessel (RPV). The deposition of stainless steel cladding during fabrication of an RPV generates residual stresses in the cladding and the heat affected zone of the under-lying base metal. In addition to residual stress, thermal strains are generated by the differential thermal expansion (DTE) of the cladding and base material due to temperature changes during normal operation. A simplified model used in the ORNL-developed FAVOR probabilistic fracture mechanics (PFM) code accounts for the clad residual stress by incorporating a stress-free temperature (SFT) approach. At the stress-free temperature (Ts-free), the model assumes there is no thermal strain, i.e., the thermal expansion stresses and clad residual stresses offset each other. For normal cool-down transients applied to the RPV, interactions of the latter stresses generate additional crack driving forces on shallow, internal surface-breaking flaws near the clad/base metal interface; those flaws tend to dominate the RPV failure probability computed by FAVOR. In a previous report from this study (PVP2015-45086), finite element analysis was used to compare the stresses and stress-intensity factors (SIF) during a cool-down transient for two cases: (1) the existing SFT model of FAVOR, and (2) directly applied RPV clad residual stress (CRS) distribution obtained from empirical (hole-drilling) measurements made at room temperature on an RPV that was never put into service. However, those analyses were limited in scope and focused on a single flaw orientation.
In this updated study, effects of CRS on the SIF histories computed for both circumferential and axial flaw orientations subjected to a cool-down transient were determined from an extended set of finite element analyses. Specifically, comparisons were made between results from applying CRS experimental data to ABAQUS two-dimensional, inner-surface flaw models and those generated by the FAVOR SFT model. It is demonstrated that the FAVOR-recommended SFT value of 488 °F produces conservatively high values of SIF relative to the use of CRS profiles in the ABAQUS models. For the vessel and flaw geometry and transient under study, the circumferential flaw (360° continuous) required a decrease of SFT down to 390 °F to match the CRS SIF histories. For the infinite axial flaw model, a decrease down to 300 °F matched the CRS SIF histories. Future plans are described to develop more general conclusions regarding the FAVOR model.
Fatigue Life and Reliability Consideration During Field Repair of Coke Drum and Piping
Authors: Chadda Tej , Anand Umakanthan 
Published: October 2017

Aging coke drums and their connected overhead piping in delayed coking units experience fatigue cracks which most commonly occur at the skirt junction and high stress pipe welds.
This paper presents 2-case studies of this new cost-effective repair methodology with fatigue resistant design upgrade. The first case study applies to coke drum weld build-up solid skirt crack repair and the second for overhead vapor line weld crack repair.
This paper presents new field repair methodology which could also improve long term fatigue resistance. It also suggests optimizing the thermal operation & thermal gradients of coke drums for further reliability improvement.
Based on FEA, successful field execution and our experience, these case studies demonstrate a long term improvement in reliability and fatigue life of the order of 2.5 to 3 or higher especially if combined with thermal operation optimization.
RSE-M Nuclear In-Service Inspection Code New Flaw Interaction Rules
Authors: Faidy Claude 
Published: October 2017

After a quick overview of RSE-M scope now used on more than 100 Nuclear Power Plants in the World, this paper will present the flaw evaluation rules of RSE-M Appendix 5.
This Appendix considers cracks, volumetric defects and thinning areas. Periodically updated on a yearly bases, the rules cover cracks in different components, for different material, in different location. It consider defects interaction in Appendix 5 - Chapter 5-1.
In the case of multiple defects, after determining the dimensions for each defect according to the rules in previous chapters, the non-interaction conditions between the elementary defects given shall be checked. If these conditions are not met, the defects shall be grouped according to a dedicated chapter.
This review is done on the basis of RSE-M 2015 edition and last modifications of 2016 edition.
Technical Basis for Proposed ASME Code Case Based on N-513 for Higher Pressure Applications
Authors: McGill Robert O. , Janowiak Ronald J. , Houston Eric J. , Shim Do Jun 
Published: October 2017

Code Case N-513 provides evaluation rules and criteria for temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. The application of the Code Case is restricted to Class 2 and 3 systems, so that safety issues regarding short-term system operation are minimized. The first version of the Code Case was published in 2000. Since then, there have been four revisions to the Code Case that have been published by ASME. The technical bases for the original version of the Code Case and the four revisions have been previously published.
There is currently work underway to employ the methods given in N-513 for a new and separate Code Case for higher pressure piping applications. This paper provides the technical basis for the proposed Code Case that includes a structural integrity evaluation and consideration of potential jet thrust forces. In addition, discussion is provided on additional Code Case requirements considering the application to higher pressure systems in order to bolster defense-in-depth. Note that the proposed Code Case still maintains the temperature limit given in N-513.
Safety Analyses for Preventing PCV Damage by FCI in Kashiwazaki-Kariwa 6,7
Authors: Hayakawa Akinori , Suehiro Shouichi , Mizuno Satoshi , Oyama Yoshihiro , Kawamura Shinichi 
Published: October 2017

The containment vessel failure mode, “molten fuel-coolant interaction outside the reactor pressure vessel” (“ex-vessel FCI”) is a phenomenon of rapid increase in the pressure inside the containment vessel or steam explosion caused by the contact between the molten reactor core and cooling water outside the reactor pressure vessel after the reactor core is damaged.
To evaluate the viability of keeping confinement function of the containment vessels of Units 6 and 7 (ABWRs) of Kashiwazaki-Kariwa Nuclear Power Station against “ex-vessel FCI,” we conducted a code-based event progression analysis. For evaluation of the rapid increase, we employed the severe accident analysis code, MAAP, after organizing the critical phenomena of the event. In addition, assuming a case of steam explosion occurrence, we conducted an analysis with employing the steam explosion analysis code, JASMINE, and the structural response analysis code, AUTODYN-2D.
As a result of the evaluation, the maximum pressure and temperature inside the containment vessels were lower than their limits. Moreover, the maximum stress applied to the lower part of the containment vessels was lower than the yield stress on support structure of the lower part of the containment vessels. Therefore, we could confirm that the containment vessels can keep their integrity.
Confinement Effects on Added Mass of Cylindrical Structures in a Potential Flow
Authors: Capanna R. , Ricciardi G. , Eloy C. , Sarrouy E. 
Published: October 2017

Efficient modelling and accurate knowledge of the mechanical behaviour of the reactor core are needed to estimate the effects of seismic excitation on a nuclear power plant. The fuel assemblies (in the reactor core) are subjected to an axial water flow which modifies their dynamical behaviour. Several fluid-structure models simulating the response of the core to a seismic load has been developed in recent years; most of them require high computational costs. The work which is presented here is a first step in order to simplify the fluid forces modelling, and thus to be able to catch the main features of the mechanical behaviour of reactor core with low computational costs. The main assumption made in this work is to consider the fluid flow as an inviscid potential flow. Thus, the flow can be described only using one scalar function (velocity potential) instead of a vector field and strongly simplifies the fluid mechanics equations, avoiding the necessity to solve Navier-Stokes equations. The pressure distribution around a cylinder is first solved in Fourier space for different values of the parameters (wavenumber, confinement size).
The method is applied to a simple geometry (cylinder in an axial flow with a variable confinement) in order to test its effectiveness. The empirical model is then compared to simulations and reference works in literature. The configuration with large confinement has been solved, and results were in agreement with Slender Body Theory. The dependency on the confinement size strongly depends on the wavenumber, but in any case added mass increases as the confinement size decreases. Finally, future perspectives to extend the model to a group of cylinders and to improve the model are discussed (i.e. add viscosity to the model).
Testing and Numerical Simulation of Elastomers: From Specimen Tests to Simulation of Seal Behavior Under Assembly Conditions
Authors: Weber Mike , Kömmling Anja , Jaunich Matthias , Wolff Dietmar , Zencker Uwe , Völzke Holger , Schulze Dietmar , Probst Ulrich 
Published: October 2017

Extended periods of interim storage are more relevant in Germany due to delays in the siting procedure to establish a deep geological repository for spent nuclear fuel, high level radioactive waste and in low/intermediate level waste container storage designated for the Konrad repository. BAM is involved in national cask licensing procedures and responsible for the evaluation of cask-related long-term safety issues.
The long-term performance of elastomer seals for lid systems of transport and storage casks, whether used as auxiliary seals in spent fuel casks or as primary seals for low and intermediate level waste packages, is an important issue in this context. The polymeric structure of these seals causes a complex mechanical behavior with time-dependent sealing force reduction.
The results of a comprehensive purpose-designed test program consisting of basic compression and tension tests as well as relaxation tests on unaged specimens of representative types of elastomers (fluorocarbon rubber (FKM) and ethylene propylene diene rubber (EPDM)) at different temperatures and strain rates are presented. They were used to identify the constitutive behavior and to obtain parameters for finite element material models provided by the computer code ABAQUS®. After estimating the influence of parameters such as Poisson’s ratio and friction coefficient by sensitivity analyses, the chosen values for the finite element simulation were validated by comparison with specimen test results.
Based on this preliminary work the simulation of a specific laboratory test configuration containing a typical elastomer seal with circular cross section is presented. The chosen finite element material model and the related parameters had to show that they are able to represent not only the specimen behavior under predominantly uniaxial load but also the more complex stress states in real components. Deviations between the measured and calculated results are pointed out and discussed.
The results from this work will be utilized in future simulations of aged elastomer behavior.
Strain Waveform Effects for Low Cycle Fatigue in Simulated PWR Water
Authors: Seppänen Tommi , Alhainen Jouni , Arilahti Esko , Solin Jussi 
Published: October 2017

In order to perform design code (ASME III, RCC-M, JSME) compatible direct strain-controlled tests in simulated PWR water, a unique environmental fatigue testing facility was previously developed. Pneumatic bellows are used to generate strain in the stainless steel specimen mid-section, while eddy current based measurement is used as a feedback signal.
The NUREG/CR-6909 report gathered a large database of test results and proposed environmental reduction factors (Fen) to account for a reduction in fatigue life in simulated LWR environment when comparing to a reference value in air. The database was composed of non-stabilized stainless steels tested using methods which are not directly comparable to those used in air to define the reference curves. Applicability of the stainless steel Fen factors has already been challenged in previous PVP papers (PVP2013-97500, PVP2014-28465, PVP2016-63294).
Results in this paper continue to show this trend of lower experimental Fen factors compared to predictions made by the NUREG report. Dual strain rate tests were performed, specifically focusing on the effect of strain waveform shape on fatigue life. Similarly to last year’s results (PVP2016-63294) a distinct effect of strain waveform, presently inadequately accounted for in Fen predictions, was observed.
New Approach on Solid State Joining of Stainless Steel Tube to Tube Sheet Joints
Authors: Roos Arne , Winkler Marcus , Wimmer Georg , Dos Santos Jorge F. , Hanke Stefanie 
Published: October 2017

Hybrid friction diffusion bonding (HFDB) is a solid-state bonding process first introduced by Helmholtz-Centre Geesthacht, Germany, to join aluminium tube-to-tube sheet joints of coil-wound heat exchangers (CWHE) for liquefaction of natural gas (LNG). This study describes how HFDB was in a first step successfully transferred to create austenitic S32100 single hole tube-to-tube sheet joints. Process parameters are presented and results from subsequent non-destructive bubble leak testing and destructive tensile pull-out testing are discussed. After pull out testing the bonded areas were further investigated using optical microscopy as well as scanning electron microscopy.
Leak tight joints were generated due to the formation of a metallic bond close to the planar friction area of the employed tools, and failure in pull-out tests occurred by ductile fracture. The results show that the HFDB approach developed for Al-alloys may well be transferred to steel, and in the future possibly to other high-temperature alloys. It thereby offers an alternative route for joining tube to tube-sheet connections in solid state, with the corresponding advantages, such as no open flames or arc, no spatter and no need for filler material.
Review and Comparison of Buckling Methodologies for ASME B&PV Code Linear Piping and Component Restraints
Authors: Wiseman Phillip , Hoch Zara Z. , Nargund Shrikant S. 
Published: October 2017

Piping and component restraints are required to follow the design requirements as mentioned in ASME Boiler and Pressure Vessel Code, Section III, Subsection NF. One of the requirements indicates the necessity of calculating the critical buckling stresses for the members that are subjected to a compressive loading. This paper discusses the prescribed requirements in the Code that specifically address the considerations of the stability and buckling load capacities of linear piping and component restraints (i.e., struts). The finite element modeling of various strut geometries and the results of the buckling analyses of slender structural members (slenderness ratio, Kl/r, greater than or equal to 100) using various finite element solution techniques are presented herein. Specifically, three types of finite element analyses are conducted in an effort to define the critical buckling load for the subject structural member. These three finite element analyses include the traditional linear (Eigen value) Euler method; the nonlinear, second order large deformation method; and finally, the nonlinear large deformation method that incorporates nonlinear elastic-plastic material behavior.
Additionally, two closed form solutions using strain energy method and Euler-Bernoulli beam theory are conducted on the same strut geometries. The results obtained from the aforementioned techniques are then compared both numerically and qualitatively with an appropriate explanation of the purpose and usefulness of each particular result with respect to the intent of the ASME B&PV Code, Section III, Subsection NF requirements. The results show significant variations based on differences in the assumptions and techniques employed in the respective analyses and simply applying the identical margin of safety to each technique does not yield consistent outcomes.
As a result of the discussion we get an insight about the axial compression allowable stress equations and factors as defined in the ASME B&PV Code and how they should be refined depending on the type of buckling analysis we choose to conduct.