bluuebunny/crossref_metadata_2025_split huggingface dataset:
Drop Testing of a Container for the Storage, Transport and Disposal of Intermediate Level Waste
Authors: Quercetti Thomas , Müller Karsten , Johnson Mark , Tait Trevor , Tso Chi-Fung , Izatt Conrad 
Published: October 2017

Impact tests were performed, as part of a corresponding container’s substantiation, during design development of a shielded container. The container will be used for storage, transport, and disposal of intermediate level waste in the UK.
The mechanical test program comprised a 9m free drop test onto an unyielding target in a container long lid edge down orientation at ambient conditions. Further, a 0,5m free drop test onto a punch target was performed. Here, the container was orientated with the lid downwards in a way that the punch, a mild steel bar, impacts a filter lid.
The test specimen was instrumented with strain gauges and accelerometers for the drop tests. Transient strains at selected points of the inner and outer container walls, at the shielding lid, as well as at the lid bolts were measured during the container’s impact. Furthermore, decelerations of the container body, container lid, and the skip were measured. The complex geometrical changes of the container due to impact were determined by optical 3d-deformation measures using the projected fringes method in combination with multi-image photogrammetry.
This paper summarizes the performance of the drop tests and various drop test results in context with the design development as well as aspects regarding the associated Finite Element (FE) analyses and post-test evaluation.
Earthquake Resistance Assessment of Contaminated Water Storage Tank and its Reinforcement Basic Design at Fukushima
Authors: Morishige Haruo , Fujita Katsuhisa , Yamashiki Yousuke 
Published: October 2017

The core melt accidents at Fukushima Daiichi Units 1-2-3 have occurred due to the tsunami of the Great East Japan Earthquake at March 11, 2011. Now a coolant of approximately 100t is injected into each Reactor Vessel (RPV) in order to cool the fuel debris, and a part of the coolant leaks out from the water. The contaminated water is collected, and it is stored in the contaminated water storage tanks more than approximately 1,000 and the total amounts of coolant exceeds 1,000,000 t. Although the contaminated water is purified by ALPS, it is still in high nuclear pollution density. According to the homepage of Nuclear Regulation Authority (NRA) in Japan, the storage tanks have been placed on the concrete bed without any anchor. Tokyo Electric Company (TEPCO) reported to NRA in 2015 that the design acceleration used for the overturning evaluation of the tanks have 0.3G. At the Fukushima Daiichi Power Station in the Great East Japan Earthquake, the maximum horizontal acceleration was 0.6G, the maximum vertical acceleration was 0.326 G at the time of the Great East Japan Earthquake. Even if an earthquake that is a fraction of this earthquake comes back, there is a possibility that this tank will fall over. The temporary special design guideline is adopted. These tanks have already stored the contaminated water for 6 years. However, it is thought that the more long term countermeasures is necessary for preventing the contaminated water from flowing out into the Pacific Ocean when another great earthquake will attack this area in future. In this paper, we propose a drastic structure to add a foundation without causing ground to the present tank.
In the past, the member fastening the steel plate of the upper structure of the tank and the concrete of the lower structure is a foundation bolt, but this time it is a stud welded to the tank side plate. This made it possible to flexibly design and reinforce the foundation itself even after tank installation.
On the Critical Boundary Conditions for Rupture of Buckled Steel Pipelines
Authors: Mohajer Rahbari Nima , Cheng J. J. Roger , Adeeb Samer 
Published: October 2017

The environmental contamination due to the leakage of energy pipelines is a serious hazard to the public property and safety. Hence, any premature rupture should be dealt with in the design and the operating mode of steel pipes. A large amount of complexity is involved in the soil-pipe interactions that makes it so challenging to discover the physical boundary conditions (BC) applied to the buried pipelines during differential ground movements. Therefore, the most critical boundary conditions of buried pipes should be conceived based on the probable mechanism of soil-pipe interactions and considered in the experimental and analytical simulations of rupture. The focus of the current research is to address the critical boundary conditions that can trigger the rupture of underground wrinkled pipelines whilst being subjected to a monotonic increase of curvature.
Finite element (FE) simulation of a full-scale bending test on a pressurized X70 line pipe specimen conducted at the University of Alberta is implemented. Cumulative fracture criterion coupled with the equivalent plastic strain to fracture for X70 steel grade is fed into the analysis to predict the ductile crack formation in the pipe’s body. The FE model is verified by the experimental data and is used to study the critical soil-pipe interactions that provoke the rupture of buckled steel pipes on the tensile side of the cross-section under increasing bending curvature. The results of this study suggest that the pipelines which are restricted from axial displacements are extremely vulnerable to experience a rupture along their post-buckling loading path. And so are the pipelines in which tensile axial force is developed due to soil-pipe interactions, e.g. pipelines in arctic regions that are installed during the summer time.
Assessment of Powder Metallurgy-Hot Isostatic Pressed Nozzle-to-Safe End Transition Joints
Authors: Sutton Ben , Gandy David 
Published: October 2017

Dissimilar metal nozzle-to-safe end welds represent a complex manufacturing operation for nuclear reactor fabrication. Transitioning from a low alloy steel nozzle to corrosion-resistant reactor coolant system piping systems requires weld buttering, an intermediate heat treatment, and a dissimilar metal weld between the nozzle buttering and safe-end. The process can be both time consuming and costly, and often provides difficulties for inspection of welds in the field. Through the use of powder metallurgy and hot isostatic pressing (PM-HIP), a series of transition joint configurations have been evaluated that could reduce the complexity of nozzle-to-safe end fabrication by eliminating fusion welding from the process.
Test coupons of SA508 low alloy steel have been joined via PM-HIP to either 316L SS or 347NG SS using various powder metal interlayer materials (Alloy 82, Nb-modified Alloy 600 (600M), Alloy 690, and 316L SS). The microstructural evolution and mechanical performance of the joints have been evaluated following a post-HIP solution anneal, quench and temper heat treatment. Particular attention was given to evaluating bond-line regions via Charpy V-notch impact testing, tensile testing, microhardness, optical microscopy, scanning electron microscopy, and energy dispersive X-ray spectroscopy. Results demonstrate that multiple transition layer configurations are capable of meeting the acceptable tensile and impact energy specifications of the two parent base material substrates. Interface impact toughness and ductility were dependent upon secondary phase precipitation within the diffusion region of the joints and presence of non-metallic inclusions. This assessment demonstrates that acceptable joint performance is achievable through proper material selection and should be considered for DMW applications in the future.
Development of Process Induced Residual Stress During Flow Forming of Tubular 15-5 Martensitic Stainless Steel
Authors: Khayatzadeh Saber , Moturu Shanmukha Rao , Kelleher Joe F. , Rahimi Salah 
Published: October 2017

Flow forming is a near net shape process for manufacturing of dimensionally accurate hollow components such as shaft in gas turbines, that is currently at its development stage for aerospace industry. The process has several advantages such as reducing material wastage, extremely fast manufacturing time, and eliminating extra manufacturing processes such as machining. Due to the nature of this complicated cold deformation process, significant magnitude of residual stress is introduced into the component. Understanding the magnitude and distribution of residual stress is essential to tailor the flow forming process to achieve parts within dimensional tolerances and desired mechanical properties. The present research is aiming to explore the generation and evolution of residual stress at various stages of flow forming process in a tubular component made from martensitic 15Cr-5Ni stainless steel, using different techniques of neutron scattering, x-ray diffraction (XRD) and hole-drilling based on electronic speckle pattern interferometry (ESPI).
Residual stress measurements were carried out in preformed and flow formed components at surface, near-surface and in the bulk of components using XRD, ESPI based hole-drilling and neutron diffraction techniques. These measurements were conducted at different levels of reduction in the thickness of the original part (i.e. after 20% and 40%), by applying identical forming parameters for all samples. The XRD results show significant change in hoop and axial residual stress levels with a reduction in the wall thickness. This is more pronounced for the axial component where the average stress switches from relatively high tensile (∼ 450MPa) in the original part to significant compressive stress (∼ −600MPa) in the formed part, after 20% of reduction. The bulk residual stress components measured in the middle of thickness of the parts, using neutron scattering, show a general increase in the magnitude of residual stress by higher level of deformation (i.e. reduction in the wall thickness). The measured bulk stress components through the thickness were tuned to tensile after reducing the wall thickness by 40%. The results of XRD and neutron diffraction stress measurements suggest that the residual stress along the length of the samples (i.e. axial direction) is consistent with ±800 MPa and ±400 MPa after 20% and 40% reduction by forming process, respectively.
The results of ESPI based hole-drilling show tensile hoop residual stress (≈600 MPa) and an abrupt fluctuation (i.e. tension-compressive-tension) in the axial residual stress near the surface of the part following flow forming. The stresses measured by ESPI based hole-drilling are complementary to the results of the XRD on surface and neutron diffraction in the bulk to reconstruct the residual stress profile form the surface through to the bulk.
Error Norm vs. Uncertainty Metric in Assessing Accuracy of the Finite Element Method
Authors: Marcal Pedro V. , Fong Jeffrey T. , Rainsberger Robert , Ma Li 
Published: October 2017

In most finite-element-analysis codes, accuracy is achieved through the use of the hexahedron hexa-20 elements (a node at each of the 8 corners and 12 edges of a brick element). Unfortunately, without an additional node in the center of each of the element’s 6 faces, nor in the center of the hexa, the hexa-20 elements are not fully quadratic such that its truncation error remains at h(0), the same as the error of a hexa-8 element formulation.
To achieve an accuracy with a truncation error of h3(0), we need the fully-quadratic hexa-27 formulation. A competitor of the hexa-27 element in the early days was the so-called serendipity cubic hexa-32 solid elements (see Ahmad, Irons, and Zienkiewicz, Int. J. Numer. Methods in Eng., 2:419-451 (1970) [1]). The hexa-32 elements, unfortunately, also suffer from the same lack of accuracy syndrome as the hexa20’s.
In recent work, we have developed methods to test the errors and the rate of convergence in FEA [2,3,4].
In this paper, we propose a new metric for determining the quality of isoparametric elements a priori. Significance of the highly accurate hexa-27 formulation and a comparison of its results with similar solutions using ABAQUS hexa-20 elements, are presented and discussed. Guidelines are proposed for selection of better elements.
Regulatory Perspectives on Fitness for Service Assessments of CANDU Pressure Boundary Components
Authors: Jin John C. , Carroll Blair 
Published: October 2017

Major pressure boundary components such as pressure tubes, feeder pipes and steam generators at some Canadian CANDU nuclear power plants are entering periods of extended operation beyond their initially assumed operating life. The Canadian Nuclear Safety Commission (CNSC) has approved their long term operations based on the assurance of fitness for service of those components which was composed of condition assessments and aging management programs carried out and implemented by the Canadian licensees. The condition assessments were conducted to demonstrate that components would be within their design basis for the period of intended long term operation and the aging management program was implemented to ensure that the conditions of the components would be maintained as evaluated in the condition assessments and to monitor if new degradation mechanisms would become active.
This paper presents the CNSC regulatory practice adopted in the course of technical reviews of fitness for service assessments of major pressure boundary components conducted by Canadian nuclear licensees to demonstrate the safe long term operation of major components can be achieved.
Considerations for the Use of Probabilistic Assessments in Regulatory Decision Making Related to Pressure Boundary Component Aging
Authors: Carroll Blair , Jin John C. 
Published: October 2017

Within the current Canadian regulatory framework, the structural integrity of pressure boundary components with detected service-induced degradation must be demonstrated using deterministic evaluation techniques. However, Canadian Nuclear Safety Commission staff has recognized that the inherent conservatism in these deterministic assessment approaches may generate overly conservative conclusions when they are applied to assess the impact of postulated service-induced degradation to establish aging management requirements for nuclear power plant pressure boundary components. This may have the unintended effect of reducing the effectiveness of aging management programs by directing resources towards activities that will have minimal benefit on improving plant safety and could result in unnecessary dose to personnel. With this in mind, CNSC staff has accepted the limited use of probabilistic assessments prepared by licensees to support aging management activities for pressure boundary components. These probabilistic assessments form a part of risk-informed decision making strategies intended to reduce excess conservatism that could arise if decisions are based solely on the results of deterministic assessments.
This paper provides an overview of CNSC staff’s experiences with the review and acceptance of licensee submissions incorporating probabilistic assessments of pressure boundary component aging for risk informed decision making.
Ellipsoidal Head Rules: A Comparison Between ASME Section VIII, Divisions 1 and 2
Authors: Seipp Trevor G. , Barkley Nathan , Wright Christopher 
Published: October 2017

In ASME Section VIII, Division 1, rules are provided for calculating the thickness of 2:1 ellipsoidal heads in UG-32. UG-32(c) also states that “an acceptable approximation of a 2:1 ellipsoidal head is a torispherical head with a spherical radius of 0.9D and a knuckle radius of 0.17D”. However, calculating the thickness of a torispherical head with those “equivalent” dimensions results in a thicker head. This result is inherently inconsistent, which starts to bring into question the so-called equivalency.
Code Case 2260 further perpetuates this equivalency by providing alternative rules for calculating the thickness of torispherical heads, and then permitting the engineer to calculate 2:1 ellipsoidal heads implementing this 90-17 equivalency.
Additionally, the calculation methodology for a 2:1 ellipsoidal head in ASME Section VIII, Division 2 uses the torispherical head calculation methodologies and directly implements this 90-17 equivalency. However, this calculation method results, for the same allowable stress basis, in a completely different thickness from the above three methods.
This paper reviews the past 90+ years of work on this topic, and presents some theoretical treatment of the different head geometries. A review of the current Code rules is presented, with a comparison of results for several sizes. A survey of head fabricators is presented to show the actual geometries produced for use in ASME pressure vessels. Finally, conclusions regarding whether or not the 2:1 ellipsoidal head is in fact equivalent to the 90-17 torispherical head are presented, and recommendations for future revisions to both ASME Section VIII, Division1 and Division 2 are provided.
The Effects of Tube Array Geometry on Fluidelastic Instability in Heat Exchanger Tube Arrays in Cross Flow
Authors: Hassan Marwan , Weaver David S. 
Published: October 2017

The shut-down of the San Onofre Nuclear Generating Station (SONGS) has been attributed to damaging streamwise Fluidelastic Instability (FEI) of the steam generator tubes, a phenomenon which has traditionally been assumed not to occur. This has generated a significant research effort to better understand this phenomenon and to develop appropriate design criteria for its prevention. Most current design codes are based on Connors criterion for FEI which neglects both streamwise FEI and the effects of tube array pattern and pitch ratio. It is becoming clear that array geometry and pitch ratio are important determining factors in FEI, especially in the streamwise direction.
This paper presents an extension of the theory of Lever and Weaver to consider arrays of flexible fluid-coupled tubes which are free to become unstable in both the transverse and streamwise directions. This simplified modelling approach has the advantages of being very tractable for numerical parametric studies and having no need for experimental data input. Previous research by the authors has shown that the predictions of this model agree very well with the available experiments for parallel triangular arrays for both transverse and streamwise FEI. In this paper, the results of such studies are presented for the both transverse and streamwise FEI for square inline and normal triangular arrays and compared with the authors’ previous results for parallel triangular arrays. It is shown that FEI is strongly influenced by array geometry, especially for small pitch ratio arrays operating at low values of the mass damping parameter. The results show good agreement with the available experimental data.
Probabilistic Structural Analysis for Aging Management of Reactor Internals Bolting
Authors: Troyer Greg , Waskey Christopher , Somers Jarrett , Fyfitch Stephen 
Published: October 2017

In the United States Pressurized Water Reactors, the reactor internals are subject to augmented inspections under the auspices of MRP-227 to ensure safe operation through periods of extended operation. Recent operating experience has revealed degradation of baffle-to-former bolts (BFBs) where emergent and lengthy shutdowns were required to replace non-functional BFBs. A pro-active approach to safely managing BFB degradation requires an integrated analysis approach that addresses uncertainties in non-destructive examination (NDE) techniques, and uncertainties in bolt degradation rate and bolt degradation location. These uncertainties present nearly unlimited permutations of potential degraded bolt conditions, requiring probabilistic treatment to demonstrate structural integrity of the reactor internals assembly over the next inspection interval.
In this paper, we present almost 10 years of experience with probabilistic structural analysis of bolted connections in nuclear reactor internals applications. Modeling techniques for bolt degradation, location of degraded bolts, and uncertainty in NDE techniques are discussed. Techniques for rapidly evaluating millions of bolt failure patterns against the structural design basis are discussed. These techniques permit definition of requirements for an initial examination, requirements for reexamination interval, and prescription replacement bolt patterns optimized to meet the operator’s financial and risk management objectives, while ensuring safety with rigorous demonstration of maintaining design basis.
Crack Analysis of a Pressure Vessel Using the API 579-1/ASME FFS-1
Authors: Carvajalino Jose de Jesus L. , Freire José Luiz F. , Paiva Vitor Eboli L. , Maneschy José Eduardo , Diaz Jorge G. , Rudolph Jürgen 
Published: October 2017

This paper presents a structural integrity evaluation of a duplex stainless steel pressure vessel containing several flaws detected in a longitudinal weld. The evaluation had the objective of determining whether the pressure vessel was suitable to continue in operation or whether it should be immediately repaired or even replaced. Due to timely issues, a first analysis was conducted in accordance with the 2007 edition of the API 579-1/ASME FFS-1 Standard [1]. A second analysis was later repeated based on the 2016 edition [1]. Results obtained from both analyses were compared and presented relevant differences caused by the other calculation procedures used to determine residual stresses generated in the longitudinal welding.
The assessment was based on the Failure Assessment Diagram (FAD). The existing indications were detected by ultrasonic examination and were located in one longitudinal weld. The assessment evaluations used stress intensity factors for the opening mode I, KI, obtained for two cases: 1) the combination of the several supposedly interacting cracks into an equivalent crack using the interaction criteria presented in [1]; 2) the allocation of the multiple cracks into a finite element model that took into consideration, more realistically, the interaction among the individual cracks. The total loads and stresses considered in the analysis resulted from a superposition of the design pressure stress and the residual stresses induced by the welding process.
Due to lack of information on the material fracture toughness for the duplex stainless steel used in the vessel, the material toughness was estimated using a lower bound value suggested in [1] for common welded stainless austenitic steels, although higher values can be predicted for duplex steels by extending the use of a transition master curve as presented and discussed elsewhere [2–7] and by employing specific Charpy test results for the vessel material.
One of the key aspects of the problem was the calculation of the residual stress distribution imposed by the welding process. Two procedures were adopted: one available in the API/ASME Standard issued in 2007, and the other in the 2016 release. The results presented in this paper have demonstrated that the limits of the Standard 2007 are conservatively satisfied when the Level 3 assessment is applied. The re-analysis of the vessel when subjected to the residual stress distribution presented in the newest 2016 edition leads to consider the vessel safe under an assessment Level 2. The overall conclusion was that the damaged pressure vessel could continue in service under restrictions of the development of an inspection plan to verify the absence of future crack growth.