bluuebunny/crossref_metadata_2025_split huggingface dataset:
Development of Mechanism for Preventing Derailing of Material Handling Equipment
Authors: Iwamoto Kosuke , Sato Yuji , Otoyo Teruyoshi , Horiuchi Munenori , Sakai Hidetoshi , Arai Haruki 
Published: December 2016

All of the traveling cranes at the port of Kobe were damaged by the strong ground motion of the Southern Hyogo Prefecture Earthquake in 1995. Seismic isolation systems for traveling cranes were developed to increase earthquake resistance against strong ground motions. In general, the isolation systems for traveling cranes for strong ground motions are very large. Quays which are not reinforced have a risk of being damaged by strong ground motions, and require traveling cranes to prevent derailing during middle strong motions, because derailing leads to collapse of the traveling cranes. A mechanism for preventing derailing of material handling equipment was developed for middle strong ground motions.
A middle rocker beam of the new mechanism is divided into two beams, and uplift motion of a leg of a crane can be absorbed by rotating motions of the two beams. This allows the mechanism for preventing derailing to be easily installed into the established equipment. Seismic analyses using an finite element model of the traveling crane and vibration tests were conducted to verify the validity of the design and derailing prevention performance. The results of analyses and vibration tests showed that the mechanism for preventing derailing can prevent wheels from derailing during middle strong earthquakes. The configuration of the mechanism and the results of the analyses and vibration tests are presented in this paper.
Fracture-Mechanics Analysis of the Reactor Pressure Vessel Beznau 1 Based on Thermal-Hydraulics Input Data From KWU-MIX and CFD Analyses
Authors: Nicak Tomas , Trewin Richard , Keim Elisabeth , Cremer Ingo , Blasset Sebastien , Hägeli Heinz 
Published: December 2016

The integrity of a reactor pressure vessel (RPV) has to be ensured throughout its entire life in accordance with the applicable regulations. Typically an assessment of the RPV against brittle failure needs to be conducted by taking into account all possible loading cases. One of the most severe loading cases, which can potentially occur during the operating time, is the loss-of-coolant accident, where cold water is injected into the RPV nearly at operating conditions. High pressure in combination with a thermal shock of the ferritic pressure vessel wall caused by the injection of cold water leads to a considerable load at the belt-line area known as Pressurized Thermal Shock (PTS).
Usually the assessment against brittle failure is based on a deterministic fracture-mechanics analysis, in which common parameters like J-integral or stress intensity factor are employed to calculate the load path for an assumed (postulated) flaw during the PTS event.
The most important input data for the fracture-mechanics analysis is the transient thermal-hydraulics (TH) load of the RPV during the emergency cooling. Such data can be calculated by analytical fluid-mixing codes verified on experiments, such as KWU-MIX, or by numerical Computational Fluid Dynamics (CFD) tools after suitable validation. In KWU-MIX, which is the standard used for TH calculations within PTS analyses, rather conservative analytical models for the quantification of mixing and, depending on the water level, condensation processes in the downcomer (including simplified stripe and plume formations) are utilized. On the contrary, the numerical CFD tools can provide best-estimate results due to the possibility to consider more realistically the stripe and plume formations as well as the geometry of the RPV in detail.
In order to quantify the safety margin inherent to the standard approach, two fracture-mechanics analyses of the RPV Beznau 1 based on thermal-hydraulic input data from KWU-MIX and CFD analyses were performed. Subsequently the resulting loading paths were compared between each other and with material properties obtained from the irradiation surveillance program of the RPV to demonstrate the exclusion of brittle-fracture initiation.
Optimization of High Pressure Components for LDPE Manufacturing
Authors: Blome Matthias , Rueckriem Stefan 
Published: December 2016

Coned and threaded type connection valves, fittings and tubing up to sizes of 9/16” are typical high pressure components used on the initiator injection system of a Low Density Polyethylene LDPE facility. The process pressures of up to 360 MPa (52,200 psi) on the reactor side require even higher pressures on the initiator injection lines. Based on the differential pressures and the challenging environment 414 MPa (60,000 psi) has been defined as an appropriate pressure rating for such components.
Valves, fittings and tubing with this pressure rating are more or less standardized and available from multiple sources on the market. However, the technical challenges in designing and manufacturing components for this duty are sometimes under-estimated and design improvements can lead to a higher process safety and a reduced downtime. This paper covers the design requirements and has a specific focus on a filter with screwed element and a check valve with ceramic ball.
The majority of high pressure in-line filters on the market have a sintered element that is press-fitted into the filter body. This design has some disadvantages in case of fast depressurizations as they occur when a safety valve or a rupture disk blows off. The sudden differential pressure may cause the element to move inside the body, and this can be accompanied by cracking of the element. Parts of the broken sintered element are a risk to the process since they would move downstream followed by a contamination of the reactor.
A new generation of such filters with screw-in elements was designed and tested. The R&D work resulted in an extended element that fits into a modified filter body with standard outside dimensions. Before performing a field test at a LDPE plant three prototype samples went through intensive testing at the manufacturer’s facilities.
Check valves are typical components used in order to maintain the flow in one direction. For liquids the metal to metal sealing ball type is the most common. Since the process pressure exposes the mating parts — ball and orifice — to high compression forces, imprints and marks on the ball surface are sometimes experienced. The material of these types of balls is typically bearing steel which has high mechanical strength but limited corrosion resistance. In order to improve the cycle life of high pressure ball check valves alternative materials were looked at. During the further product improvement a check valve with ceramic ball was designed and lab tested at the manufacturer’s facilities.
Influence of Residual Stress due to Prior Overloading on the Fatigue Life of a Notched Plate
Authors: Karpanan Kumarswamy 
Published: December 2016

During autofrettage, pressure vessels are subjected to high internal pressure, causing the internal wall to yield plastically. When the internal pressure is released, the inner wall of the vessel develops compressive residual stress. Similarly, when a subsea component is hydrotested, some of the highly stressed regions yield during hydrotesting and, when the pressure is released, these regions develop compressive residual stress. Fatigue life is greatly influenced by local stress on the component surface. Fatigue crack initiation primarily depends on the cyclic stress or strain and the residual stress state. Tensile residual stress decreases fatigue life and the compressive residual stress significantly increases fatigue life. This is true for both fatigue crack initiation and propagation.
In this paper, effects of residual stress on a notched plate are studied by subjecting it to an initial overload cycle and subsequent low loading cycles. Tensile and compressive overloads on the notched plate induce compressive and tensile residual stresses, respectively. An elastic-plastic finite element analysis (FEA) was performed to simulate the overload and low loading cycles on the notched plate. The stress and strain from the FEA is used to perform strain-based fatigue analysis. ASME VIII-3, Brown-Miller (B-M), Maximum shear strain, Socie-Bannantine, and Fatemi-Socie methods are used for calculating the fatigue life of the notched plate. Fatigue life predicted by both stress and strain methods matches well with the test fatigue data.
The Method for Structural Reliability Estimation of the Heat Exchanger Tubes of Steam Generator of WWER NPPs
Authors: Orynyak Igor , Zarazovskii Maksym , Borodii Mykhaylo 
Published: December 2016

After the replacement due to the well known problem of weld seam #111 the steam generators (SG) of the Ukrainian WWER-1000 type nuclear power plants (NPP) have been in operation for almost 25 years. So now, the main practical problem related to the SG is the leakage/breaking of heat exchanger tubes (HET). If the leakage criterion is not satisfied the NPP Unit has to be stopped until the corresponding HET is plugged, which leads to unexpected huge financial losses.
Every year at least 12.5% of the SG HETs are controlled by eddy current inspection, after which some of the HETs are plugged according to the appropriate plugging criteria (in most cases – 65% loss of tube wall thickness). Thus, every SG has its own statistics of plugging.
From the practical point of view the Utility is interested in the following questions: whether the failure of the HET happens during the normal operation mode (NOM) up to the next NPP stop or not; what is the probability of this event; what is the defect growth rate? Answers on these questions could give us the opportunity to increase the effectiveness of the HET in-service inspection and to improve the plugging criterion.
A simple statistically-based method for HET integrity assessment is proposed. The method based on an exponent distribution law and HET plugging statistics of specific SG, taking into account defect growth. Based on the history of the tubes plugging (year of operation versus quantity of pluggings) the three statistical parameters inherent to this specific SG have to be found: initial number of defects, parameter of the exponential distribution (initial defect size) and the defect growth rate.
The developed method was used for the prediction of HET leak/break number for all Ukrainian WWER-1000 SG. It is shown that for those SG which has less than 2% HET pluggings the accumulated pluggings as of the end of the next year of NOM will not exceed this value.
Based on the Ukrainian history of WWER-1000 SG operation the database of HET plugging is presented.
This method is also used for justification of pressure reduction of periodic hydrostatic test (HT) for primary circuit of WWER-1000 and WWER-440 NPPs from 24.5 MPa to 20.3 MPa and from 19.1 MPa to 15.7 MPa, correspondingly. To justify a reduction of the HT pressure, a quantitative, risk-informed assessment of HT effectiveness has been performed with taking into account the HT pressures change. The HET failure probability is calculated as the proportion of defects which exceed critical sizes for NOM and HT. Limit load models are used for the determination of the critical defects size. The variation of reliability is calculated as the difference between fracture probability during NOM after HT at routine and reduced pressures. It is shown that HT pressure reduction does not practically increase the fracture probability during operation, and satisfies the criteria of risk change.
Ratcheting With No Thermal Bending and Membrane Stress
Authors: Adibi-Asl R. , Reinhardt W. 
Published: December 2016

The ASME B&PV Code provides design by analysis rules that address failure mechanisms under cyclic loading. One of these potential failure mechanisms is incremental plastic collapse, or ratcheting. Miller presented the technical basis for the present Code requirements in a technical paper in 1959. Miller’s equations for the ratchet boundary address a beam under a cyclic through-thickness thermal gradient acting together with a steady axial mechanical load. This ratchet boundary applies approximately to a pressurized cylinder with through-thickness thermal bending stress. Conditions arise sometimes in practice where cooling or heating is applied simultaneously to the inner and outer surface of pressure boundary. The extreme case of such a scenario arises when both surfaces experience the same thermal condition such that there is a cyclic thermal stress but both zero membrane thermal stress and zero thermal bending stress The question is, could ratcheting occur in this case?
This paper derives the ratchet boundary for cases when the maximum temperature occurs mid-way through the thickness. The linearized stress due to thermal loading is zero. The solution is obtained using FE analysis and the Non-Cyclic Method (NCM) that has been proposed previously by the authors. The NCM is a generalization of the static shakedown theorem and allows the ratchet boundary to be calculated for both elastic and elastic-plastic cyclic stress states.
Effect of Adjacent Tube Expansion on the Residual Contact Pressure of Tube-to-Tubesheet Joints
Authors: Bouzid Abdel-Hakim , Zhu Linbo 
Published: December 2016

The tubular heat exchangers of nuclear, oil and gas industries often experience shutdowns due to failures of the tube-to-tubesheet joints. They are the very weak link between the heat transfer process thermos-fluids and separation boundary. The complex elasto-plastic behavior of the expanded joint and the lack of a proper design procedure of the expansion process are the two main factors contributing to the low quality and low performance of this kind of assembly. The inadequacy of a proper evaluation of the residual contact pressure achieved by the expansion process make the joint vulnerable and prone to leakage failure. Among the multiple parameters that effect the residual contact pressure is the adjacent tube-to-tubesheet expansion.
It is proposed to study the elasto-plastic behavior of tube-tube to tubesheet joints during the expansion process using numerical FEM. The numerical FE model will be validated analytically prior to study the influence of adjacent expansion. Particular emphasis will be put on the effect of adjacent tube expansion and the expansion order sequence on the residual contact pressure. The effect of ligament size and the adjacent tube expansion sequence are also investigated. Such a study would provide general guidelines on how to control the expansion process and help the design engineer to predict accurately the residual stresses that are the major source of stress corrosion cracking and leakage failure.
Pressurizer Safety Valve Data Collection, Analysis, and Performance
Authors: Esselman T. , Zysk G. , Streeter P. , Elaidi B. , Garcia H. , Khatri S. , Kewalramani M. , Brackeen E. , Mortorff A. 
Published: December 2016

The Pressurizer Safety Valves (PSVs) at Diablo Canyon Power Plant were manufactured by Crosby and are model 6M6 HB-86-BP valves. The valves have intermittingly experienced leakage as the units are heated up and pressurized to normal operating conditions. When leakage occurs, a time-consuming de-pressurization and re-pressurization process is implemented that causes significant delays in the return to power after an outage. Following the elimination of the upstream water seal, leakage has occurred on individual valves during all but one start-up.
After recent PSV leakage, instrumentation was installed to obtain information during the subsequent unit heat up and pressurization. The instruments included thermocouples on the inlet pipe, the discharge pipe, and the valve. Data was collected during a unit start-up in which two of the three installed valves leaked.
Piping and valve analyses were performed using the data to modify the valve installation procedures, to modify the process for accommodating piping thermal growth, and to modify the pressurization rate during start-up. Upon implementation of these changes, none of the PSVs leaked during the subsequent unit heat up and pressurization, saving more than 24 critical path hours.
The data, analyses, and modifications will be described along with the analyses used to support an increase in the pressurization rate for unit start-up.
Ultrasonic Inspection of Hydrogen Flakes in Large Forgings: From Qualification to Site Inspection
Authors: Moussebois Dominique , Ancrenaz Patrick 
Published: December 2016

During 2012 RPVs inspections, nearly-laminar indications were detected in the lower and upper core shells of Doel 3 and Tihange 2 RPVs. As a consequence, the Doel 3 and Tihange 2 NPPs have to stay in cold shutdown until it was proved that they can be safely operated. The performance and capability of the applied UT technique to detect and characterize hydrogen flakes had to be confirmed. To achieve this objective, extensive investigations were launched on blocks extracted from a steam generator shell known for containing hydrogen flakes.
The paper describes the UT validation performed and the results of the destructive tests which enable to demonstrate the capability of a straight ultrasonic beam to detect and size nearly laminar indications and to correctly identify ligaments between two adjacent indications. It addresses the formal extension of qualification performed by correlating UT measurements acquired on a large numbers of real hydrogen flakes, simulations of UT responses and destructive examinations.
Lessons learned from the experiments are finally described.
Overview of the Doel 3 and Tihange 2 Reactor Pressure Vessel Safety Cases
Authors: De Smet Michel , Van Vyve Jean 
Published: December 2016

During the summer outages of 2012, large numbers of indications were found inside the shell material of the Doel 3 and Tihange 2 reactor pressure vessels (RPV). Therefore both plants remained in cold shutdown with their core unloaded. A series of examinations, tests and inspections were performed leading to the conclusion that the indications are hydrogen flakes and that they do not affect the structural integrity of the RPVs, regardless the operating mode, transient or accident condition. All this was documented in the Safety Case Reports issued in December 2012 and the Safety Case Addenda issued in April 2013.
Based on those reports, the Belgian Federal Agency for Nuclear Control (FANC) authorized the restart of both units that went back on-line in June 2013. In parallel, a number of requirements from the FANC were addressed such as the qualification of the applied Ultrasonic Testing (UT) procedure, and mechanical testing of irradiated specimens containing hydrogen flakes. The preliminary tests showed unexpected results regarding the shift in RTNDT (Reference Temperature for Nil Ductility Transition) under irradiation that could not confirm the hypotheses considered in the initial Safety Case Reports. Therefore, the Licensee Electrabel decided to shut down both plants immediately.
In order to fully address this concern, the material test programme was extended including several RPV materials and covering additional irradiation campaigns. This led to a modification of the irradiation embrittlement trend curves considered in the structural integrity analysis. In addition, the qualification of the UT procedure led to an updated cartography of the flakes. The structural integrity assessments of both RPVs were revised accordingly. The final Safety Case Reports, confirming the fitness-for-continued operation of both RPVs, were submitted to the FANC in October 2015. FANC allowed restart of both units on November 17th, 2015.
The paper gives a historical overview of the Doel 3 and Tihange 2 RPV Safety Cases and explains how the roadmap was built in order to demonstrate the RPV’s structural integrity in the presence of hydrogen flakes.
Measurements of Residual Stresses Using Mechanical Strain Relaxation Methods and Mapping in a Thin-Walled Girth Welded Pipe Mock-Up
Authors: Serasli Karim , Romac Remi , Cave Douglas 
Published: December 2016

Girth welded pipes such as those located offshore on platforms in the North Sea are subjected to highly corrosive environment. The need to consider welding residual stresses in the assessment of the fitness for service and damages to these pipes when investigating local corrosion damages across a welded region is therefore important for the operators of the platforms and the manufacturers of the pipes.
This paper presents a review of work carried out to ascertain the welding residual stresses present within a thin-walled girth welded pipe mock-up made from steel API 5LX Grade 52. The mock-up was manufactured to replicate typical pipes used to convey gas, oil and water through the platforms. The mock-up was of diameter 762mm and of thickness 19mm. The incremental deep hole drilling (iDHD) technique and the contour method were applied to characterize the residual stresses in the weld and heat affected zone of the specimen. The results of these measurements are presented and compared to highlight agreements and discrepancies in the measured residual stress distributions using these different techniques.
Most residual stress measurement methods are limited in terms of their stress and spatial resolution, the number of measurable stress tensor components and their quantifiable measurement uncertainty. In contrast, finite element simulations of welding processes provide full field distributions of residual stresses, with results dependent on the quality of the input conditions available. As measurements and predictions are not often the same, the true residual stress state is therefore difficult to determine. In this paper, through-thickness residual stress measurements are made using the contour and iDHD methods and these residual stresses measured using the iDHD technique are then used as input to a residual stress mapping technique provided within a finite element analysis to reconstruct the residual stress field in the whole specimen. The technique is applied iteratively to converge to a balanced solution which is not necessarily unique. The solution can then be reused for further simulations and residual stress analyses, such as corrosion simulation. Results of the reconstruction are presented here.