Abstract A higher bolt preload is desirable for higher integrity of bolted joints. However, the bearing surface in the joints could be permanently deformed under a higher preload and the bolt preload decreases with an increase of permanent deformation. Various materials are used as clamped parts, so the permanent plastic deformation should be examined for each material clamped parts. In addition, the critical contact stresses should be examined for various clamped materials. In the previous paper, it was found that the relationship between the contact stress when the plastic deformation initiated, and the compressive proof stress of the clamped parts is linear. However, this conclusion was due to the specific bolt shape and dimensions. In the present paper, the effects of shape and dimensions of bolt head on the contact stress and deformation are examined. This applies to normal and undercut type bolts with sloped bearing surfaces. FEM calculations were used and the clamped part materials were steel and aluminum alloy. Also, a new contact area ratio at the bearing surface is defined and investigated because the nominal contact area is different from the actual contact area when the bolt bearing surface is sloped. Finally, discussion is made on an appropriate contact area, and critical contact stress at the bearing surface as well as suitable bolt shape and bolt head dimensions.

Abstract Bottom head flange leaks on coke drums are a too common occurrence for coke drum operators. This paper discusses why these complex joints leak, exploring phenomena such as bolt relaxation, flange rotation, bolt hole cracking, plastic deformation of a flange face due to cyclic thermal transients, as well as investigating the effects of gasket stress variation which can lead to gasket movement and distortion during successive drum cycles. The industry trend is to install automatic bottom un-heading devices (BUDs) to facilitate safe, reliable coke removal and to increase production through shortening un-heading operations. A case study is reviewed which shows how a new drum flange, coupled to a BUD, has been optimized using Finite Element Analysis (FEA). The findings show adequate flange thickness and optimized hub dimensions are required to combat plastic deformation of the flange ring and reduce gasket stress variation. Five designs are modelled, varying the flange thickness, outside diameter and hub geometry. Due to the close proximity of a new side nozzle, the full lower section of the drum has been modelled using quarter and half symmetry FEA with applied temperature distributions from each phase of a typical coke drum cycle; heating, coking and quench. This has allowed for nozzle loadings to be evaluated and the location of the flange weld to be optimized to give the greatest fatigue life. An explanation into why periodic re-tightening is required to keep these joints tight is provided along with recommendations on suitable joint assembly techniques using a combination of bolt load verification and alternative bolting patterns from ASME PCC-1.

Abstract Bolted joint and gasket designers currently calculate spiral wound gasket stress in one of three ways. The differences lie in gasket area determination. The first is the historical Boiler & Pressure Vessel Code calculation method of using an effective sealing width which is half or less of the full dimensional width. Found to be inadequate for creating consistently acceptable sealing performance, those promoting and affecting lower leaks rates and emissions opt to use a gasket’s full width to determine gasket area and stress. The full width of a spiral wound gasket is currently being determined in two different ways among industry participants. Most use the dimensioned outer diameter which includes the bead of winding wire that does not contact the flange surface. Some exclude the width of the bead. Inclusion or exclusion of the bead has a significant impact on gasket seating stress for sealing elements of narrow width as the width of the bead represents a large percentage of the overall dimensions. In this paper, all three methods will be discussed focusing on gasket stresses for NPS calculated from the preferred method of full width comparing the inclusion and exclusion of the bead width.

Abstract Shock wave attenuation in a straight tunnel (or pipe) can be evaluated using existing methodologies. Shock attenuation is enhanced when there are right-angle turns along the length of the tunnel over which the shock is transmitted. A repeated set of such turns is generally defined as a blast trap. Little guidance is available in the open literature regarding the blast attenuation enhancement due to a right-angle turn or a blast trap in a tunnel. This paper presents guidance for shock wave attenuation as a function of the number of right-angle turns and blast wave parameters (i.e., peak pressure and duration). Characteristic parameters are utilized in order to define shock wave properties and tunnel dimensions. The shock attenuation due to up to four consecutive right-angle turns is evaluated. The purpose of this work is to provide a database of the shock attenuation within a tunnel due to multiple right-angle turns for use in designing tunnel structural components and evaluating the response of such components to postulated transmitted shock loads.

Abstract High pressure tubes are used in many industrial applications. Examples are “waterjet cutting” and “hydrogen fuel handling”. In “high pressure application’s”, pressures are handled around 800 bar. The tube dimension 9,53 mm × 3,16 mm is a typical “high pressure” dimension. This dimension can be found in many “waterjet cutting machines”. Tube is a important component in any “high pressure application”. Tubes are typically used in a fatigue environment. The tube needs to survive a certain number of pressure cycles. It is important to increase the number of “cycles” to extend the lifetime of the tube and as a result, the lifetime of the equipment. Higher lifetime rates greatly reduce planned maintenance and reduce risk of unplanned equipment breakdowns. The investigation and development to increase the fatigue properties on existing grades (mainly TP316L is this example) is a foundation to the development of a “new high pressure grade” with specific mechanical Properties. In order to prove the fatigue properties fatigue tests have been carried out under synchronized conditions, using different tubes in comparison grades. In the production of seamless tubes, there are different production methods to create different material properties. Combinations of these methods for each grade are tested and results measured relating to grain size, defect level and surface surface condition are recorded. The test tubes are produced in different production flow with different surface conditions to develop a comparison between surface conditions and the resulting fatigue related results. Two different grades TP316L and HP120 have been used to prove the test results in combination to material grades. Different tube samples run on a fatigue test bench. The test pressure is set at 3.000 bar with a sinus curve at 6 Hz. The results give a development guideline to reach the most advanced tube product for fatigue related applications.

Abstract Thermal embrittlement of some cast austenitic stainless steels (CASS) occurs at reactor operating temperatures can lead to a reduction in the fracture toughness and increase in strength. Some aged CASS materials have the potential to have exceedingly low toughness and also show high variability due to the nature of their microstructure or compositional variation within the casting. Because of their low aged toughness with the variability, flaw evaluations of CASS material need to be done with an understanding of the materials aged condition, especially since most US PWR nuclear plants have been given plant life extensions for 60-year operation, and consideration of further extension to 80 years is underway. In this paper, a flaw evaluation procedure for CASS materials is presented using a new statistical model developed to predict the toughness of fully aged CASS using the material’s chemical composition. The new statistical model was developed based on the experimental toughness using standard 1T CT specimens (generally in the L-C orientation) at 288C to 320C and chemical compositions of the CF8m CASS materials. While the detail development of the model is beyond the scope of this paper, a brief validation of predicted toughness using chemical compositions is presented in this paper. Using the predicted toughness, a flaw evaluation procedure was developed using the Dimensionless-Plastic-Zone-Parameter (DPZP) analysis to determine when limit-load is applicable and also approximate the elastic-plastic correction factor (Z-factor) that needs to be applied to the limit-load solution to predict the failure stress for CASS pipe and fittings with a circumferential surface crack. Variability within a single casting was also determined from available test results which was included in the procedure to determine Z-factor. The procedure was then validated against several CF8m pipe test results which include various pipe diameters, crack sizes, ferrite contents, failure modes (i.e., limit load or EPFM), etc. The as-developed flaw evaluation procedure was also used to determine the Z-factors for four different pipe diameters for a database of 274 pipe/elbows in US PWR plants (whose chemical compositions were known) — essentially solving 1096 sample problems to understand what range of Z-factors might exists in US PWR plants (for CF8m CASS materials) considering all variations in pipe dimensions, ferrite contents, materials’ toughness, etc. Finally, the applicability of the CF8m-based statistical model for use with CF3 and CF8 CASS materials was also investigated by comparing the predictions with available test results.

Abstract The authors proposed a newly three-dimensional isolation system, consisting of a rubber bearing, vertical oil dampers and disc spring units, to reduce the seismic response in the vertical direction as well as horizontal direction. This isolation system is employed with a number of disc spring units to provide the vertical restoring force to the superstructure. The disc spring units are combined by three disc springs in parallels and they are are stacked in six serials. The vertical restoring force has susceptible to the variation forces for the individual disc springs because the disc spring units are combined in the six serials. The The purpose of this paper is to present two kinds of proposal to improve the quality control of our isolation system and the prediction accuracy of seismic response. The first is to create the the optimal combination method for the disc spring units using the meta-heuristic algorithm to minimize the variation of vertical vertical restoring force. The proposed optimal method was verified through the result of static loading tests using the 72 disc springs which have the half dimensions to full scale. The second is to create a newly analytical model for the friction force caused by polymeric materials. The proposed analytical model was verified by comparing the loading test results. Moreover, the seismic isolation performances were clarified by the seismic response analysis that consider the vertical restoring force of the disc spring units which were combined using the optimal method and the friction force of sliding elements which were modeled by the proposed friction model. This analytical result revealed that our isolation system can reduce the seismic response not only for the high frequency components but also the low frequency ones.

Reactor internals are components that are no typical pressure boundary but they are nevertheless very important as they hold fuel elements and all reactor control system elements and thus must ensure their safe and reliable operation during the whole reactor life under all operating and even beyond bases regimes. In principle, reactor internals can be replaced but their weight, quantity of very high activated material and cost such possibility practically excluded. Thus, evaluation of the reactor internals condition and prediction of their behavior during the whole or even extended lifetime is of high importance. Reactor internals are subjected to very high neutron irradiation that could initiated not only stress corrosion (irradiation assisted) cracking but also large embrittlement and changes in dimensions (swelling and creep). VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was initiated and co-ordinated by the Czech and was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for Nuclear Power Plants (NPPs) with WWER (Water-Water-Energetic-Reactor = PWR type) type reactors, as those codes were developed only for design and manufacture and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. Within the last upgrading and principal extending of this VERLIFE Procedure was developed within the 3-years IAEA project (in close co-operation with another project of the 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) that started in 2009 with final approval and editing in 2013”) a part dealing with the evaluation of reactor internals lifetime was elaborated.. This IAEA VERLIFE procedure for internals has been implemented into the existing Normative Technical Documentation (NTD) ASI (Czech Association of Mechanical Engineers), Section IV – Evaluation of Residual Lifetime of Components and Piping in WWER type NPPs. Main damaging mechanisms that should be taken into account in reactor internals and the procedure are described in detail with necessary formulae for materials of internals: - Radiation hardening - Radiation embrittlement - Radiation swelling - Radiation creep - Swelling under stress effect - Swelling inducing embrittlement - Irradiated assisted stress corrosion cracking - Transformation austenite-ferrite and also the method for evaluation of the resistance against non-ductile failure of postulated defect. The paper will describe these main principles and also more detailed information on the procedure for evaluation of reactor internals will be given.

In this paper, the combination rule for circumferential multiple-cracked pipe assessment is investigated using finite element damage analysis. The FE damage analysis based on the stress-modified fracture strain model is validated against limited fracture test data of two circumferential surface cracked pipes. Then systematic parametric study is performed using FE damage analysis for symmetrical surface cracked pipes. Failure bending stresses are calculated using the combination rule and the net-section collapse load approach for single crack provided in ASME BPV Code. It is found that predicted failure bending stress using the combination rule might be non-conservative when the distance between two cracks is short. To overcome the problem, a new combination criterion based on crack dimensions is proposed and compared with numerical data.

The extension of the operation period of nuclear plants requires an accurate characterisation of the vessel materials, in order to monitor their embrittlement due to neutron irradiation. This need poses a challenge, since the availability of specimens inside the vessels to characterise their evolution is rather scarce. Therefore, innovative techniques have to be applied, in order to reduce the number of tests and the volume of the specimens. In this paper, the Master Curve approach has been employed, combined with the use of small punch notched specimens. The Master Curve methodology allows to evaluate the embrittlement of the material using a single parameter: the reference temperature, T 0 . This parameter has been estimated for several steels by means of modified small punch specimens, which are characterised by their reduced dimensions: only 10 × 10 × 0.5 mm. The obtained results have been compared with those obtained by means of conventional testing and a methodology to estimate T 0 by means of small punch tests together with the Master Curve has been proposed.

In ASME Section XI Appendix C for analytical evaluation of flaws in piping, a screening procedure is prescribed to determine the failure mode and analysis method for the flawed pipe. The end-of-evaluation period flaw dimensions, temperature, material properties, and pipe loadings are considered in the screening procedure. Equations necessary to calculate components of the screening criteria ( SC ) include stress intensity factor ( K ) equations. The K -equation for a pipe with a circumferential inside surface flaw in the 2017 Edition Section XI Appendix C-4000 is for a fan-shaped flaw. Real surface flaws are closer to semi-elliptical shape. As part of Section XI Working Group on Pipe Flaw Evaluation (WGPFE) activities, revision to stress intensity factor equations for circumferential surface flaws in Appendix C-4000 has been proposed. The proposed equations include closed-form equations for stress intensity influence coefficients G 0 for membrane stress and G gb for global bending stress for circumferential inside surface flaws. The rationale for the Code changes and technical basis for the proposed stress intensity factor equations are provided in this paper.

Eurofer97 is one of leading candidates of reduced activation ferritic martensitic (RAFM) steels for first wall structural materials of early demonstration fusion power plants. During fusion plant operation, high neutron irradiation damage on first wall materials can cause irradiation embrittlement and reduce the fracture toughness of RAFM steels. Therefore, it is critical to select proper testing techniques to characterize the fracture toughness of RAFM steels with high fidelity. In this manuscript, we present the feasibility study of using pre-cracked miniature multi-notch bend bar specimens (M4CVN) with a dimension of 45mm (length) × 3.3mm (width) × 1.65mm (thickness) to characterize the transition fracture toughness of Eurofer97 steel based on the ASTM E1921 Master Curve method. The testing yielded a provisional Master Curve reference temperature T oQ of −89°C of unirradiated Eurofer97 steel heat J362A in the normalized and tempered condition. The results are within the normal scatter range of Master Curve reference temperature T 0 for Eurofer97 steel, indicating suitability of applying M4CVN specimens for characterizing the transition fracture toughness of Eurofer97 steel.

In design of fast reactor (FR) core components, seismic response must be evaluated in order to ensure the structural integrity. Thus, a core seismic analysis method has been developed to evaluate 3D core vibration behavior considering fluid structure interaction and vertical displacement (upward). Thirty seven 1/1.5 scale core element models which shape hexagonal-arrangement were used to validate the core element vibration analysis code in three dimensions (REVIAN-3D). Based on the test data, the newly incorporated analysis model has been verified to respond to strong excitation.

This work presents recent improvements in the micromechanical failure criterion based on the Weibull stress ( σ w ) concept for prediction of cleavage fracture in ferritic steels. The model is applied in SE(B) specimens extracted from an ASTM A533 pressure vessel steel having different levels of stress triaxiality at the crack tip. Nonlinear 3D finite element models with dimensions matching the tested specimens were built to provide the necessary crack tip stresses at the fracture process zone for calculation of the σ w -J evolution from wich the variation of characteristic toughness values ( J 0 ) between different cracked geometries can be estimated. Application of this methodology for the material used at this study is able to predict J 0 for SE(B) specimens with very shallow crack size ratio a / W = 0.05, short crack a / W = 0.2 and deep crack a / W = 0.4. The reported fracture toughness values for specimens having very shallow crack size ratio is an additional contribution of this study.

Regions of higher-than-normal carbon content due to carbon macro-segregation have been found in large, pressure retaining forged ferritic steel components in some nuclear reactors. Higher carbon content in ferritic steel can decrease the resistance to fracture from the presence of flaws in the material. Acceptable margins against failure of pressurized components in nuclear safety systems must be maintained throughout their service life to ensure core integrity for all operational and postulated transient loading events. Should carbon macro-segregation substantially reduce the material resistance to fracture in safety components, then the margins against through-wall flaw propagation may fall below those specified by regulatory requirements to ensure adequate component and reactor core integrity. Probabilistic fracture mechanics (PFM) analyses were performed to assess the risk and structural significance of postulated carbon macro-segregation in large, forged pressure retaining components in pressurized water reactors (PWRs). The risk assessment was performed to evaluate several forged components and two classes of loading events. The forged components include the ring and head forgings in the reactor pressure vessel (RPV), steam generator (S/G) and pressurizer. The loading events used in the risk evaluation include pressurized thermal shock (PTS) transient events and a normal RPV cooldown event. The analyses included a range of component dimensions, surface and embedded flaw distributions, various levels of carbon macro-segregation up to and beyond the maximum measured values for the components, and the effects of neutron irradiation, including the effects of potential copper and phosphorus co-segregation. The PFM analyses were performed using the software, Fracture Analysis of Vessels, Oak Ridge (FAVOR). The results from the risk assessment indicate that: acceptable margins against failure are maintained through an 80-year operating interval even if carbon macro-segregation were to be present in RPV, S/G and pressurizer ring and head forgings in PWRs; and the risk associated with the presence of carbon macro-segregation in PWR ring and head forgings is significantly lower than regulatory risk related acceptance criteria.

During tightening, the amount of torque given by the difference between the tightening torque, which is directly applied by the torque wrench, and the underhead torque, flows through the screw shank towards the threaded portion. This torque combines with the axial preload to bring about the overall stress state of the screw at tightening. Upon release of the torque wrench, a certain amount of the shank torque is released due to the elastic springback of the screw-plates system. In the literature [1], this phenomenon is just briefly treated by a few authors: they generally agree that approximately a half of the initial shank torque is released just a few seconds after torque wrench removal. This indication is given regardless of the frictional [2] and stiffness [3] parameters, which govern the joint. The present contribution aims at assessing, if there is any effect of the following parameters on the amount of shank torque being released after wrench removal: (i) the ratio between the torsional stiffness of the screw and of the plates, (ii) the friction coefficients in the underhead and in the thread. The experimentation has been run on a M20 8.8 class socket head screw, which has been instrumented by a double array of strain gauges, to simultaneously measure both the axial preload and the torque acting on its shank. Two different types of joined members have been examined: a cylindrical sleeve whose diameter is twice the screw diameter (compliant joint) and a rectangular plate whose transverse dimensions are more than ten times larger than the screw diameter (stiff joint). The underhead and thread friction coefficients have been controlled by properly selecting lubrication conditions. The main outcome of the work is that the torsional stiffness of the joined members does have an impact on the residual shank torque. A simple mathematical model has also been implemented, in order to predict the residual shank torque during the design phase.

Our objective is to evaluate precisely a life-cycle of bolted joints under an eccentric load against a bolt axis. Many approaches to achieve the objective based on a lot of theories and practices have been proposed so far [1–12]. As we can refer from their approaches, the opening of the structural interface between clamped plates of bolted joints occurs by the eccentric load, which is over a bolt preload, and then the opening gradually propagates as the eccentric load increases. In the case, nonlinearity appears remarkably on the tensional and bending stress of bolts in the axial direction. In addition to the above, the axial bolt stress larger than expected occurs due to the principle of leverage depending on the load position and the bolted joints layout in the early phase of the pull-out action. Accordingly, the stress evaluation of bolted joints under the eccentric load is very important in order to ensure the safety of industrial machines. If dimensionless quantities of the bolt stress are found out considering the influence of the structural opening and the load eccentricity, we can have a few advantages as follows. First, bolt stress evaluations can be conducted by easily converting the dimensionless quantities of the bolt stress to the physical dimension quantities in a lot of cases where the bolt preload and the load eccentricity are different. Second, the number of times of verification tests can be reduced. We are developing a lot of industrial machines which have bolted joints used under eccentric load. In such development [13], bolt stress analyses are usually conducted under the combinations of the following conditions: (i) tapped thread joints, (ii) thin clamped plates than the bolt diameter, (iii) large eccentric loads, (iv) permitting the opening of the structural interface. Therefore, we propose a concept of a normalized bolt stress considering the effect of the structural opening and the load eccentricity. We validated this concept through theoretical studies, finite element analyses, and experiments under the direct load and the centrifugal load. As a result, the dimensionless quantities of the bolt stress caused by the bolt preload and a lever ratio of bolted joints under combined conditions was determined in this study. We can easily evaluate the bolt stress by simple conversions in a lot of cases in which the bolt preload and the load eccentricity differ.

The material attributes that are fundamental for developing a candidate textured, ceramic-filled PTFE gasket, such as texture style/dimensions, filler material, thickness and so on, create a set of potential combinations that are not practical to experimentally characterize at the component-level one-by-one. Optimizing gasket performance, however, is essential to the operation of bolted connections associated with pressurized vessels that transfer media from one location to another. Gaskets are essential for these systems since they confer high levels of leak mitigation across a range of operating environments. A balance of both compressibility and sealability must be displayed in an optimal candidate gasket to be subjected to aggressive operating conditions. A novel textured PTFE material (termed textured) characterized using a miniaturized test platform. This new-to-market viscoelastic material features a dual-face, raised honeycomb pattern. Experiments on both flat (termed Flat) and textured are used to identify viscoelastic constitutive model constants associated with Burger theories. Considering that the test platform contains an elastic bolt that is tightened to a prescribe torque level, the gasket is subjected to creep relaxation. Test results on the small samples contribute to constitutive modeling. The influence of parameters such as filler material selection, torque level, dwell period, etc. are presented.

Jacketed gaskets have widespread use in the industry. The current revision of ASME B16.20-2017 provides several details of jacketed gasket construction and dimensions for use on ASME B16.5 or ASME B16.47 flanges. However, it does not currently provide a testing protocol to evaluate the sealing performance of this gasket style. This paper reports sealability and compression results with distinct sizes of jacketed gaskets tested based on the protocol proposed in ASME B16.20 - Spiral Wound Gaskets Performance Testing.

Containment vessel (CV) of nuclear power plants is an important structure to prevent a significant and sudden radioactive release, however, the safety margin of the containment vessel against the internal or external pressure are not numerically clarified. The head plate is one of the components which constitute the CV boundary. In order to develop the evaluation method of the pressure toughness of the head plate at beyond design basis events, the pressure failure tests and finite element analysis of the head plates subjected to convex side pressure were performed. In the tests, non-axisymmetric deformations with local deformation concentration were observed in post buckling behavior in the case of the thin thickness head plate. In this study, to evaluate these non-axisymmetric deformations in the test, finite element analyses using detailed 3-D solid model constructed by precise dimensions of the head plates measured by 3-D scanner were performed. Moreover, FEA using simplified model with uniform or non-uniform thickness model were performed. Through a series of FEA, it was clarified the effect of each thickness pattern on post buckling non-axisymmetric deformation.