bluuebunny/crossref_metadata_2025 huggingface dataset:
Simplified Analysis Methods for Primary Load Designs at Elevated Temperatures
Authors: Carter Peter , Sham T.-L. Sam , Jetter R. I.
Published: May 2012

The use of “simplified” (reference stress) analysis methods is discussed and illustrated for primary load high temperature design. Elastic methods are the basis of the ASME Section III, Subsection NH primary load design procedure. There are practical drawbacks with this current NH approach, particularly for complex geometries and temperature gradients. The paper describes an approach which addresses these difficulties through the use of temperature-dependent elastic, perfectly-plastic analysis. Traditionally difficulties associated with discontinuity stresses, inelastic strain concentrations and multiaxiality are addressed. A procedure is identified to provide insight into how this approach could be implemented. Though preliminary in nature, it is intended to provide a basis for further development and eventual Code adaptation.
Life Prediction of T91 Welded Joint and its Engineering Application Under Consideration of Welding Residual Stresses
Authors: Zhou Changyu , Zhang Guodong
Published: May 2012

Welding residual stresses (WRS) of welded joint have been a problem for long time. At present experiment and numerical simulation technology can be used to determine the distribution of WRS. ABAQUS codes were used to calculate WRS for welded joint of T91 tube and obtain the stress distribution under internal pressure (IP) and bending moment (BM) with WRS. ABAQUS user subroutine of the modified Kachanov-Rabotnov (K-R) constitutive equations was developed to calculate creep damage of welded joint on consideration of WRS. The distribution of stress and creep damage was obtained for the tube at different service time. Although the WRS relax at initial stage rapidly at high temperature condition, the accumulative damage caused by initial stress has much effect on the welded joint. Finally life prediction method of T91 welded joint with WRS on the basis of creep damage was proposed, which was compared with the results by DL/T 940-2005, API 579, and L-M method. The results in present paper are in good agreement with experiment result.
Ultrasound and Nonlinear Resonant Testing of Nuclear Reactor Internals Bolts
Authors: Glass S. W. , Thigpen B. , Renshaw J.
Published: May 2012

As many nuclear plants approach the end of their initial 40 year license period, inspection or replacement of their reactor internals bolts must be considered. This is consistent with the Materials Reliability Program (MRP 227/228) guideline for plant life extension [1,2]. Assurance of the internals structural integrity is essential for continued safe operation of these plants. If there is no suspicion or indication of bolt failure, simple inspection is normally more cost-effective than replacement. Inspection vendors have inspected thousands of internals bolts with conventional and Phased Array UT but different head configurations and bolt capture mechanisms mandate specific qualifications for each bolt type. In some cases, complex bolt and head geometries coupled with counter-bore and locking bar interferences render classical UT inspections difficult or impossible. A range of solutions to inspect reactor internals including these difficult-to-inspect-by-conventional-UT baffle bolts has been developed by several vendors [3]. This presentation references developments to make bolt inspection a relatively quick and easy task through adaptations to the SUSI submarine inspection platform, the extensive UT qualification work suitable for conventional UT plus more recent advanced nonlinear resonant techniques to distinguish between flawed or loose, vs. acceptable bolts where conventional UT cannot be applied. Initial evaluations show that these advanced techniques may have the ability to reliably detect smaller flaws than previously possible with conventional techniques as well as provide information on bolt tightness.
Guidelines for Repair Welding of Pressure Equipment in Refinery and Chemical Plants: Part 3—Repair Welding for Specific Materials - Carbon Steel, High Tensile Steel and Cr-Mo Steel
Authors: Kayano Rinzo , Abe Masamitsu , Hirai Yukio
Published: May 2012

It has been recognized that repair welding plays an important role in the long term, safe operation of pressure equipment. In 2009, The Japan Welding Society (JWES) published guidelines for repair welding of pressure equipment [1], to meet the great need for the safe operation and proper maintenance of aging plants. Pressure equipments made from carbon steel, high tensile steel and Cr-Mo steels are utilized for high pressure services. The subject equipments are pressure vessels, heat exchangers, piping, and storage tanks for petroleum, petrochemical and liquefied natural gas industry. This paper summarizes category and property of these steels and repair welding method including special attention. Especially, weld cold cracking for these steels could be prevented by controlling the repair welding and post-weld heat treatment process to reduce the hydrogen content, hardness and weld residual stress.
Fracture Evaluation of Ni-Based Alloy Weld Joint of Cylindrical Model Subjected to 4-Point Bending or Inner Pressure
Authors: Ogawa Kazuo , Hojo Kiminobu , Muroya Itaru , Iwamoto Youichi , Ogawa Naoki
Published: May 2012

For the purpose of establishing fracture evaluation method of nickel based alloy weld of nuclear power plants, fracture tests using pipe models (8B and 14B for bending, 12B for inner pressure) with an alloy 132 weld joint have been performed at room temperature and high temperature (325°C). The predicted loads calculated by limit load evaluation method using the measured and code regulated flow stresses were compared with the maximum test loads. And the predicted bending loads of the pipes at 325°C (8B and 14B) and at room temperature (8B) with the initial surface crack whose depth is 75% of the pipe thickness were in good agreement with the maximum test loads. Also the predicted inner pressure of the pipe at room temperature (12B) agreed with the measured maximum pressure. Only for one case of the 14B pipe subjected to the bending load at room temperature, the predicted load by limit load evaluation method has 20% unconservative difference from the measured data, on the other hand, the predicted load by J-T analysis made this difference smaller and conservative.
A Consideration of Margin on Fatigue Design Curves for Carbon and Low-Alloy Steels
Authors: Higuchi Makoto , Takanashi Masahiro , Tamura Ichiro , Takada Toshiaki
Published: May 2012

In 2007, the US NRC issued Regulatory Guide 1.207[1] and NUREG/CR-6909[2] for evaluating fatigue incorporating the life reduction due to the effects of light-water reactor environment for new reactors. NUREG/CR-6909 provides new design fatigue curves (DFC) for carbon, low-alloy and stainless steels which are different from those in the ASME Boiler and Pressure Vessel Code Section III[3] (2007 Edition). The design fatigue curves for carbon and low-alloy steels in NUREG/CR-6909 are higher than that for ferritic steels of which specified minimum tensile strength is 552 MPa (80 ksi) or less in the ASME Code Section III. The design fatigue curve for stainless steel in the ASME Code Section III was changed to the same curve as NUREG/CR-6909 in the 2009 Addenda. However, those for carbon and low-alloy steels are still different from the NUREG curves.
Development of a New Specimen to Study Crack Propagation Threshold and Non-Propagation Conditions
Authors: Bouin Pauline , Marie Ste´phane , Perez Gre´gory
Published: May 2012

The extension of nominal service-life of nuclear power plants leads to calculate more involved safety analyses of the nuclear structures for the different scenario of failures. In order to prevent catastrophic fracture, it is important to minimise the initiation of cracks or if considering that a crack can initiate, to determine if the extension of these cracks can stop or lead to the failure of the structure integrity. As a result, this work requires better understanding and characterisation of the crack propagation or non-propagation behaviour. Existing practises to determinate fatigue crack growth rate and fatigue thresholds are mainly based on compact tension (CT) specimen testing. For high loading ratio, the maximum of stress intensity factor (SIF) usually remains constant whereas the minimum of the SIF is increased to reduce the SIF range down to the threshold. On the contrary, for low loading ratio, complex loading control is required to make maximum and minimum SIF decrease. Furthermore, the results of these tests are mainly dependent on the accuracy of the piloting of the test bench, the specimen instrumentation and the force loading cell capacity. This paper presents details of adaptation of a specimen initially developed to study crack arrest problems under cleavage fracture. A bulk compact tension (BCT) specimen has been designed based on a standard CT specimen. Specifically, a reinforced heel at the back of the specimen enables a reduction of the loading at the crack front. In the case of fatigue, this reduces the SIF range while the loading condition remains constant during the test. A SIF calibration for these BCT specimens has also been established to estimate the compliance using finite element analysis with the French code, Cast3M.
Reactor Pressure Vessel: EDF R&D Program to Support Lifetime Management
Authors: Fontes Jean-Philippe , Raynaud Christelle , Martin Alain , Parrot Aurore , Dahl Anna , Moinereau Dominique , Todeschini Patrick , Churier-Bossenec Henriette , Pages Claude
Published: May 2012

The Reactor Pressure Vessel (RPV) is with the concrete containment one of the two components of a NPP whose replacement is not considered as reasonably feasible. The RPV lifetime has thus an important impact on the lifetime of the whole NPP. One of the key issues concerning RPV lifetime is the radiation effect on the RPV steel in the core zones. The vessel steel becomes indeed more brittle in the RPV core region where radiation is high. Margins have been included at design and manufacturing stages taking into account the material’s embrittlement. Moreover, operating measures have been taken to manage ageing of RPV in order to extend lifetime. The challenge is to preserve high margins and to provide the safety studies showing these margins. A large R&D program has been developed to support lifetime extension. The objective of the program is to develop tools and provide input data for the demonstration of the safe operation of the reactor pressure vessel significantly over a 40-year lifetime. The aim of the paper is to present an overview of the R&D program to support lifetime management on the fields of materials, mechanics and thermalhydraulics. Experiments are indeed performed on irradiated material in order to improve the knowledge on embrittlement for high fluences and to be able to determine embrittlement correlations for materials representative of French RPV. Actions are also planned to improve evaluation of the RPV mechanical behaviour and to describe physical phenomena such as crack arrest or warm pre-stressing effect. Last, studies are realized to improve the thermal loadings evaluations under hypothetical accidental scenarios. These studies are supported by thermalhydraulic numerical simulations whose validation is obtained by comparison to experimental results from experimental hydraulic loops representative of French RPV.
Dynamic Analysis of Intercoolers and Tubular Reactors of LDPE Plants
Authors: Fani Riccardo , Passeri Marco , Paperini Fabio
Published: May 2012

Large-scale LDPE plants involving ultra-high pressure piping and potentially high-pressure pulsations need a detailed dynamic analysis. Special attention must be paid to the vibration of intercoolers and tubular reactors that represent the most critical part of the system, being mounted on tall steel structures. The long delivery of such components requires a proper selection at an early stage of the project. This implies the assessment of plant pulsation-induced forces, the subsequent definition of layout adjustments, of piping supports locations, type and supporting structures. The study must consider the actual jacketed tube configuration to avoid internal piping vibrations, which are difficult to be monitored and able to produce unexpected failures. The final review of the plant design should bring just minor adjustments.
Long Term Stability of Compressive Residual Stress Introduced in Alloy 600 by Water Jet Peening Under Elevated Temperature Environment
Authors: Hashimoto Tadafumi , Osawa Yusuke , Kameyama Masashi , Hirano Shinro , Chigusa Naoki , Saida Kazuyoshi , Mochizuki Masahito , Nishimoto Kazutoshi
Published: May 2012

Primary water stress corrosion cracking (PWSCC) in the weld metal of alloy600 is an issue of concern in a pressurized water reactor (PWR). As a countermeasure against PWSCC, water jet peening (WJP), which can change tensile residual stress into compressive residual stress, has been applied to welded joints. Microstructure in the target area of WJP has an influence of not only WJP but welding and machining. Especially machining introduces severe plastic deformations to the materials. So microstructure in the target area might lack thermal stability due to severe plastic deformation. Additionally the region that compressive residual stress by WJP is nearly up to 1mm from the surface of the target material. As PWRs are operated at about 596K for long term, the compressive residual stress by WJP may be relieved due to creep. In order to keep operating PWRs safety, the stability of the compressive residual stress by WJP at elevated temperature has been clarified. In this work, the results were obtained written below. As a result of thermal aging test, a relaxation of compressive residual stress at specimen surface layer occurred due to recovery of the plastic deformations by machining. This stress relaxation behavior followed Johnson-Mehl equation. However residual stress relaxation due to creep was very few. Therefore it has suggested that the compressive residual stress introduced in Alloy600 by WJP is confirmed to remain stable during long term operation under elevated temperature.
Hydrogen and High Temperature Resistant V-Modified 9Cr-1Mo Heavy Plates Devoted to New Generation High Performance Petrochemical Reactors
Authors: Pillot Sylvain , Zhao Zhao , Corre Ste´phanie , Chauvy Ce´dric , Coudreuse Lionel , Toussaint Patrick
Published: May 2012

The efficiency of petrochemical reactors is intimately related to process parameters, i.e. service temperatures and pressures. Low alloyed ferritic materials, such as 2 1/4 Cr1Mo(V) and 3Cr1Mo(V) steel grades, are widely used for many years to build heavy wall reactors. This is mainly due to their good mechanical properties at high temperatures under high hydrogen partial pressures and good resistance to High Temperature Hydrogen Attack (HTHA). Depending on the grades, the ASME Code gives limitations in terms of maximum temperature that can limit the use of these low alloy grades. Moreover, above a given temperature, maximum allowable stresses are driven by the creep behaviour, leading to a strong lowering of the assumed resistance and hence to extra-thickness and weight. Many developments were done concurrently to increase the efficiency of petrochemical processes. In particular, this can lead to increase service temperatures and therefore actual pressure vessel wall temperatures. Indeed, more and more temperatures around 500°C are likely to be used, leading to reduced choice in terms of permitted steel grades. The low alloy vanadium-enhanced grades are not allowed (except using specific code case) whereas the usual grades have reduced creep allowable stresses. With a view to allowing strong improvements in admissible process parameters, a vanadium-modified 9Cr1Mo creep strength enhanced material with advanced hydrogen resistance and improved toughness was developed. Very thick plates (up to 200mm thick) were produced and tested. This contribution reports both mechanical and metallurgical assessments performed on these heavy plates. Evaluations of hydrogen resistance (HTHA) as well as creep resistance under high hydrogen pressure are also reported. The V-modified 9Cr1Mo grade exhibits an excellent behaviour in hydrogen rich environment, showing therefore some advantages in terms of service conditions. The manufacturing of heavy plates has made significant progress in the recent years, allowing thick products to be manufactured with good homogeneity and mechanical behaviour. Taking into account the maximum use temperature as well as the allowable stresses as described in the ASME BPV Code section VIII division 2, the V-modified 9Cr1Mo grade will clearly be of great interest to companies wishing to enhance the efficiency of their refining/petrochemical processes. The question of welding must also be addressed more specifically to finish validating the 9Cr1MoV option.
Survey of Ke-Factor on Strain for Strain Rate in Environmental Fatigue Evaluation
Authors: Asada Seiji
Published: May 2012

Regulatory Guide 1.207, “Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components Due to the Effects of the Light-Water Reactor Environment for New Reactors,” [1] requires environmental fatigue evaluation for LWR components and the detail evaluation method is described in NUREG/CR-6909 [2]. Strain rate is a parameter in the calculation of environmental correlation factor (Fen). When the calculated primary plus secondary stress intensity is greater than 3Sm, a Ke-factor has to be applied to Salt. However, RG 1.207 and NUREG/CR-6909 is not clear for Ke-factor on strain. The JSME Environmental Fatigue Evaluation Method [3] specifies that the Ke-factor shall not be applied to strain. This is conservative, but the consistency between stress and strain cannot be maintained for Ke-factor. A parametric study was performed for the above cases in order to confirm whether calculating strain rate with Ke-factor on strain is appropriate or not. This paper reports the results of this study.
Inadvertent Blowdown Event in a Nuclear Boiling Water Reactor
Authors: Engel Robert
Published: May 2012

On March 6th 2007, the Leibstadt Nuclear Power Plant in Switzerland experienced an automatic blowdown of eight safety/relief valves installed on the main steam lines caused by a faulty electrical manipulation while performing planned maintenance during full power operation. Due to the temperature measurements inside the reactor recirculation system and the reactor pressure vessel this event, at a first glance, appeared to be Event No. 23 (Automatic Blowdown event) as an Emergency (Service Level C) Condition in accordance with the relevant reactor pressure vessel Thermal Cycle Diagram. According to the ASME Code Section III, Service Level C limits permit large deformations in areas of structural discontinuity which may necessitate the removal of a component from service for inspection or repair. This paper presents a summary of thermal-hydraulic, stress, fatigue, and fracture mechanical evaluations as well as plant inspections performed to demonstrate the impact of the event on the reactor pressure vessel and associated components and to fulfill the requirements of the Swiss Federal Nuclear Safety Inspectorate. It is shown that the primary circuit of the plant was not inadmissibly stressed by the event and that it was acceptable from a safety-related point of view to return the plant to service. Corresponding to the 7-level International Nuclear and Radiological Event Scale this event was rated afterwards as level 1 (anomaly) by the Swiss Federal Nuclear Safety Inspectorate.
Plastic Behaviour of Small Bore Piping Bend/Elbow Under Combined Internal Pressure and Bending Moment
Authors: Duan Xinjian , Kozluk Michael
Published: May 2012

This paper presents some preliminary observations from six representative tests on the plastic behavior of small bore carbon steel piping bends/elbows with general and local wall loss subject to combined internal pressure and bending moment loading at typical CANDU® reactor operating temperature. These six tests include three materials (normalized elbow, stress relieved bend and cold worked bend), two piping sizes (NPS 2 and NPS 2 1/2), two loading modes (monotonic and cyclic) and three modes of failure (collapse, burst and leak). The test results show that a collapsed piping bend/elbow can still withstand five times design pressure without the loss of pressure boundary. The stress relieved bend appears to offer a compromise in the resistance to both environmental assisted cracking and plastic collapse relative to the normalized elbow and cold worked bend.
Inelastic Fracture Mechanics Model for Assessment of Crack-Like Flaws
Authors: Jaske Carl E. , Polasik Steven J. , Maier Clifford J.
Published: May 2012

Fitness-for-service assessment of pressure vessels and piping often involves the evaluation of existing or potential crack-like flaws to guard against fracture or leaks that could be caused by the presence of such flaws. This paper presents an inelastic fracture mechanics model that has been developed to evaluate longitudinal surface cracks in pipelines, piping and pressure vessels subjected to internal pressure loading. The model uses the J-integral parameter to predict toughness-dependent failure and an effective flaw concept to predict flow-strength dependent failure. The concepts of the model are reviewed. Then, the model is used to evaluate the results of in-service failures and full-scale burst testing of steel pipe and pressure vessel samples. Application of the model to remaining life assessment based on inspection data and hydrostatic testing results is illustrated. Stress-corrosion cracking (SCC) and fatigue are considered as possible crack-growth mechanisms. Examples of typical remaining crack-growth life calculations are presented using both deterministic and probabilistic methods. The benefits of each method are discussed. Finally, planned future additions to the model are presented.
Establishing Allowable Nozzle Loads
Authors: Koves William , Upitis Elmar , Cullotta Richard , Latif Omar
Published: May 2012

Every engineering project involving the design of pressure equipment, including pressure vessels, heat exchangers and the interconnecting piping requires that the interface loads between the equipment and piping be established for the pressure vessel nozzle design and the limitations on piping end reactions. The vessel or exchanger designer needs to know the external applied loads on nozzles and the piping designer needs to know the limiting end reactions on any connected equipment. However, the final loads are not known until the piping design is completed. This requires a very good estimate of the piping end loads prior to completing the vessel or piping design. The challenge is to develop a method of determining the optimum set of design loads prior to design. If the design loads are too low, the piping design may become too costly or impractical. If the design loads are too high the vessel nozzle designs will require unnecessary reinforcement and increased cost. The problem of the stresses at a nozzle to vessel intersection due to internal pressure and external forces and moments is one of the most complex problems in pressure vessel design. The problem has been studied extensively; however each study has its own limitations. Numerous analytical and numerical simulations have been performed providing guidance with associated limitations. The objective is to establish allowable nozzle load tables for the piping designer and the vessel designer. The loads and load combinations must be based on a technically accepted methodology and applicable to all nozzle sizes, pressure classes, schedules and vessel diameters and thicknesses and reinforcement designs within the scope of the tables. The internal design pressure must also be included along with the 3 forces and 3 moments that may be acting on the nozzle and the nozzle load tables must be adaptable to all materials of construction. The Tables must also be applicable for vessel heads. This paper presents the issues, including the limitations of some of the existing industry approaches, presents an approach to the problem, utilizing systematic Finite Element Analysis (FEA) methods and presents the results in the form of tables of allowable nozzle loads.
A Procedure to Include the Fatigue Effects of Thermal Gradients in a B31.1 Butt Weld of Dissimilar Metals and Welding End Transition Joint (TTJ) of Similar Metals for Design and Plant Life Extension
Authors: Creates David H.
Published: May 2012

Fatigue evaluation in B31.1-2007 is currently done based on B31.1 Equations 1 & 2 and generally considers only the stress due to displacement load ranges as per Equation 13. Yet, fatigue damage is also occurring due to pressure cycling and thermal gradients. To exacerbate this, power plant design pressures and temperatures are rising, new materials are being introduced, pipes and attached components are becoming increasingly thick, and owners are requiring power plants to heat-up and cool-down at faster rates. Also, power plant owners are more and more interested in extending the life of power plants beyond their original design life. This paper addresses the pressing need in today’s power plant environment for additional fatigue evaluation by providing a procedure for assessing an as-welded Butt weld of dissimilar metals and a Weld End Transition (or TTJ) (B31.1-2007 Figure 127.4.2) of similar metals to include the effects of thermal gradients calculated as per ASME Section III-2007 Subarticle NB-3600. The disadvantage of this approach is that the conservatism in the calculation of these thermal gradient stress intensities may produce unacceptable results. In that case, the assessment is a warning that something else needs to be done by way of either monitoring or modifying the thermal operation or more rigorous evaluation. The advantage of this methodology is that it will ensure a fatigue failure does not occur any sooner due to the effects of thermal gradients than would otherwise occur due to other factors. It maintains the traditional B31.1 approach to fatigue with the same limit of SA except that there is now an additional term, STG, to account for the fatigue contribution due to thermal gradients. Considering the effects of these thermal gradients in this way will further help to preserve the integrity of the piping pressure boundary and consequently, the safety of personnel in today’s power plants and into the future.
Two-Phase Flow Excitation Forces on a Vertical U-Bend Tube
Authors: Giraudeau M. , Pettigrew M. J. , Mureithi N. W.
Published: May 2012

Significant two-phase flow-induced forces were previously measured on a 20 mm diameter U-tubes over a range of homogeneous void fractions. To investigate the effect of tube diameter, new experiments were conducted using vertical 52 mm diameter U-tubes. In the experiments, two parameters were varied, the homogeneous void fraction and the homogeneous velocity. All the results were compared with those for the smaller diameter tubes. The geometrical characteristics of the smaller diameter test sections are similar. L/D = 56 and the ratio of the bend curvature radius to inner diameter: R/D = 4. For all conditions tested, we compared the vertical upward flow-patterns for four different tube diameters. The observed flow regimes agreed relatively well with those predicted from two-phase flow pattern maps, except for the slug/churn transition. A specific two-phase flow pattern map is proposed for the 52 mm diameter tube, using a more accurate transition model from slug to churn flow. The forces comparison confirms the dependency with tube diameter. The RMS forces were found to increase with tube diameter for all void fractions studied. Conversely, the dominant frequency decreases with tube diameter, linearly for void fractions of 50 and 75%. Dimensionless forces and spectra for the smaller diameter tubes show good data collapse using the proposed model. Two-phase flows in the 52 mm diameter tube generate lower dimensionless spectra. Finally, a dimensionless model is proposed for practical application.
Overview of Regulations and Codes/Standards for Inspection Intervals of Above Ground Storage Tanks
Authors: Tahara Takayasu , Sando Shinsuke
Published: May 2012

This paper summarizes survey results of regulations and Codes/Standards, such as API Standard 653, EEMUA 159 Fire Service Law in Japan. The survey relates to inspection interval of above ground storage tanks, ASTs, typically as shown in Figure 1, in several countries in recent years.
Operation and Maintenance of Hyper Compressor in LDPE Plant
Authors: Mok Abang Azian Abang , Ghani Muhd Hafiz B. Hj , Kadir Alias B.
Published: May 2012

Petlin (Malaysia) Sdn Bhd was incorporated in March 1999 and has a Low Density Polyethylene plant that was commissioned in February 2002. The plant has a capacity of 255,000 tons per annum. The aim of the plant is to generate a reasonable return by manufacturing top quality products. To realize this aim, it’s important for the operators of the plant to operate the machines in accordance to the parameters set by the original equipment manufacturer. This will also ensure the availability and reliability of the plant. It is also important to regularly monitor the performance of the plant in relation to the operation and maintenance activities in order to prepare a better plan for continuous improvement. The Hyper-Compressors are the most fragile component in the system as they’re running under fatigue conditions and high operating pressures. It is essential to have the right processes and procedures in place to ensure that safety is not compromise in meeting the production target.
The Effects of Warm Pre-Stressing of a Pre-Cracked Pressurised Thermal Shock Disk Specimen Under a LUCF Loading Cycle
Authors: Teng H. , Beardsmore D. W. , Sharples J. K. , Budden P. J.
Published: May 2012

A finite element analysis has been performed to investigate the effects of warm prestressing of a pre-cracked PTS-D (Pressurized Thermal Shock Disk) specimen, for comparison with the experimental work conducted by the Belgium SCK-CEN organisation under the European NESC VII project. The specimen was loaded to a maximum loading at −50 °C, unloaded at the same temperature, cooled down to −150 °C, and then re-loaded to fracture at −150 °C. This is a loading cycle known as a LUCF cycle. The temperature-dependant tensile stress-strain data was used in the model and the finite element software ABAQUS was used in the analysis. The finite element results were used to derive the apparent fracture toughness by three different methods: (1) Chell’s displacement superposition method; (2) the local stress matching method; and (3) Wallin’s empirical formula. The apparent fracture toughness values were derived at the deepest point of the semi-elliptical crack for a 5% un-prestressed fracture toughness of 43.96 MPam1/2 at −150 °C. The detailed results were presented in the paper.
Determination of Shakedown Limit Loads for a Cylindrical Vessel–Nozzle Intersection via a Simplified Technique
Authors: Abdalla Hany F. , Younan Maher Y. A. , Megahed Mohammad M.
Published: May 2012

In the current research, the shakedown limit loads for a cylindrical vessel–nozzle intersection is determined via a simplified technique. The cylindrical vessel–nozzle intersection is subjected to a spectrum of steady internal pressure magnitudes and cyclic in–plane bending moments on the nozzle. The determined shakedown limit loads are utilized to generate the Bree diagram of the cylindrical vessel–nozzle intersection. In addition, the maximum moment carrying capacity (limit moments) and the elastic limit loads are determined and imposed on the Bree diagram of the structure. The simplified technique outcomes showed excellent correlation with the results of full elastic–plastic cyclic loading finite element simulations.
NESC-VII: Fracture Mechanics Analyses of WPS Experiments on Large-Scale Cruciform Specimen
Authors: Yin Shengjun , Williams Paul T. , Bass B. Richard
Published: May 2012

This paper describes numerical analyses performed to simulate warm pre-stress (WPS) experiments conducted with large-scale cruciform specimens within the Network for Evaluation of Structural Components (NESC-VII) project. NESC-VII is a European cooperative action in support of WPS application in reactor pressure vessel (RPV) integrity assessment. The project aims in evaluation of the influence of WPS when assessing the structural integrity of RPVs. Advanced fracture mechanics models will be developed and performed to validate experiments concerning the effect of different WPS scenarios on RPV components. The Oak Ridge National Laboratory (ORNL), USA contributes to the Work Package-2 (Analyses of WPS experiments) within the NESC-VII network. A series of WPS type experiments on large-scale cruciform specimens have been conducted at CEA Saclay, France, within the framework of NESC VII project. This paper first describes NESC-VII feasibility test analyses conducted at ORNL. Very good agreement was achieved between AREVA NP SAS and ORNL. Further analyses were conducted to evaluate the NESC-VII WPS tests conducted under Load-Cool-Transient-Fracture (LCTF) and Load-Cool-Fracture (LCF) conditions. This objective of this work is to provide a definitive quantification of WPS effects when assessing the structural integrity of reactor pressure vessels. This information will be utilized to further validate, refine, and improve the WPS models that are being used in probabilistic fracture mechanics computer codes now in use by the NRC staff in their effort to develop risk-informed updates to Title 10 of the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G.
Proposal of a Guideline to Classify Planar/Non-Planar Flaw in Wall Thinned Straight Pipes Under Bending Load
Authors: Meshii Toshiyuki
Published: May 2012

In this paper, an idea to rationally classify planar/non-planar flaw in wall thinned straight pipes under bending load was proposed, based on the fracture mode (cracking/non-cracking) observed in experiments. This cracking phenomenon was correlated with the aspect ratio of the flaw planar configuration. Finally, a guideline to classify this planar/ non-planar flaw based on flaw aspect ratio was proposed.
Framework to Correlate Test Specimen Thickness Effect on Fracture Toughness With T33-Stress
Authors: Meshii Toshiyuki , Tanaka Tomohiro
Published: May 2012

This paper considered the test specimen thickness effect on the fracture toughness of a material Jc, in the transition temperature region, for CT and 3PB specimen. Framework to correlate test specimen thickness effect on fracture toughness with T33-stress, which is the out-of-plane elastic crack tip constraint parameter, was proposed. The results seemed to indicate a possibility of improving the existing methods to correlate the fracture toughness obtained by test specimen with the toughness of actual cracks found in the structure, in use of T33–stress.
Development of Surface Crack Analysis Program and its Application to Some Practical Problems
Authors: Shiratori Masaki , Nagai Masaki , Miura Naoki
Published: May 2012

The authors have developed a software system called “SCANP™” by which users can analyze residual lives of surface-cracked structural components such as pressure vessels and their piping systems due to fatigue or SCC. The basic concept is based upon an influence function method by which the stress intensity factor “K” of a surface crack can be calculated for arbitrarily distributed surface stresses on the cracked surface. The authors and his group have developed a great number of database of “Kij”, the influence coefficient of the stress intensity factor, for many different types of surface-cracked structural components. The database is installed into the SCANP and the K-values for one of these cracks against an arbitrarily distributed surface stress can be calculated easily through the algorithm of superposition of the surface stress and the corresponding Kij data. The fatigue crack propagation can be simulated by integrating the Paris’ law, and it is easy to estimate the residual fatigue lives up to the leakage. Further, residual lives due to SCC, stress corrosion cracking, can be simulated by following the algorithm described in the JSME Standard. In this paper it is demonstrated how the SCANP works by applying it to some practical industrial problems such as fatigue crack and SCC crack propagations into welded residual stress field, and fatigue crack propagation initiated from double-surface cracks. In the latter case the simulation was compared with the experimental results in order to evaluate the validity of the developed system. It was found that the scatter of the material data describing the Paris’ law is far larger than the errors in estimating K-values, and therefore, the choice of these material data is very important when a user wants to use this program effectively. In order to use the developed program correctly, the authors have organized “SCANP User Meeting” where only the members can use the program. In the User Meeting the users give presentations about how they applied SCANP to analyze practical problems, and discuss about the validity of the modeling, and the computed results. In this paper some of these activities will be described, and the problem of verification, validation and uncertainty quantification is discussed.
A Consideration on Nonlinear Sloshing of a Rectangular Liquid Storage Container Subjected to Seismic Excitations
Authors: Fujita Katsuhisa , Asano Sho
Published: May 2012

The liquid storage container (tank) is listed as one of the facilities, which are apt to be damaged by an earthquake in industrial facilities. As for seismic damages of containers, the sloshing phenomenon of the liquid free surface and the bulging phenomenon which is due to the interaction between the container structure rigidity and the liquid pressure become problems. Here, we investigate a nonlinear sloshing. There are many papers, in which the perturbation method using the small parameter, and the Galerkin’s method are applied as a method of the nonlinear analysis. However, in this paper, a modal method which superposes the linear solutions using Fourier series’ representation is proposed. Furthermore, this method is applied to the seismic response analysis on the nonlinear sloshing of a rectangular liquid storage container.
Influence of Fluid Viscosity and Structural Specifications on Stability in the Vicinity of Critical Velocity of an Elastic Beam Subjected to Confined Axial Flow
Authors: Fujita Katsuhisa
Published: May 2012

Evaluation methodologies for the flow-induced vibration of an elastic beam subjected to an axial flow confined in a narrow passage are reported. In this paper, by using the analytical method already proposed by the author, a parameter study is performed on the dynamic stability of an elastic beam subjected to axial flow confined a narrow passage. The effects of support conditions of structures, structural damping, and fluid characteristics on the dynamic stability are clarified. Especially, paying attention on the vicinity of critical velocity, the effects of stabilization or destabilization on flutter and divergence are investigated by changing fluid viscosity and structural damping.
Analysis of the Tightening Process and the Cyclic Stress Amplitude of Studs and Tap Bolts
Authors: Fukuoka Toshimichi , Nomura Masataka , Takeda Yosuke , Mori Uichiro
Published: May 2012

When subjected to alternating external forces, a fatigue failure of bolted joint is most likely to occur around the first bolt thread for the case of bolt-nut connections. It has been substantiated in the previous paper by three-dimensional finite element analysis, in which the numerical models of bolted joints are constructed so as to accurately take account of the effect of thread helical geometry. In the cases of bolted joints clamped by studs and tap bolts, however, fatigue failures sometimes occur at other than the first bolt thread and frequently initiate around the far end female threads. In this paper, using the FE models with correct helical thread geometry, stress amplitude distributions along the thread root including the thread run-out are evaluated for both male and female threads. It is shown that the maximum stress amplitude in the male threads occurs at the thread root slight away from the first thread and the maximum value in the female threads is generated near the far end of the engaged threads. Also shown is the contact pressure distribution pattern at the interface between the plate and the block, which is inherent to the bolted joints clamped by studs and tap bolts and may be the major source of the specific stress amplitude distributions.
Validation of Code System DRAWTHREE-FAC for Evaluation of Wall Thinning due to Flow Accelerated Corrosion by PWR Feed Water Piping Analysis
Authors: Naitoh Masanori , Uchida Shunsuke , Okada Hidetoshi , Koshizuka Seiichi
Published: May 2012

The code system DRAWTHREE-FAC for evaluation of pipe wall thinning due to flow accelerated corrosion was validated by comparison of calculations with measurements at the secondary piping of a PWR plant. Distributions of flow velocity and temperature along the whole piping were calculated with the system code RELAP5 and corrosive conditions were calculated by a N2H4-O2 reaction analysis code. Precise flow turbulence at major parts of the piping was analyzed with a 3D computational fluid dynamics (CFD) code to obtain mass transfer coefficients at structure surfaces. In the CFD calculation, the κ-ε method was applied. Since the κ-ε method can not give detailed flow behavior in a boundary layer, the results were extrapolated with a wall function, a power law, and analogy of non-dimensional numbers to obtain mass transfer coefficients in the boundary layer. Then, wall thinning rates were calculated by coupling models of static electrochemical and dynamic oxide layer growth. The wall thinning calculation was focused on T-junction portions of a PWR feed water line. The wall thickness of the PWR secondary piping was measured by the ultrasonic testing. The calculated residual wall thicknesses after thinning agreed with the measurements within ±20% difference.
Evaluation of Tightening Process of Reamer Bolts by Cooled Fitting
Authors: Fukuoka Toshimichi , Nomura Masataka , Inagaki Shun
Published: May 2012

Reamer bolts are widely used for the bolted joints subjected to large shear forces. The most important application is the case of clamping rigid flanged shaft couplings which deliver large torques. The body diameter of a reamer bolt is basically equal to the bolt hole diameter. When installing into the joint, reamer bolts are cooled in order to temporarily reduce its diameter for easy insertion. Dry ice or liquefied nitrogen is commonly used to lower the bolt temperature. It is customary in the actual tightening operation that to save working hours, the tightening torque is applied to the reamer bolt while its temperature is still well below the ambient temperature. Accordingly, the reamer bolt inevitably elongates as its temperature increases to the ambient one, which leads to the reduction of axial bolt force. In this paper, an equation of simple form is derived, which can estimate the amount of bolt force reduction occurred during the tightening operation of reamer bolts by cooled fitting. It is shown that the reduction rate in axial bolt stress, per unit of temperature difference between the reamer bolt and the fastened plate, gradually increases as the grip length is increased, ranging from about 1MPa/K to somewhat in excess of 2MPa/K. If the target joint tightened by cooled fitting has an excessive strength, the derived equation is useful from the practical point of view. It can determine the excessive torque that compensates the bolt force reduction due to the elongation during the tightening operation. The effectiveness of the derived equation is demonstrated by experiments. Based on the derived equation and the experimental results, a guide line is proposed to securely tighten reamer bolts using cooled fitting.
Optimization of Pipe Repair Sleeve Design
Authors: Arif A. F. M. , Al-Nassar Y. N. , Al-Qahtani H. , Khan S. M. A. , Anis M. , Eleiche A. M. , Inam M. , Al-Nasri N. I. , Al-Muslim H. M.
Published: May 2012

For the repair of pipeline defects, repair sleeves are the most widely used method in petro-chemical industry. The objective of this work was to optimize the thickness of non-pressure containing repair sleeve, by refining the existing design practice. Laboratory studies involving instrumentation of small-scaled repair sleeve system coupled with finite element analysis were carried out to refine the design procedure and optimize the thickness of the sleeve. Using a unified approach for finite element modeling, including failure pressure estimation and simulation of sleeve installation pressure, twenty-four cases ranging from 6–60 inches nominal diameter were investigated. In this paper, the details of the optimization approach used in this investigation have been presented.
Advances in the Design and Qualification of Main Steam and Main Feedwater Isolation Valves (MSIV and MFIV) With Type A, Gas Hydraulic Actuators
Authors: Gradle Richard J. , Bensinger Floyd A.
Published: May 2012

Flowserve and its heritage companies have supplied valves for many of the critical applications within commercial nuclear power generation plants since the beginning of commercial nuclear power generation. Two of these highly critical applications are the Main Steam Isolation Valves (MSIVs) and the Main Feedwater Isolation Valves (MFIVs). As the requirements of these two applications have evolved, so have the applicable valve and actuator designs and their qualifications. Although functional qualification standards (ASME QME-1) for power operated valve applications have been developed within the US nuclear industry, their use is relatively new within the US. In order to globally supply these valves, Flowserve has functionally qualified its MSIVs and MFIVs to this standard. In addition to the valves, the actuators have also evolved. Flowserve’s type A, gas / hydraulic actuator has been updated to improve its reliability and provide better performance for the power plants. The updated Flowserve type A, type A, gas / hydraulic actuator has completed the Environmental Qualification testing in accordance with the requirements of IEEE 323, 344 and 382. The Flowserve MSIV and MFIV designs have been selected for installation in many of the new Generation 3 and Generation 3+ nuclear power generation plants. This paper briefly discusses: • The history of the MSIV and MFIV valve applications within the nuclear power generation plants and • The Flowserve MSIV and MFIV, ASME QME-1 functional qualifications. Major paper emphasis is placed on: • The latest updates to the Flowserve type A, type A, gas / hydraulic actuator design, and • The latest results of the Flowserve type A, gas hydraulic actuator environmental qualification to IEEE 323, 344 and 382.