bluuebunny/crossref_metadata_2025 huggingface dataset:
Modelling the Mixing Process of Liquids With Concentrates in Capsules
Authors: Giannopapa C. G. , van der Linden B. J. , van Druten W. , Bongers M.
Published: July 2009

In food industry mixing of concentrates contained in capsules with liquids such as milk or water for the production of warm drinks is becoming common practice the last couple of years. This process is characterized by complicated physical phenomena: the concentrates’ viscosity is temperature dependent, the liquid is non-Newtonian and the mixing process is turbulent. The industrial objective at the end of the process is a uniform liquid end product with as little as possible left over concentrate in the capsule. The optimization of the mixing process is typically done by trial and error in laboratories, which is time consuming and expensive. Computer models can significantly reduce the manufacturing costs associated with laboratory optimization and give a better insight of the process. The objective of this paper is to create a computer simulation model that is able to capture the physical processes occurring during the production of warm drinks using finite elements. The model should be able to correctly represent the mixing of the solid concentrate with the liquid injected inside the capsule compartment. Finite element method is used to solve the flow, heat exchange and concentration problem. In the paper different shapes of the capsule and how they influence the mixing are compared and their suitability for industry according to the amount of concentrate left in the capsule at the end of the process are assessed.
Prediction of Brittle Fracture Under Generalised Elastic Plastic Loading
Authors: Lewis S. J. , Truman C. E. , Smith D. J.
Published: July 2009

This article describes an investigation into the ability of a number of different fracture mechanics approaches to predict failure by brittle fracture under general elastic/plastic loading. Data obtained from C(T) specimens of A508 ferritic steel subjected to warm pre-stressing and side punching were chosen as such prior loadings produce considerably non-proportionality in the resulting stress states. In addition, failure data from a number of round notched bar specimens of A508 steel were considered for failure with and without prior loading. Failure prediction, based on calibration to specimens in the as received state, was undertaken using two methods based on the J integral and two based on local approach methodologies.
Measurement and Modelling of Residual Stresses in Fracture Toughness Specimens Extracted From Large Components
Authors: Lewis S. J. , Hossain S. , Truman C. E. , Smith D. J. , Hofmann M.
Published: July 2009

A number of previously published works have shown that the presence of residual stresses can significantly affect measurements of fracture toughness, unless they are properly accounted for when calculating parameters such as the crack driving force. This in turn requires accurate, quantitative residual stress data for the fracture specimens prior to loading to failure. It is known that material mechanical properties may change while components are in service, for example due to thermo-mechanical load cycles or neutron embrittlement. Fracture specimens are often extracted from large scale components in order to more accurately determine the current fracture resistance of components. In testing these fracture specimens it is generally assumed that any residual stresses present are reduced to a negligible level by the creation of free surfaces during extraction. If this is not the case, the value of toughness obtained from testing the extracted specimen is likely to be affected by the residual stress present and will not represent the true material property. In terms of structural integrity assessments, this can lead to ‘double accounting’ — including the residual stresses in both the material toughness and the crack driving force, which in turn can lead to unnecessary conservatism. This work describes the numerical modelling and measurement of stresses in fracture specimens extracted from two different welded parent components: one component considerably larger than the extracted specimens, where considerable relaxation would be expected as well as a smaller component where appreciable stresses were expected to remain. The results of finite element modelling, along with residual stress measurements obtained using the neutron diffraction technique, are presented and the likely implications of the results in terms of measured fracture toughness are examined.
Validation of the European FITNET Fitness-for-Service Fracture Assessment Procedure
Authors: Hadley Isabel , Wei Liwu
Published: July 2009

The new European fitness-for-service procedure FITNET includes a hierarchy of different approaches for fracture assessment, designated Option 0 to Option 5. The choice of Option depends on the information available to the user, and can include direct calculation of crack driving force by FEA, allowance for weld strength mismatch and allowance for crack-tip constraint. The fracture assessment procedures have been extensively validated by analysing the results of large-scale fracture tests and engineering failures in accordance with the procedure, and comparing actual with predicted behaviour. This paper presents a selection of the many hundreds of analyses carried out, with particular emphasis on the more advanced Options incorporating weld strength mismatch and crack-tip constraint. Failure of the testpieces was correctly predicted by the FITNET procedure in all cases (ie, the analysis point corresponding to failure of the specimen lay outside the failure analysis line of the FAD). The safety factor associated with the analysis is, however, shown to decrease as the more advanced Options are invoked, ie the analysis becomes more accurate and less conservative, as intended.
The RCC-M Code 2007 Edition: Emphasis on New Technical and Regulatory Aspects
Authors: Grandemange Jean-Marie
Published: July 2009

In the context of the issuing of the European Pressure Equipment Directive (PED) the objective of which is to suppress trade barriers within the European Union, a new regulation has been issued in France for nuclear pressure equipment. This text makes reference to the Decree transposing the PED, while completing these provisions by supplementary requirements having the objective to provide a very high level of integrity guarantee for equipments which are the most important for safety, and to cover the prevention of radioactive release risks. Referencing the PED, and including specific provisions, the Ministerial Order implies that Codes and specifications should be updated in the following two domains: • that of the conformity of Codes and Standards used, generally inspired from the ASME Code Section III, with the essential safety requirements of the PED, • that of the respect of the complementary provisions brought by the new Order applying to nuclear pressure equipments (ESPN). This paper presents the more significant conclusions of this work and the resulting amendments of the RCC-M Code, introduced by the 2007 edition to that Code. It also presents the main evolutions integrated in the 2007 edition of the RCC-M to cover projects needs and integrate the consequences of the evolutions of standards to which the code refers. As a general rule, the RCC-M makes the largest use of European and ISO standards in order to facilitate its application in various industrial and regulatory contexts.
Alignment Criteria for Through-Wall Flaws in Plates and Pipes
Authors: Bezensek B. , Miyazaki K.
Published: July 2009

Multiple flaws in vessels and pipes are frequently assessed as a larger single flaw in accordance with the flaw alignment and combination rules. In this paper the alignment of two through-wall flaws is examined for plates in tension and pipes in bending using detailed finite element (FE) modelling. The FE model is developed using the Gurson-Tvergaard damage model and accurately describes the flaw interaction and load capacity of a random experimental sample. Results suggests that two flaws should be aligned onto the same plane for the purpose of assessment when the separation between the parallel planes containing flaws equals the flaw length for both, plates in tension and pipes in bending.
Structural Integrity Assessment of Weldments at High Temperature: A Proposed New Approach for R5
Authors: Smith N. G. , Dean D. W. , O’Donnell M. P.
Published: July 2009

The majority of problems associated with the structural integrity of components, particularly those operating at high temperature, are associated with welds. The R5 procedures provide a comprehensive methodology for the assessment of structures operating within the high temperature creep regime. This includes advice on the modifications required to the basic procedure to account for weldments in creep-fatigue crack initiation assessments. The current approach is based on the use of a Fatigue Strength Reduction Factor (FSRF) which has a value according to the particular class of welded joint. The FSRF affects the calculation of creep and fatigue damage. However, the current approach can be excessively conservative for as-welded weldments which are the main type of weldments in plant. This paper outlines the proposed changes to R5, which seek to achieve the following objectives: • to simplify and clarify the current advice for creep-fatigue initiation assessments of weldments, whilst maintaining a conservative assessment procedure; • to have a robust procedure which can be applied to complex components and loading conditions. The new approach separates the FSRF into two components which are as follows: • the geometric strain enhancement due to the weldment geometry (if applicable) and the material mis-match effect between parent material and weld metal, which is called the Weld Strain Enhancement Factor (WSEF), and • the fatigue endurance reduction effect due to the presence of small imperfections (e.g. inclusions, porosity, etc.) in the weldment constituent materials, which is called the Weld Endurance Reduction (WER). The WSEF is used to determine the stress at the start of a dwell or hold period and, because it has a lower value than the FSRF (due to the removal of the WER), results in less conservative calculations of creep damage compared to the current procedure, which uses the full FSRF. For fatigue damage predictions, the modified route is broadly similar to the current route, since the combination of the WER and the WSEF in the modified route corresponds to the FSRF used in the current route. Assessments to demonstrate the improved endurance predictions using the proposed new approach have been performed on several creep-fatigue weldment features tests and examples are provided in this paper.
A Comparative Study of a Single Damage-Parameter Life Prediction Method and a Physically-Based Damage Mechanics Method
Authors: Leen S. B. , Hyde T. H. , Peravali S. , Sun W.
Published: July 2009

This paper presents a comparison between the life predictions from a physically-based damage mechanics formulation, which captures the fracture mechanism of intergranular cavitation, via three state variables, describing intergranular creep cavitation, coarsening of carbide precipitates and primary creep softening, and a corresponding single-damage mechanics formulation, of the Kachanov-type. The comparison is made for a 0.5Cr0.5Mo0.25V ferritic steel at a temperature of 635°C across a range of stress levels on the basis of a notched bar stress-state. Simplified predictions based on the R5 reference stress method are also given for comparative purposes.
Analysis of Embedded and Surface Flaws in the Beltline of an RPV Subjected to Cyclic Loading Associated With Normal Operational Cool-Down and Heat-Up Transients: A Scoping Study
Authors: Yin Shengjun , Dickson Terry L. , Williams Paul T. , Bass B. Richard
Published: July 2009

The influence of thermal-hydraulic cyclic loading on postulated embedded and surface-breaking flaws in the beltline region of a reactor pressure vessel (RPV) are investigated numerically. Over the service life of a nuclear power plant, the RPV is expected to undergo a sequence of cool-down and heat-up thermal-hydraulic transients associated with, for example, scheduled refueling outages (RFOs) or other normal operational transients. With respect to postulated surface or embedded flaws in the RPV wall, these scheduled operational transients produce cyclic, variable-amplitude, nonlinear, multiaxial applied loadings, albeit with possibly long dwell times between the active portions of the cycles. The on-going scoping study indicates that for very large flaws, the driving force increases rapidly with cyclic hardening but tends to saturate after several loading cycles as a stabilized condition is approached; while the driving force tends to keep constant under the cyclic loading for those smaller flaws frequently used in the Pressurized Thermal Shock (PTS) study.
Cleavage Fracture Analysis of Cladded Beams With an Embedded Flaw Under Four-Point Bending
Authors: Yin Shengjun , Williams Paul T. , Bass B. Richard
Published: July 2009

Semi-large scale embedded flaw beams were tested at Nuclear Research Institute (NRI) Rez in the Czech Republic for the 6th Network for Evaluating Structural Components (NESC_VI) project. The experiments included, among others, a series of semi-large scale tests on cladded beam specimens containing simulated sub-clad flaws. Oak Ridge National Laboratory (ORNL) conducted numerical studies to analyze the constraint issues associated with embedded flaws using various fracture mechanics methods, including T-Stress, hydrostatic stress based QH stress, and the Weibull stress model. The recently developed local approach using the modified Weibull stress model combined with the Master Curve methodology was also utilized to predict the failure probability (Pf) of semi-large scale beams. For this study, the Weibull statistical model associated with the Master Curve methodology was employed to stochastically simulate the fracture toughness data using the available Master Curve reference temperature T0 for the tested base material from the “aged” WWER-440 Reactor Pressure Vessel (RPV). The study was also conducted to investigate the sensitivity of predicted probability of failure of semi-large scale beams with embedded flaw with different Weibull shape parameters, m.
Hydrogen Effects on the Fracture Toughness Properties of Forged Stainless Steels
Authors: Morgan Michael J.
Published: July 2009

The effect of hydrogen on the fracture toughness properties of Types 304L, 316L and 21-6-9 forged stainless steels was investigated. Fracture toughness samples were fabricated from forward-extruded forgings. Samples were uniformly saturated with hydrogen after exposure to hydrogen gas at 34 MPa or 69 MPa and 623 K prior to testing. The fracture toughness properties were characterized by measuring the J-R behavior at ambient temperature in air. The results show that the hydrogen-charged steels have fracture toughness values that were about 50–60% of the values measured for the unexposed steels. The reduction in fracture toughness was accompanied by a change in fracture appearance. Both uncharged and hydrogen-charged samples failed by microvoid nucleation and coalescence, but the fracture surfaces of the hydrogen-charged steels had smaller microvoids. Type 316L stainless steel had the highest fracture toughness properties and the greatest resistance to hydrogen degradation.
The Inclusion of Inner Surface Breaking Flaws in Probabilistic Fracture Mechanics Analyses of Reactor Vessels Subjected to Planned Normal Cool-Down Transients
Authors: Dickson Terry , EricksonKirk Mark
Published: July 2009

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned reactor startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, a goal of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements that reduces the conservatism and also is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). Previous publications have been successful in illustrating potential methods to provide a risk-informed relaxation to the current regulations for normal transients. Thus far, probabilistic fracture mechanics (PFM) analyses have been performed at 60 effective full power years (EFPY) for one of the reactors evaluated as part of the PTS re-evaluation project. In these previous analyses / publications, consistent with the assumptions utilized for this particular reactor in the PTS re-evaluation, all flaws for this reactor were postulated to be embedded. The objective of this paper is to review the analysis results and conclusions from previous publications on this subject and to attempt to modify / generalize these conclusions to include RPVs postulated to contain only inner-surface breaking flaws or a combination of embedded flaws and inner-surface breaking flaws.
Twice-Yield Method for Assessment of Fatigue Caused by Fast Thermal Transient According to 2007 Section VIII-Division 2 of ASME B&PV Code
Authors: Kalnins Arturs
Published: July 2009

Fatigue analysis is performed for a pipe section subjected to a repeated transient cycle of bulk temperature inside the pipe. No pressure is cycled. Finite element transient heat transfer analysis is performed. The times of the stress reversals are determined by elastic stress analysis. The number of design cycles for cyclic service of the pipe is obtained following the fatigue assessment rules for unwelded locations using the elastic-plastic Twice-Yield method of 2007 Section VIII, Division 2, of ASME Boiler and Pressure Vessel Code. Details for the use of these rules are presented.
Probabilistic Multiscale Models for Fracture Analysis of Functionally Graded Composites
Authors: Chakraborty Arindam , Rahman Sharif
Published: July 2009

This paper reports three multiscale models, including sequential, invasive, and concurrent models, for fracture analysis of a crack in a two-phase, functionally graded composite. The models involve stochastic description of the particle volume fractions, particle locations, and constituent material properties; a two-scale algorithm including microscale and macroscale analyses for determining crack-driving forces; and two stochastic methods for fracture reliability analysis. Numerical results indicate that the sequential and invasive multiscale models are the most computationally inexpensive models available, but they may not produce acceptable probabilistic characteristics of stress-intensity factors or accurate probability of fracture initiation. The concurrent multiscale model is sufficiently accurate, gives probabilistic solutions very close to those generated from the microscale model, and can reduce the computational effort of the latter model by more than a factor of two. In addition, the concurrent multiscale model predicts crack trajectory as accurately as the microscale model.
Guidance on a Defect Interaction Effect for In-Plane Surface Cracks Using Elastic Finite Element Analyses
Authors: Huh Nam-Su , Choi Suhn , Park Keun-Bae , Kim Jong-Min , Choi Jae-Boong , Kim Young-Jin
Published: July 2009

The crack-tip stress fields and fracture mechanics assessment parameters, such as the elastic stress intensity factor and the elastic-plastic J-integral, for a surface crack can be significantly affected by adjacent cracks. Such a defect interaction effect due to multiple cracks can magnify the fracture mechanics assessment parameters. There are many factors to be considered, for instance the relative distance between adjacent cracks, crack shape and loading condition, to quantify a defect interaction effect on the fracture mechanics assessment parameters. Thus, the current guidance on a defect interaction effect (defect combination rule), including ASME Sec. XI, BS7910, British Energy R6 and API RP579, provide different rules for combining multiple surface cracks into a single surface crack. The present paper investigates a defect interaction effect by evaluating the elastic stress intensity factor of adjacent surface cracks in a plate along the crack front through detailed 3-dimensional elastic finite element analyses. The effects of the geometric parameters, the relative distance between cracks and the crack shape, on the stress intensity factor are systematically investigated. As for the loading condition, only axial tension is considered. Based on the elastic finite element results, the acceptability of the defect combination rules provided in the existing guidance was investigated, and the relevant recommendations on a defect interaction for in-plane surface cracks in a plate were discussed.
Creep Crack Growth of P92 Welds
Authors: Yatomi Masataka , Fuji Akio , Kobayashi Ken-ichi I. , Tabuchi Masaaki , Yokobori Takeo , Hasegawa Yasushi , Yokobori Toshimitsu
Published: July 2009

This paper represents creep properties and creep crack growth properties for P92 welds. The CCG tests were carried out using cross-welded compact tension (C (T) specimens at several temperatures. The crack front was located at HAZ region to simulate Type IV crack. Finite element analysis was conducted to simulate multiaxiality in welded joints and compare the experimental results. The constitutive behaviour for these materials is described by a power law creep model.
Effectiveness of Stainless Steel Buffer Layer to Address Hot Cracking During Weld Overlay Repair of Dissimilar Metal Alloy 82/182 Welds With Stainless Steel Piping
Authors: Cofie Nathaniel G. , Smith Richard E. , Bax Richard L. , Lohse Christopher S. , Hermanns Bill , Valsvig Jerry , Yepez Luis D. , Parker Dixon
Published: July 2009

Many pressurized water reactor (PWR) plants have used weld overlays to mitigate the pressurizer dissimilar metal welds that are susceptible to primary water stress corrosion cracking (PWSCC). These configurations typically consist of SA-508 Class 2 low alloy steel welded to a stainless steel safe end by Alloy 82/182 weld metal. The overlay weld metal is typically Alloy 52M. In a few cases, solidification cracking (hot cracking) has been observed on the stainless steel portion of the configuration when the first weld overlay layer is deposited. To overcome this problem, a process consisting of deposition of ER308L or ER309L stainless steel buffer layer for the first layer in conjunction with a low Power Ratio welding procedure has been developed and applied successfully. The Alloy 52M weld overlay is then deposited after the buffer layer. This paper discusses the causes of the hot cracking and test programs to develop the parameters for the welding of the buffer layer and subsequent weld overlay layers. In addition the results of analysis performed to determine the impact of the buffer layer on the post weld overlay residual stresses are also discussed.
Elastic-Plastic Analysis of Thick-Walled Toroidal Pressure Vessels
Authors: Adibi-Asl R.
Published: July 2009

Piping systems in process industries and nuclear power plants include straight pipe runs and various fittings such as elbows, miter bends etc. Elbows and bends in piping systems provide additional flexibility to the piping system along with performing the primary function of changing the direction of fluid flow. Distinctive geometry of these toroidal shell components result in a structural behavior different from straight pipe. Hence, it would be useful to predict the behavior of these components with acceptable accuracy for design purposes. Analytical expressions are derived for stresses set up during loading and unloading in a toroidal shell subjected to internal pressure. Residual stresses in the component are also evaluated. The proposed solutions are then compared with three-dimensional finite element analysis at different locations including intrados, extrados and flanks.
Creep-Environment Interactions of Alloy 617 at Elevated Temperature
Authors: Kim Daejong , Jang Changheui , Ryu Woo Seog
Published: July 2009

Creep behavior and degradation of creep properties of high-temperature materials often limit the lives of components and structures designed to operate for extended period under stress at elevated temperatures. A nickel-base superalloy, Alloy 617 in particular which is considered as a prospective material for hot gas duct and intermediate heat exchanger in very high temperature gas cooled reactor, was studied for creep properties. Creep tests were carried out under various sustained tensile loadings in air and helium environments at temperature of 800°C, 900°C, and 1000°C. Times for 1% creep strain and creep rupture were taken from the short-term creep tests within 1000 hours. Effect of creep-environment interaction on creep strain and changes in viscous deformation properties by dynamic recrystallization were discussed.
Recent Development of Code Case on Use of Ductile Cast Iron for Transport and Storage Cask for Spent Nuclear Fuel
Authors: Arai Taku , Saegusa Toshiari , Hueggenberg Roland
Published: July 2009

Code Case N-670 “Use of Ductile Cast Iron Conforming to ASTM A874/A 874M-98 or JIS G5504-1992 for Transport Containments, Section III, Division 3” which permits use of ductile cast iron for transport containments of spent nuclear fuel was revised to the Code Case N-670-1, “Use of Ductile Cast Iron Conforming to ASTM A874/A 874M-98 or JIS G5504-2005 for Transport and Storage Containments, Section III, Division 3”. Items revised were as follows: (a) Scope was expanded to use for transport and storage, and changed to conform year edition of JIS G5504, (b) The elongation requirement was deleted form the code case to reflect the change of year edition of JIS G5504, (c) Temperature condition of −40 °C was clearly provided for fracture toughness test, (d) Design fatigue curve was re-established, (e) External pressure chart was re-established. Technical basis of the revised code case are described in this paper.
Effect of Welding Conditions on Residual Stress and SCC Behavior at Butt-Welding Joints of Recirculation Pipes
Authors: Katsuyama Jinya , Tobita Tohru , Itoh Hiroto , Onizawa Kunio
Published: July 2009

Stress corrosion cracking (SCC) in recirculation pipes made of low carbon austenitic stainless steel (Type 316L) has been observed near butt-welding joints. The recent SCC grows near the welding zone mainly due to high tensile residual stress by welding since the effect of the other contributing factors of material and environment decreases due to the countermeasures. Therefore, the residual stress analysis due to welding of austenitic stainless piping is important and has been already conducted by many researchers. In present work, scatters of welding conditions such as heat input and welding speed were measured experimentally by producing a series of butt-weld specimens of Type 316L pipes. Distribution and its scattering of residual stress were also measured by non-destructive and destructive methods. The effects of welding conditions on residual stress have been evaluated by parametric FEM analyses considering the variation of some parameters based on the welding experiments. The effects of welding conditions on crack growth behavior have been also evaluated by SCC growth simulations using calculated residual stress distributions and a procedure in the fitness-for-service code. Welding parameters such as heat input and welding speed have a strong influence on crack growth rate since residual stress is also affected by scatter of these welding parameters.
Effect of Surface-Machining and Butt-Welding on Residual Stress and Hardness of Type 316L Stainless Steel Pipes
Authors: Asano Wataru , Katsuyama Jinya , Onizawa Kunio , Mochizuki Masahito , Toyoda Masao
Published: July 2009

Stress corrosion cracking (SCC) near the welded zone of core internals and recirculation pipes of Type 316L stainless steel have been observed to initiate at inner surface due to tensile residual stress and hardening. Residual stress and hardening should be induced by not only welding but also surface-machining in these regions. Surface-machining is conducted before and after piping butt-welding to match the ID of pipes and to provide a smooth surface finish. Therefore, we experimentally evaluated Vickers hardness distribution and proposed a simulation method to estimate residual stress by surface-machining using a local micron scale finite element method (FEM) model in previous paper (PVP2006) [1]. In present work, residual stress and Vickers hardness distributions for thickness direction were evaluated by experimental measurement. Residual stress simulation by surface-machining and welding was carried out by proposed simulation method applied to simulate the butt-welding of full scale FEM model for weld specimen. Redistribution behavior of residual stress by welding after surface-machining was discussed based on experimental and analytical results. Estimation of Vickers hardness distribution by surface-machining was performed based on strained region map observed by Field emission scanning electron microscope (FE-SEM) equipped with electron backscatter diffraction (EBSD) and compared with experimental measurements. The effect of surface-machining before/after welding on distribution of residual stress and hardness are presented.
Complex Monitoring Programme for WWER Reactor Pressure Vessels Lifetime Assessment in the Czech Republic
Authors: Brumovsky Milan , Kytka Milos , Novosad Petr , Brynda Jiri
Published: July 2009

Lifetime of reactor pressure vessels practically depends on a level of degradation of RPV material properties during operation. The most important degradating mechanism of RPV materials is usually radiation damage, characterized by values on neutron fluence on one side and radiation embrittlement of RPV materials on the second side. WWER reactor pressure vessels in the Czech Republic are a subject of a very thorough and complex monitoring program, that includes: • Standard material surveillance program containing of WWER-440 RPV materials — base metal, weld metal, heat affected zone, but irradiated with high lead factor (13 to 18), • Supplementary surveillance program of WWER-440 RPV materials, including additionally austenitic cladding materials, IAEA reference material JRQ irradiated with low lead factor (2 to 3) with parts subjected to annealing and re-irradiation after annealing, • Modified surveillance program of WWER-1000 RPV materials — base metal, weld metal, heat affected zone, cladding materials, IAEA reference JRQ material irradiated in low lead factor (2 to 3) near RPV inner beltline region, • Integrated surveillance specimen program for WWER-1000 reactor including materials from NPP Temelin (Czech Republic), Belene (Bulgaria), Kalinin (Russia) and Ukranian NPPs, • Continous exvessel monitoring of neutron fluence on outer RPV surface for both WWER-440 and WWER-1000 plants, • Neutron fluence determination on inner RPV surface (austenitic cladding) using special technique for removal of specimens from cladding for Nb activity measurements, • Ex-vessel temperature measurements during RPV operation. All these programs serve for precision of operation conditions and determination of degradation of RPV materials for RPV integrity and lifetime assessment.
Assessment of the Influence of Plasticity and Constraint on Measured Residual Stresses Using the Contour Method
Authors: Dennis R. J. , Bray D. P. , Leggatt N. A. , Turski M.
Published: July 2009

The contour method is a relatively new relaxation method for residual stress measurement and may be seen as an evolution of established methods such as hole drilling. The general procedure when applying the contour method is cutting, measurement and calculation of residual stress normal to the cut plane using Bueckner’s principle of elastic superposition. That is the residual stresses are determined from the measured profile of a cut surface. While the contour method is simple in concept there are certain underlying issues relating to the cutting process that may lead to uncertainties in the measured results. Principally the issues are that of constraint and plasticity during cutting and the influence they have on the measured residual stresses. In this paper both issues are investigated in detail by simulating the entire contour method process using finite element techniques for two welded specimens. Constraint has been a recognised concern for the contour method with the general requirement being to hold the specimen as rigidly as possible. Both clamping and fixing bolts are routinely used however in reality these methods do not provide a fully rigid constraint. In this work a range of constraints have been examined to determine the influence on the measured residual stresses. Plasticity, as a consequence of the cutting process, has also been recognised as a factor which may affect the measured residual stresses. In this work the extent of plasticity is predicted by simulation of the cutting process. With a known initial residual stress field the effects of plasticity are directly quantifiable. This work therefore provides an extremely useful insight into some of the key issues that affect the measurement performance of the contour method.
Numerical Methods to Predict Distortion in Welded Components
Authors: Leggatt N. A. , Dennis R. J. , Bouchard P. J. , Smith M. C.
Published: July 2009

Numerical methods have been established to simulate welding processes. Of particular interest is the ability to predict residual stress fields. These fields are often used in support of structural integrity assessments where they have the potential, when accurately characterised, to offer significantly less conservative predictions of residual profiles compared to those found in assessment codes such as API 579, BS7910 and R6. However, accurate predictions of residual stress profiles that compare favourably with measurements do not necessarily suggest an accurate prediction of component distortions. This paper presents a series of results that compare predicted distortions for a variety of specimen mock-ups with measurements. A range of specimen thicknesses will be studied including, a 4mm thick DH-36 ferritic plate containing a single bead, a 4mm thick DH-36 ferritic plate containing fillet welds, a 25mm thick 316L austenitic plate containing a groove weld and a 35mm thick esshete 1250 austenitic disc containing a concentric ring weld. For each component, distortion measurements have been compared with the predicted distortions with a number of key features being investigated. These include the influence of ‘small’ vs ‘large’ strain deformation theory, the ability to predict distortions using simplified analysis methods such as simultaneous bead deposition and the influence of specimen thickness on the requirement for particular analysis features. The work provides an extremely useful insight into how existing numerical methods used to predict residual stress fields can be utilised to predict the distortions that occur as a result of the welding fabrication process.
Effect of Vertical Pipe Length on the Thermal Stratification in Nuclear Power Plant Surgeline
Authors: Kim Young-Jong , Kim Maan-Won , Lee Hyun-Soon , Ko Eunmi
Published: July 2009

During operating transients of the pressurizer, thermal stratification effect may occur especially in the horizontal parts of the surge line. US NRC requires consideration of thermal stratification in surge line as phenomenon that must be considered in the design basis of the surge line. Generally, the fatigue usage factor of the surge line is comparative high, due to its operating temperature and pressure transients and its thermal stratification loads. In this study we have performed some parametric fluid-structure interaction analyses with different length variables of the vertical part of the surge line to study the relationship between the magnitude of thermal stratification and the length of vertical part of the surge line. The conservativeness of the traditional finite element model for thermal stratification analysis based on the conservative assumption in the surge line was also discussed by comparison of the results of three-dimensional time transient fluid-structure interaction analysis of this study. Stresses calculated with three-dimensional time transient model were considerably reduced comparing with the traditional analysis.
“Adaptative” User Interface as a Support for Evaluation Analysis in the Context of Asset Management
Authors: Hai¨k P. , Parfouru S. , Fessart K. , Lonchampt J. , Remy E.
Published: July 2009

The long term management of a production asset raises several major issues among which rank the technical management of the plant, its economics and the fleet level perspective one has to adopt. Decision makers are therefore faced with the need to define long term policies (up to the end of asset operation) which take into account multiple criteria including safety (which is paramount) and performance. In this context, EDF “PWR Durability I & II” research projects have consecutively been launched, since 2001, at EDF – Research & Development in order to develop methods and tools for EDF fleet. The aim of this paper is: • to summarize and analyze the research work that has been performed by EDF – R&D in the field of decision making for nuclear power plant maintenance and operation during the past seven year; • to highlight the strong and weak points of the developed methodology and tools and to identify the research work needed in order to ensure their use by EDF decision makers; • to introduce and illustrate our last development based on the use of an “adaptative” man/machine user interface in order to allows technical/strategic experts and decision makers to consult the useful pieces of knowledge in a context dependent way and, thus, facilitate the validation of the knowledge-base content and the analysis of the processed results. As a result, in this paper, we first remind the reader of the EDF overall methodology for asset management and its adaptations to plant-level life cycle management and to fleet-level component major replacement or capital investment management. We then focus on the three software tools that implement this methodology in order to allow decision makers, in several different contexts to define, evaluate and analyze long term plant operation and maintenance policies, major component replacement policies and capital investment strategies. We also show how the methodology and the software tools were used, from 2003 to 2007, on several pilot case studies. Examples of technical and economic results obtained for plant level pilot case study is quickly described as well as the kinds of conclusions one can draw from them in order to help decision makers evaluate and analyze long term asset management strategies or compare different plants. We then present the opinion of EDF’s decision makers about the developedmethodology and tools — and their use — and our understanding of their feedback. Lastly, we illustrate, using examples of technical and economic knowledge, data and results obtained from our previous pilot case studies, how the concept of an “adaptative” man/machine user interface could be used in order to facilitate the mastering of the methodology and tools’ complexity and to support decision makers’ evaluation and analysis of long term asset management strategies.
The Prediction of Residual Stresses in an Autogenously Welded Ferritic Beams
Authors: Warren A. P. , Symington I. , Bate S. K. , Francis J. A. , Turski M. , Hurrell P. R.
Published: July 2009

The continued safe and reliable operation of plant invariably has to consider the assessment of defects in welded structural components. This often requires some estimate of the residual stresses that have developed during the welding fabrication process. Increases in the computational power available to finite element analysts have made the prediction of welding residual stresses using finite element methods an increasingly viable option. Consequently recent years have seen many advances in the field of residual stress modelling. However, relatively little work has been conducted on the accurate modelling of welded ferritic components. This is largely due to the added complications associated with the solid-state phase transformations that occur during the heating and cooling of such steels. Against this background, a programme of work has been undertaken in order to investigate the effects that phase transformations have on the development of residual stresses in ferritic components and develop methods for their simulation. To facilitate this, generic components of increasing complexity are being studied. Reported in this paper is the first phase of this programme which considers SA508 beam specimens. These specimens have been subjected to autogenous TIG welds using two different torch travel speeds. In order to predict the resultant residual stresses, simulations have been conducted using the commercial finite element package SYSWELD. These predicted stresses are then compared with residual stress measurements conducted using the neutron diffraction technique.
Liquid Sealing Materials: Sealing Gap of Liquid Gaskets
Authors: Reinhardt Stefan , Reibert Jan-Peter , Haas W.
Published: July 2009

The design of fluid sealed surface areas is a challenge for any engineer, due to mechanical, thermal and chemical influences. In the field of drive technology a further challenge is the transmission of torques onto the sealing area. Designers have to consider all these influences when making the pertinent layout. If sealing fluids are used as sealing agents, the designer has to decide whether to design the sealing area as a mere force-locked connection or as a combination of force-locked and adhesive bonded. When connections which are force-locked and join adhesive bond are chosen connections conditions and details at the sealing gap have to be known. Due to new measuring techniques and the visualization of the measuring results, it is possible to obtain all the necessary details of the surface sealing area.
EPR: Tests Performed to Confirm the Mechanical and Hydraulic Design of the Vessel Internals
Authors: Dolleans Philippe , de Monplanet Charlotte , Fontes Jean-Philippe
Published: July 2009

The EPR is an Evolutionary high-Power Reactor which is based on the best French and German experience of the past twenty years in plant design construction and operation. In the present detailed engineering phase of the plant under construction in Finland (Okiluoto 3) and in France (Flamanville 3), some actions were conducted in order to improve the knowledge of the hydraulic behavior of the innovative Reactor Pressure Vessel internals (RPV). The RPV internals are mainly derived from former French N4 or German Konvoi with some evolutions to take into account the operating experience. Design and validation of the internals were performed within AREVA’s engineering teams, which develop state of the art methods in the field of thermohydraulic testing. The experimental validation program was closely followed by EDF. Moreover, an EDF R&D project, whose results are not addressed here, was held to consolidate the RPV internals conception. The aim of the paper is to present the hydraulic tests performed on mock-ups to characterize the hydraulic behavior of the innovative EPR Reactor Pressure Vessel internals, and to introduce the role of these tests in the global conception process of the EPR RPV internals (CFD code qualification, design validation, database...). The qualification of the CFD computer codes will be described in a forthcoming paper. Three different mock-ups are presented to illustrate these tests: • JULIETTE for the reactor pressure vessel lower internals, • ROMEO for the reactor pressure vessel upper internals, • MAGALY for the design of the skeleton-type control rod guide assembly.
Use of Small Punch Notched Specimens in the Determination of Fracture Toughness
Authors: Lacalle Roberto , A´lvarez Jose´ Alberto , Gutie´rrez-Solana Federico
Published: July 2009

In recent years an increasing interest has been shown in materials characterisation techniques based on the use of reduced dimension samples. One of these techniques, first developed in the early 80’s, is Small Punch testing, in which a punch acts over a plane small specimen deforming it until fracture. So far, this test has been used as a tool for the estimation of mechanical properties, for characterising ductilebrittle transition or for knowing the evolution of these properties with ageing phenomena as neutron irradiation. The Small Punch Test has even been used in the determination of fracture toughness by means of indirect correlations or complex methodologies based on damage models and Finite Element simulations. In this paper, a direct methodology for the determination of fracture toughness from Small Punch tests is proposed. The approach is based on the use of simply modified Small Punch specimens in which a lateral notch was machined. Attending to energy considerations, the J-Δa curve for a ship-building Grade A steel, as well as for a structural steel, E690, has been obtained using this type of samples and the results have been compared with the calculated results from conventional fracture tests.
Fatigue Analysis Round Robin of Nozzle-to-Plate Juncture: Results of Analytic-Test Comparison
Authors: Jones D. P. , Adams S. A.
Published: July 2009

The ASME Boiler and Pressure Vessel Code (B&PV) Committee, Subgroup on Design Analysis, established the Task Force on Elastic-Plastic Finite Element Analysis (TF-EP-FEA) to investigate how to use EP-FEA to assess ASME B&PV Code stress limits. This paper presents results of an investigation by the TF-EP-FEA to apply EP-FEA to predict cyclic life of a nozzle-to-plate juncture tested by DeJesus et al. [1]. EP-FEA results are compared to fatigue test results for four specimens cyclically loaded from zero to maximum load. An analytical exercise was defined such that the round-robin participants were to calculate the cyclic life of the nozzle-to-plate fatigue specimens two ways; by using ASME B&PV Code with elastic methods and by using an EP-FEA program. The results and techniques used by the investigators are compared and conclusions are reached regarding which method correlated best for this problem. Commonalties in the approaches are listed so that they may be used to develop general guidelines for use of elastic-plastic FEA to show satisfaction of the fatigue design rules of the ASME B&PV Code.
Nuclear PWR Plants Life Management Reactor Pressure Vessel Strategy Evaluation for Fluence in Relation With Integrity Assessment
Authors: Bezdikian Georges
Published: July 2009

The French utility has organized the life management program of Nuclear Plants in function of several actions: • Fluence assessment, • Material characteristic and RTndt at the end of life, • Transient evaluation, • Integrity assessment, • Life management and maintenance strategy. This paper shows the studies engaged to evaluate and the monitoring on the fluence evaluation and flux factor assessment. Also this paper shows the axial and azimuthally variation of fluence studies and how these results were coupled with in-service inspection and integrity assessment. The applications were for 3-loop PWR plants and also 4-loop PWR PLANTS. The monitoring was mainly oriented on evaluation of the ratio computation and measurement on dosimetry introduced in irradiation surveillance program capsules that were withdraw after irradiation inside vessel. Some results on comparison between measurements and calculation of flux factors are presented. A discussion on good comparison on values obtained for all reactor pressures vessels from this comparison between flux factors calculations and measurements indicate that this strong methodology gives good results on RPV neutron studies.