bluuebunny/crossref_metadata_2025 huggingface dataset:
Inherent Margin in the Brittle Failure Assessment for RPV
Authors: Siegele Dieter , Varfolomeyev Igor , Nagel Gerhard
Published: July 2009

The brittle fracture assessment of the reactor pressure vessels (RPV) of the EON Kernkraft (EKK) pressurized water reactors has been updated to the state of the art. The reference temperatures of the RPV for the design fluence of 5E18 n/cm2 are at or below 0°C. The weld joint between vessel flange and cylinder as well as the hot and the cold leg nozzle corners proved to be the prominent and covering situations for the RPV. The fracture mechanics results for these configurations for operational and loss of coolant transients (LOCA) were produced rerunning the assessment procedure established for the licensing of the RPV and additionally with the application of state of the art procedure. Big margin inherent to the older procedure was disclosed with the more realistic but still conservative input. Most of the margin resulted from the application of the fracture toughness based reference temperature RTTo in place of RTNDT, from the constraint representative fracture toughness applied in the nozzle corner situation and from taking into account the results of the non destructive evaluation (NDE) during the manufacture and the in-service inspection (ISI) with defect postulates of allowable size. The results show the load paths with large distance to the fracture toughness curve, in the upper shelf temperature region of fracture toughness and/or with negligible values far below the lower shelf, both demonstrating an excessive inherent safety of the older procedure. In addition, for the LOCA, a multi barrier concept is applied. Besides the flawless state of the vessel and the preclusion of initiation from postulated flaws, with the arrest of postulated crack initiation and break preclusion for crack extension from postulated rising load after crack arrest two additional barriers against failure were proven. For all load cases the relevant load path temperatures proved to be in the upper shelf regime of fracture toughness, where ductile fracture is the failure mechanism and brittle fracture is precluded.
Damage Evolution and Life Prediction of a P91 Longitudinal Welded Tube Under Internal Pressure Creep
Authors: Ogata Takashi , Sakai Takayuki , Yaguchi Masatsugu
Published: July 2009

Clarification of creep damage mechanism and establishment of remaining life prediction methods of longitudinal welded piping of P91 steel are important subjects to maintain reliable operation of boilers in thermal power plants. Internal pressure creep tests were conducted on P91 steel longitudinal welded tubes to characterize the evolution of creep damage with time and to evaluate a life prediction method. Interrupted creep tests were utilized for damage observation in addition to rupture tests. Three dimensional FE creep analysis of the creep tested specimens were conducted to identify stress and creep strain distribution within the specimen during creep. Failure occurred at a heat affected zone without significant macroscopic deformation. It was found that initiation of creep voids had concentrated at mid-thickness region rather than surface. The creep analysis results indicated that triaxial tensile stress yielded at the mid-thickness region of the heat affected zone due to difference of creep deformation property between the base metal, heat affected zone and weld metal. It was suggested that the triaxial stress state caused acceleration of the creep damage evolution in the heat affected zone resulting in internal failure of the tube specimens. A rupture time prediction method of the welded tube is proposed based on the maximum principal stress and the triaxial stress factor. Void growth behavior in the heat affected zone was well predicted by the previously proposed void growth simulation method by introducing void initiation function to the method.
A Holistic Approach to Structural Integrity Methods Supporting Graphite Core Assessments
Authors: Elcoate C. D. , Davis G. M. , Easterbrook L. E.
Published: July 2009

An understanding of the behaviour of radially-keyed graphite cores and the internal stress and deformation of individual core components under normal operation and transient conditions is essential in assessing core geometry, component integrity, inspection strategy and consequently core integrity. When assessing graphite core integrity an in-depth understanding of the broad range of material and physical behaviours of the core during steady-state, transients and fault conditions is essential. This paper describes recent advances in the approach developed to assess Magnox reactor core structural integrity, which is one of the legs comprising a graphite core safety case, the other legs being inspection, monitoring and consequences.
Some Issues in Fitness for Service Assessment of Wall Thinned CANDU Feeder Pipes
Authors: Jin John C. , Eom Seyun , Awad Raoul
Published: July 2009

Canadian CANDU® feeder pipes experiencing pipe wall thinning due to flow accelerated corrosion (FAC) are accepted for continued service after an engineering evaluation. This evaluation is based on the assumption that FAC degradation is manageable through a comprehensive inspection program and conservative engineering evaluations. The practice of the Canadian nuclear industry is to: establish a minimum acceptable wall thickness, compare the measured thickness to predictions from the previous outage to confirm the conservatism of the predictions in a condition assessment, and predict the thickness at the next inspection and compare against the minimum acceptable value in an operational assessment. If the thickness measured during outage does not meet the pre-established thickness criteria, the feeder should be replaced, unless it is demonstrated to be fit for service through a detailed analysis. The detailed analysis usually involves more complex methodologies which are subjected to regulatory reviews. Several issues have been raised in the fitness-for-service assessments of feeder pipes relating to the definition of primary membrane stress, interpretation of minimum thickness requirements, plasticity analysis, limit load analysis and the applicability of procedures given in Code Case N-597 to Class 1 feeder pipes. This paper presents the Canadian regulatory expectations on these issues.
Recent Developments of RCC-MR Code for High Temperature Reactor and ITER Projects
Authors: Escaravage Claude , Bonne Dominique , Gelineau Odile , Sperandio Morello , Drubay Bernard
Published: July 2009

The fourth edition of the RCC-MR code has been issued in French and English versions on October 2007 by AFCEN (Association Franc¸aise pour les re`gles de Conception et de Construction des Mate´riels des Chaudie`res Electro-nucle´aires). It is the result of an important work by Areva NP and CEA to develop and improve design and construction rules of the previous edition. The objective of this paper is to emphasize the progress achieved from the last version dated 2002 such as: • Enlargement of the scope of the code to be applicable not only to high temperature structures but also to the ITER vacuum vessel. • Adaptation of the code to the European standardization, making use of the Harmonized European Standards. • Improvement of the creep-fatigue rules for shells and pipes. • Development of the “class 2 box” structures directly applicable to the ITER vacuum vessel. • Introduction of a specific appendix dealing with specific fabrication requirements of the ITER vacuum vessel. • Extension of the scope of subsections devoted to bolts. • Development of the chapter on laser welding. • Introduction of requirements in line with the new European Pressure Equipment decree and its French declination to nuclear equipment (ESPN) which is mandatory for a French site construction.
Concept for the Guarantee of the Integrity of Bolted Flanged Connections in a German Nuclear Power Plant
Authors: Schaaf Manfred , Schoeckle Friedrich , Bartonicek Jaroslav
Published: July 2009

In nuclear facilities in Germany a systematic treatment of all sealing connections which are opened during the annual shutdowns is performed since several years. Special attention is paid to a design which meets specifications of the bolted flange connections (BFC’s). Apart from the use of suitable and certified materials this includes also the calculation of the required torques in accordance to the present rules. For the calculation procedure the nuclear code KTA 3211.2 (draft 2003) is applied which allows a tightness proof of the BFC, besides a stress analysis. For this calculation, experimentally achieved gasket characteristics according to DIN 28090-1 are required, which are either supplied by the gasket manufacturers or determined in the amtec test laboratory. The geometry and material data of each BFC are stored in a special data base. In addition, the data of the operating state as well as the design and test loading cases are included in the data base, so that all inputs for a later calculation are available. The results of the calculation which must be checked and approved by an expert third-party are stored in the flange data base, too (assembly state settings of each flange connection). On this base individual flange data sheets can be provided to the flange fitter teams that include all necessary information for assembly of each BFC (bolts, nuts, gasket, geometry and material specifications, lubricant, etc.).
Application of Direct-Current Potential Difference Method to Evaluation of Various Damages in Conductive Materials
Authors: Tada Naoya
Published: July 2009

The direct-current potential difference method (DC-PDM) is known as one of the best techniques to evaluate the damage initiated in structural materials. However, it is not still very popular and has not yet been widely used. One of its main reasons is that the potential difference measured is not supported based on the physical background and the relationship between the damage and the potential difference is not clarified quantitatively. In this paper, DC-PDM is classified into several categories based on the distribution of defects, the number of potential measurement probes, and the location of current supply probes. Characteristics of all categories are described.
Fracture Assessment of Structural Component With Residual Stress on the Basis of the Weibull Stress Criterion
Authors: Yamashita Yoichi , Minami Fumiyoshi
Published: July 2009

This paper studies the method for estimating the residual stress effects on brittle fracture of structural component based on the Weibull stress criterion. Experiments show that the critical CTOD and the critical load of wide plate with welding residual stress are apparently smaller than those of wide plate without residual stress. It has been found that the critical CTODs of wide plate with and without residual stress can be predicted from the 3PB fracture toughness test results based on the Weibull stress criterion. Constraint loss effects on CTOD of wide plate with residual stress can be assessed by the equivalent CTOD ratio. The equivalent CTOD ratio β is defined as the ratio, β = δ/δWP, where δ and δWP, are CTODs of the standard fracture toughness specimen and wide plate, respectively, at the same level of the Weibull stress. Calculation results of beta are also shown for various residual stress levels and crack lengh based on the Weibull stress criterion. Fracture assessment results using β are shown within the context of CTOD design curve. An excessive conservatism observed in the conventional procedure is reasonably reduced by applying the equivalent CTOD ratio, β.
Revised LOCA Frequency Estimates From an Expert Elicitation Process
Authors: Scott Paul M. , Tregoning Robert Lee , Abramson Lee Richard
Published: July 2009

The double-ended-guillotine break (DEGB) criterion of the largest primary piping system in the plant, which generally provides the limiting condition for the emergency core cooling system requirements, is widely recognized as an extremely unlikely event. As a result, the US Nuclear Regulatory Commission (NRC) is considering a risk-informed revision of the design-basis break size requirements for commercial nuclear power plants. In support of this effort, loss-of-coolant accident (LOCA) frequency estimates were developed using an expert elicitation process by consolidating service history data and insights from probabilistic fracture mechanics (PFM) studies with knowledge of plant design, operation, and material performance. This paper describes, and presents the results for, two of the sensitivity analyses conducted as part of this effort (overconfidence adjustment and aggregation method) to examine the assumptions, structure, and techniques used to process the elicitation responses to develop group estimates of the LOCA frequency estimates.
An Experimental and Numerical Investigation of Flow Over a Hydrofoil in Transient Regimes Based on Wall-Pressure Analysis
Authors: Ducoin Antoine , Astolfi Jacques-Andre´ , Deniset Franc¸ois , Sigrist Jean-Franc¸ois , Soyer Vincent
Published: July 2009

The present study is developed within the general framework of marine structure design of lifting bodies, operating in transient regimes. The study concerns the experimental and numerical investigations of time-space distribution of the wall pressure field on a NACA66 hydrofoil in forced ramping motion. The angle of incidence varies from 0° to 15° beyond stall. Experiments in a hydrodynamic tunnel and corresponding RANSE based code calculations are carried out for various pitching velocities. Transducers are located along the chord of the hydrofoil. The numerical approach is conducted in turbulent regime using recent laminar to turbulent transition model. Global coefficients are analyzed in order to quantify the transient effects. The comparison of calculated to measured local wall pressures on the suction side leads to the identification of the effect of pitching velocity on hydrodynamic loading. Moreover, the consequences of pressure fluctuations induced by the laminar to turbulent transition are highlighted. The evolutions of transition, laminar bubble and leading and trailing edge detachments are discussed in the context of naval applications.
Comparison of the Isochronous Method and a Time-Explicit Model for Creep Analysis
Authors: Koves William , Zhao Mingxin
Published: July 2009

The design of components or structures at elevated temperature is complex. The use of rigorous time dependent material models may not be practical for many large scale industrial problems. The use of simplified methods permits the creep analysis of components that would be impractical by rigorous time dependent models. The Isochronous Stress-Strain method is an approach that has been used extensively for the creep evaluation of elevated temperature components. The method has been used for the analysis of problems containing both primary and secondary stresses. The method has also been used to evaluate creep buckling problems. Although the method has been accepted as an alternative to a full time dependent creep analysis, the limitations and accuracy of the method have not been investigated systematically and are not fully understood. This study compares the isochronous stress-strain method with a generalized time-explicit creep model for materials in high temperature applications. Analytical solutions are developed for three basic loading configurations, including uniaxial tension, pure bending, and torsion in either load or displacement controlled conditions. Deformations, stresses, and creep strains are compared between the two different methods.
Effective Lifetime Estimate of Crimped Powerline Splice Connector Operated at High Temperature
Authors: Wang John Jy-An , Lara-Curzio Edgar , King Tom , An Ke , Hubbard Camden
Published: July 2009

This paper addresses the thermal-mechanical properties and performance characteristics of full tension splice connectors under high temperature operation, in particular those used in overhead transmission and distribution lines. Due to the increase in power demand existing overhead power transmission lines often need to operate at temperatures higher than those originally considered for their design. This has led to the accelerated aging and degradation of splice connectors. The compressive residual stresses induced by the crimping process within the splice connector provide the clamping forces to secure the conductor and therefore, the determination of the state of compressive residual stresses in splice connectors is a necessary requirement to provide an accurate estimate of their service lifetime. This paper presents a protocol for integrating analytical and experimental approaches to evaluate the integrity of a full tension single-stage splice connector assembly.
Environmental Fatigue Test of CF8M With a Small Autoclave Simulating PWR Conditions
Authors: Jeong Il-Seok , Ha Gag-Hyeon , Kim Tae-Ryoung
Published: July 2009

To develop a fatigue design curve of cast stainless steel CF8M used in primary piping material of nuclear power plants, low-cycle fatigue tests have been conducted by Korea Electric Power Research Institute (KEPRI). A small autoclave simulated the environment of a pressurized water reactor (PWR), 15 MPa and 315 °C. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitudes at 0.04%/s strain rate. A small autoclave of 1 liter and cylindrical solid fatigue specimens were used for the strain-controlled low cycle environmental fatigue tests to make the experiments convenient. However, it was difficult to install displacement measuring instruments at the target length of the specimens inside the autoclave. To mitigate the difficulty displacement data measured at the shoulders of the specimen were calibrated based on the data relation of the target and shoulder length of the specimen during hot air test conditions. KEPRI developed a test procedure to perform low cycle environmental fatigue tests in the small autoclave. The procedure corrects the cyclic strain hardening effect by performing additional tests in high temperature air condition. KEPRI verified that the corrected test result agreed well with that of finite element method analysis. The process of correcting environmental fatigue data would be useful for producing reliable fatigue curves using a small autoclave simulating the operating conditions of a PWR.
Shakedown Limit Loads for 90-Degree Scheduled Pipe Bends Subjected to Constant Internal Pressure and Cyclic Bending Moments
Authors: Abdalla Hany F. , Megahed Mohammad M. , Younan Maher Y. A.
Published: July 2009

A simplified technique for determining the shakedown limit load for a long radius 90-degree pipe bend was previously developed [1, 2]. The simplified technique utilizes the finite element method and employs the small displacement formulation to determine the shakedown limit load (moment) without performing lengthy time consuming full cyclic loading finite element simulations or utilizing conventional iterative elastic techniques. The shakedown limit load is determined through the calculation of residual stresses developed within the pipe bend structure. In the current paper, a parametric study is conducted through applying the simplified technique on three scheduled pipe bends namely: NPS (Nominal Pipe Size) 10" Sch. No. 20, NPS 10" Sch. No. 40 STD, and NPS 10" Sch. No. 80. Two material models are assigned namely; an elastic-perfectly-plastic (EPP) material and an idealized elastic-linear strain hardening material obeying Ziegler’s linear kinematic hardening (KH) rule. This type of material model is termed in the current study as the KH-material. The pipe bends are subjected to a spectrum of constant internal pressure magnitudes and cyclic bending moments. The cyclic bending includes three different loading patterns namely: in-plane closing (IPC), in-plane opening (IPO), and out-of-plane (OP) bending moment loadings of the pipe bends. The shakedown limit moments output by the simplified technique are used to generate shakedown diagrams of the scheduled pipe bends for the spectrum of constant internal pressure magnitudes. A comparison between the generated shakedown diagrams for the pipe bends employing the EPP- and the KH-materials is presented. Relatively higher shakedown limit moments were recorded for the pipe bends employing the KH-material at the medium to high internal pressure magnitudes.
Using Small-Scale Testing in Integrity Assessment of Clad Pipes
Authors: Motarjemi Afshin
Published: July 2009

Utilization of clad pipes has significantly improved the performance of components used in the oil and gas industry because of their potentially higher corrosion resistance. However, a clad pipe has more complexity than a solid pipe in terms of fracture mechanics based fitness for service (FFS) assessment. To be able to properly address this issue, precise mechanical data, such as tensile properties and fracture toughness values of the both cladding and the backing materials should be known. However, as the cladding layer is normally thinner than the backing material, conventional specimens are unable to provide the required mechanical data and instead small-scale tensile and fracture toughness specimens should be used. Current paper will introduce some of the small-scale mechanical testing techniques that may be used to determine the required input data for performing a fracture-mechanics based FFS assessment of a clad pipe.
Fracture Behavior of Dissimilar Joint by Friction Welding at Elevated Temperatures
Authors: Yamazaki Masayoshi , Watanabe Takashi , Hongo Hiromichi
Published: July 2009

In this study, the tensile and creep rupture behaviors of dissimilar (SA-387 Grade 91/309S or 310S/SUS-304) joint by friction welding at high temperatures were examined. The effects of temperature and applied stress on the failure locations of the dissimilar welded joints were also investigated. Creep rupture testing was conducted at 600°C and 650°C under stresses ranging from 40 to 160 MPa and 30 to 80 MPa, respectively. For all the temperatures at which the tests were conducted, the creep rupture strengths of the friction welded dissimilar metal joints were found to be lower than those of the 9Cr-1Mo-V-Nb steel base metal and dissimilar joints welded using Gas Tungsten Arc Welding (GTAW). A new type of fracture was observed for the friction welded dissimilar joints.
Engineering Approach for the Reactor Pressure Vessel Integrity in PTS Conditions: Thermal Hydraulic and Fracture Mechanics Studies of a Safety Injection in a Three Loop Power Plant — French Benchmark With EDF and AREVA-NP Tools
Authors: Martin A. , Dahl A. , Lestang F. , Bellet S. , Beaud F. , Gilles Ph. , Guichard D. , Dolleans Ph. , Cornille S. , Maas L.
Published: July 2009

For the Reactor Pressure Vessel (RPV) assessment and Lifetime evaluation of the nuclear plants, French Utility applies a series of calculations including thermal-hydraulic, thermo mechanical and fracture mechanics studies in order to study the Pressurized Thermal Shock (PTS) in the down comer caused by the safety injection. Within the frame of the plant life time project, integrity assessments of the French 900 MWe (3-loops) series reactor pressure vessel (RPV) have been performed. We found that the modeling of thermal-hydraulics loads is a source of gain. Considering the length of local 3D calculation and the large number of cases, E.D.F and AREVA-NP decided to share the effort. However the two chains of software differ: EDF uses Code_Saturne (coupled with the thermal solid code Syrthes) and Cuve 1D and AREVA-NP uses Star_CD and CALORI codes respectively for thermal hydraulic and thermo mechanical computations. According to this approach, comparison between the two chains of tools have been performed. Moreover this action contributes to the verification and the validation of each code in accordance with the OECD Best Practise Guidelines (BPG). The study has been achieved by two independent teams from EDF and AREVA-NP. It should be emphasized that this benchmark helped to strengthen the accuracy of CFD and the adapted methodology (working progress). The investigated configuration corresponds to the injection of cold water in the vessel during a penalizing representation of a primary break transient and its impact on the solid part formed by cladding and base metal. Numerical results are given in terms of fluid temperature and velocity fields in the cold legs and in the downcomer. The obtained numerical description of the transient (internal pressure and temperature field within the vessel) is used as boundary conditions for a full mechanical computation of the stresses. This thermal mechanical transient is obtained on a 3D mesh of the vessel, covering the two core shells and their circumferential welds, as well as the internal cladding. The results show that such a complete thermal–hydraulic and mechanic 3–dimensional analysis improves the evaluation of the consequences of the loading on the stress fields and eventually the margins to fast fracture of the RPV. The good agreement observed between EDF and AREVA-NP results and their accordance with the validation computations, confirm the robustness of the approach.
Comparison of Fatigue Damage Criteria Applied to Multiaxial Fatigue
Authors: Curtit F. , Le Pecheur A. , Stephan J. M.
Published: July 2009

The mixing tees of PWR nuclear power plant are submitted to complex cyclic thermal loadings that could lead to significant fatigue damage as observed on the RHR mixing area of Civaux 1 PWR in 1998. The characteristics of associated mechanical loading have been investigated by both experimental and numerical studies. A constant loading due to mean temperature and pressure is combined with equi-biaxial variable amplitude loading in the field of high cycle fatigue. This paper compares several fatigue damage criteria applied to both low and high cycle fatigue tests on 304 stainless steel specimens under several loading conditions according to mixing zone thermal fatigue loading amplitude. In order to focus the comparison on intrinsic qualities of the models, each criterion is evaluated directly with stress and strain measured on the specimen. The relevance of each criterion for high cycle thermal fatigue is discussed taking into considerations the precision of the results, and also the “usage cost” (including identification and complexity) and the adaptability to cumulative damage rules.
Validation Test Cases for Multi-Physic Problems: Application to Magneto-Hydrodynamic Numerical Simulations
Authors: Cebron David , Sigrist Jean-Franois , Soyer Vincent , Ferrant Pierre
Published: July 2009

The present paper is concerned with the numerical simulation of magneto-hydrodynamic (MHD) problems with industrial tools. MHD has received attention some thirty to twenty years ago as a possible alternative in propulsion applications; MHD propelled ships have even been designed to that purpose. However such propulsion systems have been proved of low efficiency and fundamental researches in the area have progressively received much less attention over the past decades. Numerical simulation of MHD problem could however provide interesting solutions in the field of turbulent flow control. The development of recent efficient numerical techniques for multi-physic applications provide promising tool for the engineer for that purpose. In the present paper, some elementary test cases in laminar flow with magnetic forcing terms are analyzed; equations of the coupled problem are exposed and analytical solutions are derived in each case, highlighting the relevant non-dimensional number which drives the physics of the problem. Several analytical calculations are then proposed and discussed. The present work will serve as basis for validation of numerical tools (based on the finite element method) for academic as well as industrial application purposes.
A Fatigue Crack Driving Force Parameter
Authors: Xiong Ying , Gao Zengliang , Katsuta Junichi , Sakiyama Takeshi
Published: July 2009

Most of the previous parameters that utilized as a crack driving force were established in modifying the parameter Kop in Elber’s effective SIF range (ΔKeff = Kmax–Kop). This paper focuses on the physical meaning of compliance changes caused by plastic deformation at the crack tip, the test was carried out for structural steel under constant amplitude loading, and differences of several parameter ΔKeff in literature are analyzed quantificationally. The effect of actual stress amplitude at the crack tip on fatigue crack growth is investigated, and improved two-parameter driving force model ΔKdrive(=Kmax)n(ΔK^)1−n) has been proposed. Experimental data for several different types of materials taken from literature were used in the analyses. Presented results indicate that the parameter ΔKdrive is equally effective or better than ΔK(=Kmax-Kmin), ΔKeff(=Kmax-Kop) and ΔK*(=(Kmax)α(ΔK+)1−α) in correlating and predicting the R-ratio effects on fatigue crack growth rate.
Effect of Yield Strength to Elastic Modulus Ratio on Finite Element Based Plastic Loads for Elbows Under Bending
Authors: Lee Kuk-Hee , Kim Yun-Jae
Published: July 2009

This paper quantifies the effect of the yield strength-to-elastic modulus ratio (yield strain) on plastic loads (defined by the twice-elastic-slope according to the ASME code) for 90° elbows under in-plane and out-of-plane bending. Results are based on extensive and systematic FE limit analyses assuming elastic-perfectly plastic materials. Based on FE results, a simple approximation of plastic loads of pipe bends, incorporating the yield strength-to-elastic modulus ratio effect, is proposed. To validate the proposed approximation, predicted plastic moments are compared with published full-scale pipe test data, showing that the proposed approximation gives overall lower than the FE results and close to experimental values.
Development of a Dissimilar Temper Bead Welding Procedure for an Amine Tower Repair
Authors: Boring Matt A. , Yang Yu-Ping
Published: July 2009

An amine tower was inspected and was shown to have wall loss over the majority of the circumference. A repair plan was developed which included welding onto the tower while the tower remained in operation. The repair plan consisted of two steps the first of which was welding two Inconel 625 rings to the SA516-70N steel tower. Secondly, a stainless steel sleeve was welded to the Inconel 625 rings to cover the corroded area to contain the tower contents if the tower wall was breached and to add structural support until the next shut down. The typical requirements for welds that could possibly be exposed to amine service are outlined in the API Recommended Practice 945 document (API RP 945). API RP 945 recommends a post-weld heat treatment (PWHT) to reduce hardness and relieve stress but since the planned repair was to be made in-service, PWHT was not preferred. The objective was to develop a welding procedure to attach the Inconel 625 straps to the SA516-70N steel amine tower without using PWHT. A temper bead welding technique was used to successfully qualify a welding procedure with hardness values below 200 Brinell without PWHT, in accordance with 2006 NBIC and 2004 ASME Boiler and Pressure Vessel Code Section IX. The temper bead welding procedure required strict welding heat input control and weld bead placement. The heat input can be monitored by controlling the welding parameters or by using the run-out ratio diagram which was developed during this project. The temper bead passes need to be deposited in such a manner that the weld toe of the temper bead is within 3/32 in. (2.4 mm) of the weld toe of the initial layer. When manually welding Inconel 625 to SA516-70N with SMAW electrodes it is recommended that Inconel 182 electrodes be used rather than Inconel 112 electrodes. This recommendation is based on the increase ability of the Inconel 182 electrodes to form manganese oxides, which remove the oxygen from the weld metal. This in turn helps to avoid local brittle zones which caused the Inconel 112 weld to fail qualification.
Efficient Analytical Interatomic Potential for Metal
Authors: Park Young Ho , Hijazi Iyad
Published: July 2009

A simple empirical embedded-atom potential that includes a long range force is developed for fcc metals. The potential parameters of this model are determined by fitting lattice constant, three elastic constants, cohesive energy, and vacancy formation energy using an optimization technique. Parameters for Cu, Ag, Au, Al, Ni, Pd, Pt have been obtained. The obtained parameters are used to calculate bulk modulus, divacancy formation energy, and melting point. The predicted values are in good agreement with experimental results. We also find that the predicted total energy as a function of lattice parameter is in good agreement with the equation of state of Rose et al.
Analysis of Data on Pipe Wall Thinning Phenomena by Fluid Flow in Pressurized Water Reactor Plant
Authors: Yamakami K. , Hirano S. , Nakamura T. , Kaino E.
Published: July 2009

In August 2004, a pipe rupture accident occurred at The Kansai Electric Power Company Mihama Unit3. Japan Society of Mechanical Engineers (JSME) recognized that it was an urgent problem to settle the rules on Pipe Wall Thinning Management by Fluid Flow. JSME enacted “rules on Pipe Wall Thinning” that stipulates the basic requirements of pipe wall thinning management, considering from these situations. In 2006, JSME enacted the “rules on Pipe Wall Thinning Management for PWR Power Plants” (JSME S NG1-2006, JSME rules for PWR) that satisfies the requirements of “rules on Pipe Wall Thinning”. This JSME rules for PWR stipulates the process of inspection plan, measurement, evaluation and maintenance. This paper focuses on the technical grounds that specify the management systems, pipe components and initial FAC rate.
Application of the BEREMIN Model for Evidencing the Warm Pre-Stress Effect in an Irradiated Reactor Pressure Vessel Containing a Semi-Elliptical Subclad Defect Under Small and Large LOCA Conditions
Authors: Gilles Ph. , Izard J.-P. , Devaux J.
Published: July 2009

The nuclear power plants lifetime is strongly dependent of the guarantee of the reactor pressure vessel (RPV) integrity. Therefore, the RPV integrity has to be demonstrated under the most severe configuration, namely the Pressurized Thermal Shock induced by the Loss of Coolant Accident induced by a large break in the primary loop. For such a transient, the apparent risk of failure is maximum when the load is decreasing; the fracture resistance decreasing rate being stronger. However, such type of loading generates an increase of the fracture resistance as shown by numerous studies (Chell, 1980 – BEREMIN, 1981 – Smith et al., 2004). This is known as the warm pre-stress (WPS) effect. This beneficial effect on the resistance to brittle fracture is not accounted for in the French RCCM and RSEM codes (RCCM, 2000 – RSEM, 2005). EDF has launched several R&D actions with CEA and AREVA as well as with European partners (SMILE, 2001) to validate and model the WPS effect under RPV representative conditions. Proving the existence of this beneficial load history effect (designated as Warm Pre Stress WPS), in the case of a defective RPV in emergency and faulted conditions is the aim of the present paper. The demonstration is conducted in the case of cleavage fracture using an improved version of the BEREMIN model. As opposite to the classical Fracture Mechanics methodology, this approach allows to account for load history effects on cleavage. The study analyzes the behavior of a semi-elliptical under clad crack in the EoL core shell of a 900 MWe RPV for two loading cases: the large break Loss Of Coolant Accident transient and a small break LOCA inducing thermal fluctuations on the vessel inner wall. The WPS effect is evidenced by comparing the plasticity corrected SIF levels of two loadings for the same value of failure probability: the considered WPS loading and a virtual monotonously increasing load applied at the temperature at which the brittle fracture risk is estimated.
Calculation of B2′ Stress Indices for Elbows Using Finite Element Analysis
Authors: Venkataramana K. , Bhasin V. , Vaze K. K. , Ghosh A. K. , Kushwaha H. S.
Published: July 2009

Nuclear power plant components are designed to withstand reversed dynamic loading like earthquake loading. Such reversed dynamic loads may induce plastic deformation in nuclear power plant components like pipe elbows. Plastic deformation in nuclear power plant components is limited by equation (9) of ASME Boiler &Pressure Vessel Code, Section III, NB-3652. ASME B&PV Code was revised in the year 2000 to accommodate plastic ratcheting as a mode of failure instead of plastic collapse under reversed dynamic load. The modified Code contains B2′ index, which is given as 2/3 rd of B2 index for butt-welded elbows. In the earlier work [1] B2′ indices were determined for several elbows using quasi-static nonlinear finite element analysis. In the present work an attempt is made to determine the ratio B2/B2′ for elbows using plastic nonlinear dynamic finite element analysis. Elbows of different sizes were considered in the present study. For each elbow linear static, linear dynamic, plastic nonlinear static and plastic nonlinear transient dynamic analyses are carried out to determine B2′ index in terms of B2 index. Elastic-perfectly plastic material model is used for the elbows. Collapse loads are obtained under static and dynamic conditions. Load vs. deflection curves are obtained for elbows under linear static and nonlinear quasi-static analyses. Deflection vs. time-curves are obtained from linear dynamic and plastic nonlinear dynamic analyses. The ratio B2/B2′ is computed for elbows of different sizes. The computed stress indices are compared with the Code values.
The COMSY: Code for the Detecting of Piping Degradation Due to Flow-Accelerated Corrosion
Authors: Zander Andre´ , Nopper Helmut
Published: July 2009

For the aging and plant life management the integrity of the mechanical component and structure is one of the key objectives. In most cases, only a small percentage of mechanical components are subject to significant degradation, e.g. flow-accelerated corrosion which may affect this integrity or the function of the component. Flow-accelerated corrosion (FAC) is a degradation process resulting in wall thinning of piping, vessels, heat exchanger and further equipment made of carbon and low alloy steel. The FAC degradation mechanism occurs only locally under specific condition of flow, water chemistry, temperature and materials applied. The establishment and implementation of a long term strategy is essential to the success of a plant’s FAC program. Effective measures to reduce FAC effects are e.g.: • Optimizing the water chemical treatment (e.g. high-AVT); • Apply improved material concepts for replaced components or lines and/or sufficient wall thickness margins; • Apply qualified repair technologies (e.g. METCO spraying, sheet metal cladding). In case of changes in system operation conditions (e.g. power uprate) the effect on FAC degradation rates should be analyzed before implementation and appropriate precautions should be initiated. For this purpose AREVA NP GmbH has developed the computer code COMSY, which utilities more than 30 years of experience resulting from operational experience and research activities. The COMSY code provides the capability to establish a program guided technical documentation by utilizing a virtual plant model which includes information regarding thermal hydraulic operation, water chemical conditions and materials applied for mechanical components. It provides the option to perform a plant-wide screening for identifying system areas, which are sensitive for degradation mechanisms like FAC. If a system area is identified as being susceptible to degradation, a detailed analysis function enables the condition-oriented service life evaluation of vessels and piping systems in order to localize and conservatively quantify the effect of degradation. Based on these forecasts with COMSY, specific strategies can be developed to mitigate the effect of degradation and inspection activities can be focused on degradation sensitive areas.
Residual Stress Evaluation of Dissimilar Weld Joint Using Reactor Vessel Outlet Nozzle Mock-Up Model (Report-1)
Authors: Muroya Itaru , Iwamoto Youichi , Ogawa Naoki , Hojo Kiminobu , Ogawa Kazuo
Published: July 2009

In recent years, the occurrence of primary water stress corrosion cracking (PWSCC) in Alloy 600 weld regions of PWR plants has increased. In order to evaluate the crack propagation of PWSCC, it is required to estimate stress distribution including residual stress and operational stress through the wall thickness of the Alloy 600 weld region. In a national project in Japan for the purpose of establishing residual stress evaluation method, two test models were produced based on a reactor vessel outlet nozzle of Japanese PWR plants. One (Test model A) was produced using the same welding process applied in Japanese PWR plants in order to measure residual stress distribution of the Alloy 132 weld region. The other (Test model B) was produced using the same fabrication process in Japanese PWR plants in order to measure stress distribution change of the Alloy 132 weld region during fabrication process such as a hydrostatic test, welding a main coolant pipe to the stainless steel safe end. For Test model A, residual stress distribution was obtained using FE analysis, and was compared with the measured stress distribution. By comparing results, it was confirmed that the FE analysis result was in good agreement with the measurement result. For mock up test model B, the stress distribution of selected fabrication processes were measured using the Deep Hole Drilling (DHD) method. From these measurement results, it was found that the stress distribution in thickness direction at the center of the Alloy 132 weld line was changed largely during welding process of the safe end to the main coolant pipe.
Development of Probabilistic Fracture Mechanics Code for Primary System Pipes and Application for RI-ISI
Authors: Jang Changheui , Datta Debashis , Yang Jun-Seog
Published: July 2009

Recently, the use of risk information in regulatory decision as well as operation and maintenance of nuclear power plant has been gradually increased. Application of risk information to optimize the inservice inspection of nuclear piping system has been one of such an attempt in Korea. An accurate evaluation of the risk, or failure probability for each pipe subcomponent is essential to the application of the risk-informed inservice inspection. For this purpose, a probabilistic fracture mechanics code based on Monte Carlo simulation techniques was developed to handle various flaw shapes and orientations in nuclear piping system. Then, the failure probabilities of various subcomponents of the main coolant loop in a pressurized water reactor were calculated, and the relative risk ranking was determined using the plant’s data. Also, the effects of in-service inspection and frequency on the risk were investigated. Finally, the application of the probabilistic fracture mechanics methods to the risk-informed inservice inspection of nuclear piping system was discussed.
Evaluation of Simplified Methods for Estimating Shear Capacity Using JNES/NUPEC Low-Rise Concrete Shear Wall Cyclic Test Data
Authors: Nie Jinsuo , Braverman Joseph I. , Hofmayer Charles H. , Ali Syed A.
Published: July 2009

The simplified methods in current codes for determining the shear capacity of reinforced concrete shear walls had mostly been validated using the test results of single-element shear walls. Recently available JNES/NUPEC test data of reinforced concrete shear walls under multi-directional cyclic loadings provided a unique opportunity to investigate the adequacy of the simplified methods for use in situations with strong interaction effects. A total of 11 test specimens with aspect ratios between 0.47 and 0.87 have been used in the assessment. Two simplified methods from the ACI 349-01 standard [1] and one from the ASCE 43-05 standard [2] have been evaluated. This paper also presents the development of an adjustment factor to consider the aspect ratio and the development of two approaches to consider interaction effects for one of the simplified methods. It concludes with the insights on the applicability of the code methods when interaction effects exist.
Earthquake Resistant Design Code and Validation of Local Metal Loss Procedure Based on the Experimental Data Collected in Japan
Authors: Ohno Atsushi , Tahara Takayasu
Published: July 2009

Fitness-For-Service (FFS) assessments are performed to evaluate the components damaged in service to determine whether it is possible to continue their use. FFS assessment codes were recently standardized, and they are being used in many companies in Europe and the United States. In Japan, the regulation permits the use of FFS codes in nuclear power stations, but not yet in petroleum and petrochemical industries. The PAJ/JPCA FFS task group that consists of the members of petroleum or petrochemical companies has been studying and investigating one of the FFS Codes, API579-1/ASME FFS-1 [1], in an attempt to include it in the high pressure gas safety law [2], which regulates the pressure equipment operating at pressures greater than 1 MPa. We have now completed the adaptation of the FFS code for Japan, and it is in the process of being assessed by the authorities. It is required that the code is modified slightly because Japanese authorities and people are particularly nervous to matters regarding earthquake safety. This paper focuses on cylindrical equipment regulated by the high-pressure gas safety law. The margin for earthquake load of the actual equipment is shown, and the local metal loss assessment procedure according to API579-1/ASME FFS-1 is verified by using experimental burst test data with pressure and/or bending stress in order to determine whether or not the FFS code provides a sufficient safety margin for safe operations in Japan.
Experimental Evaluation of the Markl Fatigue Methods and ASME Piping Stress Intensification Factors
Authors: Hinnant Chris , Paulin Tony
Published: July 2009

Experimental results of ten girth weld and two unreinforced piping intersections recently tested at Paulin Research Group (PRG) are reported. Results are compared to the original Markl work [1–5] and to current ASME piping design code practices [6]. In addition, approximately 800 experimental tests from the literature [7–32] are compared to the new test results and the existing Mark1 equations. Findings indicate that the original mean girth weld equation provided by Mark1, which serves as the basis of the ASME piping codes, does not match trends expected for as-welded fatigue data. Discrepancies are apparent in the low and high cycle regimes. Finally, the new PRG test data and the existing data from literature are used to develop a new mean girth butt weld curve and to modify the existing ASME piping design codes.