bluuebunny/crossref_metadata_2025 huggingface dataset:
Benchmark of an Application of Acoustic FEA for Light Water Reactor Loss of Coolant Accident Acoustic Loads
Authors: Sommerville Daniel V.
Published: January 2011

A Recirculation Line Break Loss of Coolant Accident is a design basis event which must be considered for stress analyses of Boiling Water Reactor internal components such as Jet Pumps and Core Shrouds. This event causes acoustic and fluid drag loads on BWR internals. These loads must also be considered for fracture mechanics evaluations performed to assess allowable operating periods for flaws detected during inservice inspections. Acoustic loads methods generally utilized in the past have been 1-D or simplified 2-D models of the domain of interest. In a few cases sophisticated thermal-hydraulic codes are used to obtain the acoustic response to the LOCA event. The present paper describes the results of a benchmark study performed to validate use of acoustic finite elements available in many commercial general purpose finite element analysis software packages. Use of FEA to predict acoustic loading in the BWR is benchmarked against blow down testing performed by Pacific Northwest Laboratory on a simulated light water reactor vessel. The results of the benchmark demonstrate that use of acoustic FEA yields conservative results and can be considered a viable method for BWR LOCA acoustic load predictions.
Formulating Test Specimen Thickness Effect on Fracture Toughness With T33-Stress: Case of 3PB Test Specimen
Authors: Tanaka Tomohiro , Meshii Toshiyuki
Published: January 2011

This paper considered the test specimen thickness effect on the fracture toughness of a material, in the transition temperature region, for 3PB specimen. Then, the thickness effect on the T33-stress, which is the out-of-plane elastic crack tip constraint parameter, was studied. Finally, an experimental expression on the thickness effect on the fracture toughness by using T33-stress was proposed for 0.55% carbon steel S55C. The results seemed to indicate a possibility of improving the existing methods to correlate the fracture toughness obtained by test specimen with the toughness of actual cracks found in the structure, in use of T33-stress.
Heavy Components Replacement in Nuclear Power Plants: Guidelines and Experiences
Authors: Kang Ki Sig
Published: January 2011

Utilities are looking for ways to optimize plant lifetime, and must therefore prevent stress corrosion in primary components, while combating other phenomena, such as thermal fatigue or certain metallurgical weaknesses. The replacement of sections of the main primary system is one way of solving these problems. The increase in the number of the replacement of heavy components carried out in the reactor building on specific reactor geometries has called for major technical innovations on the replacement of heavy components. For above, the IAEA published a nuclear energy series (NES) on replacement of heavy components to propose guidance and share experiences. The major and heavy components to be considered are; 1) Steam generators for pressurized water reactor plants, 2) Reactor vessel head for PWR plants, 3) Reactor internal components for boiling water reactor plants, 4) Reactor vessel internals for PWR plants, 5) Pressurizer for PWR plants, 6) Reactor coolant piping/ recirculation piping PWR, and 7) Press Tube and feed piping for pressurized heavy water reactor. This paper is focused on heavy components replacement considered strategic aspects for nuclear power plants life management.
Simultaneous Optimal Design of the Lyapunov-Based Semi-Active Control and the Semi-Active Vibration Control Device: Inverse Lyapunov Approach
Authors: Hiramoto Kazuhiko , Matsuoka Taichi , Fukukita Akira , Sunakoda Katsuaki
Published: January 2011

We address a simultaneous optimal design problem of a semi-active control law and design parameters in a vibration control device for civil structures. The Vibration Control Device (VCD) that is being developed by authors is used as the semi-active control device in the present paper. The VCD is composed of a mechanism of a ball screw with a flywheel for the inertial resistance force and an electric motor with an electric circuit for the damping resistance force. A new bang-bang type semi-active control law referred to as Inverse Lyapunov Approach is proposed as the semi-active control law. In the Inverse Lyapunov Approach the Lyapunov function is searched so that performance measures in structural vibration control are optimized in the premise of the bang-bang type semi-active control based on the Lyapunov function. The design parameters to determine the Lyapunov function and the design parameters of the VCD are optimized for the good performance of the semi-active control system. The Genetic Algorithm is employed for the optimal design.
Investigation of Flexibility Factors and Stress Indices for Class 1 Curved Piping
Authors: Zobeiry Nima , Asadkarami Ali , Reinhardt Wolf
Published: January 2011

In the analysis of piping, consideration of the behaviour of bends and elbows is essential. Due to the ovalization phenomenon, the bends exhibit more flexibility under bending moment loading than straight pipes, while generating comparatively higher stresses. This has been recognized in the ASME Code for Class 1 piping, where flexibility factors and stress indices have been given for curved pipes. It has been understood that the values of stress indices given in the Code are conservative. As a result, extensive research has been performed to obtain lower values, as allowed by the Code. However, these studies have been performed separately from the flexibility factors, where research is generally lacking. The Code recognizes that the given flexibility factors have limited scope of use and more work on this issue has been proposed. It is evident that a more realistic assessment of pipe bends has to be based on revised stress indices and flexibility factors. In this paper, a number of compound bends are studied, where on one hand the stress indices are conservative and on the other the flexibility factors of the Code may not be applicable. For this purpose, the behaviour of these bends is simulated using 3D finite element models. Using these models and by comparison to the flexibility and stresses in straight pipes, flexibility factors and stress indices for the bends are obtained. A comparison to the values in the literature and discussion of different factors contributing to these values are presented.
Developing Response Functions for Crack Opening Displacement to Calculate Leak-Rates Through Complex Cracks in Pipes With Weld Overlays
Authors: Chakraborty Arindam , Momin Wasimreza , Miessi Angah , Jing Peihua , Qian Haiyang
Published: January 2011

Leak-Before-Break (LBB) is employed in design of nuclear power reactor piping to eliminate consideration of the dynamic effects of pipe rupture from the plant design basis for the affected piping system. LBB cannot be applied if environmental conditions that could lead to degradation by stress corrosion cracking exists. For Alloy 600/82/182 dissimilar metal welds (DMW) in pressurized water reactor plants, primary water stress corrosion cracking (PWSCC) is found to be active. Application of weld overlay (WOL) of non-susceptible Alloy 690/52/152 material has been shown to mitigate PWSCC growth in DMW. Therefore, LBB can be considered for a DMW with Alloy 690/52/152 overlay. However, WOL sizing design postulates a complex crack which is through wall in the overlay material and part through or full circumferential in the DMW base material. This significantly reduces the critical flaw size and in turn the maximum allowable flaw size for leak rate. The current industry practice conservatively ignores the full circumferential crack in the original pipe material and assumes a through wall crack along the entire pipe thickness. This assumptions leads to significantly reduced leakage due to smaller crack opening. The problem becomes more critical with small diameter pipes. The current work calculates the crack opening displacements (CODs) for a pipe with complex crack. Since it is a function of several geometry and materials parameters, response functions are generated to calculate CODs.
Method for Selecting Stress States for Use in an NB-3200 Fatigue Analysis
Authors: Meikle V Thomas L. , Cranford E. Lyles , Gray Mark A.
Published: January 2011

In ASME Code Section III NB-3222.4 fatigue evaluations, selecting stress states to determine the stress cycles according to Section NB-3216.2, Varying Principal Stress Direction, can become a challenging and complex task if the transient stress conditions are the result of multiple independent time varying stressors. This paper will describe an automated method that identifies the relative minimum and maximum stress states in a component’s transient stress time history and fulfills the criteria of NB-3216.2 and NB-3222.4. Utilization of the method described ensures that all meaningful stress states are identified in each transient’s stress time history. The method is very effective in identifying the maximum total stress range that can occur between any real or postulated transient stress time histories. In addition, the method ensures that the maximum primary plus secondary stress range is also identified, even if it is out of phase with the total stress maxima and minima. The method includes a process to determine if a primary plus secondary stress relative minimum or maximum should be considered in addition to those stress states identified in the total stress time history. The method is suitable for use in design analysis applications as well as in on-line stress and fatigue monitoring.
Steam Generator Replacement at San Onofre Nuclear Generating Station Unit 2
Authors: Chan James K. , Calhoun David J.
Published: January 2011

This paper describes the design and installation of the replacement steam generators at San Onofre Nuclear Generating Station (SONGS) Unit 2. The design improvements of the replacement steam generators are compared to the old steam generators. The difficulties encountered during the installation of the replacement steam generators are also described. Lesson learned are summarized to benefit the execution of the SONGS Unit 3 replacement steam generators installation.
Prediction of SCC Initiation in Weld Components by Multi-Scale Analysis Incorporating Crystal Plasticity
Authors: Mochizuki Masahito , Mikami Yoshiki
Published: January 2011

A multi-scale analysis method of microscopic stress is proposed to predict the occurrence of stress corrosion cracking (SCC) in the welded components in power plants. The method includes a macroscopic model and microscopic models. Calculation of the stress was first performed in the macroscopic model. Subsequent to this calculation, simulation of the microscopic model was conducted to evaluate the microscopic stress on the scale of the grains and microstructure. Then, the nodal temperatures and nodal displacements were transferred from the macroscopic model to the microscopic model as boundary conditions. The proposed multi-scale analysis was used to evaluate the weld residual stress of a bead-on-plate weld model to demonstrate the validity of the method. Good agreement was obtained between the macroscopic and microscopic models in nodal temperature, nodal displacement, and in the residual stress distribution. Following the bead-on-plate model, the multi-scale analysis method was applied to the model of an SCC test specimen of type 600 Nickel-based alloy. Crystal plasticity and inhomogeneous grain shapes were introduced into the microscopic model to consider the effect of crystal orientation. The crystal orientation was measured by electron backscattering pattern (EBSP) technique and applied to the microscopic model. The stress concentration at the grain boundaries was shown by the multi-scale analysis. In the simulated SCC tests, cracks were observed in the grain boundaries. The locations where microscopic stress concentrations occurred in the multi-scale analysis were in good agreement with the locations of cracks observed in the SCC test. The proposed multi-scale analysis method of microscopic stress distribution is thus applicable to the prediction of the locations of stress corrosion cracks.
Mechanism Study on IGSCC by Analyzing Residual Stress and Work Hardening in Welds of Low-Carbon Austenitic Stainless Steel With Surface Machining
Authors: Mochizuki Masahito , Mori Hiroaki , Katsuyama Jinya
Published: January 2011

To make clear the effects of residual stress and work hardening on intergranular stress corrosion cracking behavior in the welds of the austenitic stainless steel with surface hardening, residual stress and hardness in butt joints of pipes are evaluated by 3-dimentional thermal elastic-plastic analysis and the grain boundary sliding behavior is examined using constant strain rate tensile test. Based on these results, the mechanism of IGSCC has been clarified by the integrated knowledge between metallurgy and mechanics. The relationship between plastic strain and hardness in hard-machined surface near welds is clarified from the experimented relationship and the thermal elastic-plastic analysis. The distributions of hardness and residual stress with the actual surface machining could be evaluated. It was made clear that grain boundary sliding occurred in the steel at 561K by a constant strain rate tensile test with the numerically simulated tensile residual stress due to welding and surface machining. From the comparison of grain boundary sliding behavior between solution treated specimen and cold-rolled one, the grain boundary sliding in cold-rolled one occurs in smaller strain conditions than that in solution treated one, and the amount of grain boundary sliding in cold-rolled one increases remarkably with increase in added strain. In addition, it is clarified that the grain boundary energy is raised by the grain boundary sliding. On the basis of these results, it is concluded that the cause of intergranular stress corrosion cracking in the welds of Type 316L austenitic stainless steel with surface hardening is the increase in grain boundary energy due to grain boundary sliding accelerated by residual stress of multi pass welding and surface hardening.
Coupling Computation Between Weld Mechanics and Arc Plasma Processes
Authors: Mochizuki Masahito , Okano Shigetaka
Published: January 2011

A coupling model between arc plasma process and weld mechanics has been developed for the more accurate numerical analysis of weld distortion. Computational simulation of tungsten inert gas arc plasma based on mathematical modeling of the heat transfer from arc plasma to a welded plate is performed to obtain a more precise temperature distribution during welding. The temperature distribution obtained is used for a large-deformation thermal elastic-plastic stress analysis of weld distortion behavior. In addition, the effects of welding process conditions on weld distortion are examined considering molten pool configurations with the arc plasma process, and verified experimentally. Finally, the effectiveness of the weld distortion analysis by combining the arc plasma process with weld mechanics is evaluated.
Study on New Damping Material for Response Control Using Urethane Rubber
Authors: Furuya Osamu , Goda Kengo , Ishihana Kenta
Published: January 2011

The detached houses designed by the old structural design index in Japan were sustained enormous damage from Hyogo-ken Nanbu earthquake, and are over 47% in total detached houses in Japan even now. The upgrading for seismic safety such detached houses is urgent duties because a possibility of occurrence of huge earthquake is expected in a high probability. This study tries to apply urethane rubber material not only to seismic safety element but also to energy absorbing element for structural response control. A linearity of basic mechanical characteristics of urethane rubber is generally much stronger than viscoelastic material such as a high damping rubber. Moreover, it is possible to arrange a proper mechanical characteristics that is satisfied the design demand to fit each application cases. As a first step on the study, the mechanical characteristics were confirmed by several loading tests, and it is confirmed that the test specimen using urethane rubber indicates a high linearity in the mechanical characteristics and a high damping performance in condition with the test parameters of shaking frequency and amplitude.
Effect of Strain Rate on the Deformation-Induced Martensite Transformation and Mechanical Behavior of Austenitic Stainless Steels for Cold Stretched Pressure Vessels
Authors: Miao Cunjian , Li Yaxian , Zheng Jinyang
Published: January 2011

Cold stretching (pressure strenghtening) technique has been widely used in austenitic stainless steel pressure vessels in order to increase the proof strength and lighten the weight of the vessels. Cold stretching technique is performed by applying the strengthening pressure under a specified strain rate. Plastic deformation in the process will lead to martensite transformation and may influence material’s strength and ductility. Cold stretching tests of EN 1.4301 stainless steel are carried out at different quasi-static strain rates of 10−3 and 10−5/s, following the uniaxial tensile tests at the same strain rate of 2.5×10−3/s. The α′-martensite transformation is detected by magnetic measurement, meanwhile the work-hardening rate, the flow stress and the mechanical properties of material are studied and the results are presented.
Investigation of the Impact of Seed Record Selection on Structural Response
Authors: Houston Thomas W. , Mertz Greg E. , Costantino Michael C. , Costantino Carl J.
Published: January 2011

Time history records are typically used to define the seismic demand for critical structures for which soil structure interaction (SSI) analyses are often required. Criteria for the development of time histories is provided in ASCE 43-05. The time histories are based on a close fit of 5% damped target response spectra. Recent experience has demonstrated that for cases where the transfer functions associated with the structural response are narrow, the ASCE 43-05 criteria can under-predict peak spectral responses in the structure by as much as 70% in some frequency ranges. One potential solution for this issue is to reinstate requirements for matching target response spectra for multiple damping levels to ASCE 43-05 criteria. However, recent probabilistic seismic hazard analyses (PSHA) do not generally contain spectra for multiple damping levels. This paper proposes an approach to generate target spectra at multiple damping levels, given the 5% damped target spectrum provided by the PSHA, utilizing catalogs of recorded earthquakes. The process of fitting time histories to multiple damped spectra is effective in correcting defficiencies observed in the computed structural response when time histories meeting the ASCE 43-05 fitting criteria are used.
Effect of Radial Offset Between Welded Back to Back Elbows on Their C2 Stress Index
Authors: Balakrishnan R. , Reinhardt W.
Published: January 2011

Pipe bends and elbows in process and power plant piping systems are sometimes added to provide additional flexibility along with their primary function of changing the direction of fluid flow. In special cases, piping systems may have double bends or elbows welded back to back with or without straight pipe in between. Due to manufacturing inaccuracies, back-to-back welded elbows may have a radial offset at the welded joint. Assessment of Class 1 piping following NB-3600 requires consideration of the C2 stress index of the girth butt weld, which is to be multiplied with the elbow C2 index. As an alternative, the C2 stress index may be calculated by finite element analysis using the maximum linearized membrane plus bending stress intensity. The effect of offset between the elbows at the welded joint on the elbow C2 stress indices is studied in this paper. Double elbows with elbow angles ranging from 30° to 90° and a radial offset of 1/32″ and 3/32″ at the weld are considered. Double elbows model with no weld in between and finite element model with single elbow are analyzed for comparison. The effect of angle between the bend planes of the two back-to-back elbows is studied. Results from the analyses are compared with C2 stress indices calculated as per NB-3680.
Response of Base-Isolated Nuclear Structures for Design and Beyond-Design Basis Earthquake Shaking
Authors: Huang Yin-Nan , Whittaker Andrew S. , Kennedy Robert P. , Mayes Ronald L.
Published: January 2011

ASCE 43-05 presents two performance objectives for the design of nuclear structures: 1) 1% probability of unacceptable performance for 100% Design Basis Earthquake (DBE) shaking, and 2) 10% probability of unacceptable performance for 150% DBE shaking. To aid in the revision of the ASCE 4-98 procedures for the analysis and design of base-isolated nuclear power plants (NPPs) and meet the intent of ASCE 43-05, a series of nonlinear response-history analyses was performed to study the impact of the variability in both earthquake ground motion and mechanical properties of isolation systems on the seismic responses of base-isolated NPPs. Computations were performed for three representative sites (rock and soil sites in the Central and Eastern United States and a rock site in the Western United States), three types of isolators (lead rubber, Friction Pendulum™ and low-damping rubber bearings), and realistic mechanical properties for the isolators. Estimates were made of 1) the ratio of the 99%-ile (90%-ile) response of isolation systems computed using a distribution of spectral demands and distributions of isolator mechanical properties to the median response of isolation systems computed using best-estimate properties and 100% (150%) spectrum-compatible DBE ground motions; 2) the number of sets of three-component ground motions to be used for response-history analysis to develop a reliable estimate of the median response of isolation systems. Only the results for the rock site in the Central and Eastern United States are presented. The results of this study provide the technical basis for the revision of ASCE Standard 4-98.
Korean Regulatory Experience on PWSCC in Korean NPP
Authors: Kim Kyung-Cho , Choi Sung-bu , Chung Koo-Kab , Chung Hae-Dong
Published: January 2011

The degradation of alloy 600 and its weld material (alloy 82/182) has been reported in many nuclear power plants. In Korea, the crack induced by PWSCC was discovered in the drain nozzle of Yongkwang units 3 & 4 in 2006∼2008 and SG plug weld of Yongkwang unit 3 in 2007. In July 2007, during visual inspections of SG tube plugs at Yonggwang unit 4, boric acid deposits were observed around five Alloy 600 welded plugs. The root cause of the cracking in alloy 600 plugs was revealed to be due to the fact that the cracks were mainly caused by residual stress induced from the welding, expanding and tight-fitting. Younggwang unit 3 found the white small deposits on the drain nozzle on the 10th RFO in 2007. The root cause of the cracking in drain nozzle was revealed to be due to the initiation of a crack on the inside surface of drain nozzle and propagated to through wall cracks in the axial and circumferential direction. Younggwang unit 3 found the white widespread deposits on the upper head of a reactor vessel on the 12th RFO in 2010. Utility is trying to reveal the root cause of the cracking in the vent line of the reactor head according the KINS requirement. In this article, Korean regulatory experiences for PWSCC are introduced. After these PWSCC experiences, all SG tubes welded by Alloy 600 were replaced and all SG drain and instrumentation nozzles with Alloy 600 have been replaced into Alloy 690 material.
Sensitivity Analysis of Residual Stress Simulation of Dissimilar Metal Joint of Safe End Nozzle and Key Issues for Standard Procedure to Maintenance Rules
Authors: Hojo Kiminobu , Ogawa Kazuo , Ogawa Naoki , Muroya Itaru
Published: January 2011

For evaluation of PWSCC crack extension behavior detailed residual stress evaluation such a stress distribution on the evaluation points’ thickness is needed. In order to establish the residual stress evaluation procedure, Japan Nuclear Energy Safety Organization (JNES) had a seven years’ project from 2001 to 2007. In this project several sensitivity analyses were performed for understanding key points for the numerical simulation of residual stress. And also residual stress measurements were carried out for validation of the simulation methods. In this paper focused on the safe end nozzles, the results of the sensitivity analyses and the measurement result of the residual stress of the U-groove weld joint were introduced. Based on these result, the key issues of a guideline of a standard residual stress analysis procedure was proposed.
Comparison of Fatigue Crack Growth Curves of Japan, the US and EU Code and Standards
Authors: Hojo Kiminobu , Takahashi Yukio
Published: January 2011

There are several code and standards, handbooks and guidelines for the nuclear power plant maintenance in Japan, the US and EU. They include SCC and fatigue crack growth curve for crack growth calculation. In this paper the authors selected five kinds of codes, standards and guidelines and compared their fatigue crack growth curves for choice of the suitable curves. The feature of each curve was summarized quantitatively evaluated. JSME maintenance rule and ASME code provide the fatigue crack growth formulae for both ferritic and austenitic steels and consider the environmental effects in some cases. The FITNET curves are categorized in many kinds of metal whereas the FKM guideline and WES procedure provide the common properties applicable to steels generally.
Structural Integrity Assessment of the Pump Shaft Assembly of the APR1400 Reactor Coolant Pump
Authors: Kim Min-Chul , Bang Sang-Youn , Lee Ki-Won , Choi Young-Jin , Kim Yong-Su
Published: January 2011

The RCP (Reactor Coolant Pump) is operated in high speed and high pressure conditions. Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can drastically alter the critical speeds and stability characteristics, and can act as significant destabilizing forces. In this study, the structural integrity of the pump shaft of APR1400 RCP estimated under normal operation and accident conditions. In order to predict the vibration behavior and dynamic characteristics, modal and critical speed analysis were performed. For evaluation due to the load cases, stress characteristics were investigated. This paper shows that the 1st critical speed occurs at over rotational speed range and the calculation and analysis result of the pump shaft satisfied the code requirement.
Advanced Probabilistic Fracture Mechanics Using the R6 Procedure
Authors: Beardsmore David W. , Stone Karen , Teng Huaguo
Published: January 2011

Deterministic Fracture Mechanics (DFM) assessments of structural components (e.g. pressure vessels and piping used in the nuclear industry) containing defects can usually be carried out using the R6 procedure. The aim of such an assessment is to demonstrate that there are sufficient safety margins on the applied loads, defect size and fracture toughness for the safe continual operation of the component. To ensure a conservative assessment is made, a lower-bound fracture toughness, and upper-bound defect sizes and applied loads are used. In some cases, this approach will be too conservative and will provide insufficient safety margins. Probabilistic Fracture Mechanics (PFM) allow a way forward in such cases by allowing for the inherent scatter in material properties, defect size and applied loads explicitly. Basic Monte Carlo Methods (MCM) allow an estimate of the probability of failure to be calculated by carrying out a large number of fracture mechanics assessments, each using a random sample of the different random variables (loads, defect size, fracture toughness etc). The probability of failure is obtained by counting the proportion of simulations which lead to assessment points that lie outside the R6 failure assessment curve. This approach can give good results for probabilities greater than 10−5. However, for smaller probabilities, the calculation may be inefficient and a very large number of assessments may be necessary to obtain an accurate result, which may be prohibitive. Engineering Reliability Methods (ERM), such as the First Order Reliability method (FORM) and the Second Order Reliability Method (SORM), can be used to estimate the probability of failure in such cases, but these methods can be difficult to implement, do not always give the correct result, and are not always robust enough for general use. Advanced Monte Carlo Methods (AMCM) combine the two approaches to provide an accurate and efficient calculation of probability of failure in all cases. These methods aim to carry out Importance Sampling so that only assessment points that lie close to or outside the failure assessment curve are calculated. Two methods are described in this paper: (1) orthogonal sampling, and (2) spherical sampling. The power behind these methods is demonstrated by carrying out calculations of probability of failure for semi-elliptical, surface breaking, circumferential cracks in the inside of a pressure vessel. The results are compared with the results of Basic Monte Carlo and Engineering Reliability calculations. The calculations use the R6 assessment procedure.
Strain Rate Calculation Approach in Environmental Fatigue Evaluations
Authors: Gray Mark A. , Salac Matthew C. , Roarty David H. , Cranford E. Lyles
Published: January 2011

Fatigue usage factor evaluations including the effects of reactor water environment have been performed in numerous nuclear plant license renewal efforts. A large number of these evaluations have used the environmental fatigue penalty factor, Fen, approach prescribed in various regulatory documents. The Fen equations require input of strain rate, but the prescribing documents do not provide methodology or criteria for the quantification of the strain rate to be input. As a result, numerous approaches have been offered and studied. This paper presents an approach used by Westinghouse to include strain rate in an automated calculation of Fen based on the modified rate approach to integrated strain rate applications. The starting point of the approach is ASME Code Section III NB-3200 fatigue analysis. With environmental fatigue evaluations in new plant designs now emerging in ASME Code criteria, strain rate considerations remain part of the discussion. The intent of this paper is to provide further insight into this process.
Bending Test on an Aged Cast Duplex Stainless Steel Pipe Containing a Crack in the Heat Affected Zone: Results and Analysis
Authors: Le Delliou Patrick , Curtit Franc¸ois , Sonnefraud Christophe , Page`s Claude
Published: January 2011

Some components (elbows, pump casings and lateral connections) of the primary loop of French PWRs are made of static cast duplex stainless steels. This kind of steel may age even at relatively low temperatures (in the temperature range of PWR service conditions), depending on the material composition. An important consequence of this ageing process is the decrease in the ductility and fracture toughness of the material. It is feared that an embrittlement, associated with the occurrence of casting defects, may increase the risk of failure. In order to build the primary loops, these components are welded and the behaviour of the weld heat affected zone (HAZ) is not well known. So a specific program was launched on this topic, involving metallurgical studies, fracture mechanics tests, medium-scale experiments, and finite element analyses. This paper presents the main characteristics and results of an experiment conducted on a 6″ aged cast pipe. This pipe contained a machined notch in the heat affected zone of a butt-weld and was tested under four-point bending at 300°C. The chemical composition of the steel was chosen to obtain a fast thermal ageing and low fracture toughness properties. During the test, the defect initiated and grew subsequently by ductile tearing. The test showed that it was possible to obtain a significant amount of stable crack growth (about 4 mm) despite the low toughness properties of the aged material. A detailed fracture mechanics analysis, based on finite element calculations, was performed. These calculations fairly simulated the overall behaviour of the tested structure, gave a conservative prediction of the crack initiation pressure and well predicted the crack size associated with the maximum applied bending moment. These tests and their detailed analyses contribute to validate and justify the methodology used in the integrity assessment of in-service cast duplex stainless steel components.
A Method for the Coupled Thermal-Structural Analysis of Radioactive Material Shipping Packages in Hypothetical Accident Conditions: Part 1—NCT Thermal Load and Some HAC Mechanical Loads
Authors: Wu Tsu-te , Gupta Narendra K. , Smith Allen C. , Blanton Paul S.
Published: January 2011

The design of radioactive material (RAM) shipping packages are designed to withstand the mechanical and thermal loads specified in Title 10 Code of Federal Regulations Part 71 (10CFR71). The Hypothetical Accident Conditions (HAC) specified in 10CFR71 include a 30-foot free drop, a crush by an 1100-pound plate, a 40-inch free drop onto a round bar and a 30-minite fire. Furthermore, in accordance with the Nuclear Regulatory Commission Guide 7.8, these loads should be applied sequentially. The current common practice is to base the thermal modeling on the un-deformed (or un-damaged) packaging configuration and thermal stresses are neglected in structural analyses. This paper presents a methodology to simulate the coupled thermal and structural responses and to evaluate the accumulated damages caused by both sequentially applied thermal and mechanical loads of the Hypothetical Accident Conditions. Part 1 of the paper discusses the thermal analysis for the Normal Conditions of Transport (NCT) to establish the initial temperatures of the shipping package. It also discusses the subsequent analyses of closure-bolt tightening preload analysis and 30-foot drop analysis.
Implications of Cleavage Fracture Criteria for Structural Integrity Assessment of RPV Steels
Authors: Lewis S. J. , Truman C. E. , Smith D. J.
Published: January 2011

To accurately assess the safe operation of structures containing defects, it is necessary to consider the influences of previous load cycles on crack propagation. A number of current assessment codes contain advice to account for strain history and residual stress, but are generally known to be highly conservative which may potentially result in the unnecessary and expensive repair or replacement of infrastructure. This paper considers the results of previous investigations into cleavage fracture of an A533b RPV steel to determine the accuracy of the widely used R6 structural integrity assessment procedure for fracture following significant load history. The levels of conservatism associated with a number of assessment methods are discussed and compared with experimental data. The general trends suggested an improvement in assessment accuracy may be obtained by using local approach methods, compared to crack tip failure parameters. It is noted however that all the methods used produced some unsafe estimates of failure load, which is felt to be related to an over estimate of the characteristic material toughness.
Weld Repair for Pressure Vessels Made From Cr-Mo Steels
Authors: Kayano Rinzo , Yamamoto Eiichi , Tahara Takayasu
Published: January 2011

Pressure vessels made from Cr-Mo steels are utilized for high temperature and high pressure services including hot hydrogen services. After long term operation, there are several past experiences of damages and/or degradation of materials such as temper embrittlement, creep embrittlement, hydrogen attack and hydrogen embrittlement. This paper summarizes typical damages/degradation and examples of weld repairs including special attention to development of weld repair procedure. The subject equipments are heavy wall petroleum pressure vessels made from Cr-Mo steel with austenitic stainless steel overlay cladding. Cracking could be prevented by controlling the repair welding process to reduce the hydrogen content at the interface. After repair welding, adequate post weld heat treatment (PWHT) has to be executed. Recently, repair welding has become an important aspect as part of post construction codes for pressure equipment to keep safe and long term continuous operation of the process plants because many of the plants have been operated for more than thirty years in Japan. Responding to the needs of petroleum and chemical industries, The Chemical Plant Welding Research Committee (CPWRC) of The Japan Welding Engineering Society (JWES) established the Pressure Equipment Repair Welding Subcommittee (PERW S/C) [1]. The S/C has developed optimum repair welding methods and procedures in the guideline on November 2009, with reference to the above investigation results. This paper also introduces the repair welding guideline for the pressure vessels made from Cr-Mo steels.
Damage Accumulation and Lifetime Prediction of Carbon Fiber Composite Pressure Vessels
Authors: Bunsell Anthony R. , Barbier Franc¸oise , Thionnet Alain , Zejli Hasnae , Besanc¸on Brian
Published: January 2011

Efficient storage of hydrogen is critical to the success of transition markets for hydrogen energy. For these near-term niche markets, high pressure compressed gas storage in carbon fiber composite pressure vessels is currently the most advanced and promising technology. However, challenges remain to improve performance and reliability while still insuring the safety of these pressure vessels in service over periods of 15 to 20 years. In order to avoid overdesign and conservative use, a fundamental understanding of damage mechanisms and degradation of these materials is required to fully exploit the potential of these materials. Acoustic emission has thus been used to characterize damage accumulation and its kinetics under static and cyclic loading in carbon fiber composite samples and pressure vessels. These experimental results were complemented by finite element modeling taking into account the elastic nature of the fibers, coupled with the visco-elastic behavior of the matrix and debonding at the fiber matrix interface.
Managing Environmental Fatigue of the Real Components in a PWR
Authors: Rothenho¨fer Horst , Ko¨nig Gu¨nter
Published: January 2011

In this paper the relevance of environmental fatigue for real components of a PWR is discussed based on the ANL procedure according to NUREG/CR-6909 using the penalty factor Fen. Monitoring data of several German PWRs for the surge lines and spray lines have been evaluated and assessed for five years of operation each. Monitored loads are compared to specified loads and lab loads and the fatigue evaluation is assessed comparing traditional fatigue evaluation to the ANL approach. Taking into account environmental fatigue the calculated life time of a component in LWR environments can be reduced by a factor of 10 or even more compared to common fatigue evaluation. These tremendous reductions are based on lab results whereas experience from existent nuclear power plants has not yet shown any relevance of this influence. The conclusions of evaluations and assessments are summarized in a concept for long term operation (LTO). This concept contains procedures how to manage environmental fatigue. In the design phase conservatism in specified loads, especially numbers of load cycles, can be reduced based on the duty to monitor real loads and assess fatigue during operation. In the operating phase monitoring data should be used to record loads, evaluate and assess fatigue and subsequently reduce fatigue by optimizing operational modes.
Use of Geometrically-Accurate Models to Predict Spent Nuclear Fuel Cladding Temperatures Within a Truck Cask Under Normal and Fire Accident Conditions
Authors: Kamichetty Krishna Kumar , Greiner Miles , Venigalla Venkata V. R.
Published: January 2011

The temperature of spent nuclear fuel cladding within transport casks must be determined for both normal conditions of transport and hypothetical fire accident conditions to assure that it does not exceed certain limit conditions. In the current work a two-dimensional finite-element thermal model of a legal-weight truck cask is constructed that accurately models the geometry of the fuel rods and cover gas. Computational fluid dynamics (CFD) simulations are performed that include buoyancy induced motion in, and radiation and natural convection heat transfer across the cover gas, as well as conduction in all solid components. Separate simulations are performed using helium or nitrogen cover gas. Stagnant-gas CFD (SCFD) simulations are preformed and compared to CFD simulations to determine the effect of gas motion. For normal conditions of transport, the peak clad temperature is determined for a range of fuel heat generation rates to determine the thermal dissipation capacity based on peak cladding and surface temperature, QC and QS. These are respectively, the fuel heat generation rates that bring the peak cladding temperature to 400°C, or the peak surface temperature to 85°C (their allowed limits for normal transport). Transient fire/post fire simulations are then performed for a range of fire durations to determine the critical durations for cladding Creep Deformation or Burst Rupture, DCD or DBR. These are the fire durations that bring the cladding temperature to 570°C or 750°C, respectively. When the cladding temperature is used to select the fuel heat generation rate, the thermal dissipation capacity is 3265 W/assembly when helium is the cover gas, which is 30% higher when nitrogen is used (due to helium’s higher thermal conductivity). When nitrogen is the cover gas, the critical fire durations for creep deformation and burst rupture are, respectively, 3.3 and 7.2 hours. These durations are 18% and 14% shorter for helium (because the allowed fuel heat generation rate is higher for helium). When the fuel heat generation is chosen based on the package surface temperature, for helium, the thermal dissipation capacity is 1040 W/assembly, and the critical fire durations for creed deformation and burst rupture are, respectively, 4.7 and 11.6 hours. The values for nitrogen are all within 4% of these values. The CFD and SCFD simulations give essentially the same results. This indicates that gas motion does not significantly affect the cladding temperature, and the future calculations may not need to incur the increased computation expense required to model that motion.
Structure and Strength Analysis on End Closure Seal Using D-Section Shearing Bolts and “n”-Ring for Pressure Vessels
Authors: Chen Ping , Deng Yuting
Published: January 2011

This paper describes a new structural type of end closure with use of the metallic self-energized n-ring as its sealing gasket and the D-section shearing bolts as its load-carrying connection between the head and the top flange. The connection installs easily and quickly for pressure vessels. The strength analyses of connection component are presented.
Flamanville 3 EPR: How to Define the Pre-Service NDE Inspection Program
Authors: Champigny F. , Blin P. , Guilloteau J. L.
Published: January 2011

The last generation of PWR, the European Pressurizer Reactor, is being built both in Finland at Olkiluoto and in France at Flamanville where it will be the second unit in operation in two years. The conception of the reactor has no fundamental differences compared with the last French and German generations (i.e. N4 1450 MW and KONVOI 1300 MW). In fact the EPR is a synthesis of the best knowledge from both parts of Rhin river. Nevertheless, for what concerns the primary and secondary systems, few new issues have been implemented and they have important consequences for the pre-service inspection programme. First of it, main coolant lines and main components are declared break preclusion that means another approach for the in-service inspection and the requirements for pre-service inspection. A second one, is that the 1999 ministerial order will apply for the first time on a new unit. The inspection programme has to take it into account particularly in terms of NDE performance demonstration. In terms of structural integrity, the most important areas have been reviewed with fast fracture and fatigue criteria to determine the levels of NDE qualifications. This paper describes the important steps to reach the PSI and what is being developed in terms of NDE in relation to the structural integrity. Several examples are given to illustrate how EDF prepares the PSI of the EPR.
Mechanical Properties of SA508 Gr.4N Model Alloys as a High Strength RPV Steel
Authors: Kim Min-Chul , Lee Ki-Hyoung , Lee Bong-Sang , Kim Whung-Whoe
Published: January 2011

Demands of RPV materials with higher strength and toughness are rising to increase the power capacity and the operation life of nuclear power plants. The ASME SA508 Gr.4N specification can give a superior toughness and strength to the commercial low alloy steels such as SA508 Gr.3. However, the SA508-Gr.4N steels have not yet been used commercially due to a lack of information of the productivity and the age related properties. While the irradiation embrittlement studies are going-on, the current paper focused on the effects of alloying elements such as Ni, Cr and Mo on the fracture mechanical properties of the SA508 Gr.4N low alloy steels. Various model alloys were fabricated by changing the contents of alloying elements based on the composition range of the ASME specification. Tensile properties, Charpy impact toughness and fracture toughness of the model alloys were evaluated and those properties were discussed with the microstructural characteristics of each alloy. The strengths of the alloys were increased with increase of the Ni and Mo contents while there was no remarkable change of the yield strength with the Cr addition. The Charpy impact and fracture toughness were considerably improved with the increase of Ni, Cr contents. The Mo addition did not change the toughness properties significantly. The Cr contents were more effective on the fracture toughness through changing the carbides precipitation characteristics and the Ni contents were effective on the Charpy impact toughness through changing the effective grain size.