bluuebunny/crossref_metadata_2025 huggingface dataset:
Application of Small Punch Testing Methods for Thermal Ageing Monitoring at Primary Circuit Components in Nuclear Power Plant
Authors: Petzová Jana , Březina Martin , Kapusňák Michal , Kupča Ľudovít
Published: November 2015

This paper deals with the evaluation of material properties of safety-related components of the primary circuit of nuclear power plants (NPP) such as reactor pressure vessel (RPV), primary piping (PP) or steam generator (SG).

The main degradation mechanism of NPP’s components is radiation damage, but these processes are situated only in the core region of the reactor pressure vessel (RPV). The main mechanical loading of all individual parts of NPP’s primary circuit are the influence of high pressure at elevated temperature till 300°C. These processes lead to the fatigue damage of structural materials, which is characterized as the thermal fatigue or thermal ageing.

The reactors in Slovakia have been behind several decades in operation. Therefore, to ensure the safe and reliable operation of NPP, it is necessary to monitor and evaluate these changes.

While the monitoring of the radiation degradation is standardly performed using surveillance programs during all plant life operation, the monitoring of thermal ageing at primary circuit components was realized only after few years of all NPP units operation.

The concept of the thermal ageing assessment was divided into two possible approaches to evaluation. The first is long-term exposition of surveillance specimens. And the second approach is a direct surface sampling of heavy components after several years of operation.

The Small Punch Test (SPT) methods are mainly used for evaluation of materials actual state. By the SPT technique it is possible to evaluate the basic tensile properties as the ultimate tensile strength and the yield stress of the tested materials from a very small amount of obtained material.

The details of the original VUJE design program of thermal ageing monitoring of NPP primary circuit materials in Slovakia is described in this paper.
Applicability of the Master Curve Method to Ultra High Strength Steels
Authors: Wallin Kim , Pallaspuro Sakari , Karjalainen-Roikonen Päivi , Suikkanen Pasi
Published: November 2015

Although Ultra High Strength Steels (UHSS) with nominal strengths up to 1500 MPa have been available on the market for many years, the use of these steels in the civil engineering industry is still rather uncommon. One critical point limiting the use of UHSS steels lies in their rather poorly documented fracture properties in relation to more conventional steels covered by the codes. The major concept governing the assessment of steels is the Master Curve (MC) methodology. It provides a description for the fracture toughness scatter, size effect and temperature dependence in the ductile to brittle transition region. It enables a complete characterization of brittle fracture toughness of a material based on only a few small size specimens. The method combines a theoretical description of the scatter, a statistical size effect and an empirically found temperature dependence of fracture toughness. The fracture toughness in the brittle fracture regime is thus described with only one parameter, the transition temperature T0. At this temperature the mean fracture toughness for a 25.4 mm thick specimen is 100 MPa√m. The Master Curve method as defined in ASTM E1921-13a is applicable to ferritic structural steels with yield strength between 275 MPa and 825 MPa. Very few studies have been made with respect to the applicability of the Master Curve to Ultra High Strength Steels with yield strengths in the excess of 900 MPa. This is the topic of this work. Focusing on novel directly quenched high performance steels, the applicability of the Master Curve methodology with special emphasis on the temperature dependence will be investigated. Possible improvements to the Master Curve will be proposed for further consideration.
Investigation of Thermal Ageing in Long Term Operated RPV Materials
Authors: Eiselt Charles C. , König Günter , Hein Hieronymus , Selektor Maxim , Widera Martin
Published: November 2015

The phenomenon of thermal ageing of low alloy steels comes more into focus in terms of long term operation of nuclear power plants (NPP). Safety-relevant components such as the RPV or the pressurizer have to bear the respective loads at elevated temperatures for longer times. However the mechanical properties of the applied materials might experience certain degradations such as a decrease of the impact energy levels and a shift in the ductile to brittle transition temperature (e.g. T41) leading to higher ductile-brittle reference temperatures and a reduction of material toughness. In terms of a safe long term operation it is important to understand in how far thermal ageing alone, meaning for the RPV without the cumulative damaging effects through neutron irradiation, has detrimental influences on the respective materials of interest.

First of all an overview is provided of the current state of the art with respect to thermal ageing by describing influencing mechanisms, its implementation into different nuclear codes, standards and selected experimental investigations in this field. Following this, the test results of the thermal surveillance sets from three German PWRs are presented and discussed. The tested Charpy-V specimens, taken from representative RPV base and weld metals (22NiMoCr3-7 / NiCrMo1UP) as well as their heat affected zones, were exposed to ∼290°C for ∼30 years on the cold leg of the according plants’ main coolant loops. The obtained results are compared with the existing thermal aging data base (baseline and ∼7 years data) of the materials concerned. Finally, the role of thermal ageing particularly with respect to RPV irradiation surveillance will be assessed.
Uncertainty Characterization of the TWC_Fail Through-Wall Circumferential Crack Stability Module for xLPR
Authors: Olson Richard , Scott Paul
Published: November 2015

The US NRC/EPRI xLPR (eXtremely Low Probability of Rupture) probabilistic pipe fracture analysis program uses deterministic modules as the foundation for the calculation of the probability of pipe leak or rupture as a consequence of active degradation mechanisms, vibration or seismic loading. The circumferential through-wall crack stability module, TWC_Fail, evaluates through-wall circumferential crack stability based on the minimum crack size from the Net-Section Collapse or an EPFM J-estimation scheme analysis. Beyond the uncertainty of xLPR data inputs, each module has an uncertainty. This paper documents the module uncertainty for TWC_Fail.

Using 32 pipe fracture experiments, including: base metal, similar metal weld, and dissimilar metal weld experiments; bend only and pressure and bend loading; pipe diameters from 2-inch nominal diameter to 42-inch diameter, cracks that range from short to long, the uncertainty of the TWC_Fail methodology is characterized. Results show that TWC_Fail predictions are sensitive to the choice of J-R curve input (J-D or J-M from C(T) specimen tests) and the fit of the stress-strain data. Module uncertainty is characterized in terms mean fit and standard deviation between predictions and experimental values.
A New Methodology for Erosion Prediction Using Eulerian-Eulerian CFD Models
Authors: Messa Gianandrea V. , Ingrosso Irene , Malavasi Stefano
Published: November 2015

The erosion of a surface caused by the impact of solid particles dragged by a fluid is a serious concern in the oil&gas industry. At present, the erosion prediction is performed using algebraic erosion models which express the volume of eroded material per impact as a function of the mass of the abrasive particles as well as of fluid dynamic parameters (such as the impact velocity and the impact angle of the eroding particle) and properties of the materials involved in the process. The fluid dynamic parameters are, in turn, evaluated using Eulerian-Lagrangian CFD models which interpret the fluid phase as a continuous mean and follow the trajectories of all the particles. However, the huge computational burden makes it difficult, or even precludes, to adopt this approach in many flows of engineering interest. An innovative methodology is proposed for estimating the parameters required as input by the erosion models using computationally cheaper Eulerian-Eulerian CFD models which solve for the average properties of the ensamble of particles. The good results obtained when predicting the erosion caused the by impingement of an abrasive jet against a surface make the application of this methodology to more complex flows very promising.
Accounting for Negative R-Ratio (Crack Closure) in Fatigue Crack Growth Calculations on Stainless Steel Components
Authors: Watson Chris , Currie Chris , Emslie Julian
Published: November 2015

Negative R-ratio crack closure effects on Fatigue Crack Growth (FCG) are defined as the contribution of the compressive portion of the stress cycle to the crack extension, in addition to that contributed from the tensile portion of the cycle. Any potential decrease in FCG may be attributed to the mechanical effects of crack closure during the compressive part of the cycle. The overall effect is to decrease the crack opening portion of the stress range and to therefore reduce the crack growth rate compared to that obtained using the full stress range. This paper provides a brief overview of the treatment of negative R-ratio crack closure in FCG calculations on stainless steel components by reference to existing codes and standards. Then, using the results from crack closure tests on small cylindrical test specimens, a set of guidelines for the treatment of crack closure in the FCG assessment of stainless steel components are provided.
Crack Growth Behavior of Pipes Made From Polyvinyl Chloride Pipe Material
Authors: El-Bagory Tarek M. A. A. , Younan Maher Y. A.
Published: November 2015

The behavior of crack growth of polymeric materials is affected by several operating conditions such as, crosshead speed, specimen thickness, load line, and specimen configurations which reverse the behavior of crack from stable to unstable crack growth behavior. The main objective of the present paper is the determination of plane strain fracture toughness (KIC) for polyvinyl chloride (PVC) used in piping water transmission systems. The dimensions of the PVC pipe are outside diameter, Do=315 mm, standard dimensions ratio, SDR=13.23, ratio between outside to inside radii Ro/Ri =1.179 and pipe thickness, t =24 mm. Curved specimens are prepared from a pipe by cutting 12 mm thickness ring segments. The curved specimens are divided into two specimen configurations, namely curved three point bend, CTPB and C-shaped tension, CST specimens. All specimens are provided artificially with a pre-crack. CTPB specimen is further cut into five 72° sectors with each being centrally notched to a depth approximately a = 0.479 of the wall thickness. CST specimen configuration is characterized by the eccentricity X = 0, and 0.5W, of the loading holes from the bore surface. Linear elastic fracture mechanics theory (LEFM) is used to predict the plane strain fracture. The tests are carried out at room temperature, Ta equal 20 °C and different crosshead speeds of (10–500 mm/min). The fracture test results reveal that, the crosshead speed has been proven to affect the mode of failure and mode of fracture. At lower crosshead speeds the mode of failure is ductile, while at higher crosshead speeds it is brittle. The specimen configuration affects also on the fracture toughness. C-shaped tension specimens show higher fracture toughness in case of pin loading location X = 0.5W than X = 0 by about (12%). Transitional crosshead speed is affected by specimen geometry. C-shaped tension specimens (CST) at x= 0 and 0.5W have higher transitional crosshead speed compared with CTPB specimen configuration.
A Unified Viscoplastic Model for Creep and Fatigue-Creep Response Simulation of Haynes 230
Authors: Barrett Paul Ryan , Hassan Tasnim
Published: November 2015

A Chaboche-based unified viscoplastic constitutive model, including features of strain range dependence, strain rate-dependence, static recovery, and mean stress evolution is developed and evaluated for simulating fatigue-creep and creep responses of Haynes 230. In other words, this constitutive model attempt to simulate not only strain-controlled fatigue and fatigue-creep responses of Haynes 230, but also stress-controlled creep responses. After investigating various flow rules and kinematic hardening rules, a unified viscoplastic constitutive model is developed for simulating both the fatigue-creep and creep responses. The parameter determination for this constitutive model, however, requires a robust optimization algorithm. The proposed unified constitutive model can adequately simulate fatigue-creep responses, and creep responses up to the secondary creep regimes. However, with the introduction of damage modeling features the constitutive model can simulate the tertiary creep regime responses, but with some limitations in simulating fatigue-creep responses. Nonetheless, the unified viscoplastic constitutive model with or without damage modeling features has shown to be able to capture the stress-controlled creep responses while still maintaining high fidelity in capturing the strain-controlled fatigue and fatigue-creep responses.
Ultrasonic Evaluation of Inclusions and Surface-Breaking Defects in High Density Polyethylene (HDPE) Butt Fusion and Mitered Joints
Authors: Searfass Cliff , Milligan Jeffrey P. , Lashley Michael S.
Published: November 2015

This paper presents a method for investigating the integrity of high density polyethylene (HDPE) butt fusion joints and mitered joints. The scanning technique utilizes an ultrasonic phased array pitch-catch scanning method that provides full volumetric coverage of the fusion zone for the detection of inclusions and surface-breaking defects. Empirical results supported by numerical modeling are presented. Additionally, the effects of temperature on focalization and the proper countermeasures are discussed.
Effect of Underhead Design Configuration on Bolt Behavior Under Shear Loads
Authors: Wu Zhijun , Nassar Sayed A.
Published: November 2015

This study uses nonlinear 3-D finite element simulation to investigate the effect of bolt underhead design on its mechanical behavior under cyclic transverse shear loads. The effect of bolt underhead fillet radius and / or under-cut on bolt deformation as well as bolted joint behavior is studied in order to optimize the bolted flange performance under externally applied loads. Two levels of bolt preload and two transverse shear load cycles are studied. Bolt tensile and bending stiffnesses for various underhead designs are also evaluated and discussed.
Effect of Pressurized Water Reactor Environment on Material Parameters of 316 Stainless Steel: A Cyclic Plasticity Based Evolutionary Material Modeling Approach
Authors: Mohanty Subhasish , Soppet William K. , Majumdar Saurindranath , Natesan Krishnamurti
Published: November 2015

At present, the fatigue life of nuclear reactor components is estimated based on empirical approaches, such as stress/strain versus life (S∼N) curves and Coffin-Manson type empirical relations. In most cases, the S∼N curves are generated from uniaxial fatigue test data, which may not truly represent the multi-axial stress state at the component level. Also, the S∼N curves are based on the final life of the specimen, which may not accurately represent the mechanistic time-dependent evolution of material behavior. These discrepancies lead to large uncertainties in fatigue life estimations. We propose a modeling approach based on evolutionary cyclic plasticity that can be used for developing finite element models of nuclear reactor components subjected to multi-axial stress states. These models can be used for more accurately predicting the stress-strain evolution over time in reactor components and, in turn, fatigue life. The model parameters were estimated for 316 stainless steel material, which are widely used in U.S. nuclear reactors. The model parameters were estimated for different test conditions to understand their evolution over time and their sensitivity to particular test conditions, such as the pressurized water reactor coolant condition.
Analysis of Residual Stresses in Pipe Seam Welds and a Proposed Residual Stress Profile Estimation Method
Authors: Song Shaopin , Dong Pingsha
Published: November 2015

A recent comprehensive investigation into residual stress distributions in pipe and vessel longitudinal seam welds is presented in this paper, covering component wall thickness from 1/4” (6.35mm) to 10” (254mm), component radius to wall thickness ratio from 2 to 10, and linear welding heating input from low (50 J/mm) to high (6000 J/mm). Through the use of a residual stress decomposition technique, two key parameters that govern through-thickness residual stress distributions in terms of their membrane and bending content have been identified. One is component radius to wall thickness ratio (r/t) and the other is a characteristic heat input density (Q̂) having a unit of J/mm3. With these two parameters, a unified functional form for representing through-thickness residual stress profile behaviors in seam welds is proposed along with its solution procedure for applications in weld region. The simplicity of the proposed residual stress profile estimation scheme in functional form and demonstrated applicability for a wide range of r/t and Q̂ provides an effective framework for generating residual stress profile information for supporting defect assessment procedures in FFS codes and standards.
Correlations Between Charpy V-Notch Impact Energy and Fracture Toughness of Nuclear Reactor Pressure Vessel (RPV) Steels
Authors: Yu Meifang , Luo Zhen , Chao Y. J.
Published: November 2015

Both Charpy V-notch (CVN) impact energy and fracture toughness are parameters reflecting toughness of the material. Charpy tests are however easy to perform compared to standard fracture toughness tests, especially when the material is irradiated and quantity is limited. Correlations between the two parameters are therefore of great significance, especially for reactor pressure vessel (RPV) structural integrity assessment. In this paper, correlations between CVN impact energy and fracture toughness of three commonly used RPV steels, namely Chinese A508-3 steel, USA A533B steel, Euro 20MnMoNi55 steel, are investigated with two methods. One method applies a direct conversion using empirical formulas and the other adopts the Master Curve method. It is found that when the empirical formula is used, the difference between the predicted fracture toughness (from the CVN impact energy) and actual test data is relatively small in upper shelf, lower shelf and the bottom of transition region, while relatively large in other parts of the transition region. When the Master Curve method is adopted, whether the reference temperature T0 is estimated through temperature at 28J or 41J CVN impact energy, the predicted fracture toughness values of the three steels are consistent with actual test data. The reference temperature T0 is also estimated through the IGC-parameter correlation and through a combination of empirical formula and multi-temperature method. Both procedures show excellent agreement with test results. The mean value of T0 estimated from T28J, T41J, IGC-parameters and the combination method is denoted by TQ-ave and is then used as the final reference temperature T0 for the Master Curve determination. Accuracy of TQ-ave (and therefore the Master Curve method) is demonstrated by comparison with actual test data of the three RPV steels. It is concluded that Master Curve method provides a reliable procedure for predicting fracture toughness in the transition region utilizing limited CVN impact energy data from open literature.
Fracture Mechanics Study on the Effects of Various Notch Radii by Using FE Analysis
Authors: Hamid Zalikha Murni Abdul , Bae Keun-Hyung , Youn Gyo-Geun , Lee Dae-Young , Kim Yun-Jae , Kamaya Masayuki
Published: November 2015

This paper predicts the effects of notch root radius on the ductile fracture behavior of the structure through evaluation based on fracture mechanics concept. To understand the effects of notch radii on structure, FE analysis of J-integral for plane strain deformation fields with various size of notches were predicted. The fracture toughness of the specimens with various notch radii were determined by stress modified fracture strain damage model FE simulation that was established based on notched bar tensile test data. The simulated results were validated by comparison with the experiment data. Findings show that the value of apparent fracture toughness increases with the increase in the notch radii. Therefore, indicates the increase in the resistance to the crack propagation. Similar trend can be shown for both C(T) and M(T) specimen. Thus, shows that the proposed method can be used to obtain effects of various notch radii.
Survey of BWR Recirculation Line Break Acoustic Loads
Authors: Walter Matthew , Sommerville Daniel , Qin Minghao
Published: November 2015

There have been numerous industry communications over the past six years identifying issues pertaining to incorrectly calculated or erroneously omitted recirculation line break acoustic loads. Acoustic loads caused by a postulated recirculation line break (RLB) loss of coolant accident (LOCA) are one of the required design basis events that must be considered for stress analyses of Boiling Water Reactor (BWR) internal components such as jet pumps, core shrouds, access hole covers, and shroud support assemblies. These acoustic loads must also be considered for fracture mechanics evaluations performed to determine allowable operating periods for flaws detected during in-service inspections. Various organizations have studied the recirculation line break event and have developed methodologies to calculate the associated loads on BWR internal components. In general, loads are calculated in a plant specific and operating condition specific manner. Once these loads are available then the necessary structural evaluations can be performed. Load calculations can be expensive and time consuming. The authors have previously proposed a method for determining plant specific bounding RLB acoustic loads such that a utility can eliminate the need to periodically update the acoustic load calculations for each plant. This paper adds to the body of literature on this topic by providing a survey of BWR shroud acoustic loads, representative of the range of loads expected for the fleet, and provides a bounding load estimate that can be used for any plant in the fleet. The load survey supports previous arguments that the AC loads are expected to be substantially similar from plant to plant, based on the substantial similarity in plant operating conditions and geometry. The bounding load estimate provides significant value in that a utility engineer can rapidly obtain conservative loads to support either plant specific structural evaluations or fleet studies intended to investigate the significance of the load on inspection intervals, flaw tolerance, etc. Cost savings can be realized for utilities and fleet wide insight can be gained from understanding the similarity between RLB acoustic loads across the operating BWR fleet.
Finite Element Analysis of Cyclically-Loaded Steel Pipes During Deep Water Reeling Installation
Authors: Chatzopoulou Giannoula , Karamanos Spyros A. , Varelis George E.
Published: November 2015

Thick-walled steel pipes during their installation in deep-water are subjected to combined loading of external pressure and bending, which may trigger structural instability due to excessive pipe ovalization with catastrophic effects. The loading path followed during the reeling installation process is characterized by strong cyclic loading of the pipe material and results in residual stresses and deformations of the pipe cross-section, undermining the structural capacity of the pipe. Using advanced material tools, the present study examines the effect of reeling on the structural response and resistance of offshore pipes during the installation process.
Evaluation of Packing Drag on Knife Valves for Actuator Sizing
Authors: Garcia André F. , Veiga Jose C. , Romano Luiz P. , Girão Carlos D.
Published: November 2015

Knife Gate Valves (KGV) are one of many types of valves available in the fluid control industry. Initially used for handling slurry, KGV gradually became popular in many other engineering applications. Most of these valves are characterized by a particular stuffing box shape that require a rectangular packing installation, not so conventional in valve sealing designs where packings are installed around a cylindrical stem. The actuation methods used with KGV’s are many, from purely mechanical to pneumatic and electrically actuated. The design of those actuators is most dependent of the friction generated by the packing and, therefore, knowing more accurately how packing stress influences gate drag would be of much interest. The intent of this paper is to introduce a testing device and protocol to evaluate how packing friction behaves among distinct packing materials and to establish relations between packing gland load and actuator resistive force. Moreover, values of drag force for the different packing styles are determined with corresponding gland load.
Failure Criteria for Unirradiated PWR Cladding Subjected to Ring Compression Tests
Authors: Ruiz-Hervias Jesus , Martin-Rengel Miguel Angel , Gomez-Sanchez Francisco Javier
Published: November 2015

The ring compression test applied to nuclear fuel cladding is relatively easy to perform but difficult to interpret. It can be representative of the loading state associated to a hypothetical spent fuel assembly drop accident. This is particularly important for spent fuel cladding subjected to drying operations previous to storage and transportation, because they may produce hydride reorientation along the radial direction of cladding. In this paper, experimental testing and numerical simulations are combined to obtain operative failure criteria from the results of the ring compression tests on unirradiated pre-hydrided samples with radial hydrides, simulating drying, storage and subsequent transport conditions.
Wear Induced by Stochastic Sliding Impacts
Authors: Souilliart Thibaut , Rigaud Emmanuel , Le Bot Alain , Phalippou Christian
Published: November 2015

Vibrations of the steam generator tubes in nuclear power plants induce stochastic impacts between the tubes and their supports. As a consequence, wear is generated. A test rig is designed and used to perform impacts between two metal crossed cylinders with various incidence angles and impact velocities. The normal and tangential components of the contact load are measured during the tests. Rate and duration of impacts, instantaneous ratio between normal and tangential loads for each impact are deduced. Influence of incidence angle and impact velocity on impact duration, ratio between tangential and normal loads during impact and wear volume is highlighted.
Crack Tip Strain Distributions and Their Implications to Crack Growth Rate due to Stress Corrosion Cracking
Authors: Sung Shin-Jang , Kotasthane Nikhil , Ashida Yugo , Pan Jwo
Published: November 2015

In this paper, stress and strain distributions near a crack tip in a round compact tension specimen of elastic-plastic materials are obtained by finite element analyses. The strain distributions are used to explore the use of the crack tip strain distributions for crack growth rate models due to stress corrosion cracking in unirradiated and irradiated steels with different yield stresses and hardening behaviors. Both power-law hardening and perfectly plastic materials are considered. The computational results indicate that the critical radial distance to the tip based on the crack tip opening displacement is outside of the Hutchinson-Rice-Rosengren (HRR) dominant zone for power-law hardening materials in a round compact tension specimen under the stress intensity factor typically considered for stress corrosion cracking. For both the power-law hardening and perfectly plastic materials, the computational results show that the strain distributions are different from those of the analytical solutions for the range of the radial distance larger than the critical radial distance based on the crack opening displacement within the plastic zones. The computational results suggest that for the stress intensity factor typically considered for stress corrosion crack growth rate models, computational results are needed to estimate the strain rate for developing crack growth rate models to correlate to the experimental data.
Stress Analysis and Optimization Research for Ti Clad Steel Tube Sheet
Authors: Wang Ke , Xu Weifeng , Liu Zunchao , Liu Minshan
Published: November 2015

The models of Ti clad steel tube sheet and carbon steel tube sheet are established in this paper. By the finite element analysis software ANSYS Workbench, the disciplinarian of stress distribution is investigated under thermal and mechanical loadings of tube sheet. Through the optimization, the effects of base layer, clad layer thickness and tube wall thickness on clad steel tube sheet are obtained. Results show: the overall stress distribution of clad steel tube sheet is more complex than that of carbon steel tube sheet, and the connection surface of base and clad layer presents the phenomenon of stress concentration for clad steel tube sheet. The increase of clad layer or tube wall thickness has an unfavorable effect on the quality of clad steel tube sheet. Through reasonably decreasing thickness of clad layer or tube wall, stress concentration can be improved and performance of clad steel tube sheet can be enhanced. The results provide some reference for the design of clad steel tube sheet.
Analysis of Circumferential Welds in BWRs for Life Beyond 60
Authors: Purtscher Patrick , Sheng Simon , Dickson Terry
Published: November 2015

This paper describes the probabilistic fracture mechanics (PFM) analyzes of the conditional probability of failure (CPF) due to brittle fracture of circumferential welds (CW) from a cold overpressurize event in boiling water reactors (BWR) operated for 72 EFPY. This analysis used the Fracture Analysis for Vessels, Oak Ridge (FAVOR) computer code, developed at the Oak Ridge National Laboratory (ORNL), under United States Nuclear Regulatory Commission (NRC) funding. Two typical vessel configurations and the associated material properties for the beltline materials, CW, axial welds (AW), and plates (PL) were used. The analyses consider the potential effects of different fabrication options, shop vs field. Shop-fabrication is mainly by submerged arc weld (SAW) process, while field fabrication used the shielded metal arc weld (SMAW) process. In either case, repairs would have required the SMAW process. The calculations show that field-fabricated vessels would have a slight increase in the CPF compared to shop-fabricated vessels, but the assumed fraction of repair welds was more significant than the fabrication option. The details demonstrate the relative importance of surface-breaking flaws vs. embedded flaws for the assumed transient. The results confirm the conclusions from the original analysis from BWRVIP-05 and BWRVIP-74, the CPF for CW is orders of magnitude less than that of PL and AW regions of the vessel; therefore, the ASME Code-required volumetric examinations of the CW every 10 years as part of the in-service inspection (ISI) program does not change the overall CPF for the vessel. In all the cases analyzed, the total CPF values of the BWRs for 72 EFPY are below the goal for safe operation.
Validation of Thermal Analysis of Dry Storage Casks at Diablo Canyon
Authors: Li Jie , Liu Yung Y.
Published: November 2015

This paper is a continuation of previous work; it focuses on validating the thermal analysis of a vertical dry storage cask by using the measured temperature data and the results obtained by others in thermal modeling of a HI-STORM 100 storage cask at Diablo Canyon’s independent spent fuel storage installation (ISFSI). The cask chosen for thermal analysis contains a welded canister for 32 pressurized water reactor (PWR) used fuel assemblies in a stainless-steel basket with a total decay heat load of 17.05 kW. An effective thermal conductivity model was used to represent the used fuel assemblies with non-uniform assembly heat loads. The pressure of the canister’s helium fill gas was assumed to be 5 atm, and the ambient temperature was assumed to be 10°C.

The results showed reasonably good agreement between the calculated and measured canister axial surface temperatures. The results of ANSYS/FLUENT simulations showed that a tighter convergence criterion yielded slightly better agreement with the data; however, improvement could be obtained by adjusting the assumed ambient temperature value in the simulation. Validating the results of ANSYS/FLUENT simulation against the data (as well as the experience and additional insights gained from the validation exercise) is important to our future simulation and analysis of the thermal performance of dry storage casks, particularly for aging management and monitoring the condition and performance of dry casks during extended long-term storage at ISFSIs.
Geometrically-Accurate-Three-Dimensional Simulations of a Used Nuclear Fuel Canister Filled With Helium
Authors: Manzo Triton , Nacer Mustafa-Hadj , Greiner Miles
Published: November 2015

This paper presents preliminary results of heat transfer simulations performed in geometrically-accurate-three-dimensional model of nuclear fuel canister filled with helium. The numerical model represents a vertical canister, which relies on natural convection as its primary heat transfer mechanism, containing 24 PWR fuel assemblies. The model includes distinct regions for the fuel pellets, cladding and gas regions within each basket opening. Symmetry boundary conditions are employed so that only one-eighth of the package cross-section is included. The canister is assumed to be filled with helium at atmospheric pressure. A constant temperature of 101.7°C is employed on the canister outer surfaces, assuming the canister to be surrounded with water. These conditions of pressure and temperature were considered, in this paper, for comparison purpose with previous work. The effects of buoyancy-induced gas motion and natural convection, along with radiation and conduction through gas regions and solid are considered. Steady state simulations using ANSYS/Fluent were performed for different heat generation rates in the fuel regions. Simulations that include the effect of natural convection and others that do not include this effect are conducted. The peak cladding temperature and its radial and axial locations are reported. The maximum allowable heat generation that brings the cladding temperatures to the radial hydride formation limit (TRH=400°C) is also reported. The results of the three dimensional model simulations were compared to two dimensional model simulations for the same heat generation rate. The results showed that the two-dimensional simulations overestimate the temperature in the canister by almost 70°C.
Numerical Simulation of Screw Manufacturing Process for the Assessment of Surface Integrity
Authors: Robin Vincent , Gilles Philippe , Bosco Benoît , Mazuy Louis , Valiorgue Frédéric
Published: November 2015

Stress corrosion cracks have been observed on screws made of stainless steels grade 316 after some years of service in Pressurized Water Reactor (PWR) water environment. Grade 316 of stainless steel is not sensitive to corrosion unless it has been sensitized and/or subjected to a complex combination of factors including an important level cold work at the surface and in the bulk of the material. The tightening of the screw induces tensile stresses. This preload cannot explain the Stress Corrosion Cracking (SCC) defect appearing in the transition radius between the screw shank and its head. Thus, the question has been raised of the initial state of the screws after manufacturing. The simulation of the manufacturing processes has been carried out to have a better understanding of manufacturing process consequences on material degradation: solution annealing, cold drawing and machining. The dedicated “hybrid method”, specifically set up to simulate finish turning has been applied to obtain stress and strain states close to the surface. This method is detailed in the paper. The manufacturing process of these bolts is likely to induce high strain hardening since they have been cold drawn and then machined. It is suspected that tensile residual stress and cold work play a major role in the initiation of stress corrosion cracking of austenitic stainless steel grade of 316 type in PWR water environment. Simulation chaining method and results are highlighted in the paper with comparison with experiments. The main achievements are: the smaller the screw the less the cold work, the residual stress on the surface is mainly due to machining and the location of crack in the transition radius is well explained.
Numerical Simulation Investigation on Erosion of the Elbow Based on the Similarity Theory
Authors: Li Xiang , He Xingjian , Li Jun
Published: November 2015

Using the Discrete Phase Model (DPM) model in fluent software, a simulation study for elbow erosion situation was performed for different inlet particle concentrations, inlet velocities and curvature radiuses based on the similarity theory. The distribution of the serious erosion location and the fluid field of elbow can be determined. The results of numerical simulations show that the maximum speed is in the region of the small and big bend surfaces of the elbow. The small bend surface has a minimum pressure and the big bend surface has a maximum pressure. The most serious erosion region is mainly located in the 52°–60° region of the big bend surface near the outlet. The position and shape of the serious erosion almost does not change with the inlet concentration and velocity. Finally, based on the dimensional analysis the simulation analysis was carried out. The empirical formula on the inlet concentration, inlet velocity and bend radius ratio are obtained. The error of the formula is acceptable by verification. The research results can provide an effective help for selection of the elbow, determination of dangerous operational conditions and life prediction of the elbow.
Technical Basis for Extension of Section XI Appendix C Pipe Flaw Evaluation Procedures to Pipe Diameters Less Than NPS 4
Authors: Cipolla Russell C. , Miyazaki Katsumasa , Hasegawa Kunio
Published: November 2015

A lower limit restriction on nominal pipe size (NPS) was originally set as NPS 4 (DN 100) in ASME Section XI, Appendix C pipe flaw evaluation procedures (1983 Edition). Pipe less than NPS 4 that contained flaws could not be evaluated to the code and had to be repaired or a relief request submitted. The primary basis for this analysis restriction was that stress analysis information for small diameter lines required for flaw evaluation would be limited for a detailed flaw evaluation. The plant owner would opt to repair such flawed pipe rather than perform an evaluation. However, current evaluation procedures are generally applicability to smaller pipe sizes. Because ASME Appendix C procedures are referenced in code cases that cover small diameter pipe, and because small diameter pipe can be difficult to repair due to cost and radiation exposure, a proposed revision to ASME Section XI has extended Appendix C procedures to pipe sizes down to NPS 1 (DN 25). This improvement is now in the 2013 Edition of Section XI and gives the plant owner the option for evaluation for such situations.
Further Improvements to ASME Section XI Code Case N-806 for Evaluation of Metal Loss in Class 2 and 3 Metallic Piping Buried in a Back-Filled Trench
Authors: McGill Robert O. , Moenssens Mark A. , Antaki George A. , Scarth Douglas A.
Published: November 2015

ASME Section XI Code Case N-806, for evaluation of metal loss in Class 2 and 3 metallic piping buried in a back-filled trench, was first published in 2012. This Code Case has been prepared by the ASME Section XI Task Group on Evaluation Procedures for Degraded Buried Pipe. The Code Case addresses the nuclear industry need for evaluation procedures and acceptance criteria for the disposition of metal loss that is discovered during the inspection of metallic piping buried in a back-filled trench. A number of additional improvements have been proposed for Code Case N-806. These include expanded guidance for the determination and validation of a corrosion rate and other clarifications to improve ease of use. This paper presents an update of details of the proposed revisions to Code Case N-806 and their technical basis.
Effective Implementation of Risk Based Inspection (RBI) Approach in Asset Integrity Management of Oil and Gas Facilities
Authors: Akanni Joseph , Alonge Oladipo
Published: November 2015

Effective implementation of asset integrity management in oil and gas facilities requires a structured approach to the inspection of all equipment to get maximum value for time and resources. This is what Risk Based Inspection (RBI) strives to achieve by optimizing inspection and monitoring efforts based on degradation effects and other associated risks. The risk based approach entails a detailed assessment of all possible degradation mechanisms for equipment, operating context of the facility as well as the health, safety and environment impact of equipment failure. All these assessments and considerations extend the impact of RBI beyond its initial objective of achieving an optimised inspection frequency to also ensuring process safety, increased understanding of degradation risk and assuring asset integrity.

This paper focuses on the workflow, process, benefits and challenges of a typical RBI implementation for static equipment in facilities.
Re-Characterization of Subsurface Flaw to Surface Flaw Based on Equivalent Fatigue Crack Growth Rate
Authors: Hasegawa Kunio , Li Yinsheng , Lacroix Valery , Strnadel Bohumir
Published: November 2015

A subsurface flaw located near a component surface is transformed to a surface flaw in accordance with a flaw-to-surface proximity rule. The re-characterization process from subsurface to surface flaw is adopted in all fitness-for-service (FFS) codes. However, the criteria of the re-characterizations are different among the FFS codes. In addition, the proximity factors in the rules are defined by constant values, irrespective of flaw aspect ratios. This paper describes the stress intensity factor interaction between the subsurface flaw and component free surface, and proposes a proximity factor from the point of view of fatigue crack growth rates.
Creep Relaxation Characteristics in European Standards
Authors: Schaaf Manfred
Published: November 2015

The European Standards EN 1591-1 which is used for the calculation of bolted flanged joints and EN 13555 in which the determination of the required gasket characteristics are defined were reissued last year.

The definition of the creep/relaxation factor in EN 13555 and EN 1591-1 has been modified in these new issues of the standards. Instead of the ratio between the remaining and the initial gasket stress, now the additional gasket deflection due to creep/relaxation effects is considered in the calculation code, and therefore also the evaluation of the corresponding gasket tests has been changed.

In this paper the differences in the testing procedure and in the evaluation algorithm are shown, and also a proposal is given to simplify and to reduce the required creep/relaxation testing.
Application of the ASME Code to Radioactive Material Packaging/Cask: Lessons Learned From the 2014 DOE Training Course
Authors: Han Zenghu , Fabian Ralph , Pope Ron , Liu Yung , Shuler James
Published: November 2015

The U.S. Department of Energy (DOE) Packaging Certification Program (PCP), Office of Packaging and Transportation, Office of Environmental Management, has sponsored a suite of training courses that are conducted annually by Argonne National Laboratory (Argonne) in support of safety and security of nuclear and other radioactive material packages. One of these courses conducted by Argonne since 2000 is the Application of the ASME Code to Radioactive Material Transportation Packaging, which was expanded significantly in 2014 to include dry storage casks, resulting in a change in course title to the Application of the ASME Code to Radioactive Material Packaging/Cask. The purpose of the course is to provide guidance for the application of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code (“ASME Code”) to transportation packaging and storage cask of radioactive materials, including used (or spent) nuclear fuel and high-level waste, and to facilitate the design, fabrication, examination, and testing of packagings and casks. Both regulatory requirements in 10 CFR Parts 71 and 72 and the ASME Code requirements for transportation and storage containments are addressed, with emphasis on the Code Section III, Division 3, “Containments for Transportation and Storage of Spent Nuclear Fuel and High Level Radioactive Material and Waste.” Among the specific topics covered are the application of the ASME Code requirements to structural materials, containments, loading and design; the design of containment internal support structures and buckling analysis; fabrication, welding, examination, and test requirements; quality assurance; physical testing, structural and thermal modeling and analysis considerations; and containment, shielding, and criticality analysis considerations. Special topics covered include non-Code materials, hydrogen gas generation, and aging management for extended long-term storage of used fuel and subsequent transportation. The expanded training course was offered in June 2014 at Argonne with 27 participants representing mainly industry and government agencies. On the basis of the feedback and course evaluation by the participants, the course may be expanded from 3 to 4.5 days in the future to allow more time for in-class discussion and exercises, as well as to include additional topics related to aging management for extended long-term storage of used fuel and its post-storage transportation. The course provides insight into the DOE and the U.S. Nuclear Regulatory Commission (NRC) transportation and storage cask certification processes. The target audience is DOE, DOE contractors, other agency personnel, and commercial transportation packaging and storage cask engineering employees. Those responsible for designing, fabricating, testing, or packaging and casks, as well as preparing or reviewing the associated Safety Analysis Reports, will also benefit from the course.
Large EDF Tests on Aged Cast Duplex Stainless Steel Components: Part II — Full Scale Tests
Authors: Le Delliou Patrick , Saillet Sébastien , Bezdikian Georges
Published: November 2015

Thermal ageing of cast duplex stainless steel primary loops components (elbows, pump casings and branch connections) is a concern for long-term operation of EDF nuclear power plants. The thermal ageing embrittlement results from the microstructural evolution of the ferrite phase (spinodal decomposition), and can reduce the fracture toughness properties of the steel. In addition, it is necessary to consider manufacturing quality and the possible occurrence of casting defects such as shrinkage cavities. In a context of life extension, it is important to assess the safety margins to crack initiation and crack propagation instability.

This paper presents two tests conducted by EDF on aged cast duplex stainless steel NPP components, respectively on a full-scale elbow and a branch connection. The main characteristics of the tests are recalled, the results are presented, and finally, the lessons drawn are summarized.

These tests and their detailed analyses contribute to validate and justify the methodology used by EDF in the integrity assessment of in-service cast duplex stainless steel components.