bluuebunny/crossref_metadata_2025 huggingface dataset:
An Analytical Evaluation of the Efficacy of the Mechanical Stress Improvement Process in Pressurized Water Reactor Primary Cooling Piping
Authors: Fredette L. F. , Scott Paul M. , Brust F. W.
Published: July 2009

The Mechanical Stress Improvement Process (MSIP) has been used successfully to mitigate intergranular stress corrosion cracking in boiling water reactor (BWR) welded stainless steel piping for many years. The MSIP technique creates compressive residual stresses at the inside surface of the welded joint while producing a slight permanent deformation of the pipe on one side of the weld. A prerequisite for use of MSIP on welded pipes susceptible to primary water stress corrosion cracking (PWSCC) in pressurized water reactors (PWR) is knowledge of the efficacy of the process when applied to dissimilar metal welds common to PWR primary cooling piping. This study uses two and three dimensional finite element models to evaluate the likely residual and operating stress profiles remaining after MSIP for typical dissimilar metal weld configurations, some of which are approved for leak-before-break (LBB) applications in pressurized water reactors.
Regulatory Perspective on Advanced Finite Element Flaw Growth Analysis of Alloy 82/182 Butt Welds
Authors: Sullivan Edmund J. , Csontos Aladar A. , Lupold Timothy R. , Sheng Chia-Fu
Published: July 2009

On October 13, 2006, the Wolf Creek Nuclear Operating Corporation performed preweld overlay inspections using manual ultrasonic testing (UT) techniques on the surge, spray, relief, and safety nozzle-to-safe end dissimilar metal (DM) and safe end-to-pipe stainless steel butt welds. The inspection identified five circumferential indications in the surge, relief, and safety nozzle-to-safe end DM butt welds that the licensee attributed to primary water stress corrosion cracking (PWSCC). These indications were significantly larger and more extensive than previously seen for the case of circumferential indications in commercial pressurized water reactors. As a result of the NRC staff’s initial flaw growth analyses, the NRC staff obtained commitments from the nuclear power industry licensees to complete pressurizer nozzle DM butt weld inspections on an accelerated basis. In addition, the industry informed the NRC staff that it would undertake a task to refine the crack growth analyses using more realistic assumptions to address the NRC staff’s concerns regarding the potential for rupture without prior evidence of leakage from circumferentially oriented PWSCC in pressurizer nozzle welds. These new analyses are referred to as advanced finite element (AFE) analyses. This paper will discuss the regulatory review of the industry’s AFE analyses. This discussion will include the NRC staff’s approach to the review, the differences between the industry’s AFE analyses and the NRC staff’s confirmatory analyses, and the NRC staff’s acceptance criteria. The paper will discuss the impact of the AFE analyses on the regulatory process. Finally, the paper will discuss possible future regulatory and research applications for AFE analyses as well as additional NRC research projects intended to address some of the uncertainties in this type of analysis.
Assessment of the Reactor Pressure Vessel Integrity in PTS Conditions: Presentation and Demonstration of SECURE — A Numerical Tool Developed by EDF to Realise Automatically Integrity Assessments
Authors: Dahl A. , Messelier-Gouze C.
Published: July 2009

The reactor pressure vessel (RPV) is an essential component liable to limit the lifetime duration of PWR nuclear power plants. For the RPV assessment and Lifetime evaluation of the nuclear plants, French Utility applies a series of calculations including different approaches, more or less refined. These approaches combine different types of data : neutron physics, vessel geometry, material data, thermal and mechanical transients. SECURE, a numerical tool based on EDF finite element code, code_Aster, and full data bases was developed by EDF in order to make automatically RPV integrity assessment with simplified approaches (1D computations) and more refined approaches (2D and 3D computations). Then, with this tool, RPV integrity studies are made more quickly and with more reliability. This paper describes SECURE tool and presents an application with more and less refined methods on a 3D thermal computation of a transient. This application shows that this tool allowed to do easily refined numerical methods which improve the final margins. This tool, developed and validate by EDF Research and development division, must now be used by engineering division.
Incorporation of the New Irradiation Embrittlement Correlation Method Into the Japanese Code of Surveillance Tests for Reactor Vessel Materials
Authors: Asada Seiji , Hirano Takashi , Soneda Naoki , Yamashita Norimichi , Yonehara Akira , Tomimatsu Minoru
Published: July 2009

A new irradiation embrittlement correlation method for reactor pressure vessel (RPV) materials was developed by the Central Research Institute of Electric Power Industry (CRIEPI). In order to incorporate this new correlation method into the Japan Electric Association Code, JEAC 4201, “Method of Surveillance Tests for Structural Materials of Nuclear Reactors”, sensitivity studies for PWR and BWR reactor vessel materials were performed and the features of the new correlation method was investigated. Based on those studies, the prescription of the new correlation method in JEAC 4201 was formulated. The results of the studies and the overview of the revised JEAC 4201 are presented in this paper.
Numerical Methods for Welding Simulation: The Next Technical Step
Authors: Leggatt N. A. , Dennis R. J. , Smith M. C. , Bouchard P. J.
Published: July 2009

Numerical methods have been established to simulate welding processes, often based around the use of methods which represent the welding process in a simplified manner. Simplified methods include simultaneous deposition of weld beads and bead lumping where stringers or individual weld beads are grouped together and deposited. These approaches are widely accepted, however the requirement for simplified methods often results in compromises to the solution accuracy usually driven by limitations in data and the capability of computing hardware. In many cases this compromise in accuracy is acceptable providing it is well understood, however there are frequently cases where such simplifications are unacceptable and improved representation of the welding process is required. In practice this generally implies the requirement for a full moving heat source simulation. The transition from simplified simulation methods to the next technical step, full moving heat source simulations, is now possible for a wide variety of scenarios as will be demonstrated in this paper. This paper presents two specific cases, a 3 pass slot weld and a multipass repair weld, where full moving heat source simulations have been considered necessary. For each of these cases the reasons why moving heat source methods are necessary and the benefits that this more demanding simulation technique offers are described. Furthermore the predicted residual stress results are compared with residual stress measurements using a variety of measurement techniques. The work provides an extremely useful insight into how moving heat source methods are now considered a practical analysis method for a wide variety of real world problems. Of further consideration is the fact that in the 2 years since the work reported in this paper was undertaken computing performance would have at least doubled.
The Influence of Specimen Geometry and Test Times on High Temperature Crack Growth Behaviour in Parent and Welded 316H Stainless Steel
Authors: Davies Catrin M. , Wimpory Robert C. , Tabuchi Masaakai , Dean David W. , Nikbin Kamran M.
Published: July 2009

Experimental crack growth testing has been performed at 550 °C on a range of fracture specimens including sections taken from a 316 steel weldment. These specimens include the compact tension, C(T), and circumferentially cracked notched bar, CCB, geometries of various sizes. Results are presented from two creep crack growth (CCG) tests on a large and a small CCB weldment specimen. The creep crack initiation (CCI) and growth (CCG) behavior of the CCB weldments has been compared to that of homogeneous parent material (PM) CCB and C(T) specimens and to C(T) weldment specimen data. The data has been analyzed in terms of the C* parameter. The initiation period is found to occupy a large fraction of the test duration for weldments. The CCG rates in the larger CCB weldment test is on the order of six times faster, for a given value of C*, compared to the smaller specimen, indicating a specimen size effect. The CCI times are around an order of magnitude greater for the CCB weldment specimens compared to C(T) weldment data and are higher than that of the PM CCB data. It is recommended that further testing on weldment specimens is performed to affirm the apparent trends.
The Role of Failure Data in Plant Aging Management and Life Extension
Authors: Chockie Alan D. , Gregor Frank E.
Published: July 2009

The aging management and life extension process for the United States (US) nuclear industry has been refined and improved over the years. It has become an efficient and effective method to ensure that the nuclear plants in the US can be safely operated beyond their original 40-year operating license. The wealth of failure data from Nuclear Regulatory Commission and industry aging research programs in combination with plant specific data has provided a strong technical foundation for the aging management and long-term plant operation. This paper examines the role that generic and plant specific failure data has played in the development, application, and refinement of aging management and life extension efforts at US nuclear plants.
Comparison of Experimental and Numerical Analysis Data for BMI Mock-Up With Dissimilar Metal Welds
Authors: Yu Seung-Cheon , Chang Yoon-Suk , Kim Young-Jin , Kim Sung-Woo , Hwang Seong-Sik , Kim Hong-Pyo
Published: July 2009

During the last decade, several defects due to primary water stress corrosion cracking (PWSCC) have been reported at bottom-mounted instrumentation (BMI) and control rod drive mechanism nozzles. The exact locations were dissimilar metal weld parts which are greatly important because the cracking could lead to leakage of primary coolant. The PWSCC of BMI mock-up’s penetration with dissimilar metal welds was examined by using doped steam test method by Korea Atomic Energy Research Institute. In this work, numerical analyses are performed for the same environment condition with the doped steam test. With respect to the numerical analyses, heat transfer analyses are carried out based on thermal conduction. The welding paths are simulated by using lumped path method for conservative evaluation and model change (remove/rebirth) method. Then residual stress analyses are conducted using the heat transfer analysis results, in which annealing effect of welding process simulation is considered for resetting the plastic deformation. However, the plastic behaviour of steels during phase transformations is not considered with experimental data. In addition, the consequence of weld residual stress that is known as the cause of PWSCC is being investigated.
Uncertainty Estimate of Charpy Data Using a 5-Factor 8-Run Design of Experiments
Authors: Interrante Charles G. , Fong Jeffrey T. , Filliben James J. , Heckert N. Alan
Published: July 2009

Scatter in laboratory data with duplicates on Charpy impact tests is analyzed by identifying several sources of variability such as temperature, manganese sulfide, initial strain, mis-orientation, and notch radius in order to estimate the predictive 95% confidence intervals of the mean energy of absorption for each specific test temperature. Using a combination of real and virtual data on a high-strength pressure vessel grade steel (ASTM A517) over a range of temperatures from −40 °C (−40 °F) to 182 °C (360 °F), and the concept of a statistical design of experiments, we present an uncertainty estimation methodology using a public-domain statistical analysis software named DATAPLOT. A numerical example for estimating the mean, standard deviation, and predictive intervals of the Charpy energy at 48.9 °C (120 °F) is included. To illustrate the application potential of this methodology, we enhance it with formulas of error propagation to estimate the mean, standard deviation, and predictive intervals of the associated static crack initiation toughness, KIc. A discussion of the significance and limitations of the proposed methodology, and a concluding remark are given at the end of this paper.
A Proposed Methodology for Strain-Based Failure Criteria
Authors: Wu Tsu-Te
Published: July 2009

This paper proposes an alternative methodology to determine the failure criteria for use in dynamic simulations of radioactive material shipping packages in the events of hypothetical accident conditions. The current stress failure criteria defined in the Nuclear Regulatory Guide 7.6 [1] and the ASME Code, Section III, Appendix F [2] for Level D Service Loads are based on the ultimate strength of uniaxial tensile test specimen rather than on the material fracture in the state of multi-axial stresses. On the other hand, the proposed strain-based failure criteria are directly related to the material failure mechanisms in multi-axial stresses. In addition, unlike the stress-based criteria, the strain-based failure criteria are applicable to the evaluation of cumulative damages caused by the sequential loads in the hypothetical accident events as required by the Nuclear Regulatory Guide 7.8 [4].
Strain Rate Dependence and Short-Term Relaxation Behavior of a Thermoset Polymer at Elevated Temperature: Experiment and Modeling
Authors: McClung A. J. W. , Ruggles-Wrenn M. B.
Published: July 2009

The inelastic deformation behavior of PMR-15 neat resin, a high-temperature thermoset polymer, was investigated at 288°C. The experimental program was designed to explore the influence of strain rate changes in the 10−6 to 10−3 s−1 range on tensile loading, unloading, and strain recovery behavior, as well as on the relaxation response of the material. The material exhibits positive, nonlinear strain rate sensitivity in monotonic loading. Nonlinear, “curved” stress-strain behavior during unloading is observed at all strain rates. The strain recovery at zero stress is profoundly affected by prior strain rate. The prior strain rate is also found to have a strong influence on relaxation behavior. The rest stresses measured at the termination of relaxation tests form the relaxation boundary which resembles a nonlinear stress-strain curve. The results suggest that the inelastic behavior of the PMR-15 solid polymer at 288°C can be represented using a unified constitutive model with an overstress dependence of the inelastic rate of deformation. The experimental data are modeled with the viscoplasticity theory based on over-stress (VBO). A systematic procedure for determining model parameters is presented and the model is employed to predict the response of the material under various test histories.
Evaluation of Representative Piping Systems Designed by Implicit Fatigue Concept
Authors: Choi Shin-Beom , Kim Sun-Hye , Chang Yoon-Suk , Choi Jae-Boong , Kim Young-Jin , Lee Jin Ho , Kim Jin-Su , Chung Hae-Dong
Published: July 2009

NUREG-1801 provides generic aging lessons learned to manage aging effects that may occur during continued operation beyond the design life of nuclear power plant. According to this report, the metal fatigue, among several age-related degradation mechanisms, is identified as one of time-limited aging analysis item. The objective of this paper is to introduce fatigue life evaluation of representative surge line and residual heat removal system piping which was designed by implicit fatigue concept. For the back-fitting evaluation employing explicit fatigue concept, detailed parametric CFD as well as FE analyses results are used. The well-known ASME Section III NB-3600 procedure is adopted for the metal fatigue and NUREG/CR-5704 procedure is further investigated to deal with additional environmental water effects. With regard to the environmental effect evaluation, two types of fatigue life correction factors are considered, such as maximum Fen and individual Fen. As a result, it was proven that a thermal stratification phenomenon is the governing factor in metal fatigue life of the surge line and strain rate is the most important parameter affecting the environmental fatigue life of both piping. The evaluation results will be used as technical bases for continued operation of OPR 1000 plant.
Lessons Learned From the Application and Refinement of Risk-Informed Inservice Inspection Programs
Authors: Chockie Alan D. , Graybeal M. Robin , Kulat Scott D.
Published: July 2009

The risk-informed inservice inspection (RI-ISI) process provides a structured and systematic framework for allocating inspection resources in a cost-effective manner while improving plant safety. It helps focus inspections where failure mechanisms are likely to be and where enhanced inspections are warranted. To date, over eighty-five percent of US nuclear plants and a number of non-US plants have implemented, or are in the process of implementing, RI-ISI programs. Many are already involved in the periodic update of their RI-ISI program. The development of RI-ISI methodologies in the US has been a long and involved process. The risk-informed procedures and rules were developed to take full advantage of PRA data, industry and plant experiences, information on specific damage mechanisms, and other available information. An important feature of the risk-informed methodologies is the requirement to make modifications and improvements to the plant’s RI-ISI application as new information and insights become available. The nuclear industry, ASME Section XI, and the Nuclear Regulatory Commission have all worked together to take advantage of the lessons learned over the years to refine and expand the use of risk-informed methodologies. This paper examines the lessons learned and the benefits received from the application and refinement of risk-informed inservice inspection programs. Also included in the paper is a review of how the information and insights have been used to improve the risk-informed methodologies.
Measurement and Uncertainty of Heat Flux to a Rail-Cask Size Pipe Calorimeter in a Pool Fire
Authors: Kramer M. Alex , del Valle Marcelo A. , Greiner Miles
Published: July 2009

The goal of this work was to measure the temporally varying heat flux and surface temperature of a pipe calorimeter in a pool fire, and assess its uncertainty. Three large-scale fire tests were conducted at the Sandia National Laboratories outdoor fire test facility. In each test a cylindrical calorimeter was suspended above a water pool with JP8 fuel floating on top. The calorimeter was a 2.4 m diameter, 4.6 m long, and 2.5 cm wall thickness pipe with end-caps suspended 1 m above the 7.2 m diameter pool. 58 thermocouples were attached to the calorimeter interior surface and backed with 8 cm of insulation. The Sandia One-Dimensional Direct and Inverse Thermal (SODDIT) code was used to determine the calorimeter external surface heat flux and temperature from the measured interior surface temperature versus time. To determine the uncertainty of the SODDIT results, a simulation of the calorimeter in a fire similar to the experiments was performed using the Container Analysis Fire Environment (CAFE) computer code. In this code, a Computational Fluid Dynamics (CFD) fire model applies a temporally and spatially varying heat flux to the exterior surface of a Finite Element (FE) calorimeter model. Flux is similar but not identical to the flux in the experiment. The FE model calculates the internal calorimeter surface temperature, which is used by SODDIT to calculate heat flux which was compared to the applied values. The absorbed heat flux and surface temperature at one calorimeter location was calculated by SODDIT and then compared to the CAFE applied heat flux and surface temperature. From this comparison a base case uncertainty due to inherent inverse calculation errors and frequency smoothing methods is presented. Uncertainties in temperature measurements, calorimeter material properties and wall thickness were applied to the SODDIT calculation and iterated using the Monte Carlo method to determine the overall heat flux and surface temperature uncertainty. The total absorbed heat flux uncertainty at the one studied location is ±4.8 kW/m2 at 95% confidence. The outer surface temperature uncertainty for all data at the one studied location is ±6.6°C at 95% confidence. For all 58 measurement locations, the overall combined total absorbed heat flux uncertainty is ±13.8 kW/m2 at 95% confidence, surface temperature uncertainty is ±7.6°C. These uncertainties are valid only when the calorimeter temperature is not within the Curie temperature range of 999 to 1037K.
Robust Engineering Design for Failure Prevention
Authors: Fong Jeffrey T. , Filliben James J. , Heckert N. Alan , deWit Roland , Bernstein Barry
Published: July 2009

To advance the state of the art of engineering design, we introduce a new concept on the “robustness” of a structure by measuring its ability to sustain a sudden loss of a part without causing an immediate collapse. The concept is based on the premise that most structures have built-in redundancy such that when the loss of a single part leads to a load redistribution, the “crippled” structure tends to seek a new stability configuration without immediate collapse. This property of a “robust” structure, when coupled with a continuous or periodic inspection program using nondestructive evaluation (NDE) techniques, is useful in failure prevention, because such structure is expected to display “measurable” signs of “weakening” long before the onset of catastrophic failure. To quantify this “robustness” concept, we use a large number of simulations to develop a metric to be named the “Robustness Index (RBI).” To illustrate its application, we present two examples: (1) the design of a simple square grillage in support of a water tank, and (2) a classroom model of a 3-span double-Pratt-truss bridge. The first example is a “toy” problem, which turned out to be a good vehicle to test the feasibility of the RBI concept. The second example is taken from a textbook in bridge design (Tall, L., Structural Steel Bridge, 2nd ed., page 99, Fig. 4.3(b), Ronald Press, New York NY, 1974). It is not a case study for failure analysis, but a useful classroom exercise in an engineering design course. Significance and limitations of this new approach to catastrophic failure avoidance through “robust” design, are discussed.
A Web-Based Data Analysis Methodology for Estimating Reliability of Weld Flaw Detection, Location, and Sizing
Authors: Fong Jeffrey T. , Hedden Owen F. , Filliben James J. , Heckert N. Alan
Published: July 2009

Recent advances in computer technology, internet communication networks, and finite element modeling and analysis capability have made it feasible for engineers to accelerate the feedback loop between the field inspectors of a structure or component for critical flaws by nondestructive evaluation (NDE) and the office engineers who do the damage assessment and recommendations for field action to prevent failure. For example, field inspection data of critical flaws can be transmitted to the office instantly via the internet, and the office engineer with a computer database of equipment geometry, material properties, past loading/deformation histories, and potential future loadings, can process the NDE data as input to a damage assessment model to simulate the equipment performance under a variety of loading conditions until its failure. Results of such simulations can be combined with engineering judgment to produce a specific recommendation for field action, which can also be transmitted to the field by the internet. In this paper, we describe a web-based NDE data analysis methodology to estimate the reliability of weld flaw detection, location, and sizing by using a public-domain statistical data analysis software named DATAPLOT and a ten-step sensitivity analysis of NDE data from a two-level fractional factorial orthogonal experimental design. A numerical example using the 1968 ultrasonic examination data of weld seam in PVRC test block 251J, and the 1984 sectioning data of 251J containing 15 implanted flaws, is presented and discussed.
Advanced FEA Modeling of PWSCC Crack Growth in PWR Dissimilar Metal Piping Butt Welds and Application to the Industry Inspection and Mitigation Program
Authors: White G. , Broussard J. , Collin J. , Klug M. , Harrington C. , DeBoo G.
Published: July 2009

In late summer 2005, the U.S. pressurized water reactor (PWR) fleet imposed mandatory inspection requirements upon itself to address the challenge posed by primary water stress corrosion cracking (PWSCC) in PWR reactor coolant system (RCS) dissimilar metal (DM) piping butt welds. Under this program, the highest temperature, and thus most susceptible, locations have been addressed first. The set of highest temperature locations comprises the DM piping butt welds on the pressurizer. Within three years of promulgating the requirements, all pressurizer locations will have been inspected and nearly 90% of these locations will have been mitigated. In October 2006, several indications of circumferential flaws were reported in the pressurizer nozzles at Wolf Creek. These indications raised questions about the need to accelerate refueling outages or take mid-cycle outages at other plants. In order to address these concerns, an industry effort was undertaken to evaluate the viability of detection of leakage from a through-wall flaw in an operating plant to preclude the potential for rupture of pressurizer nozzle DM welds given the potential concern about growing circumferential stress corrosion cracks. Previous calculations of growth of PWSCC in Alloy 600 wrought materials and Alloy 82/182 weld metal materials have assumed an idealized crack shape, typically a semi-ellipse characterized by a length-to-depth aspect ratio. A key aspect of the industry effort involved developing an advanced finite-element analysis (FEA) methodology for predicting crack growth when loading conditions do not lead to a semi-elliptical flaw shape. The work also investigated an extensive crack growth sensitivity matrix to cover geometry, load, and fabrication factors, as well as the uncertainty in key modeling parameters including the effect of multiple flaw initiation sites in a single weld. Other key activities included detailed welding residual stress simulations covering the subject welds, development of a conservative crack stability calculation methodology, development of a leak rate calculation procedure using existing software tools (EPRI PICEP and NRC SQUIRT), and verification and validation studies. This paper will describe the study undertaken to model growth of circumferential weld cracks and its application to a group of nine PWRs with regard to implementation of the industry inspection and mitigation program [1]. The paper will also explore implementation progress of the industry program as the three-year mark approaches, as well as industry actions to support completion of baseline DM weld examinations.
Second Order Errors Related to Geometric Nonlinearity in Explicit Central Difference Operator
Authors: Metzger Don R. , Kim Young-Suk
Published: July 2009

Numerical analysis of nonlinear dynamic structures frequently makes use of the central difference method to step the transient forward in time. The method is particularly robust, accommodating material and geometric nonlinearities as well as contact surfaces and constraints of a very general nature. The implementation of the method is most usually performed according to [1], where velocity terms (or more generally rate quantities) are taken half a time step from the displacement and acceleration terms. It was recognized that a proper check of energy balance, requires that velocity must also be interpolated to the integer steps [2]. The stability and accuracy of the central difference method is well established, and decades of experience including its use in numerous commercial finite element codes confirms why it is the method of choice for explicit time integration of transients.
Discussion of Issues Related to Surge in LNG Pipelines at Offloading Terminals
Authors: McGuffie Sean
Published: July 2009

Due to requirements of LNG unloading terminals, the pipelines used to transport the liquid operate near the vapor pressure of the LNG. If the operational pressure in the pipeline falls below the vapor pressure, pockets of gas will develop; when these pockets collapse, very high pressure pulses can be transmitted through the pipeline, an event known as surge or waterhammer. This paper discusses transients that occur during normal and upset plant operation and how these transients can induce surge in the pipeline. The paper concludes with an overview of the methods used to evaluate whether surge will occur and the peak pressure associated with surge events, with an overview of cases that are typical candidates for analysis.
Suggested Improvements to Appendix G of ASME Section XI Code
Authors: Mehta Hardayal S. , Griesbach Timothy J. , Stevens Gary L.
Published: July 2009

This paper reviews some of the original basis documents for ASME Section XI Nonmandatory Appendix G for calculating pressure-temperature (P-T) limits and recommends areas for improvement. The original Appendix G in Section XI of ASME Code was mainly based on Welding Research Council (WRC) Bulletin 175 (WRC-175). Changes have been made to Appendix G over the past 20 years such as the use of the KIC reference toughness curve instead of KIR. However, aspects of the Appendix G method still refer back to WRC Bulletin 175. The published technical literature since the development of WRC 175 could be used to enhance the Appendix in a number of areas. One such area is stress intensity factor (K) calculation procedures for thermal gradient loading at a nozzle corner. This paper will review and evaluate the available K calculation methods for a nozzle corner crack, and develop closed-form expressions for incorporation into Appendix G. Also, the following areas will be reviewed: (1) treatment of operating stresses exceeding the material yield stress, and (2) fracture toughness criteria typically used for other than reactor pressure vessel (RPV) and piping for protection against non-ductile failure. This paper will also identify areas for future improvements in Appendix G.
Load and Resistance Factor Design for Nuclear Pipes: Benefits and Challenges
Authors: Avrithi Kleio
Published: July 2009

The Board on Nuclear Codes and Standards (BNCS) recently decided to promote standards that use risk-informed design methods. In civil engineering practice a risk informed method, namely the Load and Resistance Factor Design (LRFD), has been in usage for quite some time. It is possible to extend such methods to the design of safety-related piping as well. This paper provides a brief overview of the LRFD method. Discussion is included for load factors to be used to account for the uncertainties in piping loads (e.g., internal pressure, sustained weight, etc.) and resistance factors to be used for addressing the uncertainties in strength of piping and analysis methods. Different load factors and resistance factors can be suggested for each load type and resistance type (e.g., hoop stress, bending stress, etc.). A design example for a feed water Class 2 piping system is provided to demonstrate the benefits of LRFD. This way, benefits such as the achievement of consistent reliability levels and the facilitation of a detailed risk analysis of mechanical systems are illustrated. Finally, the challenges associated with development of the LRFD method for nuclear piping are discussed. Such challenges pertain to the selection of the appropriate target reliability indices for piping, the development of equations for components such as tees, elbows, etc.
Development and Validation of Analysis Method for Simulating Residual Stresses in Dissimilar Metal Pipe Butt Welds
Authors: Killian Douglas E. , Mahmoud Samer H.
Published: July 2009

It is now accepted practice in the commercial nuclear power industry to numerically simulate the residual stresses developed during welding of bimetallic piping components. A common example is the Alloy 82/182 shop weld between a low alloy steel nozzle and a stainless steel safe end. Piping is attached to the nozzle safe end by a separate girth butt weld in the field. The safe end to pipe weld influences stresses in the nozzle to safe end weld only for relatively short safe ends. Alloy 600 base metal material and its associated welds are susceptible to primary water stress corrosion cracking (PWSCC) at elevated levels of stress. It is therefore necessary to model each step of the fabrication process to capture the path dependent accumulation of stress and strain in the welded components, and to simulate the pre-service test loads and operating temperature and pressure loads. The fabrication process includes the build up of butter material on the nozzle prior to heat treatment and the deposition of weld metal in a single V-groove between the butter and safe end. This is followed by one or more repair welds to remove welding defects. The welded component is then subjected to a shop hydrostatic test pressure equal to 125 percent of the design pressure. Following several heatup and cooldown cycles to simulate “shakedown” of pre-service stresses, stresses in the vicinity of the weld are determined at operating conditions of temperature and pressure in order to access the susceptibility of the weld and butter to PWSCC. This paper discusses the analytical methodology used to numerically simulate welding residual stresses using the ANSYS general purpose computer code. The method is then applied to two-dimensional axisymmetric welding simulations, which are adequate to investigate most girth butt weld piping components, although it may be necessary to make conservative assumptions regarding weld repair configurations. Numerical results are used to characterize the effects of weld repair, pre-service loads, and cyclic loading on the final state of operating stresses for sustained loading conditions. The analysis method is validated using results from large scale tests performed by the European Commission Joint Research Centre as part of its Assessment of Aged Piping Dissimilar Metal Weld (ADIMEW) project. Residual stresses are predicted for a ferritic-to-austenitic dissimilar metal weld mockup and compared against results from neutron diffraction measurements.
Effect of Embrittlement Trend Curve Model on the Reference Temperature Based Screening Limits for Pressurized Thermal Shock
Authors: EricksonKirk Mark T. , Dickson Terry L.
Published: July 2009

Recently, the USNRC has proposed a voluntary alternative to the pressurized thermal shock (PTS) rule (10CFR50.61(a)) that may be used instead of the existing PTS rule (10CFR50.61) by power reactor licensees to demonstrate the safety of their RPVs. The reference temperature (RT) screening limits in 10CFR50.61(a) are less restrictive than those of 10CFR50.61; this relaxation being justified by the much more realistic models that underlie the technical basis of the revised rule. One of the many models that constitute this technical basis is used to estimate the effect of irradiation damage on the fracture toughness transition temperature. This relationship, which is called the embrittlement trend curve, or ETC, can change as time goes on due to the continued development of new knowledge (e.g., from on-going reactor vessel surveillance programs, from test reactor studies aimed at better understanding the physical mechanisms of irradiation damage, etc.). In this paper we present information that quantifies the effect of different ETCs on the through wall cracking frequencies that are calculated using the probabilistic fracture mechanics computer code that was used to establish the RT limits that have been proposed in 10CFR50.61(a).
Insights and Observations Arising From Curve-Fitting the Charpy V-Notch and Tensile Data Contained Within the United States’ Light Water Reactor Surveillance Database
Authors: EricksonKirk Mark T. , Shaikh Atif , EricksonKirk MarjorieAnn
Published: July 2009

The effect of irradiation damage on the mechanical properties of reactor pressure vessel structures is monitored in operating nuclear reactors according to the provisions of 10 CFR Part 50 Appendix H. In these surveillance programs Charpy V-notch energy and tensile data are collected. Trends in these data have and continue to be used to identify and quantify embrittlement trends, which is a key aspect to maintaining the continued structural integrity of the operating reactor fleet. This paper presents the current results from an on-going investigation aimed at assessing the effect of different curve-fitting strategies on the insights that can be gained from these data trending activities.
Numerical Simulation on Fast Filling of Hydrogen for Composite Storage Cylinders
Authors: Liu Yanlei , Zheng Jinyang , Xu Ping , Zhao Yongzhi , Li Lei , Liu Pengfei , Bie Haiyan , Chen Honggang , Huston Dryver , Sun Xiaoyan
Published: July 2009

Compressed gaseous hydrogen storage is widely used in hydrogen filling stations and fuel cell automobiles. As an effective filling is required to complete in a short time about 3 to 15 minutes, a high filling velocity is necessary which may lead to temperature rise and heterogeneity inside cylinders. In this research, a numerical method is proposed based on the finite volume method to investigate the thermal behavior such as temperature rise and distribution in the process of fast filling for the 15 L cylinder that designed by us. By numerical simulation, temperature, velocity and pressure distribution in the filling process are shown intuitively. Also, temperature and velocity gradients in the axial direction are obtained. In the simulation, parameters such as flow rates of fast filling are considered. Thus, curves of the temperature rise in the cases of different flow rates of fast filling are given. Furthermore, the limit of mass filling rate for the 15 L cylinder is obtained. The numerical results are compared with the existing experimental results and show high consistency.
Parametric CFD Analyses to Simulate Stratified Flows
Authors: Jeong Jae-Uk , Chang Yoon-Suk , Choi Jae-Boong , Kim Young-Jin , Kim Jin-Su , Lee Jin Ho , Chung Hae-Dong
Published: July 2009

CFD analysis is widely adopted for determination of design characteristics of major equipment with flexible geometry and flow. However, the accuracy of CFD analysis strongly depends on the numerical model, governing equation and simulation parameters. A thermal stratification phenomenon can lead to unanticipated damages of nuclear piping because of the stresses caused by different fluid densities due to stratified flow. In this paper, systematic CFD analyses are performed for surge line which is one of primary piping system by using representative commercial code to investigate key parameters; (1) mesh size and time step effect (2) turbulence parameter effect (3) material property approximation effect (4) conjugate heat transfer effect (5) insurge and outsurge flow effect. From numerical analysis results related to the items (1) through (3), the optimum CFD model as well as reasonable input parameters was determined. With regard to the item (4), thermal difference was bigger as 82∼208% than without considering conjugate heat transfer. On the other hand, for the item (5), stratified flows were come out clearer in outsurge flow. Based on the parametric CFD analyses to simulate stratified flows, most of numerical issues were resolved while further investigation is required for the conjugate heat transfer effect.
Contributions of Dr. Sumio Yukawa in the Development of Non-Ductile Failure Prevention Rules in the ASME Code
Authors: Mehta Hardayal S.
Published: July 2009

The objective of this paper is to review and highlight the contributions of Dr. Sumio Yukawa in the development of rules for the prevention of non-ductile failure in the ASME Boiler and Pressure Vessel Code. This includes review of his role in the development of WRC-175, Appendix G of Section III, the development of early flaw evaluation rules for carbon steel piping and in the review and evaluation of the toughness of austenitic stainless steels and nickel alloys after long-term elevated temperature exposures. The current status of these activities is briefly described.
Effect of Non-Conforming Thread Root Radius on the Fatigue Performance of Preloaded Threaded Fasteners
Authors: Nassar Sayed A. , Munn Brian S. , Yang X.
Published: July 2009

This paper presents an experimental investigation of the effect of thread root non-conformance on the fatigue performance of preloaded M12×1.5 Class 10.9 fasteners. Thread roots were dimensionally inspected using optical methods in accordance with the DIN 933 specification. Axial load fatigue tests were performed in accordance with ISO 3800. Each bolt specimen was subjected to a fixed stress amplitude of 42 MPa above and below various levels of the mean tensile stress. Test data was analyzed and fracture surfaces were examined using a Scanning Electron Microscope (SEM). Analysis of the test data showed that a non-conforming thread root radius had an adverse effect on the fatigue performance of preloaded threaded fasteners.
Influence of Stress Ratio on the Reliability of Al/Carbon-Fiber Composite Hydrogen Tanks
Authors: Bie Haiyan , Xu Ping , Zheng Jinyang , Kai Fangming , Liu Pengfei , Liu Yanlei , Chen Honggang
Published: July 2009

Hydrogen has been proposed as an important energy source for fuel cell vehicles (FCV), and the hydrogen storage tank is considered to be indispensable key equipment for hydrogen application. Currently, Al/carbon-fiber composite hydrogen tanks have become a study focus due to their outstanding properties such as high stiffness-density and strength-density ratio. In standards of this type hydrogen tanks, stress ratio is a primary parameter to be determined. Raising the stress ratio, the reliability will be increased; however, the hydrogen storage density will be decreased accordingly. In this paper, a study on the relationship between the stress ratio and reliability is carried out based on the static strength reliability analysis. The stress-strength interference model is used for the calculation of reliability. For fiber reinforced hydrogen tanks, the stress variance is mainly influenced by two classes of random variables, and the quantity of random variables is affected by the number of composite layers which depends on the stress ratio. So by analysis of stress distribution, the influence of stress ratio on the reliability is revealed, and a theoretical method is developed that can be used for selecting rational stress ratio in the standard.
Analytical and Numerical Study of the Interaction of a “Weak Shock Wave” and an Immersed Elastic Structure
Authors: Sigrist Jean-Franc¸ois
Published: July 2009

The present paper exposes an analytical and numerical study of the interaction between a weak shock wave and an immersed elastic structure. Several models of the problem are exposed and compared in the study: on the one hand, a semi-analytical method is proposed and on the other hand a numerical method is developed, based on a description of fluid-structure interaction through i) exact and ii) simplified interaction models. Comparison of the various approaches is exposed on a simple analytical case, namely that of a two-dimensional circular shell immersed in a two-dimensional compressible fluid and subjected to a spherical shock wave. Discussion of the various numerical results gives some practical guidance for the use of simplified fluid-structure interaction models in the context of industrial studies for underwater shock analysis.
Development of Fitness-for-Service Rules for the Assessment of HIC and SOHIC Damage in API 579-1/ASME FFS-1
Authors: Buchheim Gerrit M. , Osage David A. , Staats Jeremy C.
Published: July 2009

Hydrogen induced cracking (HIC) a form of wet H2S cracking that has a stepwise (through-thickness crack linkage) nature that is created from laminar (in-plane) cracking. HIC damage is initiated when atomic hydrogen generated by a surface corrosion reaction permeates into the steel and combines at non-metallic inclusions, discontinuities or other imperfections forming hydrogen molecules that are too large to diffuse out of steel. Lower strength carbon steel plate is most susceptible to hydrogen blistering and HIC damage, especially when exposed to an aqueous hydrogen environment, such as sulfide, cyanides, and hydrofluoric acid. This mechanism typically occurs below 300°F and is common in the oil and gas and petroleum refining industries, particularly with older high sulfur or phosphorous steels. HIC damage typically has not lead to breaches of containment in pressurized equipment. Stress oriented HIC (SOHIC), much rarer than HIC, is the arrangement of small HIC cracking perpendicular to the stress in a stacked array in stee and is typically located in the residual stress fields of welds. SOHIC has lead to breaches in containment. This paper focuses on the FFS assessment of HIC and SOHIC damage and describes some of the background to the new rules in Part 7 of API 579-1/ASME FFS-1 issued in 2007. It discusses laboratory tests undertaken on steel removed from a vessel that suffered severe HIC damage in service, but did not fail or leak. The results of those tests indicated that HIC damaged steel has considerable remaining strength and this lead to the new rules in API 579 Part 7. Because of the difficulty of ascertaining the presence of SOHIC when HIC damage is present, an additional crack-like flaw check is required to ensure that the HIC damaged area is acceptable as a crack-like flaw (Part 9) in addition to being acceptable as a locally thin area (Part 5) with a slight credit for remaining strength.
Analysis of Variable Amplitude Fatigue Data of the P355NL1 Steel Using the Effective Strain Damage Model
Authors: Pereira He´lder F. S. G. , DuQuesnay David L. , De Jesus Abi´lio M. P.
Published: July 2009

This paper proposes an analysis of variable amplitude fatigue data obtained for the P355NL1 steel, using a strain-based cumulative damage model. The fatigue data consist of constant and variable amplitude block loading which was applied to both smooth and notched specimens, previously published by the authors. The strain-based cumulative damage model is based on the growth and closure mechanisms of microcracks. It incorporates a parameter termed net effective strain range, which is a function of the microcrack-closure behaviour and inherent ability to resist fatigue damage. A simplified version of the model is considered which assumes crack closure at the lowest level for the entire spectrum and does not account for varying crack opening stresses. In general, the model produces conservative predictions within an accuracy range of two, for both smooth and notched geometries, demonstrating the robustness of the model.