bluuebunny/crossref_metadata_2025 huggingface dataset:
Damping of Steam Generator Tubes
Authors: Chandler Cordelia Kaye
Published: August 2008

The flow-induced vibration (FIV) analysis is one of the most critical evaluations needed to ensure long-term operation of steam generator. Original design hardware is usually based on experience and testing. Performance changes, such as an increase in power with increased feedwater flow, or tube repairs including plugs, stabilizers, and sleeves are best evaluated analytically. A critical factor in an FIV analysis is damping. The effective damping for a steam generator tube is a function of the structural support configuration, vibration amplitude, and the void fraction of the shell-side fluid. Numerous tests have been performed in the nuclear industry to determine appropriate damping ratios for the analysis of steam generator tube bundles. As part of the root-cause analysis for a recent plugged tube sever in a Once-Through Steam Generator (OTSG), Framatome ANP performed damping tests on tubes that were swollen due to internal pressure. Damping tests were performed on virgin tubes. The tests were repeated after swelling the tube into the tube supports and again after the swollen tube was severed and stabilized. Tests were performed with air or water inside the tubes. The tests results showed a damping ratio of 2% to 5% for both a virgin tube and a tube that was swollen and locked into the tubes supports. Stabilization increased the damping ratio to greater than 8%.
A New Tube/Support Impact Model for Heat Exchanger Tubes in Loose Supports
Authors: Hassan M. A. , Weaver D. S. , Dokainish M. A.
Published: August 2008

Heat exchanger tubes are usually loosely supported at intermediate points by plates or flat bars. Flow-induced vibrations result in fretting wear tube damage due to impacting and rubbing of tubes against their supports. Prediction of tube response relies on modelling the nonlinear tube/support interaction. The evaluated response is used to predict the resultant wear damage using experimentally measured wear coefficients. An accurate of prediction of impact forces and the work rate is therefore paramount. The analytical models available assume tube/support contact occurs over a single point. In this paper, a computational algorithm is proposed to describe the tube/support impact considering a finite support width. The new model provides a means of representing tube/support contact as a combination of edge and segmental contact. The proposed model utilizes a distributed contact stiffness to describe the segmental contact. The formulation also incorporates a stick/slip friction model. The model developed is utilized to simulate the dynamics of loosely-supported tubes.
On-Site Evaluation of Bending Stresses at the Root of Small Bore Pipes by Accelerometric Measurements
Authors: Guillou J. , Paulhiac L.
Published: August 2008

Several vibration-induced failures at the root of small bore piping systems occurred in French nuclear power plants in past years. The evaluation of the failure risk of the small bore pipes requires a fair estimation of the bending stress under operating conditions. As the use of strain gauges is too time-consuming in the environmental conditions of nuclear power plants, on-site acceleration measurements combined with numerical models are easier to handle. It still requires yet a large amount of updating work to estimate the stress in multi-span pipes with elbows and supports. The aim of the present study is to propose an alternate approach using two accelerometers to measure the local nozzle deflection, and an analytical expression of the bending stiffness of the nozzle on the main pipe. A first formulation is based on a static deformation assumption, thus allowing the use of a simple analog converter to get an estimation of the RMS value of the bending stress. To get more accurate results, a second method is based on an Euler Bernoulli deformation assumption: a spectral analyzer is then required to get an estimation of the spectrum of the bending stress. A better estimation of its RMS value is then obtained. An experimental validation of the methods based on strain gauges has been successfully performed.
Model Tests for Non-Eccentricity Dynamic Instability Closely Related to Folsom Dam Tainter-Gate Failure
Authors: Anami Keiko , Ishii Noriaki
Published: August 2008

This study addresses a type of dynamic instability of the Folsom Dam (radial) Tainter gates that is not related to the eccentric location of the trunnion pin about which the gate rotates. In these gates, the skinplate is concentric with the trunnion center (i.e., there is no eccentricity). This type of instability, hereafter called non-eccentricity dynamic instability, results from the coupling of the rigid body vibration of the whole gate about the trunnion pin with a streamwise bending vibration mode of the skinplate and is accompanied by a flow rate variation in the gate’s discharge. A two-dimensional 1/31-scale model of the Folsom Dam Tainter gates was constructed to verify the non-eccentricity dynamic instability mechanism. An intense instability was induced in the model, especially when the in-water natural frequency of the skinplate streamwise vibration was slightly smaller than the in-air frequency of the rigid body vibration of the whole gate about the trunnion pin. Further, a theoretical analysis of the non-eccentricity dynamic instability, in which the vibration-induced hydrodynamic forces are calculated by a potential theory developed by Rayleigh for dissipative wave radiation, was also conducted. The theoretical results are compared with the experimental results to verify the validity of the analytical model and its implicit assumptions.
A Theoretical Series Solution for the Dynamics of Finite-Length Bearings and Squeeze-Film Dampers
Authors: Antunes Jose´ , Moreira Miguel , Piteau Philippe
Published: August 2008

In this paper we develop a non-linear dynamical solution for finite length bearings and squeeze-film dampers based on a Spectral-Galerkin method. In this approach the gap-averaged pressure is approximated, in the lubrication Reynolds equation, by a truncated double Fourier series. The Galerkin method, applied over the residuals so obtained, generate a set of simultaneous algebraic equations for the time-dependent coefficients of the double Fourier series for the pressure. In order to assert the validity of our 2D–Spectral-Galerkin solution we present some preliminary comparative numerical simulations, which display satisfactory results up to eccentricities of about 0.9 of the reduced fluid gap H/R. The so-called long and short-bearing dynamical solutions of the Reynolds equation, reformulated in Cartesian coordinates, are also presented and compared with the corresponding classic solutions found on literature.
Seismic Proving Test of Ultimate Piping Strength: Simulation Analysis of Simplified Piping System Test
Authors: Suzuki Kenichi , Namita Y. , Abe H. , Ichihashi I. , Suzuki Kohei , Ishiwata M. , Fujiwaka T. , Tai K.
Published: August 2008

The six-year program for the Seismic Proving Test of Ultimate Piping Strength has been running since 1998 with the following objectives: i) to clarify the elasto-plastic response and ultimate strength of nuclear piping, ii) to ascertain the seismic safety margin of the current seismic design code for piping, and iii) to assess new allowable stress rules. To resolve outstanding technical issues before proceeding on to a seismic proving test of a large-scale piping system, a series of preliminary tests of materials, piping components and simplified piping systems is intended. A simulation analysis related to the simplified piping system test is described with a focus on the methodology of the non-linear dynamic response analysis of the whole piping system and the strain behavior of the localized critical elements, such as elbows and nozzles.
Seismic Behavior of Insulated Hollow-Brick Cavity Walls Using Quasi-Static Experimentation
Authors: Li Hong-Nan , Zhang Jing-Wei , Liu Li
Published: August 2008

This paper focuses on an experimental investigation of different types of insulated hollow-brick cavity walls, i.e. the walls with door opening, with window opening and without any opening characterized with different tie bar arrangements, subjected to slowly applied cyclic loads. The cracking and damage patterns, strength and stiffness degradation and deterioration, energy dissipation capacity and hysteretic feature were analyzed. Based on the experimental results, the formulas of calculation for cracking load and ultimate load of cavity wall are mathematically established. The evaluation equation of strength and stiffness degradation of walls is presented and its parameters are numerically given from regression results. The original characteristic curves of recovery force-displacement of actual specimens under cyclic loads were discussed, and then a standard recovery force-displacement model is suggested with convenient forms for implementation.
Shaking Table Tests of a Full Scale Steel Structure Isolated With MFPS
Authors: Tsai C. S. , Chiang Tsu-Cheng , Chen Bo-Jen
Published: August 2008

The base isolation, a kind of passive control technology, has been proved as a very efficient way to ensure the safety of a structure during severe earthquakes both from theoretical study and experimental effort. In general, the base isolation can be classified into two groups, which are sliding type and elastomeric type isolator. In this study, a new base isolator called as Multiple Friction Pendulum System (MFPS) has been proposed. The lubricant material, articulated slider and doubled concave sliding interfaces of MFPS are quite different from that proposed by V. Zayas in 1987. In this study, the MFPS isolator has been equipped beneath each column of a three-story structure at the National Center for Research on Earthquake Engineering to demonstrate its seismic resistance capability. The experimental results from shaking table tests of the 1940 El Centro, 1995 Kobe and 1999 Chi-Chi earthquake show that the proposed isolator can reduce the undesirable seismic response of the structure by lengthening the fundamental period of the structure during earthquakes. The experimental results indicate that the acceleration response of each floor can be lessened significantly as compared with those of the bare structure, and that the stress responses of structural components are limited in the certain range during severe earthquakes. Furthermore, the residual displacements of base isolators are negligible. Therefore, it is shown evidently that the proposed isolator can always bring the base-isolated structure to its initial position after an earthquake. Based on the previous observations, the proposed isolator can be adopted as an effective tool for upgrading the seismic resistibility of a structure. A finite element formulation for the MFPS is also proposed to simulate its mechanical behavior during earthquakes.
A Numerical Model of Lead Material and Its Applicability to Simulation of Isolation Devices
Authors: Matsuda Akihiro
Published: August 2008

This paper proposes a new numerical model of lead material to predict mechanical properties of isolation and vibration control devices using lead as damping material. Shear and tensile loading tests of lead were carried out to make the numerical model. Shear loading test specimen were constructed from a circumferential lead part welded at the top and bottom to steel flanges. Cyclic stress-strain relations in large strain region were obtained from shear loading test results. The elastic constants and the initial yield stress were given from tensile loading test results. Therefore a numerical model was made using both shear loading and tensile loading test results. Mechanical properties of lead dampers and isolated rubber bearings were simulated using the proposed numerical model via finite element method to show applicability of the model.
Verification of Lateral Force Distributions for Base-Isolated Structures Including Higher Mode Effects
Authors: Tsai C. S. , Chen Bo-Jen , Chiang Tsu-Cheng , Pong Wen-Shen
Published: August 2008

The seismic design of new seismically isolated structures is mainly governed by the Uniform Building Code (UBC-97). In the UBC code, the distribution formula of the inertial (or lateral) forces leads to a triangular shape in the vertical direction. It has been found to be too conservative for most isolated structures through experimental, computational and real earthquake examinations. In this paper, four simple and reasonable design formulae, based on the first mode of the base-isolated structures, for the lateral force distribution on isolated structures have been validated by a multiple-bay three-story base-isolated steel structure tested on the shaking table. Moreover, to obtain more accurate results for irregular structures in which higher mode contributions are more likely expected during earthquakes, another four inertial force distribution formulae are also proposed to include higher effects.
Seismic Design of Cylindrical and Spherical Storage Tanks According to API and Eurocode: A Difficult Merge in Design Philosophies
Authors: Botsis Ch. , Anagnostides G. , Kokavesis N.
Published: August 2008

Herein a comprehensive review and comparison of the parameters used in design of cylindrical tanks according to API 650 and Eurocodes is presented. API 650 is extensively used in many countries, including Greece, for the design of storage tanks. The European Community has developed a set of structural design codes named Eurocodes. They are the gathering and combination of existing design knowledge of many member states. Some of these codes are already mandatory in many member states, whereas others are still under discussion and improvement. The design of storage tanks is covered in the last editions of Eurocodes. It was found that the seismic design according to Eurocodes is more conservative that of API 650. As compared to API 650, the thickness of the first, second, and third courses of storage tanks needs to be increased by 15% or 20% on average, when the seismic design requirements of Eurocodes is used. Similarly the thickness of the bottom plate under the first course, must also be increased to comply with the seismic design requirements of Eurocodes. Most likely Eurocodes will be mandatory in the European Union, and therefore it is important to study and discuss the main differences between API 650 and Eurocodes. Undoubtedly API 650 is a historic and well-tested code. It has been applied in the design of storage tanks all over the world, however compliance with local and European laws is required to issue an installation license.
A Critical Comparison of Seismic Qualification Requirements for Safety Equipment in Various Codes and Standards
Authors: Abatt F. G. , Hossain Quazi , Meyer Milon
Published: August 2008

Evaluation of life safety risks to facility occupants, public, and the environment that may result from earthquake events involves both building structures and equipment supported from these structures. But, it is the seismic design of building structures that typically receive the bulk of the attention from the code committees of the national professional organizations and the regulatory authorities. For safety related equipment in nuclear facilities (e.g., Seismic Category I equipment in nuclear power plants and Seismic Performance Category 3 and 4 equipment in the Department of Energy facilities), the seismic design and analysis guidelines and acceptance criteria are well established. But, for Nonseismic Category equipment in nuclear power plants and Seismic Performance Category 1 and 2 equipment in Department of Energy facilities, these have not yet been developed to the same level of completeness and rigor. The code provisions and guidelines available today for these lower class/categories of equipment are briefly, but critically discussed here, along with a comparison of the results of the application of these code provisions.
A Direct Method on the Analysis of Steady State Residual Strains in a Steel Joint Subjected to Cyclic Loads
Authors: Yu C.-C. , Wu C.-T. , Chang T.-P.
Published: August 2008

The thick plate element of Mindlin type combined with the plane stress element is adopted in the direct analysis method. The numerical process is pretty efficient with respect to the speed of the calculation of final steady state residual quantities resulted in by cyclic loads. Two pipes joined together with a concentric reducer are employed as an example. The steady state residual stresses and strains after pipes experience cyclic anti-symmetric deflections, simulating the seismic loads, are illustrated. The diameter, the length of the pipes, and properties of material are the parameters evaluated in the present study. Results indicate that shorter pipe has higher shear effect at the fixed end. The ratio of pipe diameter to pipe length does not significantly differ the residual profile while the existence of a concentric reducer disturbs the distribution of the residual stresses and residual plastic strains dramatically.
Effect of Smear on Finite Strain Radial Consolidation for Vertical Prefabricated Drain in Compressible Soils
Authors: Chen Kuang-Hsiang , Hsieh Tzung-Hsun , Lin I.-Fan
Published: August 2008

Radial consolidation equations for vertical prefabricated drain which considers the effects of drain spacing, well resistance, extension of drain above the compressible soil layer, well peripheral smear effect, and differential vertical strain between smear zone and undisturbed zone at finite strain condition are derived. Particularly, the effects of smear as well as differential vertical strain between smear zone and undisturbed zone at finite strain condition are studied for the first time. Result indicating the time required for consolidation is increased by the effect of well peripheral smear. The effect of differential vertical strain can be ignored.
Investigation on Buckling Behavior of Cylindrical Liquid Storage Tanks Under Seismic Excitation: 1st Report — Investigation on Elephant Foot Bulge
Authors: Ito Tomohiro , Morita Hideyuki , Hamada Koji , Sugiyama Akihisa , Kawamoto Yoji , Ogo Hideyasu , Shirai Eiji
Published: August 2008

When a thin walled cylindrical liquid storage tank suffers a large seismic base excitation, buckling phenomena may be caused such as bending buckling at the bottom portion and shear buckling at the middle portion of the tank. However, the dynamic behaviors of the tanks is not fully clarified, especially those from the occurrence of buckling to some failures. In this study, bending buckling phenomena were focused which will be categorized as diamond buckling and elephant foot bulge. As ones of a series of studies, dynamic buckling tests were performed using large scale liquid storage tank models simulating thin walled cylindrical liquid storage tanks in nuclear power plants. The input seismic acceleration was increased until the elephant foot bulge occurred, and the vibrational behavior before and after buckling was investigated. In addition to the large scaled model tests, fundamental tests using small scaled tank models were also performed in order to clarify the effects of dynamic liquid pressure on the buckling threshold and deformation patterns.
Rocking Response of Unanchored Flat-Bottom Cylindrical Shell Tanks Subjected to Horizontal Base Excitation: Part II — Flexible Bottom Plate
Authors: Taniguchi Tomoyo
Published: August 2008

The mechanical analogy of the rock-translation interaction system of the tank is verified by comparing analytical results with experimental ones. To trace actual rocking behaviors of the tank, the existence of effective mass and moment inertia of liquid for a rocking motion, which is proportional to the uplift region of bottom plate, is assumed. The comparison of restoring moment defined by early investigators with overturning moment by proposed methods can identify the region of effective mass for a rocking motion in an iterative manner. Moreover, the base shear and uplift angle calculated agree with ones measured at previous shaking tests. These results corroborate the applicability of proposed methods. Finally, the sufficient friction to enter and sustain a rocking motion of the tank is discussed based on time histories of horizontal and vertical reaction forces on the pivoting edge.
Fragility Analysis for Highway Bridges in Taiwan
Authors: Liao W. I. , Loh C. H.
Published: August 2008

This paper presents the methods developed to enhance the transportation lifelines module in HAZ-Taiwan for highway bridges. The objective of this paper is to define the bridge classification and provide the fragility functions of the North-South Freeway in Taiwan that by utilize the available investigation data. The organization of this paper is as follows. First, the available bridge classes of the North-South freeway in Taiwan are reviewed and a new classification based on the available data to be implemented in Haz-Taiwan is proposed. Second, the description of failure mechanisms and criteria in different damage states adopted in this research are summarized. Third, the theoretical methodology of the fragility analysis for the new bridge classes and examples for freeway bridges are presented. Parameters for describing the fragility functions for each class are also generated and shown in the paper.
The French Approach Concerning the Protection of Shipping Casks Against Terrorism
Authors: Autrusson B. , Brochard D.
Published: August 2008

The behavior of the casks used to ship nuclear material must be assessed for various normal and accidental situations. Though not explicitly required in the French regulations, the security of the casks must be studied in the context of potential loads resulting from terrorism. The “Institut de Radioprotection et de Suˆrete´ Nucle´aire” (IRSN), in order to provide the authority with important data, has conducted analyses. These analyses have been based on both experiments and numerical models for various casks and loads. These studies have been carried out using an approach consistent with that used for facilities.
Survey of Packaging Requirements for the Transport of Highly Hazardous Materials
Authors: Leduc Daniel R. , Smith Allen C.
Published: August 2008

Among hazardous materials those which are most dangerous fall into three categories: chemical, biological, and radioactive. The DOT hazard classes for these three categories are Hazard Class 2.3 (poisonous gases) and 6.1 (toxic substances) for chemical hazards, Hazard Class 6.2 (infectious substances) for biological hazards and Hazard Class 7 for radioactive material (RAM) hazards. The packaging requirements for chemical and biological hazards are outlined and compared with RAM packaging requirements. RAM packages are found to be able to withstand much more severe performance tests than packages for other, more lethal hazards.
A Review of a Radioactive Material Shipping Container Including Design, Testing, Upgrading Compliance Program and Shipping Logistics
Authors: Celovsky Andrew , Lesco Randy , Gale Brian , Sypes Jeffrey
Published: August 2008

Ten years ago Atomic Energy of Canada developed a Type B(U)-85 shipping container for the global transport of highly radioactive materials. This paper reviews the development of the container, including a summary of the design requirements, a review of the selected materials and key design elements, and the results of the major qualification tests (drop testing, fire test, leak tightness testing, and shielding integrity tests). As a result of the testing, improvements to the structural, thermal and containment design were made. Such improvements, and reasons thereof, are noted. Also provided is a summary of the additional analysis work required to upgrade the package from a Type B(U) to a Type B(F), i.e. essentially upgrading the container to include fissile radioisotopes to the authorized radioactive contents list. Having a certified shipping container is only one aspect governing the global shipments of radioactive material. By necessity the shipment of radioactive material is a highly regulated environment. This paper also explores the experiences with other key aspects of radioactive shipments, including the service procedures used to maintain the container certification, the associated compliance program for radioactive material shipments, and the shipping logistics involved in the transport.
Proposal for Qualification of Gas-Generating Radioactive Payloads for Transportation Within a Type B Package: The Recombiner/Getter Approach
Authors: Houghtaling T. K. , Livingston R. R.
Published: August 2008

This paper proposes an alternative approach to qualifying gas generating radioactive payloads for shipment within Type B packaging through application of hydrogen recombiner/getter technology. This work compliments an earlier paper by the first author describing a direct measurement approach to achieving the same qualification goal. Specifically, this paper discusses another part of the success at the Savannah River Site in authorizing onsite transfer of legacy Radioactive Materials within the DDF-1 package. The current safety basis requires a measurement of storage can pressure and placement of a recombiner/getter product inside the package containment vessel to prevent significant hydrogen accumulation within the closed volume surrounding the storage can. These two actions are sufficient to ensure 1) that deflagration pressure of a potential flammable gas mixture is within Normal Conditions of Transport, and 2) the consequences of a detonation shock wave are within the Hypothetical Accident Conditions.
Post Fire Transient Temperature Distribution in Drum Type Packages in the Absence of Heat Generation
Authors: Smith Allen C.
Published: August 2008

This study investigates the temperature distribution in an idealized cylindrical package subjected to the HAC Fire transient, with no internal heat generation. Cases for overpack materials with thermal conductivity spanning two orders of magnitude are considered. The results show that the peak internal temperature is determined by the thermal conductivity of the overpack material, for this case. The thermal wave effect, where the interior temperature continues to rise after the end of the fire exposure, is present in all three of the test cases. For contents with no heat generation, the most desirable overpack materials would have low thermal conductivity and low heat storage capability. The study complements the parametric studies of effects of thermal properties on thermal response of packages which were previously reported.
A Life Extension Perspective for a CANDU-6 NPP: Gentilly-2 Plant Condition Assessments
Authors: Aubray Marc
Published: August 2008

To extend the life of a CANDU-6 reactor beyond its original design life requires a major investment for the replacement of reactor components (380 pressure and calandria tubes). After a preliminary technical and economical feasibility study [ref 1], Hydro-Quebec, owner of the Gentilly-2 NPP, has decided to perform a more detailed assessment to: • Get assurance that it is technically and economically viable to extend Gentilly-2 for another 20 years beyond the original design life; • Identify the detailed work to be done; • Define the overall cost and the general schedule; • Ensure an adequate licensing strategy to restart after refurbishment; • Complete all the Environmental Impact Studies required to obtain the government authorizations. Two processes have been used to assess the “health” of the station Systems, Structures and Components (SSCs): • The Plant Life Management Studies (PLIM) for approximately 10 critical SSC or families of SSC (PLIM Studies); • The Condition Assessment Studies for other SSC with a lower impact on the Plant production or safety. These two processes are briefly presented in the paper, as they were realized and applied at Gentilly-2 NPP.
Asset Management Evaluation Methods: The EDF Perspective
Authors: Bauby Catherine E. , Just Vale´ry E. , Garreau Caroline
Published: August 2008

The long term management of a production asset raises several major issues, among which rank the technical management of the plant, its economics and the fleet level perspective one has to adopt. Decision makers are therefore faced with the need to define long term policies (until the end of the life of the asset) which take into account multiple criteria including safety (which is paramount) and performance. In this paper we describe the French context where EDF (Electricite´ de France) is both Plant Owner and Operator of a fleet of 58 PWRs. We introduce a three-level methodology for asset management: the component / technical level (how to safely operate daily and invest for the future), the plant level (how to translate technical decisions into plant-wide consequences including economic performance) and the fleet level (how to manage a large number of similar assets). We then focus on the theoretical and practical links one can draw between the component level and the plant level. We describe several plant-wide indicators that are used to assess the value of the asset and we show how they can be inferred from the component-level technical and economic assessment (long-term equipment reliability, maintenance strategies, ...) by « rolling up » component level plans into a plant-wide decision process while taking into account the various sources of uncertainty associated with this assessment. We finally exemplify how this process could be applied to the life management of nuclear assets. To conclude, it appears asset management can be a major means for assessing and enhancing the long term value of a production unit while meeting everyday constraints.
Modeling Swirl Penetration and Thermal Cycling in Un-Isolable Branch Lines of PWR Plants
Authors: Keller Jeffrey D. , Bilanin Alan J. , Rosinski Stan T.
Published: August 2008

Thermal cycling has been identified as a mechanism that can potentially lead to fatigue cracking in un-isolable branch lines attached to pressurized water reactor (PWR) primary coolant piping. A significant research and development program has been undertaken to understand the mechanisms causing thermal cycling and to model the thermal-hydraulic boundary conditions for use in piping structural analysis. This paper will describe test and modeling results, with primary emphasis on predicting swirl penetration in dead-ended branch lines. Swirl penetration refers to the establishment of an axial vortical structure within the branch line, which has been shown previously to be the primary driving mechanism for thermal cycling in dead-ended branch lines. Swirl penetration testing was performed in a high Reynolds number facility to investigate the fluid dynamic and geometric parameters affecting the formation and decay of branch line swirl. Test results and preliminary model development are presented.
Data Analysis for RPV Life Management Studies of Temperature Measurements From Monitoring of Safety Injection Fluid in Spent Fuel Pit Cooling and Treatment System Storage Tank
Authors: Lagrange Dominique , Venturini Vincent , Bezdikian Georges , Salaun Jacques
Published: August 2008

In Electricite´ de France (EDF) probabilistic analyses for Reactor Pressure Vessel (RPV) life management, it’s used to take the temperature of the Reactor Spent fuel Pit cooling and treatment System storage tank as a constant. The aim of our study is to evaluate the stability of this temperature. Since 1999, we collected for several sites, and several nuclear plant units, the temperature of the Reactor Spent fuel Pit cooling and treatment System storage tank. Our results illustrate that this temperature depends on the season and the site. We first proposed to give a modelisation of this temperature dependent of the external temperature; even if this modelisation leads to a good R2, it’s not optimum. We also proposed to explain the temperature using the temperature of the essential service water or for lack the temperature of the river. In the case of extreme quantile study (meaning low temperature), we proposed to use the normal approximation, which seems to be conservative.
A Help for Lifetime Management: Defect Characterisation Through Signal Processing of Ultrasonic Data
Authors: Gautier S. , Cha˜tellier L.
Published: August 2008

So as to detect and to characterize potential defects on cast components, a program of in-service inspections is carried out for the non destructive evaluation (NDE) of the component. A first step of inspection usually consists in detection. Still, should a defect be detected, one can be asked to prove the component still stands the mechanical constraints. Then, it is very important to be able to characterize precisely the size and the position of the defect. In the studied case, we concentrate on planar defects which are to be characterized from ultrasonic. The estimation of the height consists in measuring the time of flight between two reflectors: the diffraction and the corner echoes. But in case of small defects, the two echoes overlap and it is very difficult to give a height estimation of the defect. Then, the core of our contribution deals with the signal processing method we developed for the estimation of the height of planar defects. The use of our so-called deconvolution process now enables to obtain the height estimation of the defect, when it was impossible without processing. Processing results of real data sets are given for the inspection of representative mock-ups of homogeneous austenitic welds. In conclusion, the signal processing step turns out to be an important step in the characterization of defects and can thus contributes to maintain some components in service.
Mechanical Stress Improvement Process (MSIP) Used to Prevent and Mitigate Primary Water Stress Corrosion Cracking (PWSCC) in Reactor Vessel Piping at V.C. Summer
Authors: Ray E. A. , Weir K. , Rice C. , Damico T.
Published: August 2008

During the October 2000 refueling outage at the V.C. Summer Nuclear Station, a leak was discovered in one of the three reactor vessel hot leg nozzle to pipe weld connections. The root cause of this leak was determined to be extensive weld repairs causing high tensile stresses throughout the pipe weld; leading to primary water stress corrosion cracking (PWSCC) of the Alloy 82/182 (Inconel). This nozzle was repaired and V.C. Summer began investigating other mitigative or repair techniques on the other nozzles. During the next refueling outage V.C. Summer took mitigative actions by applying the patented Mechanical Stress Improvement Process (MSIP) to the other hot legs. MSIP contracts the pipe on one side of the weldment, placing the inner region of the weld into compression. This is an effective means to prevent and mitigate PWSCC. Analyses were performed to determine the redistribution of residual stresses, amount of strain in the region of application, reactor coolant piping loads and stresses, and effect on equipment supports. In May 2002, using a newly designed 34-inch clamp, MSIP was successfully applied to the two hot-leg nozzle weldments. The pre- and post-MSIP NDE results were highly favorable. MSIP has been used extensively on piping in boiling water reactor (BWR) plants to successfully prevent and mitigate SCC. This includes Reactor Vessel nozzle piping over 30-inch diameter with 2.3-inch wall thickness similar in both size and materials to piping in pressurized water reactor (PWR) plants such as V.C. Summer. The application of MSIP at V.C. Summer was successfully completed and showed the process to be predictable with no significant changes in the overall operation of the plant. The pre- and post-nondestructive examination of the reactor vessel nozzle weldment showed no detrimental effects on the weldment due to the MSIP.
Managing the Aging of Critical Components of Nuclear Power Plants
Authors: Elder G. Gary , Llovet Ricardo , Meyer Theodore A. , Terek Edward
Published: August 2008

Managing the aging of critical components within a nuclear power plant is a challenging task that ultimately determines the value and revenue generation of the plant. This paper will provide an overview of a process for determining the critical components of the power plant and describe a long term equipment reliability and aging management program for these components. This program determines the potential failure modes and rates for each critical component part and identifies the effective repair, replacement, inspection, diagnostic, and maintenance activities. It also describes a tool for determining the optimal timing of these activities to produce the most value for the power plant. This program is currently being implemented at several power plants. As nuclear plants strive to reduce costs, extend life and maximize revenue, the aging management program and the supporting tools summarized in this paper can enable development of a long term, cost efficient plan to manage the aging of the plant’s critical components and systems.
Managing Aging Effects of the Passive Functions of Pressure Retaining Items
Authors: Sanwarwalla Mansoor H.
Published: August 2008

An overview of the known aging mechanisms is given in NUREG-6157 and NUREG 1801. These mechanisms are corrosion, corrosion assisted fatigue, embrittlement, mechanical wear and fretting, time dependent relaxation, low cycle fatigue and high cycle fatigue. A literature search was conducted and further analysis performed of these mechanisms. Large and small bore piping (including buried piping), valve bodies, pump casings, tanks, heat exchangers, accumulators and HVAC ducting were considered in this evaluation. Additional monitoring for aging effects on seismic resistance was mainly limited to small bore piping. A screening criterion for small bore piping is proposed for the two mechanisms (erosion/corrosion and low cycle fatigue) that will select a sample of the most critical locations for baseline examination. The sampling techniques are considered adequate due to the large design margin inherent in small bore piping. This paper will demonstrate that the two mechanisms, which need to be managed during life extension due to their significant effect on passive functions, are erosion/corrosion and low cycle fatigue.
The Effect of Piping System Non-Linearity on the Unstable Growth of a Circumferential Through-Wall Crack in a Pipe
Authors: Smith E.
Published: August 2008

During the last twenty-five years, considerable attention has been given to the structural integrity of steel piping systems, and in particular to the effect of circumferential cracks on their integrity. From a safety perspective, it is important that any crack, say for example a stress corrosion crack or fatigue crack, will not develop into a through-wall crack which will then propagate unstably, thus leading to a guillotine rupture and possibly a pipe whip scenario. One way of guaranteeing that this does not happen is to ensure that unstable growth of a circumferential through-wall crack is unable to occur. An appropriate methodology is based on tearing modulus concepts with the instability criterion being expressed in the form TAPP > TMAT where TAPP is the applied tearing modulus, a measure of the crack driving force, and TMAT is the material tearing modulus, a measure of the material’s crack growth resistance. With a piping system that behaves in a linear elastic manner, TAPP involves only the system’s geometry parameters and the crack size but not the magnitudes of the applied loadings or the material properties of the cracked cross-section; the behaviours of the cracked cross-section and the remainder of the piping system are therefore decoupled. If, however, the system behaves in a non-linear manner say, for example, as a result of excessive deformation arising as a consequence of large deformations, then TAPP also involves the material properties of the cracked cross-section; material and piping system geometry parameters are then not decoupled in the instability criterion. The paper illustrates this point by analysing a simple model system where the non-linearity arises from excessive deformation at a connection.
Finite Element Modeling of Blade-Rotor System in Turbomachinery
Authors: Al-Nassar Y. N.
Published: August 2008

Modal analysis of Blade-Disk system under the effect of bearing flexibility is investigated. The present study has considered soft to hard bearing flexibility. The main objective here is to read from the modal analysis results the frequencies that are carrying out some information on blade vibration. The modal analysis shows that that there are few frequencies that are changing with the change of bearing flexibility. These are shaft mode only, shaft-blade nature, disk mode only, disk-blade nature, and blade mode only. The shaft-blade modes are the ones of concern here.
Deformation Relaxation: A Finite Element Optimization Method for Tee
Authors: Wang Shijun , Zhao Jinjuan
Published: August 2008

In this paper, an optimization method for a tee is proposed. A finite element model of a tee is built and its reliability is demonstrated by experiment. Analysis models with different fillet radii are analyzed. The results show that changing the fillet radius can not distinctly change the stress intensity level. Therefore, deformation relaxation, a new method for lowering stress intensity levels, is proposed. The method’s core is that the original shape of the tee is modified by deformation displacement. The computed results show that the decrease of the highest stress intensity is approximately in proportion to the modification. When the maximum modification is up to 10% of the outer diameter of main pipe, the highest stress intensity decreases by 52.2%.
CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions for Lifetime Evaluation: Methodology and Qualification Task of the EDF Numerical Tools
Authors: Martin A. , Beaud F. , Ternon Morin F. , Veneau T.
Published: August 2008

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behaviour of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA transients. This paper presents the Research and Development program started at E.D.F on the CFD determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermalhydraulic tools N3S and Code_Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. We first explain the recent improvement of the thermalhydraulic analysis with the global definition of the SBLOCA transient and the local analysis in the downcomer. Then, the qualification task of the EDF numerical tools is described. In order to reach this purpose, we have investigated several configurations related to an injection of cold water and focused our analysis particularly in the cold leg but also in a the downcomer. Two experiment test cases have been studied. A comparison between experiment and numerical results in terms of temperature field is presented. On the whole, the main purpose of the numerical thermalhydraulic studies is to accurately estimate the distribution of fluid temperature in the downcomer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margins.
Fatigue Crack Initiation in Pressure Vessel Steels Using a Distance Parameter
Authors: Chattopadhyay Somnath
Published: August 2008

The propagation behavior of short cracks cannot be studied by linear elastic methods because of large plastic region near the crack tip, as well as a breakdown in correlation of the stress intensity factor with the fatigue crack growth rates. The proposed fatigue design approach incorporates a distance parameter in conjunction with linear elastic fracture mechanics and effectively integrates long and short crack growth test data. This distance parameter is a material constant that allows for the effects of (a) large-scale plasticity, (b) crack closure and (c) fatigue crack threshold. Furthermore, this parameter can be used to successfully predict fatigue crack growth behavior of short cracks. The practical application of this method to study fatigue crack initiation in pressure vessels rests on the concept that initiation occurs only when the material ahead of the crack tip is damaged enough by cyclic straining.
Considerations in the Design and Analysis of an ASME Section VIII, Div. 2 Reactor Support Skirt
Authors: Williams Dennis K. , Seipp Trevor G.
Published: August 2008

This paper describes the considerations employed in the finite element analysis of a relatively “short” support skirt on a hydrocarbon reactor vessel. The analysis is accomplished in accordance with ASME B&PV Code, Section VIII, Division 2 alternate rules in conjunction with the guidelines outlined in WRC Bulletin 429. This provides a sound basis for the classification of the calculated stress intensities. The support skirt is capable of sustaining the deadweight load in addition to resisting the effects of thermal displacements, wind loadings, overturning moments from external piping loads on the attached hydrocarbon reactor vessel, and friction between the skirt base plate and concrete foundation. The displacement and thermal boundary conditions are well defined and discussed in detail. The effects of multiple scenarios for the displacement boundary conditions are examined. The skirt design also employs a hot-box arrangement whereby the primary mode of heat transfer is by radiation. A discussion of the two-part analysis is included and details the interaction between the heat transfer analysis and the subsequent structural analysis. The heat transfer finite element analysis is utilized to determine the temperatures throughout the bottom of the vessel shell and head, as well as the integrally attached support skirt. Of prime importance during the analysis is the axial thermal gradient present in the skirt from the base plate up to and slightly beyond the skirt-to-shell junction. While the geometry of the subject vessel and skirt is best described as axisymmetric, the imposed loadings are a mixture of axisymmetric and non-axisymmetric. This combination lends itself to the judicious selection and utilization of the harmonic finite element and properly chosen Fourier series representation of the applied loads. Comparison of the thermally induced axial stress gradient results from the FEA to those obtained by the closed form beam-on-elastic-foundation are also tendered and discussed. Finally, recommendations are included for the design and analysis of critical support skirts for large, heavy-wall vessels.
Design and Remnant Life Analysis of a Carbon Steel Containment Jacket Operating in the Creep Range
Authors: Mair David
Published: August 2008

During routine inspection of a 35 year old steam generating plant, a large surface crack was found at a critical tee intersection. With the crack appearing to be close to a condition of sudden rupture, a probabilistic technique was used to assist in determining the likelihood of failure. This paper describes this technique and the design of a carbon steel containment jacket used to enclose the cracked area. The design of the jacket had to take into account its creep behaviour at elevated temperature. The advantage of this repair method was that it was able to be installed quickly and without having to completely de-pressure the steam system. It was later decided that the operating life of the jacket should be extended to defer a planned shutdown. A simplified remnant life analysis was then undertaken as detailed in this paper. Taking into account the operating history of the jacket, it demonstrated that the life of the jacket could be safely extended as required.