bluuebunny/crossref_metadata_2025 huggingface dataset:
Critical Soil-Structure Interaction Analysis Considerations for Seismic Qualification of Safety Equipment
Authors: Hossain Quazi A.
Published: August 2008

While developing seismic analysis models for buildings that support safety-related equipment, a number of issues should be considered to ensure that the input motions for performing seismic qualification of safety-related equipment are properly defined. These considerations are listed and discussed here with special attention to the effect and importance of the interaction among the foundation soil, the building structure, the equipment anchors, and the equipment structure. Typical industry practices are critically examined to assess their adequacy for determining the input motions for equipment seismic qualification. The features that are considered essential in a soil-structure interaction (SSI) model are described. Also, the effects of inappropriate treatment or representation of these features are discussed.
Seismic Capacity of Threaded, Brazed and Grooved Clamped Joints
Authors: Antaki George
Published: August 2008

A series of static and shake table tests were conducted on threaded, brazed and mechanical pipe joints, commonly used in fire protection systems, to understand their integrity under extreme lateral loads, of the type that would be expected in large earthquakes. This paper presents the measured loads and deflections of the joints up to the point of failure. It also describes the joints’ static and dynamic failure mode. This information can be used to model the joints’ flexibility under large lateral displacements, determine their ultimate capacity, and help understand their leak and rupture characteristics.
Study on 3-D Measurement Method in Shake Table Tests Using Image Processing Technique: 1st Report - Verification of Synthetic Performance of the Measurement System
Authors: Fujita Satoshi , Furuya Osamu , Niitsu Yasushi
Published: August 2008

This study has been doing research and development of 3-dimensional measurement method using the image processing technique for the measurement of dynamic displacement in shake table test. This measurement system is a very convenient system because it can measure the three-dimensional dynamic displacement in shake table test in which the several makers are only attached to the surface of an experimental structure. The system therefore is the most suitable measurement system for an evaluation of complex 3-dimensional behavior of experimental model. Fundamental hardware and software for the measurement system have been constructed until now. The fundamental dynamic measurement accuracy and effectiveness of the measurement system have been also confirmed from several shake table tests. This paper describes the dynamic measurement accuracy in shake table test in with the experimental model has several rotational behaviors, and also summarizes the synthetic performance of the proposed measurement system.
Study on 3-D Measurement Method in Shake Table Tests Using Image Processing Technique: 2nd Report - Measurement Accuracy in Actual Fracturing Test Using Full Scale Wooden House Model
Authors: Fujita Satoshi , Furuya Osamu , Mikoshiba Tadashi
Published: August 2008

This study has been examined about a three-dimensional measurement method that can measure a position and an attitude of test object without any contact. Fundamental hardware and software in measurement system has been constructed until now. Moreover, the fundamental dynamic measurement accuracy and effectiveness of the proposed measurement system has been confirmed from the several shake table tests. This paper describes the effectiveness of the proposed system for the measurement in fracturing test using real scale wooden house model.
French PWR Nuclear Power Plants Steam Generators and Reactor Pressure Vessel Heads Aging Management Criteria and Replacement Strategy
Authors: Bezdikian Georges
Published: August 2008

The approach used by the French utility, concerning the Aging Management system of the Steam Generators (SG) and Reactor Pressure Vessel Heads, applied on 58 PWR NPPs, involves the verification of the integrity of the component and the Life Management of each plant to guarantee in the first step the design life management and in the second step to prepare long term life time in operation, taking into account the degradation of Alloy 600 material and the replacement of these materials by components made with Alloy 690. The financial stakes associated with maintaining the lifetime of nuclear power stations are very high; thus, if their lifetime is shortened by about ten years, dismantling and renewal would be brought forward which would increase their costs by several tens of billions of Euros. The main objectives are: • to maintain current operating performances (safety, availability, costs, security, environment) in the long term, and possibly improve on some aspects; • wherever possible, to operate the units throughout their design lifetime, 40 years, and even more if possible. This paper shows the program to follow the aging evaluation with application of specific criteria for SG and for Vessel Heads, and the replacement of the Steam Generators and Vessel Heads at the best period. The strategy of Steam Generators Replacement are developed and Vessel Head program of monitoring and replacement are detailed.
Aging Management Program: Comparisons Between EDF Methodology and U.S. GALL Report
Authors: Faidy Claude
Published: August 2008

EDF has developed a special methodology to review systematically the ageing management programs of all the safety important components. After a general presentation of the methodology on the different aspects (component selection, degradation mechanism sensitivity, detailed complementary analysis), the paper will make some comparison with the US procedure developed and presented in the GALL report. The objective of the comparison is to compare and discuss the major differences between EDF and US utility approaches on the mechanical components. The major differences are connected to: • different regulation request; • different code rules; • different existing management program; • different appreciation on the potentiality of some degradation mechanism. The paper will analyse in detail some specific examples, including current maintenance program, preventive maintenance and anticipation program, damage assessment and ISI program, R&D needs...
Time Limited Aging Analyses Requirements for License Renewal for U.S. Nuclear Power Plants
Authors: Sanwarwalla Mansoor H.
Published: August 2008

One of the requirements for license renewal for US nuclear plants stated in the United States Nuclear Regulatory Commission (USNRC) regulations in the License Renewal Rule (LRR) 10CFR Part 54 (Ref. 1) is the identification and updating of Time-Limited Aging Analyses (TLAA). During the design phase for a plant, certain assumptions about the length of time the plant would be operated were made and incorporated into design calculations for several of the plant’s systems, structures and components (SSCs). Examples of TLAAs are analyses of metal fatigue, environmental qualification (EQ) of electric equipment, etc. For a renewed license, these calculations have to be reviewed to verify that they remain valid for the period of extended operation. However, the LRR does allow TLAA-associated aging effects to be managed by an aging management program. This paper discusses the USNRC regulatory requirements for TLAAs and the industry’s response for addressing the TLAAs. It also discusses the issues regarding the generic set of TLAAs that have been identified by the NRC in NUREG-1801 (Ref. 2), and how these have been addressed by all the plants that have received their renewed license. The paper also identifies certain plant specific TLAAs.
License Renewal Program for Entergy’s Fleet of Nuclear Power Plants
Authors: Young Garry G.
Published: August 2008

License renewal of operating nuclear power plants in the United States has become one of the most successful nuclear industry activities in the past few years. Entergy, the second largest nuclear plant operator in the U.S., was one of the pioneers in this new process. In 2000, Entergy submitted a license renewal application to the Nuclear Regulatory Commission (NRC) for Arkansas Nuclear One, Unit 1. By June 2001, less than 17 months later, the NRC issued a renewed license. Due in part to the efficiency and success of this first Entergy license renewal project, a dedicated team of Entergy and Areva (formerly known as Framatome-ANP) personnel was established using virtual office concepts to work on license renewal for the remaining nine Entergy nuclear units over the next decade. Since each license renewal project takes 4 to 5 years and costs $10 to $15 million to complete [1], the dedicated team has focused on improving the schedule and economics of the license renewal process. By early 2004, the dedicated team has worked on five license renewal projects and expects to begin work on at least two additional projects by 2005. The virtual team organization has established standardized processes for managing data and for performing aging management reviews, environmental reviews, and time limited aging analyses evaluations. In addition, the team has worked with the Nuclear Energy Institute (NEI) and the NRC to further improve the efficiency of the license renewal process. This paper discusses the standardized processes established, the virtual team techniques used to manage multiple license renewal projects, and the plans for further process improvements. The ultimate goal of Entergy’s license renewal work is to achieve a highly efficient and effective license renewal process that ensures the safe continued operation of its fleet of nuclear power plants for decades to come.
Lifetime Management of EDF PWR Pressure Vessels and Pipings
Authors: Hedin F. , Legendre J. C.
Published: August 2008

Lifetime management of EDF PWR vessels and pipings are one of the main technical key points of safety and competitivness. This paper describes the EDF global approach in this field, which is applied to the nuclear fleet i.e 58 nuclear power plants, and particularly to the first 34 three loops, as far as lifetime is concerned: • operating procedures and routine maintenance, special maintenance and ten years safety reassessment, • engineering analysis, based on feed back experience, scientific knowledge, degradations mechanisms, causes and consequences management, • operating loadings decrease, • complementary deterministic and cost-benefit analysis, • fit for service justifications, • anticipation strategy to prepare future, based on Non Destructive Testing investigations, ability to repair and/or to replace components, in situ expertises, ... Some examples are given: lifetime management of reactor vessels heads and bottom penetrations of pressure vessels, fit for service of cast stainless steel primary pipings, primary nozzles and auxiliary pipings special maintenance.
Reactor Pressure Vessel of Tricastin Unit 1 Core Zone Inspection: Comparison of Non-Destructive Examination Results Among Two Core Area Inspections Performed in 1999 and in 2003
Authors: Payraudeau Karl , Zamoum Karim , Pasquier Thierry
Published: August 2008

As part of the unit aging follow-up, a non.destructive examination has been designed to inspect the core zone of the Reactor Pressure Vessel (RPV). This ultrasonic process called “VPM” has been developed and qualified by Intercontroˆle in accordance with EDF specifications. The qualification has been attested by an accredited qualification body. In 1999, the VPM process was used the first time during the in-service inspection of TRICASTIN, unit 1. As a result of the RPV core zone inspection a set of under cladding flaw indications was detected and characterized with the VPM process. These flaws have been analysed as planar manufacturing flaws. In 2003, 4 cycles later, during the TRICASTIN Unit 1 outage inspection, “VPM” was used to characterize again this set of flaw indications in order to verify their dimensions. The changes of the main characteristics (specially height) between both the inspections were compared to the process accuracy. No significant dimension change has been recorded.
Molecular Dynamics Simulation of the Effect of Large Angle and Small Angle Grain Boundaries on the Near-Threshold Fatigue Crack Growth in Iron
Authors: Kubo Shiro , Misaki Masaki
Published: August 2008

In the near-threshold region of the fatigue crack growth in metals, the amount of crack growth per cycle is in the order of the atomic scale. This may suggest that the near-threshold fatigue crack growth is controlled by atomic scale events. In previous papers the present authors applied the molecular dynamics to the simulation of near-threshold fatigue crack growth. In this study, the molecular dynamics method was applied to investigate the influence of inclined grain boundary on the near-threshold mode I fatigue crack growth behavior in BCC iron. Large angle grain boundaries served as an obstacle to the fatigue crack growth. Small angle grain boundaries had the effect of decreasing crack growth when the crack approached the grain boundary. When the crack reached the grain boundary it was trapped by the grain boundary.
Fatigue Crack Modeling and Simulation Based on Continuum Damage Mechanics
Authors: Takagaki Masakazu , Nakamura Toshiya
Published: August 2008

Numerical simulation of fatigue crack propagation based on fracture mechanics and conventional finite element method requires huge amount of computational resources when cracked structure is subjected to complicated condition such as the cases of multiple site damage or thermal fatigue. The objective of the present study is to resolve this difficulty by employing the continuum damage mechanics (CDM). An anisotropic damage variable is defined to model a macroscopic fatigue crack and its validity is examined by comparing the stress distributions around the crack with those obtained by an ordinary fracture mechanics method. Together with the assumptions on crack opening/closing and damage evolution, numerical simulations are conducted for low cycle fatigue crack propagation behaviors in a plate with single and two cracks. The results show good agreement with the experiments. Finally, propagations of multiply distributed cracks under low cycle fatigue loading are simulated to demonstrate the potential applicability of the present method.
An Approach for Cost Effective Assessment in Risk-Based Maintenance as a Life-Cycle Maintenance (LCM) Model
Authors: Yatomi Masataka , Fuji Akio , Saito Noriko , Yoshida Toshiaki
Published: August 2008

For aged power plants in Japan, the life extension with retaining the safety and cost-effective beyond the original design lifetime is proposed. Therefore it is important to minimise the risk and maintenance cost to keep operating the plants. Life-Cycle Maintenance (LCM) is proposed for optimising maintenance plan with reliability in the life of the plants. Risk Based Maintenance (RBM) is included in the LCM to assess the risk of components in the plants. LCC and the investment assessment may be also conducted to decide the most cost effective maintenance strategy, if several maintenance strategies are proposed in RBM. In this paper, concept and an application of the LCM are described to optimise maintenance plan in the lifetime of a plant. It was found that the LCM is quite useful method to plan the most cost effective maintenance strategies in the lifetime of the plant.
Modeling Gas-Liquid Head Performance of Electrical Submersible Pumps
Authors: Sun Datong , Prado Mauricio
Published: August 2008

This study presents a new gas-liquid model to predict Electrical Submersible Pumps (ESP) head performance. The newly derived approach based on gas-liquid momentum equations along pump channels has improved the Sachdeva model [1, 2] in the petroleum industry and generalized the Minemura model [3] in the nuclear industry. The new two-phase model includes novel approaches for wall frictional losses for each phase using a gas-liquid stratified assumption and existing correlations, a new shock loss model incorporating rotational speeds, a new correlation for drag coefficient and interfacial characteristic length effects by fitting the model results with experimental data, and an algorithm to solve the model equations. The model can predict pressure and void fraction distributions along impellers and diffusers in addition to the pump head performance curve under different fluid properties, pump intake conditions, and rotational speeds.
Theoretical Investigation on Flow Rate at the Maximum Efficiency Point in the Design of Impeller Blade in Centrifugal Pump
Authors: Tanaka Takaharu
Published: August 2008

This paper presents a theoretical investigation of the flow rate at the maximum efficiency point in the design of impeller blade in centrifugal pump. An energy balance was performed at the trailing edge of impeller outlet in the rotating flow passage of centrifugal pump. The evaluation shows that, when the fluid particles straight forward tangential velocity is one third of the impeller blade’s peripheral velocity and the fluid particles circular forward tangential velocity is two third of the impeller blade’s peripheral velocity at the trailing edge of the impeller outlet, the maximum hydraulic energy output, that is, the maximum efficiency point is obtained.
Plant Asset Management: RIMAP (Risk-Based Inspection and Maintenance for European Industries) - The European Approach
Authors: Kauer Robert , Jovanovic Aleksandar , Angelsen Sture , Vage Gjermund
Published: August 2008

The paper presents an overview of the European project RIMAP (Risk-Based Inspection and Maintenance for European Industries) as well as a more detailed insight into its application for the power industry (RIMAP Power Workbook). RIMAP is partly financed by the European Commission for the “Growth Programme, Research Project”; Contract Number GIRD-CT-2001-03008.
High Integrity Protection Systems
Authors: Summers Angela E. , Zachary Bryan A.
Published: August 2008

The American Petroleum Institute (API) and American Society of Mechanical Engineers (ASME) provide criteria for the protection of vessels and pipeline from excess pressure. In conventional design, a pressure relief device is used as the primary means of protection and a flare is used to safely combust the gases relieved during any overpressure event. However, community and regulatory authorities are demanding that industry reduce flaring. The need to balance safety and environmental requirements has resulted in the use of an alternative approach for overpressure protection — a safety instrumented system (SIS). API 521 and Code Case 2211 of ASME Section VIII, Division 1 and 2 allow the use of an SIS in lieu of a pressure relief device as long as the SIS meets or exceeds the protection provided by the pressure relief device. As SIS, their design must follow the safety lifecycle in the international standard IEC 61511. The required risk reduction results in the need for high SIS safety availability; therefore, these systems are often called High Integrity Protections Systems (HIPS).
A Numerical Simulation of Thermal Striping in an Upper Plenum of a Liquid Metal Reactor
Authors: Choi Seok-Ki , Kim Eui-Kwang , Wi Myung-Hwan , Jeon Won-Dae , Kim Seong-O
Published: August 2008

A computational study of thermal striping in an upper plenum of KALIMER (Korea Advanced Liquid Metal Reactor) is performed. The primary objective of the present study is to obtain the distributions of the amplitude and frequency of a temperature fluctuation in the upper plenum of KALIMER. Two different treatments of the turbulent flow are considered, one is the RANS (Reynolds Averaged Navier-Stokes equation) approach and the other is the LES (Large Eddy Simulation) approach. In the RANS approach, the computation is performed using the CFX-4.2 code with the differential stress and flux turbulence model. In this computation the overall distributions of the velocity vector, temperature and temperature variance are obtained. In the LES the amplitude and frequency of the temperature fluctuation are obtained using the CFX-5.6 code. From the results of the RANS approach, the locations where thermal striping is mattered are obtained. It is found that in the KALIMER design this region is the edge of the UIS (Upper Internal Structure) bottom and the radial and circumferential distributions of the temperature fluctuation in this region are investigated. From the LES results the amplitude and frequency of the temperature fluctuation are obtained for various locations in the upper plenum of KALIMER. These results will be used for the calculation of the thermal stress caused by thermal striping.
Numerical Simualtion of Turbulence Mixing Characteristic in Various Secondary Flow Conditions at T-Junction Piping Systems
Authors: Tanaka Masa-aki , Muramatsu Toshiharu
Published: August 2008

Temperature fluctuation caused by mixing the fluids with different temperature in a T-junction pipe gives eventually thermal fatigue to structure, and this phenomenon is significant as safety issue in liquid metal cooled fast reactor (LMFBR). In Japan Nuclear Cycle Development Institute (JNC), experimental and numerical investigations have been performed to clarify the mixing phenomena in the T-junction pipe and to establish an evaluation rule for design. If the T-junction pipe is set near an elbow pipe, turbulence mixing is surly affected by the secondary flow generated in the elbow pipe and it is necessary to study the influence of the secondary flow on the temperature fluctuation in the T-junction pipe. We carried out investigation into the secondary flow effect by numerical simulation using a quasi-direct numerical simulation code. Numerical simulation is conducted on the existing experiment, in which the test section simulated the T-junction pipe with the elbow pipe in LMFBR. Major parameter in the numerical simulation is the flow direction of the branch pipe to the flow direction of the elbow pipe. We discuss the influences of the secondary flow on turbulent mixing behavior, and also clarify the mixing mechanism in T-junction pipe.
Modelling Two-Phase Flow in Pipe Bends
Authors: Supa-Amornkul S. , Steward F. R. , Lister D. H.
Published: August 2008

In order to have a better understanding of the interaction between the two-phase steam-water coolant in the outlet feeder pipes of the primary heat transport system of some CANDU reactors and the piping material, themalhydraulic modelling is being performed with a commercial CFD (computational fluid dynamics) code — Fluent 6.1. The modelling has attempted to describe the results of flow visualization experiments performed in a transparent feeder pipe with air-water mixtures at temperatures below 55°C. The CFD code solves two sets of transport equations — one for each phase. Both phases are first treated separately as homogeneous. Coupling is achieved through pressure and interphase exchange coefficients. A symmetric drag model is employed to describe the interaction between the phases. The geometry and flow regime of interest are a 73° bend in a 5.9 cm-diameter pipe containing water with a Reynolds number of ∼105–106. The modelling predicted single-phase pressure drop and flow accurately. For two-phase flow with an air voidage of 5%–50%, the pressure drop measurements were less well predicted. Furthermore, the observation that an air-water mixture tended to flow toward the outside of the bend while a single-phase liquid layer developed at the inside of the bend was not predicted. The CFD modelling requires further development for this type of geometry with two-phase flow of high voidage.
A Study of Flow-Induced Grid-to-Rod Fretting Wear in PWR Fuel Assemblies
Authors: Kim Kyu-Tae , Jang Young-Ki , Choi Joon-Hyung , Lee Sang-Soon
Published: August 2008

It is generally known that flow-induced vibration at the nuclear fuel assembly structures may cause grid-to-rod fretting wear and subsequently rod failure. The flow-induced grid-to-rod fretting wear is found to initiate at a certain critical grid-to-rod gap that is strongly correlated with the extent of flow-induced vibration. In this paper, three vibration drivers acting on the grid-to-rod fretting wear are proposed, based on various fretting wear experience in commercial reactors as well as on various flow-induced fretting wear test results. The first active vibration driver is high turbulence-induced excessive fuel rod vibration with the combination of excessive grid-to-rod gap. The second active vibration driver is self-excited fuel assembly vibration in a low frequency range caused by hydraulically unbalanced mixing vanes of the spacer grid assembly. The third active vibration driver is self-excited spacer grid strap vibration in quite a high frequency range caused by some spacer grid designs. Each vibration driver on the grid-to-rod fretting wear damage is discussed. On the other hand, simplified three formulas for the grid-to-rod fretting wear progress such as constant work rate, constant stress rate and linear stress rate have been derived and used in predicting the rod failure time. It can be said that the linear stress rate is the most effective in predicting the rod failure time, while the constant work rate the least effective. In addition, spacer grids with the larger grid-to-rod contact area seem to generate the less fretting wear rate.
Stratified Flow Experiments Pertaining to Advanced LWR Reactor Passive Safety System Designs
Authors: Liou Jim C. P. , Stephens Alan G. , Schultz Richard R.
Published: August 2008

During a loss-of-coolant-accident in advanced light water reactors, outside coolant enters the cold leg by gravity to cool the core. This coolant is at a substantially lower temperature and thus is heavier than the liquid in and from the reactor. Consequently, stratified flow may occur. A stratified flow may cause condensation-induced water hammer, and will influence the coolant flow behavior. Two sets of experiments are in progress to better understand stratified flow conditions that lead to water hammer, and the density stratification behavior. The first set uses air-oil-water as the test media. Its purposes are to conduct exploratory tests and to provide instruction an apparatus for education purposes. The second set of tests will use steam and water and, later, the refrigerant R123. This paper describes the exploratory test facility, gives a brief description of the facility that will be used for the steam-water and refrigerant tests, describes the overall test plan, and finally gives some preliminary results on the intrusion of a lighter liquid into a pipe against flow.
A RELAP5/MOD3 Model for the Estimation of Air Pull-Through for a Draining Tank
Authors: Lucas D. S. , Riemke R.
Published: August 2008

An air pull-through vortex model was implemented in RELAP5/MOD3 to investigate air-entrainment from a draining tank. Surge tanks covered by an air bubble are vulnerable to the entrainment of air into the liquid. If such a system is connected to a pump the possibility of entrained air from draining resulting in cavitation and pump degradation is possible.
The Sensibility of Rocking Motion of Flat-Bottom Cylindrical Shell Tanks to Sloshing Motion
Authors: Taniguchi Tomoyo
Published: August 2008

This paper examines the sensibility of rock motion of flat-bottom cylindrical shell tanks to the lateral force induced by the sloshing motion. Since the natural period of sloshing motion of the tank may be close to that of rocking motion of the tank, their resonance may result the serious damages of the storage systems. Therefore, it is necessary to examine their fundamental behavior beforehand. Using horizontal sinusoidal ground motion whose period matches the natural period of 1st sloshing mode of the tank, the rocking motion of the tank is numerically examined. The results imply that the sloshing behavior has the potential to make tanks rock.
Discrete Element Method for Modeling Penetration
Authors: Tavarez Federico A. , Plesha Michael E.
Published: August 2008

The Discrete Element Method (DEM) discretizes a material using rigid elements of simple shape. Each element interacts with neighboring elements through appropriate interaction laws. The number of elements is typically large and is limited by computer speed. The method has seen widespread applications to modeling particulate media and more recently to modeling solids such as concrete, ceramic, and metal. For problems with severe damage, DEM offers a number of attractive features over continuum based numerical methods, with the primary feature being a seamless transition from solid phase to particulate phase. This study illustrates the potential of DEM for modeling penetration and briefly points out its numerous advantages. A weakness of DEM is that its convergence properties are not understood. The crucial question is whether convergence is obtained as DEM element size vanishes in the limit of model refinement. The major focus of our investigation will be a careful study of convergence for modeling the degradation of a solid into fragments. Our results show that indeed convergence is obtained in several specific test problems. Moreover, elastic interelement stiffness and damping properties were proven to be particle size-independent. However, convergence in material failure due to crack growth is obtained only if the interparticle potentials are properly constructed as functions of DEM element size and bulk material properties such as elastic modulus and fracture toughness.
The BWRVIP Integrated Surveillance Program
Authors: Carter Robert G. , Griesbach Timothy J. , Hardin Timothy C.
Published: August 2008

Boiling Water Reactor (BWR) plants in the U.S. are designed with radiation surveillance programs. However, the surveillance materials in some plants do not necessarily represent the limiting plate and/or weld material of the reactor pressure vessel (RPV). Also, some plants do not have baseline data for the surveillance materials, which is needed to measure irradiation shift. In 1998 the BWR Vessel and Internals Project (BWRVIP) conceived the BWR Integrated Surveillance Program (ISP) to address these concerns. The ISP surveyed all BWR vessel limiting materials and all available BWR surveillance materials (including materials from a 1990s supplementary research program called the Supplemental Surveillance Program, or SSP). For each vessel limiting weld and limiting plate, a best representative surveillance material was assigned, based on heat number, similar chemistries, common fabricator, and the availability of unirradiated data. Many of the selected surveillance materials are good representatives for the limiting materials of multiple plants, so fewer capsules are required to be tested, reducing the overall cost of surveillance while also improving BWR fleet compliance with 10CFR50 Appendix H.
Stress Corrosion Cracking of Cold Worked Austenitic Stainless Steels in Laboratory Primary PWR Environment
Authors: Vaillant Francois , Couvant Thierry , Boursier Jean-Marie , Amzallag Claude , Rouillon Yves , Raquet Olivier
Published: August 2008

Austenitic Stainless Steels (ASS) are widespread in primary and auxiliary circuits of Pressurized Water Reactors (PWRs). Moreover, some components suffer stress corrosion cracking (SCC) under neutron irradiation. This degradation could be the result of the increase of hardness and / or the modification of chemical composition at the grain boundary by irradiation. In order to avoid complex and costly corrosion facilities, the effects of radiation hardening on the material are commonly simulated by applying a pre-strain on non-irradiated material prior to stress corrosion cracking tests. The typical features of the cracking process in primary environment at 360°C during CERTs included an initiation stage (composed of a true initiation time and a slow propagation regime leading to a crack depth lower than 50 μm), then a “rapid” propagation stage before mechanical failure. Pre-straining increased significantly CGRs and the mode of pre-straining could strongly modify the crack path. No significant cracking (< 50 μm) was obtained under a pure static loading. A dynamic loading (CERT or cyclic) was required and various thresholds (hardness, elongation, stress) for the occurrence of SCC were determined. An important R&D program is in progress to develop initiation and propagation models for SCC of austenitic SS in primary environment.
HHC (Hitachi Hyper Valve): Corrosion Resistant, Withstands Erosion and Low CO Release
Authors: Kiyotoki Yoshihisa , Chigasaki Mutuo
Published: August 2008

In order to prevent valve performance deterioration by corrosion and wear against the seat (following, the valve-seat) of the valve-disc and the valve-body at fossil and nuclear power plants, a Co-base alloy is hard-faced onto the valve-seat. However, valve-seats deteriorate, for example, by corrosion, cracking and erosion. Moreover, the drift of the set point of the safety valves also occurs, by the phenomenon of Corrosion-Bonding. A valve-seat with excellent corrosion resistance, erosion toughness and mechanical toughness is sought. And, if possible, a low Co release rate material is preferred in nuclear power plants. Therefore, to improve valve performance, we manufactured a new valve named HHV with a valve seat, whose metal microstructure is different from that of conventional valve seats, within the range of chemical compositions of RCoCr-A (AWS). We also, evaluated the performance of the HHV Valve. We verified that HHV has improved valve performance. Moreover, we applied HHV under actual operating power plant conditions. The results show that HHV is superior to conventional valves in corrosion resistance, etc.
Weldability, Thermal Aging and PWSCC Behavior of Nickel Weld Metals Containing 15 to 30% Chromium
Authors: Boursier J. M. , Vaillant F. , Yrieix B.
Published: August 2008

In 1991, a vessel head penetration was found leaking at Bugey 3 plant during the hydrotest. Metallurgical investigations confirmed that this problem was again related to primary water stress corrosion cracking of alloy 600. Moreover, the main crack initiated in the base metal of the penetration (alloy 600) has also propagated in the weld metal in alloy 182. More recently, stress corrosion cracking in alloy 182 has been found on welds of U.S. plants. SCC susceptibility of alloy 182 has been evidenced by several laboratories. In France, all original vessel heads using alloy 600 have been or will be replaced with penetrations in alloy 690 (with 30% chromium). With respect to substitution materials, ELECTRICITE´ DE FRANCE has undertaken a large R&D study focusing on the development of new weld metals. The aim of this study was to identify new materials that will be able to weld alloy 690. Weld metals containing 15 to 30% Chromium have been studied. This paper presents an overview of the main results obtained on 19% Cr, 26% Cr and 30% Cr alloys with respect to alloy 182 (15% Cr). Firstly, the weldability of weld metals has been studied focusing on the susceptibility to hot cracking. Secondly, the resistance to thermal ageing has been investigated in order to detect any long term ordering of the solid solution Ni-Cr that could induce embrittlement. Hardness tests, Charpy tests and resistivity measurements did not show any effect of ageing up to 60,000 hours at 360°C. Thirdly, stress corrosion cracking susceptibility in primary water at 360°C has been evaluated during constant load tests, RUB tests, slow strain rate tests. No cracking was observed on material containing more than 26% Cr for both initiation and propagation. Finally, a life assessment was performed for all weld materials with respect to alloy 182.
Manufacture of Advanced 2.25Cr-1Mo-V Pressure Vessel Steel for High Pressure and High Temperature Hydrogen Service
Authors: Hasegawa T. , Narita A. , Iga H. , Inoue S.
Published: August 2008

At this moment, the materials which are applied for high pressure and high temperature services are selected from conventional 2.25Cr-1Mo, 3Cr-1Mo-V and 2.25Cr-1Mo-V steels. Especially, the application of 2.25Cr-1Mo-V steel is increased extremely rather than the other steels in order to take the advantage of higher strength and better resistance to hydrogen damages such as hydrogen attack, hydrogen assisted crack growth and also hydrogen induced disbonding. In order to achieve these properties, the vanadium element is added intentionally into base material and welding consumables. Due to the addition of vanadium, the engineer have recognized that there is some possibility of cracking in the weldment due to the higher hardness and insufficient manufacturing practices caused by lack of knowledge for preheating, intermediate/final postweld heat treatment and other factors. The manufacturers have to establish the proper technologies to avoid this problem during fabrication stage while the pressure vessels are to be fabricated in their shop and/or fields. In this paper, the manufacturing technologies of 2.25Cr-1Mo-V steel pressure vessel through actual fabrication is presented herewith.
Long-Term Isothermal Aging Behavior of V-Modified 2.25Cr-1Mo Steels
Authors: Yaguchi H. , Murakami S. , Fujitsuna N. , Shinya T. , Yamada M. , Sakai T.
Published: August 2008

Long-term isothermal aging behavior up to 30,000h has been investigated in V-modified 2.25Cr-1Mo steels. When impurity elements were added, FATT in Charpy impact test vs. aging time curves showed basically of parabolic type, indicating the importance of segregation of impurities as the critical factor in controlling temper embrittlement. On the other hand, when the amount of impurities was small, FATT showed complex behavior with aging time. This suggests the importance of other factors in controlling temper embrittlement such as recovery of dislocations and enlargement of various carbide precipitates. When the degree of temper embrittlement after aging is compared in terms of the increase in FATT due to aging between the modified steels and the conventional 2.25Cr-1Mo steels, it is found to be small in the modified steel at the same impurity level (J-factor). It was also found that the modified steels showed no hydrogen embrittlement. No change was observed in FATT after hydrogen charge on the contrast to the conventional steel, which showed significant increase in FATT. This effect has been attributed to the small amount of diffusible hydrogen content in the modified steel by the presence of hydrogen trapping sites, which is present even after long-term isothermal aging.
Global Assessment System for Structural Safety in Large Size Storage Tanks
Authors: Kaita Shinichi , Shibasaki Toshikazu , Tahara Takayasu
Published: August 2008

Considering long term stable supply of oil fuel, the world largest long-term storage system of crude oil has been installed in Japan. In order to ensure safety of large size above ground storage tanks, global assessment system for structural integrity of tank considering risk level and shut down inspection interval has been developed on Risk Based Inspection, RBI and Fitness-For-Service, FFS for storage tanks of crude oil for national security reserve.
Building a Real Time System for Evaluating Oil Storage Tank Damage Due to Earthquakes in Petroleum Stockpiling Bases
Authors: Zama Shinsaku , Endo Makoto , Hatayama Ken , Yoshida Shoichi , Kawano Kazuma , Sekine Kazuyoshi , Maruyama Hiroaki
Published: August 2008

It will be difficult to prevent the damage of oil storage tanks caused by a large earthquake even if extreme care has been exercised. Therefore, rational emergency responses will be required for preventing expansion into secondary disaster. In this paper, we propose a system that can estimate ground motion distribution in a whole of petroleum stockpiling base using a seismic record and evaluate damage of oil storage tanks just after an earthquake in order to support rational emergency responses. Spectral ratios of horizontal and vertical component of microtremors were used for evaluation of relative surface soil amplification. The system can assess the hazard for circumferential shell stress, axial shell stress, seismic capacity and liquid sloshing wave height immediately using the ground motion at each tank site estimated from both the relative soil amplification factors and a seismic record at a petroleum stockpiling base just after a large earthquake.
Study on Restoring Force Models of Nonlinear Uplift Behavior of Unanchored Oil Storage Tank Under Strong Seismic Excitation
Authors: Yuan Hai , Kawano Kazuma , Yoshida Shoichi
Published: August 2008

It had been reported that some unanchored oil storage tanks had serious damages due to its uplift behavior during strong seismic excitation. In some past studies, the uplift behavior of tank was generally simulated using elastic-plastic hysteresis models. But recently, a static FEM analysis for the cyclic uplift behavior of the shell-to-bottom joint of tank has shown that the hysteresis loop of the uplift force and uplift displacement is similar to S-curve with a narrow loop area. In order to investigate the effects of S-curve Restoring Force Model (SRFM) on the nonlinear response when it is used to simulate the tank’s uplift action, in this paper, the seismic response analyses of tank are carried out by using three kinds of restoring force models including SRFM, the Elastic-Perfectly Plastic Model (EPPM) and the Bilinear Elastic-Plastic Model (BEPM), and their results are compared with each other to examine the differences. In the analyses, the tank is modeled with an equivalent mass-spring model, and three recorded earthquake accelerations scaled to 0.6G are taken as the representation of strong motions. Consequently, this paper shows that SRFM gives the maximum response displacement and ductility respectively among the three restoring force models.
FEM Stress Analysis and the Sealing Performance Evaluation in Stainless Steel Elbow and Tee Fittings Under Internal Pressure: Case Where Internal Fluid is Liquid
Authors: Sawa Toshiyuki , Shimazu Hideaki
Published: August 2008

Stainless steel fittings such as elbows, tees, nipples and so on have been widely used in mechanical structures and chemical plants, it is well known that the leakage in the fittings used sealing tapes is less than that without the sealing tapes. In a practical design, it is necessary to examine the stress states and the leakage in the fittings under internal pressure and external loads such as tensile loads, bending moments and so on. This paper deals with the FEM stress analysis of the fittings subjected to internal pressure. In the FEM calculations, the engaged screw threads are taken into consideration as helical threads in the three-dimension. The leakage tests for the fittings under internal pressure were also conducted by using liquid (oil). Using the results of the leakage tests and the calculated stress states in the fittings, the sealing performance of the fittings under internal pressure was evaluated and the effect of the tightening torque was clarified on the sealing performance. In addition, the numerical results were compared with the experimental results. As the result, the effects of the sealing tapes on the contact stress distributions were also clarified.
The Damage of the National Oil Storage Bases by the 2003 Tokachi-Oki Earthquake
Authors: Kaita Shinichi
Published: August 2008

The Tokachi-Oki earthquake, which occurred in the early morning on September 26th, brought a lot of damage to storage tanks that are located at the oil refinery and the reserve base in Tomakomai City in Hokkaido. There are two stockpiling base in Tomakomai City, National Storage base and private storage base.
Remote Monitoring Systems for Detection of Contact by Unauthorized Construction Equipment and the Cathodic Protection Condition on Buried Pipelines
Authors: Hatanaka Shozo , Tange Masahiko , Sumiyama Morio
Published: August 2008

It is necessary to avoid spreading damage to the pipe when it is struck by construction equipment. It is difficult to install indirect methods (e.g. fiber optic sensors) for detection of construction equipment in heavy traffic areas. Therefore, remote monitoring systems, which can detect contact between such equipment and the steel of polyethylene-lined pipe, were developed and installed on a gas transmission pipeline (φ 600 mm, Length 9.4 km). Detection uses the change in pipe-to-earth resistance which is derived from inner pipe current and pipe to soil potential when contact occurs. Old and new data of pipe-to-earth resistance are continuously compared. If equipment contacts the steel pipe, pipe-to-earth resistance will drop below 100 Ω. The detection systems include maximal length sequence signal (pseudo-random signal) processing. This signal processing is superior for noise reduction. These systems also have remote monitoring of the cathodic protection condition (pipe to soil potential, current of coupons (probes), and transformer rectifier output voltage and current). From the office, pipeline lining and cathodic protection conditions are monitored at all times.
Apply CFD to Design Comfortable and Energy-Efficient Buildings
Authors: Zhai Zhiqiang
Published: August 2008

CFD can play an important role in building design. This paper introduces the typical variety of CFD applications to the design of a thermally-conformable, healthy and energy-efficient building. The paper discusses the major application challenges of CFD in the building modeling and design practice. The developing trends of applying CFD for building design have been analyzed by critically reviewing the proceedings of one of the most influential symposium in building simulation field (International Conference on Building Simulation).
On Effective Combination of Heuristic Algorithms and Dynamic Meshes in CFD Applications
Authors: Vatandas¸ Ergu¨ven , O¨zkol Ibrahim
Published: August 2008

In the study presented, dynamic mesh method applications on Heuristic Algorithms are investigated in the design of wing geometry. In our application dynamic mesh technique is implemented on a real coded genetic algorithm to demonstrate gain in computational time as well as in higher performance for optimized parameters. Since the differences developed at each step in geometries of new members are not much, it is possible to use dynamic mesh methods for new members. In this work, because the population members are obtained by modifying the pervious ones, each member is considered as one step of geometry-change of a deforming body, for example a wing inflating, deflating or cambering. Obviously the strategy applied here can be used for any slowly deforming complex geometries, as long as an effective combination of the genetic algorithm and dynamic mesh be succeeded.
Performance Analysis of a Low-Pressure Three-Stage Axial Compressor
Authors: Subramanian Sivakumar , Prasad B. V. S. S. S. , Krishnan S. , Janakamma C.
Published: August 2008

Prediction of three-dimensional flow through a multi-stage axial compressor involving multiple frames of reference is one of the challenging tasks in CFD. When the axial gap between the stationary and rotating blade rows is reduced, the blade row interactions become important. Therefore, a detailed knowledge of flow features is essential for the optimum design of multi stage compressor. As the design and conduct of experiments and evaluation of compressor performance is expensive and time consuming, many aerospace industries prefer to obtain the same information by the computational efforts. In this context, a number of CFD codes for modelling and analysis of turbomachinery flows are used. The most exigent aspect of simulating multi-stage compressor is representing the interactions between the rotor and stator. The present work is to find out the best-suited method for the analysis of a low-pressure three-stage compressor that gives reliable results. The motivation for this effort is derived from the inability to consistently compare predicted performance parameters obtained from using the interface models with the experimental results, which is especially true for off-design operation.