bluuebunny/crossref_metadata_2025 huggingface dataset:
French Procedure for Ageing Management Program of Safety Components
Authors: Faidy Claude
Published: August 2008

Managing ageing and remaining lifetime of an industrial facility is a concern that must be taken in account by utility as soon as possible in daily activities. The corresponding actions engaged in France are based on 3 major step that are described in the paper: • routine maintenance, • exceptional maintenance, • systematic and periodic review of safety important components and structures sensitive to ageing to assure the effectiveness of the maintenance actions and maintain a high safety level of the plant with a good competitiveness. Following different on-going programs on ageing management of different components, EDF developed its own approach, based on IAEA guidelines, in order to systematically review all the ageing management programs implemented on its 3-loop plants. The methodology is done in 3 steps: • selection of components and justification, • degradation mechanism analysis, • synthesis and consequences on maintenance programs. After a presentation of each step of the procedure a quick overview of the status of application in France is done. Comparison of the methodology with similar methodology used in different other countries is done to close the paper.
French Four-Loop PWR 1300 MWe Reactor Pressure Vessel: First Results for Integrity Assessment and Life Manangement
Authors: Bezdikian Georges , Ternon-Morin Franc¸oise , Churier-Bossennec Henriette , Moinereau Dominique , Martin Alain
Published: August 2008

The process used by the French utility, concerning the Reactor Pressure Vessel (RPV) integrity assessment, applied on 34 PWR NPPs 3-loop Reactors, involved the verification of the integrity of the component under the most severe conditions of situation, was engaged several years ago and the result obtained was the justification of the 900 MWe RPV life management for at least 40 years and to prepare the projection for beyond 40 years. Since 2000, in the continuity of this results, the studies was carried out on the 20 PWR NPPs 4-loop 1300 MWe Reactor Pressure Vessels, and the recent results obtained shows the demonstration of the integrity of the RPV, in the most severe conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and considering major parameters particularly the severity of the transient. This approach, is based on specific mechanical safety studies on the 1300 MWe RPV, to demonstrate the absence of risk of failure by brittle fracture. For these mechanical studies the major input data are necessary: 1 - the fluence distribution and the values of 4-loop RPV RTNDT during the lifetime in operation, 2 - the temperature distribution values in the downcomer and the PTS evaluation. The main results must show significant margins against initiation of the brittle fracture. The flaws considered in this approach are shallow flaws beneath the cladding (subclad flaws) or in the first layer of cladding. The major tasks and expertises engaged by EDF are: • better knowledge of the vessel material properties, including the effect of radiation, • more precise assessment of the fluence and neutronic calculations, • the NDE inspection program based on the inspection of the vessel wall, with a special NDE tool. The principal actions conducted during recent years are: • the optimisation of the fuel management and the new development for the fluence evaluation, • the data gathered from radiation specimen capsules, removed from the vessels (4loop reactor), within the framework of the radiation surveillance program, and the thermal-hydraulic-mechanical calculations based on finite element thermal-hydraulic computations and three dimensional elastic-plastic mechanical analyses.
Integrated Plant Life Cycle Management: The IAEA Contribution
Authors: Kang Ki-Sig , Clark Claude Russell , Juhn Poong Eil
Published: August 2008

For the past couple of decades there has been a change of emphasis in the world nuclear power from that of building new Nuclear Power Plants (NPP) to that of taking measures to optimize the life cycle of operational plants. National approaches in many countries showed an increase of interest in Plant Life Management (PLIM), both in terms of plant service life assurance and in optimizing the service or operational life of NPP. A strong convergence of views is emerging from different National approaches, particularly in the area of the economic aspects of NPP operation and in the evolution in the scope of NPP PLIM. The latter can directly affect the cost of electricity from NPP in an increasingly competitive environment. The safety considerations of a NPP are paramount and those requirements have to be met to obtain and to extend/renew the operating license. To achieve the goal of the long term safe, economic and reliable operation of the plant an Integrated Life Cycle Management Programme (ILCMP) is necessary. Some countries already have advanced PLIM Programmes while others still have none. The ILCMP objective is to identify all that factors and requirements for the overall plant life cycle. The optimization of these requirements would allow for the minimum period of the investment return and maximum of the revenue from the sell of the produced electricity. Recognizing the importance of this issue and in response to the requests of the Member States the IAEA Division of Nuclear Power implements the Sub-programme on “Engineering and Management Support for Competitive Nuclear Power”. Four projects within this sub-programme deal with different aspects of the NPP life cycle management with the aim to increase the capabilities of interested Member States in implementing and maintenance of the competitive and sustainable nuclear power. Although all four projects contain certain issues of PLIM there is one specific project on guidance on engineering and management practices for optimization of NPP service life including decommissioning. This particular project deals with different specific issues of NPP life management including aspects of ageing phenomena and their monitoring, issues of control and instrumentation, maintenance and operation issues, economic evaluation of NPP life cycle management including guidance on its earlier shut down and decommissioning. The paper describes in detail the full scope IAEA activities on different issues of NPP life management and some of its achievements in this field during the nearest past as well as plans for the future.
A Modified Technique for Crack Formation on Nuclear Steam Generator Tubing
Authors: Lee T. H. , Oh Y. J. , Hwang I. S. , Chung H. S. , Park J. Y.
Published: August 2008

For the aging management and safety assurance of an operating pressurized water reactor (PWR), the integrity of steam generator (SG) tubes receives increasing attention. Non-destructive examination by eddy current technique (ECT), leak rate measurement and burst pressure evaluation constitute key elements. SG tubes containing cracks with physical and microstructural characteristics similar to those of field-aged tubes are needed for the effort. We explored a radial dent loading method to introduce axial intergranular cracks using sensitized alloy 600 tubes in laboratories as a procedure to generate a library of laboratory degraded tubes (LDT) for ECT qualification. Based on three-dimensional finite element analysis and preliminary experimental work that the method is shown to be more useful than the internal pressurization method for the production of cracks with high aspect ratio. In addition, direct current potential drop (DCPD) method applied with array probes has been developed for use in accurate monitoring and control of crack size and shape. In order to minimize plastic deformation of region containing cracks, it is desired to use the radial dent loading only for shallow crack initiation. Then, subsequent growth of crack depth can be made by internal pressurization method.
The Sensitivity of Counterweighted Swing Check Valve Performance to Counterweight Location
Authors: Ezekoye L. I.
Published: August 2008

Check valves are used to minimize flow reversal. In general, the two primary design objectives of installing a check valve in a system include quick opening in forward flow and fast closure in reverse flow. The fast response requirements in both opening and closing directions are challenging. In the opening direction, the concern is to minimize forward flow resistance and, in the reverse direction, the objective is to minimize flow reversal and avoid water hammer. Check valve manufacturers have often used counterweights to permit quick opening or quick closing. The drawback of forward flow counterweight check valves is that in the flow reverse direction, the counterweights may retard valve closure. The location of the counterweight could further complicate the performance of the check valve. Misaligning the counterweight can also affect check valve performance. The use of quick closing counterweights present similar challenges. This paper examines the interaction of counterweight location and alignment on the performance of check valves.
Pressure Equipment Integrity at a Large Chemical Process Facility
Authors: Antaki George
Published: August 2008

The paper provides an overview of an integrity program for pressure equipment implemented at a large chemical process facility. It addresses the challenges encountered and solutions achieved in the practical implementation of the ASME Boiler and Pressure Vessel Code and the National Board NB-23 Code.
Development of the Guideline on Inelastic Analysis for Design
Authors: Tanaka Yoshihiko , Shibamoto Hiroshi , Inoue Kazuhiko , Kasahara Naoto , Ando Masanori , Morishita Masaki
Published: August 2008

The guideline on inelastic analysis for design, one of the key items of Fast Reactor Design Standard (FDS), is being developed. The basic policies of this guideline are as follows: (a) to emphasis conservative analysis output rather than nominal value representing actual behavior, (b) to clarify the applicable area for assurance of conservative results. With such concepts, it would be possible that the guideline provides useful explanations on the manner of analysis and estimation in the form of concrete examples of design as well as general rules (somehow vague). As the first step of the guideline development, the following five issues to be solved were extracted: 1) applicable area, 2) selection of constitutive equation, 3) modeling method of the load history, 4) ratchet strain and creep fatigue damage evaluation methods by inelastic analysis and 5) example design problems to check users’ analysis quality and to complement the general rules. In parallel, inelastic analyses with the promising constitutive equations were applied by way of trial to obtain rough presumption on their effects on structural design of the components. As a result, all inelastic analyses provided smaller cumulative strains and equivalent strain ranges than the existing design method based on elastic analysis, suggesting advantage of introducing them into actual design.
Plastic Analysis of a Reactor Pressure Vessel Support Pad With Different Methods From Appendix F, Faulted Conditions, Comparison of Allowable Loads
Authors: Schramm Konrad
Published: August 2008

The present paper shows the results of different plastic analysis methods stipulated in the ASME-code appendix F for faulted conditions. Results will be compared and differences with an elastic analysis are discussed. The analyzed structure is a support pad of the RPV (reactor pressure vessel) of a German PWR. There are four support pads equally distributed around the outer surface of the flange of the RPV. Under normal and faulted conditions these support pads serve as a link between the RPV and RPV support construction, which is in the German design a ring girder. A vertical force is applied onto the structure in an incremental manner until failure will occur. Real measured true stress strain curve is used. In the context of the European research project LISSAC [7] special attention is paid on local effects near holes or notches causing non-uniform stress and strain distributions. The chosen support pad provides such a hole. Allowable loads achieved with elastic analysis, plastic analysis, limit load or plastic analysis collapse load are compared. Details as well as problems of the numerical analysis were discussed. Another aspect is the strain distribution in a plastic analysis, especially near local discontinuities. As an outlook these plastic strains were compared with local failure strains found in the LISSAC project.
Analysis of Proportional and Nonproportional Loading for Low Cycle Fatigue Design
Authors: Itoh Takamoto , Sakane Masao
Published: August 2008

This paper proposes a simple method for defining uniaxial and multiaxial strain and stress states by using principal strains and stresses. It also proposes a simple method for determining the principal and shear strain/stress ranges in proportional and nonproportional loading, introducing an angle rotation of the maximum principal strain and stress. The comparison of the proposed methods and that used in the pressure vessel code in Japan was discussed. The suitability of strain and stress parameters for predicting proportional and nonproportional low fatigue lives was also discussed.
Stress and Fracture Analysis of Ceramic Lined, Composite or Steel Jacketed Pressure Vessels
Authors: Underwood John H. , Parker Anthony P.
Published: August 2008

Stress and fracture analysis of ceramic-lined cannon pressure vessels is described, for a Si3N4 or SiC liner and A723 steel or carbon-epoxy jacket and with an initial residual interface pressure between liner and jacket. Room temperature stresses for a steel jacket over ceramic are similar to those for a carbon-epoxy jacket, but both radial and hoop jacket stresses can exceed typical carbon-epoxy strength values. Elevated temperature liner stresses are reduced for a carbon-epoxy jacket, due to the effective increase in interface pressure caused by differential thermal expansion. Critical crack size for brittle fracture is larger for Si3N4 than SiC due to lower liner stresses and higher fracture toughness with a Si3N4 liner.
Fundamental Study on the Super-Long-Period Active Isolation System
Authors: Minagawa Keisuke , Fujita Satoshi
Published: August 2008

Since the Hanshin-Awaji Earthquake Disaster, the number of isolated structures has been greatly increased. The natural period of the isolation system is designed around 3 seconds, because predominate period of observed seismic waves is usually 0.1 to 1 second. However, relatively long period seismic waves have been observed in various earthquakes, and the resonance of long-period structures, such as high-rise buildings, during earthquakes have been reported at the same time. Therefore the natural period needs to be extended. When extending the natural period of the isolated structure using rubber bearings, its stiffness needs to be reduced. It is more difficult to extend the natural period of the isolation system than the conventional system because of its buckling problem. Therefore we propose a super-long-period active seismic isolation system as a new method for extending the natural period of the isolated structure. This system consists of rubber bearings and actuators. In this study, we designed a control system by using the model-matching-method. This is one of the classical control system design methods. We investigated the isolation performance by numerical analysis. In addition, we selected the optimal variables of transfer function using genetic algorithm.
Three-Dimensional Analysis of Crack in Centrally Perforated Photoelastic Cylinders Under Internal Pressure
Authors: Liu C. T. , Smith C. W.
Published: August 2008

In this study, the frozen stress photoelastic method was used to investigate the three-dimensional effect on the crack growth behavior in a centrally perforated circular cylinder under internal pressure. The inner surface of the cylinder had a star shape, which consisted of six fins. The specimens were capped at the ends and pressurized internally above critical temperature after real cracks were introduced at the fin tip. After growing to a desired size, the pressure was reduced to stop crack growth and held through cooling. Two different types of cracks, part-through crack and long crack with the crack length nearly equal to the length of the cylinder, were considered. The experimental data were analyzed, and the results are discussed.
Comparison of Fracture Methodologies for Flaw Stability Analysis of Storage Tanks
Authors: Lam P.-S. , Sindelar R. L.
Published: August 2008

Fracture mechanics methodologies for flaw stability analysis of a storage tank were compared in terms of the maximum stable through-wall flaw sizes or “instability lengths.” The comparison was made at a full range of stress levels at a specific set of mechanical properties of A285 carbon steel and with a specific tank configuration. The two general methodologies, the J-integral-tearing modulus (J-T) and the failure assessment diagram (FAD), and several specific estimation schemes were evaluated. A finite element analysis of a flawed tank was also performed for validating the J estimation scheme with a curvature correction and for constructing the finite element-based FAD. The calculated instability crack lengths show that the J-T methodology based on a center-cracked panel solution with a curvature correction, and the material-specific FAD, most closely approximate the result calculated with finite element analysis for the stresses at the highest fill levels in the storage tanks (less than 124 MPa or 18 ksi). The results from the other FAD methods show instability lengths less than the J-T results over this range.
A Numerical Model of Solidification of a Massive Casting From Malleable Cast-Iron
Authors: Kavicka Frantisek , Stetina Josef , Stransky Karel , Dobrovska Jana , Sekanina Bohumil , Heger Jaromir
Published: August 2008

This numerical model of the temperature field of solidifying castings aims to achieve two general goals: directed solidification (as the primary condition for a healthy casting) and optimization of the technology of casting together with the preservation of optimum utility properties of the product. A specific goal of this model is the selection and optimization of the method of cooling to shorten the solidification time in order to obtain a spherical graphite structure with good nodular properties and with a sufficient density of graphite spheres (cells). The speed of cooling during solidification and cooling in the mould is therefore a significant quantity influencing the formation of the structure. The achievement of these goals depends on the ability to analyze and, successively, to control the effect of the main factors which characterize the solidification process or accompany it. The analysis of the quantities is focused on determining the causes of the formation of the heterogenic temperature field within the actual casting during casting, considering the phase and structural changes. It is also focused on the thermokinetics of the formation of shrinkage porosities and cavities and on the prediction of their formation. This leads to the optimization of the shape and sizes of the risers, the method of insulation, the treatment of the level, etc. The model is applicable to various shapes of castings. The software is capable of analyzing the temperature field of the actual casting, as well as the temperature field of the mould and cores, including the dependence of their material. It is also capable of considering non-linearity, i.e. the dependence of the thermophysical properties—namely the material of the casting and mould on the temperature, as well as the dependence of the heat transfer coefficients on the surface and interface temperature. The model is also equipped with an original network generator (pre-processing) as well as graphical output (post-processing).
Effect of Length on Stress Concentration for Short Cylindrical Members With Holes
Authors: Troyani Nando , Sterlacci Gaetano , Jaimes Nelson , Gomes Carlos J.
Published: August 2008

As is well known stress concentration will appear in any application where some form of geometric discontinuity is present. To deal with such situations the so called Stress Concentration Factors (SCF) concept was developed and is widely used in both analysis and design of loaded components especially when subjected to fatigue, usually the working condition of vessels. However, recent observations suggest that the influence of member length on the magnitude of the stated SCF’s was not considered. In this work, this observation was studied in the context of cylindrical vessels and it was found that in this case, as well, length could be a critical factor when computing stresses developed as a result of externally applied loads. The values of the Finite Element (FE) calculated Theoretical Stress Concentration Factors (TSCF’s) are computed, for the case of short circular cylinders with circular holes subjected to axial tension, and presented in a fashion similar to existing published results. It is shown that significantly larger stress concentrations appear for short members. The novel concept of transition length, that defines the threshold between long plates and short plates, is discussed in the context of this study and reported as well.
Thermal Stress Analysis and Sealing Performance Evaluation of Pipe Flange Connections With Spiral Wound Gaskets Under Elevated Temperature
Authors: Sawa Toshiyuki , Maezaki Wataru
Published: August 2008

In this paper, the thermal stress distributions at the interfaces between pipe flanges and the gasket under elevated temperature and internal pressure were calculated by using the finite element method (FEM) taking into account hysteresis in the stress-strain curves of spiral wound gasket. Leakage tests were performed using helium gases. In addition, the effects of temperature on the sealing performance were examined by using an actual pipe flange connection with spiral wound gasket under elevated temperature. By using the calculated contact stress distributions and the results of the leakage tests, the sealing performance was evaluated.
Stress Analysis of a 46-Inch Reactor Feed/Effluent Exchanger
Authors: Alexander Chris , Jagodzinski John , Biel Richard
Published: August 2008

Stress Engineering Services, Inc. performed a series of analyses for Brown Fintube to determine the mechanical behavior of its high performance 46-inch reactor feed/effluent exchanger. Initial efforts focused on the global behavior of the exchanger subjected to pressure and thermal loads, while the ultimate objective addressed potential leakage in the main body flange. Using finite element analyses incorporating shell, solid, and continuum models, Stress Engineering was able to demonstrate that the sealing load of the flanges could be maintained even with elevated bolt temperatures up to 400°F. Using the methods permitted by the ASME Boiler & Pressure Vessel Code, Section VIII, Division 2, linearized stresses were calculated. All calculated stress intensities were less than the Division 2 allowable stresses. Based upon the results of the analyses, the design of the main body flange also met the stress design criteria per Division 1 of the Code. This project was a clear demonstration on how analysis methods can be used to solve complex engineering problems that include a wide range of variables and operating conditions. The conservative calculation techniques of Division 1 of the Code were also confirmed in this work.
Efficient Assembly of Pressure Vessel Bolted Joints
Authors: Brown Warren
Published: August 2008

This paper details the results of recent testing and experience into the assembly of bolted joints. Pressure boundary bolted joint assembly procedures are examined and details are presented that result in a reduction in the required effort to assemble any given joint by more than 50%. The selection of correct assembly load levels is addressed and suggestions for joint integrity improvement are made. Experimental results are also presented on the use of bolt lubricants and the effect of temperature on the lubricant properties.
Comprehensive Thermal Solution Procedures for Laser Forming
Authors: Cao Zhenning , Yang Zhishang , Yang Yuping
Published: August 2008

Comprehensive thermal solution procedures (CTSP) are developed to predict temperature history during laser forming process. The CTSP can provide not only fast and accurate results, but also many useful features, such as convection boundary conditions and distributed water cooling schemes. By means of interface through the virtual fabrication technology (VFT®) software, the CTSP gives direct temperature results for FEA stress/distortion analysis.
Fracture Assessment of Through-Wall and Surface Cracked Pipes by BS 7910 and API 579 Assessment Procedures: A Comparative Study
Authors: Motarjemi Afshin K.
Published: August 2008

Fracture assessment procedures such as BS 7910 and API 579 are formulated based on the Fracture Mechanics concept for assessing integrity of structures such as pipelines, pressure vessels, etc. In the current study those procedures are applied to through-wall and surface cracked pipe geometry under four-point bending. The predicted maximum tolerable applied loads are then compared with pipe full-scale fracture testing results published by Miura et al (2002). Other assessment schemes namely, GE/EPRI, Net-section plastic collapse, LBB.NRC and finally LBB.ENG2, as reported in the same publication are also included in the current paper for sake of comparison. The comparative study showed that BS 7910 and API 579 predict similar maximum tolerable load for through-wall pipes but different value for surface-cracked pipes. Difference in predictions for the latter geometry is owing to the use of different stress intensity factor/reference stress solution by BS 7910 than API 579. However, both procedures provided conservative results compared with the experimental data as well as other engineering routes mentioned in Miura et al (2002).
Factors Influencing SCC and IASCC of Stainless Steels in High Temperature Water
Authors: Andresen Peter L.
Published: August 2008

SCC growth studies were performed in high temperature, high purity water on various grades and various conditions of stainless steel. The synergistic effects of corrosion potential, sensitization, cold work (yield strength), temperature and irradiation were evaluated, and their implications to interpreting and modeling SCC in unirradiated and irradiated structures are discussed.
Study on the SCC Growth of Alloy 182 Under Constant Displacement Condition in High Temperature Oxygenated Water
Authors: Suzuki Shunichi , Kumagai Katsuhiko , Namatame Satoshi , Kikuchi Masaaki , Itow Mikiro , Futami Tsuneo
Published: August 2008

SCC initiates and propagates along the fusion line or in the weld metal in BWR and many SCC initiation & propagation studies have been performed so far (Saito, et al. (1997), Kikuchi, et al. (1997), Itow, et al. (1997, 2000), Suzuki (1999), Namatame, et al. (2001)). SCC growth behavior can be evaluated by conjunction of SCC growth rates and the residual stress of the welded component, which consists of tension/compression region. Especially, thick components such as core shrouds have increasing and decreasing tensile stress profile under constant displacement. In general, SCC growth rates are obtained from CT specimens under constant loading condition. This study shows that SCC growth rates depend on dK/dt as well as on K and that their growth rates under constant displacement with decreasing K are lower than those under constant loading condition with increasing K.
Evaluation of IGSCC Growth Behavior of 316NG PLR Piping in BWR
Authors: Kumagai Katsuhiko , Suzuki Shunichi , Mizutani Jun , Shitara Chikashi , Yonekura Kazuyoshi , Masuda Minoru , Futami Tsuneo
Published: August 2008

Through inspections of Japanese boiling water reactors (BWRs) from 2002, a high occurrence rate of Inter Granular Stress Corrosion Cracking (IGSCC) was confirmed in core shrouds and re-circulation (PLR) pipings made of low-carbon (L-grade) stainless steels (Okamura, et al., 2003), and the evaluation of IGSCC growth behavior especially in PLR pipe weld joints has been recognized as one of the urgent and important issues for applying the Japanese maintenance code for BWRs. Adjacent to the weldment of PLR pipe joints, work-hardened area due to weld shrinkage is generated, which enhances SCC growth rates in BWR environment (high purity, 288°C water). However, the effect of hardening on SCC growth rate has not been considered in the present flaw evaluation codes. On the other hand, detailed examinations on actual cracked PLR joints showed that SCC growth in the depth direction seems to be arrested or, at least, decelerated around the fusion boundary. Therefore, it is not appropriate to apply a enhanced growth rate of work-hardened materials to the entire SCC path in PLR weld joints. In this paper, a new concept to evaluate SCC growth considering observed characteristics of SCC behavior is proposed, followed by tentative analyses using latest crack growth rate data measured in specimens taken from the real-size pipe joint mock-ups.
A Method of Detecting Fatigue Crack Initiation and an Application of an Overload Using Local Strains
Authors: Makabe Chobin , Kaneshiro Hideo
Published: August 2008

A method of detecting a fatigue crack which initiates from a notch and an application of an overload during a stress cycle were investigated. A crack initiated from a notch can be detected by analyzing the waveform of strain function which was calculated from the local strains in the vicinity of the notch. The local compliance of the material changed after initiation of the crack. This is the reason why the crack initiation can be detected by the strain waveform. This technique can also be applied to detection of an overload application during stable crack growth. The waveform shape of strain function changed after overloading. Especially, the present method is useful for cases of crack propagation under negative stress ratio conditions.
Effects of Water Flow Rate on Fatigue Life of Carbon and Stainless Steels in Simulated LWR Environment
Authors: Hirano Akihiko , Yamamoto Michiyoshi , Sakaguchi Katsumi , Shoji Tetsuo
Published: August 2008

Fatigue tests in simulated LWR environment of carbon and stainless steels were performed under high water flow rates between 7 to 10 m/s. For carbon steel, high flow rate of water clearly mitigated the environmental effect on a fatigue life at the high sulfur concentration of 0.016 wt% which caused high environmental effect on a fatigue life. On the contrary, high flow rate of water slightly enhanced the environmental effect at the low sulfur concentration at or less than 0.008 wt% which caused very low environmental effect. These results suggested that the environmental fatigue life under various flow rate conditions should be determined by the combination between the mitigating effect caused by flushing of locally severe environment and the enhancing effect caused by increase in corrosion potential. To understand those effects, effects of sulfur concentration on fatigue life for various DO condition were formulated. And corrosion potential under low and high flow rate condition was measured during the fatigue test. Environmental correction factor, Fen, which is the ratio of fatigue lives derived from the fatigue life at room temperature in air divided by that in water to be used for the fatigue life prediction at high flow rate condition was assumed based on the MITI guideline equation and considering the hypothetical fatigue life under sulfur free condition and high corrosion potential condition. This assumption was agreed very well with the test data. For stainless steel, flow rate had little effect on a fatigue life of type 316 stainless steel. It suggested that there was no role of water flushing. For type 304 stainless steel, fatigue life has a tendency to decrease with increase in water flow rate. Fatigue lives of type 304 stainless steel under high flow rate of 7 to 10 m/s were shorter than those predicted by MITI guideline equation. This effect should be considered in an evaluation of environmental fatigue.
Effect of Hydrogenated Water on Fatigue Crack Growth for Austenitic Stainless Steel in Simulated BWR Water Environment
Authors: Wang Li H.
Published: August 2008

Fatigue crack growth rates (FCGR) of sensitized austenitic stainless steel (SS) were measured in simulated BWR water at 288 °C using compact tension specimens under different cyclic loading modes, including saw-tooth, trapezoidal and constant loading pattern. This study tested sensitized SS in normal water chemistry (NWC) and hydrogen water chemistry (HWC) respectively, and attempted to clarify the effect of low electrochemical corrosion potential on the FCGR of sensitized stainless steel. Significant environment effects on FCGR of sensitized stainless steel were observed in both water chemistries when compared with air fatigue curve. The pronounced suppression effect of HWC on crack growth in statically sustained load was not observed in cyclic loading condition. ASME curve doesn’t seem to be conservative and could not bound all the FCGR data tested in this study. In contrast, all of the measured FCGR data were bound by the JSME disposition curve. PLEDGE model proposed by General Electric reasonably predicted the FCGR of sensitized SS in NWC, but underestimated the FCGR in HWC. ANL’s superposition model successfully estimated the FCGR measured in both water chemistries. The fractography exhibited transgranular fracture mode during the crack initiation and growth stage. No differences in the appearance of fracture surface were observed in HWC and NWC. Only in very high DO environments, the sensitized 304 SS exhibited the mixed mode of intergranular and transgranular during growth stage.
First Publication of the Nuclear Materials Code for Nuclear Facilities in Japan
Authors: Murakami Hiroyoshi
Published: August 2008

The Subgroup on Nuclear Materials of JSME (Japan Society of Mechanical Engineers) has been developed a new nuclear materials code for nuclear facilities. The current nuclear materials code was published as a part of the Ministry of Economy, Trade and Industry Code (METI) Notification No. 501 [1]. The new materials code shall come out as a separate publication. It shall publish in 2004. The new nuclear materials code contains 58 ferrous material specifications of Japanese Industry Standards (JIS), 15 nonferrous material specifications of JIS and 14 nuclear specified materials. The code has widely consistent with JIS such as JIS B8265 [2] and B8266 [3], JSME code such as the Rules on Design and Construction for Thermal Power Generation Facilities [4], and ASME Boiler and Pressure Vessel Section II [5]. This paper introduces outlines on the Rules on Nuclear Materials for Nuclear Facilities.
Technical Changes in ASME B31.3, 2004 Edition
Authors: Becht Charles
Published: August 2008

Significant changes have been included in the 2004 edition of the ASME B31.3, Process Piping Code. This presentation provides some highlights of the changes.
Assessment of the Reactor Pressure Vessel Integrity in PTS Conditions: ThermalHydraulic and ThermoMechanical Studies of a Safety Injection in a Four Loop PWR Plant
Authors: Martin A. , Bugat S. , Dahl A. , Lestang F. , Beaud F.
Published: August 2008

This paper explains the Research and Development program started at E.D.F about the cooling phenomena of a PWR vessel after a Pressurized Thermal Shock. The numerical results are obtained with the thermalhydraulic code Code_Saturne coupled with the thermal-solid code SYRTHES to take into account the conjugate heat transfer on the cooling of the vessel. The geometry used represents a four loop PWR plant. In this calculation, the simulated geometry takes into account as much as possible the exact geometry of the lower plenum such as its columns and plates instrumentation. The configuration investigated is related to the injection of cold water in the vessel during a penalizing operating transient and its impact on the solid part formed by cladding and base metal. Numerical results are given in terms of temperature field in the cold legs and in the down comer. The obtained numerical description of the transient (internal pressure and temperature field within the vessel) is used as boundary conditions for a full mechanical computation of the stresses. This thermal–mechanical transient is obtained by F.E. simulation using the F.E. code Code_Aster on a 3D mesh of the vessel, covering the two core–shells and their circumferential welds, as well as the internal cladding. Based on an analytical method specially established for underclad flaws, the corrected stress intensity factor Kβ during the transient is evaluated for an hypothetical flaw, by extracting the stresses along a radial segment. The severity of the flaw with respect to the transient is quantified by the minimum of the ratio KIc/Kβ, where KIc refers to the base metal fracture toughness for brittle initiation. The evolution of the severity with the position of the hypothetical flaw is studied and compared with the results given by a classical uni–dimensional method. The results show that such a complete thermal–hydraulic and mechanical 3–dimensional analysis allows to reduce considerably the severity of the flaws, thus improving the integrity of the RPV.
Influence of Constraint on JR Curves and on Maximum Load of Circumferentially Surface-Cracked Pipes
Authors: Novak Jiri
Published: August 2008

Assessment of the maximum load of circumferentially surface-cracked pipes is usually performed on the basis of JR curves measured in experimental arrangement with high constraint of plastic flow in the crack tip region, i.e. on the basis of JR curves determined conservatively. For lowering the conservatism, it is necessary to use JR curve corresponding to conditions in the pipe. Correlation between JR curves and parameter B/√a is proposed; here B is biaxiality factor defined by Leevers and Radon and a is (initial) crack length. This parameter depends on geometry only and represents — up to a numerical factor — T-stress normalized by the stress intensity factor KI. Determination of the maximum load of circumferentially surface-cracked pipes is based on JR curves corresponding to the constraint in the pipe and on the Failure Assessment Diagram defined in the R6 method. Size of the effect depends on the pipe and crack geometry (including size) and on mechanical and fracture properties of the material (including constraint dependency of JR curves). Numerical examples use published data concerning properties of steels A533B and HY-100.
A Two-Parameter Description of the Behaviour of a Process Zone at a Crack Tip: Part II — Determination of the Parameters
Authors: Smith E.
Published: August 2008

An earlier paper (Part I) has shown how key parameters associated with the uniform stress process zone model of a crack: crack tip opening displacement, process zone size, crack opening area and the effective opening area of the process zone, depend upon parameters that are associated with the relevant terms in the expansion of the expression, for the purely elastic situation, for the relative displacement of the crack faces or the stress ahead of an elastic crack. The earlier paper focussed upon the case where the non-linear (with regards to applied stress) contributions to the crack-process zone parameters were determined to the first two terms in increasing powers of the applied loading stress parameter. These terms depend upon the first two terms in the expressions for the crack face relative displacement on the stress ahead of the crack in the elastic situation. The first of these terms is related to the stress intensity factor. In this paper we show how the parameter g0, which defines the second term, can be determined for some idealised situations.
Simulation of Buckling Instability of a Large Deformation Twisting Operation
Authors: Metzger D. R. , Kim Y. S. , Han S.
Published: August 2008

The twisting of flat bars to create the shape of an auger has been used by blacksmiths for many centuries. Easily recognizable examples of such work include twisted square rods used in wrought iron gates, and hand crafted twist drills. To this day, a large variety of long drills used in the mining industry are created by twisting flat bars of appropriate cross-section. In all of these examples, the edges of the material form a helical pattern which may be characterized by the diameter to pitch ratio. In practice, a pitch ratio of unity is typical, as this ensures sufficient axial motion of material. This pitch ratio is obtained without difficulty as long as the cross-section of the bar is not very slender. In an effort to manufacture augers intended for moving granular material rather than for drilling, thin and wide sections are desirable. This is mainly to save weight and material cost. However, such slender geometry may be prone to non-uniform deformations prior to achieving the desired pitch ratio. Apparently, for a given thickness, the ability to produce an acceptable final product is diminished as the width is increased. Nonuniform deformation in the form of folding or curling as shown in Figure 1 is the main concern for the wide and thin bars. The folding response may be recognized as instability of the system under the action of the applied torsion. Buckling of rods due to torsional loading has been studied theoretically by Ziegler [ 1 ] for the case of circular cross-sections and elastic deformation. Thin walled beam sections are considered by Timoshenko and Gere [2], and Bazant and Cedolin [3] with deformations limited to small strain elasticity. Twisting an auger necessarily requires very large strains and plastic response of the material. Furthermore, the twisting operation involves several parameters, such as temperature, axial tension and twisting rate. The success of the process and the range of its applicability may be improved if the optimum conditions are established. For this purpose, numerical modelling of the process is applied to many variations of the parameters to establish an understanding of the limitations, and how to best control the operation.
Advanced General-Purpose Finite Element Solid Analysis System Adventure_Solid on the Earth Simulator: Its Application to Full-Scale Analysis of Nuclear Pressure Vessel
Authors: Shioya Ryuji , Ogino Masao , Kawai Hiroshi , Yoshimura Shinobu
Published: August 2008

We have been developing an advanced general-purpose computational mechanics system, named ADVENTURE, which is designed to be able to analyze a model of arbitrary shape with a 10–100 million degrees of freedom (DOFs) mesh, and additionally to enable parametric and non-parametric shape optimization. Domain-decomposition-based parallel algorithms are implemented in pre-processes (domain decomposition), main processes (system matrix assembling and solutions) and post-process (visualization), respectively. Especially the hierarchical domain decomposition method with a preconditioned iterative solver (HDDM) is adopted in one of the main modules for solid analysis, named ADVENTURE_Solid. The employed preconditioner is the Balancing Domain Decomposition (BDD) type method. The ADVENTURE_Solid has been successfully implemented on a single PC, PC clusters and massively parallel processors such as Hitachi SR8000/MPP. In this study, this solid analysis module is implemented with minor modification on the Earth Simulator consisting of 256 nodes, i.e. 2,048 vector-type processing elements of theoretical peak performance of 16 TFLOPS (Tela FLoating point Operations Per Seconds), and succeeded in solving an elastostatic problem of a nuclear pressure vessel model of 100 million DOFs mesh in 8.5 minutes with 5.1 TFLOPS, which is 31.8% of the peak performance and over 80% parallel efficiency. As the purpose of demonstration of virtual mock-up test, the ADVENTURE_Solid is applied to solve a precise model of the ABWR vessel subjected to two kinds of loading conditions, i.e. (1) quasi-static seismic loading and (2) hydrostatic internal pressure.
Out-of-Plane Deflection of Nonprismatic Curved Beam Structures Considering the Effect of Shear Deformation Solved by DQEM
Authors: Chen Chang-New
Published: August 2008

The development of differential quadrature element method out-of-plane deflection analysis model of curved nonprismatic beam structures considering the effect of shear deformation was carried out. The DQEM uses the differential quadrature to discretize the governing differential equation defined on each element, the transition conditions defined on the inter-element boundary of two adjacent elements and the boundary conditions of the beam. Numerical results solved by the developed numerical algorithm are presented. The convergence of the developed DQEM analysis model is efficient.
An Efficient Algorithm for Damage-Coupled Visco-Plastic Fatigue Model
Authors: Zhao A. H. , Chow C. L.
Published: August 2008

The paper describes the development of an efficient and robust numerical algorithm for a damage-coupled visco-plastic-fatigue material model. The material chosen for the investigation is a eutectic material, Sn-Pb solder, exhibiting strain-softening behavior. The numerical algorithms employs a modified explicit method with adaptive sub-stepping based on the local error control for which the stress (constitutive) Jacobian explicit solution is derived. The algorithm is implemented in a commercial finite element (FE) code ABAQUS (Version 6.2) via its user-defined material subroutine. The validity of the algorithm is examined with several numerical examples, including (i) single-element simulations for uniaxial test, tensile creep, and fatigue simulations to attain an optimized algorithm, and (ii) two three-dimensional analyses of a miniature specimen under monotonic tensile loading and fatigue loading. The numerical examples illustrate the effectiveness of the modified explicit algorithm in predicting cyclic thermoviscoplastic behavior of a solder material. The algorithm is considered a generalized methodology that can be readily applied characterize thermoviscoplastic behavior and fatigue life of similar materials.
ES-3100: A New Generation Shipping Container for Bulk Highly Enriched Uranium and Other Fissile Materials
Authors: Arbital Jeffrey G. , Byington Gerald A. , Tousley Dean R.
Published: August 2008

The U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) is shipping bulk quantities of surplus fissile materials, primarily highly enriched uranium (HEU), over the next 15 to 20 years for disposition purposes. The U.S. Department of Transportation (DOT) specification 6M container is the package of choice for most of these shipments. However, the 6M does not conform to the Type B packaging requirements in the Code of Federal Regulations (10CFR71) and, for that reason, is being phased out for use in the secure transportation system of DOE. BWXT Y-12 is currently developing a package to replace the DOT 6M container for HEU disposition shipping campaigns. The new package is based on state-of-the-art, proven, and patented insulation technologies that have been successfully applied in the design of other packages. The new package, designated the ES-3100, will have a 50% greater capacity for HEU than the 6M and will be easier to use. Engineering analysis on the new package includes detailed dynamic impact finite element analysis (FEA). This analysis gives the ES-3100 a high probability of complying with regulatory requirements.