bluuebunny/crossref_metadata_2025 huggingface dataset:
Simulation and Measurements of Electron Beam Welding Distortions
Authors: Martin Antoine , Robin Vincent , Gilles Philippe , Bouyer Fabrice
Published: July 2010

Welding problems encountered in the nuclear industry have been mainly addressed by weldability tests analysis, development of new techniques or improvements through lessons learnt. For the last decade, AREVA has developed a complementary approach based on numerical simulation. Residual stresses inherent in the manufacturing of reactor components do not constitute a major problem at the design stage; even though they may have a strong impact on some types of damage. Numerical welding simulation in the nuclear industry has mainly focused on residual stress prediction. Distortion may also be a source of concern in component design: some structures are slender in spite of their thickness. In the frame of GEN IV project like sodium fast reactor or ITER, the distortion problem has gained in importance. For industrial project considering large structure, DCNS has developed electron beam welding process and built several mock-ups to validate the manufacturing steps. AREVA NP carried out welding simulations on one of these mock-ups (1m × 1m) made of austenitic steel plates and representative of nuclear components. The aim of the numerical simulations was to check the quality of the distortion prediction. Electron beam (EB) welding offers significant advantages for structures with many long welds and high requirements on the final shape of the welded structure. Since distortion values are very low (between 0.1 mm and 1.6 mm), prediction of distortions induced by electron beam welding is difficult. Compared to the size of the mock-up (1 m), distortions represent only 0.1%. Distortions induced by each weld have been computed using a simplified approach (local/global method [1]), which aims at modelling long and numerous welding operations with an acceptable calculation time. After each welding sequence, DCNS measured distortions at some representative points of the mock-up. This paper comments on the difficulties of the validation of the numerical predictions by comparison with measurements: • Measurements are made after each welding step, requiring removing the parts from the chamber. • Distortion values are very low compared to the size of the mock-up that may increase the risk of measurement error. In spite of these problems, the main trends of the experimental deformed shape are well represented by the simulation.
Constraint Corrected Cleavage Fracture Assessment Based on Tests of Standard and Small Scale Cruciform Specimens
Authors: Hohe Jo¨rg , Hardenacke Volker , Siegele Dieter
Published: July 2010

The objective of the present study is the determination and validation of a constraint corrected probabilistic cleavage fracture model considering both, the nucleation and the possible instability of potentially critical microdefects. For this purpose, the local mechanical conditions at the cleavage triggering point are analyzed by a numerical simulation of different fracture mechanics experiments using C(T), M(T) and SE(B) specimens with deep and shallow cracks. The data base is extended by experiments on novel small scale cruciform bending specimens allowing the application of an additional stress acting longitudinally to the crack front. Based on the results, an enhanced local cleavage fracture model is proposed, which accounts for the nucleation of microdefects in terms of a combined parameter of the local accumulated plastic strain and the local stress triaxiality ratio.
Instability of Pressure Relief Valves in Water Pipes
Authors: Moussou Pierre , Gibert Rene´ Jean , Brasseur Gilles , Teygeman Christophe , Ferrari Je´roˆme , Rit Jean-Franc¸ois
Published: July 2010

Pressure relief valves in water pipes are known to sometimes chatter when the inlet pressure slightly exceeds the maximum allowable working pressure (MAWP) value. Though these devices are responsible for numerous fatigue issues in process industries, there is a relatively low number of technical publications describing well-established facts about them, especially for heavy fluids. The present study is intended as a contribution to the understanding of pressure relief valve dynamics, taking into account fluid-structure interactions. A series of tests was performed with a water relief valve upon a test rig. Adjusting the MAWP of the valve to about 30 bars, an upstream pressure varying from 20 to 35 bars was imposed, so that the valve opened and the water flow varied from a few m3/h to about 80 m3/h. During the tests, the pipe was equipped upstream and downstream of the valve with static pressure sensors and a flowmeter, the disc lift was measured with a laser displacement sensor, and the spring force was recorded simultaneously. Several fluctuating pressure sensors were also arranged in the inlet pipe. Static instability is investigated by comparing the spring force to the hydraulic force. Dynamic instability is observed and it is shown that the resonant behavior of the disc generates an apparent negative pressure drop coefficient at some frequencies. This negative pressure drop coefficient can trigger a dynamic instability in a manner similar to the negative damping effect in leakage-flow vibrations.
The Damage of a Plunger Operating in Carbamide Environment
Authors: Zaha´lka Frantisˇek , Houdkova´ Sˇa´rka , Kasˇparova´ Michaela
Published: July 2010

The surface of plungers, operating in demanding conditions, is exposed to the effects of pumped media, which may be chemically aggressive, in combination with mechanical wear caused by the interaction between the plunger, the counterpart and possibly the solid particles included in the pumped media. To increase of plunger’s lifetime, their surfaces can be protected by different technologies of a surface treatment. One of them is the progressive technology of thermal spraying. A wide range of sprayed materials is one of the reasons for using these versatile coatings as the protection against corrosion, wear or high temperature exposition in order to increase the parts lifetime, reliability and safety. In this case, the TAFA Praxair JP-5000 spraying equipment was used for the creation of the WC-based hardmetal coating on the plungers operating in carbamide environment. Even though the spraying process and the check sample were without any sign of defects the plungers showed a critical damage after several hours of running. The damage mechanism, its possible causes and the proposed solution are discussed in this paper.
The Effect of Transient Operation on Gasket Contact Pressure in Steam Generator Primary Side Obround Manway
Authors: Idvorian Nick , Sun Nansheng
Published: July 2010

The Steam Generators used in CANDU Nuclear Power Plants employ a primary-side obround manway design with two cover plates on each side of the opening. During transient operation of the units (heat up, start up, operation, shut down and cool down) a complex interaction between the manway components occurs, which not only causes time-varying gasket contact pressure but also induces a non-uniform gasket contact pressure along the gasket width. As a result, the leak tightness of the joint may be affected. The effect of different stud preloads on the gasket contact pressure during transient operation is investigated by the 3D FE Analysis of a primary-side obround manway of CANDU’s steam generator. The results are used to justify the seal performance of the manway structure under conditions of different stud pretension loads.
Finite Element Simulation of Laser Beam Welding Induced Residual Stresses and Distortions in a T-Joint Configuration for Aeronautic Structures
Authors: Zain-ul-abdein Muhammad , Ne´lias Daniel , Jullien Jean-Franc¸ois , Deloison Dominique
Published: July 2010

Laser beam welding has found its application in the aircraft industry for the fabrication of fuselage panels in a T-joint configuration. However, the inconveniences like distortions and residual stresses are inevitable consequences of welding. The effort is made in this work to experimentally measure and numerically simulate the distortions induced by laser beam welding of a T-joint with industrially used thermal and mechanical boundary conditions on the thin sheets of aluminium 6056-T4. Several small scale experiments were carried out with various instrumentations to establish a database necessary to verify the simulation results. Finite element (FE) simulation is performed with Abaqus and the conical heat source is programmed in FORTRAN. Heat transfer analysis is performed to achieve the required weld pool geometry and temperature fields. Mechanical analysis is then performed with industrial loading and boundary conditions so as to predict the distortion and the residual stress pattern. A good agreement is found amongst the experimental and simulation results.
Considerations of Alloy 617 Application in the Gen IV Nuclear Reactor Systems: Part I—Mechanical Property Challenges
Authors: Ren Weiju
Published: July 2010

Alloy 617 is currently considered as a leading candidate material for high temperature components in the Gen IV Nuclear Reactor Systems. Because of the unprecedented severe working conditions beyond its commercial service experience required by the Gen IV systems, the alloy faces various challenges in both mechanical and metallurgical properties. This paper, as Part I of the discussion, is focused on the challenges and issues in the mechanical properties of Alloy 617 for the intended nuclear application. Considerations are given in details in its mechanical property data scatter, low creep strength in the desired high temperature range, lack of long-term creep curves, high loading rate dependency, and preponderant tertiary creep. Some research and development activities are suggested with discussions on their viability to satisfy the Gen IV Nuclear Reactor Systems needs in near future and in the long run.
Determination of Residual Stress Profiles of Pipe Girth Weld Using a Unified Parametric Function Form
Authors: Teng Huaguo , Bate Steve K. , Beardsmore David W.
Published: July 2010

In this paper we present an improved analysis of residual stress data of a pipe girth weld by applying the developed heuristic method to one set of high-quality residual stress measurement data. The through-thickness residual stress is expressed as a parametric function form which is a combination of three stress components: membrane, bending and self-equilibrating. This parametric function form provides not only a clear physical basis for the residual stress profile, but is also closely related to two important governing parameters, i.e. the pipe geometry and the welding heat input. The residual stress profiles obtained are also compared with results predicted by the Bayesian method as well as the profiles from the UK R6 procedure and the US API 579 code.
Analysis of Contact Pressure Distribution on 3-Bolt Self-Energized Connector Seals
Authors: Madazhy Rajeev , Mathews Sheril , Howard Erik
Published: July 2010

A novel design using 3 bolts for a self-energized seal connector is proposed for quick assembly applications. Contact pressure distribution on the surface of the seal ring during initial bolt-up and subsequent operating pressure is analyzed for 3″ and 10″ connectors using Finite Element Analysis. FEA is performed on a 3″ and 10″ ANSI RF flange assembly and contact pressure distribution on the RF gasket is compared with the tapered seal ring assemblies. Hydrostatic tests are carried out for the tapered seal and ANSI bolted connectors to evaluate maximum pressure at which leak occurs for both size assemblies.
Experimental Study of Mode II Fracture of Hybrid Composite and Metal-Wire Joints
Authors: Kwon Y. W. , Schultz W. A. , Loup D. C. , Rasmussen E. A.
Published: July 2010

In order to join a composite structure to a metallic structure, metal-wire mats were co-cured with composite laminas using the Vacuum Assisted Resin Transfer Molding (VARTM) technique. Then, interface fracture toughness was measured for Mode II fracture for various lay-up and interface conditions. The study included interfaces of metal-wire to composite, composite to composite, and metal-wire to metal-wire. In addition, the lay-up orientations of metal-wire mats were either 0 or 90 degrees. The study also examined the crack propagation from a composite to a metal/composite interface. During the test, the digital image correlation technique was applied to capture the strain field around the crack tip. The results suggested that a hybrid metal-wire/composite laminate would be effective to connect a composite structure to a metallic structure as long as some critical interface conditions were avoided.
Shakedown Analysis of Axisymmetric Nozzles Under Primary and Secondary Cyclic Loads
Authors: Vlaicu Dan
Published: July 2010

In this paper, the finite element method is used to develop the lower bound limit for the elastic shakedown analysis of axisymmetric nozzles under periodic loading conditions. The Nonlinear Superposition Method is employed to calculate the lower bound shakedown loads by quoting Melan’s theorem in a nonlinear finite element analysis. The calculation is divided into two separate iterations which are blended with a technique that matches the elastic-plastic part of the analysis with the linear part. In the first part of the calculation, the cyclic load is applied as a static load to generate an elastic stress field in the structure. The same cyclic load is subsequently combined with the constant fraction of the load in the second part of the calculation, and the total load is applied in an elastic-plastic analysis that exceeds the yield limit. For each solution increment, the residual stress is generated from the superposition of the elastic stress field scaled through the applied cyclic load and the shakedown stress field calculated from the nonlinear analysis. The results obtained from the lower bound method are compared with the full cyclic loading analyses based on nonlinear material properties, and this paper discusses the choice of the global shakedown in terms of the radial strain, and the local through thickness shakedown defined by the hoop strain. Furthermore, this paper presents the development of a generic model that emulates the behavior of the finite element model under cyclic loads in a simplified form, with the statistical representation based on a sampling of base-model data for a variety of test cases. The probabilistic method takes variations of the geometrical dimensions, nonlinear material properties, and pressure load as the input parameters, whereas the response variable is defined in terms of the lower bound of the shakedown loads.
Reliability Assessment of Piping With Cracks Using Advanced First-Order Second-Moment Method
Authors: Yoshida Shinji , Machida Hideo
Published: July 2010

This paper describes applicability of the 2 parameter assessment method using a reference stress method from the viewpoint of reliability. The applicability of the reference stress method was examined comparing both the GE-EPRI method. As a result, J-integral and limit load at the time of fracture evaluated by the reference stress method is almost equivalent to that by the GE-EPRI method. Furthermore, the partial safety factor (PSF) evaluated by reliability assessment has little difference between two methods, and the required safety factor is enveloped by the safety factor for Service Level-A and B defined in fitness for service (FFS) codes. These results show that of the reference stress method is applicable for J-integral calculation in fracture assessment.
FSI-Based Assessment of FAC-Caused Wall Thinned Piping
Authors: Kim Sun-Hye , Chang Yoon-Suk , Kim Young-Jin
Published: July 2010

Lots of investigations on failures of wall thinned piping have been carried out since the accident of Surry unit 2 in USA. From these preceding efforts, flow accelerated corrosion (FAC) which is a kind of wall thinning phenomenon is revealed main factor of failure of pipes in nuclear power plants. However, there are a few researches which directly take into account of flow characteristics and geometric changes for stress assessment of FAC-caused wall thinned piping. In this paper, structural integrity assessment employing a fluid-structure interaction (FSI) analysis scheme is performed on pipes representing secondary piping system of PWR which consists of straight pipes and elbows of various bend angles. Prior to the assessment, CFD analyses are conducted to predict plausible wall thinning location by considering flow and geometric parameters such as bend angle and radius of elbow. Then, for typical pipe geometry, detailed limit load analyses are performed to calculate maximum stress caused by turbulence and velocity of flow near the wall thinned part. Through these kinds of detailed parametric analyses, effects of FSI were observed, which should be considered for assessment of FAC-caused wall thinned piping.
Nuclear PWR Plants Mitigation of Reactor Coolant System Areas to Consider Plants Aging Management and/or Life Extension
Authors: Thevenet Re´mi
Published: July 2010

Eliminating the risk of Alloy 600 Primary Water Stress Corrosion Cracking (PWSCC) in any part of the Reactor Coolant System (RCS) is a major challenge to allow facing Plants aging and even consider a possible life extension. Many of the most sensitive areas of the RCS have already faced PWSCC and remediation was applied, thanks to repairs or main components replacements-Steam Generators Replacement (SGR) against A600 tubing cracks, Reactor Vessel Head Replacement (RVHR) against A600/182/82 CRDM nozzles cracks, Pressurizer replacement (PZRR) against cracks in heaters nozzles, Surge line Dissimilar Weld (DMW) and/or Safety/Relief lines DMWs. Other areas, expected to support longer production time without PWSCC risk, may also become sensitive while considering life extension. Additional inspections are now required by National Authorities in order to secure this assessment. Most of this “new” sensitive areas are located where it is hardly foreseen to replace the affected component (Reactor Vessel, for instance), in terms of technical feasibility or economical consideration. For these areas, Mitigation techniques have been developed in order to eliminate the risk of PWSCC. Two basic processes are employed: • Lowering the sensitive material tensile stresses (residual and in-service), under a limit where the material, in contact with the primary water, should not become sensitive; • Isolating the sensitive material from the primary water with a coating made from non-sensitive material. The presentation will list and describe the Mitigation processes, already applied, as well as those potentially available, or under development for most of the RCS A600 areas under consideration, which are not presently managed through Repair or Replacement. Examples of worldwide actual applications will be given. The presentation will conclude on the limits of these Mitigation techniques, when — even if economically preferred — they should be replaced by “classical” remediation, when these are technically achievable.
Analytical Study of the Relaxation of Welding Residual Stress by Excessive Loading for Austenitic Stainless Steel Piping Welds
Authors: Katsuyama Jinya , Onizawa Kunio
Published: July 2010

Welding residual stress is one of the most important factors of stress corrosion cracking (SCC) for austenitic stainless steel in pressure boundary piping in nuclear power plants. The effect of excessive loading, such as an earthquake, on the residual stress was evaluated by three-dimensional analyses based on finite element method (FEM). The FEM analyses were performed using three-dimensional model for a 250A piping butt weld of low carbon stainless steel of Type 316L. A welding simulation method used in this work is based on the moving heat source with the double ellipsoid model and was confirmed by comparing with the experimental measurements. After conducting welding residual stress simulation, several loading patterns of bending moment and uni-axial displacements have been applied to a model by varying amount of moment and displacement. The analyses indicated that higher loading to bending and axial stresses caused higher relaxation of welding residual stress near piping welds. The difference in the effect of loading direction was observed for both cases. It is concluded that the SCC growth rate might be decreased as loading level increased.
Abrasive-Waterjet Drilling of High Temperature Jet Engine Parts
Authors: Hashish Mohamed
Published: July 2010

An experimental investigation was performed to demonstrate the AWJ process capability for precision drilling of small diameter cooling holes in TBC-coated samples and full scale parts used in jet engines. All holes were drilled from the TBC side. Both qualitative and quantitative hole results issues were addressed. These include chipping of TBC, gouging inside the hole at the ceramic-metal interface, feathering at the exit side of the hole, tear drop shape, hole size and its consistency, overall drilling time, and potential systems productivity. It was observed that the AWJ produces high quality holes free from chipping or gouging. However, feathering may occur depending primarily on whether the substrate material is cast or rolled and can be substantially reduced or eliminated by adjusting the drilling parameters. It was also demonstrated that holes of about 0.025-ich can be drilled at 30 with less than 0.001-inch standard deviation. Air flow measurement results have also confirmed that less than 6% variation can be obtained from one set of holes (270 holes) to another. Drilling times below 10 seconds hole to hole were observed for drilling 0.060-inch thick metal substrates with 0.020-inch thick TBC thickness at 30 degrees. This time can be reduced to 3–4 seconds with improved systems. Although AWJ is still slower than laser, produced holes are of superior quality and do not need any additional processing. Future research efforts will focus on improving the drilling time.
Extension of the Simplified Method for the Interaction of Primary and Secondary Stresses
Authors: James P. M. , Hooton D. G. , Madew C. J. , Sharples J. K. , Jackson M.
Published: July 2010

The R6 defect assessment procedure, used commonly in the UK nuclear industry to assess the significance of defects in structures, uses the Failure Assessment Diagram to evaluate limiting parameters whilst accounting for the effects of plasticity. The interaction of primary and secondary stress is accounted for within R6 through the use of the ρ, or an equivalent V, term. Previous work has developed an alternate method to that in R6 and has been termed the “Simplified Method”. This Simplified Method has the benefit of being less conservative than the current R6 method and of not requiring ρ (or V factors, and hence not requiring the use of the ‘Look-up’ tables. This paper builds upon the work presented previously and considers the reduction in displacement controlled secondary stress through primary stress induced plasticity by proposing a new function, g(), which is based upon an Option 2 failure assessment curve. Comparisons are also made upon a range of cracked geometries using both the current R6 Method and the derived Simplified Method including g(). The comparisons are made upon both a thermally induced bending stress and a weld residual stresses obtained from a detailed weld simulation. The range of cases provided is to further validate the method and provide strong evidence not only for different geometries but also for realistic residual stress fields. The application of the Simplified Method to the thermally induced bending stress field show improved estimated of KJ over the R6 Method and provide added value to both the Simplified Method and g(). Application to the weld residual stress field shows potential non-conservatism (of both the R6 and Simplified Method) and potential reasons are discussed.
Prediction of Liquid Droplet Impingment Erosion (LDI) Trend in Actual NPP
Authors: Ohira Taku , Morita Ryo , Tanji Kazuhiro , Akiba Shinji , Niiyama Koichi , Inada Fumio , Yoneda Kimitoshi
Published: July 2010

Liquid droplet impediment erosion (LDI) is a pipe wall thinning phenomena that a droplet accelerated by steam flow attacks pipe surface. It is difficult to evaluate LDI behavior in the system of NPP. Therefore, in current pipe wall thinning management, the pipe element is replaced conservatively when it is recognized to become thinner by pipe wall thinning measurement result. If LDI behavior at each point of the system can be estimated, it is possible to measure the pipe wall thickness and replace the pipe element at an appropriate time, in accordance with reasonable pipe wall measurement schedule. CRIEPI has constructed the evaluation system for LDI by using flow dynamics. This paper is concerned with the following items: • the applicability of the LDI evaluation system by a comparative verification between pipe wall thinning rate calculated by LDI system and that by actual measurement result; • establishment of remaining life assessment method of pipe in LDI environment by using LDI model.
Seismic Waves Attenuation With Distance in the Bohemian Massif
Authors: Maly Jan , Prochazkova Dana
Published: July 2010

For ensuring the seismic safety of nuclear facilities there is very important to know not only the size of earthquakes or other induced seismic phenomena in the foci, but also the attenuation of seismic energy with a distance in order that it was possible to determine destructive potential of phenomena in a given site for nuclear facility. The paper presents relationships for seismic energy drop with a distance derived from macroseismic and instrumental data.
VERLIFE: Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs During Operation—Updating and Further Development
Authors: Brumovsky Milan
Published: July 2010

VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program project “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure has to fill the gap in original Soviet/Russian Codes and Rules for WWER type NPPs, as these codes were developed only for design and manufacture and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. To assure that VERLIFE Procedure will remain a living document, new 3-year IAEA project (in close co-operation with another project 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) has started in 2009. Within this project, upgrading/updating of the VERLIFE procedure is prepared together with the extension by following procedures: - Lifetime of reactor pressure vessel internals. - Leak-before-break concept for WWER NPPs. - No-break-zone for WWER NPPs. - Probabilistic approach for failure and lifetime assessment of WWER reactor pressure vessels and primary piping (including Risk-informed ISI). - Thermal fatigue evaluation. Final document, after its approval by expert groups of the IAEA and NULIFE, will be issued as “IAEA Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs”. The paper will describe these main principles and also future plans.
Good Practice Handbook for PTS Evaluation (Result of the IAEA Co-Ordinated Project on PTS)
Authors: Pistora Vladislav , Brumovsky Milan , Taylor Nigel
Published: July 2010

Integrity and lifetime of reactor pressure vessels are practically determined by their behavior during “pressurized thermal shock” (PTS) emergency regimes as the most severe regimes during reactor operation. Assessment of these potential regimes is carried out mostly in deterministic way but used procedures are different in different countries. Proper and reliable evaluation of these PTS regimes depends on many parameters and approaches used during computations. During the period 2005 – 2008, the Coordinated Research Project 9 (CRP 9) “Review and Benchmark of Calculation Methods for Structural Integrity Assessment of RPVs During PTS” was organised by the IAEA. The overall objective of this Coordinated Research Project was to perform benchmark deterministic calculations of a typical pressurised thermal shock (PTS) regime and finally to recommend the best practice for PTS assessment. This paper describes main results and collected experience within this project that were bases for the preparation of the “Good Practice Handbook for Deterministic Evaluation of the Integrity of a Reactor Pressure Vessel during a Pressurised Thermal Shock” that will be issued as an IAEA TECDOC. Main parameters discussed in this handbook are: - selection of overcooling sequences; - thermal-hydraulics analyses; - temperature and stress field calculations; - crack tip loading incl.K estimations; - integrity assessment; - analyses of nozzles; - national practices; - results from sensitivity studies. Finally, recommendations for reliable and correct PTS evaluation are given.
A Preliminary Evaluation of a Generation IV Liquid Metal Fast Reactor Due to Sloshing and Thermal Stress
Authors: Vezzoni Barbara , Lo Frano Rosa , Forasassi Giuseppe
Published: July 2010

The aim of this paper is a preliminary evaluation of the load bearing capability of a generation IV molten metal cooled reactor (LMFR), with reference, as an example, to the ELSY project in the event of a safe shutdown earthquake (SSE). Two relevant safety aspects were considered and analyzed by means finite element (FEM) codes: the fluid-structure interaction and the high thermal loads due to the residual heat levels of the possible high burnup fuel. The first issue depends on the presence of a liquid free surface that especially in seismic accident conditions, allows for fluid motion (“sloshing”) and may have a significant influence on the reactor structures stress levels. The second aspect deals with the irradiated core heat source thermal effects after the reactor shutdown. In order to evaluate the reactor and internal components structural response capability for both high temperature and hydrodynamic loads, appropriate 3-D models were set up. Moreover preliminary thermal analyses were carried out by means of MCNPX and Origen2.2 codes in order to determine the transient input decay power after shutdown. The obtained preliminary results allowed to individuate the structures mostly affected by the assumed loading conditions in order to make possible an upgrading of the design geometry, should any be necessary, for the considered reactor design. The mentioned activities was a part of a CE funded 6th FWP.
Improvements of the Containment Measuring System
Authors: Zdarek Jiri , Bulak Jan , Vomacka Petr , Brumovsky Milan
Published: July 2010

Proper pre-stressing of concrete containment is a strict requirement to assure its reliable and safe behavior during potential emergency conditions within the nuclear island. Thus, it was decided to start the project concentrated on experimental verification of measuring systems and methods applied during in-service inspection as well as during potential replacement of tendons. Aim of this project is creating of experimental facility for research of the specifics in loading system of pre-stressing of the containment in WWER-1000 NPP. This facility is determined for modeling of stress and strain state of pre-stressed tendons of the containment and also for a more precise determination of forces along their length. Results obtained during tests of pre-stressed tendon will be used for precision of computation of stress and strain state in the containment during its long term operation. These results will be also used in development and application of expert system for evaluation of the stress situation in containment for NPP with WWER-1000 reactors. This paper describes the experimental facility and working plan to ensure the required task.
Comparison of Pipe Bend Ratchetting/Shakedown Test Results With the Shakedown Boundary Determined via a Simplified Technique
Authors: Abdalla Hany F. , Megahed Mohammad M. , Younan Maher Y. A.
Published: July 2010

Scarce experimental verification exits in the open literature concerning determination of the shakedown boundary for pipe bends subjected to steady internal pressure and cyclic bending loading. The objective of the present paper is to test the capability of a simplified technique presented by the authors in recent ASME JPVT publications and PVP conferences [1–4] in adequately predicting the shakedown boundary obtained through experimental testing. Recently, Chen et al. [5] published experimental and finite element (FE) simulation results on ratchetting of low-carbon steel pressurized 90-degree pipe bend specimens subjected to cyclic reversed in-plane bending forces. Chen et al. [5] performed experimental testing on a pipe bend specimen subjected to a steady internal pressure magnitude of 20.0 MPa. Through FE simulations employing a modified form of the Ohno-Wang non-linear kinematic hardening (KH) rule, Chen et al. [5] predicted a shakedown boundary for a steady internal pressure spectrum ranging from 10.0 to 25.0 MPa. Chen et al. [5] experimental and FE outcomes are utilized for comparison with the simplified technique outcomes. The simplified technique outcomes showed very good correlation with Chen et al. [5] shakedown boundary predictions for the 18.0 – 25.0 MPa steady internal pressure spectrum. On the contrary, noticeable disagreement was found for the lower magnitudes of steady internal pressure. Reasons behind the discrepancy are discussed.
Prediction of Potential for the Ingression of Crude Oil Covering the Sea Water Surface Into the Sea Water Cooling System of Nuclear Power Plant
Authors: Jo Jong Chull , Roh Kyung Wan , Kim Hho Jhung , Kim Kyun Tae
Published: July 2010

When an incident of crude oil spill from an oil carrier occurs in the sea near the nuclear power plants, the spilled oil can be transported to the intake pit, where the NSCW (nuclear service cooling water) pumps locate, by sea current and wind drift (induced) current. The NSCW pumps take the essential service water from the sea being used as the ultimate heat sink and supply to the component cooling water heat exchangers to remove the decay heat generated from reactor. The NSCW system shall provide sufficient cooling capacity during the reactor normal operation, transients, and loss-of-coolant accidents (LOCAs). In this regard, it is very important to confirm if the cooling function of the NSCW system can be threaten due to spilled oil which may overspread the surface of sea water in the intake area. Thus, in this work, when an incident of crude oil spill occurs in the sea near a nuclear power plant using sea water as the ultimate heat sink, possibility of crude oil ingression into the component cooling water heat exchangers through the NSCW pumps has been evaluated in a conservative manner. To do this, for a flow field surrounding a NSCW pump equipped in an intake pit where a limited volume of sea water is initially contained and its surface is covered with spilled crude oil, a numerical simulation is performed using a CFD (Computational Fluid Dynamics) code. The objective of the present simulation is to find out the critical sea water level at or below which oil or air floating on the surface of sea water begins to be sucked in the pump inlet nozzle. The numerical simulation results show that in general, the oil covering the free surface of sea water in the intake can hardly flow into the pump inlet until the floating oil changes into tars which are heavier than oil as long as the sea water level does not fall below the limit value specified in the Technical Specification of each nuclear plant. It is also shown that the critical sea water level increases as the pump flowrate increases. This is physically plausible considering the effect of gravitational force.
PTS Evaluation for Czech Nuclear Power Plants of WWER Type
Authors: Pistora Vladislav , Kral Pavel
Published: July 2010

There are two NPPs operated in the Czech Republic: NPP Dukovany of WWER 440/213 type and NPP Temelin of WWER 1000/320 type. Four units of NPP Dukovany have started their operation during 1985 – 1987, two units of NPP Temelin have been put into operation during 2002 – 2003. Pressurised-thermal shock evaluation, i.e. assessment of resistance of reactor pressure vessel (RPV) against fast fracture, was for both NPPs performed during the plant operation. For NPP Dukovany the PTS evaluation started in 1996 and the project was completed in 2004. For NPP Temelin the PTS project was conducted during 2001 – 2004. At the beginning of the project for NPP Dukovany the PTS evaluation was performed according to the IAEA “Guidelines on PTS Analysis for WWER NPP” [1]. At later phases of this project and for the whole project of NPP Temelin, the appropriate PTS evaluations were performed according to the IAEA PTS Guidelines, Revision 1 [2] and the “Unified Procedure VERLIFE” [3]. The PTS analyses consisted of two main parts: 1) Thermal hydraulic (TH) analyses: The system TH analyses were performed mainly with using the RELAP5 computer code. Reactor coolant system, secondary circuit, and emergency core cooling system were modelled in a very detailed way according to the actual NPP configuration, signals specific to the NPP were built into the models and operator actions were taken into account for some PTS transients. For relevant cases, the behaviour of media in such components as hermetic confinement (NPP Dukovany), containment (NPP Temelin), sumps and ECCS coolers, was investigated by MELCOR or RELAP5 code with primary objective — to determine the exact temperature of water from safety injection system. The detailed TH mixing analyses were performed for all non-symmetric cases using CATHARE (with 2-D model of reactor downcomer) or REMIX/NEWMIX codes. 2) Structural Analyses: All computations were focused on circumferential welds in the beltline zone. The overall approach to the analyses (for the final stages of the projects) was as follows: Temperature and stress fields were computed by finite element method (FEM) on RPV model with crack included in the mesh. SYSTUS FEM code was used for this purpose, with elastic-plastic modelling of material properties. After performing the stress analysis, the energy release rate G was calculated using the postprocessor of SYSTUS code. Finally, the stress intensity factors KI were computed based on values of G. The defects were finally postulated as semielliptical underclad cracks, both axial and circumferential, with aspect ratios a/c = 0.3 and 0.7 and with the depth 1/10 of the wall thickness. Maximum allowable critical temperature of brittleness Tka was determined by comparison of KI and [KIC]3 (the allowable value of stress intensity factor) curves. Some examples of results are presented in the paper.
Study on Residual-Stress Redistributions During the Process of Manufacture of a Vessel Penetration Set-On Joint
Authors: Yanagida Nobuyoshi , Ogawa Kazuo , Saito Koichi , Kingston Ed
Published: July 2010

The stress-redistribution phenomenon in a vessel penetration set-on joint due to post-weld heat treatment (PWHT) was studied using finite element (FE) analyses and mocked-up experiments. The mocked-up consisted of a nickel-based alloy (NCF600) tube welded onto an alloy-82 cladded, low-alloy steel plate (SQV2A) using an alloy-182 butt weld. The angle of the tube to the plate surface was 45 degrees, simulating a side hill, a control rod drive (CRD), and a stub-tube nozzle attachment used in boiling-water reactor (BWR) plants. PWHT at a temperature of 625 °C was conducted after welding and then the inner surface of the tube was machined. Three-dimensional FE modeling was performed to simulate the cladding, the butt weld, the PWHT, and the inner-surface machining of the tube. Thermal elasto-plastic and thermal elasto-plastic creep analyses were conducted to simulate the process of residual-stress build up and its redistribution by PWHT. To validate the FE analysis, the residual stresses in the mocked-up specimen were experimentally measured using the deep-hole-drilling (DHD) and sectioning methods. The analytical and experimental results revealed that residual-stress redistributions in the mocked-up specimen were different in circumferential positions. High-residual stresses in the low-alloy steel plate were particularly mitigated during the PWHT. The stress relief in the low-alloy steel plate primarily controlled the global stress balance between the cladding, the weld metal, and the stub tube.
Numerical Simulation of Pressure Transients in a PWR Main Steam Line System Due to Quick Operations of Pressure Relief Valve
Authors: Jo Jong Chull
Published: July 2010

At a PWR (pressurized water reactor) nuclear power plant in Korea which was under being retrieved from the zero reactor power after a periodic overhaul in 2005, manual operation for opening the PORV (power operated pressure relief valve) was tried at a low speed to reduce pressure in the MSL (main steam line). Initially, one MSIV (main steam isolation valve) and five MSSVs (main steam safety valves) installed to the MSL remained all closed. To speed up the pressure reduction, the reactor operator in charge abruptly switched the operation mode from “Manual” to “Auto.” Just after the valve was opened fully in a very short time, rapid reduction of pressure in the MSL occurred. A little later on, the operator closed the valve quickly to mitigate the unexpected transient occurrence. In consequence of the result of the sudden discharge of high pressure steam to the atmosphere through the PORV, a curved pipe spool in the downstream was separated from both connections and blasted away into air. Then it collided with the outer wall of the RWST (refueling water storage tank) located about 50 m away, resulting in mechanical damage on it with the maximum permanent deformation of about 60 mm in the radial direction at a local part of wall. This paper describes a numerical simulation of the transient pressure responses to the quick opening and subsequent closing operations of the PORV on the PWR MSL initially maintained as a closed system at a high pressure, which has been performed to illustrate the incident scenario as well as to demonstrate its root cause. For computational efficiency, the pressure field to be solved has been modeled with a two-dimensional mathematical formulation. As the result, the simulation has shown realistically how the incident was initiated and developed.
Use of HDPE Piping in the Callaway Nuclear Plant Essential Service Water System
Authors: Abel Shannon L. , Brandes Matthew D. , Corley Lonnie J. , Fortman Joseph L. , Musto Thomas M. , O’Sullivan James
Published: July 2010

Callaway Nuclear Plant is the first nuclear plant in the United States to utilize high density polyethylene (HDPE) piping in a nuclear safety-related application. HDPE is being installed in buried sections of the plant’s ASME Section III, Class 3 Essential Service Water (ESW) system. Due to its resistance to erosion, corrosion and microbiologically induced corrosion (MIC), HDPE is well suited to raw water system applications. As with any other first of a kind project, the use of HDPE piping in the Callaway ESW system has presented challenges in all phases of the project. Design, qualification and installation considerations for thermally-fused HDPE in an ASME Class 3 system differ significantly from those for traditional metallic materials. This paper will examine the challenges and lessons learned in the design, qualification testing, installation, examination and pressure testing of the HDPE piping at Callaway Nuclear Plant.
High Pressure Testing Hazards
Authors: Edel Matthew , Ketchum Donald , Falcon Jasen
Published: July 2010

Pressure vessels that operate at high pressures (greater than 10,000 psi [69 MPa]) pose several potential hazards to nearby personnel. Some of these hazards include projectile launch, water jetting, and blast loads that may occur as a result of a pneumatic or hydrostatic pressure vessel failure. In order to provide personnel protection, pressure vessels may be placed inside hardened test enclosures designed to contain these hazards. Some of the key response mechanisms that should be considered when designing such enclosures include applied blast loads for pneumatic testing), localized barrier perforation, global or gross barrier response, and generation of secondary debris from damage to a barrier, such as back-face spalling. Some of these hazards and shielding responses have been evaluated experimentally in the High Pressure Testing Safety Research Joint Industry Program [1, 2, 3]. This paper describes some of these hazards and some considerations for providing personnel protection.
Mechanical Characteristic of Multiple Direction Optimized-Friction Pendulum System
Authors: Tsai C. S. , Lin Yung-Chang , Su H. C.
Published: July 2010

Presented in this paper is the detailed mechanical characteristic of the multiple direction optimized-friction pendulum system. Based on the equilibrium equations, mathematical formulations describing the sliding motions of the multiple direction optimized-friction pendulum system have been derived. The mechanical behavior of the multiple DO-FPS isolation system under cyclic loadings has been investigated. The parametrical study for the multiple DO-FPS isolation bearing has also been carried out to realize its various characteristics under different parameters. It is demonstrated that the natural period and damping effect of the multiple DO-FPS isolator are a function of the sliding direction, displacement and coefficient of friction.
Heterogeneous Microstructure Effect on Residual Stress and Fatigue Crack Resistance in Dual-Phase Materials
Authors: Mochizuki Masahito , Mikami Yoshiki
Published: July 2010

The effect of transformation-induced microscopic residual stress on fatigue crack propagation behaviour of ferrite-martensite lamellar steel was discussed. Fatigue tests of prestrained and non-prestrained specimens were performed. Inflections and branches at ferrite-martensite boundaries were observed in the non-prestrained specimens. On the other hand, less inflections and branches were found in the prestrained specimens. The experimental results showed that the transformation induced microscopic residual stress has influence on the fatigue crack propagation behaviour. To estimate the microscopic residual stress distribution, a numerical simulation of microscopic residual stress induced by martensitic transformation was performed. The simulation showed that compressive residual stress was generated in martensite layer, and the result agree with the experimental result that inflections and branches were observed at ferrite-martensite boundaries. In addition, the change in the microscopic residual stress distribution by prestraining was also calculated to show the compressive residual stress changed to tensile by prestraining. This also agree with the experimental result of the observation of fatigue crack path.
Influence of Symmetrical Boundary Conditions on Free Vibration of Functionally Graded Cylindrical Shells With Ring Support
Authors: Isvandzibaei M. R. , Najafizadeh M. M. , Khazaeinejad P.
Published: July 2010

In the present work, the free vibration of thin cylindrical shells with ring support made of functionally graded materials under various symmetrical boundary conditions is presented. Temperature and position dependent material properties are varied linearly through the thickness of the shell. The functionally graded cylindrical shell has ring support which is arbitrarily placed along the shell and imposed a zero lateral deflection. The third order shear deformation theory is employed to formulate the problem. The governing equations of motion are derived using the Hamilton’s principle. Results are presented on the frequency characteristics and influence of the boundary conditions and the locations of the ring support on the natural frequencies. The present analysis is validated by comparing the results with those available in the literature.