bluuebunny/crossref_metadata_2025_split huggingface dataset:
Study of Crack Propagation Under Fatigue Equibiaxial Loading
Authors: Bradaï S. , Gourdin C. , Gardin C. 
Published: November 2014

In some nuclear power plant components, the fatigue loading may be equibiaxial because of thermal fatigue. As a consequence, the potential impact of multiaxial loading on the fatigue life of components is a major concern. However, few experimental data are available about austenitic stainless steels. It is essential to improve the fatigue assessment methodologies taking into account the potential equibiaxial fatigue damage. It implies obtaining experimental data on the considered material, with a strain tensor corresponding to equibiaxial tension. Many multiaxial criteria, based on thermal fatigue tests or uniaxial fatigue tests, were proposed by Von Mises, Zamrik and other researchers [cited in 1]. Yet, additional thermal fatigue tests and isothermal equibiaxial fatigue tests must be carried out. In this context, a new experimental fatigue device FABIME2 [cited in 2] is developed, in the LISN, in collaboration with EDF and AREVA, in order to characterize accurately the possible equibiaxiality effect on the fatigue behavior of the 316L austenitic stainless steel.
This paper is focused on the identification of a new criterion in order to be able to take into account the possible effect to the equibiaxial loading on fatigue life. It will be identified by a comparison of experimental results and numerical predictions. In this aim, a test campaign with several experiments was realized on the new FABIME2 device, in order to determine the crack growth rate from an equibiaxial fatigue loading. Propagation tests were carried out on 316L austenitic stainless steel CT specimens, at room temperature, in order to obtain the Paris law [cited in 3]: (crack growth rate da/dN versus the applied stress intensity factor range ΔK). The crack growth is followed throughout equibiaxial test. This allows obtaining, the da/dN evolution, and the corresponding ΔK values. A finite element modeling done with CASTEM®, will be proposed to simulate the crack propagation in the FABIME2 specimen using a 3-dimensional approach, considering a semi-elliptical crack shape.
Cleavage Crack Path Prediction in a PWR Vessel Steel
Authors: Yang Xiaoyu , Marie Stéphane , Jacquemoud Clémentine 
Published: November 2014

Cleavage crack propagation has been tested for three different geometries of Compact Tensile (CT) specimens: CT25, CT50 and extended CT25 (CT25 with CT50 width) (Figure 3). The experimental results show that the crack paths are straight for CT25 and CT50, but they are unstable and curved for extended CT specimens (Figure 5 to 7).
Numerical computation had been performed by extended finite element method (XFEM) in CAST3M software. 2D modeling was used in order to predict the crack path. The analysis was based on a local non-linear dynamic approach with a RKR fracture stress criterion depending on temperature and strain rate.
In order to simulate the curvature of the cracks path, a statistical effect was introduced in the model to take into account the spatial distribution of cleavage initiators, which is the characteristic of cleavage fracture. At each step of propagation during the modeling, the direction was randomly chosen, according to a uniform defects distribution.
The numerical results show a good agreement with experience. The different crack paths were curved in extended CT25, but remained almost straight in CT25 and CT50 specimens, despite of the instability introduced in the modeling in the propagation direction. These results show that the statistics of micro-defects can induce, jointly with the geometry of specimen, a large scatter of crack propagation paths.
Process Optimization of Ultrasonic Shot Peening by Realistic Modeling of Beads Motion
Authors: Brosse Alexandre , Le Moal Simon , Lacroix Rémi , Mourgue Philippe , Robin Vincent , Gilles Philippe 
Published: November 2014

In the nuclear industry surface mechanical treatments are used in order to improve the surface integrity of the component, which increases their lifetime regarding corrosion and fatigue damages. A good understanding of these processes and their consequences is required to ensure the efficiency and perpetuity of such mitigation treatment.
This study focuses on the ultrasonic shot peening process. It consists in shooting at high speed small steel beads on the part to be treated by using a high frequency vibration device. Parameters such as the number and the size of beads, the input frequency and the dimensions of the chamber can induce large ranges of impact velocity and coverage. In order to help manufacturers to control the treatment applied on their components, a numerical model has been developed. It accounts for the shocks of the beads against the walls of the chamber, the peening head and between beads, describing their motions accurately.
In this paper, we will introduce the numerical model developed to simulate the motions of beads in the peening chamber. Special attention will be taken to the determination of the restitution rates related to the different materials. Results of the model will be shown for different process parameter (e.g. the number of beads), and a thorough analysis of their effects on the workpiece will be presented, including a comparison with some experimental results.
Internal Crack Growth Simulation Using S-Version FEM
Authors: Kikuchi Masanori , Yamada Shinya , Serizawa Ryosuke , Yulong Li 
Published: November 2014

In nuclear power plant, there is a proximity rule to evaluate multiple inner cracks [1]. Inner cracks are generated inside of the structure in different manners. There are many parameters which affects the growing processes of inner cracks. They are; locations, slant angles, aspect ratio of each inner crack and distances between adjacent inner cracks. When multiple inner cracks are detected, proximity rules are proposed. But due to the complexity of the problem, it is necessary to verify proposed proximity rules. But experimental study is very difficult due to existence of many parameters, and crack growth occurs inside of the structure. Numerical simulation is needed for this purpose.
This problem is simulated using S-version FEM [2]. Using S-FEM, inner crack is modeled independently from global structure, and crack growth is easily simulated. In maintenance code of nuclear power plant, initial defects are modeled as elliptical cracks in a normal plane to tension loading direction, and growth rate is estimated on this plane. But by using S-FEM, realistic defect shape is modeled, and crack growth by fatigue is simulated. Usually, such small defects are subjected to multi-axial loading, and crack growth behaviors are very complicated. Finally, detect shape becomes elliptical or circular crack in a plane normal to tension loading direction in the structure. Fatigue cycles for these growing processes are calculated, and conservativeness of this maintenance code is discussed. Parametric studies are conducted for this problem, and proximity rules are verified with numerical results.
Estimation of Crack Growth Rate by J-Integral Range for Type SUS304 Stainless Steel Plate in Creep-Fatigue Tests at High Temperature
Authors: Funai Madoka , Watanabe Osamu , Matsuda Akihiro 
Published: November 2014

In structures having stress concentration under cyclic loading, a small crack initiates and it grows and propagates. Evaluating crack growth is important to estimate the remaining life of cracked components.
The present paper shows the estimation of crack growth rate under creep-fatigue loading with some patterns of strain holding times.
Creep-fatigue tests of the perforated plate having initial crack were conducted with the different strain holding time under strain-controlled loading at 550°C. The crack growth was observed from the photographs taken at each cycle. The crack growth per one cycle of creep-fatigue loading was evaluated by the creep-fatigue crack propagation law which used increment J-integral range. Comparing the crack growth rates of experimental with those of predicted, the crack growth rates were predicted by using increment J-integral range with the accuracy was factor of 2.
Study on Strength of Thin-Walled Tee Pipe for Fast Breeder Reactors Under Seismic Loading
Authors: Watakabe Tomoyoshi , Tsukimori Kazuyuki , Otani Akihito , Moriizumi Makoto , Kaneko Naoaki 
Published: November 2014

In recent years, earthquakes over design condition were observed in Japan. Confirming the ultimate strength and design safety margin of mechanical components is important for the seismic integrity. This study focused on piping components, and it was one of the most important mechanical components for protecting boundary of coolant. Failure tests of thick-walled piping components for Light Water Reactors (LWRs) described previously in the literature. According to these tests, the failure mode of thick-walled piping components under seismic cyclic loading was low cycle fatigue. However, failure tests have scarcely been performed on thin-walled piping components pressurized at low levels for Fast Breeder Reactors (FBRs).
This paper presents dynamic failure tests of thin-walled piping components in FBRs. Based on the test results, the failure mode, the ultimate strength, and the elastic-plastic behavior are discussed.
Statistical Analysis of Seismic Effects of Low Seismic Class Equipment Based on Damage Data of Nuclear Power Plants
Authors: Morita Ryo , Inada Fumio , Ohtori Yasuki , Nanpo Mitsuhide , Naradate Koji , Yamaguchi Syuhei , Takeuchi Masataka , Yamaguchi Tatsuya , Numata Takeshi , Tamura Ichiro , Hosotani Terushige , Kimura Yusuke , Amemiya Mitsuhiko , Taguchi Toyonobu , Fukushi Naomi , Yamaguchi Atsushi , Kojima Nobuyuki 
Published: November 2014

Structural strengths of the piping and components in NPPs have been designed with seismic margin. They are classified seismically S, B and C class in terms of the influence rate to nuclear safety. For the highest seismic class (Class S) equipment, it is clarified that they have enough seismic margins against design seismic conditions by shaking table tests or numerical simulations. However, for the lower seismic class (Class B and C) equipment, their seismic margins have not been clarified quantitatively.
In this paper, in order to evaluate seismic robustness of the lower seismic class equipment with no clarification of seismic margin, seismic influences of the lower seismic class equipment in NPPs damaged by actual large earthquakes have been surveyed and sorted as a database, and the integrity of the lower class equipment have been discussed.
Seismic effects on 24 plants damaged by the recent large 6-earthquakes are surveyed, sorted as a database, and investigated. As a result, a total of 29 cases of function deterioration or loss were observed. Considering the total number of components and piping, the frequency of those cases in class B and C components and piping was low. And also, as it is found there are a few cases of degradation or loss of function in the equipment installed on the bedrock or in the buildings.
Reheat Cracking Sensitivity of CGHAZ in Vanadium-Modified 2.25Cr1Mo Welds
Authors: Han Yichun , Chen Xuedong , Fan Zhichao , Bu Huaquan , Zhou Yu 
Published: November 2014

Over the past few years, some reheat cracks have been detected in the welding coarse grained heat affected zones (CGHAZ) of Vanadium-modified (V-mod) 2.25Cr1Mo reactor. Focusing on this problem, a series of Gleeble high temperature tensile tests were carried out on the simulation CGHAZ specimens of V-mod 2.25Cr1Mo at different temperature and strain rate, the reheat cracking sensitivity of CGHAZ has been assessed according to the reduction of area (RoA) of the specimens. Microstructural observation, hardness test and fractographic analysis also have been employed to ascertain possible causes for the effects of temperature and strain rate on the reheat cracking sensitivity.
The results of tests show that the reheat cracking sensitivity of CGHAZ increases with a decreasing strain rate, it first increases and then decreases with a rising test temperature. The RoA presents a minimum value around 675°C, corresponding to the most sensitive temperature to reheat cracking. The comprehensive analysis reveals that the ductility of the simulated CGHAZ at elevated temperature is closely related to the tensile strength and the hardness of fractured tensile specimen. In the intermediate temperature range or at a slower strain rate, a relatively high hardness and relatively low tensile strength together results in a low RoA of the specimen and high reheat cracking sensitivity of CGHAZ metal.
The Rocking and Stability of Spent Fuel Casks
Authors: Ananth Raju , Day Shari , Bose Mrinal 
Published: November 2014

This paper describes an analytical and numerical study performed to benchmark LS-DYNA computer analysis of the response of a stacked spent fuel cask system to seismic base excitation. The LS-DYNA solution of the rocking block problem and stability of the block under the action of a base acceleration pulse are compared to the inverted pendulum solution first published by George W. Housner in 1963. Housner’s work has been cited by a number of subsequent investigations exploring the physics behind the rocking block problem, including Yim and Chopra [2]. The analytical formulation of the rocking block problem is derived in 2D in this paper. As is the case in many of the previous formulations, the solutions are provided after linearization of the governing equations.
The stability of the unanchored block subjected to an acceleration pulse is solved using the energy method. The solution displays key differences relative to Housner’s minimum acceleration required to overturn. An alternate equation for effective viscous damping is also presented, which differs from the formulation given in Appendix A of ASCE 43-05 [9]. Simulation of the rocking block problem using LS-DYNA is shown to faithfully reproduce the classical solution, albeit at lower levels of energy loss. Effective damping and minimum acceleration to overturn from the LS-DYNA benchmark agree with predictions of the classical solution even after linearization.
Validation of Thermal Stress Ratcheting Criteria
Authors: McKillop Suzanne , Reinhardt Wolf , Reddy Vangala , Koski William 
Published: November 2014

Design of components against incremental deformation or “ratcheting” under cyclic loading conditions is addressed in Article NB-3200 of Section III of the ASME Boiler and Pressure Vessel Code. The ratcheting rules, based on the Bree diagram, relate primary stress and secondary stress ranges that are calculated elastically and aim to approximate elastic-plastic material behaviors under cyclic loading. The Bree diagram was developed for cases with through-thickness thermal bending and constant primary membrane stress. It does not account for high thermal membrane stress that can occur near gross structural or thermal discontinuities. Cyclic thermal membrane stress combined with sustained stress can lead to ratcheting that is not accounted for in the current design rules.
This paper discusses the validation of proposed criteria for evaluating thermal stress ratcheting under high thermal membrane stress using an elastic analysis. These proposed criteria are confirmed by an analysis of a nozzle that is attached by a partial penetration weld to a vessel head and subjected to severe thermal cycling. Linearized stresses from an elastic analysis under pressure and thermal loadings typical for a nuclear power plant are compared to limits in NB-3200 for thermal stress ratcheting. Additionally, an elastic-perfectly plastic analysis is used to evaluate if the component will shakedown.
This analysis demonstrates that the proposed rules prevent ratcheting of a typical geometry with typical operating loads in a nuclear plant. The current thermal stress ratcheting rules evaluated on an elastic basis are enhanced to cover cases with high thermal membrane stress while not removing conservatism. Additionally, the evaluation of the simplified elastic-plastic rules for thermal stress ratcheting are simplified.
Evaluation of Mass Transfer Coefficient Under Swirl Flow Generated by the Combination of Pipe Elements: Part 3 — Analysis of Geometry Factor Under Swirl Flow on Pipe Element
Authors: Matsushita Mikiya , Yamamoto Kei , Nakamura Takao 
Published: November 2014

Flow accelerated corrosion (FAC) is a major pipe wall thinning phenomena in power plant. The management of pipe wall thinning is carried out in pipe elements such as elbow, T-tube, orifice, etc. of piping system in power plants. Wall thinning estimation methods in the management is usually applied with pipe wall thickness measurement in Japan.
In piping system, it is known that several pipe elements that are connected in series generate swirl flow. In Mihama Unit 3 pipe failure accident in 2004, it is pointed out the swirl flow caused by the piping layout might influence thinning rate behind orifice. Therefore the arrangement of pipe elements is considered to affect thinning phenomena seriously. So the behavior of swirl flow is one of the major research items to improve FAC model.
The purpose of this paper is to study the influence of swirl flow on wall thinning phenomena generated by the combination of pipe elements. This study focuses on the conditions and parameters (combination, distance, curvature radius of elbow, etc. of pipe elements) that affect the generation of swirl flow and analyzes the effect of such parameters in mass transfer coefficient and flow field of main-stream of swirl flow behind pipe elements. It is expected that axial velocity of main-stream of swirl flow affects distribution of geometry factor under swirl flow on pipe element.
The results of this study will be reflected to the evaluation of fluid-dynamic factor of FAC and can improve the management of wall thinning.
Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities: Part 2 — Application of CCFS Method to the Multiply Supported Systems
Authors: Tai Koichi , Sasajima Keisuke , Fukushima Shunsuke , Takamura Noriyuki , Onishi Shigenobu 
Published: November 2014

This paper provides a part of series of “Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities”. Paper is focused on the seismic evaluation method of the multiply supported systems, as the one of the design methodology adopted in the equipment and piping system of the seismic isolated nuclear power plant in Japan.
Many of the piping systems are multiply supported over different floor levels in the reactor building, and some of the piping systems are carried over to the adjacent building. Although Independent Support Motion (ISM) method has been widely applied in such a multiply supported seismic design of nuclear power plant, it is noted that the shortcoming of ignoring correlations between each excitations is frequently misleaded to the over-estimated design.
Application of Cross-oscillator, Cross-Floor response Spectrum (CCFS) method, proposed by A. Asfura and A. D. Kiureghian[1] shall be considered to be the excellent solution to the problems as mentioned above. So, we have introduced the algorithm of CCFS method to the FEM program.
The seismic responses of the benchmark model of multiply supported piping system are evaluated under various combination methods of ISM and CCFS, comparing to the exact solutions of Time History analysis method. As the result, it is demonstrated that the CCFS method shows excellent agreement to the responses of Time History analysis, and the CCFS method shall be one of the effective and practical design method of multiply supported systems.
Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities: Part 7 — Breaking Test Plan and Development of Test Machine for Full-Scale Lead Rubber Bearings
Authors: Kosugi Shinji , Kanazawa Kenji , Tsudome Hidekazu , Sato Nobuhisa , Jimbo Masakazu , Sato Kunihiko , Umeki Yoshito , Imaoka Tetsuo 
Published: November 2014

This paper is a part of the series “Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities.” This part presents a break test plan and development of a test machine for a full-scale lead rubber bearing (LRB). Application of seismic base-isolation systems using LRBs of 1600 mm in diameter to reactor building has been considered for the purpose of enhancing seismic safety. It is important to obtain the ultimate properties of isolators in order to estimate the seismic safety margin of seismic base-isolated structures against a beyond design-basis earthquake events. Recent studies reveal that the scaled effect appears on the ultimate properties of seismic rubber bearings. However, because of the limitation of the loading capabilities of loading machines, the ultimate property of such a large scale LRBs has not been confirmed.
In this study, the break tests for LRBs of 1600 mm in diameter is planned on the basis of estimation that refers to previous studies on break tests for small-scale LRBs and natural rubber bearings. The world-largest class test machine is designed and constructed to conduct static break tests for the full-scale LRBs. Furthermore, the performance of the test machine is evaluated from test results including those for break tests.
Fracture Toughness in the Ductile-Brittle Transition and Thermal Ageing Behavior of Decarburized Heat Affected Zone of Alloy 52 Dissimilar Metal Welds of Nuclear Components
Authors: Joly Pierre , Yescas Miguel , Keim Elisabeth 
Published: November 2014

Dissimilar metal welds (DMW) are used in nuclear power plants between the nozzles of main components in low alloy steel and stainless steel pipes, or safe-ends connected to the main coolant line pipes. AREVA proposes for EPR™ an improved design of DMW involving narrow gap welding without buttering between the low alloy steel nozzles and the stainless steel safe-ends, and the use of a corrosion resistant weld filler metal (Alloy 52).
AREVA performed a thorough characterization of this type of welds, which shows a particular microstructure close to the fusion line between the low alloy steel and the nickel base alloy, where the heat affected zone of the low alloy steel is decarburized.
This paper presents results of fracture toughness tests performed with the crack tip located in this area, in the ductile to brittle transition in the as post-welded heat treated condition and after thermal ageing. The results show an excellent fracture toughness behavior of this particular area, compared to that of low alloy steel parent metal.