Analytical results of the vodo-vodyanoi energetichesky reactor- (VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Institute Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10 %) between BUGLE-96 and ALPAN VII.0 libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15 % with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements.
{"title":"VVER-440 and VVER-1000 Reactor Dosimetry Benchmark —BUGLE-96 Versus ALPAN VII.0","authors":"Jose I. Duo","doi":"10.1520/JAI104131","DOIUrl":"https://doi.org/10.1520/JAI104131","url":null,"abstract":"Analytical results of the vodo-vodyanoi energetichesky reactor- (VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Institute Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10 %) between BUGLE-96 and ALPAN VII.0 libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15 % with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements.","PeriodicalId":15057,"journal":{"name":"Journal of Astm International","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2012-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76209960","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Although heat transfer coefficient characterization of quench severity is not new, there continues to be a need for the rapid and relatively simple calculation of heat transfer coefficients from time-temperature cooling curve data files obtained via test methods such as ASTM D6200, D6482, D6549, and D7646, which utilize relatively small cylindrical test probes with diameters of ≤12.5 mm. One method that may be readily used is Kobasko’s computational method for effective heat transfer coefficients, which is based on time-temperature data obtained at the geometric center of small test probes during cooling curve analysis. A description of the step-by-step procedure for performing these calculations on actual experimental data is provided here.
{"title":"Calculation of Kobasko's Simplified Heat Transfer Coefficients from Cooling Curve Data Obtained with Small Probes","authors":"R. Otero","doi":"10.1520/JAI104304","DOIUrl":"https://doi.org/10.1520/JAI104304","url":null,"abstract":"Although heat transfer coefficient characterization of quench severity is not new, there continues to be a need for the rapid and relatively simple calculation of heat transfer coefficients from time-temperature cooling curve data files obtained via test methods such as ASTM D6200, D6482, D6549, and D7646, which utilize relatively small cylindrical test probes with diameters of ≤12.5 mm. One method that may be readily used is Kobasko’s computational method for effective heat transfer coefficients, which is based on time-temperature data obtained at the geometric center of small test probes during cooling curve analysis. A description of the step-by-step procedure for performing these calculations on actual experimental data is provided here.","PeriodicalId":15057,"journal":{"name":"Journal of Astm International","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2012-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88267540","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
When the combined loads on ball bearings include forces in the radial or moment directions, the balls will not orbit the bearing at one common speed. This ball speed variation (BSV) might result in additional bearing friction forces if the variation in speed is large enough to allow the balls to spread out from their normal spacing by an amount that exceeds the cage pocket clearances. In this paper we present a model used to estimate the cage forces and friction torques caused by BSV and compare the model predictions with measured friction torque test data. The model first analyzes the forces on a single ball by determining the distance over which a ball must slip during a single orbit of the bearing center. The ball to race force is determined by equating the energy lost during this slip with the energy input through traction at the inner race interface. The component of the ball to cage force normal to the ball–pocket interface is then determined by balancing this force with the ball to race force. The cage to land force is similarly determined by balancing the collective forces at all the ball to race and ball to cage interfaces. Finally, the BSV drag torque is written as the sum of contributions from these three drag sources. In order to validate the model, test data were obtained using single 204 size bearings operating under applied thrust and radial loading. The bearing drag torque was found to depend on the degree of misalignment and cage pocket clearance, as predicted by the model.
{"title":"A Model to Estimate Separator Forces during Ball Speed Variations","authors":"A. Leveille, P. Frantz, Garry Rosene","doi":"10.1520/JAI104209","DOIUrl":"https://doi.org/10.1520/JAI104209","url":null,"abstract":"When the combined loads on ball bearings include forces in the radial or moment directions, the balls will not orbit the bearing at one common speed. This ball speed variation (BSV) might result in additional bearing friction forces if the variation in speed is large enough to allow the balls to spread out from their normal spacing by an amount that exceeds the cage pocket clearances. In this paper we present a model used to estimate the cage forces and friction torques caused by BSV and compare the model predictions with measured friction torque test data. The model first analyzes the forces on a single ball by determining the distance over which a ball must slip during a single orbit of the bearing center. The ball to race force is determined by equating the energy lost during this slip with the energy input through traction at the inner race interface. The component of the ball to cage force normal to the ball–pocket interface is then determined by balancing this force with the ball to race force. The cage to land force is similarly determined by balancing the collective forces at all the ball to race and ball to cage interfaces. Finally, the BSV drag torque is written as the sum of contributions from these three drag sources. In order to validate the model, test data were obtained using single 204 size bearings operating under applied thrust and radial loading. The bearing drag torque was found to depend on the degree of misalignment and cage pocket clearance, as predicted by the model.","PeriodicalId":15057,"journal":{"name":"Journal of Astm International","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2012-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84684726","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
U.S. Regulatory Guide 1.99 Revision 2 (U.S. Nuclear Regulatory Commission, 1988, “Radiation Embrittlement of Reactor Vessel Materials,” Regulatory Guide 1.99, Revision 2, Washington, D.C.) provides for the use of two substantially different methods for determining through-wall fluence in nuclear reactor pressure vessels. One method is a generic attenuation curve based on a simplistic exponential decay equation. Partly due to the simplicity of its application, the generic attenuation method is predominantly used for licensing calculations. However, it has a limitation in that at increasing distances away from the core beltline, it becomes increasingly less accurate because it cannot account for neutron streaming effects in the cavity region surrounding the pressure vessel. The other attenuation method is based on a displacement per atom (dpa) calculation specific to the reactor vessel structure. The dpa method provides a more accurate representation of fluence attenuation through the reactor pressure vessel (RPV) wall at all elevations of the pressure vessel because it does account for neutron streaming in the cavity region. A requirement for using the dpa method, however, is an accurate flux solution through the RPV wall. This requirement has limited the use of traditional transport methods, such as discrete ordinates, that are limited by their treatment of cavity regions (i.e., air) outside the pressure vessel wall. TransWare Enterprises, under the sponsorship of EPRI and BWRVIP, has developed an advanced three-dimensional transport methodology capable of producing fully converged flux solutions throughout the entire reactor system, including in the cavity region and primary shield structures. This methodology provides an accurate and reliable determination of through-wall fluence in boiling water reactor (BWR) and pressurized water reactor (PWR) pressure vessels, thus allowing the dpa method to be implemented with high reliability. Using this advanced 3-D methodology, this paper presents comparisons of the generic and dpa attenuation methods at critical locations in both BWR and PWR pressure vessel walls.
{"title":"Comparison of Regulatory Guide 1.99 Fluence Attenuation Methods","authors":"E. Jones","doi":"10.1520/JAI104028","DOIUrl":"https://doi.org/10.1520/JAI104028","url":null,"abstract":"U.S. Regulatory Guide 1.99 Revision 2 (U.S. Nuclear Regulatory Commission, 1988, “Radiation Embrittlement of Reactor Vessel Materials,” Regulatory Guide 1.99, Revision 2, Washington, D.C.) provides for the use of two substantially different methods for determining through-wall fluence in nuclear reactor pressure vessels. One method is a generic attenuation curve based on a simplistic exponential decay equation. Partly due to the simplicity of its application, the generic attenuation method is predominantly used for licensing calculations. However, it has a limitation in that at increasing distances away from the core beltline, it becomes increasingly less accurate because it cannot account for neutron streaming effects in the cavity region surrounding the pressure vessel. The other attenuation method is based on a displacement per atom (dpa) calculation specific to the reactor vessel structure. The dpa method provides a more accurate representation of fluence attenuation through the reactor pressure vessel (RPV) wall at all elevations of the pressure vessel because it does account for neutron streaming in the cavity region. A requirement for using the dpa method, however, is an accurate flux solution through the RPV wall. This requirement has limited the use of traditional transport methods, such as discrete ordinates, that are limited by their treatment of cavity regions (i.e., air) outside the pressure vessel wall. TransWare Enterprises, under the sponsorship of EPRI and BWRVIP, has developed an advanced three-dimensional transport methodology capable of producing fully converged flux solutions throughout the entire reactor system, including in the cavity region and primary shield structures. This methodology provides an accurate and reliable determination of through-wall fluence in boiling water reactor (BWR) and pressurized water reactor (PWR) pressure vessels, thus allowing the dpa method to be implemented with high reliability. Using this advanced 3-D methodology, this paper presents comparisons of the generic and dpa attenuation methods at critical locations in both BWR and PWR pressure vessel walls.","PeriodicalId":15057,"journal":{"name":"Journal of Astm International","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2012-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75020207","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron–photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation.
{"title":"Development and Experimental Validation of a Calculation Scheme for Nuclear Heating Evaluation in the Core of the OSIRIS Material Testing Reactor","authors":"F. Malouch","doi":"10.1520/JAI104026","DOIUrl":"https://doi.org/10.1520/JAI104026","url":null,"abstract":"The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron–photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation.","PeriodicalId":15057,"journal":{"name":"Journal of Astm International","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2012-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74051440","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the beltline region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locations and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure.
{"title":"Ex-Vessel Neutron Dosimetry Analysis for Westinghouse 4-Loop XL Pressurized Water Reactor Plant Using 3D Parallel Discrete Ordinates Code RAPTOR-M3G","authors":"J. Chen, F. Alpan, G. Fischer, A. Fero","doi":"10.1520/JAI104030","DOIUrl":"https://doi.org/10.1520/JAI104030","url":null,"abstract":"Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the beltline region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locations and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure.","PeriodicalId":15057,"journal":{"name":"Journal of Astm International","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2012-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84791639","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Anastasiadis, K. Senetakis, K. Pitilakis, Chrysanthi Gargala, Iphigeneia Karakasi
The paper examines the small-strain dynamic properties of mixtures composed of sandy soils with recycled tire rubber. For this purpose, the experimental results stemming from a torsional resonant column testing program on twenty four [24] saturated and dry specimens are analyzed. The percentages of rubber used range between 0 and 35 % by mixture weight. GO values increase whereas DTO values decrease systematically as the content of rubber decreases and the mean confining pressure increases. Based on the experimental results we propose an analytical relationship for the estimation of GO, which is expressed in terms of an equivalent void ratio that considers the volume of rubber solids as part of the total volume of voids, along with an analytical relationship for the estimation of DTO. Finally, the effect of the specimen’s size and the duration of confinement on the initial shear modulus and damping ratio of the mixtures are also discussed.
{"title":"Dynamic Behavior of Sand/Rubber Mixtures. Part I: Effect of Rubber Content and Duration of Confinement on Small-Strain Shear Modulus and Damping Ratio","authors":"A. Anastasiadis, K. Senetakis, K. Pitilakis, Chrysanthi Gargala, Iphigeneia Karakasi","doi":"10.1520/JAI103680","DOIUrl":"https://doi.org/10.1520/JAI103680","url":null,"abstract":"The paper examines the small-strain dynamic properties of mixtures composed of sandy soils with recycled tire rubber. For this purpose, the experimental results stemming from a torsional resonant column testing program on twenty four [24] saturated and dry specimens are analyzed. The percentages of rubber used range between 0 and 35 % by mixture weight. GO values increase whereas DTO values decrease systematically as the content of rubber decreases and the mean confining pressure increases. Based on the experimental results we propose an analytical relationship for the estimation of GO, which is expressed in terms of an equivalent void ratio that considers the volume of rubber solids as part of the total volume of voids, along with an analytical relationship for the estimation of DTO. Finally, the effect of the specimen’s size and the duration of confinement on the initial shear modulus and damping ratio of the mixtures are also discussed.","PeriodicalId":15057,"journal":{"name":"Journal of Astm International","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2012-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80548858","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
V. Klupák, L. Viererbl, Z. Lahodová, M. Marek, M. Vinš
The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)
{"title":"Photon spectrum behind biological shielding of the LVR-15 research reactor","authors":"V. Klupák, L. Viererbl, Z. Lahodová, M. Marek, M. Vinš","doi":"10.1520/JAI104077","DOIUrl":"https://doi.org/10.1520/JAI104077","url":null,"abstract":"The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)","PeriodicalId":15057,"journal":{"name":"Journal of Astm International","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2012-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83389714","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The fatigue-crack propagation at stress concentrations is a topic of significant importance in a number of engineering applications. Further, it is recognized that the fatigue limit of notched components is dictated by the critical condition for either initiation or propagation of a small crack at the root of a notch. Moreover, because most fatigue cracks spend the vast majority of their lives as short cracks, the behavior of such a flaw is of significant importance. In the literature, McEvily and co-workers [McEvily, A. J., Eifler, D., and Macherauch, E., “An analysis of the Fatigue Growth of Short Fatigue Cracks,” Eng. Fract. Mech., Vol. 40, No. 3, 1991, pp. 571–584] developed a modified linear elastic fracture mechanics (LEFM) approach to tackle a number of fatigue problems, including the growth and threshold behavior of small fatigue cracks. In this study, a further extension is presented to deal with notch effects in fatigue. In this method, the elastic–plastic behavior and the crack closure are taken into account, as the major factors responsible for the peculiar behavior of small fatigue cracks emanating from notches. In the present paper, the notch effect in fatigue is systematically investigated by making use of a mechanism-based computational framework. A series of parametric studies demonstrate the predictive capability of the proposed framework. Based on the thorough investigation for notch-fatigue problem, the novelty of present study is illustrated.
应力集中下的疲劳裂纹扩展问题在许多工程应用中具有重要意义。此外，人们认识到，缺口部件的疲劳极限取决于缺口根部小裂纹的萌生或扩展的临界条件。此外，由于大多数疲劳裂纹以短裂纹的形式度过其寿命的绝大部分，因此这种裂纹的行为是非常重要的。在文献中，McEvily和他的同事[McEvily, A. J.， Eifler, D.， and Macherauch, E.]，“短疲劳裂纹的疲劳扩展分析”，英。打破。动力机械。[j]， Vol. 40, No. 3, 1991, pp. 571-584]开发了一种改进的线弹性断裂力学(LEFM)方法来解决许多疲劳问题，包括小疲劳裂纹的扩展和阈值行为。在本研究中，提出了进一步的扩展，以处理缺口效应的疲劳。在这种方法中，考虑了弹塑性行为和裂纹闭合，作为造成缺口产生的小疲劳裂纹的特殊行为的主要因素。本文采用基于力学的计算框架，系统地研究了缺口效应在疲劳中的作用。一系列的参数研究证明了该框架的预测能力。通过对缺口疲劳问题的深入研究，说明了本文研究的新颖性。
{"title":"Analysis of the Notch Effect in Fatigue","authors":"K. Yanase, M. Endo","doi":"10.1520/JAI103944","DOIUrl":"https://doi.org/10.1520/JAI103944","url":null,"abstract":"The fatigue-crack propagation at stress concentrations is a topic of significant importance in a number of engineering applications. Further, it is recognized that the fatigue limit of notched components is dictated by the critical condition for either initiation or propagation of a small crack at the root of a notch. Moreover, because most fatigue cracks spend the vast majority of their lives as short cracks, the behavior of such a flaw is of significant importance. In the literature, McEvily and co-workers [McEvily, A. J., Eifler, D., and Macherauch, E., “An analysis of the Fatigue Growth of Short Fatigue Cracks,” Eng. Fract. Mech., Vol. 40, No. 3, 1991, pp. 571–584] developed a modified linear elastic fracture mechanics (LEFM) approach to tackle a number of fatigue problems, including the growth and threshold behavior of small fatigue cracks. In this study, a further extension is presented to deal with notch effects in fatigue. In this method, the elastic–plastic behavior and the crack closure are taken into account, as the major factors responsible for the peculiar behavior of small fatigue cracks emanating from notches. In the present paper, the notch effect in fatigue is systematically investigated by making use of a mechanism-based computational framework. A series of parametric studies demonstrate the predictive capability of the proposed framework. Based on the thorough investigation for notch-fatigue problem, the novelty of present study is illustrated.","PeriodicalId":15057,"journal":{"name":"Journal of Astm International","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2012-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88570841","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The Jules Horowitz Reactor (JHR) is an experimental reactor under construction at the French Nuclear Energy and Alternative Energies Commission (CEA) facility at Cadarache. It will achieve its first criticality by the end of 2014. Experiments that will be conducted at JHR will deal with fuel, cladding, and material behavior. The JHR will also produce medical radio-isotopes and doped silicon for the electronic industry. As a new irradiation facility, its instrumentation will benefit from recent improvements. Nuclear instrumentation will include reactor dosimetry, as it is a reference technique to determine neutron fluence in experimental devices or characterize irradiation locations. Reactor dosimetry has been improved with the progress of simulation tools and nuclear data, but at the same time the customer needs have increased: Experimental results must have reduced and assessed uncertainties. This is now a necessary condition to perform an experimental irradiation in a test reactor. Items improved, in the framework of a general upgrading of the dosimetry process based on uncertainty minimization, will include dosimeter, nuclear data, and modelling scheme.
{"title":"Jules Horowitz Reactor, a New Irradiation Facility: Improving Dosimetry for the Future of Nuclear Experimentation","authors":"G. Grégoire, D. Beretz, C. Destouches","doi":"10.1520/JAI104127","DOIUrl":"https://doi.org/10.1520/JAI104127","url":null,"abstract":"The Jules Horowitz Reactor (JHR) is an experimental reactor under construction at the French Nuclear Energy and Alternative Energies Commission (CEA) facility at Cadarache. It will achieve its first criticality by the end of 2014. Experiments that will be conducted at JHR will deal with fuel, cladding, and material behavior. The JHR will also produce medical radio-isotopes and doped silicon for the electronic industry. As a new irradiation facility, its instrumentation will benefit from recent improvements. Nuclear instrumentation will include reactor dosimetry, as it is a reference technique to determine neutron fluence in experimental devices or characterize irradiation locations. Reactor dosimetry has been improved with the progress of simulation tools and nuclear data, but at the same time the customer needs have increased: Experimental results must have reduced and assessed uncertainties. This is now a necessary condition to perform an experimental irradiation in a test reactor. Items improved, in the framework of a general upgrading of the dosimetry process based on uncertainty minimization, will include dosimeter, nuclear data, and modelling scheme.","PeriodicalId":15057,"journal":{"name":"Journal of Astm International","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2012-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90142428","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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