Journal of Astm International最新文献_第7页

VVER-440 and VVER-1000 Reactor Dosimetry Benchmark —BUGLE-96 Versus ALPAN VII.0 VVER-440和VVER-1000反应堆剂量测定基准-BUGLE-96与ALPAN VII.0

Journal of Astm International Pub Date : 2012-04-01 DOI: 10.1520/JAI104131

Jose I. Duo

{"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":"1 1","pages":"1-6"},"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}

引用次数: 0

Calculation of Kobasko's Simplified Heat Transfer Coefficients from Cooling Curve Data Obtained with Small Probes 从小探头获得的冷却曲线数据计算Kobasko的简化传热系数

Journal of Astm International Pub Date : 2012-04-01 DOI: 10.1520/JAI104304

R. Otero

引用次数: 9

A Model to Estimate Separator Forces during Ball Speed Variations 球速度变化时分离器力的估计模型

Journal of Astm International Pub Date : 2012-04-01 DOI: 10.1520/JAI104209

A. Leveille, P. Frantz, Garry Rosene

{"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":"35 3 1","pages":"1-15"},"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}

引用次数: 1

Comparison of Regulatory Guide 1.99 Fluence Attenuation Methods 监管指南1.99 Fluence衰减方法的比较

Journal of Astm International Pub Date : 2012-04-01 DOI: 10.1520/JAI104028

E. Jones

{"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":"25 1","pages":"1-7"},"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}

引用次数: 0

Development and Experimental Validation of a Calculation Scheme for Nuclear Heating Evaluation in the Core of the OSIRIS Material Testing Reactor OSIRIS材料试验堆堆芯核加热评估计算方案的开发与实验验证

Journal of Astm International Pub Date : 2012-04-01 DOI: 10.1520/JAI104026

F. Malouch

引用次数: 14

Ex-Vessel Neutron Dosimetry Analysis for Westinghouse 4-Loop XL Pressurized Water Reactor Plant Using 3D Parallel Discrete Ordinates Code RAPTOR-M3G 基于三维平行离散坐标代码RAPTOR-M3G的西屋4环XL压水堆堆前中子剂量学分析

Journal of Astm International Pub Date : 2012-04-01 DOI: 10.1520/JAI104030

J. Chen, F. Alpan, G. Fischer, A. Fero

引用次数: 0

Dynamic Behavior of Sand/Rubber Mixtures. Part I: Effect of Rubber Content and Duration of Confinement on Small-Strain Shear Modulus and Damping Ratio 砂/橡胶混合物的动力特性。第一部分:橡胶含量和约束时间对小应变剪切模量和阻尼比的影响

Journal of Astm International Pub Date : 2012-04-01 DOI: 10.1520/JAI103680

A. Anastasiadis, K. Senetakis, K. Pitilakis, Chrysanthi Gargala, Iphigeneia Karakasi

引用次数: 25

Photon spectrum behind biological shielding of the LVR-15 research reactor LVR-15研究反应堆生物屏蔽后的光子光谱

Journal of Astm International Pub Date : 2012-04-01 DOI: 10.1520/JAI104077

V. Klupák, L. Viererbl, Z. Lahodová, M. Marek, M. Vinš

引用次数: 2

Analysis of the Notch Effect in Fatigue 疲劳状态下缺口效应分析

Journal of Astm International Pub Date : 2012-04-01 DOI: 10.1520/JAI103944

K. Yanase, M. Endo

{"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":"79 1","pages":"103944"},"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}

引用次数: 5

Jules Horowitz Reactor, a New Irradiation Facility: Improving Dosimetry for the Future of Nuclear Experimentation 朱尔斯·霍洛维茨反应堆，一种新的辐照设备:为未来的核实验改进剂量学

Journal of Astm International Pub Date : 2012-04-01 DOI: 10.1520/JAI104127

G. Grégoire, D. Beretz, C. Destouches

{"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":"5 1","pages":"1-7"},"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}

引用次数: 7