O. Grytsenko, S. Pugach, V. L. Diemokhin, V. Bukanov, M. Marek, S. Vandlík
Results of determination of irradiation conditions for vessel internals of VVER-440 reactor No. 1 and 2 at Rovno Nuclear Power Plant, obtained by specialists at Institute for Nuclear Research Kyiv, Ukraine, and Nuclear Research Institute Rez, Czech Republic, are presented. To calculate neutron transport, detailed calculation models of these reactors were prepared. Distribution of neutron flux functionals on the surface of reactor VVER-440 baffle and core barrel for different core loads was studied. Agreement between results obtained by specialists at Institute for Nuclear Research and at Nuclear Research Institute is shown.
{"title":"Exposure Conditions of Reactor Internals of Rovno VVER-440 Nuclear Power Plant Units 1 and 2","authors":"O. Grytsenko, S. Pugach, V. L. Diemokhin, V. Bukanov, M. Marek, S. Vandlík","doi":"10.1520/JAI104050","DOIUrl":"https://doi.org/10.1520/JAI104050","url":null,"abstract":"Results of determination of irradiation conditions for vessel internals of VVER-440 reactor No. 1 and 2 at Rovno Nuclear Power Plant, obtained by specialists at Institute for Nuclear Research Kyiv, Ukraine, and Nuclear Research Institute Rez, Czech Republic, are presented. To calculate neutron transport, detailed calculation models of these reactors were prepared. Distribution of neutron flux functionals on the surface of reactor VVER-440 baffle and core barrel for different core loads was studied. Agreement between results obtained by specialists at Institute for Nuclear Research and at Nuclear Research Institute is shown.","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":"80696131","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}
In the 1980's the dosimetry community embraced the need for a high fidelity quantification of uncertainty in nuclear data used for dosimetry applications. This led to the adoption of energy-dependent covariance matrices as the accepted manner of quantifying the uncertainty data. The trend for the dosimetry community to require high fidelity treatment of uncertainty estimates has continued to the current time where requirements on nuclear data are codified in standards such as ASTM E 1018. This paper surveys the current state of the dosimetry cross sections and investigates the quality of the current dosimetry cross section evaluations by examining calculated-to-experimental ratios in neutron benchmark fields. In recent years more nuclear-related technical areas are placing an emphasis on uncertainty quantification. With the availability of model-based cross sections and covariance matrices produced by nuclear data codes, some nuclear-related communities are considering the role these covariance matrices should play. While funding within the dosimetry community for cross section evaluations has been very meager, other areas, such as the solar-related astrophysics community and the US Nuclear Criticality Safety Program, have been supporting research in the area of neutron cross sections. The Cross Section Evaluation Working Group (CSEWG) is responsible for the creation and maintenancemore » of the ENDF/B library which has been the mainstay for the reactor dosimetry community. Given the new trends in cross section evaluations, this paper explores the path forward for the US nuclear reactor dosimetry community and its use of the ENDF/B cross-sections. The major concern is maintenance of the sufficiency and accuracy of the uncertainty estimate when used for dosimetry applications. The two major areas of deficiency in the proposed ENDF/B approach are: 1) the use of unrelated covariance matrices in ENDF/B evaluations and 2) the lack of 'due consideration' of experimental data in the evaluation. (authors)« less
在20世纪80年代，剂量学界接受了对用于剂量学应用的核数据的不确定度进行高保真量化的需要。这导致采用能量相关协方差矩阵作为量化不确定性数据的公认方式。剂量学界要求对不确定性估计进行高保真度处理的趋势一直持续到目前，对核数据的要求已写入ASTM E 1018等标准。本文综述了目前剂量学截面的现状，并通过检验中子基准场的计算与实验比，探讨了目前剂量学截面评估的质量。近年来，越来越多的核相关技术领域开始重视不确定性的量化。随着核数据编码产生的基于模型的截面和协方差矩阵的可用性，一些核相关社区正在考虑这些协方差矩阵应该发挥的作用。虽然剂量学界对截面评估的资助非常少，但其他领域，如太阳相关天体物理学界和美国核临界安全计划，一直在支持中子截面领域的研究。截面评估工作组(CSEWG)负责创建和维护更多的ENDF/B库，该库一直是反应堆剂量学社区的支柱。鉴于截面评估的新趋势，本文探讨了美国核反应堆剂量学界及其使用ENDF/B截面的前进道路。主要关注的是在剂量学应用中保持不确定度估计的充分性和准确性。提出的ENDF/B方法的两个主要缺陷是:1)在ENDF/B评估中使用不相关的协方差矩阵;2)在评估中缺乏对实验数据的“适当考虑”。(作者)«更少
{"title":"Path Forward for Dosimetry Cross Sections","authors":"P. Griffin, C. Peters","doi":"10.1520/JAI104114","DOIUrl":"https://doi.org/10.1520/JAI104114","url":null,"abstract":"In the 1980's the dosimetry community embraced the need for a high fidelity quantification of uncertainty in nuclear data used for dosimetry applications. This led to the adoption of energy-dependent covariance matrices as the accepted manner of quantifying the uncertainty data. The trend for the dosimetry community to require high fidelity treatment of uncertainty estimates has continued to the current time where requirements on nuclear data are codified in standards such as ASTM E 1018. This paper surveys the current state of the dosimetry cross sections and investigates the quality of the current dosimetry cross section evaluations by examining calculated-to-experimental ratios in neutron benchmark fields. In recent years more nuclear-related technical areas are placing an emphasis on uncertainty quantification. With the availability of model-based cross sections and covariance matrices produced by nuclear data codes, some nuclear-related communities are considering the role these covariance matrices should play. While funding within the dosimetry community for cross section evaluations has been very meager, other areas, such as the solar-related astrophysics community and the US Nuclear Criticality Safety Program, have been supporting research in the area of neutron cross sections. The Cross Section Evaluation Working Group (CSEWG) is responsible for the creation and maintenancemore » of the ENDF/B library which has been the mainstay for the reactor dosimetry community. Given the new trends in cross section evaluations, this paper explores the path forward for the US nuclear reactor dosimetry community and its use of the ENDF/B cross-sections. The major concern is maintenance of the sufficiency and accuracy of the uncertainty estimate when used for dosimetry applications. The two major areas of deficiency in the proposed ENDF/B approach are: 1) the use of unrelated covariance matrices in ENDF/B evaluations and 2) the lack of 'due consideration' of experimental data in the evaluation. (authors)« less","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":"77809146","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 paper summarizes the attenuation coefficient of the neutron fluence with E > 0.5 MeV through a reactor pressure vessel for vodo-vodyanoi energetichesky reactor (VVER) reactor types measured and/or calculated for mock-up experiments, as well as for operated nuclear power plant (NPP) units. The attenuation coefficient is possible to evaluate directly only by using the retro-dosimetry, based on a combination of the measured activities from the weld sample and concurrent ex-vessel measurement. The available neutron fluence attenuation coefficients (E > 0.5 MeV), calculated and measured at a mock-up experiment simulating the VVER-440-unit conditions, vary from 3.5 to 6.15. A similar situation is used for the calculations and mock-up experiment measurements for the VVER-1000 RPV, where the attenuation coefficient of the neutron fluence varies from 5.99 to 8.85. Because of the difference in calculations for the real units and the mock-up experiments, the necessity to design and perform calculation benchmarks both for VVER-440 and VVER-1000 would be meaningful if the calculation model is designed adequately to a given unit.
{"title":"Comparison of Attenuation Coefficients for VVER-440 and VVER-1000 Pressure Vessels","authors":"M. Marek, J. Rataj, S. Vandlík","doi":"10.1520/JAI104055","DOIUrl":"https://doi.org/10.1520/JAI104055","url":null,"abstract":"The paper summarizes the attenuation coefficient of the neutron fluence with E > 0.5 MeV through a reactor pressure vessel for vodo-vodyanoi energetichesky reactor (VVER) reactor types measured and/or calculated for mock-up experiments, as well as for operated nuclear power plant (NPP) units. The attenuation coefficient is possible to evaluate directly only by using the retro-dosimetry, based on a combination of the measured activities from the weld sample and concurrent ex-vessel measurement. The available neutron fluence attenuation coefficients (E > 0.5 MeV), calculated and measured at a mock-up experiment simulating the VVER-440-unit conditions, vary from 3.5 to 6.15. A similar situation is used for the calculations and mock-up experiment measurements for the VVER-1000 RPV, where the attenuation coefficient of the neutron fluence varies from 5.99 to 8.85. Because of the difference in calculations for the real units and the mock-up experiments, the necessity to design and perform calculation benchmarks both for VVER-440 and VVER-1000 would be meaningful if the calculation model is designed adequately to a given unit.","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":"84709379","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}
P. Casoli, N. Authier, J. Laurec, E. Baugé, T. Granier
In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energymore » causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)« less
{"title":"Measurements of Actinide-Fission Product Yields in Caliban and Prospero Metallic Core Reactor Fission-Neutron Fields","authors":"P. Casoli, N. Authier, J. Laurec, E. Baugé, T. Granier","doi":"10.1520/JAI104018","DOIUrl":"https://doi.org/10.1520/JAI104018","url":null,"abstract":"In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energymore » causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)« less","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":"79905742","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}
Y. R. Takeuchi, S. Davis, M. Eby, Jerome K. Fuller, D. L. Taylor, Michael J. Rosado
Bearings are used in a number of spacecraft applications, ranging from minimal motion devices, such as pointing mechanisms, to high-speed components, such as control moment gyroscopes and reaction and momentum wheels. Terrestrial applications include pumps, axles, and tooling. Heat-transfer modes for rotational systems in a vacuum environment differ from their terrestrial counterpart. In space, with the absence of air, heat transfer consists of radiation from the rotating system and conductance through the bearings themselves. Depending on the application, conductance could dominate the effects on bearing temperatures and thermal gradients. Accurate thermal predictions are important because they can drive life and performance requirements. To accurately predict bearing temperatures, basic bearing thermal conductance data was needed. However, bearing thermal conductance tends to be the most significant unknown in a rotational system in the space environment. To address this shortcoming, this paper explores a new vacuum test rig designed to measure bearing conductance under simulated operational conditions. Experimental variables include control of the bearing rotational speed, applied axial load, and average bearing temperature and temperature gradient via an applied heat source/heat sink mechanism. All tests are conducted in vacuum. The experimental variables studied herein allowed parametric studies to be conducted under controlled thermal and mechanical conditions, permitting the exploration of the influences of those operational variables on bearing thermal conductance. This paper will describe the test method, the use of uncertainty analysis to design the experiment, and a verification study.
{"title":"Bearing Thermal Conductance Measurement Test Method and Experimental Design","authors":"Y. R. Takeuchi, S. Davis, M. Eby, Jerome K. Fuller, D. L. Taylor, Michael J. Rosado","doi":"10.1520/JAI104233","DOIUrl":"https://doi.org/10.1520/JAI104233","url":null,"abstract":"Bearings are used in a number of spacecraft applications, ranging from minimal motion devices, such as pointing mechanisms, to high-speed components, such as control moment gyroscopes and reaction and momentum wheels. Terrestrial applications include pumps, axles, and tooling. Heat-transfer modes for rotational systems in a vacuum environment differ from their terrestrial counterpart. In space, with the absence of air, heat transfer consists of radiation from the rotating system and conductance through the bearings themselves. Depending on the application, conductance could dominate the effects on bearing temperatures and thermal gradients. Accurate thermal predictions are important because they can drive life and performance requirements. To accurately predict bearing temperatures, basic bearing thermal conductance data was needed. However, bearing thermal conductance tends to be the most significant unknown in a rotational system in the space environment. To address this shortcoming, this paper explores a new vacuum test rig designed to measure bearing conductance under simulated operational conditions. Experimental variables include control of the bearing rotational speed, applied axial load, and average bearing temperature and temperature gradient via an applied heat source/heat sink mechanism. All tests are conducted in vacuum. The experimental variables studied herein allowed parametric studies to be conducted under controlled thermal and mechanical conditions, permitting the exploration of the influences of those operational variables on bearing thermal conductance. This paper will describe the test method, the use of uncertainty analysis to design the experiment, and a verification study.","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":"82841619","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}
S. Bourganel, M. Soldevila, A. Ferrer, G. Grégoire, C. Destouches, D. Beretz
Interpretation of reactor dosimetry experiments with C/E comparison requires precise knowledge of parameters involved in modeling. Some parameters have more weight than others on the calculated values. So, sensitivity studies should be conducted to verify the importance of these parameters. The conclusions of these studies are used to refine the experiment modeling, or to correct uncertainty calculations. The results of these sensitivity studies allow a post-irradiation analysis, which can justify the discarding of some atypical C/M values. Derived uncertainties may be improved by the sensitivity analyses. Beyond classical parameters as geometry or composition, this paper describes some specific sensitivity studies conducted for dosimetry irradiation in reactor, and presents conclusions. These studies are based on dosimeters irradiated in the EOLE reactor facility at Cadarache CEA center. Conclusions drawn from these studies are generic and can be applied to any dosimetry study. Calculations performed for these studies were realized using TRIPOLI-4 Monte Carlo code.
{"title":"Sensitivity Studies Associated with Dosimetry Experiment Interpretation","authors":"S. Bourganel, M. Soldevila, A. Ferrer, G. Grégoire, C. Destouches, D. Beretz","doi":"10.1520/JAI104011","DOIUrl":"https://doi.org/10.1520/JAI104011","url":null,"abstract":"Interpretation of reactor dosimetry experiments with C/E comparison requires precise knowledge of parameters involved in modeling. Some parameters have more weight than others on the calculated values. So, sensitivity studies should be conducted to verify the importance of these parameters. The conclusions of these studies are used to refine the experiment modeling, or to correct uncertainty calculations. The results of these sensitivity studies allow a post-irradiation analysis, which can justify the discarding of some atypical C/M values. Derived uncertainties may be improved by the sensitivity analyses. Beyond classical parameters as geometry or composition, this paper describes some specific sensitivity studies conducted for dosimetry irradiation in reactor, and presents conclusions. These studies are based on dosimeters irradiated in the EOLE reactor facility at Cadarache CEA center. Conclusions drawn from these studies are generic and can be applied to any dosimetry study. Calculations performed for these studies were realized using TRIPOLI-4 Monte Carlo code.","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":"83292636","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}
S. H. Jalbani, M. Ackerman, E. M. Crown, M. Keulen, G. Song
Use of hot water has become extensive, especially in bitumen extraction from oil sands and in the production of heavy oil. The hot water is often under pressure and is 80–90°C, which is well above temperatures that result in immediate, potentially severe burn injuries. The ASTM F2701-08 apparatus consists of a funnel through which hot liquid is hand-poured to produce a 10 s exposure. Two 40 mm diameter copper calorimeters, mounted in an insulating sheet are positioned beneath the funnel outlet and are intended to measure the energy transfer through the fabric from the hot liquid. For this research, changes were made to the apparatus and procedures to more closely simulate low pressure hot water streams found in the oil industry and to improve reproducibility. The funnel producing the liquid splash was replaced with a small pipe directly fed by a circulating hot water bath via a small pump, through a hose and valve system, allowing for consistent application of a given quantity of water at a consistent temperature and flow rate. Water temperature, flow rate, and pressure can be altered as desired. A series of fabrics varying systematically on several parameters were tested with the modified equipment. Resulting heat transfer data suggest the system differentiates well among both semi-permeable and impermeable fabrics. Specifications for hot water protection are proposed.
{"title":"Apparatus for Use in Evaluating Protection from Low Pressure Hot Water Jets","authors":"S. H. Jalbani, M. Ackerman, E. M. Crown, M. Keulen, G. Song","doi":"10.1520/JAI104098","DOIUrl":"https://doi.org/10.1520/JAI104098","url":null,"abstract":"Use of hot water has become extensive, especially in bitumen extraction from oil sands and in the production of heavy oil. The hot water is often under pressure and is 80–90°C, which is well above temperatures that result in immediate, potentially severe burn injuries. The ASTM F2701-08 apparatus consists of a funnel through which hot liquid is hand-poured to produce a 10 s exposure. Two 40 mm diameter copper calorimeters, mounted in an insulating sheet are positioned beneath the funnel outlet and are intended to measure the energy transfer through the fabric from the hot liquid. For this research, changes were made to the apparatus and procedures to more closely simulate low pressure hot water streams found in the oil industry and to improve reproducibility. The funnel producing the liquid splash was replaced with a small pipe directly fed by a circulating hot water bath via a small pump, through a hose and valve system, allowing for consistent application of a given quantity of water at a consistent temperature and flow rate. Water temperature, flow rate, and pressure can be altered as desired. A series of fabrics varying systematically on several parameters were tested with the modified equipment. Resulting heat transfer data suggest the system differentiates well among both semi-permeable and impermeable fabrics. Specifications for hot water protection are proposed.","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":"83828628","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 development of surrogate headforms with similar dimensions and weight to that of a human head has allowed researchers to collect dynamic impact response data for impact reconstructions and injury assessment. These headforms are relied upon to deliver accurate and repeatable dynamic impact response data for setting helmet certification standards as well as head injury reconstruction. With recent research demonstrating the importance of measuring three dimensional dynamic impact response characteristics, the Hybrid III headform is a potentially a good candidate for use in standards testing and impact reconstructions. Currently, this headform is validated with a single 37.6-cm drop to the front region of the headform with an acceptance window of 50 g. Therefore, the purpose of this study was to compare the dynamic impact response of two Hybrid III headforms and verify repeatability, compare dynamic impact response, and determine how closely the two headforms correlate across different impact conditions. Two Hybrid III headforms were dropped from nine heights at two impact locations (front and side). Results of this study show that the two headforms are highly correlated across drop heights. Significant differences in terms of dynamic impact response were found between the two headforms across impact conditions. This study showed that two Hybrid III headforms produce similar mean peak linear acceleration for front centric impacts, however, differ significantly for mean peak angular response.
{"title":"Comparison between Hybrid III Headforms by Linear and Angular Dynamic Impact Response Characteristics","authors":"M. Kendall, T. Hoshizaki","doi":"10.1520/STP104270","DOIUrl":"https://doi.org/10.1520/STP104270","url":null,"abstract":"The development of surrogate headforms with similar dimensions and weight to that of a human head has allowed researchers to collect dynamic impact response data for impact reconstructions and injury assessment. These headforms are relied upon to deliver accurate and repeatable dynamic impact response data for setting helmet certification standards as well as head injury reconstruction. With recent research demonstrating the importance of measuring three dimensional dynamic impact response characteristics, the Hybrid III headform is a potentially a good candidate for use in standards testing and impact reconstructions. Currently, this headform is validated with a single 37.6-cm drop to the front region of the headform with an acceptance window of 50 g. Therefore, the purpose of this study was to compare the dynamic impact response of two Hybrid III headforms and verify repeatability, compare dynamic impact response, and determine how closely the two headforms correlate across different impact conditions. Two Hybrid III headforms were dropped from nine heights at two impact locations (front and side). Results of this study show that the two headforms are highly correlated across drop heights. Significant differences in terms of dynamic impact response were found between the two headforms across impact conditions. This study showed that two Hybrid III headforms produce similar mean peak linear acceleration for front centric impacts, however, differ significantly for mean peak angular response.","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":"83940367","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}
In this study, recycled asphalt shingles (RAS) were evaluated for potential use as structural fill in highway embankments or backfills behind retaining walls. Bottom ash (BA) and foundry slag (FS) were also investigated as additives to RAS to enhance its mechanical properties. The engineering properties of RAS:BA/FS mixtures including compaction characteristics, hydraulic conductivity, compressibility, shear strength, and the coefficient of lateral earth pressure at rest were evaluated in a systematic manner. The results show that the addition of bottom ash and foundry slag significantly reduces the compressibility of RAS while increasing the drainage capacity and shear strength. The RAS:BA/FS mixtures are a favorable light weight material for use as embankment fills or backfill behind retaining walls.
{"title":"Recycled Asphalt Shingles Mixed with Granular Byproducts as Structural Fills","authors":"A. Soleimanbeigi, T. Edil, C. Benson","doi":"10.1520/JAI103766","DOIUrl":"https://doi.org/10.1520/JAI103766","url":null,"abstract":"In this study, recycled asphalt shingles (RAS) were evaluated for potential use as structural fill in highway embankments or backfills behind retaining walls. Bottom ash (BA) and foundry slag (FS) were also investigated as additives to RAS to enhance its mechanical properties. The engineering properties of RAS:BA/FS mixtures including compaction characteristics, hydraulic conductivity, compressibility, shear strength, and the coefficient of lateral earth pressure at rest were evaluated in a systematic manner. The results show that the addition of bottom ash and foundry slag significantly reduces the compressibility of RAS while increasing the drainage capacity and shear strength. The RAS:BA/FS mixtures are a favorable light weight material for use as embankment fills or backfill behind retaining 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":"80169979","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 Indira Gandhi Centre for Atomic Research (IGCAR) procedure for predicting the master curve (MC) reference temperature T0 has been further validated via its application to ORNL thermal shock experiment (TSE) 5 steel, an A508 Cl.2 steel that has been widely characterized in the literature. The procedure also was applied to ASTM A203D 3.5% Ni steel tested at IGCAR in virgin and irradiated states. The IGCAR procedure involves the application of the inverse Wallin strain rate equation to TQSchdy, the dynamic reference temperature obtained from Charpy tests using the modified Schindler procedure corresponding to a stress intensity factor rate of ∼106 MPa√m.s−1. A direct correlation of TQSchdy to T0 is also used. The larger of the estimates from these two, namely, TQSchW, provides an accurate and acceptably conservative estimate of the reference temperature TQ-est for steels with TQSchdy ≤ 60°C; for steels with TQSchdy>60°C, TQ-est=the larger of TQSchW and TQM2. The value of TQ-est obtained via the IGCAR procedure is termed TQ-IGC. Mostly, TQM2 gives the most conservative estimate, and TQBT (based on the value of TD obtained from instrumented Charpy V-notch impact tests) has a tendency toward accuracy, provided a robust estimate of TD can be obtained (which is not always the case). The IGCAR procedure was compared with two other conservative lower-bound (LB) procedures (Schindler-LB and Merkle-LB procedures), and it was shown that the IGCAR procedure with a 1% MC provides an LB curve without the difficulties involved in using the more tortuous calculations necessary for applying the Schindler- and Merkle-LB procedures. For the irradiated A203D steel, the shift in the estimated T0 was ΔTQ-IGC=340°C.
{"title":"Characterization of the Master Curve Based Fracture Toughness of ORNL TSE5 Steel and Unirradiated and Irradiated ASTM A203D 3.5 % Ni Steel by the IGCAR Procedure","authors":"P. Sreenivasan","doi":"10.1520/JAI103774","DOIUrl":"https://doi.org/10.1520/JAI103774","url":null,"abstract":"The Indira Gandhi Centre for Atomic Research (IGCAR) procedure for predicting the master curve (MC) reference temperature T0 has been further validated via its application to ORNL thermal shock experiment (TSE) 5 steel, an A508 Cl.2 steel that has been widely characterized in the literature. The procedure also was applied to ASTM A203D 3.5% Ni steel tested at IGCAR in virgin and irradiated states. The IGCAR procedure involves the application of the inverse Wallin strain rate equation to TQSchdy, the dynamic reference temperature obtained from Charpy tests using the modified Schindler procedure corresponding to a stress intensity factor rate of ∼106 MPa√m.s−1. A direct correlation of TQSchdy to T0 is also used. The larger of the estimates from these two, namely, TQSchW, provides an accurate and acceptably conservative estimate of the reference temperature TQ-est for steels with TQSchdy ≤ 60°C; for steels with TQSchdy>60°C, TQ-est=the larger of TQSchW and TQM2. The value of TQ-est obtained via the IGCAR procedure is termed TQ-IGC. Mostly, TQM2 gives the most conservative estimate, and TQBT (based on the value of TD obtained from instrumented Charpy V-notch impact tests) has a tendency toward accuracy, provided a robust estimate of TD can be obtained (which is not always the case). The IGCAR procedure was compared with two other conservative lower-bound (LB) procedures (Schindler-LB and Merkle-LB procedures), and it was shown that the IGCAR procedure with a 1% MC provides an LB curve without the difficulties involved in using the more tortuous calculations necessary for applying the Schindler- and Merkle-LB procedures. For the irradiated A203D steel, the shift in the estimated T0 was ΔTQ-IGC=340°C.","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":"80665559","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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