A series of Rossi-alpha measurements was performed on the University of New Mexico`s AGN-201 reactor to measure the effective delayed neutron fraction {beta} and the mean prompt-neutron generation time of the system A{sub m}. An example of one of the Rossi-alpha measurements is shown in Fig. 1. Because the reactor is reflected, multiple prompt-neutron decay modes were observed.
{"title":"Preliminary results of a Rossi-alpha experiment on the University of New Mexico`s AGN-201 reactor","authors":"R. Busch, G. Spriggs","doi":"10.2172/10163021","DOIUrl":"https://doi.org/10.2172/10163021","url":null,"abstract":"A series of Rossi-alpha measurements was performed on the University of New Mexico`s AGN-201 reactor to measure the effective delayed neutron fraction {beta} and the mean prompt-neutron generation time of the system A{sub m}. An example of one of the Rossi-alpha measurements is shown in Fig. 1. Because the reactor is reflected, multiple prompt-neutron decay modes were observed.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"36 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87139207","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}
Research work is currently under way on the design and analysis of advanced gaseous core nuclear rocket concepts. The potentially very high operating temperatures encountered in the gas core nuclear reactor require detailed computational analysis of the fluid dynamics, heat transfer, and neutronics characteristics of such an assembly. Among the most important parameters needed for analysis are the optical and thermophysical properties of the uranium fuel; a detailed set of values is needed as a function of both temperature and pressure to correctly model the conditions that would exist in the gas core reactor.
{"title":"Optical and thermophysical properties of high-temperature gaseous uranium for nuclear rocket applications","authors":"V. Banjac, A. Heger","doi":"10.2514/6.1994-2898","DOIUrl":"https://doi.org/10.2514/6.1994-2898","url":null,"abstract":"Research work is currently under way on the design and analysis of advanced gaseous core nuclear rocket concepts. The potentially very high operating temperatures encountered in the gas core nuclear reactor require detailed computational analysis of the fluid dynamics, heat transfer, and neutronics characteristics of such an assembly. Among the most important parameters needed for analysis are the optical and thermophysical properties of the uranium fuel; a detailed set of values is needed as a function of both temperature and pressure to correctly model the conditions that would exist in the gas core reactor.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"54 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-06-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80380724","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}
Shutdown margin is an important safety parameter in reload core design. In order to search for the most reactive control rod, it is necessary to calculate the worth of each rod, which will require much computing time if three-dimensional full-core calculation is performed. Recently, Smith developed a one-group model for fast shutdown margin calculations in SIMULATE-3. In this paper, a perturbation method with regionwise flux expansion is proposed for fast shutdown-margin calculations. Because it was observed that in cold conditions, with no voids present, the axial distribution of neutron flux will not change drastically when a single control rod is withdrawn from its full-in to full-out position, a two-dimensional model is adopted in this study.
{"title":"Fast shutdown-margin calculation using perturbation theory with regionwise flux expansion","authors":"Jengjung Fang, Y. Liu, Pin-Wu Kao","doi":"10.13182/NSE94-A19812","DOIUrl":"https://doi.org/10.13182/NSE94-A19812","url":null,"abstract":"Shutdown margin is an important safety parameter in reload core design. In order to search for the most reactive control rod, it is necessary to calculate the worth of each rod, which will require much computing time if three-dimensional full-core calculation is performed. Recently, Smith developed a one-group model for fast shutdown margin calculations in SIMULATE-3. In this paper, a perturbation method with regionwise flux expansion is proposed for fast shutdown-margin calculations. Because it was observed that in cold conditions, with no voids present, the axial distribution of neutron flux will not change drastically when a single control rod is withdrawn from its full-in to full-out position, a two-dimensional model is adopted in this study.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"4 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81762083","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}
This report focuses on the prediction of materials performance for the carbon steel corrosion-allowance container as a function of thermal loading for the potential repository at Yucca Mountain. Low, intermediate and high thermal loads were evaluated as to their performance given assumptions regarding the temperature and humidity changes with time and the resultant depth of corrosion penetration. The reference case involved a kinetic relation for corrosion that was utilized in a sensitivity analysis to examine the impacts of time exponent, pitting, and mirobiologically-influenced corrosion. As a result of this study, the high thermal load appears to offer the best corrosion performance. However, other factors must be considered in making the final thermal loading decision.
{"title":"Impact of thermal loading on waste package material performance","authors":"D. Stahl, J. K. Mccoy, R. D. Mccright","doi":"10.1557/PROC-353-671","DOIUrl":"https://doi.org/10.1557/PROC-353-671","url":null,"abstract":"This report focuses on the prediction of materials performance for the carbon steel corrosion-allowance container as a function of thermal loading for the potential repository at Yucca Mountain. Low, intermediate and high thermal loads were evaluated as to their performance given assumptions regarding the temperature and humidity changes with time and the resultant depth of corrosion penetration. The reference case involved a kinetic relation for corrosion that was utilized in a sensitivity analysis to examine the impacts of time exponent, pitting, and mirobiologically-influenced corrosion. As a result of this study, the high thermal load appears to offer the best corrosion performance. However, other factors must be considered in making the final thermal loading decision.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"56 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1994-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87044945","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}
Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1Bmore » and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests.« less
{"title":"Direct containment heating experiments in Zion Nuclear Power Plant geometry using prototypic materials","authors":"J. L. Binder, L. McUmber, B. W. Spencer","doi":"10.2172/140597","DOIUrl":"https://doi.org/10.2172/140597","url":null,"abstract":"Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1Bmore » and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests.« less","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"47 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78870917","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}
Sandia National Laboratories, at the request of the Savannah River Technology Center (SRTC), studied the interactions between single drops of molten aluminum-lithium alloys and water. Most experiments were performed with ``B`` alloy (3.1 w/o Li, balance A1). Objectives were to develop experimental procedures for preparing and delivering the melt drops and diagnostics for characterizing the interactions, measure hydrogen generated by the reaction between melt and water, examine debris recovered after the interaction, determine changes in the aqueous phase produced by the melt-water chemical reactions, and determine whether steam explosions occur spontaneously under the conditions studied. Although many H{sub 2} bubbles were generated after the drops entered the water, spontaneous steam explosions never occurred when globules of the ``B`` alloy at temperatures between 700 and 1000C fell freely through water at room temperature, or upon or during subsequent contact with submerged aluminum or stainless steel surfaces. Total amounts of H{sub 2} (STP) increased from about 2 to 9 cm{sup 3}/per gram of melt as initial melt temperature increased over this range of temperatures.
{"title":"Interactions between drops of molten Al-Li alloys and liquid water","authors":"M. Hyder, L. S. Nelson, P. M. Duda, D. Hyndman","doi":"10.2172/10107057","DOIUrl":"https://doi.org/10.2172/10107057","url":null,"abstract":"Sandia National Laboratories, at the request of the Savannah River Technology Center (SRTC), studied the interactions between single drops of molten aluminum-lithium alloys and water. Most experiments were performed with ``B`` alloy (3.1 w/o Li, balance A1). Objectives were to develop experimental procedures for preparing and delivering the melt drops and diagnostics for characterizing the interactions, measure hydrogen generated by the reaction between melt and water, examine debris recovered after the interaction, determine changes in the aqueous phase produced by the melt-water chemical reactions, and determine whether steam explosions occur spontaneously under the conditions studied. Although many H{sub 2} bubbles were generated after the drops entered the water, spontaneous steam explosions never occurred when globules of the ``B`` alloy at temperatures between 700 and 1000C fell freely through water at room temperature, or upon or during subsequent contact with submerged aluminum or stainless steel surfaces. Total amounts of H{sub 2} (STP) increased from about 2 to 9 cm{sup 3}/per gram of melt as initial melt temperature increased over this range of temperatures.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"86 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77995730","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}
To make comparative assessments of competing technologies, consistent ground rules must be applied when developing cost estimates. This document provides a uniform set of assumptions, ground rules, and requirements that can be used in developing cost estimates for advanced nuclear power technologies. 10 refs., 8 figs., 32 tabs.
{"title":"Cost estimate guidelines for advanced nuclear power technologies","authors":"J. Delene, C. Hudson","doi":"10.2172/10176857","DOIUrl":"https://doi.org/10.2172/10176857","url":null,"abstract":"To make comparative assessments of competing technologies, consistent ground rules must be applied when developing cost estimates. This document provides a uniform set of assumptions, ground rules, and requirements that can be used in developing cost estimates for advanced nuclear power technologies. 10 refs., 8 figs., 32 tabs.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"15 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83457363","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 in-tank precipitation (ITP) process will be utilized as one of the steps to prepare high-level radioactive liquid wastes for vitrification in the Defense Waste Processing Facility at the Savannah River site. Hydrogen (H2) and benzene (C6H6) will be generated in the ITP waste tanks (tanks 48 and 49) as a result of radiolysis and decomposition. The structural response of these tanks to a hypothetical deflagration accident is being analyzed as part of the ITP waste tank safety analysis, and this work was performed to define the range of potential gas pressure loadings. The ITP waste tanks are equipped with a nitrogen purge gas system that removes combustible gases from the tank's vapor space and displaces oxygen. The deflagration accident scenario assumes that the purge gas system has failed and that a combustible gas mixture accumulates because of the buildup of combustible gases and inleakage of air.
{"title":"Evaluation of the gas pressure resulting from an ITP waste tank deflagration","authors":"J. K. Thomas, S. Hensel","doi":"10.2172/10188865","DOIUrl":"https://doi.org/10.2172/10188865","url":null,"abstract":"The in-tank precipitation (ITP) process will be utilized as one of the steps to prepare high-level radioactive liquid wastes for vitrification in the Defense Waste Processing Facility at the Savannah River site. Hydrogen (H2) and benzene (C6H6) will be generated in the ITP waste tanks (tanks 48 and 49) as a result of radiolysis and decomposition. The structural response of these tanks to a hypothetical deflagration accident is being analyzed as part of the ITP waste tank safety analysis, and this work was performed to define the range of potential gas pressure loadings. The ITP waste tanks are equipped with a nitrogen purge gas system that removes combustible gases from the tank's vapor space and displaces oxygen. The deflagration accident scenario assumes that the purge gas system has failed and that a combustible gas mixture accumulates because of the buildup of combustible gases and inleakage of air.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"14 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76746441","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}
Nuclear electric Propulsion (NEP) systems are expected to play a significant role in the exploration and exploitation of space. Unlike nuclear thermal propulsion (NTP) systems in which the hot reactor coolant is directly discharged from nozzles to provide the required thrust, NEP systems include electric power generation and conditioning units that in turn are used to drive electric thrusters. These thrusters accelerate sub atomic particles to produce thrust. The major advantage of NEP systems is the ability to provide very high specific impulses ([approximately]5000 s) that minimize the requirement for propellants. In addition, the power systems used in NEP could pro vide the dual purpose of also providing power for the missions at the destination. This synergism can be exploited in shared development costs. The NEP systems produce significantly lower thrust that NTP systems and are generally more massive. Both systems have their appropriate roles in a balanced space program. The technology development needs of NEP systems differ in many important ways from the development needs for NTP systems because of the significant differences in the operating conditions of the systems. The NEP systems require long-life reactor power systems operating at power levels that are considerably lower than those formore » NTP systems. In contrast, the operational lifetime of an NEP system (years) is orders of magnitude longer than the operational lifetime of NTP systems (thousands of second). Thus, the critical issue of NEP is survivability and reliable operability for very long times at temperatures that are considerably more modest than the temperatures required for effective NTP operations but generally much higher than those experienced in terrestrial reactors.« less
{"title":"Critical technologies for reactors used in nuclear electric propulsion","authors":"S. Bhattacharyya","doi":"10.2172/10171767","DOIUrl":"https://doi.org/10.2172/10171767","url":null,"abstract":"Nuclear electric Propulsion (NEP) systems are expected to play a significant role in the exploration and exploitation of space. Unlike nuclear thermal propulsion (NTP) systems in which the hot reactor coolant is directly discharged from nozzles to provide the required thrust, NEP systems include electric power generation and conditioning units that in turn are used to drive electric thrusters. These thrusters accelerate sub atomic particles to produce thrust. The major advantage of NEP systems is the ability to provide very high specific impulses ([approximately]5000 s) that minimize the requirement for propellants. In addition, the power systems used in NEP could pro vide the dual purpose of also providing power for the missions at the destination. This synergism can be exploited in shared development costs. The NEP systems produce significantly lower thrust that NTP systems and are generally more massive. Both systems have their appropriate roles in a balanced space program. The technology development needs of NEP systems differ in many important ways from the development needs for NTP systems because of the significant differences in the operating conditions of the systems. The NEP systems require long-life reactor power systems operating at power levels that are considerably lower than those formore » NTP systems. In contrast, the operational lifetime of an NEP system (years) is orders of magnitude longer than the operational lifetime of NTP systems (thousands of second). Thus, the critical issue of NEP is survivability and reliable operability for very long times at temperatures that are considerably more modest than the temperatures required for effective NTP operations but generally much higher than those experienced in terrestrial reactors.« less","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"44 1","pages":"17908"},"PeriodicalIF":0.0,"publicationDate":"1993-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75322916","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 accelerator transmutation of waste (ATW) concept is aimed at destroying key long-lived radionuclides (both actinides and fission products) in nuclear wastes, thereby reducing the long-term risks associated with the storage of such wastes. This technology could evolve into an approach to the production of fission power, utilizing abundant natural fuels and producing minimal long-lived nuclear waste. An ATW system would consist of the following components: 1. proton accelerator; 2. heavy-metal target; 3. moderating blanket; 4. thermal-to-electric power conversion plant; and 5. chemical separation facility. The linear accelerator provides a medium-energy, high-current proton beam that is directed at a heavy-metal target. The target converts the proton beam through spallation reactions into an intense neutron flux that is thermalized in the blanket region surrounding the target. The radioactive material to be transmuted is circulated through the blanket, where it undergoes neutron-induced reactions. Long-lived fission products undergo (n, [gamma]) reactions followed by beta decay, producing short-lived or stable products. The actinides are fissioned, producing additional neutrons and an assortment of fission products to reduce parasitic absorption in the blanket and to prevent further activation of these materials to long-lived radionuclides.
{"title":"The Accelerator Transmutation of Waste (ATW) concept overview","authors":"H. Dewey","doi":"10.2172/10190589","DOIUrl":"https://doi.org/10.2172/10190589","url":null,"abstract":"The accelerator transmutation of waste (ATW) concept is aimed at destroying key long-lived radionuclides (both actinides and fission products) in nuclear wastes, thereby reducing the long-term risks associated with the storage of such wastes. This technology could evolve into an approach to the production of fission power, utilizing abundant natural fuels and producing minimal long-lived nuclear waste. An ATW system would consist of the following components: 1. proton accelerator; 2. heavy-metal target; 3. moderating blanket; 4. thermal-to-electric power conversion plant; and 5. chemical separation facility. The linear accelerator provides a medium-energy, high-current proton beam that is directed at a heavy-metal target. The target converts the proton beam through spallation reactions into an intense neutron flux that is thermalized in the blanket region surrounding the target. The radioactive material to be transmuted is circulated through the blanket, where it undergoes neutron-induced reactions. Long-lived fission products undergo (n, [gamma]) reactions followed by beta decay, producing short-lived or stable products. The actinides are fissioned, producing additional neutrons and an assortment of fission products to reduce parasitic absorption in the blanket and to prevent further activation of these materials to long-lived radionuclides.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"72 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1993-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84737825","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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