This paper reports on revised cross-section evaluations for {sup 134}Ba, {sup 149}Sm, {sup 154}Eu, {sup 155}Eu, {sup 160}Dy, {sup 161}Dy, {sup 162}Dy, {sup 163}Dy, and {sup 164}Dy. The evaluations for {sup 134}Ba, {sup 154}Eu, and {sup 1554}Eu were previously revised for ENDF/B-VI. The other 6 evaluations, carried over from ENDF/B-V, were completed in the 1974--1980 time period. The evaluations for the dysprosium isotopes go back to ENDF/B-IV. Newer experimental data, not considered for the current ENDF/B-VI evaluations, was used in all of the revised evaluations. In the present work the primary emphasis was placed on the resolved and unresolved resonance regions, but newer measured data were also used for energies above the unresolved resonance region. Elastic, capture, and total cross sections are revised. Some important parameters from the revised evaluations are given in Table 1; corresponding values from the ENDF/B-VI evaluations are also given.
{"title":"Revised evaluations of fission-product cross sections","authors":"R. Q. Wright","doi":"10.2172/663274","DOIUrl":"https://doi.org/10.2172/663274","url":null,"abstract":"This paper reports on revised cross-section evaluations for {sup 134}Ba, {sup 149}Sm, {sup 154}Eu, {sup 155}Eu, {sup 160}Dy, {sup 161}Dy, {sup 162}Dy, {sup 163}Dy, and {sup 164}Dy. The evaluations for {sup 134}Ba, {sup 154}Eu, and {sup 1554}Eu were previously revised for ENDF/B-VI. The other 6 evaluations, carried over from ENDF/B-V, were completed in the 1974--1980 time period. The evaluations for the dysprosium isotopes go back to ENDF/B-IV. Newer experimental data, not considered for the current ENDF/B-VI evaluations, was used in all of the revised evaluations. In the present work the primary emphasis was placed on the resolved and unresolved resonance regions, but newer measured data were also used for energies above the unresolved resonance region. Elastic, capture, and total cross sections are revised. Some important parameters from the revised evaluations are given in Table 1; corresponding values from the ENDF/B-VI evaluations are also given.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"473 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1998-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90189789","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}
An approach to delayed critical experiment was performed in 1981 at Pacific Northwest Laboratory with a cylindrical tank of plutonium-uranium nitrate solution. During this experiment, various methods to determine the critical height were used, including (1) extrapolation of the usual plot of inverse count rate vs. height, which estimates the delayed critical height (DCH); (2) the inverse count rate vs. height divided by count rate, which corrects somewhat for the change in inherent source size as the height changes; (3) ratio of spectral densities vs. height, which extrapolates to DCH; (4) extrapolations of prompt neutron decay constant vs. height, which extrapolates to the prompt critical height (PCH); and (5) inverse kinetics rod drop (IKRD) methods, which measure {Delta}k/k{Beta} very accurately for a particular solution height. The problem with some of the extrapolation methods is that the measured data are not linear with height, but, for lack of anything better, linear extrapolations are made. In addition to the measurements to determine the delayed critical height subcriticality measurements by the {sup 252}Cf source driven frequency analysis method were performed for a variety of subcriticality heights. This paper describes how all these methods were applied to obtain the critical height of a cylindrical tank of plutonium nitrate solution and how the subcritical neutron multiplication factor was obtained.
{"title":"Extrapolations to critical for systems with large inherent sources","authors":"J. Mihalczo, Wyatt","doi":"10.2172/633994","DOIUrl":"https://doi.org/10.2172/633994","url":null,"abstract":"An approach to delayed critical experiment was performed in 1981 at Pacific Northwest Laboratory with a cylindrical tank of plutonium-uranium nitrate solution. During this experiment, various methods to determine the critical height were used, including (1) extrapolation of the usual plot of inverse count rate vs. height, which estimates the delayed critical height (DCH); (2) the inverse count rate vs. height divided by count rate, which corrects somewhat for the change in inherent source size as the height changes; (3) ratio of spectral densities vs. height, which extrapolates to DCH; (4) extrapolations of prompt neutron decay constant vs. height, which extrapolates to the prompt critical height (PCH); and (5) inverse kinetics rod drop (IKRD) methods, which measure {Delta}k/k{Beta} very accurately for a particular solution height. The problem with some of the extrapolation methods is that the measured data are not linear with height, but, for lack of anything better, linear extrapolations are made. In addition to the measurements to determine the delayed critical height subcriticality measurements by the {sup 252}Cf source driven frequency analysis method were performed for a variety of subcriticality heights. This paper describes how all these methods were applied to obtain the critical height of a cylindrical tank of plutonium nitrate solution and how the subcritical neutron multiplication factor was obtained.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"28 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1997-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88790703","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}
Drift flux models are commonly used to describe two-phase flow systems when explicit representation of the relative phase motion is not required. In these models, relative phase velocity is typically described by flow-regime-dependent, semi-empirical models. Although they are a somewhat simple description of the two-phase conditions that might be expected in nuclear power systems, drift flux models can still be expected to give reasonable results in a significant range of operating conditions and can be useful in incorporating thermal-hydraulic feedback into steady-state and transient neutronics calculations. In this paper, we examine the numerical stability associated with the finite difference solution of the mixture drift flux equations. We assume a standard semi-implicit discretization on a staggered spatial mesh, where the drift flux terms are evaluated purely explicitly.
{"title":"Numerical stability of the mixture drift flux equations","authors":"J. Doster, J. Kauffman","doi":"10.13182/NSE99-A2051","DOIUrl":"https://doi.org/10.13182/NSE99-A2051","url":null,"abstract":"Drift flux models are commonly used to describe two-phase flow systems when explicit representation of the relative phase motion is not required. In these models, relative phase velocity is typically described by flow-regime-dependent, semi-empirical models. Although they are a somewhat simple description of the two-phase conditions that might be expected in nuclear power systems, drift flux models can still be expected to give reasonable results in a significant range of operating conditions and can be useful in incorporating thermal-hydraulic feedback into steady-state and transient neutronics calculations. In this paper, we examine the numerical stability associated with the finite difference solution of the mixture drift flux equations. We assume a standard semi-implicit discretization on a staggered spatial mesh, where the drift flux terms are evaluated purely explicitly.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"35 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1997-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74551728","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}
Many of the wastes from processing fissile materials contain metals which may serve as nuclear criticality poisons. It would be advantageous to the criticality evaluation of these wastes to demonstrate that the poisons remain with the fissile materials and to demonstrate an always safe poison-to-fissile ratio. The first task, demonstrating that the materials stay together, is the job of the chemist, the second, calculating an always safe ratio, is an object of this paper.
{"title":"Metal Poisons for Criticality in Waste Streams","authors":"T. G. Williamson, A. Goslen","doi":"10.2172/636037","DOIUrl":"https://doi.org/10.2172/636037","url":null,"abstract":"Many of the wastes from processing fissile materials contain metals which may serve as nuclear criticality poisons. It would be advantageous to the criticality evaluation of these wastes to demonstrate that the poisons remain with the fissile materials and to demonstrate an always safe poison-to-fissile ratio. The first task, demonstrating that the materials stay together, is the job of the chemist, the second, calculating an always safe ratio, is an object of this paper.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"64 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1996-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76840972","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 nuclear weapons accident is an extremely unlikely event due to the extensive care taken in operations. However, under some hypothetical accident conditions, plutonium might be dispersed to the environment. This would result in costs being incurred by the government to remediate the site and compensate for losses. This study is a multi-disciplinary evaluation of the potential scope of the post-accident response that includes technical factors, current and proposed legal requirements and constraints, as well as social/political factors that could influence decision making. The study provides parameters that can be used to assess economic costs for accidents postulated to occur in urban areas, Midwest farmland, Western rangeland, and forest. Per-area remediation costs have been estimated, using industry-standard methods, for both expedited and extended remediation. Expedited remediation costs have been evaluated for highways, airports, and urban areas. Extended remediation costs have been evaluated for all land uses except highways and airports. The inclusion of cost estimates in risk assessments, together with the conventional estimation of doses and health effects, allows a fuller understanding of the post-accident environment. The insights obtained can be used to minimize economic risks by evaluation of operational and design alternatives, and through development of improved capabilities for accident response.
{"title":"Site Restoration: Estimation of Attributable Costs From Plutonium-Dispersal Accidents","authors":"D. Chanin, W. B. Murfin","doi":"10.2172/249283","DOIUrl":"https://doi.org/10.2172/249283","url":null,"abstract":"A nuclear weapons accident is an extremely unlikely event due to the extensive care taken in operations. However, under some hypothetical accident conditions, plutonium might be dispersed to the environment. This would result in costs being incurred by the government to remediate the site and compensate for losses. This study is a multi-disciplinary evaluation of the potential scope of the post-accident response that includes technical factors, current and proposed legal requirements and constraints, as well as social/political factors that could influence decision making. The study provides parameters that can be used to assess economic costs for accidents postulated to occur in urban areas, Midwest farmland, Western rangeland, and forest. Per-area remediation costs have been estimated, using industry-standard methods, for both expedited and extended remediation. Expedited remediation costs have been evaluated for highways, airports, and urban areas. Extended remediation costs have been evaluated for all land uses except highways and airports. The inclusion of cost estimates in risk assessments, together with the conventional estimation of doses and health effects, allows a fuller understanding of the post-accident environment. The insights obtained can be used to minimize economic risks by evaluation of operational and design alternatives, and through development of improved capabilities for accident response.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"43 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1996-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89391169","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 recently developed exponential discontinuous spatial differencing scheme for the discrete-ordinate equations has been extended to x-y-z geometry with hexahedral cells. This scheme produces strictly positive angular fluxes given positive discrete-ordinate sources. The exponential discontinuous scheme has been developed and implemented into the three-dimensional, discrete-ordinate code. THREEDANT. Numerical results are given which show that the exponential discontinuous scheme is very accurate for deep-penetration transport problems with optically thick spatial meshes.
{"title":"An exponential discontinuous scheme for X-Y-Z geometry transport problems","authors":"T. Wareing, R. Alcouffe","doi":"10.2172/224949","DOIUrl":"https://doi.org/10.2172/224949","url":null,"abstract":"The recently developed exponential discontinuous spatial differencing scheme for the discrete-ordinate equations has been extended to x-y-z geometry with hexahedral cells. This scheme produces strictly positive angular fluxes given positive discrete-ordinate sources. The exponential discontinuous scheme has been developed and implemented into the three-dimensional, discrete-ordinate code. THREEDANT. Numerical results are given which show that the exponential discontinuous scheme is very accurate for deep-penetration transport problems with optically thick spatial meshes.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"37 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1996-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81579327","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 method of monitoring in situ the dose rate contributed from the gaseous radionuclide {sup 41}Ar was developed using a portable gamma-ray spectrometer. The instrument was immersed in a sealed chamber filled with {sup 41}Ar to determine the calibration factor for the dose rate converted from the photopeak count rate. The calibrated spectrometer was applied in situ to measure the noble gas {sup 41}Ar dose rate both inside and outside the containment of a research reactor operating at full power. The continuously monitored dose rate from {sup 41}Ar was contour-mapped around the nuclear reactor facility, and the nonuniform distribution of {sup 41}Ar was correlated with the prevailing diffusion as well as dispersion conditions.
{"title":"Rapid measurement of the gaseous {sup 41}Ar released from a nuclear reactor","authors":"C. Chung, Chen-Yi Chen, Cheng-Jong Lee","doi":"10.13182/NT96-A35214","DOIUrl":"https://doi.org/10.13182/NT96-A35214","url":null,"abstract":"A method of monitoring in situ the dose rate contributed from the gaseous radionuclide {sup 41}Ar was developed using a portable gamma-ray spectrometer. The instrument was immersed in a sealed chamber filled with {sup 41}Ar to determine the calibration factor for the dose rate converted from the photopeak count rate. The calibrated spectrometer was applied in situ to measure the noble gas {sup 41}Ar dose rate both inside and outside the containment of a research reactor operating at full power. The continuously monitored dose rate from {sup 41}Ar was contour-mapped around the nuclear reactor facility, and the nonuniform distribution of {sup 41}Ar was correlated with the prevailing diffusion as well as dispersion conditions.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"46 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1996-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87465236","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}
Pub Date : 1995-12-31DOI: 10.13182/FST96-A11963162
M. Zucchetti
Passive design features are more reliable than operator action of successful operation of active safety systems. Passive safety has usually been adopted for fission. The achievement of an inventory-based passive safety is difficult if the fusion reactor uses neutronic reactions. Ignitor is a high-magnetic field tokamak designed to study the physics of ignited plasmas. The safety goal for Ignitor is classification as a mobility-based passively safe machine.
{"title":"Fusion reactor passive safety and ignitor risk-based regulation","authors":"M. Zucchetti","doi":"10.13182/FST96-A11963162","DOIUrl":"https://doi.org/10.13182/FST96-A11963162","url":null,"abstract":"Passive design features are more reliable than operator action of successful operation of active safety systems. Passive safety has usually been adopted for fission. The achievement of an inventory-based passive safety is difficult if the fusion reactor uses neutronic reactions. Ignitor is a high-magnetic field tokamak designed to study the physics of ignited plasmas. The safety goal for Ignitor is classification as a mobility-based passively safe machine.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"43 1","pages":"1501-1505"},"PeriodicalIF":0.0,"publicationDate":"1995-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72655916","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}
As computing power has increased, so too has the ability to model and simulate complex systems and processes. In addition, virtual reality technology has made it possible to visualize and understand many complex scientific and engineering problems. For this reason, a virtual dosimetry program called Virtual Radiation Fields (VRF) is developed to model radiation dose rate and cumulative dose to a receptor operating in a virtual radiation environment. With the design and testing of many facilities and products taking place in the virtual world, this program facilitates the concurrent consideration of radiological concerns during the design process. Three-dimensional (3D) graphical presentation of the radiation environment is made possible through the use of IGRIP, a graphical modeling program developed by Deneb Robotics, Inc. The VRF simulation program was designed to model and display a virtual dosimeter. As a demonstration of the program`s capability, the Hanford tank, C-106, was modeled to predict radiation doses to robotic equipment used to remove radioactive waste from the tank. To validate VRF dose predictions, comparison was made with reported values for tank C-106, which showed agreement to within 0.5%. Graphical information is presented regarding the 3D dose rate variation inside the tank. Cumulative dose predictions were made for the cleanup operations of tank C-106. A four-dimensional dose rate map generated by VRF was used to model the dose rate not only in 3D space but also as a function of the amount of waste remaining in the tank. This allowed VRF to predict dose rate at any stage in the waste removal process for an accurate simulation of the radiological conditions throughout the tank cleanup procedure.
{"title":"Virtual radiation fields for ALARA determination","authors":"T. Knight","doi":"10.2172/672123","DOIUrl":"https://doi.org/10.2172/672123","url":null,"abstract":"As computing power has increased, so too has the ability to model and simulate complex systems and processes. In addition, virtual reality technology has made it possible to visualize and understand many complex scientific and engineering problems. For this reason, a virtual dosimetry program called Virtual Radiation Fields (VRF) is developed to model radiation dose rate and cumulative dose to a receptor operating in a virtual radiation environment. With the design and testing of many facilities and products taking place in the virtual world, this program facilitates the concurrent consideration of radiological concerns during the design process. Three-dimensional (3D) graphical presentation of the radiation environment is made possible through the use of IGRIP, a graphical modeling program developed by Deneb Robotics, Inc. The VRF simulation program was designed to model and display a virtual dosimeter. As a demonstration of the program`s capability, the Hanford tank, C-106, was modeled to predict radiation doses to robotic equipment used to remove radioactive waste from the tank. To validate VRF dose predictions, comparison was made with reported values for tank C-106, which showed agreement to within 0.5%. Graphical information is presented regarding the 3D dose rate variation inside the tank. Cumulative dose predictions were made for the cleanup operations of tank C-106. A four-dimensional dose rate map generated by VRF was used to model the dose rate not only in 3D space but also as a function of the amount of waste remaining in the tank. This allowed VRF to predict dose rate at any stage in the waste removal process for an accurate simulation of the radiological conditions throughout the tank cleanup procedure.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"26 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78838459","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 proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW{sub f} nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U{sub 3}Si{sub 2}-Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of {sup 235}U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119{degrees}C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400{degrees}C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered.
{"title":"Neutronic analysis of three-element core configurations for the Advanced Neutron Source Reactor","authors":"J. Gehin","doi":"10.2172/205866","DOIUrl":"https://doi.org/10.2172/205866","url":null,"abstract":"The proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW{sub f} nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U{sub 3}Si{sub 2}-Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of {sup 235}U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119{degrees}C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400{degrees}C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"32 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89077384","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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