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}
A method for the boundary coupling of forward and adjoint Monte Carlo radiation transport calculations, with no statistical error propagation due to the coupling procedures, has been developed. This method has direct application in the analysis of radiation streaming through multileg ducts with, in principle, no limitation to the number of legs.
{"title":"Forward-adjoint Monte Carlo coupling with no statistical error propagation","authors":"S. N. Cramer","doi":"10.13182/NSE96-A17919","DOIUrl":"https://doi.org/10.13182/NSE96-A17919","url":null,"abstract":"A method for the boundary coupling of forward and adjoint Monte Carlo radiation transport calculations, with no statistical error propagation due to the coupling procedures, has been developed. This method has direct application in the analysis of radiation streaming through multileg ducts with, in principle, no limitation to the number of legs.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-12-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86509558","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}
Determination of the appropriate number and placement of air monitors in the workplace is quite subjective and is generally one of the more difficult tasks in radiation protection. General guidance for determining the number and placement of air sampling and monitoring instruments has been provided by technical reports such as Mishima, J. These two documents and other published guidelines suggest that some insight into sampler placement can be obtained by conducting airflow studies involving the dilution and clearance of the relatively inert tracer gas sulfur hexafluoride (SF{sub 6}) in sampler placement studies and describes the results of a study done within the ITRI alpha inhalation exposure laboratories. The objectives of the study were to document an appropriate method for conducting SF{sub 6} dispersion studies, and to confirm the appropriate number and placement of air monitors and air samplers within a typical ITRI inhalation exposure laboratory. The results of this study have become part of the technical bases for air sampling and monitoring in the test room.
{"title":"Use of sulfur hexafluoride airflow studies to determine the appropriate number and placement of air monitors in an alpha inhalation exposure laboratory","authors":"G. J. Newton, Hoover","doi":"10.2172/381353","DOIUrl":"https://doi.org/10.2172/381353","url":null,"abstract":"Determination of the appropriate number and placement of air monitors in the workplace is quite subjective and is generally one of the more difficult tasks in radiation protection. General guidance for determining the number and placement of air sampling and monitoring instruments has been provided by technical reports such as Mishima, J. These two documents and other published guidelines suggest that some insight into sampler placement can be obtained by conducting airflow studies involving the dilution and clearance of the relatively inert tracer gas sulfur hexafluoride (SF{sub 6}) in sampler placement studies and describes the results of a study done within the ITRI alpha inhalation exposure laboratories. The objectives of the study were to document an appropriate method for conducting SF{sub 6} dispersion studies, and to confirm the appropriate number and placement of air monitors and air samplers within a typical ITRI inhalation exposure laboratory. The results of this study have become part of the technical bases for air sampling and monitoring in the test room.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"81 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80835171","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}
T. Sampson, T. Kelley, T. Cremers, T. R. Konkel, R. Friar
We describe the new capability of and.present measurement results from the PC/FRAM plutonium isotopic analysis code. This new code allows data acquisition from a single coaxial germanium detector and analysis over an energy range from 120 keV to above I MeV. For the first time we demonstrate a complete isotopic analysis using only gamma rays greater than 200 keV in energy. This new capability allows the measurement of the plutonium isotopic composition of items inside shielded or heavy-walled containers without having to remove the items from the container. This greatly enhances worker safety by reducing handling and the resultant radiation exposure. Another application allows international inspectors to verify the contents of items inside sealed, long-term storage containers that may not be opened for national security or treaty compliance reasons. We present measurement results for traditional planar germanium detectors as well as coaxial detectors measuring shielded and unshielded samples.
{"title":"PC/FRAM: New capabilities for the gamma-ray spectrometry measurement of plutonium isotopic composition","authors":"T. Sampson, T. Kelley, T. Cremers, T. R. Konkel, R. Friar","doi":"10.2172/114469","DOIUrl":"https://doi.org/10.2172/114469","url":null,"abstract":"We describe the new capability of and.present measurement results from the PC/FRAM plutonium isotopic analysis code. This new code allows data acquisition from a single coaxial germanium detector and analysis over an energy range from 120 keV to above I MeV. For the first time we demonstrate a complete isotopic analysis using only gamma rays greater than 200 keV in energy. This new capability allows the measurement of the plutonium isotopic composition of items inside shielded or heavy-walled containers without having to remove the items from the container. This greatly enhances worker safety by reducing handling and the resultant radiation exposure. Another application allows international inspectors to verify the contents of items inside sealed, long-term storage containers that may not be opened for national security or treaty compliance reasons. We present measurement results for traditional planar germanium detectors as well as coaxial detectors measuring shielded and unshielded samples.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"7 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89765773","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}
MOCUP is a system of external processors that allow for a limited treatment of the temporal composition of the user-selected MCNP cells in a time-dependent flux environment. The ORIGEN2 code computes the time-dependent compositions of these individually selected MCNP cells. All data communication between the two codes is accomplished through the MCNP and ORIGEN2 input/output files, the MOCUP Processor Output files, and two user supplied tables. MOCUP is either command line or interactively driven. The interactive interface is based on the portable XII window environment and the Motif tool kit. MOCUP was constructed so that no modifications to either MCNP or ORIGEN2 were necessary. Section 4 of the writeup contains the input instructions needed to set up the MOCUP run. MOCUP is extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices.
{"title":"MOCUP: MCNP-ORIGEN2 coupled utility program","authors":"R. Moore, B. Schnitzler, C. Wemple","doi":"10.2172/130667","DOIUrl":"https://doi.org/10.2172/130667","url":null,"abstract":"MOCUP is a system of external processors that allow for a limited treatment of the temporal composition of the user-selected MCNP cells in a time-dependent flux environment. The ORIGEN2 code computes the time-dependent compositions of these individually selected MCNP cells. All data communication between the two codes is accomplished through the MCNP and ORIGEN2 input/output files, the MOCUP Processor Output files, and two user supplied tables. MOCUP is either command line or interactively driven. The interactive interface is based on the portable XII window environment and the Motif tool kit. MOCUP was constructed so that no modifications to either MCNP or ORIGEN2 were necessary. Section 4 of the writeup contains the input instructions needed to set up the MOCUP run. MOCUP is extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"95 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80432638","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}
Continuous-energy Monte Carlo Codes are not generally suited for adjoint coupled electron-photon transport. Line radiation (e.g., fluorescence) is especially difficult to implement in adjoint mode with continuous-energy codes. The only published work on adjoint electron Monte Carlo transport is Jordan. The adjoint capability of his NOVICE code is expedited by a multigroup approximation. More recently, a Boltzmann-Fokker-Planck (BFP) Monte Carlo technique has been developed for adjoint electron transport. As in NOVICE, particle transport with BFP Monte Carlo is neither entirely continuous energy nor entirely multigroup. The BFP method has been tested in the multigroup version of MCNP and is being integrated into the ITS code package. Multigroup data produced by the CEPXS cross-section-generating code is needed to operate the BFP codes in adjoint electron-photon mode. In this paper, we present adjoint electron-photon transport results obtained with a new version of CEPXS and a new multigroup version of ITS.
{"title":"Adjoint ITS calculations using the CEPXS electron-photon cross sections","authors":"L. Lorence, R. Kensek, J. Halbleib","doi":"10.2172/90095","DOIUrl":"https://doi.org/10.2172/90095","url":null,"abstract":"Continuous-energy Monte Carlo Codes are not generally suited for adjoint coupled electron-photon transport. Line radiation (e.g., fluorescence) is especially difficult to implement in adjoint mode with continuous-energy codes. The only published work on adjoint electron Monte Carlo transport is Jordan. The adjoint capability of his NOVICE code is expedited by a multigroup approximation. More recently, a Boltzmann-Fokker-Planck (BFP) Monte Carlo technique has been developed for adjoint electron transport. As in NOVICE, particle transport with BFP Monte Carlo is neither entirely continuous energy nor entirely multigroup. The BFP method has been tested in the multigroup version of MCNP and is being integrated into the ITS code package. Multigroup data produced by the CEPXS cross-section-generating code is needed to operate the BFP codes in adjoint electron-photon mode. In this paper, we present adjoint electron-photon transport results obtained with a new version of CEPXS and a new multigroup version of ITS.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"45 16 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77947377","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}
L. S. Nelson, P. M. Duda, D. Hyndman, D. K. Allison, M. Hyder
Seventeen steam explosion experiments were performed with 2 to 10 g drops of molten, high-purity Al. Seven were successfully initiated with underwater exploding bridgewires. At melt release temperatures up to 1400{degrees}C (1673 K) only moderate thermal-type explosions occurred that produced bubbles with volumes up to approximately 1 L. Bubble growth intensified in the melt temperature range 1400-1525{degrees}C (1673--1798 K) as threshold ignition of Al set in. In this range, one of the explosions emitted a flash of light and generated a bubble that grew very rapidly to approximately 14 L, broke through the water surface, and destroyed the test chamber. We attribute the behavior of this latter bubble, which grew as fast as one produced by the underwater firing of a 0.6 g explosive detonator, to an ignition-type steam explosion. Aluminum oxides could not be detected visually in the debris recovered from either typical thermal-type or the ignition-type explosions, and only traces could be detected by X-ray diffraction. In the ignition-type explosion, it is possible however that some oxidic material, probably the smaller particles, was lost during the flooding that occurred as the chamber failed. Both bubble analyses and the absence of appreciable oxide in the debris suggest that themore » ignition-type steam explosion was not very efficient, probably involving the combustion of only a small fraction of the original molten aluminum globule.« less
{"title":"Thermal- and ignition-type steam explosions of single drops of molten aluminum","authors":"L. S. Nelson, P. M. Duda, D. Hyndman, D. K. Allison, M. Hyder","doi":"10.2172/147716","DOIUrl":"https://doi.org/10.2172/147716","url":null,"abstract":"Seventeen steam explosion experiments were performed with 2 to 10 g drops of molten, high-purity Al. Seven were successfully initiated with underwater exploding bridgewires. At melt release temperatures up to 1400{degrees}C (1673 K) only moderate thermal-type explosions occurred that produced bubbles with volumes up to approximately 1 L. Bubble growth intensified in the melt temperature range 1400-1525{degrees}C (1673--1798 K) as threshold ignition of Al set in. In this range, one of the explosions emitted a flash of light and generated a bubble that grew very rapidly to approximately 14 L, broke through the water surface, and destroyed the test chamber. We attribute the behavior of this latter bubble, which grew as fast as one produced by the underwater firing of a 0.6 g explosive detonator, to an ignition-type steam explosion. Aluminum oxides could not be detected visually in the debris recovered from either typical thermal-type or the ignition-type explosions, and only traces could be detected by X-ray diffraction. In the ignition-type explosion, it is possible however that some oxidic material, probably the smaller particles, was lost during the flooding that occurred as the chamber failed. Both bubble analyses and the absence of appreciable oxide in the debris suggest that themore » ignition-type steam explosion was not very efficient, probably involving the combustion of only a small fraction of the original molten aluminum globule.« less","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"117 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86815845","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 response to a request from Thomas Grumbly, Assistant Secretary of Energy for Environmental Management, the Hanford Site contractors developed a set of risk-based cleanup strategies that (a) protect the public, workers, and environment from unacceptable risks; (b) are executable technically; and (c) fit within the currently expected annual funding profile. These strategies were developed because (1) the U.S. Department of Energy (DOE) and Hanford site budgets are being reduced (2) stakeholders are dissatisfied with the perceived rate of cleanup (3) the U.S. Congress and the DOE are increasingly focusing on risk and risk-reduction activities (4) the present strategy is not integrated across the site and is inconsistent in its treatment of similar hazards (5) the present cleanup strategy is not cost-effective from a risk-reduction or future land-use perspective (6) the milestones and activities in the Tri-Party Agreement cannot be achieved with an anticipated funding of $1.05 billion, or less, annually.
{"title":"Development of a risk-based approach to Hanford site cleanup","authors":"W. Hesser","doi":"10.2172/120008","DOIUrl":"https://doi.org/10.2172/120008","url":null,"abstract":"In response to a request from Thomas Grumbly, Assistant Secretary of Energy for Environmental Management, the Hanford Site contractors developed a set of risk-based cleanup strategies that (a) protect the public, workers, and environment from unacceptable risks; (b) are executable technically; and (c) fit within the currently expected annual funding profile. These strategies were developed because (1) the U.S. Department of Energy (DOE) and Hanford site budgets are being reduced (2) stakeholders are dissatisfied with the perceived rate of cleanup (3) the U.S. Congress and the DOE are increasingly focusing on risk and risk-reduction activities (4) the present strategy is not integrated across the site and is inconsistent in its treatment of similar hazards (5) the present cleanup strategy is not cost-effective from a risk-reduction or future land-use perspective (6) the milestones and activities in the Tri-Party Agreement cannot be achieved with an anticipated funding of $1.05 billion, or less, annually.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"65 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77316546","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 revised ENDF/B-VI data set is discussed. The low capture resonance integral is primarily responsible for the large eigenvalue trend with epithermal fission rate, and a modification of the uranium 235 evaluation will yield a significantly improved evaluated data set.
{"title":"Homogeneous critical Monte Carlo eigenvalue calculations with revised ENDF/B-VI data sets","authors":"A. Kahler","doi":"10.2172/48679","DOIUrl":"https://doi.org/10.2172/48679","url":null,"abstract":"The revised ENDF/B-VI data set is discussed. The low capture resonance integral is primarily responsible for the large eigenvalue trend with epithermal fission rate, and a modification of the uranium 235 evaluation will yield a significantly improved evaluated data set.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"27 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"1995-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74899034","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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