Post-irradiation uranium isotopic atomic densities within the core of the High Flux Isotope Reactor (HFIR) were calculated and compared to uranium mass spectrographic data measured in the late 1960s and early 70s [1]. This study was performed in order to validate a Monte Carlo based depletion methodology for calculating the burn-up dependent nuclide inventory, specifically the post-irradiation uranium
{"title":"Validation of a Monte Carlo Based Depletion Methodology Using HFIR Post-Irradiation Measurements","authors":"D. Chandler, G. Maldonado, Trent Primm","doi":"10.2172/1052595","DOIUrl":"https://doi.org/10.2172/1052595","url":null,"abstract":"Post-irradiation uranium isotopic atomic densities within the core of the High Flux Isotope Reactor (HFIR) were calculated and compared to uranium mass spectrographic data measured in the late 1960s and early 70s [1]. This study was performed in order to validate a Monte Carlo based depletion methodology for calculating the burn-up dependent nuclide inventory, specifically the post-irradiation uranium","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"7 1","pages":"696-698"},"PeriodicalIF":0.0,"publicationDate":"2009-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73099903","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 : 2009-01-01DOI: 10.1007/978-1-84800-384-2_9
J. Ha, P. Seong
{"title":"HUPESS: Human Performance Evaluation Support System","authors":"J. Ha, P. Seong","doi":"10.1007/978-1-84800-384-2_9","DOIUrl":"https://doi.org/10.1007/978-1-84800-384-2_9","url":null,"abstract":"","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"94 1","pages":"197-229"},"PeriodicalIF":0.0,"publicationDate":"2009-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78547841","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 U.S. Nuclear Regulatory Commission (NRC) has undertaken a program to lay the technical foundation for defining proactive actions so that future degradation of materials in light water reactors (LWRs) is limited and, thereby, does not diminish either the integrity of important LWR components or the safety of operating plants. This technical letter report was prepared by staff at Pacific Northwest National Laboratory in support of the NRC Proactive Management of Materials Degradation (PMMD) program and relies heavily on work that was completed by Dr. Joseph Muscara and documented in NUREG/CR-6923. This report concisely explains the basic principles of PMMD and its relationship to prognostics, provides a review of programs related to PMMD being conducted worldwide, and provides an assessment of the technical gaps in PMMD and prognostics that need to be addressed. This technical letter report is timely because the majority of the U.S. reactor fleet is applying for license renewal, and many plants are also applying for increases in power rating. Both of these changes could increase the likelihood of materials degradation and underline, therefore, the interest in proactive management in the future.
{"title":"Proactive Management of Materials Degradation - A Review of Principles and Programs","authors":"L. Bond, S. Doctor, T. Taylor","doi":"10.2172/939352","DOIUrl":"https://doi.org/10.2172/939352","url":null,"abstract":"The U.S. Nuclear Regulatory Commission (NRC) has undertaken a program to lay the technical foundation for defining proactive actions so that future degradation of materials in light water reactors (LWRs) is limited and, thereby, does not diminish either the integrity of important LWR components or the safety of operating plants. This technical letter report was prepared by staff at Pacific Northwest National Laboratory in support of the NRC Proactive Management of Materials Degradation (PMMD) program and relies heavily on work that was completed by Dr. Joseph Muscara and documented in NUREG/CR-6923. This report concisely explains the basic principles of PMMD and its relationship to prognostics, provides a review of programs related to PMMD being conducted worldwide, and provides an assessment of the technical gaps in PMMD and prognostics that need to be addressed. This technical letter report is timely because the majority of the U.S. reactor fleet is applying for license renewal, and many plants are also applying for increases in power rating. Both of these changes could increase the likelihood of materials degradation and underline, therefore, the interest in proactive management in the future.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"31 1","pages":"414-415"},"PeriodicalIF":0.0,"publicationDate":"2008-08-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88529899","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}
Argonne National Laboratory and Idaho National Laboratory researchers are analyzing the electrochemical and thermal-fluid behavior of solid oxide electrolysis cells (SOECs) for high temperature steam electrolysis using computational fluid dynamics (CFD) techniques. The major challenges facing commercialization of steam electrolysis technology are related to efficiency, cost, and durability of the SOECs. The goal of this effort is to guide the design and optimization of performance for high temperature electrolysis (HTE) systems. An SOEC module developed by FLUENT Inc. as part of their general CFD code was used for the SOEC analysis by INL. ANL has developed an independent SOEC model that combines the governing electrochemical mechanisms based on first principals to the heat transfer and fluid dynamics in the operation of SOECs. The ANL model was embedded into the commercial STAR-CD CFD software, and is being used for the analysis of SOECs by ANL. The FY06 analysis performed by ANL and reported here covered the influence of electrochemical properties, SOEC component resistances and their contributing factors, SOEC size and inlet flow conditions, and SOEC flow configurations on the efficiency and expected durability of these systems. Some of the important findings from the ANL analysis are: (1) Increasing the inlet mass flux while going to larger cells can be a compromise to overcome increasing thermal and current density gradients while increasing the cell size. This approach could be beneficial for the economics of the SOECs; (2) The presence of excess hydrogen at the SOEC inlet to avoid Ni degradation can result in a sizeable decrease in the process efficiency; (3) A parallel-flow geometry for SOEC operation (if such a thing be achieved without sealing problems) yields smaller temperature gradients and current density gradients across the cell, which is favorable for the durability of the cells; (4) Contact resistances can significantly influence the total cell resistance and cell temperatures over a large range of operating potentials. Thus it is important to identify and avoid SOEC stack conditions leading to such high resistances due to poor contacts.
{"title":"Thermal-fluid and electrochemical modeling and performance study of a planar solid oxide electrolysis cell","authors":"B. Yildiz, T. Sofu","doi":"10.2172/934425","DOIUrl":"https://doi.org/10.2172/934425","url":null,"abstract":"Argonne National Laboratory and Idaho National Laboratory researchers are analyzing the electrochemical and thermal-fluid behavior of solid oxide electrolysis cells (SOECs) for high temperature steam electrolysis using computational fluid dynamics (CFD) techniques. The major challenges facing commercialization of steam electrolysis technology are related to efficiency, cost, and durability of the SOECs. The goal of this effort is to guide the design and optimization of performance for high temperature electrolysis (HTE) systems. An SOEC module developed by FLUENT Inc. as part of their general CFD code was used for the SOEC analysis by INL. ANL has developed an independent SOEC model that combines the governing electrochemical mechanisms based on first principals to the heat transfer and fluid dynamics in the operation of SOECs. The ANL model was embedded into the commercial STAR-CD CFD software, and is being used for the analysis of SOECs by ANL. The FY06 analysis performed by ANL and reported here covered the influence of electrochemical properties, SOEC component resistances and their contributing factors, SOEC size and inlet flow conditions, and SOEC flow configurations on the efficiency and expected durability of these systems. Some of the important findings from the ANL analysis are: (1) Increasing the inlet mass flux while going to larger cells can be a compromise to overcome increasing thermal and current density gradients while increasing the cell size. This approach could be beneficial for the economics of the SOECs; (2) The presence of excess hydrogen at the SOEC inlet to avoid Ni degradation can result in a sizeable decrease in the process efficiency; (3) A parallel-flow geometry for SOEC operation (if such a thing be achieved without sealing problems) yields smaller temperature gradients and current density gradients across the cell, which is favorable for the durability of the cells; (4) Contact resistances can significantly influence the total cell resistance and cell temperatures over a large range of operating potentials. Thus it is important to identify and avoid SOEC stack conditions leading to such high resistances due to poor contacts.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"60 1","pages":"84-85"},"PeriodicalIF":0.0,"publicationDate":"2008-06-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84005600","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 "Minor Actinide Doppler Coefficient Measurement Assessment" was a Department of Energy (DOE) U-NERI funded project intended to assess the viability of using either the FLATTOP or the COMET critical assembly to measure high temperature Doppler coefficients. The goal of the project was to calculate using the MCNP5 code the gram amounts of Np-237, Pu-238, Pu-239, Pu-241, AM-241, AM-242m, Am-243, and CM-244 needed to produce a 1E-5 in reactivity for a change in operating temperature 800C to 1000C. After determining the viability of using the assemblies and calculating the amounts of each actinide an experiment will be designed to verify the calculated results. The calculations and any doncuted experiments are designed to support the Advanced Fuel Cycle Initiative in conducting safety analysis of advanced fast reactor or acceoerator-driven transmutation systems with fuel containing high minor actinide content.
{"title":"Minior Actinide Doppler Coefficient Measurement Assessment","authors":"N. Hertel, D. Blaylock","doi":"10.2172/927795","DOIUrl":"https://doi.org/10.2172/927795","url":null,"abstract":"The \"Minor Actinide Doppler Coefficient Measurement Assessment\" was a Department of Energy (DOE) U-NERI funded project intended to assess the viability of using either the FLATTOP or the COMET critical assembly to measure high temperature Doppler coefficients. The goal of the project was to calculate using the MCNP5 code the gram amounts of Np-237, Pu-238, Pu-239, Pu-241, AM-241, AM-242m, Am-243, and CM-244 needed to produce a 1E-5 in reactivity for a change in operating temperature 800C to 1000C. After determining the viability of using the assemblies and calculating the amounts of each actinide an experiment will be designed to verify the calculated results. The calculations and any doncuted experiments are designed to support the Advanced Fuel Cycle Initiative in conducting safety analysis of advanced fast reactor or acceoerator-driven transmutation systems with fuel containing high minor actinide content.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"16 1","pages":"114-115"},"PeriodicalIF":0.0,"publicationDate":"2008-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89060959","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}
Radiation damage to structural and cladding materials, including austenitic stainless steels, ferritic steels, and zirconium alloys, in nuclear reactor environments results in significant mechanical property degradation, including yield strength increases, severe ductility losses and flow localization, which impacts reliability and performance. Generation IV and advanced fuel cycle concepts under consideration will require the development of advanced structural materials, which will operate in increasingly hostile environments. The development of predictive models is required to assess the performance and response of materials in extreme Gen IV reactor operating conditions (temperature, stress, and pressure), to decrease the time to rapidly assess the properties of new materials and insert them into technological applications (Gen IV and Advanced Fuel Cycle Operations).
{"title":"Dislocation - radiation obstacle interactions : Developing improved mechanical property constitutive models","authors":"B. Wirth, I. Robertson","doi":"10.2172/920995","DOIUrl":"https://doi.org/10.2172/920995","url":null,"abstract":"Radiation damage to structural and cladding materials, including austenitic stainless steels, ferritic steels, and zirconium alloys, in nuclear reactor environments results in significant mechanical property degradation, including yield strength increases, severe ductility losses and flow localization, which impacts reliability and performance. Generation IV and advanced fuel cycle concepts under consideration will require the development of advanced structural materials, which will operate in increasingly hostile environments. The development of predictive models is required to assess the performance and response of materials in extreme Gen IV reactor operating conditions (temperature, stress, and pressure), to decrease the time to rapidly assess the properties of new materials and insert them into technological applications (Gen IV and Advanced Fuel Cycle Operations).","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"568 1","pages":"473-473"},"PeriodicalIF":0.0,"publicationDate":"2007-11-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91429212","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}
During the summer of 2007, engineers at the Idaho National Laboratory (INL) conducted a two-day evaluation of commercially available battery powered, wireless, self-forming mesh nodes for use in emergency response. In this paper, the author describes the fundamentals of this emerging technology, applciations for emergency response and specific results of the technology evaluation conducted at the Idaho National Laboratory.
{"title":"EVALUATION OF RUGGED WIRELESS MESH NODES FOR USE IN EMERGENCY RESPONSE","authors":"K. Young, A. M. Snyder","doi":"10.2172/924509","DOIUrl":"https://doi.org/10.2172/924509","url":null,"abstract":"During the summer of 2007, engineers at the Idaho National Laboratory (INL) conducted a two-day evaluation of commercially available battery powered, wireless, self-forming mesh nodes for use in emergency response. In this paper, the author describes the fundamentals of this emerging technology, applciations for emergency response and specific results of the technology evaluation conducted at the Idaho National Laboratory.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"11 1","pages":"160-162"},"PeriodicalIF":0.0,"publicationDate":"2007-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74308428","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}
Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions.
{"title":"Scattered Neutron Tomography Based on A Neutron Transport Inverse Problem","authors":"W. Charlton","doi":"10.2172/915225","DOIUrl":"https://doi.org/10.2172/915225","url":null,"abstract":"Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"1 1","pages":"489-490"},"PeriodicalIF":0.0,"publicationDate":"2007-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82265113","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 fuel cycle economic analysis was performed on four fuel cycles to provide a baseline for initial cost comparison using the Gen IV Economic Modeling Work Group G4 ECON spreadsheet model, Decision Programming Language software, the 2006 Advanced Fuel Cycle Cost Basis report, industry cost data, international papers, the nuclear power related cost study from MIT, Harvard, and the University of Chicago. The analysis developed and compared the fuel cycle cost component of the total cost of energy for a wide range of fuel cycles including: once through, thermal with fast recycle, continuous fast recycle, and thermal recycle.
{"title":"Advanced fuel cycle economic sensitivity analysis","authors":"D. Shropshire, K. Williams, J. Smith, Brent Boore","doi":"10.2172/911885","DOIUrl":"https://doi.org/10.2172/911885","url":null,"abstract":"A fuel cycle economic analysis was performed on four fuel cycles to provide a baseline for initial cost comparison using the Gen IV Economic Modeling Work Group G4 ECON spreadsheet model, Decision Programming Language software, the 2006 Advanced Fuel Cycle Cost Basis report, industry cost data, international papers, the nuclear power related cost study from MIT, Harvard, and the University of Chicago. The analysis developed and compared the fuel cycle cost component of the total cost of energy for a wide range of fuel cycles including: once through, thermal with fast recycle, continuous fast recycle, and thermal recycle.","PeriodicalId":23138,"journal":{"name":"Transactions of the American Nuclear Society","volume":"2 1","pages":"172-172"},"PeriodicalIF":0.0,"publicationDate":"2006-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79788929","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}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：