Mohamed Sawan - Academia.edu (original) (raw)
Papers by Mohamed Sawan
Fusion Engineering and Design, 2015
Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 1998
As part of an ongoing exploration of final transport schemes based on plasma channels, a point de... more As part of an ongoing exploration of final transport schemes based on plasma channels, a point design of a final focus and reactor system is being developed. Six MJ of 3–4GeV Pb ions are delivered to a target in two opposing current-carrying plasma channels within a modified HYLIFE II reactor filled with 5Torr of Xe gas. The transition from the
Silicon carbide was first proposed as a low activation fusion reactor material in the mid 1970s. ... more Silicon carbide was first proposed as a low activation fusion reactor material in the mid 1970s. However, serious development of this material did not begin until the early 1990s, driven by the emergence of composite materials that provided enhanced toughness and an implied ability to use these typically brittle materials in engineering application. In the decades that followed, SiC composite
2 . Such effects will likely limit the performance and dominate failures in FNST components and m... more 2 . Such effects will likely limit the performance and dominate failures in FNST components and must be addressed to meet the extreme reliability and lifetime requirements of Demo fusion components. Data from experiments in the true fusion environment can give higher order insight into the evolution of multi-scale, multi-phenomena, multi-material synergies, including possible unanticipated effects and failure modes that
Fusion Science and Technology, 2005
Testing blanket concepts in the integrated fusion environment is one of the principal objectives ... more Testing blanket concepts in the integrated fusion environment is one of the principal objectives of ITER. Blanket test modules will be inserted in ITER from Day 1 of its operation and will provide the first experimental data on the feasibility of the D-T cycle for fusion. With the US rejoining ITER, the US community has decided to have strong participation
IEEE Transactions on Plasma Science, 2000
JD Sethian, DG Colombant, JL Giuliani, Jr., RH Lehmberg, MC Myers, SP Obenschain, AJ Schmitt, J. ... more JD Sethian, DG Colombant, JL Giuliani, Jr., RH Lehmberg, MC Myers, SP Obenschain, AJ Schmitt, J. Weaver, MF Wolford, F. Hegeler, M. Friedman, AE Robson, A. Bayramian, J. Caird, C. Ebbers, J. Latkowski, W. Hogan, WR Meier, LJ Perkins, K. Schaffers, S. Abdel ...
Fusion Engineering and Design, 2006
MHD flow and heat transfer have been analyzed for a front poloidal channel in the outboard module... more MHD flow and heat transfer have been analyzed for a front poloidal channel in the outboard module of a Dual Coolant Lithium Lead (DCLL) blanket, with a flow channel insert made of a silicon carbide composite. The US reference DCLL blanket module .] has been considered. Effectiveness of the insert as insulator was assessed via numerical simulations based on a 2D model for a fully developed flow and on a 3D model for heat transport. Parametric studies were performed at σ = 5-500 ( m) −1 and k = 2-20 W/m K. Parameters resulting in a reasonably low MHD pressure drop and almost no heat leakage from the breeder into the cooling helium flows have been identified.
Fusion Engineering and Design, 2004
Within the magnetic fusion energy program in the US, a program called APEX is investigating the u... more Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration, and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840 MW of fusion power of which 767 MW is in the form of energetic particles (alpha power) and 3073 MW is in the form of neutrons. The alpha plus auxiliary power total 909 MW of which 430 MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.
Fusion Engineering and Design, 2006
There is no practical external source of tritium for fusion energy development beyond ITER and al... more There is no practical external source of tritium for fusion energy development beyond ITER and all subsequent fusion systems have to breed their own tritium. To ensure tritium self-sufficiency, the calculated achievable tritium breeding ratio (TBR) should be equal to or greater than the required TBR. The potential of achieving tritium self-sufficiency depends on many system physics and technology parameters. Interactive physics and technology R&D programs should be implemented to determine the potential of realizing those physics and technology options and parameters that have large effects on attaining a realistic "window" for tritium self-sufficiency. The ranges of plasma and technology conditions that need to be met, in order to ensure tritium self-sufficiency, are identified.
Fusion Engineering and Design, 2006
An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) br... more An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC f /SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 • C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.
Fusion Engineering and Design, 2001
This study, called APEX, is exploring novel concepts for fusion chamber technology that can subst... more This study, called APEX, is exploring novel concepts for fusion chamber technology that can substantially improve the attractiveness of fusion energy systems. The emphasis of the study is on fundamental understanding and advancing the underlying engineering sciences, integration of the physics and engineering requirements, and enhancing innovation for the chamber technology components surrounding the plasma. The chamber technology goals in APEX include: (1) high power density capability with neutron wall load \ 10 MW/m 2 and surface heat flux \ 2 MW/m 2 , (2) high power conversion efficiency ( \ 40%), (3) high availability, and (4) simple technological and material constraints. Two classes of innovative concepts have emerged that offer great promise and deserve further research and development. The first class seeks to eliminate the solid ''bare'' first wall by flowing liquids facing the plasma. This liquid wall idea evolved during the APEX study into a number of concepts based on: (a) using liquid metals (Li or Sn-Li) or a molten salt (Flibe) as the working liquid, (b) utilizing electromagnetic, inertial and/or other types of forces to restrain the liquid against a backing wall and control the hydrodynamic flow configurations, and (c) employing a thin ( 2 cm) or thick ( 40 cm) liquid layer to remove the surface heat flux and attenuate the neutrons. These liquid wall concepts have some common features but also have widely different issues and merits. Some of the attractive features of liquid walls include the potential for: (1) high power density capability; (2) higher plasma b and stable physics regimes if liquid metals are used; (3) increased disruption survivability; (4) reduced volume of radioactive waste; (5) reduced radiation damage in structural materials; and (6) higher availability. Analyses show that not all of these potential advantages may be realized simultaneously in a single concept. However, the realization of only a subset of these advantages will result in remarkable progress toward attractive fusion energy systems. Of the many scientific and engineering issues for liquid walls, the most important are: (1) plasma-liquid interactions including both plasma-liquid surface and liquid wall-bulk plasma interactions; (2) hydrodynamic flow configuration control in complex geometries including penetrations; and (3) heat transfer at free surface and temperature control. The second class of concepts focuses on ideas for extending the capabilities, particularly the power density and operating temperature limits, of solid first walls. The most promising idea, called EVOLVE, is based on the use of a high-temperature refractory alloy (e.g. W -5% Re) with an innovative cooling scheme based on the use of the heat of vaporization of lithium. Calculations show that an evaporative system with Li at 1 200°C can remove the goal heat loads and result in a high power conversion efficiency. The vapor operating pressure is low, resulting in a very low operating stress in the structure. In addition, the lithium flow rate is about a factor of ten lower than that required for traditional self-cooled first wall/blanket concepts. Therefore, insulator coatings are not required. Key issues for EVOLVE include: (1) two-phase heat transfer and transport including MHD effects; (2) feasibility of fabricating entire blanket segments of W alloys; and (3) the effect of neutron irradiation on W.
Fusion Engineering and Design, 2010
Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lit... more Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li (~650°C to 700°C) from the RAF/M structure, which has a corrosion temperature limit of ~480°C. The RAF/M material must also operate at temperatures above 350°C but less than 550°C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R&D needs to address key areas. This paper is a summary report on the progress and results of recent DCLL TBM development activities.
Fusion Engineering and Design, 1998
An integrated study of compact stellarator power plants, ARIES-CS, has been conducted to explore ... more An integrated study of compact stellarator power plants, ARIES-CS, has been conducted to explore attractive compact stellarator configurations and to define key research and development (R&D) areas. The large size and mass predicted by earlier stellarator power plant studies had led to cost projections much higher than those of the advanced tokamak power plant. As such, the first major goal of the ARIES-CS research was to investigate if stellarator power plants can be made to be comparable in size to advanced tokamak variants while maintaining desirable stellarator properties. As stellarator fusion core components would have complex shapes and geometry, the second major goal of the ARIES-CS study was to understand and quantify, as much as possible, the impact of the complex shape and geometry of fusion core components. This paper focuses on the directions we pursued to optimize the compact stellarator as a fusion power plant, summarizes the major findings from the study, highlights the key design aspects and constraints associated with a compact stellarator, and identifies the major issues to help guide future R&D.
Within the ARIES-CS project, design activities have focused on developing the first compact devic... more Within the ARIES-CS project, design activities have focused on developing the first compact device that enhances the attractiveness of the stellarator as a power plant. The objectives of this paper are to review the nuclear elements that received considerable attention during the design process and provide a perspective on their successful integration into the final design. Among these elements are the radial build definition, the well-optimized in-vessel components that satisfy the ARIES top-level requirements, the carefully selected nuclear and engineering parameters to produce an economic optimum, the modeling-for the first time ever-of the highly complex stellarator geometry for the three-dimensional nuclear assessment, and the overarching safety and environmental constraints to deliver an attractive, reliable, and truly compact stellarator power plant.
The US blanket design activity has focused on the developments and the analyses of a solid breede... more The US blanket design activity has focused on the developments and the analyses of a solid breeder blanket concept for ITER. The main function of this blanket is to produce the necessary tritium required for the ITER operation and the test program. Safety, power reactor relevance, low tritium inventory, and design flexibility are the main reasons for the blanket selection.
Fusion Engineering and Design, 2015
Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 1998
As part of an ongoing exploration of final transport schemes based on plasma channels, a point de... more As part of an ongoing exploration of final transport schemes based on plasma channels, a point design of a final focus and reactor system is being developed. Six MJ of 3–4GeV Pb ions are delivered to a target in two opposing current-carrying plasma channels within a modified HYLIFE II reactor filled with 5Torr of Xe gas. The transition from the
Silicon carbide was first proposed as a low activation fusion reactor material in the mid 1970s. ... more Silicon carbide was first proposed as a low activation fusion reactor material in the mid 1970s. However, serious development of this material did not begin until the early 1990s, driven by the emergence of composite materials that provided enhanced toughness and an implied ability to use these typically brittle materials in engineering application. In the decades that followed, SiC composite
2 . Such effects will likely limit the performance and dominate failures in FNST components and m... more 2 . Such effects will likely limit the performance and dominate failures in FNST components and must be addressed to meet the extreme reliability and lifetime requirements of Demo fusion components. Data from experiments in the true fusion environment can give higher order insight into the evolution of multi-scale, multi-phenomena, multi-material synergies, including possible unanticipated effects and failure modes that
Fusion Science and Technology, 2005
Testing blanket concepts in the integrated fusion environment is one of the principal objectives ... more Testing blanket concepts in the integrated fusion environment is one of the principal objectives of ITER. Blanket test modules will be inserted in ITER from Day 1 of its operation and will provide the first experimental data on the feasibility of the D-T cycle for fusion. With the US rejoining ITER, the US community has decided to have strong participation
IEEE Transactions on Plasma Science, 2000
JD Sethian, DG Colombant, JL Giuliani, Jr., RH Lehmberg, MC Myers, SP Obenschain, AJ Schmitt, J. ... more JD Sethian, DG Colombant, JL Giuliani, Jr., RH Lehmberg, MC Myers, SP Obenschain, AJ Schmitt, J. Weaver, MF Wolford, F. Hegeler, M. Friedman, AE Robson, A. Bayramian, J. Caird, C. Ebbers, J. Latkowski, W. Hogan, WR Meier, LJ Perkins, K. Schaffers, S. Abdel ...
Fusion Engineering and Design, 2006
MHD flow and heat transfer have been analyzed for a front poloidal channel in the outboard module... more MHD flow and heat transfer have been analyzed for a front poloidal channel in the outboard module of a Dual Coolant Lithium Lead (DCLL) blanket, with a flow channel insert made of a silicon carbide composite. The US reference DCLL blanket module .] has been considered. Effectiveness of the insert as insulator was assessed via numerical simulations based on a 2D model for a fully developed flow and on a 3D model for heat transport. Parametric studies were performed at σ = 5-500 ( m) −1 and k = 2-20 W/m K. Parameters resulting in a reasonably low MHD pressure drop and almost no heat leakage from the breeder into the cooling helium flows have been identified.
Fusion Engineering and Design, 2004
Within the magnetic fusion energy program in the US, a program called APEX is investigating the u... more Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration, and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840 MW of fusion power of which 767 MW is in the form of energetic particles (alpha power) and 3073 MW is in the form of neutrons. The alpha plus auxiliary power total 909 MW of which 430 MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.
Fusion Engineering and Design, 2006
There is no practical external source of tritium for fusion energy development beyond ITER and al... more There is no practical external source of tritium for fusion energy development beyond ITER and all subsequent fusion systems have to breed their own tritium. To ensure tritium self-sufficiency, the calculated achievable tritium breeding ratio (TBR) should be equal to or greater than the required TBR. The potential of achieving tritium self-sufficiency depends on many system physics and technology parameters. Interactive physics and technology R&D programs should be implemented to determine the potential of realizing those physics and technology options and parameters that have large effects on attaining a realistic "window" for tritium self-sufficiency. The ranges of plasma and technology conditions that need to be met, in order to ensure tritium self-sufficiency, are identified.
Fusion Engineering and Design, 2006
An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) br... more An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC f /SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 • C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.
Fusion Engineering and Design, 2001
This study, called APEX, is exploring novel concepts for fusion chamber technology that can subst... more This study, called APEX, is exploring novel concepts for fusion chamber technology that can substantially improve the attractiveness of fusion energy systems. The emphasis of the study is on fundamental understanding and advancing the underlying engineering sciences, integration of the physics and engineering requirements, and enhancing innovation for the chamber technology components surrounding the plasma. The chamber technology goals in APEX include: (1) high power density capability with neutron wall load \ 10 MW/m 2 and surface heat flux \ 2 MW/m 2 , (2) high power conversion efficiency ( \ 40%), (3) high availability, and (4) simple technological and material constraints. Two classes of innovative concepts have emerged that offer great promise and deserve further research and development. The first class seeks to eliminate the solid ''bare'' first wall by flowing liquids facing the plasma. This liquid wall idea evolved during the APEX study into a number of concepts based on: (a) using liquid metals (Li or Sn-Li) or a molten salt (Flibe) as the working liquid, (b) utilizing electromagnetic, inertial and/or other types of forces to restrain the liquid against a backing wall and control the hydrodynamic flow configurations, and (c) employing a thin ( 2 cm) or thick ( 40 cm) liquid layer to remove the surface heat flux and attenuate the neutrons. These liquid wall concepts have some common features but also have widely different issues and merits. Some of the attractive features of liquid walls include the potential for: (1) high power density capability; (2) higher plasma b and stable physics regimes if liquid metals are used; (3) increased disruption survivability; (4) reduced volume of radioactive waste; (5) reduced radiation damage in structural materials; and (6) higher availability. Analyses show that not all of these potential advantages may be realized simultaneously in a single concept. However, the realization of only a subset of these advantages will result in remarkable progress toward attractive fusion energy systems. Of the many scientific and engineering issues for liquid walls, the most important are: (1) plasma-liquid interactions including both plasma-liquid surface and liquid wall-bulk plasma interactions; (2) hydrodynamic flow configuration control in complex geometries including penetrations; and (3) heat transfer at free surface and temperature control. The second class of concepts focuses on ideas for extending the capabilities, particularly the power density and operating temperature limits, of solid first walls. The most promising idea, called EVOLVE, is based on the use of a high-temperature refractory alloy (e.g. W -5% Re) with an innovative cooling scheme based on the use of the heat of vaporization of lithium. Calculations show that an evaporative system with Li at 1 200°C can remove the goal heat loads and result in a high power conversion efficiency. The vapor operating pressure is low, resulting in a very low operating stress in the structure. In addition, the lithium flow rate is about a factor of ten lower than that required for traditional self-cooled first wall/blanket concepts. Therefore, insulator coatings are not required. Key issues for EVOLVE include: (1) two-phase heat transfer and transport including MHD effects; (2) feasibility of fabricating entire blanket segments of W alloys; and (3) the effect of neutron irradiation on W.
Fusion Engineering and Design, 2010
Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lit... more Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li (~650°C to 700°C) from the RAF/M structure, which has a corrosion temperature limit of ~480°C. The RAF/M material must also operate at temperatures above 350°C but less than 550°C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R&D needs to address key areas. This paper is a summary report on the progress and results of recent DCLL TBM development activities.
Fusion Engineering and Design, 1998
An integrated study of compact stellarator power plants, ARIES-CS, has been conducted to explore ... more An integrated study of compact stellarator power plants, ARIES-CS, has been conducted to explore attractive compact stellarator configurations and to define key research and development (R&D) areas. The large size and mass predicted by earlier stellarator power plant studies had led to cost projections much higher than those of the advanced tokamak power plant. As such, the first major goal of the ARIES-CS research was to investigate if stellarator power plants can be made to be comparable in size to advanced tokamak variants while maintaining desirable stellarator properties. As stellarator fusion core components would have complex shapes and geometry, the second major goal of the ARIES-CS study was to understand and quantify, as much as possible, the impact of the complex shape and geometry of fusion core components. This paper focuses on the directions we pursued to optimize the compact stellarator as a fusion power plant, summarizes the major findings from the study, highlights the key design aspects and constraints associated with a compact stellarator, and identifies the major issues to help guide future R&D.
Within the ARIES-CS project, design activities have focused on developing the first compact devic... more Within the ARIES-CS project, design activities have focused on developing the first compact device that enhances the attractiveness of the stellarator as a power plant. The objectives of this paper are to review the nuclear elements that received considerable attention during the design process and provide a perspective on their successful integration into the final design. Among these elements are the radial build definition, the well-optimized in-vessel components that satisfy the ARIES top-level requirements, the carefully selected nuclear and engineering parameters to produce an economic optimum, the modeling-for the first time ever-of the highly complex stellarator geometry for the three-dimensional nuclear assessment, and the overarching safety and environmental constraints to deliver an attractive, reliable, and truly compact stellarator power plant.
The US blanket design activity has focused on the developments and the analyses of a solid breede... more The US blanket design activity has focused on the developments and the analyses of a solid breeder blanket concept for ITER. The main function of this blanket is to produce the necessary tritium required for the ITER operation and the test program. Safety, power reactor relevance, low tritium inventory, and design flexibility are the main reasons for the blanket selection.