The semi-catalyzed-deuterium Light-Water Hybrid Reactor (LWHR) comprises a lithium-free light-wat... more The semi-catalyzed-deuterium Light-Water Hybrid Reactor (LWHR) comprises a lithium-free light-water-moderated blanket with U/sub 3/Si fuel driven by a deuterium-based fusion-neutron source, with complete burn-up of the tritium but almost no burn-up of the helium-3 reaction product. A one-dimensional model for a neutral-beam-driven tokamak plasma is used to determine the operating modes under which the fusion energy multiplication Q/sub p/ can be equal to or greater than 0.5. Thermonuclear, beam-target, and energetic-ion reactions are taken into account. The most feasible operating conditions for Q/sub p/ approximately 0.5 are tau/sub E/ = 2 to 4 x 10/sup 14/ cm/sup -3/s, = 10 to 20 keV, and E/sub beam/ = 500 to 1000 keV, with approximately 40% of the fusion energy produced by beam-target reactions. Illustrative parameters of LWHRs are compared with those of an ignited D-T reactor.
Schematics of D-D based fusion fuel cycles 4 2 Effect of PCA and H 2 O volume fraction on the tri... more Schematics of D-D based fusion fuel cycles 4 2 Effect of PCA and H 2 O volume fraction on the tritium breeding per D-T neutron. 3 He volume fraction is 10% 12 3 Effect of PCA and H 2 O volume fraction on the tritium breeding per D-D neutron. 3 He volume fraction is 10Z. 12 4 Effect of 3 He volume fraction on the tritium breeding per D-T neutron. PCA volume fraction is 10% 13 5 Effect of 3 He volume fraction on the tritium breeding per D-D neutron. PCA volume fraction is 10% 13 6 3 He requirements for attaining a given tritium breeding ratio per D-T neutron. 15 7 3 Ke requirements for attaining a given tritium breeding ratio per D-D neutron 15 8 Effect of 3 He-to-PCA volume ratio on the energy multiplication of D-T (e,j_ T) and D-D [ejy_ D ] neutrons for PCA:H 2 O: 3 He blankets 19 9 Effect of tritium breeding ratio on the energy multiplication (e) attainable from PCA:H 2 O: 3 He blankets 21 Effect of the fraction of 3 He escaping fusion (a) on the deuteron density (n D), major radius (R) and maximal magnetic field strength (] of TBR-0.4 tokamaks. .« 31 Effect of the fraction of 3 He escaping fusion (a) on the D-T neutron first wall loading (Tp-x) and fusion power density of TBR-0.4 tokamaks 31 Effect of the fraction of 3 He escaping fusion (a) on the total capital cost (TCC) and cost of electricity (COE) of TBR-0.4 tokamaks 32 13 Selected characteristics of TCD tokaoaks, including major radius (R); maximal magnetic field for B-10Z (B max); fraction of 3 He escaping fusion (a); tritium breeding ratio (TBR); cost of electricity (COE); deuterons density (n D); confinement parameter (n-r); D-T neutrons first wall loading 3r D _ T); ratio of D-T to D-D neutrons (D-T/D-Dn); and average electron temperature (T e)» 3 He recycling coefficient R 3-0.99 32
This report is intended• as a User' s:'.~Manual for "SWAN"-a code written for perturbation theory... more This report is intended• as a User' s:'.~Manual for "SWAN"-a code written for perturbation theory ' analysis and optimization of the nucleonic characteristics of fusion reactor blankets. SWAN is also applicable to any problem described by the inhomo-.geneous linear transport equation, e.g., radiation shield problems • The optimization method is based on variational techniques. The variables of the optimization are •. . • material densi.t~es, with no restriction on their .. number. One joint restraint may be imposed in addition to upper and lower limits on each density. The param-et~r to be extremized and the restraint may be either a weight-type (linear) or nucleonic (bilinear) functional. The transport calculations for SWAN are performed ! with the one-dimensional di•scr-et~-ordinate code ANISN.
Assuming that the release, transport, and deposition of fissile material stored in a repository w... more Assuming that the release, transport, and deposition of fissile material stored in a repository will establish a critical configuration, the energy yield may be of concern if it has a positive temperature reactivity coefficient. This work shows that a heterogeneous distributions of highly enriched uranium in rock plus water can have a significant positive temperature coefficient of reactivity due to spectrum hardening.
An advanced assembly code system that can efficiently and accurately analyze various designs (cur... more An advanced assembly code system that can efficiently and accurately analyze various designs (current and advanced) proposed for plutonium disposition is being developed by {open_quotes}marrying{close_quotes} two existing state-of-the-art methodologies-GTRAN2 and SCALE 4.2. The resulting code system, GT-SCALE, posses several unique characteristics: exact 2D representation of a complete fuel assembly, while preserving the heterogeniety of each of its pin cells; flexibility in the energy group structure, the present upper limit being 218 groups; a comprehensive cross-section library and material data base; and accurate burnup calculations. The resulting GT-SCALE is expected to be very useful for a wide variety of applications, including the analysis of very heterogeneous UOâ fueled LWR fuel assemblies; of hexagonal shaped fuel assemblies as of the Russian LWRs; of fuel assemblies for HTGRs; as well as for the analysis of criticality safety and for calculation of the source term of spe...
Molten salt reactors (MSRs) are liquid-fueled reactors that can be used for burning actinides, pr... more Molten salt reactors (MSRs) are liquid-fueled reactors that can be used for burning actinides, production of electricity, production of hydrogen, and production of fissile fuels (breeding). Fissile, fertile, and fission products are dissolved in a high-temperature molten fluoride salt with a very high boiling temperature (-1400EC). The molten salt serves as both the reactor fuel and the coolant. Heat is generated in the reactor core and transported by the fuel salt to heat exchangers before returning to the reactor core. The MSR is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development and is the only fluid-fuel concept. Liquid-fueled reactors have several unique capabilities in terms of burning actinide elements and thus offer the potential to reduce the long-term radiotoxicity of the waste from production of electricity. This is the basis for much of the renewed interest in MSRs. MSRs (compared with s...
This study investigates the economic prospects of T-assisted D-based Fusion Breeders (FB's) ... more This study investigates the economic prospects of T-assisted D-based Fusion Breeders (FB's) in comparison with a conventional D-T FB.The D-based driver uses the Tritium-Catalyzed-Deuterium (TCD) mode of operation where the ³He leaking out from the plasma is converted into T in the blanket (by neutron capture). The T produced is then added back to the plasma. TCD operation is selected because it offers a significantly higher neutron yield and support-ratio than the standard Catalyzed-D. It offers also higher neutron yield and better plasma energy balance than the standard Semi-Catalyzed-D mode.
An economic analysis is performed to calculate the levelized unit cost of electricity (COE) for a... more An economic analysis is performed to calculate the levelized unit cost of electricity (COE) for a pressurized water reactor (PWR) retrofitted with a range of potential U (45wt.%)–ZrH1.6 hydride and UO2 oxide fueled geometries (i.e., combinations of rod diameter and pitch) supported by traditional grid spacers (square array) and wire wrap spacers (hexagonal array). The time frame considered in computing
The classic approach to the recycling of Pu in PWR is to use mixed U-oxide Pu-oxide (MOX) fuel. T... more The classic approach to the recycling of Pu in PWR is to use mixed U-oxide Pu-oxide (MOX) fuel. The mono-recycling of plutonium in PWR transmutes less than 30% of the loaded plutonium, providing only a limited reduction in the long-term radiotoxicity and in the inventory of TRU to be stored in the repository. The primary objective of this study is to assess the feasibility of plutonium recycling in PWR in the form of plutonium hydride, PuH2, mixed with uranium and zirconium hydride, ZrH1.6, referred to as PUZH, that is loaded uniformly in each fuel rod. The assessment is performed by comparing the performance of the PUZH fueled core to that of the MOX fueled core. Performance characteristics examined are transmutation effectiveness, proliferation resistance of the discharged fuel and fuel cycle economics. The PUZH loaded core is found superior to the MOX fueled core in terms of the transmutation effectiveness and proliferation resistance. For the reference cycle duration and reference fuel rod diameter and pitch, the percentage of the plutonium loaded that is transmuted in one recycle is 53% for PUZH versus 29% for MOX fuel. That is, the net amount of plutonium transmuted in the first recycle is 55% higher in cores using PUZH than in cores using MOX fuel. Relative to the discharged MOX, the discharged PUZH fuel has smaller fissile plutonium fraction – 45% versus 60%, 15% smaller minor actinides (MA) inventory and more than double spontaneous fission neutron source intensity and decay heat per gram of discharged TRU. Relative to the MOX fuel assembly, the radioactivity of the PUZH fuel assembly is 26% smaller and the decay heat and the neutron yield are only 3% larger. The net effect is that the handling of the discharged PUZH fuel assembly will be comparable in difficulty to that of the discharged MOX assembly while the proliferation resistance of the TRU of the discharged PUZH fuel is enhanced.
The economic implications of designing BWR cores with hydride fuels instead of conventional oxide... more The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010$/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the
The neutronic properties of U–ZrH1.6 fuelled PWR cores are investigated and compared against thos... more The neutronic properties of U–ZrH1.6 fuelled PWR cores are investigated and compared against those of the currently used UO2 fuelled cores. In the first part of this work a parametric study is performed to quantify the neutronically achievable burnup for both hydride and oxide fuels at a number of enrichment levels and for a large number of geometries covering a wide design space of fuel rod outer diameter, D, and lattice pitch, P. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are calculated for hydride fuel with 5% and 12.5% enriched uranium. For this purpose a simplified procedure was developed that can, using single unit cell or assembly calculations, (1) account for non-linear burnup dependent k∞ and thus to adequately predict the discharge burnup; (2) estimate the burnup dependent soluble boron concentration and; (3) estimate the reactivity coefficients; all of the above for a multi-batch core. In the second part of this work a detailed neutronic analysis is carried out for the six most economical geometries of both oxide and hydride fuels, with the purpose of designing the U–ZrH1.6 fueled PWR cores to have negative reactivity coefficients. The preferred design found is replacement of 25 v/o of the ZrH1.6 by thorium hydride, along with addition of some IFBA burnable poison. It is also found that the conversion from oxide to hydride fueled PWR cores could be done without modifications in the control system.
ABSTRACT This paper summarizes the neutronic part of a study of the feasibility of designing BWR ... more ABSTRACT This paper summarizes the neutronic part of a study of the feasibility of designing BWR cores to have enhanced power density and simplified fuel bundle by using hydride instead of oxide fuel. A 3D fuel bundle neutronic analysis is performed for a limited number of geometries to determine attainable discharge burnup, pin-by-pin power distribution, axial power distribution, reactivity coefficients, reactivity worth of control elements and burnable absorber effects. It is found that hydride fuel bundle design can be simplified by eliminating water rods and partial length fuel rods and by reducing the volume of water in-between the fuel bundles. Both an ideal and more practical bundle designs are examined. A companion study of the thermal–hydraulic and vibration characteristics of BWR cores predicts that the increase in the number of fuel rods per given core volume enables increasing the BWR power density by up to ∼30% relative to oxide fuelled core design. The net outcome is expected to be improved BWR economics even though hydride fuel requires higher uranium enrichment to compensate for its reduced uranium loading.
A preliminary assessment is conducted for the autocatalytic criticality phenomena for the Japanes... more A preliminary assessment is conducted for the autocatalytic criticality phenomena for the Japanese repository concept, where the vitrified high-level waste (HLW) from reprocessed fuel of commercial light-water reactors is assumed to be disposed of in fractured granitic rock saturated with water. By a static neutronic analysis, the mass of 235U in 12%-enriched uranium required to make the transition from an under-moderated to an over-moderated critical system in water-saturated granite of a porosity of 0.3 is obtained 35 kg in a spherical core with a radius of 90 cm. The radionuclide transport analysis indicates that in any case the only fissile nuclide of concern for the autocatalytic criticality scenario is 235U. Other fissile nuclides are too short-lived, or are accompanied with a strong neutron absorber. This preliminary study indicates that the transport model can be simplified for uranium isotopes, with which a statistical uncertainty analysis can be performed to finally determine whether or not the autocatalytic criticality event can be screened out from the scenarios for the generic performance assessment of the conceptual repository.
The effect of a uniform magnetic field on the heat transfer of liquid metal forced flows in strai... more The effect of a uniform magnetic field on the heat transfer of liquid metal forced flows in straight rectangular channels subjected to a uniform heat flux on one of the walls parallel to the field is experimentally studied. The experiments covered the domain 3.8 Q U d 18 cm/s; 0 d B d 0.9 T; 0 d M d 300 and 4 X 10' Q Re < 2 X 104. It is found that the application of the magnetic field improves the heat transfer in channels made of both non-conducting walls and of conducting walls. Whereas in the former case this improvement results from the creation of enhanced anisotropic turbulence across he entire channel, in the conducting channel the improvement results from an increase in the velocity near the heated wall as well as enhancement of turbulence in the vicinity of this wall.
The cost-effectiveness of new polyethylene-based materials for shielding against 14 MeV neutrons ... more The cost-effectiveness of new polyethylene-based materials for shielding against 14 MeV neutrons is investigated. The new materials consist of a powder of a medium-density, low-activation, low-cost material which is bonded by polyethylene. These new "polynated" materials are found to be more attractive than polyethylene-based commercial shielding materials; they enable to design more compact, and lower cost shields. Of the materials studied, "poly-marble" was identified the most cost-effective; it uses CaCO 3 for the powder.
The semi-catalyzed-deuterium Light-Water Hybrid Reactor (LWHR) comprises a lithium-free light-wat... more The semi-catalyzed-deuterium Light-Water Hybrid Reactor (LWHR) comprises a lithium-free light-water-moderated blanket with U/sub 3/Si fuel driven by a deuterium-based fusion-neutron source, with complete burn-up of the tritium but almost no burn-up of the helium-3 reaction product. A one-dimensional model for a neutral-beam-driven tokamak plasma is used to determine the operating modes under which the fusion energy multiplication Q/sub p/ can be equal to or greater than 0.5. Thermonuclear, beam-target, and energetic-ion reactions are taken into account. The most feasible operating conditions for Q/sub p/ approximately 0.5 are tau/sub E/ = 2 to 4 x 10/sup 14/ cm/sup -3/s, = 10 to 20 keV, and E/sub beam/ = 500 to 1000 keV, with approximately 40% of the fusion energy produced by beam-target reactions. Illustrative parameters of LWHRs are compared with those of an ignited D-T reactor.
Schematics of D-D based fusion fuel cycles 4 2 Effect of PCA and H 2 O volume fraction on the tri... more Schematics of D-D based fusion fuel cycles 4 2 Effect of PCA and H 2 O volume fraction on the tritium breeding per D-T neutron. 3 He volume fraction is 10% 12 3 Effect of PCA and H 2 O volume fraction on the tritium breeding per D-D neutron. 3 He volume fraction is 10Z. 12 4 Effect of 3 He volume fraction on the tritium breeding per D-T neutron. PCA volume fraction is 10% 13 5 Effect of 3 He volume fraction on the tritium breeding per D-D neutron. PCA volume fraction is 10% 13 6 3 He requirements for attaining a given tritium breeding ratio per D-T neutron. 15 7 3 Ke requirements for attaining a given tritium breeding ratio per D-D neutron 15 8 Effect of 3 He-to-PCA volume ratio on the energy multiplication of D-T (e,j_ T) and D-D [ejy_ D ] neutrons for PCA:H 2 O: 3 He blankets 19 9 Effect of tritium breeding ratio on the energy multiplication (e) attainable from PCA:H 2 O: 3 He blankets 21 Effect of the fraction of 3 He escaping fusion (a) on the deuteron density (n D), major radius (R) and maximal magnetic field strength (] of TBR-0.4 tokamaks. .« 31 Effect of the fraction of 3 He escaping fusion (a) on the D-T neutron first wall loading (Tp-x) and fusion power density of TBR-0.4 tokamaks 31 Effect of the fraction of 3 He escaping fusion (a) on the total capital cost (TCC) and cost of electricity (COE) of TBR-0.4 tokamaks 32 13 Selected characteristics of TCD tokaoaks, including major radius (R); maximal magnetic field for B-10Z (B max); fraction of 3 He escaping fusion (a); tritium breeding ratio (TBR); cost of electricity (COE); deuterons density (n D); confinement parameter (n-r); D-T neutrons first wall loading 3r D _ T); ratio of D-T to D-D neutrons (D-T/D-Dn); and average electron temperature (T e)» 3 He recycling coefficient R 3-0.99 32
This report is intended• as a User' s:'.~Manual for "SWAN"-a code written for perturbation theory... more This report is intended• as a User' s:'.~Manual for "SWAN"-a code written for perturbation theory ' analysis and optimization of the nucleonic characteristics of fusion reactor blankets. SWAN is also applicable to any problem described by the inhomo-.geneous linear transport equation, e.g., radiation shield problems • The optimization method is based on variational techniques. The variables of the optimization are •. . • material densi.t~es, with no restriction on their .. number. One joint restraint may be imposed in addition to upper and lower limits on each density. The param-et~r to be extremized and the restraint may be either a weight-type (linear) or nucleonic (bilinear) functional. The transport calculations for SWAN are performed ! with the one-dimensional di•scr-et~-ordinate code ANISN.
Assuming that the release, transport, and deposition of fissile material stored in a repository w... more Assuming that the release, transport, and deposition of fissile material stored in a repository will establish a critical configuration, the energy yield may be of concern if it has a positive temperature reactivity coefficient. This work shows that a heterogeneous distributions of highly enriched uranium in rock plus water can have a significant positive temperature coefficient of reactivity due to spectrum hardening.
An advanced assembly code system that can efficiently and accurately analyze various designs (cur... more An advanced assembly code system that can efficiently and accurately analyze various designs (current and advanced) proposed for plutonium disposition is being developed by {open_quotes}marrying{close_quotes} two existing state-of-the-art methodologies-GTRAN2 and SCALE 4.2. The resulting code system, GT-SCALE, posses several unique characteristics: exact 2D representation of a complete fuel assembly, while preserving the heterogeniety of each of its pin cells; flexibility in the energy group structure, the present upper limit being 218 groups; a comprehensive cross-section library and material data base; and accurate burnup calculations. The resulting GT-SCALE is expected to be very useful for a wide variety of applications, including the analysis of very heterogeneous UOâ fueled LWR fuel assemblies; of hexagonal shaped fuel assemblies as of the Russian LWRs; of fuel assemblies for HTGRs; as well as for the analysis of criticality safety and for calculation of the source term of spe...
Molten salt reactors (MSRs) are liquid-fueled reactors that can be used for burning actinides, pr... more Molten salt reactors (MSRs) are liquid-fueled reactors that can be used for burning actinides, production of electricity, production of hydrogen, and production of fissile fuels (breeding). Fissile, fertile, and fission products are dissolved in a high-temperature molten fluoride salt with a very high boiling temperature (-1400EC). The molten salt serves as both the reactor fuel and the coolant. Heat is generated in the reactor core and transported by the fuel salt to heat exchangers before returning to the reactor core. The MSR is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development and is the only fluid-fuel concept. Liquid-fueled reactors have several unique capabilities in terms of burning actinide elements and thus offer the potential to reduce the long-term radiotoxicity of the waste from production of electricity. This is the basis for much of the renewed interest in MSRs. MSRs (compared with s...
This study investigates the economic prospects of T-assisted D-based Fusion Breeders (FB's) ... more This study investigates the economic prospects of T-assisted D-based Fusion Breeders (FB's) in comparison with a conventional D-T FB.The D-based driver uses the Tritium-Catalyzed-Deuterium (TCD) mode of operation where the ³He leaking out from the plasma is converted into T in the blanket (by neutron capture). The T produced is then added back to the plasma. TCD operation is selected because it offers a significantly higher neutron yield and support-ratio than the standard Catalyzed-D. It offers also higher neutron yield and better plasma energy balance than the standard Semi-Catalyzed-D mode.
An economic analysis is performed to calculate the levelized unit cost of electricity (COE) for a... more An economic analysis is performed to calculate the levelized unit cost of electricity (COE) for a pressurized water reactor (PWR) retrofitted with a range of potential U (45wt.%)–ZrH1.6 hydride and UO2 oxide fueled geometries (i.e., combinations of rod diameter and pitch) supported by traditional grid spacers (square array) and wire wrap spacers (hexagonal array). The time frame considered in computing
The classic approach to the recycling of Pu in PWR is to use mixed U-oxide Pu-oxide (MOX) fuel. T... more The classic approach to the recycling of Pu in PWR is to use mixed U-oxide Pu-oxide (MOX) fuel. The mono-recycling of plutonium in PWR transmutes less than 30% of the loaded plutonium, providing only a limited reduction in the long-term radiotoxicity and in the inventory of TRU to be stored in the repository. The primary objective of this study is to assess the feasibility of plutonium recycling in PWR in the form of plutonium hydride, PuH2, mixed with uranium and zirconium hydride, ZrH1.6, referred to as PUZH, that is loaded uniformly in each fuel rod. The assessment is performed by comparing the performance of the PUZH fueled core to that of the MOX fueled core. Performance characteristics examined are transmutation effectiveness, proliferation resistance of the discharged fuel and fuel cycle economics. The PUZH loaded core is found superior to the MOX fueled core in terms of the transmutation effectiveness and proliferation resistance. For the reference cycle duration and reference fuel rod diameter and pitch, the percentage of the plutonium loaded that is transmuted in one recycle is 53% for PUZH versus 29% for MOX fuel. That is, the net amount of plutonium transmuted in the first recycle is 55% higher in cores using PUZH than in cores using MOX fuel. Relative to the discharged MOX, the discharged PUZH fuel has smaller fissile plutonium fraction – 45% versus 60%, 15% smaller minor actinides (MA) inventory and more than double spontaneous fission neutron source intensity and decay heat per gram of discharged TRU. Relative to the MOX fuel assembly, the radioactivity of the PUZH fuel assembly is 26% smaller and the decay heat and the neutron yield are only 3% larger. The net effect is that the handling of the discharged PUZH fuel assembly will be comparable in difficulty to that of the discharged MOX assembly while the proliferation resistance of the TRU of the discharged PUZH fuel is enhanced.
The economic implications of designing BWR cores with hydride fuels instead of conventional oxide... more The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010$/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the
The neutronic properties of U–ZrH1.6 fuelled PWR cores are investigated and compared against thos... more The neutronic properties of U–ZrH1.6 fuelled PWR cores are investigated and compared against those of the currently used UO2 fuelled cores. In the first part of this work a parametric study is performed to quantify the neutronically achievable burnup for both hydride and oxide fuels at a number of enrichment levels and for a large number of geometries covering a wide design space of fuel rod outer diameter, D, and lattice pitch, P. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are calculated for hydride fuel with 5% and 12.5% enriched uranium. For this purpose a simplified procedure was developed that can, using single unit cell or assembly calculations, (1) account for non-linear burnup dependent k∞ and thus to adequately predict the discharge burnup; (2) estimate the burnup dependent soluble boron concentration and; (3) estimate the reactivity coefficients; all of the above for a multi-batch core. In the second part of this work a detailed neutronic analysis is carried out for the six most economical geometries of both oxide and hydride fuels, with the purpose of designing the U–ZrH1.6 fueled PWR cores to have negative reactivity coefficients. The preferred design found is replacement of 25 v/o of the ZrH1.6 by thorium hydride, along with addition of some IFBA burnable poison. It is also found that the conversion from oxide to hydride fueled PWR cores could be done without modifications in the control system.
ABSTRACT This paper summarizes the neutronic part of a study of the feasibility of designing BWR ... more ABSTRACT This paper summarizes the neutronic part of a study of the feasibility of designing BWR cores to have enhanced power density and simplified fuel bundle by using hydride instead of oxide fuel. A 3D fuel bundle neutronic analysis is performed for a limited number of geometries to determine attainable discharge burnup, pin-by-pin power distribution, axial power distribution, reactivity coefficients, reactivity worth of control elements and burnable absorber effects. It is found that hydride fuel bundle design can be simplified by eliminating water rods and partial length fuel rods and by reducing the volume of water in-between the fuel bundles. Both an ideal and more practical bundle designs are examined. A companion study of the thermal–hydraulic and vibration characteristics of BWR cores predicts that the increase in the number of fuel rods per given core volume enables increasing the BWR power density by up to ∼30% relative to oxide fuelled core design. The net outcome is expected to be improved BWR economics even though hydride fuel requires higher uranium enrichment to compensate for its reduced uranium loading.
A preliminary assessment is conducted for the autocatalytic criticality phenomena for the Japanes... more A preliminary assessment is conducted for the autocatalytic criticality phenomena for the Japanese repository concept, where the vitrified high-level waste (HLW) from reprocessed fuel of commercial light-water reactors is assumed to be disposed of in fractured granitic rock saturated with water. By a static neutronic analysis, the mass of 235U in 12%-enriched uranium required to make the transition from an under-moderated to an over-moderated critical system in water-saturated granite of a porosity of 0.3 is obtained 35 kg in a spherical core with a radius of 90 cm. The radionuclide transport analysis indicates that in any case the only fissile nuclide of concern for the autocatalytic criticality scenario is 235U. Other fissile nuclides are too short-lived, or are accompanied with a strong neutron absorber. This preliminary study indicates that the transport model can be simplified for uranium isotopes, with which a statistical uncertainty analysis can be performed to finally determine whether or not the autocatalytic criticality event can be screened out from the scenarios for the generic performance assessment of the conceptual repository.
The effect of a uniform magnetic field on the heat transfer of liquid metal forced flows in strai... more The effect of a uniform magnetic field on the heat transfer of liquid metal forced flows in straight rectangular channels subjected to a uniform heat flux on one of the walls parallel to the field is experimentally studied. The experiments covered the domain 3.8 Q U d 18 cm/s; 0 d B d 0.9 T; 0 d M d 300 and 4 X 10' Q Re < 2 X 104. It is found that the application of the magnetic field improves the heat transfer in channels made of both non-conducting walls and of conducting walls. Whereas in the former case this improvement results from the creation of enhanced anisotropic turbulence across he entire channel, in the conducting channel the improvement results from an increase in the velocity near the heated wall as well as enhancement of turbulence in the vicinity of this wall.
The cost-effectiveness of new polyethylene-based materials for shielding against 14 MeV neutrons ... more The cost-effectiveness of new polyethylene-based materials for shielding against 14 MeV neutrons is investigated. The new materials consist of a powder of a medium-density, low-activation, low-cost material which is bonded by polyethylene. These new "polynated" materials are found to be more attractive than polyethylene-based commercial shielding materials; they enable to design more compact, and lower cost shields. Of the materials studied, "poly-marble" was identified the most cost-effective; it uses CaCO 3 for the powder.
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Papers by Ehud Greenspan