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01298 On the nuclear response of the water-cooled Pb-17Li test blanket module for ITER-FEAT
05 Nuclear fuels (scientific, technical) 04/01297 Numerical simulation of a free-falling liquid sodium droplet combustion Okano, Y. and Yamaguchi, A. Annals of Nuclear Energy, 2003, 30, (18), 1863-1878. Droplet combustion is considered one of the combustion figurations in sodium-fire events; the detailed combustion mechanism by which it occurs is evaluated by a developed computational fluid dynamic (CFD) simulation code, called COMET, in which the extended MAC method is employed for calculating the reacting compressible flow coupled with the multi-component diffusion of chemical species. A single droplet combustion in steady-airflow was simulated using the COMET to analyse: (1) the spatial distributions of heat rate and temperature, and (2) the formation, decomposition, and transfer of combustion products. Next, a free-falling droplet combustion experiment was simulated for code validation, where the evaluated falling-velocity and burnt-mass showed good agreement with the experimental values.
04/01298 On the nuclear response of the water-cooled Pb17Li test blanket module for ITER-FEAT Chiovaro, P. et al. Fusion Engineering and Design, 2003, 69, (1-4), 469477. Within the European Fusion Technology Programme, the WaterCooled Lithium Lead (WCLL) D E M O breeding blanket line was selected in 1995 as one of the two EU lines to be developed in the next decades, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be tested in ITER-FEAT. The present paper is focused on the study of the WCLL-TBM nuclear response in ITERFEAT, being specifically oriented to the investigation of the local effects due to the typical C-shaped tubes of the breeder zone, since they could play a pivotal role in the module-relevant t h e r m o mechanical design. A 3D heterogeneous model of the WCLL-TBM, realistically simulating its new lay out and taking into account 9% Cr rnartensitic steel as reference structural material, has been set-up. A particular attention has been paid to the simulation of the characteristic 'C' shape of the breeder zone double walled tubes, which have been realistically reproduced. The WCLL-TBM model has been inserted into an existing ITER-FEAT 3D semi-heterogeneous model accounting for a proper D - T neutron source. Analyses have been performed by means of MCNP-4C code running On a cluster Of four workstations through the implementation of a parallel virtual machine. For each analysis a large number of histories (> 10 000 000) have been simulated, obtaining statistical uncertainties on the results lower than 3%. The main features of the WCLL-TBM nuclear response have been determined focusing the attention on power deposition density, material damage through DPA and He and H production rate, daily tritium production and tritium production rate radial distribution in the module. The obtained results are presented and critically discussed.
04/01299
On the use of silicon as thermal neutron filter
Adib, M. et al. Annals of Nuclear Energy, 2003, 30, (18), 1905-1917. A simple formula is given which allows to calculate the contribution of the total neutron cross-section including the Bragg scattering from different (hkl) planes to the neutron transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy. A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500 peV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given.
04/01300 Optical absorption and luminescence of 14-MeV neutron-irradiated CaF2 single crystals Cooke, D. W. and Bennett, B. L. Journal of Nuclear Materials', 2003, 321, (2-3), 158-164. The effects of 14-MeV neutron irradiation (1.1 x 1019 n/m 2) on crystalline CaF2 have been examined by optical absorption and luminescence techniques to evaluate its suitability as a window material for fusion energy applications. For comparison, similar studies were done on un-irradiated and X-irradiated samples. It is confirmed that pristine CaFa exhibits excellent optical transmission in the spectral region 200-1000 nm. X and neutron irradiation induce similar optical absorption spectra with maximum absorption coefficients approximately 1.6 and 0.8 cm -1, respectively. Thermally stimulated luminescence glow curves are induced by X-ray (11.55 kGy) and neutron exposures; peaks occur at 423, 534, 596 and 479, 550, 605 K, respectively. Thermal annealing experiments show that the major
178
Fuel and Energy Abstracts
May 2004
absorption peaks decay in concert with appearance of the first glow peak, which is attributed to an electron trap. Thus, the major absorption bands are associated with F and F-aggregate centres. The relative ease with which these centres are produced strongly suggests that CaF2 is not a good final optic window material for fusion energy applications.
04/01301 Reaction layer growth and reaction heat of U-Mo/ AI dispersion fuels using centrifugally atomized powders Ryu, H. J. et al. Journal of Nuclear Materials, 2003, 321, (2-3), 210220. The growth behavionr of reaction layers and heat generation during the reaction between U - M o powders and the A1 matrix in U-Mo/AI dispersion fuels were investigated. Annealing of 10 vol.% U-10Mo/AI dispersion fuels at temperatures from 500 to 550°C was carried out for 10 min to 36 h to measure the growth rate and the activation energy for the growth of reaction layers. The concentration profiles of reaction layers between the U-10Mo vs A1 diffusion couples were measured and the integrated inter-diffusion coefficients were calculated for the U and A1 in the reaction layers. Heat generation of U-Mo/AI dispersion fuels with 10-50 vol.% of U - M o fuel during the thermal cycle from room temperature to 700°C was measured employing the differential scanning calorimetry. Exotherrnic heat from the reaction between U Mo and the AI matrix is the largest when the volume fraction of U - M o fuel is about 30 vol.%. The un-reacted fraction in the U - M o powders increases as the volume fraction of U - M o fuel increases from 30 to 50 vol.%.
04/01302 Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels Bondarenko, A. V. et al. Journal of Nuclear Materials, 2003, 319, (1), 159-165. This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve a high degree of plutonium incineration in oncethrough cycle. In this paper two fuel compositions were considered. In both compositions weapons' grade plutonium is used as fissile material. Spinel (MgAI204) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of ROXfuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastic change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption crosssection of plutonium as compared to uranium, (iv) 3elf is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is to identify a variant of the fuel composition and the reactor layout, which would neutralize the negative effect of the above-mentioned distinctive features.
04/01303
SP100 space reactor design
Demuth, S. F. Progress in Nuclear Energy, 2003, 42, (3), 323-359. The SP100 space nuclear reactor was designed for use as an orbital power supply, lunar or Martian surface power station, and power supply for nuclear electric propulsion, with a scaleable power range of tens of kWe to hundreds of kWe. The original mission was an orbital power supply for the United State's (US) Strategic Defense Initiative (SDI) of the 1980s. Although the original sponsors were a consortium of the US Department of Defense, US Department of Energy, and the US National Aeronautics and Space Administration (NASA), as the SDI effort diminished with the demise of the Soviet Union the mission evolved more toward the needs of NASA. Eventually, as the grandiose missions of NASA came into question in the early 1990s and less extravagant missions became more palatable, the SP100 programme was discontinued. At the time of programme discontinuation a complex infrastructure of industry and national laboratories had been established, a n d approximately US$1 billion invested in design and development. The SP100 was not intended as a low cost one-time-use device; but rather, a highly flexible power supply that realized a cost advantage by being capable of multiple missions, based on a common design with the flexibility to evolve. The design and development team made major progress successfully fabricating and testing technology features that were essential to meeting the stringent safety, performance, life-time and reliability requirements of proposed missions. The
04/01298 On the nuclear response of the water-cooled Pb17Li test blanket module for ITER-FEAT Chiovaro, P. et al. Fusion Engineering and Design, 2003, 69, (1-4), 469477. Within the European Fusion Technology Programme, the WaterCooled Lithium Lead (WCLL) D E M O breeding blanket line was selected in 1995 as one of the two EU lines to be developed in the next decades, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be tested in ITER-FEAT. The present paper is focused on the study of the WCLL-TBM nuclear response in ITERFEAT, being specifically oriented to the investigation of the local effects due to the typical C-shaped tubes of the breeder zone, since they could play a pivotal role in the module-relevant t h e r m o mechanical design. A 3D heterogeneous model of the WCLL-TBM, realistically simulating its new lay out and taking into account 9% Cr rnartensitic steel as reference structural material, has been set-up. A particular attention has been paid to the simulation of the characteristic 'C' shape of the breeder zone double walled tubes, which have been realistically reproduced. The WCLL-TBM model has been inserted into an existing ITER-FEAT 3D semi-heterogeneous model accounting for a proper D - T neutron source. Analyses have been performed by means of MCNP-4C code running On a cluster Of four workstations through the implementation of a parallel virtual machine. For each analysis a large number of histories (> 10 000 000) have been simulated, obtaining statistical uncertainties on the results lower than 3%. The main features of the WCLL-TBM nuclear response have been determined focusing the attention on power deposition density, material damage through DPA and He and H production rate, daily tritium production and tritium production rate radial distribution in the module. The obtained results are presented and critically discussed.
04/01299
On the use of silicon as thermal neutron filter
Adib, M. et al. Annals of Nuclear Energy, 2003, 30, (18), 1905-1917. A simple formula is given which allows to calculate the contribution of the total neutron cross-section including the Bragg scattering from different (hkl) planes to the neutron transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy. A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500 peV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given.
04/01300 Optical absorption and luminescence of 14-MeV neutron-irradiated CaF2 single crystals Cooke, D. W. and Bennett, B. L. Journal of Nuclear Materials', 2003, 321, (2-3), 158-164. The effects of 14-MeV neutron irradiation (1.1 x 1019 n/m 2) on crystalline CaF2 have been examined by optical absorption and luminescence techniques to evaluate its suitability as a window material for fusion energy applications. For comparison, similar studies were done on un-irradiated and X-irradiated samples. It is confirmed that pristine CaFa exhibits excellent optical transmission in the spectral region 200-1000 nm. X and neutron irradiation induce similar optical absorption spectra with maximum absorption coefficients approximately 1.6 and 0.8 cm -1, respectively. Thermally stimulated luminescence glow curves are induced by X-ray (11.55 kGy) and neutron exposures; peaks occur at 423, 534, 596 and 479, 550, 605 K, respectively. Thermal annealing experiments show that the major
178
Fuel and Energy Abstracts
May 2004
absorption peaks decay in concert with appearance of the first glow peak, which is attributed to an electron trap. Thus, the major absorption bands are associated with F and F-aggregate centres. The relative ease with which these centres are produced strongly suggests that CaF2 is not a good final optic window material for fusion energy applications.
04/01301 Reaction layer growth and reaction heat of U-Mo/ AI dispersion fuels using centrifugally atomized powders Ryu, H. J. et al. Journal of Nuclear Materials, 2003, 321, (2-3), 210220. The growth behavionr of reaction layers and heat generation during the reaction between U - M o powders and the A1 matrix in U-Mo/AI dispersion fuels were investigated. Annealing of 10 vol.% U-10Mo/AI dispersion fuels at temperatures from 500 to 550°C was carried out for 10 min to 36 h to measure the growth rate and the activation energy for the growth of reaction layers. The concentration profiles of reaction layers between the U-10Mo vs A1 diffusion couples were measured and the integrated inter-diffusion coefficients were calculated for the U and A1 in the reaction layers. Heat generation of U-Mo/AI dispersion fuels with 10-50 vol.% of U - M o fuel during the thermal cycle from room temperature to 700°C was measured employing the differential scanning calorimetry. Exotherrnic heat from the reaction between U Mo and the AI matrix is the largest when the volume fraction of U - M o fuel is about 30 vol.%. The un-reacted fraction in the U - M o powders increases as the volume fraction of U - M o fuel increases from 30 to 50 vol.%.
04/01302 Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels Bondarenko, A. V. et al. Journal of Nuclear Materials, 2003, 319, (1), 159-165. This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve a high degree of plutonium incineration in oncethrough cycle. In this paper two fuel compositions were considered. In both compositions weapons' grade plutonium is used as fissile material. Spinel (MgAI204) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of ROXfuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastic change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption crosssection of plutonium as compared to uranium, (iv) 3elf is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is to identify a variant of the fuel composition and the reactor layout, which would neutralize the negative effect of the above-mentioned distinctive features.
04/01303
SP100 space reactor design
Demuth, S. F. Progress in Nuclear Energy, 2003, 42, (3), 323-359. The SP100 space nuclear reactor was designed for use as an orbital power supply, lunar or Martian surface power station, and power supply for nuclear electric propulsion, with a scaleable power range of tens of kWe to hundreds of kWe. The original mission was an orbital power supply for the United State's (US) Strategic Defense Initiative (SDI) of the 1980s. Although the original sponsors were a consortium of the US Department of Defense, US Department of Energy, and the US National Aeronautics and Space Administration (NASA), as the SDI effort diminished with the demise of the Soviet Union the mission evolved more toward the needs of NASA. Eventually, as the grandiose missions of NASA came into question in the early 1990s and less extravagant missions became more palatable, the SP100 programme was discontinued. At the time of programme discontinuation a complex infrastructure of industry and national laboratories had been established, a n d approximately US$1 billion invested in design and development. The SP100 was not intended as a low cost one-time-use device; but rather, a highly flexible power supply that realized a cost advantage by being capable of multiple missions, based on a common design with the flexibility to evolve. The design and development team made major progress successfully fabricating and testing technology features that were essential to meeting the stringent safety, performance, life-time and reliability requirements of proposed missions. The