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01506 Creep surge effect under cyclic irradiation ofstructural materials for fusion reactors
05 Nuclear fuels (scientific, technical) mixed process of interracial-diffusion mechanism is proposed to be the rate determining step for ZrNb(l%)O(0.13%) oxidation in this environment.
05101506 Creep surge effect under cyclic irradiation of structural materials for fusion reactors Tsepelev, A. B. Fusion Engineering and Design, 2004, 70, (2), 79-85. Experimental investigations were performed on the radiation-creep behaviour of austenitic stainless steels, a-Fe and Fe-Cr alloys. The effect of an abrupt increase of creep rate early after either switching on and switching off irradiation (radiation-induced creep surge) was revealed. The effect of electron irradiation on the thermal-activation parameters of creep was examined. The radiation creep surge effect is discussed in terms of a dynamic preference of dislocations under nonstationary conditions typical of the formation of new radiationequilibrium flows of radiation induced point defects.
05101507 Deformation of zirconium irradiated by 4.4 MeV protons at 347 K Chow, C. K. et al. Journal of Hazardous Materials, 2004, 328, (1), 1-10. Proton irradiation tests have been performed on two cold-worked zirconium specimens using 4.4 MeV protons with controlled displacement damage rates between 1.4 and 6.9 x 10 7 dpa s -1. The specimens had different Fe contents, nominally pure (NP): 64-70 ppm by weight, and ultra-high purity (UP): 3-5 ppm by weight. A stress of 50 MPa was applied during the tests, and the test temperature was 347 K.
05/01508
Design integration of liquid surface divertors
Nygren, R. E. et aI. Fusion Engineering and Design, 2004, 72, (1-3), 223-244. The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. The authors have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected, Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. The authors have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium, sodium and beryllium fluorides, that has some potential because of its lower melting temperature. Sn and Sn-Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3D M H D flows cannot yet be fully modelled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied.
05•01509 Design of actively cooled flat toroidal limiter with CuCr heat sink for the HT-7 superconducting tokamak Hu, J. et al. Fusion Engineering and Design, 2005, 72, (4), 377-390. Long plasma discharges require in-vessel integration of actively cooled plasma facing components (PFC) in order to operate the HT-7 tokamak at the steady state. To promote power removing capacity, two flat toroidal limiters with high heat removal capability have been designed and installed in the HT-7 tokamak. The high performance toroidal limiters, 1.7 m 2 area, were located at top and bottom of the HT-7 vessel, covering 265 ° and 285 °, respectively, in the toroidal direction. The bolted structure for the graphite tile has been adopted, Employing flexible graphite sheet interlayer between the graphite tile and copper alloy heat sink has been demonstrated to be an effective way to improve heat transfer. The water cooling in the heat sink has been specially designed with large channel. The geometry of the array bars in the water cooling channel accelerates the turbulent flow of the cooling water and provided higher heat removing capacity. The experiences accumulated in the process of design of the HT-7 toroidal limiter will be helpful to design high thermal conductive plasma facing components with easy maintenance. Its heat removal capacity has been increased by a factor of 5 compared with the previous the HT-7 poloidal limiter with stainless steel heat sink, which were served as the main limiter.
05•01510 Development of plasma stored energy feedback control and its application to high performance discharges on JT-60U Oikawa, T. et al. Fusion Engineering and Design, 2004, 70, (2), 175-183.
The real-time feedback control of the plasma stored energy has been developed for control of the plasma MHD stability in the JT-60U tokamak. The plasma stored energy can be detected with high accuracy in real-time by a function parameterization method for various plasmas available in JT-60U, such as Ohmic plasmas, the L-mode, the H-mode, the high poloidal beta mode and the reversed shear mode over a wide range of the plasma parameters. By manipulating the neutral beam injection power, the plasma stored energy has been successfully controlled along the pre-programmed reference waveform. Especially in the reversed shear mode, this feedback control scheme has improved the reproducibility of the formation of the internal transport barrier, and MHD activities could be suppressed keeping the normalized beta in a stable region. A DT equivalent fusion amplification gain of 0.5 was sustained for 0.8 s in a reversed shear plasma by employing this feedback control scheme.
05•01511 Development of solidification techniques for radioactive sludge produced by a research reactor Plecas, I. B. et al. Progress in Nuclear Energy, 2004, 44, (1), 43-47. In the last 40 years in the 'Vinca' Institute, as a result of the operation of two research reactors, named R A and RB, and as a result of the application of radionuclides in medicine, industry and agriculture, radioactive waste materials of different levels of specific activity were generated. As a temporary solution, it is proposed that radioactive waste materials be stored in two interim storages. Radwaste materials that were immobilized in the inactive matrices are to be placed into concrete containers, for further manipulation and disposal. The present paper reports the results on the preliminary removal of sludge from the bottom of the spent fuel storage pool in the R A reactor, mechanical filtration of the pool water and sludge conditioning and storage.
05/01512 Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons Klix, A. et al. Fusion Engineering and Design, 2004, 70, (4), 279-287. At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate, which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different 6Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation.
05•01513 Drying characteristics of thorium fuel corrosion products Smith, R.-E. Journal of Nuelear Materials, 2004, 328, (2-3), 215-224. The open literature and accessible US Department of Energysponsored reports were reviewed for the dehydration and rehydration characteristics of potential corrosion products from thorium metal and thorium oxide nuclear fuels. Mixed oxides were not specifically examined unless data were given for performance of mixed thoriumuranium fuels. Thorium metal generally corrodes to thorium oxide. Physisorbed water is readily removed by heating to approximately 200°C. Complete removal of chemisorbed water requires heating above 1000°C. Thorium oxide adsorbs water well in excess of the amount needed to cover the oxide surface by chemisorption. The adsorption of water appears to be a surface phenomenon; it does not lead to bulk conversion of the solid oxide to the hydroxide. Adsorptive capacity depends on both the specific surface area and the porosity of the thorium oxide. Heat treatment by calcination or sintering reduces the adsorption capacity substantially from the thorium oxide produced by metal corrosion.
05•01514 Effect of prior thermal treatment on the microchemistry and crack propagation of proton-irradiated AISI 304 stainless steels Wang, L. H. et al. Journal of Hazardous Materials, 2004, 328, (l), 11 20.
Fuel and Energy Abstracts
July 2005 227
05101506 Creep surge effect under cyclic irradiation of structural materials for fusion reactors Tsepelev, A. B. Fusion Engineering and Design, 2004, 70, (2), 79-85. Experimental investigations were performed on the radiation-creep behaviour of austenitic stainless steels, a-Fe and Fe-Cr alloys. The effect of an abrupt increase of creep rate early after either switching on and switching off irradiation (radiation-induced creep surge) was revealed. The effect of electron irradiation on the thermal-activation parameters of creep was examined. The radiation creep surge effect is discussed in terms of a dynamic preference of dislocations under nonstationary conditions typical of the formation of new radiationequilibrium flows of radiation induced point defects.
05101507 Deformation of zirconium irradiated by 4.4 MeV protons at 347 K Chow, C. K. et al. Journal of Hazardous Materials, 2004, 328, (1), 1-10. Proton irradiation tests have been performed on two cold-worked zirconium specimens using 4.4 MeV protons with controlled displacement damage rates between 1.4 and 6.9 x 10 7 dpa s -1. The specimens had different Fe contents, nominally pure (NP): 64-70 ppm by weight, and ultra-high purity (UP): 3-5 ppm by weight. A stress of 50 MPa was applied during the tests, and the test temperature was 347 K.
05/01508
Design integration of liquid surface divertors
Nygren, R. E. et aI. Fusion Engineering and Design, 2004, 72, (1-3), 223-244. The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. The authors have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected, Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. The authors have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium, sodium and beryllium fluorides, that has some potential because of its lower melting temperature. Sn and Sn-Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3D M H D flows cannot yet be fully modelled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied.
05•01509 Design of actively cooled flat toroidal limiter with CuCr heat sink for the HT-7 superconducting tokamak Hu, J. et al. Fusion Engineering and Design, 2005, 72, (4), 377-390. Long plasma discharges require in-vessel integration of actively cooled plasma facing components (PFC) in order to operate the HT-7 tokamak at the steady state. To promote power removing capacity, two flat toroidal limiters with high heat removal capability have been designed and installed in the HT-7 tokamak. The high performance toroidal limiters, 1.7 m 2 area, were located at top and bottom of the HT-7 vessel, covering 265 ° and 285 °, respectively, in the toroidal direction. The bolted structure for the graphite tile has been adopted, Employing flexible graphite sheet interlayer between the graphite tile and copper alloy heat sink has been demonstrated to be an effective way to improve heat transfer. The water cooling in the heat sink has been specially designed with large channel. The geometry of the array bars in the water cooling channel accelerates the turbulent flow of the cooling water and provided higher heat removing capacity. The experiences accumulated in the process of design of the HT-7 toroidal limiter will be helpful to design high thermal conductive plasma facing components with easy maintenance. Its heat removal capacity has been increased by a factor of 5 compared with the previous the HT-7 poloidal limiter with stainless steel heat sink, which were served as the main limiter.
05•01510 Development of plasma stored energy feedback control and its application to high performance discharges on JT-60U Oikawa, T. et al. Fusion Engineering and Design, 2004, 70, (2), 175-183.
The real-time feedback control of the plasma stored energy has been developed for control of the plasma MHD stability in the JT-60U tokamak. The plasma stored energy can be detected with high accuracy in real-time by a function parameterization method for various plasmas available in JT-60U, such as Ohmic plasmas, the L-mode, the H-mode, the high poloidal beta mode and the reversed shear mode over a wide range of the plasma parameters. By manipulating the neutral beam injection power, the plasma stored energy has been successfully controlled along the pre-programmed reference waveform. Especially in the reversed shear mode, this feedback control scheme has improved the reproducibility of the formation of the internal transport barrier, and MHD activities could be suppressed keeping the normalized beta in a stable region. A DT equivalent fusion amplification gain of 0.5 was sustained for 0.8 s in a reversed shear plasma by employing this feedback control scheme.
05•01511 Development of solidification techniques for radioactive sludge produced by a research reactor Plecas, I. B. et al. Progress in Nuclear Energy, 2004, 44, (1), 43-47. In the last 40 years in the 'Vinca' Institute, as a result of the operation of two research reactors, named R A and RB, and as a result of the application of radionuclides in medicine, industry and agriculture, radioactive waste materials of different levels of specific activity were generated. As a temporary solution, it is proposed that radioactive waste materials be stored in two interim storages. Radwaste materials that were immobilized in the inactive matrices are to be placed into concrete containers, for further manipulation and disposal. The present paper reports the results on the preliminary removal of sludge from the bottom of the spent fuel storage pool in the R A reactor, mechanical filtration of the pool water and sludge conditioning and storage.
05/01512 Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons Klix, A. et al. Fusion Engineering and Design, 2004, 70, (4), 279-287. At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate, which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different 6Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation.
05•01513 Drying characteristics of thorium fuel corrosion products Smith, R.-E. Journal of Nuelear Materials, 2004, 328, (2-3), 215-224. The open literature and accessible US Department of Energysponsored reports were reviewed for the dehydration and rehydration characteristics of potential corrosion products from thorium metal and thorium oxide nuclear fuels. Mixed oxides were not specifically examined unless data were given for performance of mixed thoriumuranium fuels. Thorium metal generally corrodes to thorium oxide. Physisorbed water is readily removed by heating to approximately 200°C. Complete removal of chemisorbed water requires heating above 1000°C. Thorium oxide adsorbs water well in excess of the amount needed to cover the oxide surface by chemisorption. The adsorption of water appears to be a surface phenomenon; it does not lead to bulk conversion of the solid oxide to the hydroxide. Adsorptive capacity depends on both the specific surface area and the porosity of the thorium oxide. Heat treatment by calcination or sintering reduces the adsorption capacity substantially from the thorium oxide produced by metal corrosion.
05•01514 Effect of prior thermal treatment on the microchemistry and crack propagation of proton-irradiated AISI 304 stainless steels Wang, L. H. et al. Journal of Hazardous Materials, 2004, 328, (l), 11 20.
Fuel and Energy Abstracts
July 2005 227