The Experts below are selected from a list of 171 Experts worldwide ranked by ideXlab platform
Bilge Albayrak Ceper - One of the best experts on this subject based on the ideXlab platform.
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Study on spent fuel rejuvenation in PROMETHEUS Fusion Reactor
Energy Conversion and Management, 2006Co-Authors: Huseyin Yapici, Nesrin Demir, Gamze Genc, Bilge Albayrak CeperAbstract:Abstract This study presents the spent fuel rejuvenation potential of the PROMETHEUS-H Fusion Reactor. For this purpose, three different spent fuels were selected, i.e., (1) CANDU, (2) PWR-UO 2 and (3) PWR-MOX spent fuels. The spent fuel (volume fraction of 60%), spherically prepared and cladded with SiC (volume fraction of 10%), was located in the fuel zone (FZ) in the blanket of the modified PROMETHEUS-H Fusion Reactor. The FZ was cooled with high pressure helium gas (volume fraction of 30%) for the nuclear heat transfer. The neutronic calculations were performed by solving the Boltzmann transport equation with the help of the neutron transport code XSDRNPM-S/SCALE 4.3. The calculations of the time dependent atomic densities of the isotopes were performed for an operation period (OP) of up to 4 years with a 75% plant factor ( η ) under a first wall neutron load ( P ) of 4.7 MW/m 2 . The temporal variations of the atomic densities of the isotopes in the spent fuel composition and other physical parameters were calculated for a discrete time interval (Δ t ) of 1/12 year (one month) by using the interface program (code). In all investigated spent fuel cases, the tritium self sufficiency is maintained for the DT Fusion driver along the OP. The CANDU spent fuel becomes usable in a conventional CANDU Reactor after a regeneration time of ∼5.5 months. The CFFE value approaches 3.5% in the blanket fuelled with the PWR-UO 2 and PWR-MOX spent fuels after 41 and 35 months, respectively. The plutonium component can never reach a nuclear weapon grade quality during the spent fuel rejuvenation. Consequently, the modified PROMETHEUS-H Fusion Reactor has high neutronic performance for the rejuvenation of the spent fuels.
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Minor Actinide Transmutation Potential of Modified PROMETHEUS Fusion Reactor
Journal of Fusion Energy, 2004Co-Authors: Hüseyin Yapııcıı, Nesrin Demir, Gamze Genc, Bilge Albayrak CeperAbstract:This study presents the investigation of the burning and/or transmutation (B/T) of minor actinides (MAs) in the modified PROMETHEUS-H Fusion Reactor. The calculations were performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a neutron wall load (P) of 4.7 MW/m^2. In order to incinerate and transmute the MAs effectively, the transmutation zone (TZ) containing the mixture of MA nuclides, discharged from the pressured water Reactor (PWR)-MOX spent fuel, was located in the modified blanket of the PROMETHEUS-H Fusion Reactor. Two different blanket modifications were considered, (the Models A and B). The MA mixture was spherically prepared and cladded with SiC to prevent fission products from contaminating coolant and the MAs from contacting coolant. Helium was selected for the nuclear heat transfer in the TZ. The effect of the MA volume fraction in the TZ on the B/T was also investigated. The results bring out that the MAs are converted by the transmutation rather than the incineration. Both␣models can reduce significantly the effective half-lives of the MA nuclides by burning and/or transmuting these nuclides, and at the same time produce substantial electricity in␣situ .
Huseyin Yapici - One of the best experts on this subject based on the ideXlab platform.
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Study on spent fuel rejuvenation in PROMETHEUS Fusion Reactor
Energy Conversion and Management, 2006Co-Authors: Huseyin Yapici, Nesrin Demir, Gamze Genc, Bilge Albayrak CeperAbstract:Abstract This study presents the spent fuel rejuvenation potential of the PROMETHEUS-H Fusion Reactor. For this purpose, three different spent fuels were selected, i.e., (1) CANDU, (2) PWR-UO 2 and (3) PWR-MOX spent fuels. The spent fuel (volume fraction of 60%), spherically prepared and cladded with SiC (volume fraction of 10%), was located in the fuel zone (FZ) in the blanket of the modified PROMETHEUS-H Fusion Reactor. The FZ was cooled with high pressure helium gas (volume fraction of 30%) for the nuclear heat transfer. The neutronic calculations were performed by solving the Boltzmann transport equation with the help of the neutron transport code XSDRNPM-S/SCALE 4.3. The calculations of the time dependent atomic densities of the isotopes were performed for an operation period (OP) of up to 4 years with a 75% plant factor ( η ) under a first wall neutron load ( P ) of 4.7 MW/m 2 . The temporal variations of the atomic densities of the isotopes in the spent fuel composition and other physical parameters were calculated for a discrete time interval (Δ t ) of 1/12 year (one month) by using the interface program (code). In all investigated spent fuel cases, the tritium self sufficiency is maintained for the DT Fusion driver along the OP. The CANDU spent fuel becomes usable in a conventional CANDU Reactor after a regeneration time of ∼5.5 months. The CFFE value approaches 3.5% in the blanket fuelled with the PWR-UO 2 and PWR-MOX spent fuels after 41 and 35 months, respectively. The plutonium component can never reach a nuclear weapon grade quality during the spent fuel rejuvenation. Consequently, the modified PROMETHEUS-H Fusion Reactor has high neutronic performance for the rejuvenation of the spent fuels.
Nesrin Demir - One of the best experts on this subject based on the ideXlab platform.
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Study on spent fuel rejuvenation in PROMETHEUS Fusion Reactor
Energy Conversion and Management, 2006Co-Authors: Huseyin Yapici, Nesrin Demir, Gamze Genc, Bilge Albayrak CeperAbstract:Abstract This study presents the spent fuel rejuvenation potential of the PROMETHEUS-H Fusion Reactor. For this purpose, three different spent fuels were selected, i.e., (1) CANDU, (2) PWR-UO 2 and (3) PWR-MOX spent fuels. The spent fuel (volume fraction of 60%), spherically prepared and cladded with SiC (volume fraction of 10%), was located in the fuel zone (FZ) in the blanket of the modified PROMETHEUS-H Fusion Reactor. The FZ was cooled with high pressure helium gas (volume fraction of 30%) for the nuclear heat transfer. The neutronic calculations were performed by solving the Boltzmann transport equation with the help of the neutron transport code XSDRNPM-S/SCALE 4.3. The calculations of the time dependent atomic densities of the isotopes were performed for an operation period (OP) of up to 4 years with a 75% plant factor ( η ) under a first wall neutron load ( P ) of 4.7 MW/m 2 . The temporal variations of the atomic densities of the isotopes in the spent fuel composition and other physical parameters were calculated for a discrete time interval (Δ t ) of 1/12 year (one month) by using the interface program (code). In all investigated spent fuel cases, the tritium self sufficiency is maintained for the DT Fusion driver along the OP. The CANDU spent fuel becomes usable in a conventional CANDU Reactor after a regeneration time of ∼5.5 months. The CFFE value approaches 3.5% in the blanket fuelled with the PWR-UO 2 and PWR-MOX spent fuels after 41 and 35 months, respectively. The plutonium component can never reach a nuclear weapon grade quality during the spent fuel rejuvenation. Consequently, the modified PROMETHEUS-H Fusion Reactor has high neutronic performance for the rejuvenation of the spent fuels.
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Minor Actinide Transmutation Potential of Modified PROMETHEUS Fusion Reactor
Journal of Fusion Energy, 2004Co-Authors: Hüseyin Yapııcıı, Nesrin Demir, Gamze Genc, Bilge Albayrak CeperAbstract:This study presents the investigation of the burning and/or transmutation (B/T) of minor actinides (MAs) in the modified PROMETHEUS-H Fusion Reactor. The calculations were performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a neutron wall load (P) of 4.7 MW/m^2. In order to incinerate and transmute the MAs effectively, the transmutation zone (TZ) containing the mixture of MA nuclides, discharged from the pressured water Reactor (PWR)-MOX spent fuel, was located in the modified blanket of the PROMETHEUS-H Fusion Reactor. Two different blanket modifications were considered, (the Models A and B). The MA mixture was spherically prepared and cladded with SiC to prevent fission products from contaminating coolant and the MAs from contacting coolant. Helium was selected for the nuclear heat transfer in the TZ. The effect of the MA volume fraction in the TZ on the B/T was also investigated. The results bring out that the MAs are converted by the transmutation rather than the incineration. Both␣models can reduce significantly the effective half-lives of the MA nuclides by burning and/or transmuting these nuclides, and at the same time produce substantial electricity in␣situ .
Gamze Genc - One of the best experts on this subject based on the ideXlab platform.
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Study on spent fuel rejuvenation in PROMETHEUS Fusion Reactor
Energy Conversion and Management, 2006Co-Authors: Huseyin Yapici, Nesrin Demir, Gamze Genc, Bilge Albayrak CeperAbstract:Abstract This study presents the spent fuel rejuvenation potential of the PROMETHEUS-H Fusion Reactor. For this purpose, three different spent fuels were selected, i.e., (1) CANDU, (2) PWR-UO 2 and (3) PWR-MOX spent fuels. The spent fuel (volume fraction of 60%), spherically prepared and cladded with SiC (volume fraction of 10%), was located in the fuel zone (FZ) in the blanket of the modified PROMETHEUS-H Fusion Reactor. The FZ was cooled with high pressure helium gas (volume fraction of 30%) for the nuclear heat transfer. The neutronic calculations were performed by solving the Boltzmann transport equation with the help of the neutron transport code XSDRNPM-S/SCALE 4.3. The calculations of the time dependent atomic densities of the isotopes were performed for an operation period (OP) of up to 4 years with a 75% plant factor ( η ) under a first wall neutron load ( P ) of 4.7 MW/m 2 . The temporal variations of the atomic densities of the isotopes in the spent fuel composition and other physical parameters were calculated for a discrete time interval (Δ t ) of 1/12 year (one month) by using the interface program (code). In all investigated spent fuel cases, the tritium self sufficiency is maintained for the DT Fusion driver along the OP. The CANDU spent fuel becomes usable in a conventional CANDU Reactor after a regeneration time of ∼5.5 months. The CFFE value approaches 3.5% in the blanket fuelled with the PWR-UO 2 and PWR-MOX spent fuels after 41 and 35 months, respectively. The plutonium component can never reach a nuclear weapon grade quality during the spent fuel rejuvenation. Consequently, the modified PROMETHEUS-H Fusion Reactor has high neutronic performance for the rejuvenation of the spent fuels.
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Minor Actinide Transmutation Potential of Modified PROMETHEUS Fusion Reactor
Journal of Fusion Energy, 2004Co-Authors: Hüseyin Yapııcıı, Nesrin Demir, Gamze Genc, Bilge Albayrak CeperAbstract:This study presents the investigation of the burning and/or transmutation (B/T) of minor actinides (MAs) in the modified PROMETHEUS-H Fusion Reactor. The calculations were performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a neutron wall load (P) of 4.7 MW/m^2. In order to incinerate and transmute the MAs effectively, the transmutation zone (TZ) containing the mixture of MA nuclides, discharged from the pressured water Reactor (PWR)-MOX spent fuel, was located in the modified blanket of the PROMETHEUS-H Fusion Reactor. Two different blanket modifications were considered, (the Models A and B). The MA mixture was spherically prepared and cladded with SiC to prevent fission products from contaminating coolant and the MAs from contacting coolant. Helium was selected for the nuclear heat transfer in the TZ. The effect of the MA volume fraction in the TZ on the B/T was also investigated. The results bring out that the MAs are converted by the transmutation rather than the incineration. Both␣models can reduce significantly the effective half-lives of the MA nuclides by burning and/or transmuting these nuclides, and at the same time produce substantial electricity in␣situ .
Yasushi Seki - One of the best experts on this subject based on the ideXlab platform.
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Overview of the Japanese Fusion Reactor studies programme
Fusion Engineering and Design, 2000Co-Authors: Yasushi SekiAbstract:Fusion Reactor design studies conducted in Japan since the fifth IAEA Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology in 1993 are introduced. During the 5 years since the last IAEA-TCM, the Fusion Reactor studies in Japan became more focused concentrating on tokamaks, helical devices and laser Fusion. There are four tokamak Reactor studies, namely, A-SSTR, CREST, DREAM and IDLT. There are helical Reactor studies such as FFHR and one laser Fusion Reactor study on KOYO. Compared with the situation of 5 years ago having many types of confinement schemes being studied, the confinement schemes are now limited to three and each of the effort is larger involving greater number of scientists. The cooperation and exchanges between the design groups have increased through various committees, meetings and symposia.
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Impact of low activation materials on Fusion Reactor design
Journal of Nuclear Materials, 1998Co-Authors: Yasushi Seki, T. Tabara, Isao Aoki, Shuzo Ueda, Satoshi Nishio, R. KuriharaAbstract:The following impact of low activation materials to the Fusion Reactor design are described based on the design of five Fusion power Reactors with different structural material/coolant combinations. (1) Reduce the radioactive impact to the environment in case of severe accidents. (2) Reduce the radioactive impact to the environment during normal operation. (3) Reduce the decay heat during the maintenance and in case of loss of cooling accidents. (4) Reduce the gamma-ray dose during the maintenance. (5) Reduce the amount and lower the level of radioactive waste from replaced components and at the decommissioning of a Fusion Reactor. In order to reduce environmental impact in case of severe accidents to the level such as to enable construction of a Fusion Reactor near big cities, the low activation material must be of very low activity such as may only be achievable by SiC/SiC composites.
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Development of Dust Removal System for Fusion Reactor
Journal of Fusion Energy, 1997Co-Authors: Y. Oda, Yasushi Seki, Isao Aoki, Shuzo Ueda, T. Nakata, T. Yamamoto, R. KuriharaAbstract:In the future Fusion Reactor, dust control may become more important for safety, than for existing Fusion facilities. Some estimations show more than hundreds kg/yr dust will be generated from the plasma facing materials in the vacuum vessel. If we consider continuously operating plant, dust should be monitored and removed during the operation time. Optical monitoring methods and electrostatic removal methods are useful approaches to accomplish this. An investigation of the development of the dust removal system for a Fusion Reactor is reported in this paper.
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Fusion Reactor design studies
Fusion Engineering and Design, 1994Co-Authors: Yasushi SekiAbstract:Abstract This paper is a summary of Fusion Reactor design studies presented and discussed at the IAEA Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology, held in September 1993 at the University of California, Los Angeles. This summary does not represent all the Reactor design studies carried out in the world since the last meeting in Yalta in 1986. The Reactor designs presented are mostly D–T fuel cycle with both magnetic confinement and inertial confinement. As a D–3He, only the ARIES-III was presented. Steady state and pulsed tokamak studies, stellarator and reversed field pinch Reactor studies are summarized. Six recent D–T inertial confinement Fusion Reactors presented at the meeting are summarized. A short summary on D–3He Reactors and Fusion-fission hybrid Reactor designs is given.
