The Experts below are selected from a list of 312 Experts worldwide ranked by ideXlab platform
E Bilgen - One of the best experts on this subject based on the ideXlab platform.
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an improved process for h2so4 decomposition step of the sulfur iodine cycle
Energy Conversion and Management, 1995Co-Authors: Ilhan Ozturk, A Hammache, E BilgenAbstract:This study presents a new design, and thermodynamic and engineering analyses of the H2SO4 decomposition section (Section II of the GA sulfur-iodine process flow sheet) of the thermochemical hydrogen producing cycle. Thermodynamic (energy and exergy) and cost analyses have been carried out using thermodynamic data and costs in the literature. The results show that energetic and exergetic efficiencies are 76.0 and 75.6%, respectively, and the typical cost is 2.2S (1990) per kmol SO2 for 4S (1990)/GJ Nuclear Heat cost.
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A new process for oxygen generation step for the hydrogen producing sulphur-iodine thermochemical cycle
Chemical Engineering Research & Design, 1994Co-Authors: I. T. Öztürk, A Hammache, E BilgenAbstract:This study presents a new design, and thermodynamic and engineering analyses of the H 2 SO 4 decomposition section of the sulphur-iodine thermochemical cycle for producing hydrogen. Excess oxygen is used as an energy vector in various direct contact adiabatic equipment and shell and tube Heat exchangers are eliminated as much as possible. Thermodynamic (energy and exergy) and cost analyses have been carried out. The results show that energetic and exergetic efficiencies are 64.2% and 64.0% respectively and typical cost is 3.3$(1990) per kmol SO 2 for 4$US(1990)/GJ Nuclear Heat cost
Huseyin Yapici - One of the best experts on this subject based on the ideXlab platform.
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neutronic analysis of denaturing plutonium in a thorium fusion breeder and power flattening
Energy Conversion and Management, 2005Co-Authors: Huseyin Yapici, Mustafa BayrakAbstract:Abstract The purpose of this study is to denature Nuclear weapon grade quality plutonium in a thorium fusion breeder. Ten fuel rods containing the mixture of ThO 2 and PuO 2 are placed in a radial direction in the fissile zone where ThO 2 is mixed with variable amounts of PuO 2 to obtain a quasi-constant Nuclear Heat production density. The plutonium composition volume fractions in the fuel rods are gradually increased from 0.1% to 1% by 0.1% increments. The fissile fuel zone is cooled with four various coolants with a volume fraction ratio of 1 ( V coolant / V fuel = 1). These coolants are helium gas, flibe “Li 2 BeF 4 ”, natural lithium and eutectic lithium “Li 17 Pb 83 ”. Nuclear weapon grade quality 239 Pu in the fuel composition is denatured due to the accumulation of the 240 Pu isotope in the fissile zone after 18 months of plant operations. Under a first wall fusion neutron current load of 2.222 × 10 14 (14.1 MeV n/cm 2 s), which corresponds to 5 MW/m 2 , by a plant factor of 100%, at the end of the plant operation, the fissile fuel enrichment quality between 6.0% and 10% is obtained depending on the coolant types. During the plant operation, the tritium breeding ratio (TBR) should be at least 1.05. In the selected blanket, only the flibe coolant is already self sustaining at start up. The TBR increases steadily due to the higher neutron multiplication rate during the plant operation period. The highest TBR is obtained for the eutectic lithium coolant 1.4035, followed by the flibe coolant 1.3095, helium gas coolant 1.2172 and natural lithium coolant 1.0553 at the end of the operation period of 48 months. The energy multiplication factor M changed between 2.1731 and 6.6241 depending on coolant type during the operation period. The peak to average fission power density ratio Γ in the blanket decreases by ∼15%, which allows a more uniform power generation in the fissile zone. The isotopic percentage of 240 Pu reaches higher than 5% in all coolant types. This is very important for international safety.
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determination of the optimal plutonium fraction in transuranium discharged from pressured water reactor pwr spent fuel for a flat fission power generation in the force free helical reactor ffhr along the transmutation period
Annals of Nuclear Energy, 2003Co-Authors: Huseyin YapiciAbstract:Abstract Transmutation of transuranium (TRU) discharged from PWR spent fuel and the possibility of a flat fission power (FFP) generation along the transmutation process have been investigated in the force-free helical reactor (FFHR), which is a demo relevant helical-type D-T fusion reactor, for an operation period (OP) of up to 10 years by 75% plant factor (η) under a neutron wall load (P) of 1.5 MW/m 2 . For this purpose, the TRUs discharged from four different pressured water reactor (PWR) spent fuels have been selected as fuel, which include minor actinides (MAs) and plutonium isotopes, and in each spent fuel, the MA has been mixed with the plutonium by using various fractions. The MA–Pu mixture has been spherically prepared, and cladded with SiC to prevent the TRU nuclides from contacting coolant and the fission products from contaminating coolant. The mixture has been located in transmutation zone (TZ) of the FFHR to incinerate and/or transmute the TRU nuclides effectively, and helium has been used for the Nuclear Heat transfer in the TZ. The calculations have brought out that the fission power generation profiles are quasi-linear for the mixtures with small Pu fractions, in the range of 0–30%, along the transmutation period. The optimal Pu fraction in the mixture for the FFP generation has been determined individually for each mixture by applying interpolation method to the fission rate (R F ) values obtained for the relevant mixture with various the Pu fractions along the transmutation period. In consequence of the calculations, the optimal Pu fractions are obtained as 18.94, 19.90, 18.10 and 11.85% for the fuel types A, B, C and D, respectively. In the mixture types A, B, C and D with their optimal Pu fractions, the R F s are quasi-constant and about 0.84, 0.73, 0.94 and 0.79, respectively. At the end of operation period (EOP), the averages of effective half-lives of 237 Np, 239 Pu, 241 Pu, 241 Am and 243 Am can decrease to 9.8, 9.4, 9.9 and 9.6 years in the mixture types A, B, C and D, respectively. The average net transmutation fractions (TFs) for the whole TRUs and MAs are about 31 and 52%, respectively.
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study on transmutation of minor actinides discharged from high burn up pwr mox spent fuel in the force free helical reactor
Annals of Nuclear Energy, 2003Co-Authors: Huseyin YapiciAbstract:Abstract The force-free helical reactor (FFHR) is a demo relevant helical-type D-T fusion reactor. In this study, the burning and/or transmutation (B/T) of minor actinides (MAs) has been investigated for an operation period (OP) of up to 10 years in the FFHR by 75% plant factor (η) under a neutron wall load (P) of 1.5 MW/m2. In order to incinerate and transmute the MAs effectively, transmutation zone (TZ), containing the mixture of MA nuclides discharged from high burn-up pressured water reactor (PWR)-MOX spent fuel, has been located in the blanket of the FFHR. The MA mixture has been spherically prepared, and cladded with SiC to prevent fission products from contaminating coolant and the MAs from contacting coolant. Helium has been selected for the Nuclear Heat transfer in the TZ. Effect of the MA volume fraction in the zone on the B/T has been also investigated. The energy multiplication ratio (M), which is one of the main parameters in a fusion-fission hybrid reactor and relates to fission rate (RF), is quite high and increases from 5.8 to 9.4 in the case of a MA fraction of 10% and from 15.3 to 31.8 in that of 20% depending on the OP. The neutron multiplication coefficient (keff), also relating to the RF, is less than 0.9 in all investaged cases during the OP. Tritium breeding ratio (TBR) value is greater than 1.1 for all investigated cases so that tritium self-sufficiency is maintained for (D,T) fusion driver. Its value reaches to 3.2 in the case of a MA fraction of 20% at the end of operation period (EOP). The spatial non-uniformity of the fission energy density can be expressed with the help of the peak-to-average fission power density ratio (Γ). The Γ value varies in the range of 1.05–1.14 depending on the MA fraction and the OP. These values show that the TZ has a good flat fission power density profile in all investigated cases during the OP. At the beginning of operation period (BOP), the total transmutation rate (TR) values of 237Np, 241Am and 243Am are 1489 and 1206 kg/GWthyr in the cases of 10 and 20% MA, respectively. The TRs of 237Np, 241Am and 243Am decrease exponentially in all investigated cases during the OP. At the EOP, in the case of a MA fraction of 20%, the effective half-lives of 237Np, 241Am and 243Am decrease to 5.4, 5.1 and 6.8 years, repectively, and the net transmutation fractions (TFs) for the whole TRUs and MAs are obtained as 40 and 60%, respectively. These consequences bring out that the blanket carries out the B/T of MA effectively.
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proliferation hardening and power flattening of a thorium fusion breeder with triple mixed oxide fuel
Annals of Nuclear Energy, 2001Co-Authors: Sumer şahin, Veysel Ozceyhan, Huseyin YapiciAbstract:Abstract The proliferation hardening of the 233 U fuel in a thorium fusion breeder has been realised successfully with a homogenous mixture of ThO 2 , natural-UO 2 and CANDU spent Nuclear fuel in the form of a triple mixed oxide (TMOX) fuel. The new 233 U component will be successfully hardened against proliferation with the help of the 238 U component in the natural-UO 2 and spent fuel. The plutonium component remains non-prolific through the presence of the 240 Pu isotope in the spent CANDU fuel due to its high spontaneous fission rate. A (D,T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The fissile fuel zone, containing 10 fuel rod rows in the radial direction, covers the cylindrical fusion plasma chamber. A quasi-constant power density in the fissile zone has been achieved by reducing the ThO 2 component in the rods continuously in the radial direction (from 91 down to 64%). Three different coolants (pressurised helium, natural lithium and Li 17 Pb 83 eutectic) are selected for the Nuclear Heat transfer out of the fissile fuel breeding zone with a volume ratio of V coolant V fuel =1 in the fissile zone. The fissile fuel breeding occurs through the neutron capture reaction in the 232 Th (ThO 2 ), in the 238 U (natural-UO 2 and CANDU spent fuel) isotopes. The fusion breeder increases the Nuclear quality of the spent fuel, which can be defined with the help of the cumulative fissile fuel enrichment (CFFE) grade of the Nuclear fuel calculated as the sum of the isotopic ratios of all fissile materials ( 233 U+ 235 U+ 239 Pu+ 241 Pu) in the TMOX fuel. Under a first-wall fusion neutron current load of 10 14 (14.1 MeV n/cm 2 s), corresponding to 2.25 MW/m 2 and by a plant factor of 100%, the TMOX fuel can achieve an enrichment degree of ∼1% after ∼12–15 months. A longer irradiation period (∼ 30 months) increases the fissile fuel enrichment levels of the TMOX towards much higher degrees (∼ 2%), opening new possibilities for utilisation in advanced CANDU thorium breeders. The selected TMOX fuel remains non-prolific over the entire period for both uranium and plutonium components. This is an important factor with regard to international safeguarding.
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neutronic analysis of a thorium fusion breeder with enhanced protection against Nuclear weapon proliferation
Annals of Nuclear Energy, 1999Co-Authors: Sumer şahin, Huseyin YapiciAbstract:The fissile breeding capability of a (D,T) fusion-fission (hybrid) reactor fueled with thorium is analyzed to provide Nuclear fuel for light water reactors (LWRs). Three different fertile material compositions are investigated for fissile fuel breeding: (1) ThO2; (2) ThO2 denaturated with 10% natural-UO2 and (3) ThO2 denaturated with 10% LWR spent fuel. Two different coolants (pressurized helium and Flibe ‘Li2BeF4’) are selected for the Nuclear Heat transfer out of the fissile fuel breeding zone. Depending on the type of the coolant in the fission zone, fusion power plant operation periods between 30 and 48 months are evaluated to achieve a fissile fuel enrichment quality between 3 and 4%, under a first-wall fusion neutron energy load of 5 MW/m2 and a plant factor of 75%. Flibe coolant is superior to helium with regard to fissile fuel breeding. During a plant operation over four years, enrichment grades between 3.0 and 5.8% are calculated for different fertile fuel and coolant compositions. Fusion breeder with ThO2 produces weapon grade 233U. The denaturation of the 233U fuel is realized with a homogenous mixture of 90% ThO2 with 10% natural-UO2 as well as with 10% LWR spent Nuclear fuel. The homogenous mixture of 90% ThO2 with 10% natural-UO2 can successfully denaturate 233U with 238U. The uranium component of the mixture remains denaturated over the entire plant operation period of 48 months. However, at the early stages of plant operation, the generated plutonium component is of weapon grade quality. The plutonium component can be denaturated after a plant operation period of 24 and 30 months in Flibe cooled and helium cooled blankets, respectively. On the other hand, the homogenous mixture of 90% ThO2 with 10% LWR spent Nuclear fuel remains non-prolific over the entire period for both, uranium and plutonium components. This is an important factor with regard to international safeguarding.
Ehud Greenspan - One of the best experts on this subject based on the ideXlab platform.
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recycling independent core design for the enhs fuel self sustaining reactor
Nuclear Science and Engineering, 2009Co-Authors: Massimiliano Fratoni, Lanfranco Monti, Marco Sumini, Ehud GreenspanAbstract:- The feasibility of indefinite recycling in the Encapsulated Nuclear Heat Source (ENHS) core without changing the pitch-to-diameter (P/D) ratio, while maintaining a nearly zero burnup reactivity swing, is investigated. The P/D ratio required to achieve a nearly burnup-independent k eff over the life of the ENHS core was found sensitive to the initial composition of the transuranium (TRU) loaded and to the number of recycles this fuel underwent. The longer the cooling time is of the TRU from light water reactor (LWR) spent fuel, the larger the optimal P/D ratio becomes. Whereas the optimal P/D ratio of the reference ENHS core that is fueled with TRU from LWR spent fuel discharged at 50 GWd/t heavy metal (HM) and cooled for 10 yr is 1.36, it is 1.54 for the equilibrium core that features a substantially smaller concentration of 241 Pu as well as of 242 Pu, a larger concentration of 239 Pu, and a substantially larger concentration of minor actinides. It was found that by increasing the cooling period of the above LWR TRU to ∼32 yr, the optimal first core P/D ratio is that of the equilibrium core. The burnup reactivity swing of the subsequent cores fueled with successive recycling of the ENHS discharged HM is satisfactory. There is no need to adjust the core P/D ratio from recycle to recycle. The power level that can be removed by natural circulation from the P/D = 1.54 core is ∼36% higher than that of the reference ENHS core. The physical phenomena affecting the observed trends are discussed, and the neutronic characteristics of the equilibrium cores identified are summarized.
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the encapsulated Nuclear Heat source enhs reactor core design
Nuclear Technology, 2005Co-Authors: Ser Gi Hong, Ehud Greenspan, Yeong Il KimAbstract:A once-for-life, uniform composition, blanket-free and fuel-shuffling-free reference core has been designed for the Encapsulated Nuclear Heat Source (ENHS) to provide the design goals of a nearly zero burnup reactivity swing throughout {approx}20 yr of full-power operation up to the peak discharge burnup of more than 100 GWd/t HM. What limits the core life is radiation damage to the HT-9 structural material. The temperature coefficients of reactivity are all negative, except for that of the coolant expansion. However, the negative reactivity coefficient associated with the radial expansion of the core structure can compensate for the coolant thermal expansion. The void coefficient is positive but of no safety concern because the boiling temperature of lead or lead-bismuth is so high that there is no conceivable mechanism for the introduction of significant void fraction into the core. The core reactivity coefficients, reactivity worth, and power distributions are almost constant throughout the core life.It was found possible to design such once-for-life cores using different qualities of Pu and transuranics as long as U is used as the primary fertile material. It is also feasible to design ENHS cores using nitride rather than metallic fuel. Relative to the reference metallic fuel core, nitride fuelmore » cores offer up to {approx}25% higher discharge burnup and longer life, up to {approx}38% more energy per core, a significantly more negative Doppler reactivity coefficient, and less positive coolant expansion and coolant void reactivity coefficient but a somewhat smaller negative fuel expansion reactivity coefficient. The pitch-to-diameter ratio (1.45 of the nitride fuel cores using enriched N) is larger than that (1.36) for the reference metallic fuel core, implying a reduction of the coolant friction loss, thus enabling an increase in the power level that can be removed from the core by natural circulation cooling.It is also possible to design Pu-U(10Zr) fueled ENHS-type cores using Na as the primary coolant with either Na or Pb-Bi secondary coolants. The Na-cooled cores feature a tighter lattice and are therefore more compact but have spikier power distribution, more positive coolant temperature reactivity coefficients, and smaller reactivity worth of the control elements.« less
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power flattening options for the enhs encapsulated Nuclear Heat source core
Progress in Nuclear Energy, 2005Co-Authors: Ser Gi Hong, Ehud GreenspanAbstract:Abstract The feasibility of power flattening while maintaining a nearly constant keff over the core life is assessed for the Encapsulated Nuclear Heat Source (ENHS). A couple of approaches are considered — using different fuel dimensions and using different enrichment levels across the core. Three new cores with flattened power distribution are successfully designed: Design-I uses different fuel rod diameters but uniform fuel composition; Design-II uses different fuel enrichment in the radial direction but uniform fuel rod dimensions; Design-III is similar to Design-II but uses enrichment splitting also in the axial direction. Relative to the reference ENHS core, the BOL peak-to-average channel power ratio is reduced from 1.50 to 1.15, 1.22 and 1.15 and the average discharge burnup increases by 8.5%, 27.9% and 41.2% for, respectively, Design-I, -II and -III. The corresponding burnup reactivity swings over 20 years of full power operation are 0.37%, 0.52% and 0.60% relative to 0.22% of the reference design. Design-II and -III have a negative coolant expansion reactivity defect while in the reference design this defect is positive. The radial power flattening increases the reactivity worth of the peripheral absorbers of the three new designs while the central absorber reactivity worth is reduced but their sum is nearly maintained. The newly designed cores have slightly more positive coolant void reactivity worth than the reference ENHS core.
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the long life core encapsulated Nuclear Heat source enhs generation iv reactor
2002Co-Authors: Ehud Greenspan, E Feldman, A Barak, D Saphier, Neil W Brown, Q Hossain, J Buongiorno, Lawrence E Conway, Milorad B Dzodzo, James J SienickiAbstract:The long-life core for the Encapsulated Nuclear Heat Source (ENHS) reactor has been redesigned so as to provide for fuel rod clad integrity up to the discharge burnup design goal. It was found feasible to design a nearly zero burnup reactivity swing long-life core that will maintain the fuel rod integrity up to the peak discharge burnup while enabling to handle the rated power using natural circulation. The core life is limited by radiation damage to its structural material. The core power shape is exceptionally constant throughout the core life. The new reference core design can deliver 125 MW{sub th} while having very generous margins for maximum acceptable temperatures or temperature differences. Using a cover-gas lift-pump it may be possible to design an ENHS module to deliver {approx}50% more power than the set goal. Briefly reviewed are unique features of the ENHS reactor along with the potential of this reactor to meet the goals set for Generation IV reactors. (authors)
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enhs the encapsulated Nuclear Heat source a Nuclear energy concept for emerging worldwide energy markets
10th International Conference on Nuclear Engineering (ICONE 10) Arlington VA (US) 04 14 2002--04 18 2002, 2002Co-Authors: D C Wade, Ehud Greenspan, E Feldman, James J Sienicki, Tanju Sofu, A Barak, D Saphier, Neil W Brown, Q Hossain, Mario D CarelliAbstract:A market analysis is presented which delineates client needs and potential market size for small turnkey Nuclear power plants with full fuel cycle services. The features of the Encapsulated Nuclear Heat Source (ENHS) which is targeted for this market are listed, and the status of evaluation of technological viability is summarized.
Charles Forsberg - One of the best experts on this subject based on the ideXlab platform.
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sustainability by combining Nuclear fossil and renewable energy sources
Progress in Nuclear Energy, 2009Co-Authors: Charles ForsbergAbstract:Abstract The energy industries face two sustainability challenges: the need to avoid climate change and the need to replace traditional crude oil as the basis of our transport system. Radical changes in our energy system will be required to meet these challenges. These challenges may require tight coupling of different energy sources (Nuclear, fossil, and renewable) to produce liquid fuels for transportation, match electricity production to electricity demand, and meet other energy needs. This implies a paradigm shift in which different energy sources are integrated together, rather than being considered separate entities that compete. Several examples of combined-energy systems are described. High-temperature Nuclear Heat may increase worldwide light crude oil resources by an order of magnitude while reducing greenhouse gas releases from the production of liquid fossil fuels. Nuclear–biomass liquid-fuels production systems could potentially meet world needs for liquid transport fuels. Nuclear–hydrogen peak power systems may enable renewable electricity sources to meet much of the world's electric demand by providing electricity when the wind does not blow and the sun does not shine.
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An Air-Brayton Nuclear-Hydrogen Combined-Cycle Peak- and Base-Load Electric Plant
Volume 6: Energy Systems: Analysis Thermodynamics and Sustainability, 2007Co-Authors: Charles ForsbergAbstract:A combined-cycle power plant is proposed that uses Heat from a high-temperature Nuclear reactor and hydrogen produced by the high-temperature reactor to meet base-load and peak-load electrical demands. For base-load electricity production, air is compressed; flows through a Heat exchanger, where it is Heated to between 700 and 900°C; and exits through a high-temperature gas turbine to produce electricity. The Heat, via an intermediate Heat-transport loop, is provided by a high-temperature reactor. The hot exhaust from the Brayton-cycle turbine is then fed to a Heat recovery steam generator that provides steam to a steam turbine for added electrical power production. To meet peak electricity demand, after Nuclear Heating of the compressed air, hydrogen is injected into the combustion chamber, combusts, and Heats the air to 1300°C - the operating conditions for a standard natural-gas-fired combined-cycle plant. This process increases the plant efficiency and power output. Hydrogen is produced at night by electrolysis or other methods using energy from the Nuclear reactor and is stored until needed. Therefore, the electricity output to the electric grid can vary from zero (i.e., when hydrogen is being produced) to the maximum peak power while the Nuclear reactor operates at constant load. Because Nuclear Heat raises air temperatures above the auto-ignition temperatures of the hydrogen and powers the air compressor, the power output can be varied rapidly (compared with the capabilities of fossil-fired turbines) to meet spinning reserve requirements and stabilize the grid. Copyright © 2007 by ASME.
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advanced csic composites for high temperature Nuclear Heat transport with helium molten salts and sulphur iodine thermochemical hydrogen process fluids
Nuclear Science, 2004Co-Authors: Charles Forsberg, Per F Peterson, Paul S PickardAbstract:This paper discusses the use of liquid-silicon-impregnated (LSI) carbon-carbon composites for the development of compact and inexpensive Heat exchangers, piping, vessels and pumps capable of operating in the temperature range of 800 to 1100°C with high-pressure helium, molten fluoride salts, and process fluids for sulfur-iodine thermochemical hydrogen production. LSI composites have several potentially attractive features, including ability to maintain nearly full mechanical strength to temperatures approaching 1400°C, inexpensive and commercially available fabrication materials, and the capability for simple forming, machining and joining of carbon-carbon performs, which permits the fabrication of highly complex component geometries. In the near term, these materials may prove to be attractive for use with a molten-salt intermediate loop for the demonstration of hydrogen production with a gas-cooled high temperature reactor. In the longer term, these materials could be attractive for use with the moltensalt cooled Advanced High Temperature Reactor, molten salt reactors, and fusion power plants.
Pascal Anzieu - One of the best experts on this subject based on the ideXlab platform.
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a general survey of the potential and the main issues associated with the sulfur iodine thermochemical cycle for hydrogen production using Nuclear Heat
Progress in Nuclear Energy, 2008Co-Authors: X Vitart, Philippe Carles, Pascal AnzieuAbstract:Abstract The thermochemical sulfur–iodine cycle is studied by CEA with the objective of massive hydrogen production using Nuclear Heat at high temperature. The challenge is to acquire by the end of 2008 the necessary decision elements, based on a scientific and validated approach, to choose the most promising way to produce hydrogen using a generation IV Nuclear reactor. Amongst the thermochemical cycles, the sulfur–iodine process remains a very promising solution in matter of efficiency and cost, versus its main competitor, conventional electrolysis. The sulfur–iodine cycle is a very versatile process, which allows lot of variants for each section which can be adjusted in synergy in order to optimise the whole process. The main part of CEA's program is devoted to the study of the basic processes: new thermodynamics data acquisition, optimisation of water and iodine quantity, optimisation of temperature and pressure in each unit of the flow-sheet and survey of innovative solutions (membrane separations for instance). This program also includes optimisation of a detailed flow-sheet and studies for a hydrogen production plant (design, scale, first evaluations of safety issues and technico-economic questions). This program interacts strongly with other teams, in the framework of international collaborations (Europe, USA for instance).
