Nuclear heat and its use for Hydrogen production

Transcription

1 IEA/HIA TASK 25: HIGH TEMPERATURE HYDROGEN PRODUCTION PROCESS Nuclear heat and its use for Hydrogen production Generation I 1 In this period, because of the nonmaturity of the uranium enrichment process, these reactors had to work with natural uranium. Generation II Coming on stream in the eighties, these reactors (most of all the PWRs - Pressurized Water Reactor - and BWRs - Boiling Water Reactors) constitute the majority of the world reactor fleet in operation today. Generation III These evolutionary reactors are the industrial engineered versions of the second generation water reactors incorporating advanced safety specifications. Generation IV Generation IV will reach technical maturity by This generation of nuclear reactors will need to deliver the following advantages: Minimised waste and optimised natural resource usage Competitive economics Advanced nuclear safety Socially responsible answers to nuclear non-proliferation and physical protection issues 2 Generation IV systems will have the potential for the electricity and heat generation, hydrogen from water production and seawater desalination. Temperature of nuclear reactors Hydrogen production plant LFR/SFR/ SCWR MSR GFR VHTR 550 C 750 C 850 C 1000 C 500 C 800 C 1300 C Considered T min for High Temperature Production of H 2 Avg. T max for High Temperature Production of H 2 Theoretical safe T max obtained by nuclear technology US DOE NERI, 2006

2

3 IEA/HIA TASK 25: HIGH TEMPERATURE HYDROGEN PRODUCTION PROCESS GENERATION IV neutron spectrum Coolant T( C) Pressure Fuel Size range (MWe) GFR fast helium 850 high U LFR MSR fast fast or thermal lead or Pb-Bi fluoride salts low U low UF in salt 1000 SFR fast sodium 550 low U-238 & MOX Uses electricity & hydrogen electricity & hydrogen electricity & hydrogen electricity SCWR thermal or fast water very high UO electricity VHTR thermal helium high UO 2 prism or pebbles electricity & hydrogen 3 GENERATION IV SYSTEMS Sodium-cooled Fast Reactor (SFR) Description Sodium-cooled Fast Reactors are split in two different concepts: Liquid Metal Fast breeder Reactor (LMFBR), Integral Fast reactor (US project cancelled in 1994). The fuel cycle employs a full actinide recycling with two proposed variants: An intermediate-size ( MW e ) sodiumcooled reactor with uranium-plutonium-minoractinide-zirconium metal alloy fuel, supported by a fuel cycle based on pyrometallurgical reprocessing in facilities integrated with the reactor. A 500 to 1500 MW e sodium-cooled reactor with mixed uranium-plutonium oxide fuel, supported by a fuel cycle based upon advanced aqueous processing at a central location serving a number of reactors. Both of them can provide an outlet temperature of roughly C. 4,5 Moreover, nuclear waste longevity has been identified as a major concern, thus attention has been focused on the capability of SFRs to "burn" or to transmute the longest-lived components of the nuclear wastes, i.e. the trans-uranic elements. 6 Version 1 2

4

5 IEA/HIA TASK 25: HIGH TEMPERATURE HYDROGEN PRODUCTION PROCESS Lead-cooled Fast Reactor (LFR) Description The LFR system is similar to the SFR, where lead or the eutectic, lead-bismuth replaces sodium. A major difference is that the melting point of lead is higher than sodium (327 C vs 98 C) while that of lead-bismuth is only 125 C. On the one hand, the relatively high melting point of lead leads to the requirements of heating of all the coolant systems to prevent the coolant freezing. On the other hand, bismuth produces 210 Po which is a very active element; moreover the abundance of bismuth on earth is not sufficient to provide a wide-scale deployment of LFRs. 6 The fuel is composed of fertile uranium and transuranics, and is metal or nitride based. The plant can be large and monolithic with a factory manufactured battery of 1,200 MWe, or it could be a modular system with MWe,or it could be a small battery of MWe that would be more difficult to refuel. 7 The LFR is cooled by natural convection with a reactor outlet coolant temperature of 550 C, Super Critical Water Cooled Reactor (SCWR) possibly ranging up to 800 C with advanced materials. The high temperature would potentially enable the production of hydrogen by thermochemical processes. Description The Super Critical Water-cooled Reactor system works at high temperature and very high pressure, and is cooled by water above its thermodynamic critical point (T c = 374 C, P c = 22.1 MPa). The reference system is 1700 MWe where the SC water, at 510 C (up to 550 C) and 25 MPa directly drives the turbine, without any secondary steam system. 5 The fuel is UO 2. Passive safety features are incorporated similar to those of simplified boiling water reactors (SBRs). Moreover, SC water enables a thermal efficiency about one-third higher than current light-water reactors. 4,7 However, SC water is much more corrosive than water under LWR conditions, chemistry control and the suppression of radiolytic dissociation are also more difficult. 6 Further issues for development include corrosion and stress corrosion cracking (SCC), radiolysis as a function of temperature and fluid density, and water chemistry, dimensional and micro-structural stability and strength, embrittlement and creep resistance. The effects of neutrons, gamma radiation and impurities introduced into the primary system on water radiolysis needs to be studied. Water flow could affect the criticality safety of the system, since cold water would have a higher moderating ability possibly leading to a power surge. Version 1 3

6

7 IEA/HIA TASK 25: HIGH TEMPERATURE HYDROGEN PRODUCTION PROCESS GENERATION IV SYSTEMS Temperature 800 C Molten Salt Reactor (MSR) Description The MSR produces fission power in a circulating molten salt fuel. MSRs are fuelled with uranium or plutonium fluorides dissolved in Na and Zr fluorides as the primary option. The operating temperatures of MSRs range from the melting point of eutectic fluorine salts (about 450 C) to below the chemical compatibility temperature of nickel-based alloys (about 800 C). 4 Technology base MSRs were first developed in the late 1940s and 1950s for aircraft propulsion. The 8 MW th Molten Salt Reactor Experiment (MSRE) demonstrated attractive features. Outlet temperature The system operates at low pressure (< 0.5 MPa) and has a coolant outlet temperature above 700 C. Affording improved thermal efficiency 800 C is envisaged. 8 Gas cooled fast reactor (GFR) Description The GFR system features a fast-spectrum heliumcooled reactor and closed fuel cycle. Through the combination of a fast-neutron spectrum and full recycle of actinides, GFRs minimize the production of long-lived radioactive waste isotopes. The GFR fast spectrum also enables use of available fissile and fertile materials, including depleted Application for hydrogen production MSRs have the longer-term potential for thermochemical hydrogen production because of the heat delivered at high temperature and low pressure. Requirements are similar to those for the first MSR: the Aircraft Reactor Experiment (T out = 860 C). 9 uranium from enrichment plants. This technology is twice as efficient as thermal spectrum gas reactors with once-through fuel cycles. 4 Technology base General Atomics worked on the design in the 1970s (but not as fast reactors); none has so far been built. 10 An 80 MWt experimental demonstration GFR, AL- LEGRO, is planned by Euratom by A conceptual design of an entire GFR prototype system will be developed by GFR is estimated to be deployable by Outlet temperature The reference reactor is a 600-MW th /288-MW e, helium-cooled system operating with an outlet temperature of 850 C using a direct Brayton cycle gas turbine for high thermal efficiency. 1 Application for hydrogen production The GFR uses a direct-cycle helium turbine for electricity and can produce a secondary steam for thermochemical production of hydrogen. 11 Version 1 4

8

9 IEA/HIA TASK 25: HIGH TEMPERATURE HYDROGEN PRODUCTION PROCESS High Temperature Gas cooled Reactor (HTGR) Description HTGR technology has been developed over the last 50 years and so far seven plants have been designed, constructed and operated. The HTGR concept evolved from early air and CO 2 cooled reactors. 12 The use of helium as a coolant, along with a graphite moderator, offers enhanced neutronic and thermal efficiencies. There are currently two small-scale, operational HTGRs in the world, with five more, of a much larger scale, in development. Very High Temperature Reactors (VHTR), modern version of HTGR, are currently in the design and component testing stage, and aim at being operational around HTTR coupled with hydrogen production process 13 Reactor Location Power He T MWt In/Out ( o C) Fuel element Operation years Coupled HTPs HTR-10 China /950 Spherical Since 2000 HTTR Japan /950 Hexagonal Since 1998 SI HTR-PM China /750 Spherical In development PBMR SA / USA (Westinghouse) /950 Spherical In development HyS/HTE ANTARES France (AREVA NP) /850 Hexagonal In development SI/HTE GT-MHR USA (GA) / Russia /950 Hexagonal In development SI/HTE GTHTR300 Japan (JAEA) /850 Hexagonal In development SI HTGR plants operated and those in development 2 The VHTR is in fact a next step in the evolutionary development of HTGRs. The VHTR can generate electricity with theoretical high efficiency, over 50 % at 1000 C, compared with 47 % at 850 C in the GT-MHR or PBMR. Co-generation of heat and power makes the VHTR an attractive heat source for large industrial complexes. The VHTR can be deployed in refineries and petrochemical industries to substitute large amounts of process heat at different temperatures, as for steel, aluminium production. 4 Application for hydrogen production The HYTHEC and RAPHAEL European projects, specialised in the SI cycle and VHTR development, respectively, have worked in conjunction on the coupling between them. 14 A 600 MW th VHTR dedicated to hydrogen production could yield over 2 million normal cubic meters per day. The Korean Atomic Energy Research Institute (KAERI) earlier submitted a VHTR design to the Generation IV International Forum with a view to massive hydrogen production from it (300 MW th modules each producing 30,000 tonnes of hydrogen per year). 3 Version 1 5

10

11 IEA/HIA task 25: High Temperature Hydrogen Production Process Nuclear heat energy and its use for hydrogen production Contacts: Sabine POITOU, References [1] CEA website: [2] R. Elder and R. Allen, Nuclear heat for hydrogen production: Coupling a very high/high temperature reactor to a hydrogen production plant, Progress in Nuclear Energy, Volume 51, Issue 3, April 2009, Pages , [3] World nuclear website: [4] Overview of Generation IV Technology Roadmap: [5] Generation IV Nuclear Reactors July 2006, [6] A. Tim, S. Ion, Generation-IV nuclear power: A review of the state of the science, Energy Policy, Vol. 36, Issue 1 (2008) pp , [7] Generation IV International Forum LFR website: [8] A technology roadmap for Generation IV Nuclear Energy Systems 2002, [9] P. Pickard Dr. C. Forsberg Molten Salt Reactors (MSRs) Presentation for 2002 American Nuclear Society Winter Meeting Washington D.C. November 18, 2002, [10] The gas cooled fast reactor system, Hussein Khalil, [11] E. Bogusch, F. Carre, J.U. Knebel, K. Aoto, Synergies in the design and development of fusion and generation IV fission reactors Fusion Engineering and Design 83 (2008) , [12] M.P. LaBar, A.S. Shenoy, W.A. Simon, E.M. Campbell, The gas-turbine modular helium reactor. Nuclear Energy 43, , [13] Onuki, Y. Inagaki, R. Hino, and Y. Tachibana, R&D On Nuclear Hydrogen Production using HTGR at JAERIK Japan Atomic Energy Research Institute COE- Institute COE-INES International SymposiumTokyo, November 1, 2004, [14] D. de Lorenzo, J. Cedillo, J. Borgard, C. Corgnale, HYTHEC D6: I S Cycle VHTR reactor coupling with energy recovery. First Safety Assessment. FP6-HYTHEC-D6. Version 1 6

12