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Mark-­‐1 PB-­‐FHR Technical Description Technical Description of the “Mark 1” Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactor (PB-FHR) Power Plant Charalampos “Harry” Andreades Anselmo T. Cisneros Jae Keun Choi Alexandre Y.K. Chong Massimiliano Fratoni Sea Hong Lakshana R. Huddar Kathryn D. Huff David L. Krumwiede Michael R. Laufer Madicken Munk Raluca O. Scarlat Nicolas Zweibaum Ehud Greenspan Per F. Peterson UCBTH-14-002 September 30, 2014 Department of Nuclear Engineering University of California, Berkeley This research is being performed using funding received from the U.S. Department of Energy Office of Nuclear Energy’s Nuclear Energy University Programs. Technical Description of the Mark-1 PB-FHR Power Plant 1 | 153 Executive Summary This report describes the results of work at the University of California, Berkeley (UCB) to develop an initial pre-conceptual design for a small, modular 236-MWth pebble-bed fluoride- salt-cooled, high-temperature reactor (PB-FHR). This design study contributes to a larger U.S. Department of Energy Integrated Research Project (IRP) collaboration with the Massachusetts Institute of Technology and the University of Wisconsin, Madison to establish the technical basis to design, license, and commercially deploy FHRs. The Mark-1 (Mk1) PB-FHR design described here differs from previous FHR designs developed and published by UCB and others. It uses a nuclear air-Brayton combined cycle (NACC) based upon a modified General Electric 7FB gas turbine, designed to produce 100 MWe of base-load electricity when operated with only nuclear heat, and to increase this power output to 242 MWe using gas co-firing for peak electricity generation. Due to the high thermal efficiency of the NACC system, the steam-bottoming condenser of the Mk1 PB-FHR requires only 40% of the cooling water supply that is required for a conventional light water rector (LWR), for each MWh of base-load generation. As with conventional natural-gas combined cycle (NGCC) plants, this makes the efficiency penalty of using dry cooling with air-cooled condensers much smaller, enabling economic operation in regions where water is scarce. The primary purpose of the Mk1 design, with its co-firing capability, is to provide a new value proposition for nuclear power. The new value proposition for NACC arises from additional revenues earned by providing flexible grid support services to handle the ever- increasing demand for dispatchable peak power, in addition to traditional base-load electrical power generation. Because under base-load operation NACC power conversion has lower fuel costs than NGCC, and under peaking operation has higher efficiency in converting natural gas to electricity than NGCC, NACC plants will always dispatch before conventional NGCC plants. The reference configuration for the Mk1 site uses 12 Mk1 units, as shown in Fig. ES-1, capable of producing 1200 MWe of base load electricity, and ramping to a peak power output of 2900 MWe. The Mk1 design uses the same steel-plate composite wall modular construction methods as the Westinghouse AP1000, and its modular components can be manufactured in the same factories. A Mk1 reactor uses 10 structural modules, so the total number of structural modules needed to build a 12-unit station is quite similar to the ~120 structural modules used to build an AP1000 reactor. Estimated quantities of steel and concrete needed to construct a Mk1 station compare favorably, per MWe, with requirements for LWRs. The major difference between construction of a Mk1 station and an AP1000 is the highly repetitive construction tasks for the Mk1 station, arising from the construction of 12 identical units. PB-FHR fuel pebbles are 3.0 cm in diameter, smaller than golf balls (4.3 cm). Four Mk1 pebbles can provide electricity for a full year for an average U.S. household, which in 2011 consumed 11.3 MWe-hr. These four small pebbles are far less than the 8.1 tons of anthracite coal, or 17 tons of lignite coal, needed to produce the same amount of electricity using a coal power plant. Each Mk1 pebble contains 1.5 g of uranium encapsulated inside 4730 coated particles. This is only slightly more than the 0.9 g of uranium in 4150 particles in cylindrical, 1.25-cm diameter, 2.54-cm long Advanced Gas Reactor fuel compacts that have provided outstanding performance Technical Description of the Mark-1 PB-FHR Power Plant 2 | 153 in recent irradiation tests in the Idaho National Laboratory Advanced Test Reactor. Due to the high power density of the Mk1 core compared to modular helium-cooled reactors (MHRs), Mk1 pebbles reach full depletion in 1.4 years, compared to 2.5 years in a MHR and over 3 years for LWR fuel, so fuel testing and qualification can be performed more rapidly than for conventional LWR and MHR fuels. This fact has important programmatic implications, because it means that fuel testing and qualification, which would normally be time consuming, is unlikely to occupy the critical path for development of FHRs that use pebble fuel. 1) Mk1 reactor unit (typ. 12)! 2) Steam turbine bldg (typ. 3)! 3) Switchyard! 4) Natural gas master isolation! 5) Module assembly area" 5! 6) Concrete batch plant" 3! 7) Cooling towers (typ. 3)! 2! 4! 18! 1! 17! 6! 8! 9! 16! 10! 11! 12! 14! 13! 19! 7! 15! 20! 8) Dry cask storage! 9) Rad. waste bldg! 15) Main admin bldg! 10) Control room bldg! 16) Warehouse" 11) Fuel handling bldg! 17) Training" 12) Backup generation bldg" 18) Outage support bldg! 13) Hot/cold machine shops" 19) Vehicle inspection station! 14) Protected area entrance! 20) Visitor parking! Figure ES-1 Reference site arrangement for a 12-unit PB-FHR plant, capable of producing 1200 MWe base load and 2900 MWe peak. FHRs produce significantly more tritium than current LWRs, but less than current heavy water reactors. Analysis of the Mk1 design, with its well-defined core and heat exchanger surface areas, shows that graphite surfaces will be the most important sink to recover and remove this tritium. This conclusion generalizes to all FHRs, which will have much larger graphite surface areas than molten-salt-fueled reactors. The key goal for FHR tritium control is to limit tritium losses to heat exchangers. In the Mk1 design, tritium diffusion through heat exchangers is controlled by the use of an aluminized metal coating on the outside of heat exchanger tubes, which forms a impermeable aluminum oxide layer that self-heals via oxidation with the high- temperature air contacting the outside of the tubes. The Mk1 design implements several other innovative features. The Mk1 PB-FHR does not use an intermediate coolant loop and instead directly heats the power conversion fluid. It eliminates the conventional reactor guard vessel used in sodium fast reactors and instead uses a refractory reactor cavity liner system. All components for the Mk1 design are rail-transportable. As with previous FHR conceptual designs, the new Mk1 PB-FHR design presented here provides a basis to develop safety models, perform fuel cycle analyses, and assess the potential economics of FHRs compared to other reactor technologies. Technical Description of the Mark-1 PB-FHR Power Plant 3 | 153 Contents 1 Mark-1 PB-FHR Design Overview ...................................................................................... 12 1.1 Mk1 Reactor Design Overview .................................................................................................. 15 1.2 Mk1 Reactor Building Arrangement ......................................................................................... 16 1.3 Mk1 Reference 12-Unit Site Arrangement ................................................................................ 20 1.4 Mk1 Modular Design and Construction ................................................................................... 23 1.5 Mk1 Materials Quantities ........................................................................................................... 26 1.6 Mk1 Structural Materials Selection .......................................................................................... 29 1.6.1 Mk1 structural design approach ............................................................................................. 30 1.6.2 Mk1 structural material selection criteria .............................................................................. 31 1.7 Mk1 Fuel Development and Qualification ................................................................................ 32 1.8 Comparison with Other Reactors .............................................................................................. 33 1.9 Organization of Report ............................................................................................................... 35 2 Mark -1 Nuclear Heat Supply ............................................................................................... 36 2.1 Reactor Internals System ............................................................................................................ 36 2.1.1 Fuel and reflector pebbles ...................................................................................................... 36 2.1.2 Core design ............................................................................................................................ 39 2.1.3 Pebble injection subsystem .................................................................................................... 41 2.1.4 Graphite structures ................................................................................................................
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