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Fluoride-Salt Cooled High Temperature Reactors (FHR)

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Fluoride-Salt Cooled High Temperature Reactors (FHR) Fluoride-Salt Cooled High Temperature Reactors (FHR) Raluca O. Scarlat Nuclear Engineering, UW Madison [email protected] | HEATandMASS.ep.wisc.edu Engineering Physics Colloquium UW Madison 21 April 2015 FOR PUBLIC DISTRIBUTION Outline 1. Overview of FHR technology 2. Tritium management in the FHR 3. Tritium transport in the salt-graphite system 4. Salt freezing and overcooling transients 5. Future work Raluca Scarlat | HEATandMASS.ep.wisc.edu 2 FHRs: Solid Fuel, Salt Cooled Reactors Liquid fluoride salt coolants Coated particle fuel Excellent heat transfer (TRISO Fuel) Transparent, clean fluoride salt Boiling point ~1400ºC Reacts very slowly in air No energy source to pressurize containment But high freezing temperature (459oC) And industrial safety required for Be FHRs have uniquely large fuel thermal margin Raluca Scarlat | HEATandMASS.ep.wisc.edu 3 900 MWth 400 MWth PB-FHR PBMR Nuclear Air-Brayton Combined Cycle (NACC) Fluoride Salt Coolants Were Conventional combined cycle gas Developed plants achieve efficiencies of 50-60% for the Aircraft Nuclear Propulsion Program ! Aircraft Reactor Experiment, 1954 Nuclear Air-Brayton Combined Cycle (NACC) Modified GE 7FA Turbine for Co-fired NACC Base-load power: 42% efficiency (100 MWe) Gas co-firing: 66% efficiency (242 MWe) FHR Physical Plant Arrangement Coiled Tube Air Heaters (CTAHs) P.V. Gilli et al., “Radial Flow Heat Exchanger,” U.S. Patent Number 3712370, Filing date: Sept. 22, 1970, Issue date: 1973 PB-FHR Mark 1 Core Design Andreades, C., Cisneros, A. T., Choi, J. K., Chong, A. Y. K., Fratoni, M., Hong, S., … Peterson, P. F. (2014). Technical Description of the “Mark 1” Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactor (PB-FHR) Power Plant. Outline 1. Overview of FHR technology 2. Tritium management in the FHR 3. Tritium transport in the salt-graphite system 4. Future work Raluca Scarlat | HEATandMASS.ep.wisc.edu 10 Tritium Source Term ! The main sources of tritium production are Li-6 and Li-7 1 6 3 4 7 1 4 1 3 0 n+3Li→1H +2 He 3 Li+0 n→2 He+0 n+1H ! Natural Li isotopic composition: 7.5% 6Li – 92.5% 7Li ! Target 6Li start-up concentration: 60 ppm 6Li ! Steady state 6Li concentration: 8 ppm 6Li ! Lithium-6 is continuously created: 1 9 6 4 6 6 0 0 n+ 4 Be→2 He+ 2 He 2 He→3 Li+ −1e ! Natural Be isotopic composition: 100% 9Be Cross Sections for (n,α) Reactions 10000$ 1000$ 100$ 10$ 1$ Cross%Sec)on%(b)% 0.1$ 0.01$ 0.001$ 0.0001$ 0.00001$ 0.0001$ 0.001$ 0.01$ 0.1$ 1$ 10$ 100$ 1000$ 10000$ 100000$ 1000000$10000000$ Energy%(eV)% Tritium Source Term Estimated Tritum Production Rate in the Mk1 PB-FHR Core 14 0.18 0.16 12 0.14 10 0.12 8 0.10 6 0.08 0.06 EFPY) 4 0.04 T/EFPD) (mol 2 0.02 0 0.00 0 1 2 3 4 5 6 7 8 9 10 11 Tritium Production Rate (Ci/ Production Tritium Effective Full Power Years g/year per GWe Ci/year per GWe PB-FHR 176 1.7 x 106 PWR (average) 0.08 0.73 x 103 CANDU (Darlington) 576 8.0 x 106 PBMR 512 4.9 x 106 MSR 92 0.65 x 106 1. http://www.nrc.gov/reactors/operating/ops-experience/tritium/faqs.html#normal 2. Ohashi, Hirofumi and Sherman, Steven R. Tritium Movement and Accumulation in the NGNP System Interface and Hydrogen Plant. s.l. : Idaho National Laboratory, 2007. INL/EXT-07-12746 FHR Tritium Emission Target (Based on tritium production at steady state flibe isotopic composition, per GWe) ! FHR production: (1.7)106 Ci/yr (176 g/yr) ! If 100% released to air: 10-5 Ci/kg air (0.5x10-7 NRC limit) ! Reduce by a factor of 200 ! PWR emissions: (0.7)103 Ci/yr (0.075 g/yr) ! Reduce by a factor of 2000 Ohashi, Hirofumi and Sherman, Steven R. Tritium Movement and Accumulation in the NGNP System Interface and Hydrogen Plant. s.l. : Idaho National Laboratory, 2007. INL/EXT-07-12746 Tritium Sinks and Sources Raluca Scarlat | HEATandMASS.ep.wisc.edu 14 Alternative Tritium Sinks ! Fuel elements (pebbles: 100 days to full BU) ! Reflector pebbles ! Graphite particle absorber: annular cartridges of ~2.0-mm carbon spheres ! Gas sparging ! Double-wall heat exchangers Tritium Permeation Barriers for CTAH tubes Barrier Base Metal PRF SS316, MANET, 10 to Al O TZM, Ni, 2 3 >10,000 Hastalloy-X TiC, TiN, SS316, MANET, 3 to TiO2 TZM, Ti >10,000 Cr2O3 SS316 10 to 100 Si Steels 10 BN 304SS 100 N Fe 10 to 20 Er2O3 Steels 40 to 700 Barriers to release through CTAH • 4x lower mass convection coefficient than the pebble bed • Aluminum Oxide Tritium permeation barrier: Kanthal AF or Alkrothal 14; ~0.1 mm • Ensure low tritium concentration in the flibe at CTAH inlet: 10-12 to 10-23 wt fraction (10-5 to 10-17 g/m3) 1. Fluoride-Salt-Cooled High Temperature Reactor (FHR) Materials, Fuels and Components White Paper. UCBTH-12-003. Berkeley, CA. fhr.nuc.berkeley.edu SINAP Gas Sparging Studies Raluca Scarlat | HEATandMASS.ep.wisc.edu 17 Outline 1. Overview of FHR technology 2. Tritium management in the FHR 3. Tritium transport in the salt-graphite system 4. Future work Raluca Scarlat | HEATandMASS.ep.wisc.edu 18 Tritium Transport Scales of Interest Core Fuel Element Raluca Scarlat | HEATandMASS.ep.wisc.edu 19 Perfect Sink Calculations Table I. Perfect Sink Calculation Input and Outputs Inputs Inputs Parameter Value Units Parameter Value Units Salt Average Carbon Molecular 923.15 K 12 g/mol Temperature Weight Tritium Molecular Reynolds Number 500 - 3 g/mol Weight Diffusivity (Tritium in 3.90E-09 m2/s Graphite Density 1.70E+06 g/m3 FLiBe) Pebble Diameter 0.03 m Pebble Life 1.4 yr Salt Viscosity 6.78E-03 kg/(m*s) FLiBe Volume 46.82 m3 Salt Density 1.96E+03 kg/m3 Outputs Mass Transfer Salt Kinematic Viscosity 3.45E-06 m2/s 5.74E-05 m/s Coefficient Bulk Salt Schmidt Number (ν/D) 885.25 - Concentration Steady 2.38E-06 mol/m3 State Bulk Salt Steady State Production 2.66E-07 mol/s Concentration 1.15E-05 mol/m3 Rate Startup Graphite Saturation Startup Production Rate 1.28E-06 mol/s 0.27 g T/g C Steady State μ Graphite Saturation Pebble Surface Area 1945.27 m2 1.28 μg T/g C Startup Raluca Scarlat | HEATandMASS.ep.wisc.edu 20 Background Information: MSRE ! ORNL-4865: Graphite stringers (CGB) from MSRE were analyzed for tritium with depth profiling. Contains information on FP distributions and transport. ! ORNL-5011: Small graphite samples were exposed to T2 gas and analyzed. ! Briggs (1972): Measured and calculated distributions of tritium in the MSRE. ! Does this data lend itself to useful benchmark exercises? ! If so, what assumptions about the MSRE and its operations limit application to FHR? 21 What data is relevant? ! 750 C for 6.5 hr in P_T2 = 0.14 Pa on Strehlow, R. A. (1986). Chemisorption of tritium on graphites 1 cm^3 specimens. at elevated temperatures. Journal of Vacuum Science & Technology A: Vacuum, Surfaces, and Films, 4(1986), 1183. ! POCO AXF-5Q graphite taken as the doi:10.1116/1.573434 example (non-oxidized). Michael Young 22 Modeling of MSRE Depth Profile Trapping Model Raluca Scarlat | HEATandMASS.ep.wisc.edu 23 We need to understand the transport and trapping mechanisms ! Porous diffusion with adsorption occurring on the edge of crystallites near grain boundaries. ! Intra-granular diffusion ! Intracrystalline sorption ! Chemisorption on pore or grain surfaces Redmond, J. P., & Walker, P. L. (1960). Hydrogen sorption on graphite at elevated temperatures. The Journal of Physical Chemistry, 64(9), 1093–1099. Kanashenko, S. L., Gorodetsky, A. E., Chernikov, V. N., Markin, A. V., Zakharov, A. P., Doyle, B. L., & Wampler, W. R. (1996). Hydrogen adsorption on and solubility in graphites. Journal of Nuclear Materials, 233-237, 1207–1212. doi:10.1016/ j.jnucmat.2009.03.033 Atsumi, H., Tanabe, T., & Shikama, T. (2011). Hydrogen behavior in carbon and graphite before and after neutron irradiation – Trapping, diffusion and the simulation of bulk retention–. Journal of Nuclear Materials, 417(1-3), 633–636. doi:10.1016/ j.jnucmat.2010.12.100 Michael Young 24 Matrix Graphite and Nuclear Graphite Raluca Scarlat | HEATandMASS.ep.wisc.edu 25 Irradiation Effects Raluca Scarlat | HEATandMASS.ep.wisc.edu 26 MITR FHR Irradiations ! Two in-core irradiations completed: 300 and 1000 hours at 700°C ! Double-encapsulation with nickel inner vessel, graphite liner, titanium cold-wall ! 100-300g of MSRE secondary flibe ! Nuclear heating with ±3°C temperature control (He/Ne) ! Separate gas flows in each containment, tritium collected " Future irradiations for IRP-2 o Tritium uptake from flibe into irradiated graphite (reflector, fuel matrix, etc.) o Transport and release of gaseous tritium and activation products (from flibe and corrosion) o Tritium diffusion from salt through metals (heat exchangers) o Incorporate electrical heating for temperature control independent of reactor power 27 Summary: FHR Tritium Transport - Ongoing Work ! Goals: ! How much tritium will be retained in an FHR fuel pebble? ! How long will it take for the pebble to reach equilibrium? ! Studies: ! What is the mechanism of tritium transport and trapping in matrix graphite? ! What data from nuclear graphite is relevant? ! What data from higher temperature is relevant? ! Characterize irradiation effects ! What is the effect of salt intrusion? ! What are the desorption characteristics – important for fuel handing Michael Young 28 Outline 1. Overview of FHR technology 2. Tritium management in the FHR 3. Tritium transport in the salt-graphite system 4.
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