Fast Reactor Operating Experience in the U.S.

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1 Fast Reactor Operating Experience in the U.S. Harold F. McFarlane Deputy Associate Laboratory Director for Nuclear Science and Technology 3 March 2010 [insert optional photo(s) here]

2 Thanks to John Sackett and Jim Cahalan for providing the material for this summary Fast Reactors Database, IAEA International Working Group on Fast Reactors, see also IAEA-TECDOC-866. R. Loftness, Nuclear Power Plants, Van Nostrand Co., L. J. Koch, Experimental Breeder Reactor-II, An Integrated Experimental Fast Reactor Nuclear Power Station, Argonne National Laboratory. (Available on-line at G. Billuris, et al., SEFOR Plant Design, Fast Reactors National Topical Meeting, American Nuclear Society, San Francisco, April, A Summary Description of the Fast Flux Test Facility, HEDL-400, Hanford Engineering Development Laboratory, December, S. H. Fistedis, Ed., The Experimental Breeder Reactor-II Inherent Safety Demonstration, Nucl. Eng. Des., Vol. 101, No. 1,

3 US Fast Reactors Built and Operated Facility Location Mission Dates Power (MWt) EBR-I Idaho R&D EBR-II Idaho R&D Fermi-1 Michigan Power SEFOR Arkansas Safety Test FFTF Washington Fuel & Material Test Experimental Breeder Reactor-I (EBR-I) was built to demonstrate fuel breeding Experimental Breeder Reactor-II (EBR-II) was built to demonstrate closure of the metallic fuel cycle and recycling of reactor fuel Fermi-1 was built as a metallic uranium-fueled reactor on a utility grid Southwest Experimental Fast Oxide Reactor (SEFOR) was built to demonstrate the safety properties of the Doppler feedback for oxide fuel Fast Flux Test Facility (FFTF) was built to test fuel and cladding materials for the Liquid Metal Fast Breeder Reactor (LMFBR) program 3

4 EBR-I produced electricity 21 December 1951

5 Experimental Breeder Reactor-1 (EBR-I) From 1951 to 1963, EBR-I was operated with four core designs to demonstrate breeding and to develop an understanding of liquid metal fast reactor performance Pu breeding demonstrated by February, 1952 Testing platform for reactor physics, fluid dynamics, and power generation. Generated 200 kw of electricity for NRTS NaK cooled, U-235 (94% enriched) metal fuel in stainless steel cladding 217 pin locations, in. OD, 4.25 in. core height 227 o C inlet, 316 o C outlet, 20 psig Mark III core used hexagonal tubes and wire wraps for fuel pin positioning Mark IV core used Pu fuel (1962) In November, 1955, during a test to investigate a prompt positive component of the power coefficient, an unanticipated power excursion resulted in fuel melting. Core was replaced and operation continued through

6 Assembly of EBR-I Core Inner Tank 6

7 Mark III Inner Tank Assembly Outer blanket 7

8 Experimental Breeder Reactor-II (EBR-II) From 1964 through 1994, EBR-II operated as a prototype breeder power station demonstrating fuel cycle closure Dry critical September 1961; wet critical November MWt with 3 MWe to INEL in August, 1964; 50 MWt August 1965; 62.5 MWt September 1969 Sodium cooled, 371 o C inlet, 473 o C outlet, 47 psig Fuel pins 0.17 in. OD, 13.5 in. core height; metal fuel in SS cladding First fuel processed in Fuel Cycle Facility in September 1964; recycled fuel irradiation in April 1965 Mission oriented to irradiation testing in 1969; supporting FFTF and CRBRP oxide fuel testing and development Integral Fast Reactor (IFR) program began in mid 1980 s Testing and demonstration of high burnup metallic fuels Shutdown Heat Removal Test series ; natural circulation decay heat removal and passive shutdown in ATWS events (unprotected loss-of-flow and loss-of-heat-sink) Operated through

9 EBR-II Site 9

10 EBR-II pool layout

11 Fermi-I 200 MWt power station located on the western shore of Lake Erie south of Detroit Designed by Atomic Power Development Associates (APDA) and constructed by Power Reactor Development Co. (PRDC) for Detroit Edison Critical August 1963, first power August 1966 Sodium cooled, 288 o C inlet, 427 o C outlet, 120 psia Metal fuel, Zr cladding in. OD, 31 in. height, square pin pitch Subassembly flow blockage and fuel melting accident during power ascension on October 5, 1966 Metal fuel core removed and replaced with oxide core; full power 1969 Operation ceased in

12 Fermi-I site

13 Fermi-I primary system

14 Southwest Experimental Fast Oxide Reactor (SEFOR) Southwest Experimental Fast Oxide Reactor (SEFOR) was designed and built by General Electric for the USAEC, with participation by EURATOM, Kernforschungszentrum Karlsruhe, and seventeen US utilities 20 MWt sodium cooled transient test reactor Designed for power oscillations, sub-prompt critical excursions, and super-prompt critical excursions to demonstrate the safety characteristics of the Doppler coefficient for oxide fuel Located 20 miles south of Fayetteville, Arkansas. Built in Began operation in Mission completed in fuel assemblies, each with six fuel rods and a spacer rod Fuel rods 0.97 in. OD, core height 36 9/16 in. 20% PuO 2 -UO 2 fuel, clad with SS Axially segmented core to reduce transient axial expansion Coolant inlet 371 o C, outlet 438 o C Ex-vessel moveable reflector for reactivity control 14

15 Construction of SEFOR

16 Fast Flux Test Facility (FFTF) FFTF was built as a fuels and materials test facility for the US breeder reactor program, principally supporting oxide fuel development for the Clinch River Breeder Reactor Plant (CRBRP) 400 MWt, sodium cooled, 360 o C inlet, 527 o C outlet, 133 psig nominal MOX fueled, SS cladding, 0.23 in. OD pin: 217 pin fuel assembly First critical February 1980, full power December 1980, shutdown December 1993 Provided verification of the CRBRP fuel design Test bed for Investigation of advanced, low-swelling fuel cladding materials Verification of large-scale component designs (pumps, heat exchangers) Mission extension: safety testing Natural circulation shutdown heat removal Passive power reduction in unprotected loss-of-flow sequence 16

17 FFTF Site Hanford, Washington 17

18 FFTF Containment Building View 18

19 Thermal/Hydraulic Safety Demonstrations Sodium fast reactor thermal/hydraulic safety margins were demonstrated in two test programs conducted at EBR-II and FFTF The EBR-II Shutdown Heat Removal Test (SHRT) series was conducted in Fifty-eight tests of reactor and balance-of-plant transients Focus on natural circulation cooling margins Sequence initiators ranged from normal shutdown to total loss of forced coolant flow at full power without reactor scram Test results showed that EBR-II, and plants with similar design features, are capable of inherent self-protection The FFTF Inherent Safety Test series was conducted in 1986 Reactivity feedback characteristics were measured in static tests, followed by loss of forced flow test from 10% to 50% power Focus on inherent reactor power reduction due to loss of forced flow Test results verified that FFTF design features provided inherent self-protection 19

20 Liquid Metal Fast Reactor Experience Liquid sodium was initially selected as the coolant for fast reactors on the basis of 1) low pumping power requirement, 2) high volumetric heat capacity, and 3) high thermal conductivity. Other favorable properties include low neutron absorption and very good compatibility (negligible corrosion) with structural materials Many reactor years (~300) of experience in the US and internationally have demonstrated the effectiveness of the liquid metal fast reactor concept for electricity generation and fuel production, as originally conceived by Fermi, Szilard, and Wigner The US reactor development program has also demonstrated that liquid metal cooling contributes to excellent inherent safety reactor performance, through Natural circulation decay heat removal Passive reactor power reduction in beyond-design-basis 20

21 U.S. and International Experience in Fast Reactor Operation Has Shown Many Benefits Fast reactor fuel is reliable and safe, whether metal or oxide. Cladding failure does not lead to progressive fuel failure. High burnup of fast reactor fuel is achievable, whether metal or oxide. Sodium is not corrosive to stainless steel or components within it. Leakage in steam generating systems with resultant sodium-water reactions does not lead to serious safety problems. Such reactions are not catastrophic, as previously believed, and can be detected, contained and isolated. Leakage of high-temperature sodium coolant, leading to a sodium fire, is not catastrophic and can be contained, suppressed and extinguished. There have been no injuries form sodium leakage and fire. Fast reactors can be self-protecting against Anticipated Transients Without Scram.

22 Lessons From Fast Reactor Operation (contd.) Passive transition to natural convective core-cooling and passive rejection of decay heat has been demonstrated. Reliable control and safety-system response has been demonstrated. Low radiation exposures are the norm for operating and plant maintenance personnel. Emissions are quite low from sodium cooled fast reactors, in part because sodium reacts chemically with many fission products if fuel cladding is breached. Maintenance and repair techniques are well developed and straightforward.

23 Lessons From Fast Reactor Operation (contd.) Several aspects of design and operation have highlighted challenges in fast reactor designs Thermal shock to structures is a major design challenge. Problems with handling fuel in sodium systems have occurred. Failure of in-sodium components without adequate means for removal and repair has resulted in costly and time-consuming repair. Reactivity anomalies have occurred in a number of fast reactors, requiring careful attention to core restraint system design. Operational problems have been encountered at the sodium/covergas interface, resulting from formation of sodium-oxide deposits and can lead to binding of rotating machinery, control-rod drives and contamination of the sodium coolant

24 Conclusion: U.S. Experience has advanced progress toward Generation-IV goals ~ 60 Years of U.S Experience, coupled with International experience, has demonstrated reliability of fast reactor technology Much has been learned through the U.S. technology development programs that support advanced fast reactor designs. Reliable, sustainable and safe operation has been demonstrated and the impact of design choices clarified. Areas for further development include: Development and demonstration of integrated fuel recycle technology Reduction in costs

Trends in Transmutation Performance and Safety Parameters Versus TRU Conversion Ratio of Sodium-Cooled Fast Reactors

Trends in Transmutation Performance and Safety Parameters Versus TRU Conversion Ratio of Sodium-Cooled Fast Reactors The Tenth OECD Nuclear Energy Agency Information Exchange Meeting on Actinide and Fission