7. Fast Breeder Reactors - FAST BREEDER REACTORS M. Ragheb...

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Unformatted text preview: FAST BREEDER REACTORS M. Ragheb 10/9/2007 INTRODUCTION Energy production by fast breeding remains the main goal of the Liquid Metal Fast Reactor (LMFR) to ensure a sustainable long term fissile fuel supply. In addition, the use of LMFRs allows the recycling of the Minor Actinides content of nuclear waste burning them to produce energy and reduce the amounts of disposed waste. Another advantage of the LMFR is its higher thermal efficiency compared with water-cooled reactors. The sustainable, environmentally clean long term use of nuclear power can be achieved with fast reactors, since thermal reactors are capable of burning less than 1 percent of the uranium fuel. It is surmised that the known reserves of uranium will fuel thermal reactors for only a few decades. Fast reactors burn most of the uranium fuel extending the power producing capability of the uranium reserves into the hundreds of years, making the recoverable energy resource from uranium larger than from coal. Fig. 1: The Experimental Breeder Reactor I, EBR-I, in the Idaho desert turned into a museum. It used a Na-K eutectic alloy that was liquid at room temperature as a coolant. 238 The fast reactor first transforms the 99.3 percent in the original ore abundant isotope U 239 into Pu , then burning it to produce to produce 50 to 60 times as much energy per metric tonne of uranium ore as a thermal reactor. Recycling becomes an essential part of the fast reactor system since the fuel is recycled through the reactor several times. The first nuclear electricity was produced on December 20, 1951 in the Experimental Breeder Reactor I or EBR-I at Idaho. It was turned into a museum after an accident, and was superseded by the Experimental Breeder Reactor II, or EBR-II which successfully produced electricity for more than 25 years. Fig. 2: The Experimental Breeder Reactor I, EBR-I, lighted up a string of light bulbs with the first produced nuclear electricity on December 20, 1951. Fig. 3: The Experimental Breeder Reactor II, EBR-II, at Idaho used a sodium coolant. Fig. 4: EBR-II Flow diagram. Fig. 5: EBR-II Reactor building. Fig. 6: EBR-II sodium pump. Fig. 7: EBR-II pressure vessel configuration. Fig. 8: EBR-II core and blanket assembly. Fig. 9: EBR-II control rod assembly. In the earliest days emphasis was placed on the breeding of fissile material. The increasing availability of cheap fossil fuels in the 1960s shifted the emphasis to include other uses for fast reactors, particularly for the control of plutonium stocks and the treatment of radioactive wastes. In spite of these additional functions the main long term importance of fast reactors as breeders, essential to world energy supplies, remains unchanged. Another advantage of the LMFBR concept is its operation at high temperature offering a high thermal efficiency. In addition it operates at low pressure so that a pressurization system is not required, and the loss of cooling through depressurization is not a safety consideration. This makes the fast reactor as an attractive inherently safe system. HISTORY In the 1940s it was realized that fast reactors would have potential advantages over thermal reactors because the excess neutrons available. These could be used for breeding fissile material from the fertile isotopes. This is the key to utilizing the 238 enormous world-wide energy reserve represented by U . The development of civilian fast reactors in the late 1940s involved test reactors such as Clementine and the Experimental Breeder Reactor I (EBR-I) in the USA and BR2 in the USSR. There were also low-power experimental assemblies such as Zephyr and Zeus in the UK. Demonstration reactors followed such as the EBR-II in the USA, BOR-60 in the USSR, Rapsodie in France and DFR in the UK were constructed in the 1950s and 1960s. This eventually led to prototype power reactors such as Phénix in France, PFR in the UK and BN-350 in Kazakhstan, and finally full scale power plants like the Super Phénix or SPX in France and BN-600 in Russia. FAST REACTOR POWER PLANTS Prototype fast reactors have been built with a power level of 250 MWe such as Phénix in France and the Prototype Fast Reactor (PFR) in Dounreay, UK. The KNK2 reactor produced 20 MWe and the 300 MWe Schneller Naturiumgekülter Reaktor (SNR300) in Germany. A full power commercial plant is the 1,200 MWe Super Phénix-1 (SPX1) plant in France. It was built as a collaborative effort between Germany, Belgium, the Netherlands, Italy and the UK. It is owned and operated by three utilities: Electricité de France (EDF), Schnell-Brüter-Kernkraftwerks-Gesellschaft (GSF) and ENEL. The Mixed Oxide Fuel (MOX) a mixture of PuO and UO is the preferred fuel 2 2 with sodium as a coolant. Two design concepts: the pool type and the loop type configurations have been considered, with the pool type gaining a preference. Both approaches use a primary and a secondary heat transport systems followed by a steam generation loop to the turbine plant. SAFETY CONSIDERATIONS The reactor core, primary coolant pump and the intermediate heat exchanger are contained in the main reactor tank in the pool design. The liquid sodium metal is contained in a simple double walled tank without penetrations below the sodium surface level and operating at atmospheric pressure. The loss of primary coolant becomes so unlikely as to be incredible. The primary sodium has such a large thermal heat capacity that it can survive the loss of decay heat cooling after the reactor has been shut down for about 10 hours. There exist substantial margins between normal operating temperatures and the coolant boiling temperature. Fig. 10: Super Phénix fast reactor, France. The concept possesses a strong negative power coefficient of reactivity associated with the fuel: when the temperature of the fuel rises, the power goes down in the absence of any control action. There exist no significant positive coefficients below the sodium boiling point. This implies that such a reactor is completely stable under normal operation. It can survive some hypothetical fault conditions for which the design intent of fast automatic shutdown is assumed not to take place. For instance, after a hypothesized loss of all pumping power to the coolant flow in the secondary circuits, there would be a reduction of power following the negative reactivity effects arising from thermal expansion, to a degree that can be removed by the emergency decay heat removal loops. As engineered safety features, highly reliable shutdown and decay heat removal systems are provided. In the prototypes used reliable shutdown systems never experienced ant failures. Two separate systems are used to remove the decay heat following shutdown. The first system uses the conventional steam system associated with the turbine. The second uses dedicated heat exchangers immersed in the primary coolant transferring heat to a Na or a Na-K alloy that is liquid at room temperature, from which heat is removed to the atmosphere. The plant temperatures can be maintained at safe levels by natural air flow in needed. Fig. 11: Super Phénix plant configuration. The double walled reactor tank and its roof provide a strong primary containment structure that can deal with a wide range of hypothetical accidents. A secondary containment is provided around the primary circuit with the ability to deal with an internal pressure with a low leak rate. With the loss of coolant made incredible, a highly reliable shutdown and decay heat removal systems, the reactor design has practically no faults within the design basis down to 1 in a million, which produces a possibility for any active release from the fuel. The low operating pressure of the sodium and the enormous affinity of liquid 131 sodium for fission products such as I provides a second barrier. Fig. 12: Super Phénix coolant circuit. COMPONENTS DESIGN Since the sodium circuit operates at virtually atmospheric pressure, the main stresses on the components are due to the temperature gradients with creep and fatigue being the main considerations. The steam generators receive particular attention because the leaks of water or steam into the sodium could produce secondary damage and shorten the components lifetimes. Once through units with helically wound tubes have been used and ferritic materials such as 9Cr 1Mo alloys have been adopted. Inservice inspection methods using ultrasound to monitor the components immersed in sodium are meant to extend the life of the structures. FUEL AND CLADDING COMPOSITION The choice fuel is the Mixed Oxide Fuel (MOX) which is a mixture of PuO and 2 UO . Burnup is the design parameter of most significance for the fuel cycle cost. Higher 2 levels of burnup imply more favorable fuel costs and are achievable at a level of 10-20 percent. The fuel burnup is defined as: A 10 percent fuel burnup corresponds to 95,000 MWth.days / metric tonne of fuel of energy release. It is expected that the fuel cycle costs would be lower than for thermal reactors compensating for a higher capital cost component for fast reactors. Potential new cladding and wrapper materials can increase the fuel burnup such as 10Cr 25Ni, Inconel 706, Nimonic PE16 alloys, dispersion strengthened ferritic steels for the cladding, and martensitic materials for the wrappers. Fig. 13: Super Phénix reactor vessel. Fig. 14: Super Phénix core configuration. Fig. 15: Super Phénix fuel handling system. Fig. 16: Super Phénix fuel assembly. Fig. 17: Super Phénix steam turbine. REFERENCES 1. A. Judd, “Fast Breeder Reactors,” Pergamon Press, Oxford, 1981. 2. J. G. Yevick, Ed., “Fast Reactor Technology: Plant Design,” MIT Press, 1966. 3. W.H.Hannum, Guest editor, “The Technology of the Integral Fast Reactor and its Associated Fuel Cycle,” in “Progress in Nuclear Energy,” Volume 31, No. 1/2, 1997, 4. Alan E. Waltar and Albert B. Reynolds, "Fast Breeder Reactors," Pergamon Press, 1980. 5. Alan E.Waltar, "America the Powerless, facing our Nuclear Energy dilemma,” Cogito Books, Madison, Wisconsin, 1995. ...
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This note was uploaded on 06/16/2010 for the course NPRE 402 taught by Professor Ragheb during the Spring '08 term at University of Illinois at Urbana–Champaign.

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