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Home Dr. John Sackett Operating and Test Experience for the Experimental Breeder Reactor II (EBR-II)

Operating and Test Experience for the Experimental Breeder Reactor II (EBR-II)

Abstract – This paper describes the contributions of EBR-II to sodium-cooled fast reactor technology. The reliability of the EBR-II system was demonstrated through 30 years of operation and sodium maintenance technology was well established. As an irradiation test facility, Oxide, Metal, Carbide and Nitride fuels were developed. As an operational-safety test facility, the self-protecting response of a metal-fueled reactor to upsets was demonstrated and the benefits to safety were quantified in a PRA. The safety of operation with breached fuel was also demonstrated. As the Integral Fast Reactor (IFR) prototype, non-aqueous reprocessing and recycle of fuel was demonstrated and fuel containing minor actinides was fabricated and irradiated.  Waste forms for geologic storage of waste from fuel reprocessing were developed and qualified. When decommissioned, draining and reaction of the sodium to produce an acceptable form for disposal was accomplished, including passivation of residual sodium. The EBR-II experience and test program has established the viability of sodium-cooled fast reactor power plants.

I. INTRODUCTION

Fast reactors are an important element of a global renaissance of nuclear power because they can use as fuel the waste that thermal reactors produce (primarily long-lived minor actinides). Studies have indicated that approximately 20% of the fleet should be fast reactors to effectively manage this waste. Further, fast reactors can greatly extend the fuel supply (approaching a factor of 100). Extending fuel supply was the promise of fast-reactor development at the dawn of the nuclear age. EBR-I, a fast-reactor, was the first reactor in the world to produce electricity (December of 1951). EBR-II then followed, producing power in 1964 and operating for 30 years as a complete power plant (1,2). The EBR-II test program addressed all of the fundamental issues associated with fuel performance, fuel recycle, sodium technology and reactor safety. Based on this and extensive international experience (3,4), sodium-cooled fast-reactor technology has been shown to be successful.



II. EBR-II Design Approach

EBR-II suffered few of the negatives that have beset some of the fast reactor projects around the world and was able to develop technology that led to success of plants to follow. The reason for this success was based on design choices, attention to details of construction, disciplined operation and maintenance, and an aggressive test program that developed a deeper understanding of the technology(5).

EBR-II was a complete power plant, producing 20 MWe with a conventional steam-turbine (with superheat). The reactor produced 62.5 MWt. (See Table 1.)

Table 1. EBR-II Operating Parameters

Power Output, thermal    62.5MWt
Power Output, electric    20 MWe
Reactor Inlet Temperature    700 F
Reactor Outlet Temperature    883 F
Flow-rate Through Core    9,000 gpm
Volume of Primary Sodium    89,000 gal
Sodium Temperature to Superheaters    866 F
Sodium Temperature from evaporators    588 F
Steam Temperature    820 F
Steam Pressure    1,250 psig
Feedwater Temperature    550 F
Fuel    Metal 63% enriched
Primary System Configuration    Piped Pool
Steam Generator Design    Duplex Tube

EBR-II was a sodium-cooled reactor with a piped-pool configuration. That is, the coolant was drawn from a tank of sodium by two centrifugal pumps and then piped to a plenum at the bottom of the core. The sodium, after flowing through the core, was piped to an intermediate heat exchanger where heat was transferred to the secondary sodium system. This configuration allowed for leakage at the connections at the outlet of the pumps and at the intermediate heat exchanger, since primary sodium would simply leak back to the tank from which it was drawn. This also allowed for “ball and socket” connections at the pumps which facilitated their removal and replacement. The tank, which was a right circular cylinder, was maintained at a nearly uniform and constant temperature (700F), which limited thermal stress. Another important feature was that the tank included no penetrations through the wall; all penetrations were through the top. This also limited the risk for thermal stress, weld failure and sodium leakage.

A guard tank surrounded the primary tank with an annulus between them which allowed for detection of sodium leakage. The guard tank was in turn surrounded by concrete shielding which acted as a final containment vessel. Were leakage to occur in both the primary and guard tank, the core would not be uncovered and would be adequately cooled.

An inert gas (argon) filled the space between the tanks and their cover. Since there were penetrations though the cover for rotating machinery (pump shafts, fuel handling systems) and control rods, much attention was paid to seals to prevent ingress of oxygen which would result in formation of sodium-oxide. Sodium oxide as a deposit on this equipment would cause binding of the machinery and contamination of the sodium coolant with particulate. Much attention was paid to instrumentation for detection of oxygen ingress and remained a priority through the life of EBR-II operation.

Heat was removed from the primary sodium by three systems, 1) the secondary sodium loop which transfers heat to the steam generators, 2) thimbles immersed in the primary sodium and filled with sodium-potassium, which removed decay heat by natural convection, and 3) forced flow of air through the annulus between the primary tank and its guard tank, which also removed decay heat. Because decay heat removal did not depend on the secondary sodium loop, sodium in that loop could be drained to a storage tank for maintenance or in the event of a sodium leak. The secondary sodium loop was designed such that a severe reaction between sodium and steam would not endanger the reactor. The steam generators themselves were double-wall (a tube in a tube) to minimize the potential for leakage. The tube sheets were configured such that there was a plenum between the two tube sheets at each end which provided a path for sodium or steam to travel if one of the tubes were to fail, facilitating detection.

The EBR-II containment building was a domed cylinder designed to withstand a pressure of 24 psig. The design pressure was determined from analysis of a massive sodium fire, assuming that primary sodium was somehow sprayed as aerosol to containment (such an event is hypothetical). Interestingly, the extent of the fire is limited by available oxygen so a smaller containment is better, but its size must be great enough to allow for fuel handling, removal of major primary system components and other activities. The result was a rather large containment building. It was a welded steel structure lined with concrete to provide the ability to withstand high temperatures associated with a sodium fire in the building. The building was pressurized slightly to ensure cleanliness of the atmosphere in the building and all exhaust was through HEPA filters. Periodic pressure tests were conducted to verify leak-tightness.

One of the more unique aspects of the EBR-II containment was a fuel transfer tunnel that attached the building to an adjacent Fuel Cycle Facility. Spent fuel from EBR-II was transferred from the holding basket in the primary tank to a shielded cask which was then lowered through a hatch to the tunnel. The cask was then moved through this tunnel on rails where it was then mated with a transfer hatch at the Fuel Cycle Facility. Many thousands of transfers of fuel were made in this manner and during the first five years of operation, ~35,000 reprocessed fuel pins were returned to the reactor.

To facilitate fuel handling, a fuel storage basket was placed in the primary tank, capable of storing a large fraction of the core. Because the containment building and the primary tank of sodium were accessible during reactor operation, the storage basket would be pre-loaded with fresh fuel and when the reactor was shut down for refueling, fresh fuel would be exchanged for spent fuel in the storage basket. This greatly facilitated core unloading/loading, which typically took 3 days. Spent fuel could then be transferred on a schedule determined by cooling requirements for decay heat generation.

The fuel handling system was one of the more complicated features of EBR-II design. Because fuel handling is done in the blind (sodium is opaque), the equipment had to be precise with many checks and interlocks to ensure that transfers were being made properly. The main fuel assembly gripper was a straight pull through a penetration in a rotating plug in the top of the primary tank cover. This plug was one of two, placed within a larger plug at an eccentric position which allowed positioning of the gripper over any core location. The control-rods also penetrated this smaller rotating plug which required that they be disconnected and withdrawn before rotation of the plugs. When a fuel assembly was withdrawn by the gripper, it was captured by a transfer arm which positioned the assembly above a desired location in the fuel-storage basket. The basket could be rotated, and then raised to accept the assembly, which was then detached from the transfer arm.

Sealing of penetrations to the atmosphere above the primary tank was given special attention in the EBR-II design. The large rotating plug, in particular represented a special challenge. To provide a seal while the plug was rotated, a dip ring was immersed in an alloy of tin-bismuth, which was heated until molten for rotation of the plug. When the plug was secured, the alloy was cooled, sealing the interface. This arrangement created many problems for operation and maintenance. Frequent manual cleaning of this trough was necessary to avoid sticking the rotating plug, which would have created serious problems for recovery.

The intermediate heat exchanger (sodium to sodium) was a conventional shell and tube design, with primary sodium flowing to the shell at the top and exiting at the bottom while the secondary sodium flowed in tubes from the bottom. An electromagnetic pump was immersed in the primary sodium, providing forced flow for a smooth transition from forced to convective primary flow in event of a loss of power to the primary pumps (a feature later determined to be unnecessary).

Purification of the sodium was accomplished by on-line cold traps which cooled sodium to the point that sodium-oxide would solidify and collect on stainless-steel wire mesh. Later, special graphite traps were added to clean the sodium of Cesium, a fission product associated with extensive run-beyond-cladding breach testing in EBR-II. Both systems worked well.

Sodium leak detectors were installed throughout the plant and were of two main types, smoke detectors and “spark-plug” type detectors that would sense the presence of liquid sodium. In the steam generator building, acoustic monitors and hydrogen detectors were installed to detect a sodium-water reaction. In the event of a sodium-water reaction, blow-out diaphragms and panels were installed to relieve pressure away from the reactor building, and fast actuating valves would dump the secondary sodium to a storage tank.

Control of the reactor required two operators, one controlling sodium-flow in the secondary system (to maintain a constant reactor inlet temperature) and the second operator controlling reactor power through control-rod movement. Primary coolant flow was held constant. (More on this later; it was found that the reactor would load follow easily, responding through reactivity feedback as inlet coolant temperature changed in response to changes in power demand. No operator action is required in such a case.)

III. EBR-II Missions

The mission of EBR-II went through five distinct phases. The first was as a complete power plant with co-located fuel reprocessing. The second was as an irradiation facility, testing fuels for following fast reactors, primarily oxide fuel for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor Project (CRBRP). The third was as an operational safety testing facility, subjecting fuel and the plant to off normal conditions such as operation of fuel with breached cladding and ultimately, the reactor inherent-safety demonstration tests which subjected the reactor to anticipated transients without scram. The fourth was as the Integral Fast Reactor prototype, including demonstration of non-aqueous reprocessing and recycle technology. The fifth was decommissioning, yielding important information about the technology of sodium processing for disposal.

III-A Power Plant Operation

The power plant operated reliably for 30 years. Capacity factors approached 80% even with an aggressive testing program. Maintenance techniques were proven, with exposure to personnel less than 10% of that for a comparable Light Water Cooled Reactor (LWR). Effective sodium management was demonstrated, including successful suppression of a fire from a major sodium leak early in EBR-II’s operation. The steam generators operated quite well, with no failures or leaks in the systems, a testament to the duplex-tube design.

Fuel reprocessing was also very successful, with over 35,000 fuel pins reprocessed and recycled to the reactor in the first five years of operation. This demonstrated the viability of remote casting of metallic fuel elements and non-aqueous reprocessing of spent fuel, using a simple melt refining process.

Several key features of design and characteristics of the system contributed to the excellent performance of EBR-II as a power plant. The first was the sodium coolant. Sodium is compatible with the reactor materials in the primary circuit, with no corrosion found after 35 years at temperature. Sodium also has a high boiling point (>1640 F at atmospheric pressure) allowing the primary and secondary systems to be low-pressure. Consequently, there was no potential for high-pressure ejection of coolant. This feature is important for maintenance activities and is a major reason that there were no injuries from sodium leakage over the course of EBR-II operation. Already mentioned was the low exposure to maintenance personnel.

The pool-type primary system also provided distinct advantages. Its large thermal capacity limited the severity of thermal transients and therefore limited stress on the primary tank and components submerged within it. The piped pool configuration allowed for the majority of the primary sodium to be at reactor inlet temperature, further increasing the capacity of the sodium to absorb heat in the event of an upset. All primary system components were submerged in this relatively cold pool of sodium, which proved to be beneficial for their operating reliability and ease of removal for maintenance or repair. It also minimized the potential for leakage of primary sodium, since all penetrations were through the top of the vessel. The only leakage encountered was in smaller systems, such as sodium sampling and purification, which were external to the primary tank and which contained small inventories of sodium.

III-B Fuel Development

After the initial demonstration of EBR-II as a complete power plant, the reactor core was reconfigured to enhance its capability as an irradiation test facility. The inner blanket surrounding the core was replaced with a stainless-steel reflector which increased the flux levels and provided a smaller flux gradient across the core. The irradiation testing mission was directed primarily at development of oxide fuel for FFTF and CRBR, but also important was improvement of the performance of EBR-II metal fuel. In addition, nitride and carbide fuels were tested.

EBR-II metal driver fuel was significantly improved over the course of the 30 year operating life of the reactor (6, 7). Early in the development of metal fuel, failures in the cladding were seen at burnups as low as 1%. The reason was that the fuel would swell against the cladding, exerting enough force to cause it to fail. The solution was quite simple; the gap was increased between the fuel pin and the cladding which allowed the fuel to swell until the fission gas was released from the pin, stopping the swelling. Fission gas was released at about 2 atom % burnup when the gas bubbles in the fuel would interconnect, creating a porous fuel structure that allowed the fission gas to be released. This interconnected porosity would then backfill with sodium further enhancing the thermal conductivity of the fuel and resulting in very low fuel-centerline temperatures. The second modification to the fuel pin design was to increase the gas-plenum volume to accommodate the fission gas that was released. With these design changes, burnups of in excess of 20 atom % were achieved. In fact the limit was not reached but it is likely to be related to filling of the fuel porosity with solid fission products sufficient to again initiate fuel swelling.

Perhaps one of the most important aspects of metal fuel was its ease of fabrication. Metal fuel pins were produced in 100 pin lots by simple injection casting. Glass molds were lowered into molten fuel and then the system was pressurized, forcing fuel into the molds. The molds were removed, allowed to cool, the molds removed and then the pins were cut to length. They were then placed into the cladding tubes which contained a small amount of molten sodium as a thermal bond and a cap was then welded to close the tube. This process produced ~150,000 fuel pins and was carried out both at the EBR-II Fuel Manufacturing Facility (FMF) and by commercial vendors. It was also accomplished remotely and because of its simplicity was done without difficulty.

In addition to metal fuel development for EBR-II, eight full-sized assemblies (1800 pins) of metal fuel were irradiated to high burn-up in FFTF without failure. This work was done as part of a plan to convert the FFTF core from oxide fuel to metal, but the reactor was shut down before the conversion could be accomplished. Those assemblies have been returned to the INL where they are available for examination.

A full range of metal fuel compositions was tested, including uranium-zirconium and uranium-zirconium plutonium mixtures, with and without additions of minor actinides. Peak cladding temperatures reached 620 degrees C with maximum in-reactor exposures to 5 years. An important conclusion is the metal is a versatile and “forgiving” fuel design, able to accommodate a wide range of compositions.

As noted earlier, oxide fuel was also demonstrated to be viable, operating to high burnup and achieving the smear densities and power ratings desired. The major difference is  reactor safety. Metal fuel provides a large degree of self-protection in response to off normal events; oxide fuel does not as explained in the following discussion.

III-C Operation Safety Testing.

When the FFTF began operation, taking on a major role in irradiation-testing of fuels and materials, EBR-II was able to conduct more aggressive operational-safety tests. These involved integral plant-safety tests as well as fuel-safety tests. The interest for fuel was its performance with breached cladding under both steady state and transient over-power conditions. A particular concern for oxide fuel is formation of sodium-oxide as a reaction product once fuel is exposed to the coolant. Sodium-oxide is less dense than the fuel and can tend to split the cladding, causing progressive failure. However, the EBR-II program of run-beyond-clad breach testing supported the safety case for oxide fuel for both the MONJU reactor in Japan and the CRBRP reactor in the US. The testing was extensive and included operational transients in EBR-II as well as more aggressive tests in TREAT. The result of this work was data which demonstrated the safety of continued operation of oxide fuel with breached cladding, forming the safety basis for MONJU.

The question also arose about the performance of metal fuel with breached cladding, since testing of oxide fuel in the reactor would mask failure of cladding for metal fuel. The EBR-II driver fuel had to be qualified to operate safely for extended periods with breached cladding. Metal fuel has an advantage in that it is chemically compatible with the sodium. (Sodium is used in the fuel pin to enhance thermal conductivity between the metal fuel and the cladding). Extensive tests, including both steady state and transient over-power conditions, demonstrated that metal fuel was completely compatible with the sodium coolant and a breach in cladding would not “grow”. The safety case was made that breached cladding in metal fuel could be safely accommodated; no fuel loss would be expected.

A modification made to EBR-II to accommodate fission-gas release was installation of the cover-gas cleanup system which captured the noble gases, Xe and Kr. (Chemically active fission products, like Cs and I were captured in the sodium and subsequently cleaned by the sodium-cleanup systems). The cover-gas cleanup system used cryogenic cooling to capture these gases in an activated charcoal bed, working very well over the remaining life of the plant.

The most dramatic of the safety tests were those involving the whole plant, leading to the inherent safety demonstration tests conducted in April of 1986 (8, 9, 10, 11). The EBR-II plant was subjected to all of the Anticipated Transients without Scram (ATWS) events without damage, demonstrating the self protecting characteristics of a metal-fueled fast reactor.

The first of these was loss of all pumping power with failure to scram, simulating a station blackout with failure to scram. The reactor was brought to 100 % power and the pumps were turned off, allowing them to coast-down and coolant flow to transition from forced to natural-convective flow. Testing and analysis over the previous 4 years had been conducted to accurately model the reactor for this event, and the system responded as expected. Special in-core temperature monitoring had been provided as a safety system to scram the reactor if temperatures rose to unexpectedly high levels, but they did not.

Temperatures initially rose rapidly as the cooling flow decayed, but the increase in temperature introduced sufficient negative reactivity feedback that the power was also reduced rapidly, resulting in peak core coolant temperatures that were higher than for normal operation (~1300 degrees F vs. 890 degrees F at normal operation) but not high enough to damage the fuel. There was also significant margin to sodium boiling temperature, which would occur at ~1640 degrees F. The system power came down rapidly, reducing peak core temperature until they equilibrated at an average temperature very close to that at normal operation.

A point to be emphasized is that there was no fuel or core damage with this event, unlike what would occur in a conventional reactor system. In fact, this was the 45th test of ATWS events on this core and the reactor was restarted for a subsequent test that same afternoon.

The second test subjected the reactor to loss-of-heat-sink without scram. The reactor was brought to 100 % power and flow was stopped in the secondary heat transfer system, blocking the transfer of heat to the steam generators. As the reactor inlet temperature rose, negative reactivity feedback reduced power such that the temperature difference across the core was reduced; peak coolant temperature never increased. The reactor temperatures equilibrated at an average temperature very close to that at normal operation.
This behavior results because of the very strong negative feedback associated with neutron leakage as the coolant temperature increases and the lack of a strong positive reactivity feedback from Doppler as fuel centerline temperature falls. Metal fuel operates with a very low centerline temperature and therefore little Doppler feedback reactivity. For loss of cooling or loss of heat sink events, coolant temperature rises, power falls and if one has a high Doppler coefficient of reactivity in the system (such as for oxide fuel), the positive reactivity feedback will delay power reduction with the result that the sodium will boil. Boiling sodium will insert significant positive reactivity, likely leading to a severe overpower transient. For this reason, a metal fuel core is self-protecting against under-cooling events while an oxide-fuel core is not.

It was earlier thought that a high Doppler coefficient of reactivity was important to protect against severe overpower events, so tests were done in TREAT with metal fuel to determine its performance under such conditions. Many metal fuel pins from EBR-II were subjected to severe over-power events which took the pins to failure. It was found that the relatively low melting temperature of the fuel was important, since it softened and then flowed like toothpaste in a tube before breaching the cladding. This flow of fuel occurred rapidly and would be effective in introducing large negative feedback on severe transients, acting as an effective self-protecting mechanism. Also, metal fuel cladding failures typically occurred at 4 times nominal power, higher power than typical for oxide fuel which typically fails at 3 times nominal power.

Further tests were conducted to determine the load-following characteristics of the reactor, which are very good (12). Metal fuel is not adversely affected by cyclic changes in power and temperature, which coupled with its strong tendency to maintain a constant average core temperature, greatly facilitates its ability to load follow. EBR-II could be easily controlled by fixing control-rod position and controlling power demand at the steam turbine. A full range of safety and load following tests were conducted, including for example, rapid run-up of the primary pumps to their maximum capacity which would cool the core, insert positive reactivity and raise power level. No damage occurred to the fuel or core through all of these tests.

A level-1 Probabilistic Risk Assessment (PRA) was completed to quantify these results (13). It was shown that risks associated with EBR-II operation were substantially lower than typical LWR plants, an order of magnitude less. The EBR-II risks would have been lower still except for its seismic response. (Subsequent plants employ seismic isolation to mitigate even this risk).

III-D Prototype for the Integral Fast Reactor (IFR)

With all that was learned through of EBR-II operation, the results were integrated into an approach to fast-reactor design termed the Integral Fast Reactor (IFR) (14, 15). A new feature of the approach was a reprocessing technology that accommodated fuel containing actinides and that offered proliferation resistance (16). The Fuel Cycle Facility (FCF) was refurbished and equipment was installed to conduct the work. The heart of the process is an electro-refiner into which the chopped EBR-II spent fuel was placed. The potassium/lithium chloride salt in the electro-refiner was maintained at 500 C, into which the fuel dissolved leaving the cladding hulls behind. The anode for the electro-refiner was the fuel basket from which the fuel was dissolved and two types of cathodes were employed, a solid cathode upon which uranium was deposited and a liquid-cadmium cathode within which a mixture of uranium and transuranics were deposited. The reason that the system offers proliferation resistance is that it is virtually impossible to cleanly separate Pu. Through a quirk of nature, the free energies of Pu and the minor actinides in the salt are so closely aligned that it is virtually impossible to adjust electro-refiner voltages to distinguish between them for transport of material.

Spent fuel was first chopped and then loaded into the anode basket. Uranium was then electro-transported to a solid cathode. Subsequently, Pu and a mixture of minor actinides were transported to a liquid-cadmium cathode. From there, material was taken to a cathode processor where clinging salt was distilled from the product, to be returned to the electro-refiner. The fuel product was then consolidated into an ingot for subsequent casting into fuel.

An important aspect of the fuel cycle was the production of waste forms suitable for geologic storage. One was ceramic and the other metallic. To produce the ceramic waste form the electro-refiner, salt was cleaned of active fission products by flowing it over a zeolite bed, which was then consolidated into a ceramic waste-form after the addition of glass frit. To form the metallic waste-form, the cladding hulls and noble metals were recovered from the anode basket and the bottom of the electro-refiner and cast into a metal ingot. Extensive leach tests were conducted on these waste-forms which were qualified for long-term geologic disposal. Because the waste could be free of actinides, required storage times are on the order of hundreds, not thousands of years.

This work was overseen by a special committee of the National Academy of Sciences which issued a final report supportive of the technology.

III-E Decommissioning EBR-II

The first phase of decommissioning involved the removal of all fuel and blanket assemblies, 637 in all. This involved cleaning each assembly of sodium and transfer to a hot-cell facility for disassembly and repackaging, then transfer to interim storage. As a final step, EBR-II fuel is being reprocessed for recovery of uranium and production of waste forms suitable for geologic storage, as described previously. Defueling was accomplished over 14 months without difficulty.

The next phase concentrated on the technology for dealing with sodium coolant to produce a waste-form suitable for land-fill disposal. It was also important that residual sodium left in the coolant systems after draining be fully reacted so as not to pose a long-term hazard. The 89,000 gallons of primary sodium was thoroughly cleaned of fission products (especially Cs-137) and sodium oxide. It was then transferred to a sodium process facility where it was reacted with water to produce sodium hydroxide at a concentration of 73% by weight. The sodium hydroxide at this concentration is a solid product that could be stored in drums at a DOE land-fill. In addition to the EBR-II sodium, sodium drained from the Fermi-1 reactor was reacted and disposed of in this manner.

One of the more interesting challenges was passivating the residual sodium that remained in the primary system in order to place the system in a radiologically and industrially safe condition. After a number of laboratory tests, a solution was found. Moist CO2 was introduced to the primary system at a controlled rate, monitoring the reaction rate through the evolution of hydrogen. A question for this process was that since pools of sodium form a “scab” at the surface, whether the reaction would continue. It was found that over time, the CO2 would permeate this surface layer and the reaction would continue to completion, although the process could take several years. The volume of residual sodium has now been reduced to a point that it would be safe to flood the primary tank with water.

An important observation after the sodium was drained from the primary tank was that the condition of the tank and the components submerged in sodium was pristine. There was absolutely no corrosion of the stainless steel after 35 years in contact with hot sodium.

IV. CONCLUSIONS

The extensive program of operation and testing at EBR-II has established sodium-cooled fast reactors as a viable technology to support a nuclear renaissance. The ability of fast-reactors to manage nuclear materials for waste and fuel has been demonstrated, along with advantages for safety, operability and reliability. Cost remains the major issue but there are opportunities for significant cost reductions by taking advantage of the self protecting nature of the reactor system to simplify design. The major conclusions are that

•    A pool-type, metal-fueled LMR nuclear generating station can be reliably operated and maintained with large margins of safety to workers and the public.
•    Sodium system maintenance is straightforward and safe, facilitated by low pressure in operating systems.
•    Sodium-spills and fires are manageable, principally because of lack of high pressure driving fluid.
•    Sodium is highly compatible with reactor materials, facilitating long life.
•    Attention must be given to maintaining purity of the inert gas covering sodium systems to avoid sodium-oxide buildup on systems penetrating the interface.
•    Personnel radiation exposure levels are very low, typically <10% of those for LWR systems.
•    A metal-fueled LMR nuclear generating station can be passively safe, offering self protection against anticipated transients even without safety-system action.
•    Safety benefits have been quantified by PRA, which demonstrates very low levels of risks.
•    Metal-fuel offers exceptional benefits for reprocessing and recycle, conversion, load following, passive safety and benign behavior in degraded condition.
•    Fuel handling systems require much design, operating and maintenance attention to ensure reliability.

ACKNOWLEDGMENTS

Ron King, Yoon Chang, Leon Walters, Dave Wade
NOMENCLATURE

EBR-II  Experimental Breeder Reactor II
IFR        Integral Fast Reactor
FFTF     Fast Flux Test Facility
CRBR    Clinch River Breeder Reactor

REFERENCES

1.    J. I. Sackett, “Operating and Test Experience with EBR-II, The IFR Prototype” , Progress in Nuclear Energy, Vol. 31, No. ½, pp.111-129, (1997).
2.    J .I. Sackett, ”EBR-II Test and Operating Experience”, prepared for the U.S. Nuclear Regulatory Commission, (December 2008)
3.     D. C. Wade, et.al, “Application of World-Wide Lessons Learned to Advanced Fast Reactor Plant Design and Operation”, ANL/RE-97/1, (September 1994).
4.    J. Guidez and L. Martin (CEA-France), S. C. Chetal, P. Chellapandi and Baldev Raj, “Lessons Learned from Sodium-Cooled Fast Reactor operation and their Ramifications for Future Reactors with Respect to Enhanced Safety and Reliability”  (Indira Gandhi Centro for Atomic Research (India), Nuclear Technology, Vol. 164, Nov. 2008
5.    L.J. Koch, “EBR-II Design and Construction”, ANS – 2007)
6.    G. L. Hoffman, L. C. Walters, and T. H. Bauer, “Metallic Fast Reactor Fuels”, Progress in Nuclear Energy, Vol.31, No.1/2, pp. 83-110, 1997
7.    D. C. Crawford, D. L. Porter, and S. L. Hayes, “Fuels for Sodium Cooled Fast Reactors: US Perspective”, Journal Nuclear Materials, 371, 202 (2007)
8.    G. H. Golden, et.al “Evolution of thermal-hydraulics testing in EBR-II”, Nuclear Engineering and Design Vol.101, No. 1 (1987).
9.    H. P. Planchon et.al. “The Experimental Breeder Reactor II inherent shutdown and heat removal tests – Results and analysis”, Nuclear Engineering Design, 91 (1986) 287-296
10.    C. E. Lahm, et.al, “EBR-II driver fuel qualification for loss-of-flow and low-of-heat-sink tests without scram”, Nuclear Engineering and Design, Vol.101, No.1 (1987).
11.    H. P. Planchon, J. I. Sackett, G. H. Golden and R. H. Sevy, “Implications of the EBR-II Inherent Safety Demonstration Test Nuclear”, Engineering and Design, Vol.101, No.1 (1987).
12.    L. K. Chang, D. Mohr, H. P. Planchon, E. E. Feldman and N. C. Messick,  “Demonstration of EBR-II Power Maneuvers Without Control Rod Movement”, North-Holland Physics Publishing, May, 1988
13.    D. J, Hill, A. Ragland, J. Roglands, “The EBR-II Probabilistic Risk Assessment: Lessons Learned Regarding Passive Safety Reliability Engineering and System Safety, Vol. 62, No. 1, Oct. 1998, pp. 43-50.
14.    C. E. Till, Y .I. Chang and W. H. Hannum, “The Integral Fast Reactor-An Overview”, Progress in Nuclear Energy, Vol. 31, No. ½, pp.3-11, 1997
15.    D.C. Wade et.al, “The Safety of the IFR”, Progress in Nuclear Energy, Vol. 31, No. ½, pp. 63-82, 1997.
16.    H. F. McFarlane and M. J. Lineberry, “The IFR Fuel Cycle Demonstration”, Progress in Nuclear Energy, Vol. 31, No. ½, pp.155-173, 1997

John I. Sackett
Idaho National Laboratory
4762 Aspen Lane, Bozeman, MT, 59715 USA
406 585 3372

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Dr. John Sackett


SUMMARY OF EXPERIENCE

My 34-year career in advanced reactor and fuel-cycle development has led to international recognition in the areas of passive nuclear safety and advanced reactor design.  I am widely published in the technical community.
My technical achievements include developing the approach to reactor safety now considered the standard for advanced nuclear systems, namely design approaches which emphasize passive safety. I was the principal author of the program of Operational Safety Testing at the EBR-II Reactor and was a principle in developing the Integral Fast Reactor Program, the major nuclear R&D program for DOE in the 1980s and 1990s

EDUCATION
Ph.D., Nuclear Engineering, University of Arizona, 1970
B.S., Mechanical Engineering, University of Idaho, 1965

CHRONOLOGICAL EXPERIENCE SUMMARY


Retired from Argonne National Laboratory Feb. 2004 as the Associate Laboratory Director.
Currently hold part time positions as an adjunct faculty member of Idaho State University. Also am a part time employee as a senior advisor with Idaho National Laboratory.

These duties involve research and consulting as a nuclear engineer with particular expertise in fast-reactor design, fuel cycle and safety.

Associate Laboratory Director, Argonne National Laboratory, 2002 to 2004
As Associate Lab Director, I had responsibility for Engineering Research at Argonne National Laboratory, with emphasis on developing advanced nuclear power systems.  I was responsible for 1,150 employees and a budget of approximately $150M, emphasizing nuclear energy, space nuclear systems, and energy conversion technologies, including hydrogen and batteries 

Deputy Associate Laboratory Director for ANL-West,  1991-2002.

Was responsible for management of the ANL-W site in Idaho, including 13 major nuclear facilities and supporting laboratories.

Director, EBR-II Division,  1989-1994.

Was responsible for the operation and utilization of the Experimental Breeder Reactor – II (EBR-II), in support of the Integral Fast Reactor Program.

University of Chicago / Argonne National Laboratory.

Joined Argonne National Laboratory, operated by the University of Chicago, in 1970 as a staff engineer with the EBR-II project. Rose through the ranks to become Associate Laboratory Director in 2002.

AWARDS RELEVANT TO NUCLEAR POWER

•    Walker Cissler International Award for outstanding contributions to fast-reactor design.
•    Fellow, American Nuclear Society.
•    U of Chicago Award for Distinguished Performance at ANL.
•    American Nuclear Society Special Award for Outstanding Contributions to Reactor Safety.
•    Outstanding Idaho Engineer. 
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