Experimental Breeder Reactor II

Abstract

The Experimental Breeder Reactor II (EBR-II) operated from 1964 to 1994. EBR-II was a sodium-cooled fast reactor operating at 69 MWth producing 19 MWe. Rather than using a loop approach for the coolant, EBR-II used a pool arrangement where the reactor core, primary coolant piping, and primary reactor coolant pumps were contained within the pool of sodium. Also contained within the pool was a heat exchanger where primary coolant, which is radioactive, transferred heat to secondary, nonradioactive, sodium. The nuclear power plant included a sodium boiler building where heat from the secondary sodium generated superheated steam, which was delivered to a turbine/generator for electricity production. EBR-II fuel was metallic uranium alloyed with various metals providing significant performance and safety enhancements over oxide fuel. The most significant EBR-II experiments occurred in April 1986. Relying on inherent physical properties of the reactor, two experiments were performed subjecting the reactor to loss of primary coolant flow without reactor SCRAM and loss of the secondary system heat removal without reactor SCRAM. In both experiments, the reactor experienced no damage. This chapter provides a description of the most important design features of EBR-II along with a summary of the landmark reactor safety experiments.

1. Introduction

The worldwide nuclear power industry is currently dominated by light water reactor technology. However, U-235 fissile material resource utilization challenges are likely to drive the need for non-light water reactor technologies when one considers timelines extending beyond the next half century. Many alternative reactor technologies that are capable of addressing the resource constraints of light water reactors are currently being pursued.

It is frequently worthwhile to look to the past as a means of guiding the path for the future. The first demonstration nuclear power plant was influenced by the expectation that limited supplies of fissile material will necessitate breeding fissile material. The Experimental Breeder Reactor I (EBR-I) achieved initial power production operation on December 20, 1951 (see Figure 1) and produced the first significant amounts of electrical energy generated by nuclear fission. EBR-I was a sodium-potassium cooled fast neutron spectrum reactor capable of breeding more fissile material than it consumed. The reactor was part of a power plant design that included steam generation and a turbine/generator system.

Figure 1.

Chalk message at EBR-I [1].

Following the success of EBR-I, the Experimental Breeder Reactor II (EBR-II) was constructed near EBR-I on the high-altitude arid Snake River Plain of southeastern Idaho in the western United States. Like EBR-I, EBR-II was a complete power plant demonstration, and it also included an attached fuel cycle facility to reprocess spent fuel using a melt refining process (see Figure 2). The reactor was a sodium cooled fast reactor (SFR) capable of producing more fissile material than it consumed. EBR-II achieved initial criticality in 1964 and operated until 1994. The reactor produced 19 MWe and supported decades of sodium cooled fast reactor development activities. The success of EBR-II provides insight into the potential benefit of future widespread use of sodium cooled fast reactors as a means of addressing fissile material resource limitation issues. It should also be noted that numerous other sodium cooled fast reactors have been developed including, but not limited to, Fermi I and the Fast Flux Test Facility in the US, Phénix and Super Phénix in France, Joyo and Monju in Japan, BN-350 in Kazakhstan, BN-600 and BN-800 in Russia, as well as sodium cooled fast reactors in India and China.

Figure 2.

Experimental breeder reactor II [2].

From an industry perspective, there is a resurgence of interest into sodium cooled fast reactors. Two commercial entities have proposed the use of sodium cooled fast reactors. The TerraPower company is pursuing a sodium cooled fast reactor coupled with a molten salt heat storage capability. The reactor is capable of producing 345 MWe as well as boosting the output to 500 MWe by using heat stored in molten salt. The reactor is called Natrium, which is Latin for sodium. In October 2020, the US Department of Energy awarded TerraPower funding to demonstrate the Natrium technology. TerraPower is targeting 2023 for submission of a construction permit from the US Nuclear Regulatory Commission. The planned location for the reactor will be one of four prospective sites in the state of Wyoming in the western United States. Furthermore, the Oklo Power Company has a sodium cooled fast reactor design which produces 4 MWth and integrates significant inherent safety features into the design. Oklo Power submitted the first-ever combined construction and operation license application to the US Nuclear Regulatory Commission in March 2020.

From a US Government perspective, the US Department of Energy is pursuing the Versatile Test Reactor (VTR). The VTR is a sodium cooled fast reactor that will operate at 300 MWth. The purpose of the VTR is to provide a very high neutron flux (4 x 10¹⁵ n/cm² sec) which will be used to test fuels and components for a wide range of advanced reactor concepts. The VTR project received Critical Decision–1 approval in September of 2020, allowing the project to proceed to preliminary design.

With this information in mind, it is worthwhile to reflect on the design and performance EBR-II since it provides tremendous knowledge and potential direction for sodium cooled fast reactors moving forward.

2. Power plant and reactor design

EBR-II was a complete power plant along with an attached fuel cycle facility. The reactor containment was centered between the sodium boiler building and the turbine/generator building. The reactor was an SFR which acted as a breeding facility and test bed for liquid metal fast breeder reactors [3]. Along with this, EBR-II produced electricity as part of its overall demonstration. Being a fast neutron spectrum reactor, the neutron chain reaction was driven primarily by fast neutrons. Fast neutrons often invalidate many assumptions commonly assumed for light water reactors. The long neutron mean free path associated with a fast neutron spectrum is indicative that much of the core is coupled, meaning there are relatively few localized reactivity effects. This often helps prevent localized peaking. The long mean free path of neutrons also means that negative reactivity insertion due to control rods in a few sections of the core provides the necessary means to shut down the reactor.

EBR-II was a pool-type SFR, meaning the core, and all supporting structures, were contained in a double walled vessel comprised of 86,000 gallons of primary sodium [4]. Due to this design, leaks in any of the primary system piping would drain into the primary coolant. This would result in a loss of plant efficiency but would not leak primary sodium outside the vessel. This design is unlike loop type reactors (i.e. Fast Flux Test Facility, Monju, SuperPhenix), where a leak in the primary coolant had the potential to cause a sodium fire and release activated sodium and would likely cause prolonged outages for repairs.

From a reactor operating perspective, sodium couples four very important properties: 1) extremely high boiling point (870 C) at atmospheric pressure, 2) outstanding heat transfer properties owing to its metallic nature, 3) relatively high atomic weight compared to neutrons leading to limited neutron moderation, and 4) a low neutron absorption cross section along with a relatively short neutron activation half-life of 15 hrs. These properties allow sodium to be used as an outstanding fast reactor coolant. The most obvious drawback of using sodium metal as a reactor coolant is the fact that it reacts with water and evolves hydrogen in the reaction process. The sodium-water reaction can be violent especially when the evolved hydrogen combines with oxygen. The reaction between sodium and water follows two primary schemes forming sodium hydroxide and sodium oxide as shown in Eqs. (1) and (2). In both reactions, hydrogen is also produced which presents a flammability and explosion hazard. It is important to keep in mind that a leak of high temperature sodium to an air atmosphere will result in dense white smoke which makes leak identification simple.

$$Na+H_{2}O\to \mathit{NaOH}+\frac{1}{2}{H}_2$$E1

$$2Na+H_{2}O\to N{a}_{2}O+H_2$$E2

The primary coolant arrangement for EBR-II can be seen in Figure 3. This highlights the major components associated with the primary coolant. Cold coolant (~370 C) was drawn in via two primary pumps, each of which supplied ~18,000 liters per minute of coolant and was split into a high-pressure and lower pressure inlet plenum at the bottom of the core. Of special note, the two primary coolant pumps were single-stage centrifugal mechanical pumps: a first of their kind for liquid metal coolant at the time. After flowing through the core, hot coolant (~480 C) then flowed into a shared upper plenum with a single outlet (shown as a “Z” in both figures). The hot coolant then entered the heat exchanger and was discharged back into the primary coolant pool. To filter out impurities, a cold-trap system continually filtered primary coolant by reducing the sodium temperature to reduce the solubility limits and precipitate out impurities. Above the sodium was ~12 in. of argon gas providing a protective inert cover for the sodium coolant.

Figure 3.

Primary coolant system for EBR-II [3].

The secondary system extracted heat from the primary system which was then used to drive a Rankine cycle for power generation [4]. The sodium flow rate for the secondary system was 23,000 liters per minute, with an inlet temperature of 310 C and an outlet temperature of 460 C. Transferring heat from the radioactive primary sodium to non-radioactive secondary sodium provided a safety enhancement and the ability to place much of the secondary system in a separate sodium boiler building, which was physically separate from the main reactor building. This separation reduced the time required in containment and reduced the potential for radioactive impurities to cause exposure. The sodium boiler building design incorporated a sacrificial plastic wall located away from the reactor building. The sacrificial wall would fail in the event of a catastrophic sodium water reaction in the sodium boiler building thereby directing the reaction energy away from the reactor building.

For the Rankine cycle, superheated steam was generated at 450 C with a pressure of 9000 kPa: this powered an off-the-shelf 20 MW turbine generator. The ability to use off the shelf components, helped reduce cost in the secondary system (one of the primary objectives of EBR-II). The secondary system allowed for a steam by-pass to continually dump heat despite any electrical needs. The overall EBR-II heat transfer pathway is shown in Figure 4.

Figure 4.

EBR-II heat transfer pathway [2].

In addition to the primary and secondary systems, an auxiliary pump was used to ensure a low-pressure flow rate was always present, despite normal power failure. The auxiliary pump was attached to a DC battery system, which would last long enough to allow the EBR-II system time to convert from forced cooling to natural circulation. To aid in the natural circulation, two shutdown coolers penetrated the primary coolant tank and allowed for heat removal directly to the atmosphere. The shutdown coolers contained sodium-potassium which extracted heat from the primary system and was exposed to an air-cooled heat exchanger.

The EBR-II core used 637 hexagonal subassemblies that made up the driver, inner blanket, and outer blanket regions. Figure 5 shows the top of the reactor core prior to the introduction of sodium coolant. The driver region was where a majority of the neutron flux was generated, which meant that a majority of the power was generated in this region. In terms of an equivalent cylinder, EBR-II had a diameter of ~20 in. and a height of ~14 in. Subassemblies were generally broken up into a few major categories: driver, blanket, control, reflector, and experiment [5].

Figure 5.

EBR-II reactor Core [4].

Subassembly types shared many characteristics, the most notable being the outer dimensions which allowed for subassemblies to be moved throughout the core, depending on the specific needs. Each assembly was hexagonal in shape, and had an outside flat-to-flat distance of 5.82 cm with a flow duct wall thickness of 0.10 cm. All subassemblies also had an upper adapter (this allowed for subassemblies to be placed and removed from the core), and a lower adapter. The lower adapters had slightly different configurations to ensure subassemblies were placed in the correct location.

Driver fuel assemblies were comprised of, in general, a lower adapter, fuel pin grid, and upper preassembly. Coolant flowed from the inlet plenum into the lower adapter, through fuel pin grid (where heat was transferred to the coolant), and out the upper preassembly into the outlet plenum. Multiple driver fuel designs were used throughout the lifetime of EBR-II, and as such, a brief description of the MK-II fuel assembly design is given [5]. Since these were used throughout the life of the reactor. Comprised within the fuel pin grid were 91 fuel pins in a hexagonal lattice with a fuel pitch of 0.56 cm. Fuel pins are described further in a later section. Half-worth driver assemblies where nearly identical to driver fuel assemblies, however, half of the fuel pins were replaced with stainless steel pins; this reduced the reactivity of the fuel assembly. Half-worth driver assemblies were typically placed near the center of the core to dampen peaking effects.

Blanket assemblies were used throughout the life of EBR-II, where they were initially inserted around the core to breed plutonium. Blanket assemblies contained 19 fuel pins comprised of a fuel slug (outer diameter (OD) 1.1 cm), sodium bond (OD 1.16 cm), and a stainless-steel cladding (OD 1.25 cm). Blanket fuel pins were much larger than their driver counterparts due to the lower power density and a desire to increase the fuel to sodium ratio to promote breeding in the pins. Blanket fuel pins were 1.43 m long.

EBR-II, like many SFRs, used full assembly positions for the safety and control rods (denoted control assemblies from here on). Control assemblies had an inner hexagonal duct (flat-to-flat diameter of 4.90 cm) which contained a fuel region with 61 fuel pins which could be brought into the plane of the driver fuel to add reactivity to the core. Some control rods (designated high worth control rods) had a region comprised of seven B₄C pins directly above the fuel, which acted as an additional poison to ensure the reactor could shut down and remain shut down.

Reflector assemblies did not contain a pin grid section, but instead contained stacks of stainless-steel hexagonal blocks. These blocks were used to reflect neutrons back into the core and were typically placed in the periphery.

Experimental assemblies were unique in both design and contents. These assemblies maintained the hexagonal duct but could contain fuel, material, monitor, etc. experiments. Experimental assemblies are described in greater detail in a subsequent section.

Fuel pins consisted of a metallic fuel slug (OD of 0.33 cm), sodium bond (OD 0.38 cm) and stainless-steel cladding (OD 0.44 cm). The total length of the fuel pin was 62.04 cm, where the fuel slug had a length of 34.29 cm. Above the fuel slug was a helium plenum to capture fission product gasses and was often tagged with trace amounts of xenon to allow for the determination of burst fuel pins. Each fuel pin was surrounded by a wire-wrap with a diameter of 0.125 cm and an axial pitch of 15.24 cm. The wire wrap was used to ensure fuel pins did not come in contact with each other and provided additional coolant mixing to encourage heat transfer. Throughout the lifetime of EBR-II, the fuel pins changed slightly in dimensions, however, the dimensions presented provide a reasonable representation of a typical fuel slug. Figure 6 shows an arrangement of driver fuel pins along with the wire-wrap.

Figure 6.

Fuel pin arrangement [6].

The fuel slugs in Mk-II subassemblies comprised a uranium-fissium alloy (95 wt. % uranium 5 wt. % fissium), meaning that the fuel was metallic in nature, compared with the typical ceramic fuel (uranium-oxide) found in light water reactors. The uranium in the fuel was enriched to between 45 wt. % and 67 wt. % U-235, again in stark contrast to the typical 5 wt. % light water reactor fuel. Fissium was comprised of elements to simulate dominant mid-fuel cycle fission products. The short-highly-enriched fuel for EBR-II created a very short-flat core, which provided multiple inherent safety benefits, described in greater detail later.

One other noteworthy feature of the EBR-II design involved a fuel storage basket located within the primary tank. The fuel storage basket contains 75 indexed storage tubes in three concentric rings. Each tube could accommodate a single fuel assembly. The fuel storage basket was accessed essentially anytime by operators including when the reactor is operating at full power. The fuel storage basket provided great operational flexibility. During reactor operation, spent fuel assemblies stored in the basket could be removed one at a time and transferred out of the reactor facility and delivered to a hot cell facility for storage and disassembly. Fresh fuel and experimental assemblies could also be loaded into the basket during reactor operation. When the reactor was shut down, operators could then quickly move spent fuel assemblies from the core into the fuel storage basket and move fresh fuel from the basket into the core making the refueling outage time as short as possible. Since the driver region of the core contained roughly 100 assemblies, the 75-assembly fuel storage basket provided ample capacity for staging fresh fuel assemblies as well as holding spent fuel assemblies removed from the core. With the fuel storage basket located within the primary tank, the sodium coolant provides sufficient heat transfer capacity to ensure the spent fuel assemblies are adequately cooled prior to their removal.

3. Experiments

Experiments were not placed in specific assembly locations in the core. This is unlike many light water test reactors which have specific ports or testing locations. Instead, experiments were often placed in the same hexagonal duct as a typical driver fuel assembly. This meant multiple experiments could be placed in the same assembly, experiments could be intermixed with fuel pins, or experiments could be placed in an assembly with dummy stainless-steel pins. The placement of an experiment in the core was largely determined by the conditions required for the experiments. If an experiment needed a large flux of high energy neutrons in a short period of time, it could be placed in the center of the core. On the other hand, if an experiment needed to experience a large neutron fluence over a long period of time, it could be placed in the periphery of the core. Overall, an experiment could likely be placed in any assembly position within the core, with the exception of the control/safety assemblies. To compensate for any loss of reactivity due to adding experimental assemblies, additional driver assemblies were placed in the periphery of the core.

EBR-II also examined multiple endurance type testing for both fuel and cladding [7]. In the 1970’s, a series of experiments examined running fuels to cladding breach (RTCB) and running fuel beyond cladding breach (RBCB). These experiments were used to help increase the burnup capabilities for fuels and determine neutron fluence limits for these fuels. To accomplish this, an additional cover-gas cleanup system (GGCS) was installed to help remove radioisotopes that leaked from the fuel and into the argon cover gas.

3.1 Dry/wet critical experiment

In April of 1961, before EBR-II was used as a power producing or breeding facility, it underwent a series of zero power experiments (designated as less than 1 kW of power) before the primary system was filled with sodium [8, 9, 10]. To perform the dry critical experiment, fuel and blanket assemblies that would be used for normal operations were loaded into the core in a similar configuration to when sodium would be added. For this, additional driver assemblies (~87 driver assemblies compared with ~56 driver assemblies for a sodium filled core) were required achieve criticality since the lack of sodium increased neutron leakage in the core. These experiments were able to take place while construction work was being performed elsewhere in the plant.

The basis of these experiments was twofold. The first was used to determine the performance of the system without sodium, which allowed them to subsequently identify sodium effects on system neutronics. The second gathered operational data to determine if modifications or improvements were required prior to adding sodium. To gather this information, four major experiments were conducted. The first was to determine the strength of the neutron source and the neutron detector responses to ensure an adequate relationship between the two. The second was an approach to critical to verify the ability to insert assemblies and determine the dry critical mass. The dry critical mass could then be compared with the wet critical mass to determine the total reactivity worth of the sodium. The third aspect examined the neutron flux distribution and fission distribution throughout the core and provided a power calibration. The final aspect that was examined was a series of reactivity measurements. This included seven measurements ranging from the total worth of the control rods, individual control rods, to the dry isothermal temperature coefficient of reactivity.

3.2 Connected fuel cycle

EBR-II was originally designed as a power-producing facility which would be able to produce more fuel (in the form of plutonium) than it consumed. To accomplish this, blanket subassemblies were placed around the periphery of the core, where neutrons which leaked out would be absorbed by U-238 to produce plutonium. In addition to creating a core design which was favorable for generating fuel, additional facilities were constructed on-site to allow for fuel/experiment examination and fuel reprocessing.

The fuel cycle facility (FCF) was built to allow for post-irradiation examination of experiments placed in the core [11]. FCF allowed for experiments to be removed from one subassembly and placed in a new subassembly for further irradiation if necessary. Along with this, FCF was used to reprocess spent EBR-II fuel using a crude melt refining technique rather than a complicated and large solvent extraction process. Melt refining involved melting the spent fuel elements and mechanically separating fission products and slag from the uranium. The uranium (or other actinides) was then used to fabricate additional fuel.

The last decade of operations for EBR-II was focused on the Integral Fast Reactor (IFR) concept [12, 13]. This project encompassed nearly all aspects of life for a nuclear reactor. The IFR concept was meant to overcome many obstacles such as proliferation concerns, waste generation concerns, and reactor safety concerns. The IFR concept was meant to provide the United States (and the world) with a nuclear energy concept that could provide a nearly inexhaustible energy supply for the future. Unfortunately, in 1994, the IFR concept and indeed EBR-II was terminated, and the full realization of the IFR concept never came to pass.

3.3 High burnup

One of the many advantages of fast reactor technology is the ability to “burn” to a greater extent than thermal reactor. The average burnup for a typical light water reactor is 45,000 MWD/MTHM. EBR-II demonstrated 20 atom % burnup which is the equivalent of 190,000 MWD/MTHM. These burnups are possible primarily because of the fast neutron spectrum present in the reactor. Along with the energy extracted from the fission of U-235, the fast spectrum transmutes the U-238 to higher order actinides. Those elements are subsequently fissioned, releasing energy rather than creating a problematic waste issue. The transmutation process does happen in thermal spectrum reactors, but to a far lesser extent. Given this, the extractable energy from fast reactors is fundamentally limited by the structural materials of the fuel and how long they can serve the engineering requirements under significant irradiation.

3.4 Inherent safety

April 3rd 1986 is a date that is unknown to the general public and to large portions of the nuclear industry. The reason was that nothing newsworthy happened that day. The EBR-II functioned as designed without any damage, everyone working in the facility went home that day, and in general it was like any other day in southeast Idaho. Despite nothing being widely reported that day, one of the most significant achievements in nuclear reactor technology was demonstrated. The EBR-II was intentionally placed into an accident scenario that would have melted down any light water reactor. The accident scenario far exceeded that of Three Mile Island. The scenario was to operate the EBR-II at 100% power, disable the primary coolant pumps (for the first experiment) and the secondary cooling pumps (for the second experiment). Both experiments were conducted without SCRAM the reactor. To achieve the plant conditions listed above, EBR-II was modified to create the conditions but still remain in control in case unpredictable behavior occurred. An example of a modification was the cooling pumps. They were not directly disabled; the pump controllers were modified to simulate coast down function shapes, one of which simulated station blackout. Nominally the presented scenario would be a guaranteed melt-down for the typical US nuclear power plant. The EBR-II design, however, managed to achieve a temperature profile shown in Figure 7.

Figure 7.

EBR-II driver temperature predicted and measured [14].

Figure 7 demonstrates that given a catastrophic failure of major safety mechanisms, including failure to SCRAM following the loss of primary reactor coolant pumps or secondary coolant pumps, the peak temperature remained well below the sodium coolant boiling temperature of 870 C. Additionally, the peak temperature only lasted tens of seconds before reducing to a temperature less than that of 100% power. The inherent properties of the reactor design drove the reactor response rather than any engineered active systems. In short, the large thermal mass of the primary coolant pool, the thermal expansion of the core upon heating and the properties of the metal fuel all worked together to cause the reactor to become subcritical before fuel damage occurred following termination of coolant pump operation even without reactor SCRAM. The current fleet of light water reactors subjected to a similar experiment would melt down without active cooling because the water coolant would eventually boil and the heat removal would be insufficient to prevent fuel melting.

Removal of the heat from the fuel elements and transporting that heat to the outside required several design layers. The first layer starts with the fuel elements, the metallic uranium, sodium bond, and stainless steel 316 cladding which provides an uninterrupted metallic conduction path from the uranium slugs to the sodium coolant. Sodium has one thousand times the heat conduction of water and in EBR-II’s design, allowed for the decay heat to be transported rapidly to the sodium pool. Figure 8 shows the uninterrupted metallic conduction path, the sodium is the green color.

Figure 8.

Thermal conduction path [15].

The second layer was the large sodium pool that could absorb a significant amount of heat without changing temperature. Even without active cooling, the natural convection of the sodium over the fuel elements was enough to circulate cool sodium in from the pool and inject hot sodium back in the pool. Given the 337,000 liters of sodium in the pool, it would take many weeks for the pool to reach a temperature where the sodium would begin to boil.

The last layer was the natural convection heat exchanger that led pool sodium to a chimney that naturally exhausted to the outside. The heat exchanger functioned solely on the temperature differential of the pool to the outside and required no external power. The natural convection heat exchanged moderated the temperature in the pool to keep the sodium from boiling away.

In summary, the solution to a run-away heating event was to increase the thermal conduction from the fuel slugs to the outside to the point where the heat generated could not exceed the bandwidth of the heat removed to the outside.

The previous sections describe how EBR-II removed the decay heat from the fuel elements, mitigating a meltdown event. This mitigation only covered long term inherent safety, not short term. Short term transients also require mitigation due to their rapid onset. Large reactivity insertions can cause localized heating that cannot be conducted away fast enough leading to fuel melting. An example is, during fuel shuffle operations, an assembly falls into the pool. Mitigation of these events (aside from not causing them in the first place) requires a negative feedback mechanism to compensate for the reactivity change. In reactors, these are called negative reactivity coefficients. They are a result of the inherent physics of a reactor’s design and are nominally passive. For example, as a legal requirement in the US, light water reactors have a negative temperature coefficient. Meaning, the hotter the fuel, the less fission occurs, thus preventing a cascade event where heating creates more fission which creates more heating. For EBR-II several of these coefficients kept the reactor in a 100% negative feedback regime.

The first of these and most effective was the expansion of the sodium inside of the core region. The liquid sodium density reduced due to thermal expansion. Given that sodium has a moderating effect on fast neutrons, the decrease in moderation led to an overall negative reactivity feedback due to sodium temperature increases. This proved invaluable in the safety heat removal tests because as the temperature increased, there was a greater the reduction in fissions.

Second, EBR-II’s core construction allowed for thermal expansion in the core. As temperature increased the fuel assemblies were pushed away from each other. The core grid plate that locked the bottom of the assemblies would expand due to temperate having the effect of increasing the pitch. Fast reactors in general are very sensitive to geometry changes due to their high-power densities. Any expansion increases the leakage of neutrons due to the increase in effective surface area with the same neutron population.

These two negative reactivities constitute 99% of the reactivity coefficients. They kept the reactor from running away in a thermal transient allowing for thermal conduction to occur. The long-term conductive mechanisms of EBR-II then kept the reactor from melting down. With these passive mechanisms in place, the severe accident scenario described in the previous section could happen without any real consequences. It was due to the inherent safety mechanisms of EBR-II that made April 3rd 1986 just another day in southeast Idaho.

4. Conclusion

EBR-II was arguably the most significant, meaningful, and successful sodium cooled fast reactor power plant demonstration in the history of nuclear power. It must be emphatically stated that the success of EBR-II was the result of actual demonstration rather than simulation and modeling or claims of future success based on short-lived small past experiments. Over a 30-year operating lifetime, the reactor demonstrated all aspects necessary for using a sodium cooled fast reactor for power production. Numerous technological advancements were made using EBR-II. Foremost among the advancements were 1) the demonstration of a pool type primary coolant arrangement with all primary piping and pumps located within the pool, 2) the ability to conduct fuel handling activities in opaque molten sodium, 3) the ability to transfer fuel into and out of the primary sodium pool while the reactor was operating at full power, 4) the ability to safely operate a system where heat is transferred from molten sodium to water, 5) the development of metallic fuel, 6) the demonstration of tremendous fuel burnup, and 7) the demonstration of compact on-site fuel reprocessing. The most significant accomplishment of EBR-II was the demonstration of the inherent safety associated with the overall reactor design and material properties that allowed the reactor to survive the most severe accident scenarios, loss of flow without SCRAM and loss of heat sink without SCRAM, with no fuel damage.

It is hoped that the success of EBR-II will not only influence the design of future sodium cooled fast reactors, but that it will be identified as an example of the true feasibility of such designs. This chapter is dedicated to the memory of Len Koch who was present for the startup of EBR-I and served as one of the principal EBR-II designers.