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By Charles Rhodes, P.Eng., Ph.D.

In 2018 most existing nuclear power reactors are water moderated. The term "water moderated" means that water flows through the core of the reactor and absorbs most of the kinetic energy from free neutrons before they are reabsorbed. Water moderation is easy to understand but it has several distinct disadvantages including high operating pressure, potential void instabilities, wasteful use of natural uranium and large scale production of long lived nuclear waste.

Liquid sodium cooled Fast Neutron Reactors (FNRs) are a more modern type of nuclear reactor that avoids these disadvantages by using liquid sodium rather than water for heat removal. Liquid sodium cooled Small Modular FNRs (SM-FNRs) offer many important performance and safety advantages as compared to existing water cooled nuclear power reactors. These advantages include:
More than 100 fold less natural uranium consumption per kWh;
More than 1000 fold less long lived nuclear waste production per kWh;
No decommissioning waste;
Low primary sodium pressure;
Higher primary sodium temperatures;
High secondary sodium pressure;
Disposal of existing spent water cooled reactor fuel;
Passive high temperature shutdown allowing safe reactor siting within major cities;
Rapid output power slew rate;
Assembly from factory built truck transportable modules;
Long operating life;
High capacity factor due to multiple independent heat/electricity output ports;

If a nuclear reactor's neutron spectrum contains primarily fast neutrons with kinetic energies of the order of 2 MeV it is known as a Fast Neutron Reactor (FNR). If a FNR is assembled from truck transportable modules it is known as a Small Modular Fast Neutron Reactor (SM-FNR).

If a FNR or a SM-FNR is cooled with liquid sodium it is known as a liquid sodium cooled FNR or liquid sodium cooled SM-FNR. All FNRs and SM-FNRs described on this web site are liquid sodium cooled. Hence on this web site the term FNR or SM-FNR implies liquid sodium cooling. The discharge temperature of the primary liquid sodium is typically 440 degrees C to 450 degrees C. If the primary sodium surface temperature reaches to 500 degrees C that surface appears faint red.

To put this primary sodium discharge temperature in perspective the melting points of some common metals are:
tin = 232 C, cadmium = 321 C, lead = 327 C and Zinc = 419.5 C.

Fast neutron reactors discussed on this web site obtain their prime energy from the abundant uranium isotope U-238. The FNR nuclear process converts U-238 into Pu-239 and then fissions the Pu-239. During each FNR fuel cycle the Pu fraction in the core fuel decreases from 20% to 12.5% by weight corresponding to 15% core fuel burnup and 7.5% conversion of U into Pu in the core fuel. Most of the fission products have short half lives. The spent fuel recycling process breeds more core fuel rod material than is consumed, allowing ongoing expansion of the FNR fleet.

A practical fast neutron reactor used for thermal or electrical power generation superficially looks like an olympic diving pool filled with liquid sodium.

There a 25.4 m long X 18.4 m wide X 13.5 m deep pool of primary liquid sodium. There is an octagonal assembly of vertical closely spaced (1 / 2) inch outside diameter closed end steel tubes known as fuel tubes 6.0 m high X 12.8 m in diameter that is centrally positioned within that liquid sodium pool. The liquid sodium surface is 1 m below the pool deck. The depth from the liquid sodium surface to the top of the steel fuel tubes is nominally 3.0 m. There is nominally about 3.0 m of liquid sodium depth below the bottom of the fuel tubes.

The pool walls are stabilized by external fill embankments. At each end of the 36 m high pool building is a separate steam generator-turbogenerator-condenser building. The two turbogenerator buildings each contain 16 X 8 MWe modular steam turbo-electric generator units. The steam condensers feed captured heat to 130 m high dry cooling towers. At maximum thermal output the air temperature in these towers is about 20 degrees C above the outside air ambient temperature. At urban reactor sites the cooling tower vents may be concealed inside 40 floor office buildings.

The fuel tubes emit heat at a variable rate that keeps the primary liquid sodium surface at 440 C to 450 C. The mechanism that controls the fuel tube heat emission rate is thermal expansion of the liquid sodium. Due to its lower density and hence buoyancy hot liquid sodium naturally rises vertically between the fuel tubes. When the upper 7 m of primary liquid sodium reach 450 C heat production stops. From a thermal engineering perspective a SM-FNR acts as a nearly constant temperature pool of primary liquid sodium. The reactor thermal power is controlled by controlling the rate at which heat is extracted from the primary liquid sodium pool.

The heat extraction system is designed to limit the maximum heat extraction rate while the reactor is operating to the maximum rated thermal output power of the fuel tubes. Otherwise the core fuel and active fuel tubes could potentially overheat and melt. Similarly during a reactor cold start the safe rate of liquid sodium pool warmup is limited by the rated thermal output power of the core fuel tubes.

When there is no external thermal load the heat extraction rate is zero and the primary sodium temperature rises to 450 degrees C which stops the nuclear chain reaction. Hence ideally with no thermal load the primary liquid sodium pool temperature stabilizes at 450 degrees C. However, there may be some temperature overshoot due to fission product decay. When there is no thermal load the fission product decay heat must be dumped via the cooling towers. This decay heat dump should occur by natural circulation without reliance on any electrically powered fans or pumps.

The top surface of the liquid sodium is covered by steel floats. The purpose of these floats is to minimze the exposed liquid sodium surface area. In the event that air leaks into the reactor argon cover gas the reduced exposed sodium surface area minimizes spontaneous sodium oxidation and the fire risk.

In the middle of the sodium pool is a grid of 532 vertical steel indicator tubes that project 0.5 m to 1.5 m above the liquid sodium surface. These indicator tubes convey important active fuel bundle control portion position, temperature and radiation level data to overhead monitoring and control instrumentation.

Immersed in the hot liquid sodium at each end of the liquid sodium pool are vertical tube single pass intermediate heat exchange bundles that are used to extract heat from the primary liquid sodium when the reactor is producing heat or are used to add heat to the primary liquid sodium during reactor cold startup or when the reactor is shut down for prolonged service. The isolated heat transfer fluid is non-radioactive secondary sodium. For safety reasons the secondary liquid sodium pressure is kept higher than both the primary liquid sodium pressure and the thermal load (steam) pressure. This heat transfer arrangement is safe in the presence of both intermediate heat exchange bundle failures and steam generator heat exchange bundle failures.

There is an overhead gantry crane that is used to add, reposition or remove fuel tube bundles and to add or remove intermediate heat exchange tube bundles.

There are an air-vacuum-argon locks at each end of the reactor building that are used to bring equipment into the reactor building or remove equipment from the reactor building without loss of argon and without mixing air and argon.

There are also rotating door type airlocks for personnel entry and exit. These personnel air locks use oxygen absorbing chemicals to minimize the amount of oxygen that enters the reactor space.

The argon cover gas above the primary sodium is at atmospheric pressure. Its normal operating temperature is about 450 degrees C. When the reactor is shut down for service this temperature can be lowered to about 120 degrees C provided that sufficient additional argon molecules are added to the cover gas gas in the reactor building to maintain the 1 atmosphere argon pressure so as to prevent a wall or overhead ceiling collapse. Each FNR has two redundant liquid argon storage, release and recovery facilities.

When the reactor is operating a small fraction of the liquid sodium (Na-23) absorbs neutrons and forms Na-24. The Na-24 naturally decays by electron emission with a half life of 15 hours to become stable Mg-24. In the decay process a 1.389 MeV gamma photon is emitted. After a reactor shutdown it takes about a week for this radiation to drop by a factor of 2000. If any of the fuel tubes are leaking or if the primary sodium purity or filtering is inadequate there could be other radioactive emissions. For certainty with respect to radiation safety, reactor security and sodium containment a FNR is completely surrounded by a concrete shell about 1 m thick. The impurity concentration in the primary liquid sodium is less than one part in 10^5.

The fuel tube assembly is completely surrounded by a 2.8 m thick jacket of liquid sodium. This sodiuum jacket absorbs all neutrons that escape from the fuel tubes. Hence no neutrons reach the primary sodium pool containment walls, the intermediate heat exchange fuel bundles, the primary sodium pool floor or overhead structures. This prevention of neutron activation and neutron damage outside the fuel bundle assembly enables a very long facility working life and prevents formation of decommissioning waste. Absence of neutron activation also minimizes the complications involved in intermediate heat exchange bundle replacement.

In FNR core fuel the trans-uranium actinides, instead of simply capturing neutrons as in a water moderated reactor, preferentially fission. With appropriate periodic fuel reprocessing a FNR yields at least 100 fold more energy per kg of natural uranium than does a heavy water moderated CANDU reactor. During fuel reprocessing the fission products are extracted from the balance of the fuel. About 95% of the extracted fission products decay to safe levels in 300 years. Hence on a per kWh basis the rate of FNR spent fuel long term waste production is about 2000 fold less than for a CANDU reactor.

The best method of CANDU reactor spent fuel disposal is to reprocess the CANDU spent fuel into new FNR fuel and then consume it in a FNR.

Another important feature of FNRs is efficient load following. Unlike the almost fixed thermal output of a water moderated nuclear reactor the thermal output of a FNR can be rapidly increased or decreased to follow electricity grid net load changes arising from rapid variations in load and unconstrained renewable generation.

In summary, liquid sodium cooled power FNRs can provide sufficient energy to sustainably displace fossil fuels with almost no production of long lived nuclear waste. FNRs can also be used to safely dispose of spent fuel from CANDU and other water moderated nuclear reactors.

When a Pu-239 atom absorbs a fast neutron and fissions on average it emits 3.1 fast neutrons. If the probability of other Pu-239 atoms absorbing these fast neutrons and fissioning is greater than (1 / 3.1) then a rapid nuclear chain reaction will occur liberating large numbers of neutrons and a large amount of thermal energy. Otherwise the thermal energy emission by spontaneous fissioning of Pu-239 atoms is extremely small.

A Fast Neutron Reactor (FNR) consists of a pancake shaped core zone about 0.35 m thick containing uniformly distributed U-238 - Pu-239 - Zr alloy core fuel rods sandwitched between two 1.2 m thick blanket zones containing uniformly distributed U-238 - Zr alloy blanket fuel rods. The Pu-239 in the core zone fissions and emits fast neutrons. The U-238 in the core and blanket zones absorbs excess fast neutrons and makes more Pu-239. There is an ongoing flux of fast neutrons flowing from the core zone into the blanket zones. Hence there is ongoing production of Pu-239 in the blanket zones.

The Pu-239 concentration in the core zone, the core zone thickness and the blanket zone thicknesses are chosen so that slightly more than (1 / 3.1) of the neutrons emitted by fission of Pu-239 atoms in the core zone are captured by other Pu-239 atoms in the core zone. Hence a chain reaction occurs because the probability of an emitted neutron being captured by a Pu-239 atom is slightly greater than (1 / 3.1).

In the core zone the rate of loss of Pu-239 by fissioning is partially offset by the rate of production of Pu-239 via neutron capture by U-238. Reactor criticality at the desired operating temperature is maintained through the operating life of fuel bundles via small controlled changes in fuel bundle geometry. These geometry changes are accomplished by using a hydraulic actuator to change the vertical overlap between each active fuel bundle control portion and its corresponding surround portion.

If due to nuclear heat release the temperature of the materials increases thermal expansion of the materials in three dimensions causes the fraction of fission neutrons diffusing from the core zone into the blanket zones to increase, decreasing the probability of Pu-239 atoms in the core zone capturing fission neutrons. Hence the nuclear chain reaction stops.

Liquid sodium has an unusually high thermal coefficient of expansion (TCE) which enhances this effect. An increase in primary liquid sodium temperature above the chosen temperature setpoint will turn the chain reaction off. A subsequent decrease in primary liquid sodium temperature below that temperature setpoint will increase the primary liquid sodium density causing restart of the nuclear chain reaction. The TCE of the other reactor core materials is smaller than for sodium but further contributes to this effect. Hence, a liquid sodium pool type nuclear reactor automatically maintains the pool of liquid sodium at the chosen temperature setpoint.

Each active fuel bundle has a vertically sliding portion known as the control portion with hydraulic insertion/withdrawal actuator that provides control of the fuel bundle's primary liquid sodium discharge temperature set point.

The reactor thermal power is set by the rate of extraction of heat from the primary liqud sodium pool. This heat extraction rate is a function of the difference between the secondary liquid sodium temperature and the thermal load temperature as well as the secondary liquid sodium flow rate. The rate of heat extraction must be kept within fuel tube design limits. Otherwise the reactor fuel tubes could overheat.

The pressure in each steam generator is controlled by a motorized steam discharge valve which maintains a constant pressure (11.2 MPa) in the steam generator. That pressure, via the pressure-temperature relationship for saturated steam, determines the water temperature in the steam generator (320 C). The difference between the liquid sodium primary discharge temperature and the water temperature in the steam generator, less two heat exchange wall temperature drops, determines the change in temperature across the secondary sodium loop. Thus controlling the secondary liquid sodium flow rate controls the amount of steam delivered to the correspnding turbo-generator.

If there is a step change in FNR fuel bundle geometry the fission rate and hence the fission gamma photon flux and the prompt neutron flux respond almost instantly. However, the liquid sodium temperature, which limits the fission power, takes longer to respond.

To prevent uncontrolled explosive power growth FNRs must always remain subcritical with respect to prompt fission neutrons, which constitute about 99.8% of the total neutron flux. The remaining 0.2% of the total neutron flux consists of delayed neutrons emitted by fission fragments approximately 3 seconds after the corresponding nuclear fission. Provided that most of the delayed neutrons participate in reactor power control, the rate of fission power growth is safely limited by the rate of delayed neutron production. This time delay in reactor power growth allows sufficient time for the liquid sodium temperature to rise and suppress the core reactivity to safely control the fission power in a FNR.

There is a requirement that the insertion rate of the active fuel bundle control portion used to ajust each active fuel bundle's discharge temperature be sufficiently low to prevent the FNR becoming critical on prompt neutrons. This insertion rate limit is achieved via important flow orifices on the hydraulic cylinder positioning valves. Hence the insertion of the active fuel bundle control portion into the active fuel bundle surround portion is very slow and is carefully controlled. By contrast on an emergency shutdown the withdrawal rate of active fuel bundle control portion is very fast. To enable fast control portion withdrawal a parallel connected full port hydraulic cylinder drain valve is used to achieve rapid fuel bundle shutdown. These valves must be rated for continuous use with liquid sodium at the highest possible liquid sodium operating temperature. To achieve the required temperature isolation these valves should be argon pressure actuated. On loss of argon pressure gravity should cause the full port valve to open. The argon pressure is controlled by an electrical transducer located in a cool environment.

Liquid sodium coolant enters the bottom of a FNR fuel tube bundles at about 340 C, flows upwards through many flow channels between the active fuel tubes, and at full load emerges from the top of the active tube bundles at 440 C. The liquid sodium discharge temperature setpoint of each active fuel bundle is controlled by the amount of active fuel bundle control portion insertion into the active fuel bundle surround portion. Withdrawing an active fuel bundle control portion from its active fuel bundle surround portion reduces the active fuel bundle discharge temperature setpoint.

The lowest density hot liquid sodium rises to the top surface of the liquid sodium pool, flows across the top of the liquid sodium pool and at the pool ends is cooled by the intermediate heat exchange bundles, causing the circulating primary sodium to increase in density and sink. The cooler higher density primary liquid sodium flows along the bottom of the primary liquid sodium pool to a point underneath the reactor fuel tube bundle and then rises again through the reactor fuel tube bundle.

The FNR geometry is chosen so that about (1 / 3) of the fission neutrons are absorbed by plutonium in the core zones to sustain the nuclear chain reaction and the remaining (2 / 3) of the fission neutrons are absorbed by depleted U-238 located in both the core and blanket zones for the purpose of breeding more Pu-239 for future use. A small fraction of the fission neutrons are absorbed by steel in the fuel tube assembly and by liquid sodium.

As the fuel bundle temperature rises to its setpoint the sodium density decreases allowing a larger fraction of fission neutrons to escape from the core zone into the adjacent blanket zones, which locally turns off the nuclear chain reaction.

Similarly, as the fuel bundle temperature falls below its setpoint the sodium density increases confining a larger fraction of the emitted neutrons in the core zone, which turns on the nuclear chain reaction.

If the reactor's external heat load is less than the reactor thermal power output the excess heat is absorbed by the liquid sodium causing the liquid sodium temperature to increase, turning off the chain reaction and hence causing the reactor's thermal power output to drop to zero.

If an external heat load removes heat from the liquid sodium the liquid sodium temperature decreases and the nuclear chain reaction restarts.

Thus when the active fuel bundle control portions are correctly positioned the FNR thermal power output automatically tracks the external thermal load.

As FNR fuel ages its plutonium concentration slowly decreases, its concentration of neutron absorbing fission products slowly increases and due to fuel tube swelling the liquid sodium flow rate through the reactor slowly decreases. Compensation for these long term changes is achieved by fine adjustment of the positions of the active fuel bundle control portions with respect to the surround portions to maintain a full load fuel bundle discharge temperature of 440 degrees C.

Full withdrawal of the active fuel bundle control portions causes a total shutdown of fission chain reactions regardless of the liquid sodium temperature.

Almost all of the neutrons that are not consumed by the fission chain reaction convert U-238 to U-239 which via two spontaneous electron emissions soon converts to Pu-239. Periodically the reactor fuel rods from both the core and blanket zones are reprocessed to extract fission products and to move newly formed Pu-239 from blanket fuel rods into core fuel rods.

The fission product mass extracted from core fuel rods is replaced by an equal mass of core rod alloy obtained by reprocessing blanket rod material. The consumed blanket rod material is replaced by an equal mass of new depleted U-238. The extracted fission products should be safely stored in isolation for about 300 years to allow their radio toxicity to naturally decay down to the level of natural uranium.

The reactor fuel tube assembly consists of many thousands of vertical 0.5 inch OD steel tubes X .065 inch wall, 6.0 m high that form a square lattice spaced (5/8) inch center to center. These steel fuel tubes are closed at both ends and contain the reactor core and blanket rods as well as internal liquid sodium to enhance thermal contact between the fuel rods and the steel tubes and to chemically absorb corrosive fission product gases such as F, Cl, Br and I. For structural and transportation purposes the steel fuel tubes are assembled into square tube bundles.

Each fuel tube has a 3.2 m high plenum which stores inert gas fission products and spare internal liquid sodium and also serves as a chimney to enhance primary liquid sodium natural circulation.

Each active tube bundle, including its shroud, is nominally 15.50 inches face to face X 20 feet high and contains 476 X 0.5 inch OD HT-9 steel tubes. The tube to tube spacing is maintained by horizontal (1 / 16) inch diameter steel rods. The steel tube lattice bottom spacing is fixed by the fuel bundle bottom gratings which support and position the fuel tubes and permit vertical liquid sodium coolant flow.

Each active 0.5 inch OD steel fuel tube contains 4 X .600 m long blanket (B) rods and 1 X .35 m long core (C) rod. Each passive fuel tube contains 5 X .600 m long blanket (B) rods. For the 532 active bundles the fuel rod types from stack bottom to stack top are B, B, C, B, B. For the 272 perimeter passive tube bundles the fuel rod types from stack bottom to stack top are: B, B, B, B, B.

During prolonged reactor operation the core fuel rods swell from 0.35 m long to about 0.400 m long. Each fuel tube contains a measured amount of liquid sodium. The top 3.2 m of each steel fuel tube are nominally empty to provide sufficient plenum volume to relieve pressure stress and to store sufficient sodium to compensate for fuel tube material swelling.

Each fuel tube bundle is structurally supported by a square steel support pipe. The vertical insertion/withdrawal of each active fuel bundle control portion is controlled by liquid sodium pressure applied to a piston type hydraulic actuator located inside the square support pipe. The control portion's vertical position is indicated by a vertical indicator tube attached to the top of the control portion. This indicator tube projects above the surface of the liquid sodium. At the bottom inside of the indicator tube is a mercury pool that maintains a mercury vapor pressure inside the indicator tube corresponding to the liquid sodium temperature at the bottom of the indicator tube. This temperature is the fuel bundle's liquid sodium discharge temperature.

This temperature is indicated by a coiled Bourdon tube attached to the top of the indicator tube. The Bourdon tube moves a mirror that changes the angle of reflection of a scanning laser beam.

The core fuel rods are initially by weight: 10% zirconium; 20% plutonium, U-235 and fissionable transuranium actinides; and 70% U-238.

The blanket rods are initially by weight: 10% zirconium and 90% U-238.

The purpose of the zirconium in both the core and blanket rods is to prevent plutonium from forming a low melting point eutectic with the steel fuel tube material.

The square fuel bundle support pipe is welded to the steel frame on the bottom of the liquid sodium pool. This steel frame position stabilizes the positions of the bottoms of the fuel bundles. Hydraulic pressure lines routed through the frame and the support pipes provide the controlled liquid sodium pressure that raises or lowers each active fuel bundle control portion.

Heat is removed from the fuel tube assembly by primary liquid sodium which flows upwards via natural convection through the fuel tube support gratings and then up through the gaps between the HT-9 steel fuel tubes. The support gratings are fitted with bottom filters to trap any particles with dimensions over (1 / 32) inch.

It is necessary to maintain the FNR near the threshold of fission criticality. This constraint in combination with the fuel geometry sets the nominal core zone height with new fuel at 0.35 m. As the fuel ages the required core zone height gradually increases to about 0.40 m.

The reactor core zone maximum outside diameter is a function of the liquid sodium pool width. That width is constrained by practical structural issues related to reactor roof construction. With a 9.6 m diameter reactor core the sodium pool width is about 18.4 m . It is necessary to have a roof covered 5 m wide perimeter strip around the liquid sodium pool for the insulating lava rock, access and air cooling, and gamma ray absorbing concrete.

The total core zone height and fuel tube assembly diameter establish the active heat transfer area of the fuel tubes.

There is also a heat transfer limit relating to the rate at which liquid sodium will naturally circulate between the fuel tubes.

There are other practical limitations related to modular component transport that also come into play at a FNR core zone diameter of over 9.6 m. At a smaller FNR core zone diameter the economies of scale related to the required liquid sodium pool volume are lost. At larger core zone diameters the FNR roof cost quickly rises.

During reactor operation the reactivity of the core zone is at its critical point so there is ongoing fission in the core zone and the core zone acts as a source of neutrons. During reactor operation the blanket zones act as net neutron sinks.

As the reactor runs there is a gradual accumulation of fission products, primarily in the core fuel rods. Neutrons emitted by fission of Pu-239 are absorbed by U-238 in both the core and the blanket rods. The resulting U-239 spontaneously converts via Np-239 into Pu-239 within about one week. Further neutron absorption by Pu-239 that does not fission causes formation of Pu-240. This Pu-240 spontaneouly decays in a manner which prevents FNR fuel from being used for atom bomb production.

After the reactor has run for some time (~ 40 years) the accumulated fission products in the core fuel rods and the decrease in Pu-239 concentration reduce the core zones' reactivity and the fuel tubes may swell reducing the primary liquid sodium flow. The fuel tube bundle is then moved to the perimeter zone of the liquid sodium pool to allow dissipation of short lived fission product decay heat. Over a fuel cycle period of about 40 years the fuel and blanket bundles are gradually replaced and the fuel bundle material is reprocessed and recycled.

After fission product decay heat dissipation the tube bundles are removed from the liquid sodium pool and transported to a reprocessing facility where the fuel tubes are extracted and disassembled. The fuel tube material, the core rods, the blanket rods, the liquid sodium and the sodium salts are each reprocessed differently. The fuel reprocessing involves selective removal of uranium, selective removal of fission products, selective separation of zirconium and reforming the remaining fuel rod residue into new core rods. The selectively removed uranium plus some new U-238 is formed into new blanket rods. New steel tube bundles are assembled and the entire FNR fuel cycle process is repeated. The steel tubes are formed from an iron-chromium alloy known as HT-9 that has a low nickel and low carbon content. The fuel tube material recycling process involves selective titanium and chromium extraction.

Due to neutron activation of the materials the entire fuel recycling process is carried out by robotic equipment.

The reprocessing of the used fuel rods and blanket rods yields more Pu-239 and trans-uranium actinides than the FNR needs. Hence the excess Pu-239 and actinide inventory can be accumulated and used to start another FNR.

A major constraint on the rate of implementation of liquid sodium cooled FNRs is the available supply of FNR start fuel. This start fuel may be obtained from plutonium and other actinides in spent water cooled reactor fuel and by plutonium breeding by FNRs. It may take over 70 years for one FNR to breed enough surplus plutonium to start another identical FNR. Thus in the near term the maximum rate of FNR deployment will be primarily a function of the available supply of spent water cooled reactor fuel. It appears that due to the FNR start fuel constraint the world will have to live with a mixed fleet of both water cooled reactors and FNRs for at least the next century.

In liquid sodium cooled Fast Neutron Reactors (FNRs) the thermal power output is proportional to the fast neutron flux incident upon the fuel, as compared to water moderated thermal neutron nuclear reactors in which the thermal power output is proportional to the slow neutron flux incident upon the fuel. Fast neutron reactors are characterized by rapid fuel bundle thermal power output changes near each fuel bundle's discharge temperature set point. For safe power control Fast Neutron Reactors rely on thermal expansion to reduce each fuel bundle's reactivity as its temperature increases. The vertically sliding active fuel bundle control portion positions are adjusted so that all the active fuel bundles operate at the same discharge temperature. These vertically sliding active fuel bundle control portions are also used to achieve a reactor cold shutdown.

One of the issues in FNR design is ensuring that no matter what adverse circumstances occur in an emergency gravity will cause the active fuel bundle control portions to fall into a safe cold shutdown position.

The major advantages of liquid sodium cooled fast neutron reactors (FNRs) over CANDU reactors are:
1. The argon cover gas above the FNR liquid sodium surface is at atmospheric pressure. Apart from the fuel tubes components subject to pressure stress are not subject to excitation or material degradation due to the neutron flux.

2. High pressure steam and hydrogen can be safely vented to the atmosphere because they never contain radioactive isotopes.

3. The FNR radio isotope containment system never has to deal with high steam pressures, condensation or hydrogen production.

4. FNRs yield about 100 fold more energy per kg of mined natural uranium;

5. FNRs reduce long lived nuclear waste production about 1000 fold as compared to a CANDU reactor;

6. FNRs can easily track rapid changes in grid load:
7. The primary liquid sodium coolant in a FNR typically runs at 340 degrees C to 440 degrees C as compared to the 260 degree C to 300 degree C primary coolant temperature in a water moderated reactor. The higher primary coolant temperature allows efficient use of steam turbines without direct lake water cooling;
8.The greatly reduced cooling water requirement of an FNR reduces its impact on marine ecology.
9. Due to the reduced requirement for cooling water a FNR can be sited much further above the local water table and surrounding bodies of water than a direct water cooled reactor, thus enhancing system safety in rare but severe events such as floods, earthquakes, hurricanes, meteorite strikes, tsunamis, etc.
10. The primary liquid sodium coolant in a FNR operates at a low pressure rather than a high pressure, which simplifies many reactor design, construction, operation and maintenance issues;
11. In a liquid sodium cooled FNR the structural components such as the sodium pool walls and the heat exchangers are not exposed to the neutron flux, enabling a facility service life of hundreds of years rather than just 60 years as with a CANDU reactor. The secondary sodium flow velocity within the intermediate heat exchangers is chosen to minimize heat exchange tube internal erosion while maintaining a sufficiently turbulent liquid sodium flow for good heat transfer.
12. The long reactor service life and the liquid sodium guard band minimize the rate of formation of radioactive decommissioning waste.
13. The intermediate heat exchange bundles, which are subject to internal pressure stress and internal erosion, are not exposed to the neutron flux. Hence the service life of the heat exchange bundles is enhanced and when they do need replacing a simple exterior surface cleaning will remove all radio activity allowing economic material recycling. Isolating the intermediate heat exchangers from the neutron flux enables use of a high performance nickel alloy (Inconel 600) in the intermediate heat exchanger;
14. The steel components of a FNR that are exposed to a high neutron flux are replaced and recycled at the same time as the reactor fuel. Part of the iron is transmuted into chromium.
15. A FNR should be designed with multiple independent secondary heat transport systems so that there is absolute certainty relating to removal of decay heat by natural circulation after fission reaction shutdown;
16. Unlike water cooled and moderated reactors a liquid sodium cooled FNR does not produce high pressure radioactive steam. Hence a FNR does not need a pressure rated enclosure for radioactive steam containment. If the pressure in the steam generator gets too high the steam can be safely vented to the atmosphere because it is not radio active. The liquid sodium pool must enormously over heat before the sodium vapor pressure becomes structurally significant.
17. A major non-obvious advantage of FNRs is a nearly infinite fuel supply. Since a FNR consumes natural uranium at less than 1% of the consumption rate of a CANDU reactor an FNR can economically utilize natural uranium resouces that have very low uranium concentrations. That feature greatly increases the total available natural uranium resource.
18. The scientific issues related to FNRs have been well understood since the late 1960s. The practical metallurgy issues were resolved by about 1990.
19. FNRs can be assembled from modules. All the modules are replaceable. There is no practical limit to the working life of a FNR facility.

A disadvantage of a liquid sodium cooled FNR is that the sodium reacts violently on contact with water and above 200 degrees C sodium burns in air. The liquid sodium pool requires both a floating steel cover and an argon gas cover to be safe. The biggest issues in safe siting and operation of liquid sodium cooled FNRs are certain exclusion of water and continuous maintenance of an argon gas cover.

Local fire department personnel must be trained to NEVER use water to fight a fire in a FNR facility.

A significant security issue with FNRs is that they operate with a large Pu-239 inventory. During FNR operation first-in first-out fuel bundle replacement should be followed to maintain a sufficient Pu-240 concentration in the fuel that the Pu-239 cannot be selectively chemically extracted and used to fabricate fission bombs.

The object is to implement a nuclear reactor fuel cycle that will simultaneously achieve several important objectives:
1. Optimally utilize the inventory of spent water cooled reactor fuel to start future FNRs;
2. Achieve a major increase in energy yield by breeding U-238 in the spent CANDU fuel bundles into Pu-239 and fissionable trans-uranium actinides and then fissioning these actinides;
3. Effectively utilize existing CANDU reactors for interim on-going electricity generation;
4. Achieve a fuel cycle supervision, transportation and storage arrangement that will prevent nuclear weapon proliferation;
5. Achieve a fast neutron reactor fuel composition that meets the safety requirements.

The concept of recycling spent CANDU fuel through a fast neutron reactor was originally developed by Professor Peter Ottensmeyer and a group of students at the University of Toronto. The process relies on a chemical/mechanical process that first selectively extracts UO2 from spent CANDU fuel and then makes the residue metallic and separates low atomic weight atoms (fission products) from high atomic weight atoms (remaining uranium, plutonium and trans-uranium actinides).

The low atomic weight atoms are placed in isolated storage where over 300 years their radio toxicity naturally decays to the level of natural uranium. Then, subject to selective Se-79 and Sn-126 chemical extraction, this low atomic weight waste can either be recycled or buried in existing depleted uranium mines.

The weight of the fission product atoms removed from the core fuel is replaced by an equal weight of transuranium actinides extracted from the spent CANDU fuel inventory. New core rods and blanket rods are fabricated and then run through a fast neutron reactor.

This fuel cycle is repeated over and over again until the entire inventory of spent CANDU fuel is exhausted. It is estimated that during each fuel cycle about 15% of the core fuel rod weight will be converted from high atomic weight atoms to low atomic weight atoms. It is estimated that the Ottensmeyer Plan realizes more than 100 fold more energy per kg of mined natural uranium than does the present CANDU fuel cycle that operates with natural uranium in a slow neutron environment with no fuel recycling. Hence the existing inventory of spent CANDU fuel may be sufficient to power Canadian Fast Neutron Reactors for centuries to come.

During each fuel cycle sufficient Pu-239 is produced to sustain the reactor reactivity during the following fuel cycle and to provide yet more Pu-239 for starting other FNRs. A storage facility comparable to the Jersey-Emerald mine is needed to securely store the fission products for about three centuries and to store the small fraction of long lived nuclear waste for about one million years.

Obstacles to immediate implementation of FNRs are political willingness to accept transportation, storage and processing of material derived from spent CANDU nuclear fuel bundles. The spent fuel, instead of gradually diminishing in radioactivity as the years pass, would be chemically and mechanically reprocessed and reused about every thirty years. Hence on-going access to a secure naturally dry spent fuel storage facility, such as the Jersey-Emerald mine complex in British Columbia, is an important part of FNR implementation. It is contemplated that the initial FNR fuel reprocessing site would be at Chalk River, Ontario, which is far from any urban center. Ideally the fission product storage facility should be located close to the fuel rod reprocessing facility. One of the lessons to be learned from experience in France is that if the facilities are not properly located highly radioactive materials wind up being transported to and fro across the country.

Another potential site for a fuel rod processing facility is Trail, British Columbia where the local population has for generations been accustomed to large scale management of highly toxic materials.

Once all of the reprocessing issues related to the spent fuel bundles have been resolved there is no obvious reason why analogous techniques could not also be applied to recycling the medium level nuclear waste that OPG currently contemplates burying in yet another deep geological repository.

A blunt reality that humans must face is that fossil hydrocarbons must be left in the ground. Sustainable replacement of fossil hydrocarbons requires fast neutron power reactors. Fast neutron power reactors require about 20% Pu in their core fuel rods to operate. Hence any treaty, legislation or regulation that only permits lower fractions of Pu in nuclear fuel is not compatible.

The important issue in prevention of nuclear non-proliferation is maintianing a sufficient Pu-240 to Pu-239 ratio in the fuel to prevent the plutonium being suitable for making fission bombs. This ratio is maintained by doing all necessary to ensure that FNR fuel bundles are reprocessed in a first in-first out sequence.

In the USA a 20 MWe fully functional prototype liquid sodium cooled FNR known as the EBR-2 was built and successfully operated from about 1964 to 1994. Under the Bill Clinton administration the USA took a huge step backwards when it cancelled funding of its fast neutron reactor program.

In Russia a 600 MWe fully functional prototype liquid sodium cooled FNR known as the BN600 was built and successfully operated from about 1984 to 2016. See 600 MWe LIQUID SODIUM COOLED POWER REACTOR

This web page last updated March 9, 2018

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