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FAST NEUTRON REACTOR DESCRIPTION:
A typical Fast Neutron Reactor (FNR) is an enclosed cylindrical pool of liquid sodium 20.0 m in diameter X 15.0 m deep. Immersed in middle of this pool are about 1300 bundles of vertical 1.27 cm OD X 6 m steel fuel tubes that, via a passive nuclear process with no moving parts, maintain the liquid sodium pool surface temperature at a constant 502 degrees C.
Immersed in the liquid sodium around the pool perimeter are heat exchange bundles which extract heat from the liquid sodium pool to make steam for production of electricity and supply of district heat.
When there is no thermal load the liquid sodium pool surface temperature remains at 502 degrees C.
The fuel and heat exchange bundles used in a FNR are road truck portable. In principle a FNR can be sized to meet almost any desired thermal load, although for economic reasons some reactor sizes make more sense than others. If FNRs are made too small economies of scale are lost. If FNRs are made too large economies related to energy transmission and enclosure size are lost.
For electricity supply reliability it is usually better for a utility to have multiple smaller power reactors than one large power reactor. A FNR is usually assembled from truck portable modules.
Please go to for additional FNR description.
FAST NEUTRON REACTOR (FNR) HIGHLIGHTS:
FNRs provide the only economic and sustainable zero CO2 emitting means of fully displacing fossil fuels from global energy production, especially in the circumpolar countries.
It is not possible to fully displace fossil fuels with renewable energy. Most of the available economic hydroelectric power is already harnessed. There is no economic means of generating, transmitting and storing sufficient intermittent wind and solar energy to reliably meet electric power requirements.
In the circumpolar countries such as Canada and Russia there is insufficient sunlight in the winter and there are protracted periods of low wind. Absent abundant local hydro-electric energy storage there is no other non-fossil energy storage technology that can economically meet power and energy requirements during extended low wind and low sun periods.
In the circumpolar countries heating is one of the largest energy usage categories. Efficiently meeting the heat load with non-fossil energy requires district heating systems that are fed by Small Modular nuclear Reactors (SMRs). These SMRs should be geographically distributed within urban centers to economically supply both heat and electricity.
The history of a successful Small Fast Neutron Reactor in the USA is well summarized by the video:
The Nuclear Option. The public should realize that since about 1980 the fossil fuel industry has spent billions of dollars on publicity campaigns and government lobbying aimed at preserving fossil fuel market share by preventing wider deployment of nuclear energy. In spite of a few highly publicized but relatively minor accidents, nuclear energy has shown itself to be by far the safest and least expensive means of reliable non-fossil bulk energy production.
In the USA there is gradual realization that the US government policy has been politically dominated by fossil fuel industry influence to the detriment of the environment.
In 2019 most existing nuclear power reactors are water moderated. The term "water moderated" means that the core of the reactor contains water which absorbs most of the kinetic energy from energetic neutrons emitted during the fission process before these neutrons are reabsorbed by the fuel. Water moderated power reactors were initially developed for military submarine applications and are primarily fuelled by the relatively rare uranium isotope U-235. However, the mineable U-235 resource is insufficient to permit sustained full displacement of fossil fuels using water moderated reactors.
Water moderation has several other distinct disadvantages for land based electric power generation including: high fluid operating pressure, essential continuous mechanical coolant pumping, potential void related power instabilities, wasteful use of natural uranium, large scale production of long lived nuclear waste and relatively frequent shutdowns for refuelling.
Modern liquid sodium cooled Fast Neutron Reactors (FNRs) avoid these disadvantages by using naturally circulated liquid sodium rather than pumped water for reactor core heat removal. In nuclear applications the combined chemical and nuclear properties of sodium have many advantages. However, the chemical incompatibility of sodium and water make liquid sodium cooled nuclear reactors unsuitable for marine applications.
A major issue with FNRs is full automation of fuel bundle production and fuel reprocessing. A typical power FNR has about 500,000 fuel tubes containing about 3,000,000 fuel rods. In order to make FNRs economic all aspects of FNR fuel bundle production and fuel recycling must be fully automated. This automation requires a major capital commitment.
A major advantage of Small Modular liquid sodium cooled Fast Neutron Reactors (SM-FNRs) is that they can be designed for safe unattended operation within a city. If there is loss of control power, loss of thermal or electrical load or loss of intermediate coolant pumping the reactor defaults to a safe shutdown state with fisson product decay heat removal by natural convection.
FAST NEUTRON REACTOR DEFINITIONS:
If a nuclear reactor's neutron spectrum in its core zone contains mainly neutrons with kinetic energies of the order of 2 MeV it is known as a Fast Neutron Reactor (FNR). If a FNR is cooled with liquid sodium it is known as a liquid sodium cooled FNR. A Small Modular FNR (SM-FNR) is a FNR that is assembled from a collection of factory fabricated and tested modules, each of which is truck portable.
The FNR described on this web site is liquid sodium cooled and is assembled from road truck portable modules. Hence on this web site the term FNR implies both liquid sodium cooling and road truck portable modules.
The design discharge temperature of the primary liquid sodium in a FNR is 502 degrees C. When the primary sodium surface temperature reaches 500 degrees C that surface glows faint red. To put this temperature in perspective the melting points of some common metals are:
sodium = 98 C, tin = 232 C, cadmium = 321 C, lead = 327 C, zinc = 419.5 C, iron = 1538 C and chromium = 1907 C.
Fuel tube properties indicate that in principle FNRs can be operated at 650 degrees C. Advantages of higher temperature operation include increased efficiency in electricity generation and reduced embrittlement of fuel tubes. However, at higher temperatures there may be significant material problems related to plutonium in the core fuel rod alloy melting at 640 deg C and chemically interacting with iron in the fuel tube alloy. A thermal analysis which takes into account the temperature drop across the core fuel radius indicates that 502 degrees C is a more prudent liquid sodium working temperature.
The FNRs contemplated on this web site have both core and blanket zones. The core zone is where the nuclear chain reaction takes place. The blanket zone surrounds the core zone and captures excess neutrons emitted by the chain reaction in the core zone. These excess neutrons react with the blanket fuel rod material and via spontaneous nuclear decay reactions form Pu-239, Pu-240, U-233 or H-3. The fissionable isotopes are recovered via fuel rod reprocessing and are used to make new FNR core fuel rod material. The H-3 and its decay product He-3 are recovered for use as fusion fuels. The near term importance of fusion is that it can potentially be used to accelerate the rate of Pu-239 inventory growth required to expand the FNR reactor fleet.
FNR POWER CONTROL:
The core zone of a FNR has a pancake geometry which gives it a large [(surface area) / (volume)] ratio. In a FNR core zone a major cause of neutron loss is neutron diffusion out of the core zone. The neutron diffusion loss rate increases with core zone temperature. In a FNR operating at its setpoint temperature the neutron production rate precisely balances the neutron loss rate. If the core zone temperature drops the fission reaction rate increases to restore the core zone temperature to its set point. If the core zone temperature is above its set point fission reaction rate decreases. Hence a FNR acts to maintain its core zone temperature setpoint.
As the FNR fuel ages the inner core zone thickness is gradually increased to maintain the desired core zone setpoint temperature. This inner core zone thickness increase is realized by increasing the insertion of the mobile square active fuel bundles into a matrix of fixed octagonal active fuel bundles.
Thermal power control in a FNR is realized by use of a lower temperature heat sink and by controlling the flow rate of intermediate liquid sodium flow between the high reactor temperature and the low load temperature. Usually the load is a steam generator. A liquid sodium recirculation pump serving the lower portion of each steam generator is required to reduce thermal stress on the steam generator heat exchange tubes that are immersed in super heated water.
As compared to water moderated reactors properly designed liquid sodium cooled FNRs offer many important performance and safety advantages in land based power reactor applications. 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;
Passive high temperature chain reaction shutdown allowing safe reactor siting within major cities;
Passive cold shutdown on loss of control power allowing safe reactor siting within major cities;
Two independent systems for cold shutdown in response to an equipment fault;
Natural circulation of secondary heat transport fluid sufficient for removal of fission product decay heat;
Redundant natural draft dry cooling towers for certain rejection of fission product decay heat;
High output power ramp rate;
Module delivery schedule certainty;
Rapid assembly from factory built and tested truck transportable modules;
Rapid repair by module replacement;
Long operating life;
Infrequent shutdowns for refuelling;
High capacity factor due to multiple independent heat transport systems and electricity generation systems;
Disposal of existing spent water cooled reactor fuel;
Breeding extra core start fuel for other FNRs:
Potential for forming fission fuel U-233 from thorium and fusion fuel H-3 from Li-6.
FNR DESIGN CONCEPTS:
FNR design concepts have been tested in small research reactors. However, the power output of these research reactors is small as compared to full size modern electricity utility power reactors. The design concepts of a FNR are significantly different from a water moderated nuclear power reactor. These design concepts are reviewed below:
1) There is a large swimming pool size double wall primary coolant containment tank operating at atmospheric pressure.
2) For safety there are no penetrations through the tank bottom or the vertical tank side walls;
3) The primary coolant naturally density stratifies so that the hottest coolant is at the top;
4) A FNR contains mobile square active fuel bundles that actuators move in and out of the matrix of fixed octagonal fuel bundles to set the reactor core zone operating temperature;
5) Natural circulation causes the primary coolant to circulate;
6) In a FNR fast neutrons travel much further through the primary coolant than they travel through water before they lose their kinetic energy. Hence the walls and floor of the primary coolant pool are protected from neutron damage. The protective barrier, in addition to absorbing neutrons, provides thermal mass that limits the overall rate of change of primary coolant pool temperature;
7) The positions of the mobile active fuel bundles should be set to keep the discharge temperature uniform across the reactor core zone;
8) As the primary coolant warms up and thermally expands the FNR reactivity decreases, which reduces the thermal power output. Similarly when the primary coolant cools and contracts the reactor thermal power increases. Hence, in normal FNR operation when the primary coolant temperature is high the reactor thermal power output is low and as the primary coolant temperature decreases the reactor thermal power output increases;
9) A FNR automatically seeks an equilibrium temperature at which the rate of heat generation equals the rate of heat removal.
10) As long as the thermal load is sufficient to remove fission product decay heat and the core zone is dimensionally stable then a FNR is passively stable;
11) For safety the reactor cooling system should be designed so that natural fluid circulation is sufficient to remove fission product decay heat;
12) For practical primary sodium pool maintenance there must be available storage containers into which the primary coolant can easily be transferred;
13) The primary coolant pool mass and the intermediate heat exchange components must be rated to safely accommodate worst case power transients;
14) In normal operation reactor temperature is nearly constant. Power reduction is achieved by reducing the intermediate coolant flow rate.
15) A distinct advantage of FNRs is that they are almost unaffected by slow neutron poisons. Hence the thermal power of a FNR can ramp quickly to follow a rapidly changing grid electricity load.
16) Cold shutdown of a FNR is achieved by withdrawal of the mobile square active fuel bundles from the matrix of fixed octagonal active fuel bundles.
17) Normal hot shutdown of a FNR is achieved by thermal expansion of the primary coolant;
18) After either a hot or cold reactor shutdown there must be sufficient primary and intermediate coolant natural circulation to safely remove fission product decay heat.
The liquid sodium cooled fast neutron reactors (FNRs) contemplated herein are primarily fuelled by the abundant uranium isotope U-238. The main FNR nuclear process converts the uranium isotope U-238 into the plutonium and then fissions the plutonium. Most of the resulting fission products have short half lives. There are excess fission neutrons that:
a) if absorbed by an abundant U-238 atom in the blanket yield a fissionable Pu atom;
b) if absorbed by a Pu-239 atom in the blanket may yield a Pu-240 atom which plays an essential role in preventing proliferation of nuclear weapons;
c) if absorbed by an abundant thorium (Th-232) atom in the blanket yield a fissionable uranium (U-233) atom; or
d) if absorbed by a lithium isotope (Li-6) atom in the blanket yield a mix of tritium (H-3), rare helium isotope (He-3) and common helium (He-4) atoms.
At about six year intervals part of the fuel in a FNR is removed and reprocessed to:
a) extract fission products;
b) replace the extracted fission product mass by an equal mass of new U-238;
c) to transfer Pu, U-233 and other transuranium actinides from the innermost reactor blanket fuel rods to new reactor core fuel rods; and
d) to harvest H-3 and He-3 atoms for use in fusion power systems.
The time required for complete reactor fuel replacement is known as a fuel cycle and is typically about 30 years.
During each FNR fuel cycle the Pu fraction in the core fuel rods 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. With U-238 blanket fuel the fuel recycling process yields excess core fuel rod material which enables ongoing expansion of the FNR fleet. With Th-232 blanket fuel the fuel recycling process can sustain the existing FNR fleet but does not allow significant fleet growth.
FAST NEUTRON REACTOR SIDE ELEVATION:
For clarity in the above diagram the fuel bundle support tubes, the overhead gantry crane and the air locks are not shown.
FAST NEUTRON REACTOR PLAN VIEW
In the above diagram only one quadrant of active fuel bundles is fully populated and only 4 of 32 intermediate heat exchange bundles are shown.
FAST NEUTRON REACTOR PHYSICAL DESCRIPTION:
A practical FNR used for thermal and electrical power generation is formed from an octagonal assembly of about 500,000 vertical closely spaced 0.500 inch outside diameter closed end steel fuel tubes 6.0 m high X 14 m in diameter that is centrally positioned in an upright cylindrical pool of primary liquid sodium 20 m inside diameter by 15.0 m deep. The primary liquid sodium surface is 1 m below the pool deck. The depth from the primary liquid sodium surface to the top of the steel fuel tubes is 6.0 m. There is 3.0 m of primary liquid sodium depth below the bottom of the fuel tubes.
This primary sodium pool is lined with stainless steel and is thermally insulated from its surroundings by a 2 m thick layer of fire brick. There are two further steel liquid sodium containment layers followed by a 1 m wide ventilated air space stabilized with structural steel which in turn is enclosed by a ~ 1 m thick concrete wall. The concrete wall is externally stabilized by fill embankments of sufficient height, density and dryness to safely contain the liquid sodium even in the event of combined sodium pool and concrete wall structural failures. The external embankments also protect the FNR from impact by a large airplane.
The primary sodium pool must always be above the maximum possible elevation of the local water table during worst possible flood conditions.
Outside the reactor enclosure is a ring of up to 32 steam generators mounted in isolated spaces on an upper level. On a lower level below the steam generators are up to 32 steam turbogenerators and condensers. The condensers feed captured heat to a district heating system and to four independent natural draft dry cooling towers. These cooling towers are each sized for safe rejection of the reactor's maximum fission product decay heat by natural circulation when there is no electricity. At urban reactor sites these cooling towers may appear to be nearby high rise buildings.
The top surface of the primary 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 cover gas the reduced exposed sodium surface area minimizes the rate of spontaneous sodium oxidation and the fire risk.
There is an overhead gantry crane that is used to reposition fuel tube bundles and to replace intermediate heat exchange tube bundles.
In the middle of the primary sodium pool, poking up through holes in the metal floats, is a grid of vertical steel indicator tubes that project 0.3 m to 1.5 m above the liquid sodium surface. These indicator tubes convey to overhead monitoring and control instrumentation important data relating to mobile square fuel bundle position, discharge temperature and gamma ray production. The indicator tube attachment fitting also allows lifting of individual fuel bundles by the overhead gantry crane.
Immersed in the hot liquid sodium around the perimeter 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 while the reactor is shut down for prolonged service. The isolated heat transfer fluid is non-radioactive intermediate sodium. For safety reasons either the intermediate liquid sodium pressure is kept higher than the primary liquid sodium pressure or the intermediate sodium and the steam generator water are isolated via a double isolated heat exchanger. This heat transfer arrangement is safe in the presence of either intermediate heat exchange tube bundle failures or steam generator heat exchange tube bundle failures.
There are 3 m diameter air-vacuum-argon locks that that used to bring equipment into the reactor space or remove equipment from the reactor space without loss of argon and without mixing of air and argon.
There is also an airlock for personnel entry and exit. This air lock uses chemicals to minimize the amount of air which mixes with the argon cover gas.
The argon cover gas above the primary sodium is at atmospheric pressure. Its normal operating temperature is about 500 degrees C. When the reactor is shut down for service this temperature decreases to about 120 degrees C. Each FNR has redundant bladder tanks in external silos to accommodate thermal expansion-contraction of the argon cover gas.
To the extent possible all work in the reactor enclosure is done using robotic equipment. Any entry by service persons requires both a reactor shutdown and a thermally insulated and cooled suit with a closed circuit breathing apparatus. Such entry might be necessary to replace a defective intermediate heat exchange bundle.
The fuel tube assembly is completely surrounded by a ~ 3.0 m thick guard band of high purity liquid sodium. This sodium guard band absorbs all neutrons that escape from the assembly of fuel tubes. These absorbed neutrons convert natural sodium (Na-23) into 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 radio isotope emissions. For certainty with respect to gamma radiation safety, reactor physical security and sodium containment a FNR is surrounded by a ~ 1 m thick concrete wall.
Due to the sodium guard band 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 long intermediate heat exchanger working life, a very long facility working life and prevents formation of radioactive decommissioning waste. Absence of neutron activation also minimizes the safety complications involved in intermediate heat exchange bundle: alloy selection, repair and replacement. Generally the intermediate heat exchange bundles are fabricated from high nickel steel (INCONEL).
The fuel tubes emit heat at a variable rate which keeps the primary liquid sodium in the reactor core zone at about 502 deg C. The mechanism which controls the fuel tube heat emission rate is thermal expansion of the fuel and liquid sodium. Due to its lower density and hence buoyancy hot liquid sodium naturally rises vertically between the fuel tubes. When the upper 10.2 m of primary liquid sodium reach 502 deg C with no thermal load the primary sodium natural circulation stops and hence heat production stops. From a thermal engineering perspective a FNR acts as a nearly constant temperature source of flowing 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 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 active fuel tubes.
When there is no external thermal load the heat extraction rate is zero. When the top 10.2 m of primary sodium reach 502 degrees C the nuclear chain reaction stops. Hence with no thermal load the primary liquid sodium pool surface temperature stabilizes at 502 degrees C. However, there may still be fission product decay heat. If the primary liquid sodium pool temperature exceeds 510 degrees C heat is rejected via natural circulation and the natural draft cooling towers without any reliance on electrically powered fans or pumps.
IMPORTANT FNR FEATURES:
In FNR core fuel the trans-uranium actinides preferentially fission, instead of simply capturing neutrons as in a water moderated reactor. With appropriate periodic fuel reprocessing a FNR yields about 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 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 long lived spent fuel waste production is about 2000 fold less than for a CANDU reactor.
The best method of spent CANDU fuel disposal is to reprocess the spent CANDU fuel into new FNR fuel and then consume it in a FNR.
Another important feature of FNRs is efficient electrical load following. Unlike the nearly 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 net load due to connected intermittent renewable energy 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.
FNR OPERATION PRINCIPLE:
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 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 atoms is extremely small.
A Fast Neutron Reactor (FNR) consists of a pancake shaped core zone 10 m diameter X 0.5 m thick containing uniformly distributed U-238 - Pu-239 - Zr alloy core fuel rods sandwiched between two 1.8 m thick blanket zones containing uniformly distributed U-238 - Zr alloy blanket fuel rods. The Pu 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. There is an ongoing flux of fast neutrons flowing from the core zone into the blanket zones. Hence there is ongoing production of Pu in the blanket zones.
The Pu concentration in the core zone and the core zone thickness are chosen so that slightly more than (1 / 3.1) of the neutrons emitted by fission of Pu atoms in the core zone are captured by other Pu atoms in the core zone. Hence a chain reaction occurs because the probability of an emitted neutron being captured by a Pu atom is slightly greater than (1 / 3.1).
In the core zone the rate of loss of Pu by fissioning is partially offset by the rate of production of Pu via neutron capture by U-238. Reactor criticality at the desired operating temperature is maintained through the operating life of fuel bundles via small incremental changes in fuel bundle geometry. These geometry changes are accomplished by using a liquid sodium 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 zone to increase, decreasing the probability of Pu atoms in the core zone capturing neutrons and then fissioning. 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 reduce the primary liquid sodium density which will turn the chain reaction off. A subsequent decrease in primary liquid sodium temperature below the 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, subject to an appropriate fuel mix and geometry, a liquid sodium pool type FNR automatically maintains the core zone temperature of the liquid sodium at the chosen temperature setpoint.
The FNR chain reaction maintains a constant liquid sodium temperature in the core zone. However, the thermal power switch from chain reaction full on to chain reaction off is gradual due to the gradual increase in primary sodium natural circulation as the temperature difference between the reactor core zone and the lower portion of the liquid sodium pool outside the fuel bundle assembly increases. During the switch from chain reaction full on to chain reaction off the liquid sodium stratification level changes by about 4.2 m. If the chain reaction is initially off due to high sodium pool temperature the amount of heat that must be extracted from the primary liquid sodium to bring the reactor to full power is:
4.2 m X Pi (8 m)^2 X 927 kg / m^3 X 160 deg C X 1.27 KJ / kg-deg C X 1 KWt-s / kJ
= 159,067,910 KWt-s
= 159,068 MWt-s
Since the reactor has a ~ 500 MWt average thermal output capacity during this period its nominal step response time is:
159,068 MWt-s / 500 MWt
= 318 seconds
= 5.3 minutes
Thus the thermal mass of the sodium prevents the fuel tubes from being exposed to thermal shock relating to very rapid changes in net electricity grid load. However, the intermediate sodium loop components are exposed to thermal expansion and contraction related to following rapid changes in net grid electricity load. Hence the heat exchange tubes in both the intermediate heat exchanger and the steam generator are made from Inconel.
Each mobile active fuel bundle has a liquid sodium hydraulic insertion/withdrawal actuator which sets its vertical position. This arrangement provides incremental control of the fuel bundle's liquid sodium discharge temperature set point.
The reactor thermal power output is set by the rate of extraction of heat from the primary liquid sodium pool. This heat extraction rate is a function of the difference between the secondary liquid sodium supply temperature and the thermal load temperature as well as the secondary liquid sodium flow rate. The maximum rate of heat extraction must be kept within reactor fuel tube assembly design limits. Otherwise the reactor fuel tubes could overheat.
The pressure in each steam generator is controlled by a motorized steam discharge valve which in normal operation 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 intermediate sodium loop. Thus controlling the intermediate liquid sodium flow rate controls the flow of constant temperature steam delivered to the corresponding turbo-generator.
The elevation of the steam generators is at least 10 m above the intermediate heat exchangers to provide sufficient natural circulation of the intermediate liquid sodium for fission product decay heat rejection when the intermediate sodium induction pumps are not energized. The steam discharged from the steam generators is ducted to turbogenerators and steam condensers located underneath the steam generator space. The steam generator space is fitted with rupture panels which in the event of sodium-water contact will vent the resulting hydrogen to the atmosphere.
FNR TEMPERATURE SETPOINT CONTROL:
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 from fission fragments emitted 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 normally 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 during a cold startup the insertion rate of the active fuel bundle control portion used to ajust each active fuel bundle's discharge temperature setpoint must be sufficiently low to prevent fuel rod melting and to prevent the FNR fuel bundle becoming critical on prompt neutrons. This insertion rate limit is ensured via flow orifices on the hydraulic actuator positioning valves as well as by programmed actuator position slew rate limits. Hence the insertion of the active fuel bundle control portion into the active fuel bundle surround portion is very slow and is very carefully controlled. By contrast on loss of control power the withdrawal rate of the active fuel bundle control portions is very fast. To enable fast active fuel bundle control portion withdrawal a parallel connected full port hydraulic actuator drain valve is used to achieve rapid fuel bundle cold shutdown. The hydraulic control 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 drain valve to open. The argon pressure to each actuator drain valve is controlled by an electronic transducer located in a cool environment outside the reactor space.
LIQUID SODIUM COOLED FNR OPEATION DESCRIPTION:
When the reactor is at ful rated power liquid sodium coolant enters the bottom of a FNR fuel tube bundles at about 340 C, flows upwards through the flow channels between the active fuel tubes, and emerges from the top of the active tube bundles at 502 C. The operating temperature setpoint of each active fuel bundle is controlled by the amount of mobile fuel bundle insertion into the matrix of fixed fuel bundles. Withdrawing a mobile fuel bundle from the matrix of fixed fuel bundles reduces the mobile fuel bundle discharge temperature setpoint.
The low 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 edges is cooled by the intermediate heat exchange bundles, causing the circulating primary sodium to increase in density and sink to the bottom of the primary sodium pool. This cooler higher density primary liquid sodium flows along the bottom of the primary liquid sodium pool to a point underneath the reactor fuel tube bundles and then rises again through the reactor fuel tube bundles. The intermediate heat exchange bundles have baffles to enhance this flow pattern.
The FNR geometry is chosen so that about (1 / 3) of the fission neutrons are absorbed by plutonium in the core zone to sustain the nuclear chain reaction and the remaining (2 / 3) of the fission neutrons are absorbed by 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 and sodium temperatures rise up to the fuel bundle temperature setpoint the material densities decrease allowing a larger fraction of fission neutrons to escape from the core zone into the adjacent blanket zone, which turns off the nuclear chain reaction.
Similarly, as the fuel bundle and sodium temperatures fall below the fuel bundle temperature setpoint the material densities increase 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 outside the fuel bundle assembly to increase, reducing the primary sodium natural circulation which turns off the chain reaction and hence causes 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 outside the fuel bundle decreases which restarts primary liquid sodium natural circulation and hence the nuclear chain reaction restarts. The core zone runs at a constant temperature of 440 C to 450 C. The rate of heat extraction from the core zone varies with the elevation of the liquid sodium thermal stratification layer.
Thus when the active fuel bundle control portions are correctly positioned the FNR thermal power output automatically varies to meet the external thermal load.
As FNR fuel ages its core zone plutonium concentration gradually decreases and its core zone concentration of neutron absorbing fission products gradually increases. Compensation for these long term changes is achieved by fine adjustment of the positions of the active fuel bundle control portions with respect to their 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 becomes Pu-239. Periodically the fuel rods from both the core and the blanket zones are reprocessed to extract fission products. Then the fission product mass extracted from core fuel rods is replaced by an equal mass of new core rod alloy obtained by reprocessing blanket rod material. Then the consumed blanket rod material is replaced by an equal mass of new depleted uranium-zirconium alloy.
The extracted fission products should be safely stored in isolation for about 300 years to allow their radio toxicity to naturally decay to below the level of natural uranium.
FUEL TUBE DETAIL:
The FNR fuel tube assembly consists of 605 active fuel tube bundles surrounded by up to 764 passive and cooling fuel tube bundles. Each fuel tube bundle has 1.5 m long lower legs and 0.25 top extensions which give it a total height of 7.75 m.
Each fuel bundle is composed 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.
Each square fuel tube bundle contains 312 X 0.5 inch OD HT-9 steel tubes. Each octagonal fuel tube bundle contains 460 0.5 inch OD HT-9 steel tubes.
The fuel tube to fuel tube spacing is maintained by horizontal (1 / 16) inch diameter steel rods. The tube lattice bottom spacing is fixed by the fuel bundle bottom gratings which support and position the fuel tubes and which permit vertical liquid sodium coolant flow. Some fuel tubes are missing from the lattice to allow for fuel tube bundle structural steel elements. The fuel tube bundles are positioned on a square grid 0.3333 m center-to-center.
Each active fuel tube contains 3 X ~.60 m long blanket fuel rods, 2 X .45 m long core fuel rod and then another 3 X ~ 0.600 m long blanket fuel rods. Each passive fuel tube contains 7 X ~ .600 m long blanket fuel rods.
During prolonged reactor operation the core fuel rods swell from 0.45 m long to about 0.5 m long. Each fuel tube contains a measured amount of liquid sodium. The top 1.4 m of each active fuel tube are nominally empty to provide sufficient plenum volume to relieve pressure stress resulting from formation of inert gas fission products and to store sufficient spare sodium to compensate for fuel tube material swelling. The plenum tube length also serves as a flow guide to enhance natural circulation of primary liquid sodium.
Each octagonal fuel tube bundle is supported and stabilized its 1.5 m high legs that are in turn supported by a 1.5 m high open steel lattice. The vertical insertion/withdrawal of each square fuel bundle is controlled by liquid sodium pressure applied to a piston type hydraulic actuator located within the steel lattice. The square bundle's vertical position is indicated by a vertical indicator tube attached to the top of the square bundle. This indicator tube projects above the surface of the liquid sodium.
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.
Hydraulic pressure lines routed through the steel lattice provide the controlled liquid sodium pressure that raises or lowers each mobile fuel bundle.
The mobile fuel bundle actuators are periodically automatically cycled to ensure that none of them stick in their fixed fuel bundle matrix guides.
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. There is space behind the filters to allow for liquid sodium cross flow so as to ensure liquid sodium can flow past all the active fuel tubes even if a particular filter section is obstructed.
FNR THERMAL OUTPUT LIMITS:
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 thickness with new fuel at ___ m. As the fuel ages and the fuel swells the core zone thickness is gradually increased.
The reactor core zone maximum outside diameter is a function of the liquid sodium pool inside diameter. That diameter is constrained by practical structural issues related to the reactor enclosure roof. With a 10 m diameter reactor core zone the sodium pool inside diameter is 20.0 m . It is necessary to have a roof covered 4 m wide perimeter strip around the liquid sodium pool for the insulating brick, maintenance access and air cooling, and gamma ray absorbing concrete. Hence the nominal reactor enclosure outside diameter is 28 m.
The reactor core zone height, the fuel tube assembly diameter and the fuel tube geometry 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 with a 160 degree C temperature rise.
At FNR core zone diameters smaller than 10 m the economies of scale related to the required liquid sodium pool volume are reduced. At FNR core zone diameters greater than 10 m the FNR enclosure cost quickly rises due to the length of the unsupported roof span and the required volume of liquid sodium.
LIQUID SODIUM COOLED FAST NEUTRON REACTOR OPERATION:
During reactor operation there is ongoing fission in the core zone so 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 an active fuel bundle has operated for some time (~ 30 years) the accumulated fission products in the core fuel rods and the decrease in Pu-239 concentration reduce the core zones' reactivity. In addition the fuel tubes may swell reducing the primary liquid sodium flow. The fuel tube bundle is then moved to the perimeter of the liquid sodium pool for 6 years to allow dissipation of short lived fission product decay heat. Over a fuel cycle period of 30 years the active and passive fuel bundles are gradually replaced at 5 X 6 year staggered intervals 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 bundles are disassembled. The fuel tube material, the core rods, the blanket rods, the liquid sodium within the fuel tubes and the sodium salts are each reprocessed differently. The fuel reprocessing involves successive selective removal of sodium and sodium salts, selective removal of uranium, selective extraction of fission products from the high atomic weight elements, selective extraction of zirconium from the low atomic weight elements 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-12% 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 fuel bundle materials much of the fuel reprocessing is carried out by robotic equipment.
The reprocessing of the used fuel rods and blanket rods yields more Pu and other trans-uranium actinides than the FNR needs. Hence the excess Pu and actinide inventory can be accumulated and used to provide fuel for starting other FNRs.
A major constraint on the rate of implementation of liquid sodium cooled FNRs is the available supply of FNR start fuel containing 20% (plutonium plus actinides). This start fuel may be obtained by reprocessing spent water cooled reactor fuel or by plutonium breeding by FNRs. It may take as much as 200 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 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. For safe power control Fast Neutron Reactors rely on thermal expansion to reduce each fuel bundle's reactivity as its core zone 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 FNR enclosure must be sufficintly robust that an enclosure collapse or a crane collapse sufficient to crush the matrix of fixed fuel bundles is not a credible accident.
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 either neutron activation or material degradation.
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 more than 1000 fold as compared to a CANDU reactor;6. FNRs can easily track rapid changes in grid load:
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. For fire safety the liquid sodium pool of a FNR requires both a floating steel cover and an argon gas cover. Each FNR should have an on-site cryogenic argon production facility.
The liquid sodium pool of a FNR must be located above the highest possible future water flood level. Hence usually FNRs reject heat to the environment by air cooling rather than by water cooling. In district heating systems surplus heat can be rejected via roof-top fan-coil units at customer owned premises.
Liquid sodium cooled FNRs are inherently unsuitable for marine applications.
Local fire department personnel must be trained to NEVER use water to fight a fire in a FNR facility. A preferred extinguishing agent is Na2CO3.
A significant security issue with FNRs is that they operate with a large Pu inventory. During FNR operation first-in first-out fuel bundle replacement should be followed to maintain a sufficient Pu-240 concentration in the fuel to prevent selective extraction of Pu-239 for fabrication of fission bombs.
Since FNRs are not located near large bodies of water they require adjacent small (~ 8% of reactor capacity) natural draft dry cooling towers for reliable rejection of fission product decay heat. In normal operation the FNR heat output is absorbed by a district heating system. In an urban environment excess heat can be rejected by rooftop mounted fan-coil units distributed across a city.
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 water moderated reactor fuel bundles into Pu and other fissionable trans-uranium actinides and then fissioning these actinides;
3. Effectively utilize existing water moderated reactors for interim on-going electricity generation;
4. Prevent nuclear weapon proliferation;
5. Achieve a fast neutron reactor fuel composition that meets the required performance and safety requirements.
BASIC FUEL RECYCLING CONCEPT - THE OTTENSMEYER PLAN:
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, then makes the residue metallic and then 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 below the level of natural uranium. Then, subject to selective radio isotope 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 about 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 should be sufficient to power Canadian Fast Neutron Reactors for centuries to come.
During each fuel cycle sufficient Pu is produced to sustain the reactor reactivity during the following fuel cycle and to provide yet more Pu 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 a small fraction of long lived nuclear waste for about one million years.
Obstacles to immediate implementation of FNRs are political resistance to 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 50 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 fuel reprocessing facility is 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 mine waste.
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.
FISSIONABLE ISOTOPE FORMATION CONSTRAINTS:
1) A CANDU reactor takes about 1.7 years to convert 1% of its fuel weight into fission products.
2) One fuel cycle time is the time period between successive reprocessing of FNR fuel bundles. During one ~ 30 year fuel cycle time a FNR converts about 15% of its core fuel weight or 75% of its plutonium into fission products. However, with each fission it genertes about 0.6 atoms of plutonium in addition to the two atoms of plutonium required to usstain its own operation.
3) The Pu-239 doubling time is the time required to double the total
available Pu-239 mass. Since each Pu-239 fission in a FNR produces
about 0.6 new Pu-239 atoms in addition to those necessary to sustain the FNR operation in one fuel cycle the amount of net new plutonium produced is:
0.75 (number of original plutonium atoms)(0.6) = 0.45 (number of original plutonium atoms)
Hence the Pu-239 doubling time is given by:
(fuel cycle time) / 0.45 = 2.22 fuel cycle times
Since one fuel cycle time is about 30 years the time required to double that amount of available Pu-239 is about:
2.22 X (30 years) = 67 years
1) Presently there is no recognition by either the Canadian or US governments that the Pu-239 shortage threatens the very existence of the human species. Without sufficient Pu-239 there is no sustainable substitute for fossil fuels.
2) Making Pu-239 from U-235 requires consumption of one atom of U-235 for every 1.5 atoms of Pu-239 produced. It takes much more natural uranium to start a FNR as it does to fuel a CANDU reactor of similar thermal output capacity.
3) If we contemplate quadrupling the the world nuclear reactor capacity over the next 40 years using breeder reactors to achieve sustainability we are committing the entire known mineable natural uranium resource. If we fail to do this as fossil fuels are exhausted there will be no economic energy left but renewable energy.
4) The only strategy that can mitigate these problems is conservation of Pu-239 and U-235. Consuming Pu-239 in water moderated reactors or burying spent water moderated reactor fuel containing these isotopes in the ground is worse than stupid.
5) All the new reactor designs that do not net breed new fuel should simply be discarded as a waste of critical resources. The regulatory authorities should do all necessary to to accelerate approval and funding of new breeder reactor designs.
6) From an electricity market perspective all new breeder reactor capacity should have the highest priority for electricity grid access. The existing market mechanisms will just have to be changed to make that happen. The high school core curriculum should have a section that discusses the crucial role of Pu in future energy production and that sufficient Pu will not exist unless breeder reactors are both funded and operated today irrespective of present natural gas prices.
7) The above observations are dictated by the laws of physics. There are all kinds of claims based on market models that have no foundation in physical reality. The human species as we now know it will live or die in accordance with the natural physical laws.
8) There is a possibility of harvesting natural uranium from the oceans. However, the concentration of uranium in sea water is only about 3 parts per billion so the cost of uranium recovery from the ocean is very high. Possibly there might be some marine life species that naturally concentrates uranium in the ocean which might make harvesting of uranium from the ocean more practical.
NUCLEAR WEAPON NON-PROLIFERATION:
A blunt reality that humans must face is that fossil hydrocarbons must remain in the ground. Sustainable production of reliable non-fossil power requires fast neutron power reactors. Fast neutron power reactors initially 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 maintaining 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 irradiated 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. Today the Russians also have an 800 MWe FNR and are working on a 1200 MWe FNR. Realistically, as compared to North America, the Russians have at least a 30 year lead in FNR technology. This lead is a direct result of fossil fuel industry led corruption of the US and Canadian governments.
1. The components of a FNR that are exposed to a high neutron flux are routinely replaced and recycled along with the reactor fuel.
2. A major issue in practical liquid sodium cooled FNR operation is complete exclusion of water and air. The reactor cover gas is argon. At high sodium temperatures exposure to the atmosphere or water will cause a spontaneous fire.
3. A significant limitation on the rate of deployment of FNRs is the time required for one FNR to breed enough plutonium to start another identical FNR. This period is known as the plutonium doubling time.
4. The scientific issues related to FNRs are well understood. However, due to governmental corruption by the fossil fuel industry, in North America there is little power FNR operating experience and North American electricity utilities have no pressing financial motivation to adopt FNRs.
This web page last updated October 19, 2019
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