Why in the world aren't we doing this ???
Breeders "can" produce Plutonium ..
Not a good thing if you're the 'cop'
in the non-proliferation business ..
Part II .... Breeder reactors ....
"What the heck is a Breeder Reactor and how does it work? That is a very complicated question because there are at least six different reactor designs that qualify as Breeder Reactors, however it boils down to one aspect. By definition a breeder reactor is any reactor that contains more fissionable fuel mass in its core at the end of its fuel cycle than it contained at the beginning of the fuel cycle. In other words a breeder makes more fissionable fuel than it consumes through fission.
In order to do this the reactor must have an ample supply of a fertile fuel in its reaction chamber which is converted into fissile fuel as the reactor operates. The classic list of fissile fuels are U-233, U-235, Pu-239 and Pu-241. These four isotopes are classified as fissile because they will undergo fission more than 50% of the time if they capture a neutron. By that definition U-232 is also a fissile fuel, it undergoes fission 57% of the time when it captures a neutron. U-232 is also a fertile fuel, in the 43% of the cases where it captures a neutron without undergoing fission it becomes U-233, a very highly fissionable isotope. The classic list of fertile fuel isotopes are Th-232, U-238 and Pu-240 which give rise to U-233, Pu-239 and Pu-241 respectively. Additional fertile fuels are Pa-232, U-234, Pu-238 which result in U-233, U-235 and Pu-239 respectively.
While most highly fissionable isotopes are odd numbered some even numbered isotopes like U-232 are more than 50% likely to fission and others like Am-242 are 90% likely to fission when they capture a neutron.
As demonstrated early in this thread in a typical LWR the U-235 isotope is about 80% likely to fission when it captures a neutron and about 20% of the time it transmutes into U-236. In those cases where the U-235 fissions it releases on average 2.43 neutrons however when the U-236 result neutrons are subtracted the total neutrons released per U-235 neutron absorbed averages 2.06 . In order for the chain reaction to continue one of these neutrons must induce another fissile atom to fission and at an 80% success rate per absorption this requires on average 1.25 neutrons of the 2.06 total available, leaving .81 for all other effects. Most of that .81 available are absorbed by fertile isotopes which make up the bulk of the core fuel in a reactor. In today's reactors this amounts to absorption by U-238 ultimately resulting in Pu-239 isotopes. Because at most .81 neutrons are available from U-235 in a thermal fission reaction it is impossible for U-235 to breed more Pu-239 than the amount of U-235 fissioned. A best case scenario for U-235 in a standard LWR as fueled today is that all of the excess neutrons released in the LWR will breed Pu-239 from U-238 giving a conversion ratio of .81 . As shown in the first post on this thread the average for current reactors is closer to .69, which is very good considering that control rods and many types of fission waste absorb neutrons preferentially, that is they absorb neutrons much more easily than U-238 does.
Pu-239 is the second most common type of fission fuel currently used by today's reactors, it actually is the source of one third of the energy released by the fission process over the life of the fuel inside the reactor core. Pu-239 releases significantly more neutrons per thermal fission event than U-235, 2.90, but it suffers from the fact that Pu-239 does not fission as often as U-235 when it captures a thermal neutron. As a consequence Pu-239 scores higher than U-235, but not by very much averaging 2.10 neutrons per thermal neutron absorbed. Only about 70% of thermal neutrons absorbed by Pu-239 result in a fission so it takes on average 1.42 of the 2.10 neutrons released to continue the chain reaction leaving only .68 for potential breeding of additional fuel in a thermal reactor.
Given these facts how is it possible to make a breeder reactor that operates on thermal neutrons? Obviously you can not use U-235 or Pu-239 as the fissile fuel for a thermal breeder, they just do not release enough neutrons in the thermal fission reaction. What about Pu-241? Pu-241 is unsuitable for two main reasons, the only way to manufacture it is to expose Pu-240 to neutrons, but the only way to get Pu-240 it to have Pu-239 absorb neutrons without fissioning. This makes Pu-241 manufacture cost prohibitive. Additionally separating Pu-241 from the Pu-238, Pu-239, Pu-240, Pu-242 in the spent fuel is extremely difficult because they are all so similar in mass to one another.
What about U-233? Well in that case the news is not only much better, it is excellent. All of the successful thermal spectrum breeders built and tested to date have been U-233 breeders that operate by converting Th-232 into U-233. While U-235 has a max potential of .81 conversion and Pu-239 has a potential of .68 the numbers for U-233 are much better. First of all U-233 releases an average of 2.50 neutrons per thermal fission and because it is 89% likely to fission it averages 2.27 neutrons released per neutron absorbed. Because it fissions so easily it only takes an average of 1.12 neutrons to continue the chain reaction which leaves 1.14 neutrons available to transmute Th-232 into U-233. Another key fact is that in a thermal spectrum Th-232 has a capture cross section of 7.4, while U-238 only has a capture cross section of 2.7. Th-232 is therefore 2.74 times more likely to capture a thermal neutron than U-238. While the control rods, structure, fission waste and coolant all still absorb some of the neutrons released in the fission a light water reactor has been fueled successfully with a U-235/Th-232 fuel mixture and at the end of the fuel cycle it contained slightly more U-233 than the U-235 consumed. Fueling the same design with U-233/Th-232 would give even better results because the U-233 releases more neutrons and fissions more easily, both factors in the conversion ratio.
For information on the USA LWBR project see LWBR
The second type of thermal spectrum breeder reactor extensively studied was the Molten Salt Breeder Reactor. This system uses liquid core made up of several different metalic flouride salts including primarily Lithium, Berylium, Thorium and Uranium. The liquid salts can not 'melt down' they are fter all already melted, and in most designs they only acheive self sustaining fission when the liquid salt circulates through a graphite core block. Because the core is a liquid and operates at very high temperatures the gasseous fission wastes like Kryton and Xenon bubble out of the fuel in a chemical seperator step and Iodine is also easily removed at this stage. By preventing the buildup of these specific kinds of fission waste it is possible for the reactor to operate with a very low level of fuel enrichment, in this case the fuel only has to be at half the enrichment of natural Uranium ore in order to operate. While the LWBR with a U-235/Th-232 core acheived a breeding ratio of 1.013 the MSBR acheived 1.05 and also demonstrated the abillity to use any fissionable actinide as a portion of its fuel mass, specifically the demonstration reactor in the USA used Plutonium as part of the fissile fuel while breeding U-233 from Th-232. With a graphite core block the MSBR is a thermal reactor, that is the neutrons it operates on are slowed down to give them a greater chance of interacting with the fissile and fertile fuel materials.
Another type of Molten Salt Reactor eliminates the graphite block and only uses the molten salt itself as a moderating agent. This does not slow the neutrons down as much, the spectrum is 'harder' and is described as Epithermal in nature. The Epithermal spectrum above 10,000 eV for neutrons has many advantages over the thermal spectrum, above 10,000 eV neutrons that are captured by Pu-239 yeild a higher average number of resulting neutrons than U-233, which means that at an Epithermal energy or harder Pu-239 becomes a better fission fuel than U-233.
For breeder reactors with greater than Epithermal spectrum neutrons we come to what are called 'fast' breeders. Fast in this case refers to the spectrum of the neutrons, not the speed at which the reactor operates. Several designs of 'Fast' reactors have been built and tested and even more designs have been proposed. The most common type of fast reactor is the Liquid Sodium or Liquid Potassium reactor which use liquified light metals as their coolant. Sodium and Potassium are both easily melted and offer excellent heat conductivity, they transfer heat very well from the core to the steam generators that convert the heat into electricity. On the con side both Sodium and Potassium are very chemically reactive, they will burn on contact with water and can catch fire easily in ordinary air which makes it paramount that leaks be prevented and opertunities for contacting water and air be limited as much as possible.
An additional liquid metal reactor design are the molten lead and molten lead-bisimuth alloy reactors designed in the FSU and used to power several Russian naval reactor designs. They have the advantage of high heat conductivity and low pressure like the Liquid Sodium and Potassium reactors without the fire danger, but there are problems caused by the fact that the Lead and Lead-Bisimuth alloys are very dense, which can cause erosion of the piping as they are pumped from the core to the heat exchangers and back.
A third fast reactor design avoids all of the problems these four liquid metal coolants have by simply eliminating the metal coolants all together, in this case by using a gaseous coolant. The coolant is usually Helium but can be any inert gas like Argon or Neon, or an inert compound gas like CO2.
The reason for using any fast reactor design has to do with the effect of an energetic neutron spectrum on the fuel. In the case of a 'fast' neutron fissioning the three main fission fuels you get the following increased neutron released per fast neutron absorbed ratio's. U-233; 2.60, U-235; 2.18, Pu-239; 2.74. For comparison the thermal ratio's are U-233; 2.27, U-235; 2.06, Pu-239; 2.10.
Consequently fissioning any fissile fuel in a fast spectrum demonstrably releases a greater ratio of neutrons which gives a greater opertunity for excess neutrons to breed fertile materials into fissile. In addition the fast spectrum has a much greater chance of causing fission in any actinide isotope that captures it including Pu-238 which only has a 6% chance of fissioning in a thermal spectrum and U-232 which only has a 57% chance of fissioning in a thermal spectrum."
Triff |
