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Fission FAQ v 1.5

Discussions of conventional and alternative energy production technologies.

Fission FAQ v 1.5

Unread postby Tanada » Wed 25 Jan 2006, 20:46:32

I am attempting to put together a Fission FAQ file so we do not have to keep regurgitating the same argument over and over. Please let me know of any mistakes I make and I will edit in corrections. It would also help if this were a Sticky topic and stays where people can find it right off the bat.

First the math behind Fission. Fission is a process that takes place on a nucleus level where fundamental particles interact to rearrange the building blocks of chemical isotopes into new chemical isotopes. Because these processes take place inside the nucleus of chemical isotopes they are dubbed 'nuclear', however both Fission, where the Nucleus splits and Fusion where the Nucleus grows are Nuclear, they rearrange the building blocks to change a chemical from one isotope into another isotope.

When a nucleus of any element heavier than Iron is fissioned and the fragments are larger than Iron the result is a release of energy. This is a result of the fact that Iron is at the peak of the binding curve of energy, the particles that make up the nuclease of Iron are packed as tightly together as physical laws allow nucleons to be in normal matter. While the difference in the binding energy between Iron and Hydrogen is the greatest difference on the curve, the difference between the Actinide elements and Iron is great enough to release copious amounts of energy. An atom of Uranium 235 has a binding energy of about 218.89 MeV, Iron Fe-58 has a binding energy of 505+-2 MeV.
When a nucleus of U-233, U-235, Pu-239 or Pu-241 fissions the binding energy is expressed as kinetic energy as the fission fragments speed away from each other, because the fission can result in a large range of fragment isotopes and a variable number of neutrons the energy release is averaged over a spectrum of possible results. U-233 averages 197.9 MeV, U-235 averages 202.5 and Pu-239 averages 207.1 Plugging these averages into formulas give an energy yield as follows;

For U-233 one gram fissioned yields 24.22654024 MWh(t)
For U-235 one gram fissioned yields 24.78966346 MWh(t)
For Pu-239 one gram fissioned yields 25.35278668 MWh(t)

To convert Thermal MWh into Electric MWh the heat released has to go through some sort of conversion process, typically this involves boiling water and using the steam to drive a turbine which in turn spins a generator to produce electricity. Not all Fission plants operate by simply heating water and using steam, some of the more efficient reactor designs heat a gaseous fluid like Helium or Carbon Dioxide which is used to directly drive a gas turbine, then used to heat water into steam and drive a steam turbine. This double use of the heat yields a considerable improvement in energy efficiency. An average fission plant converting heat directly into steam is about 33% efficient, a gas cooled reactor is closer to 50% efficient.

Assuming the average reactor world wide produces 1000 MWe and is 33% efficient at converting heat to electricity it will produce 3030 MWt by Fissioning 122 grams of U-235 or 119 grams of Pu-239 per hour of operation. When a fission reactor fueled with Uranium starts up for the first time all of the fission takes place in the U-235 in the fresh fuel, however as the reactor continues to run a small percentage of the Uranium 238 which makes up the bulk of the fresh fuel is converted into Pu-239. Each 12 to 18 months after the first start of the reactor the system is shut down for heavy maintenance and refueling. During the refueling cycle about one third of the used or 'spent' fuel is removed and placed in a large cooling pool for temporary storage. The remaining two thirds of the fuel is examined for damaged elements and rearranged in the core of the reactor where it is joined by a one third core of fresh fuel. When the reactor maintenance and refueling is completed and the reactor is started back up only the fresh fuel is pure Uranium, the remaining two thirds has a mix of Uranium, fission products, and Plutonium. After another 12 to 18 months the reactor again shuts down for heavy maintenance and refueling and again one third of the fuel is replaced with fresh fuel. From that point on until the reactor is decommissioned the core will be a mixture of one third fresh(first cycle), one third second cycle fuel and one third, third cycle fuel. The fresh fuel used in most reactors is all Uranium Oxide, the third cycle fuel has more Plutonium than U-235 undergoing fission in it.

Many European reactors now use a Mixed Oxide fuel loading where 30% of the fuel elements contain recycled Plutonium instead of enriched Uranium. For these reactors when one third of the fuel is replaced 10% of the total fuel is replaced with MOX elements. Any standard light water reactor can use this much MOX fuel without any modification to the reactor itself, however because of the differences in the way that Plutonium fissions compared to Uranium if the MOX exceeds 50% of the total fuel loading the reactor must either be designed for MOX or modified to operate correctly with it. The MOX fuel currently being used in Europe is made by recycling ‘spent’ fuel which has been used in a reactor for three to five years to recover the Uranium and Plutonium in it. On average ‘spent’ fuel has 1% mixed Plutonium isotopes in it that can be chemically separated from the Uranium and fission waste and recycled as MOX fuel. Most of this recovered Plutonium is currently being mixed with depleted Uranium in a concentration of about 7% vs. 93% Uranium but several other mixtures are being explored as options for future fueling systems. In the USA and FSU each 34 tons of weapons grade Plutonium is being mixed in 5% vs. 95% Uranium MOX elements and is to be used as fuel for civilian fission power reactors. Two other mixtures under study are 10% vs. 90% for Plutonium recovered from ‘spent’ MOX fuel elements, and a mix of 1% Plutonium, 5% U-235, 94% U-238, both of which are intended to allow Plutonium to be recycled repeatedly in standard reactors until it is all consumed. A final fuel under study in South Korea is TMOX-RG which has a mixture of reactor grade Plutonium, depleted Uranium and Thorium. TMOX-RG has the benefit of consuming Plutonium while producing Uranium-233 and is designed to be used as a replacement fuel for standard reactors.

EDIT ONE

More fuel information for those who remain interested. When a standard Generation II PWR of 1000 MWe is studied some assumptions are made. These include not only the MWt and conversion efficiency for heat to electricity but also the fuel enrichment ratio and 'burn time' which is calculated in terms of MW days per ton of Actinide metal. For a standard 1000 MWe Generation II reactor fresh fuel is assumed to be enriched to 3.5% and fuel is assumed to be 'burned' for 35,000 MWd/t. Recently these figures have been changed as reactors have been 'uprated' to operate more efficiently, use a more highly enriched fuel, and 'burn' the fuel longer between refueling cycles.

With the old standard of 3.5% U-235 enrichment the fuel would be 'burned' in the reactor for 3 years during which time the U-235 enrichment would fall from 3.5% to .84%. That would imply to the inexperienced that 2.66% U-235 was fissioned, however this figure is not correct. Nearly 20% of the time when a U-235 nucleus captures a neutron it does not fission, instead it is transmuted into U-236. U-236 is not a good nuclear fuel in thermal reactors, instead it has the tendency that when a neutron is captured it almost always transmutes into U-237 instead. U-237 is highly radioactive with a short half life, it almost always decays into Neptunium-237, the most stable isotope of Neptunium, before it interacts with another Neutron. Np-237 like U-236 does occasionally fission when it captures a neutron but predominantly it becomes Np-238 which is like U-237 in that it decays rapidly into another chemical, in this case Plutonium-238. Pu-238 has only a 6% chance of fissioning when it captures a neutron, the remaining 94% of the time it simply transmutes into Pu-239, which is a good fission fuel but not as good as U-235. In the case of U-235 about 81% of the time it fissions, for Pu-239 the figure is close to 62% of the time fission occurs.

Because standard reactor fuel starts out 3.5% U-235 and 96.5% U-238 most of the Pu-239 and heavier isotopes come from the U-238+n=U-239~~Np-239~~Pu-239 chain of reactions, but if you add up the U-236, Np-237 and Pu-238 in spent fuel you get .52%, or about 20% of the 2.66% U-235 which has been consumed during the fuel cycle. 2.66%*.20=.532%
So that explains where 2.66%-.52%=2.14% of the fission waste comes from, it was the U-235 fissioned during the fuel cycle. But when 'Spent' fuel is removed from the reactor it typically has 4% fission waste and 1% Plutonium and higher actinide isotopes in it. The remaining 1.86% fission waste comes from the Plutonium bred in the reactor which has undergone fission during the operation of the reactor. The operation of the reactor breeds not only the 1.86% of the fuel mass Plutonium fissioned, it breeds enough that about .9% plutonium remains in the fuel at the end of its life in the reactor. 1.86%+.9%=2.76% Uranium converted into Plutonium. 1.86% Plutonium fissioned + 2.14% U-235 fissioned=4. 2.76/4=.69, which is the conversion ratio of fuel fissioned to fuel bred for this spent fuel assay. Because this PWR reactor replaces 69% of its own fuel consumption while operating it is able to run for a much longer time. If you replaced the 96.5% U-238 in the fresh fuel with an inert matrix like SiC (Silicon Carbide) the reactor would not be able to use the fuel for 3 to 5 years per cycle, instead the fuel U-235 would have to supply all of the fission fuel and would run out in about 14 months. Only the fact that Fertile fuels in the core undergo conversion into fissile fuels make it economical to operate NPP's.
Last edited by Tanada on Sun 12 Feb 2006, 10:32:39, edited 5 times in total.
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Re: Fission FAQ v 1.0

Unread postby GoIllini » Wed 25 Jan 2006, 21:48:38

Nuclear Safety

The worst nuclear disaster to occur at a commercial facility occurred at Three Mile Island, near Harrisburg Pennsylvania, in the spring of 1979.

The problem was caused by a coolant pump failure in the reactor loop. Three Mile Island was a Pressurized Water Reactor that had a loop that took in cool, outside water as a way to condense water vapor coming out of the generator, a loop to run the generator, and a loop to take heat away from the reactor and use it to boil water in the generator loop.

Seconds after the pump failed, the control rods were automatically inserted and the nuclear reaction shut down. The reactor went from producing 3000 MW Thermal to 150 MW Thermal. The 150 MW Thermal was residual heat that sticks around at that level for several hours after a reaction shuts down.

Reactor operators got confused and were worried about a leak. This prevented them from turning on the Emergency Core Coolant system- which is required to come with every western reactor, in order to cool the reactor.

The reactor continued to heat up, until, finally, the ECC was turned on and cold water flowed into the reactor vessel. This caused some of the fuel elements to shatter, which is responsible for a majority of the "52% meltdown".

Had the ECC not been turned on in anywhere from 1/2 hour to several hours of when it was, it's possible the reactor vessel might have melted. Had this happened, the uranium would have gotten out of the control of operators. It's possible the uranium might have then melted through the floor of the containment building, and resulted in a below ground breach of containment. According to the WASH study, a below-ground breach would have only affected the water table and caused a slight increase in background radiation, resulting in the deaths of 10 people to cancer in a large population.

In order to make an area around the plant uninhabitable, there would have had to have been an above-ground breach of containment. There was some concern about a hydrogen build-up in the reactor that *might* have resulted in an explosion, but most engineers conclude that this would have not resulted in an above-ground breach.

According to the WASH study, a meltdown or near-meltdown, like the one experienced at TMI, can be expected about once every 10,000 reactor years. A meltdown involving an above-ground breach of containment can be expected about once every million to five-million reactor years. Two and a half thousand reactor-years of operation in the U.S. at least confirm that these figures are close to being on-target.


Chernobyl
The meltdown, or more accurately, steam explosion at Chernobyl in the Ukraine happened in 1986. The accident occurred due to a number of operator errors, and a number of design flaws prevented the design of the plant from protecting the public in the face of these operator errors.

Namely, the unit was involved in a safety test of its generators. I don't remember the exact details of the test, but it involved powering down the reactor. The electrical engineer conducting it violated numerous safety rules, and ultimately, removed most of the control rods on a very unstable reactor. The removal of the control rods caused a power surge, which turned coolant water in the reactor into steam. The coolant water's conversion into steam was a form of positive feedback that increased the power of the reactor and caused a steam explosion that blew the weak containment roof off the reactor.

Chernobyl's main engineering flaws were:

-Not having a combined coolant and moderator in the RMBK design, as in western reactors. A combined coolant and moderator actually causes negative feedback if the reactor loses coolant. Specifically, it shuts the reaction down.

-Having a weak containment. Chernobyl's RMBK design was created to allow online refueling. This feature was needed in case the Soviet government wanted to generate plutonium for its weapons program.

An RMBK would NEVER be approved in the U.S, the West, or just about anywhere in the world. The design was made in the atmosphere of Russian cold-war paranoia, combined with a repressive regime. Indeed the only 10 RMBK reactors ever built were constructed in the former Soviet Union, and many of them have been shut down.

Chernobyl had 33 directly attributable deaths (that is, 33 people were killed due to an explosion or radiation exposure), and the U.N's high estimate for total deaths due to cancer as a result of the accident as 5,000. It affected agriculture as far away as Norway, and rendered 20 sq. km of land around the plant site uninhabitable for ~30-50 years.
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Re: Fission FAQ v 1.1

Unread postby Tanada » Sat 28 Jan 2006, 08:05:27

Fissionable material supply:

Currently the world has more than 400 NPP's consuming mostly Uranium in operation and it is expected that this number will increase in the future. If each plant is a standard PWR design the natural uranium must go through an enrichment process to concentrate the isotopes that fission easily in a thermal spectrum. Most of the fuel used today is Low Enriched Uranium (LEU) that has 2.0% to 7.7% fission fuel in it. The predominant method of increasing the fission fuel ratio is to take Natural Uranium with a ratio of .71% U235, dissolve it with Flourine into a gas (UF6) and then run the gas through a series of high speed centerfuges. By centerfuging the UF6 the processor is able to concentrate the desired lighter mass U-235 and seperate out the heavier U-238. The process requires energy to spin the centerfuges, expensive chemical handling equipment and careful controlls and it leaves some of the U-235 in the 'tail' product of seperated out U-238. The ammount of U-235 in the tail is chosen by the processor for economic reasons, in theory you can keep centerfuging the Uranium until the tails fall to .1% or even .05% U-235. To do so however takes more time, effort and money, so long as natural Uranium is cheap it is not considered worth the cost. When the price of NatU was $40.00/kg in the 1970's the enrichment tails were kept between .15% and .2% U-235 but as the price fell in the early 1980's this was relaxed to .3% and even .35% in some cases. This difference has resulted in a large stockpile of 'depleated Uranium' in the .3%+ that is still in storage mostly at the enrichment facillities world wide. If the price of NatU climbs very high so that the enrichment cost is lower than the fresh Uranium cost some or all of the partially depleted stockpile will be put back through the enrichment process.
EDIT, fixed math error!
Another way to think of it goes like this, one ton of natural Uranium has .71% U-235 mixed in it. That works out to 7100 grams per ton, enough to run a 33% efficient fission reactor 58 hours if you could fission all of it. However in a light water reactor your fresh fuel feed is enriched, the natural Uranium has too much fertile fuel to fission fuel in the ratio to operate. The fuel goes through enrichment where you know the final product must have a ratio of at least 2%, but in most cases greater than 3.5%. (The fuel we import from Russia made from weapons Uranium is enriched to 4.4%) There are only two ways to increase the fissionable to fertile ratio in the natural Uranium, you can add fissionable material like Highly Enriched Uranium (HEU) or Plutonium from weapons that have been dismantaled or from reprocessed fission fuel, or you can enrich the NatU by seperating out much of the U-238 as depleated tails. Fresh reactor fuel at 3.5% U-235 has 35000 grams U-235 per ton, the fuel imported from Russia under the Megatons to Megawatts program has 44000 grams/ton. The fresh MOX fuel used in France and many other countries has up to 7% or 70000 g/ton Plutonium of which 1209 g/t is Pu-238; 41941 g/t is Pu-239; 17338 g/t Pu-240; 5914 g/t Pu-241; 3595 g/t Pu-242. The fissionable isotopes in a LWR are 41491 g/t Pu-239 plus 5914 g/t Pu-241 or 47405 g/t fissionable mass.

EDIT ONE
To enrich NatU from .71% to 3.5% with a tail assay of .2% you need 5470 kg per ton of enriched fuel. A useful online calculator for figuring out various ratios can be found HERE

Playing with the calculator we can quickly determine that at a high price for Uranium you can upgrade your tails from .3% assay to .15% assay you need 22,333 kg to get 1,000 kg of 3.5% enriched fuel, but the advantage comes from the fact that this semi depleted Uranium has already been mined, milled, transported and stored at the enrichment facillity making all of those steps sunk costs. Another result of higher NatU prices is that recycled uranium recovered from 'spent' fuel reprocessing goes back through enrichment taking it from .84% back up to 3.5%. Going from .84 back to 3.5% with a depleated assay of .15% takes in 4855 kg per metric ton of fuel enriched, at .2% assay it takes 5156. The 103 currently active USA NPPs put out about 3,100,000 kg of spent fuel per year. By reenriching this 'spent' fuel the USA could reduce its natural uranium demand by 3,822,000 kgs. Currently the USA is planning to store 70,000,000 kgs at Yucca mountain depository, by recycling and reenriching this spent fuel they could produce 14,418,000 kg of fresh fuel, enough to replace all the NatU demand for 4.65 years. Recycled Uranium has the same advantages as semi-depleted uranium in that it has already been mined and milled and is currently stored, but it requires chemical seperation to remove the fission waste products and actinides before it can be reenriched. The only figure I have seen for this process gives a price of $26.00 per kilogram of recovered Uranium from the reprocessing plant. If anyone has current figures from say France or the UK I would be interested in using them for further calcs.

EDIT TWO

From DEFFEYES TESTIMONY before congress Dr. Deffeyes gave the following testimony;

REP. BARTLETT: I get widely divergent estimates of how much fissionable uranium is left in the world, from 30 years to 200 years. Before we can really have an effective dialogue about how to address this problem, we need to have an agreement on what the problem is. And there is just so much difference of opinion out there, and I talked to the National Academy of Sciences. They would be delighted. We need to find the money for them. We need an honest broker somewhere that tells us roughly what the truth is because we have widely divergent opinions now as to how much fissionable uranium is out there.

MR. DEFFEYES: I suggest you look at the Scientific American for January 1980, Deffeyes and MacGregor, on the world uranium supply.

REP. BARTLETT: And how much is there, sir?

MR. DEFFEYES: Every time you drop the ore grade by a factor of 10, you find about 300 times as much uranium, so that going down to the ore grade of - going down through the ore grades continues to increase the supply. But just about the time we were writing that Scientific American article, these enormously rich deposits, and big deposits in Australia and Canada sort of blew away our early estimates and we had to quickly increase the estimates. There are deposits in Saskatchewan so rich that the miners can’t be in the same room as the uranium, where the uranium is being mined. They mine it by remote control. So at the moment we’re swimming in uranium, but the Deffeyes-MacGregor piece, which comes out with a Hubbard-like curve, says that, no, we can go on down, and specifically we don’t need a breeder reactor.



Last edited by Tanada on Fri 08 Dec 2006, 02:22:42, edited 4 times in total.
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Re: Fission FAQ v 1.2

Unread postby Tanada » Sun 29 Jan 2006, 04:28:38

Making Reprocessing Pay:

Much information both pro and con reguarding the cost and benefits of reprocessing 'spent' nuclear fuel has been bandied about. The pro side cites low costs, the con side cites high costs and all of the studies done are funded by the two sides creating a very clear case of bias, each side chooses numbers which support their own position.

In the past and even today in most of the world Reprocessing is done either by or for the federal governments of the countries where the reprocessing takes place. Therefore the real costs of reprocessing are obscured and only those isotopes of interest to the contracting governments have been recovered.

It has now been made clear however that the reprocessing facilities in Japan will be of a different nature, their primary driver will be economics, not government fiat. To be sure reprocessing in a general sense will be mandated by the government, however the details of the process are being left up to the contracted agent.

If you review

Link to image

and also

Image

you can clearly see that fission waste isotopes fall into two groups centered around masses of 95 and 139. The lower peak is where economics come to the forefront, in this portion of the curve you have yeilds of Ruthenium, Rhodium, Palladium and Silver. 10 years after removal from the reactor these isotopes fall to a low level of radioactivity with a yeild of 9% Ru, 3.5% Rh, 7.9% Pa and 1% Ag. A ton of spent fuel is 3% fission products so this works out to 2.7 kg Ru, 1.05 kg Rh, 2.37 kg Pa and .3 kg Ag. From SPOT PRICE I see that Rh is $3095.00 per troy ounce, Pa is $273.00 per troy ounce and Ag is $9.57 per troy ounce. There are 26.4 troy ounces per kg, so the Rh recovered from 10 year old spent fuel ammounts to $85,793.40 per ton of fuel reprocessed, which is more money than the entire cost of mining, milling, converting, enriching, fissioning and reprocessing the fuel!

To read a fascinating article about potential industrial applications for fission platinoids recovered from fission waste see FP APPLICATIONS The difference in the yeild cited between the article and the graph above has to do with the fission time in the reactor of the 'spent fuel'. The 33 GWd/t figure is for older spent fuel currently stored while the higher yeild graph cited above is for modern fuel fissioned for 55 GWd/t or more. If the article about FP is correct then the projected 2030 supply of 1000 t fission waste Palladium is very significant, that ammounts to 26,400,000 troy ounces per year. Current world production is 8,300,000 troy ounces per year so in effect reprocessing spent fuel would quadruple the supply of palladium. Naturally this would have the effect of driving down the price of Palladium on the market, but at the same time it would open up a huge new set of applications which are currently too costly for palladium use.
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Re: Fission FAQ v 1.4

Unread postby Tanada » Mon 30 Jan 2006, 21:20:02

Section; 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.
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Re: Fission FAQ v 1.4

Unread postby cube » Tue 07 Feb 2006, 13:52:08

Tanada wrote: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.
As we all know - energy is always conserved, it cannot be created or destroyed.

So how is it possible for a breeder reactor to "produce" more fuel then it consumes?

here's an example:
Imagine a lock box filled with coal. If you take a lump of coal and use it to burn a hole into the lock box, giving you access to the larger quantity of coal inside, you've just "created" more fuel then what you've started off with. Perhaps a more accurate statement would be, you now have greater access to more fuel/energy then what you started off with. The laws of physics have NOT been violated.

A breeder reactor works the same way.

U-238 can be thought of as the "lock box filled with coal" and U-235 can be thought of as the lump of coal used to gain access to the "lock box". U-235 may be in short supply but this planet has a generous supply of U-238, so a future shortage of nuclear fuel would be highly unlikely.

This may seem like a half-baked analogy but at least it helps to dispell the myth that somehow breeder reactors violate the 1st law of thermodynamics which it certainly does NOT. I lost count how many people who have proclaimed that breeder reactors don't work.
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Re: Fission FAQ v 1.4

Unread postby Tanada » Tue 07 Feb 2006, 18:56:05

cube wrote:
Tanada wrote: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.
As we all know - energy is always conserved, it cannot be created or destroyed.

So how is it possible for a breeder reactor to "produce" more fuel then it consumes?

here's an example:
Imagine a lock box filled with coal. If you take a lump of coal and use it to burn a hole into the lock box, giving you access to the larger quantity of coal inside, you've just "created" more fuel then what you've started off with. Perhaps a more accurate statement would be, you now have greater access to more fuel/energy then what you started off with. The laws of physics have NOT been violated.

A breeder reactor works the same way.

U-238 can be thought of as the "lock box filled with coal" and U-235 can be thought of as the lump of coal used to gain access to the "lock box". U-235 may be in short supply but this planet has a generous supply of U-238, so a future shortage of nuclear fuel would be highly unlikely.

This may seem like a half-baked analogy but at least it helps to dispell the myth that somehow breeder reactors violate the 1st law of thermodynamics which it certainly does NOT. I lost count how many people who have proclaimed that breeder reactors don't work.


Your analogy might make sense to you but it would leave me scratching my head in confusion.

EDITED FOR CLARITY

Let me try your analogy translated like this. You have a big pile of coal in your furnace, but your matches just won't get it too burn. Therefore you take a wooden stick and shave it into a pile of small flinders. You put the flinders of kindling in a pile, then carefully arrange the coal on top of it. Your matches do succeed in starting the wood kindling afire, and they in turn burn long and hot enough to start the coal on fire as well. Once the coal is burning you can keep the fire going by simply adding more coal. Right now we are piling the wood on top of the coal and only a little of the coal is burning with the wood, and we get a lot of ash which smothers the fire.

Replace the Wood in the Analogy with U-235 and the Coal with Th-232 or U-238. Our current system uses up half the 'wood' and only manages to burn 1% of the 'coal', but in a breeder you will burn up not only the U-235, you will also be able to burn up the U-238 and Th-232 which ammount to 540 times as much fuel as the U-235. Therefore with a breeder you effectively stretch the fuel supply from 100 years to 54,000 years.
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Re: Fission FAQ v 1.4

Unread postby Frank » Sat 11 Feb 2006, 06:53:49

Thank you for creating this post. Can/will you please comment on CANDU style reactors - light vs. heavy water, fuel handling, efficiency, cost, etc.?

TIA
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Re: Fission FAQ v 1.5

Unread postby Tanada » Sun 12 Feb 2006, 10:31:02

CANDU Reactors;

CANDU or CANadian Deuterium Uranium, a reactor designed in Canada to use natural Uranium with .71% U-235 moderated by Heavy Water to produce heat and electricity.
EDIT
Heavy water is water in which the average ‘light’ hydrogen, Protium, has been replaced with Heavy Hydrogen, Deuterium. Heavy water can be manufactured by electrically separating the Hydrogen and Oxygen in the water, then putting the Hydrogen through an enrichment process to concentrate the Deuterium which is about 1 part in 6000 of average surface water on Earth. This is the method which was used during WW II in European reactor research. It is much easier to enrich Hydrogen than it is to enrich Uranium or other heavy isotopes because of simple physics. H-1 makes up about 5999 of every 6000 atoms while H-2 makes up just one, but the H-2 masses nearly exactly twice as much, a 100% difference. To enrich Uranium you are separating U-235 from U-238, which only have a mass difference of .0127%. Once the Deuterium is separated the gas is burned in pure Oxygen, the vapor is condensed into water and shipped in water tanks to where it is needed. Heavy water is not radioactive and you can bathe in it without risk, but some studies show that drinking too much of it can have detrimental health effects. However any spills of Heavy Water tend to be mixed quickly with surface waters and diluted back to the natural level of concentration quickly.
To read about how AECL of Canada enriched Deuterium today read the bottom page of this LINK


Light water reactors use regular fresh water which slows or thermalizes neutrons very rapidly because the Protium H-1 isotope has nearly the same mass as the Neutron involved in the collision. To reduce the energy of a fission neutron from the average 2 MeV it has at release down to .025 eV it has to collide with 18 Protons. Because it masses twice as much Deuterium does not absorb as much energy per collision, therefore it requires 25 collisions for Deuterium to slow a fast fission neutron to thermal energy levels. Because of this simple fact of physics it requires more volume of Heavy Water to slow neutrons all the way to thermal energy than Light Water. For other chemicals used in reactors the number of collisions is 43 for Helium, 86 for Beryllium, 114 for Carbon and 2172 for Uranium. These numbers are important because Helium is used for gas cooled reactors, Beryllium is used for Molten Salt reactors, Carbon is used for Graphite moderated reactors and Uranium has similar numbers to all the Actinides and is used in almost all reactors fast or thermal. Given that CANDU reactors use thermal neutrons to fission heavy metals it would seem that Light Water would do a better job, however Protium also has the unfortunate effect that a few percent of the collisions can lead to the neutron being absorbed forming a Deuterium H-2 isotope. This effect robs the reactor core of neutrons needed to continue the chain reaction, in an average light water reactor this effect is compensated for by enriching the Uranium from .71% U-235 to at least 2.0% U-235. The longer the fuel is intended to remain in the core the higher the enrichment needs to be. You can in theory fuel a LWR with natural Uranium, however within a short time the fuel no longer has enough reactivity to operate and you typically have to shut the reactor down to refuel it. Refueling for a LWR as currently designed can not be done while the reactor is producing power and requires about 21 days, which costs the utility considerable expense.

CANDU reactors side step the whole issue in two ways, they use Heavy Water which has an exceedingly low probability of absorbing neutrons needed to maintain the chain reaction, and they are designed specifically to be refueled while producing full power. This is made possible by the fact that the CANDU reactor is designed so that the fuel bundles are placed in pressure tubes inside the water tank holding the heavy water moderator, similar to the way a tubular boiler is designed for a fossil fueled steam plant. By exchanging fuel in one tube at a time the reactor is able to continue operating without disrupting the operation of the system. When fueled with natural Uranium a typical CANDU reactor exchanges 15 fuel bundles per 24 hours of operation, out of a total of about 4500 bundles in the core. Fuel bundles of natural Uranium spend up to 470 days in the core of the reactor this is achieved by moving some of the 15 removed to different fuel pressure tubes within the core. This is done to keep the core power curve ‘flat’, that is all areas inside the core produce close to the average number of fissions and therefore heat as the other area’s. Fresh fuel has no neutron absorbing fission fragments in it therefore it has a higher reactivity than older fuel, by mixing fresh fuel bundles with partially spent bundles throughout the core this higher reactivity is averaged into the mixture instead of creating a power spike in one area. In a typical LWR the power spike produced by fresh fuel inserted in the core is dealt with through the addition of neutron absorbers, which also has the effect of decreasing the conversion ratio and wasting potential fuel in the process.

Fresh fuel for a CANDU reactor is typically natural Uranium with a fresh fuel assay of .71% U-235 and a spent fuel assay of .2% U-235 and .3% Pu-239 & Pu-241. Because it is relatively simple to mix fuel bundles in the core a CANDU is very flexible in its fuel loading. Tests have been done using Thorium, Enriched Uranium, Plutonium/Thorium, Plutonium/Uranium, Plutonium/Silicon-Carbide, and spent LWR fuel known as DUPIC (Direct Use PWR spent fuel In CANDU). Because of these extensive testing programs it is now known that a CANDU reactor can operate safely on any combination of these fuels as well as the ability to operate on natural Uranium. Any of the fuel types with a higher than natural Uranium level of reactivity are able to spend a longer time in the core producing more power per bundle. The Plutonium/Silicon Carbide bundles are designed to burn Plutonium in an inert matrix where it can not breed more Plutonium in the process these bundles are pure consumers and consequently have to be of a higher reactivity level than natural Uranium if they are intended to remain in the core for the typical 470 days of full power production. By using the inert matrix or the Plutonium/Thorium option either weapons grade or reactor grade Plutonium can be used as fuel in the CANDU and in the process eliminate stockpiled Plutonium from both decommissioned nuclear weapons and reprocessing of spent reactor fuel. If spent natural Uranium CANDU fuel were to be reprocessed the Plutonium in 236 bundles would be enough to fuel 100 fresh Pu/Th or Pu/U bundles. Put another way by reprocessing spent CANDU fuel you can 42% more energy out of the fuel which saves that much mining and milling of natural Uranium ore. So long as natural Uranium remains cheap and reprocessing expensive there is no economic incentive to do so and for now the spent fuel is stored.
Last edited by Tanada on Fri 08 Dec 2006, 02:36:59, edited 2 times in total.
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Re: Fission FAQ v 1.5

Unread postby ChumpusRex » Sun 12 Feb 2006, 17:37:30

Heavy water is manufactured by electrically separating the Hydrogen and Oxygen in the water, then putting the Hydrogen through an enrichment process to concentrate the Deuterium which is about 1 part in 6000 of average surface water on Earth.


That's not how Atomic Energy of Canada (AECL) do it. While deuterium can be separated by electrolysis, it is too energy intensive an d expensive if bulk deuterium is required. Similarly, distillation is also too energy intensive.

Most heavy water is prepared by a chemical method - utilising the fact that different isotopes have slightly different chemical properties. The Girdler process is most commonly used - it works on the principle that the equilibrium point between hydrogen sulphide and water is different for deuterium and protium. This method avoids the need for expensive catalysts and energy.

In practice, the Girdler process is used to achieve enrichment of about 25% deuterium. Distillation or electrolysis is used to finish the job and achieve the 99% required for reactor use.
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Re: Fission FAQ v 1.5

Unread postby Frank » Wed 15 Feb 2006, 15:25:43

Thank you for the explanation. I do know that there used to be a CANDU reactor in Karachi, Pakistan but I don't know if it's still operational. My impression always was that the extra cost for this style of reactor was more than made up for by the cheaper fuel, online fueling, etc. Some of Ontario Hydro's plants had uptimes of 99+% if I recall correctly, for years on end.
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Re: Fission FAQ v 1.5

Unread postby sch_peakoiler » Thu 16 Feb 2006, 13:37:27

a layman's question on fast breeders.

Everywhere where I read about those breeders there are arguments being made that the reactors are unprofitable, unreliable and things. But as of now i did not find any document that would clearly say: "YES there are breeders running, that actually turned U238 into fissable fuel and thus actually generated more fissable fuel than consumed, however unprofitable and unreliable they might have been. Provided that they did not consume more energy than generated of course".

Can anybody of you proffessional guys say a definite YES or NO to that question? I am really puzzled, i even read that info about the russian BN-600 here http://eng.rosatom.ru/?razdel=160
a pretty nice description, but does not answer the question whether that thing can fastbreed.

Thanks in advance, I think this question should really be included into the FAQ, but it is my opinion.
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Re: Fission FAQ v 1.5

Unread postby ChumpusRex » Sat 18 Feb 2006, 08:56:00

There have been functioning fast reactors that have generated electricity and produced more fissile fuel than they used.

E.g the Phenix reactor at Marcoule, France. This is a demonstration reactor, for testing of the FBR concept. It has operated as a breeder, with a proven breeding gain of 16% (for every 1 kg of plutonium burned, 1.16 kg are formed). It produced approx 20 TWh of electricity.

See the IAEA's document: Operational and Decommissioning Experience with Fast Reactors
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Re: Fission FAQ v 1.5

Unread postby J_S_Bokchoy » Sun 02 Apr 2006, 14:46:53

I'm trying to figure out how expensive natural uranium will get before enrichment facilities find it economical to go back and extract more u235 from their depleted uranium. For back-of-envelope calculation purposes, I assume 400 nuclear power plants each using 30 tons fuel annually for 25 years, with each ton of fuel burned leaving 7 tons of depleted uranium behind at the enrichment plant, with 2/10ths of a percent U235 still left in it. That would be 400 x 30 x 25 x 7 x .002, or 4200 tons U235 that once separated could be blended at a 24 to 1 ratio with U238 to make 105,000 tons of burnable fuel, or 105000/(400x30)=8.75 years supply, long enough to bring plenty of new production into the market. What I can't even guess at, is how much more marginally expensive it becomes to centrifuge out that last ounce of U235, when the decisions have already been made once that it's not worth it. Any ideas? When I tried to guess the price where power companies would find it cheaper to switch to natural gas, it came out around $500 a pound, so maybe it's a moot point if recovery from depleted U works out even higher.
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Re: Fission FAQ v 1.5

Unread postby Tanada » Sun 02 Apr 2006, 20:32:01

J_S_Bokchoy wrote:I'm trying to figure out how expensive natural uranium will get before enrichment facilities find it economical to go back and extract more u235 from their depleted uranium. For back-of-envelope calculation purposes, I assume 400 nuclear power plants each using 30 tons fuel annually for 25 years, with each ton of fuel burned leaving 7 tons of depleted uranium behind at the enrichment plant, with 2/10ths of a percent U235 still left in it. That would be 400 x 30 x 25 x 7 x .002, or 4200 tons U235 that once separated could be blended at a 24 to 1 ratio with U238 to make 105,000 tons of burnable fuel, or 105000/(400x30)=8.75 years supply, long enough to bring plenty of new production into the market. What I can't even guess at, is how much more marginally expensive it becomes to centrifuge out that last ounce of U235, when the decisions have already been made once that it's not worth it. Any ideas? When I tried to guess the price where power companies would find it cheaper to switch to natural gas, it came out around $500 a pound, so maybe it's a moot point if recovery from depleted U works out even higher.


Take a look at Enrichment and play with the numbers a bit. Keep in mind that most reactors are now using an enrichment level around 4.4% U-235 (which is the grade the FSU sells us downblended weapons Uranium) because it allows the plant to burn the fuel for 8-12 months longer, saving costly refueling down time periods. With a tails assay of .3% (which has been common with cheap Uranium availible) and a product assay of 4.4% you need 300 tons of natural Uranium feed to get 30 tons of enriched fuel. If you store that 270 tons of depleted tails, which is what they do now, you can later re-enrich those tails, but the energy cost is substantially higher. To enrich your 30 tons of fuel the first time you used about 182000 SWU (Seperative work units). A couple of point on this, the calculator is old and assumes you are using Gaseous Diffusion instead of Centerfuge enrichment, if you centerfuge you save a heck of a lot of SWU energy.

The specific energy consumption is 2300-3000 kWh/SWU for Gaseous Diffusion, versus 100-300 kWh/SWU for gas centrifuge. The number of stages required to produce LEU is about 30 times larger in the diffusion plant than in the centrifuge plant. The corresponding equilibrium time is significantly longer in diffusion plants (months) as compared to centrifuge plants (hours). This effect, more intensive when the diffusion plant processes Uranium with higher enrichments, makes difficult and time consuming any significant change of the modus operandi of a gaseous diffusion plant. The large in-process inventory in the diffusion plant (a few tons in a small-scale diffusion plant) indicates the importance of closing the Uranium balance in this facility. On the other hand, for centrifuge plants, the small equilibrium time, small in-process inventory and the flexibility to change the cascade design (parallel to series) determine the importance of verifying that the plant is operating as declared.
(From CLICK )

Note that for worst case numbers a gas centerfuge is one tenth the cost per SWU in energy terms which needless to say substantially reduces the cost of enrichment. Because of this fact that 270 tons of .3% tails stored from a years worth of fuel production can go back through enrichment again, either retaining a moderate depletion level and making just a little more fuel, or being depleted to .12% which is the lowest level recorded for regular enrichment and which was done the last time Uranium prices were high. Depleting the tails from .3% to .25% only gains you 3.25tons of fuel enriched to 4.4%, depleting the tails all the way to .12% gains you 11.35 tons of fuel at 4.4%, or another third of a year worth of fuel from stored tails, without the cost of mining, milling and refining the metal which are sunk costs for the tails. To do this costs almost 209000 SWU, but when the cost has been reduced by 90% due to advances in the technology it is cheaper in economic terms than the orriginal enrichment preformed decades earlier with the old gas diffusion plants.

Another kink is that the spent fuel reprocessed in France and other countries seperates out the Uranium portion and stores it. This 'spent fuel' Uranium is about .84% U-235 and about .42% U-236 and can be sent back through enrichment to be reused, however it contains trace quantities of U-232 which is a gamma emmiter and requires much more stringent safety protocols during re-enrichment than does fresh natural Uranium or tails being upgraded. Currently this 'spent' Uranium is being stored because the price of natural Uranium was too low to make it worth the extra precautions needed, but as the price goes up re-enrichment becomes more attractive.

BTW all of this is covered in more detail on page 1 of this FAQ.
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Re: Fission FAQ v 1.5

Unread postby J_S_Bokchoy » Mon 03 Apr 2006, 00:24:41

So choosing a midpoint of 200 kwhr per SWU times 209000 @ 7 cents/kwhr would cost about 8 million for a third of a year's fuel. If the choice is between making fuel out of natural U or re-enrichment of stored tails, let the breakeven price equal B and solve for B as follows: $24 million = 10 (natural to fuel ratio) X 30 (tons a year) X 2000 lbs per ton X B, so B= $40 a pound. That's where it's at now, so I guess they might as well just leave their ore in the ground for 8 years until the million or so tons of tails worldwide get used up! At least it seems like that could inhibit the price from going much higher. But then if the value chosen for kwhr per SWU is 300, then B goes to $60. What other constraints could hold prices down, besides another TMI?
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Re: Fission FAQ v 1.5

Unread postby J_S_Bokchoy » Tue 04 Apr 2006, 01:12:47

AND please disregard preceeding inquiry, I forgot to include power expense on the right hand side. After that my hypothetical breakeven point was only $20 so the question must be pointless if the price is double that already. Maybe hedge fund speculation, panic inventory building or even the limited enrichment capacity now in operation washes out mere cost accounting.
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Re: Fission FAQ v 1.5

Unread postby Tanada » Tue 04 Apr 2006, 05:31:23

J_S_Bokchoy wrote:AND please disregard preceeding inquiry, I forgot to include power expense on the right hand side. After that my hypothetical breakeven point was only $20 so the question must be pointless if the price is double that already. Maybe hedge fund speculation, panic inventory building or even the limited enrichment capacity now in operation washes out mere cost accounting.


You are forgetting a couple key ingredients in the whole deal, so to speak. One factor is time and inertia, most Uranium is bought on long term contracts with the mines, not on the spot market because none of the power companies is going to risk sitting around with an expensive to build power plant and no fuel on hand. As those long term contracts expire new contracts are written based on more current market conditions, this raises the contract price more slowly than the spot market price, but the rise is less subject to short term uncertainty. As the long term price goes up the producers are incentivized to invest in greater production which will most likely exceed short term demand and drive the price back down.

For going on 15 years now the price of Uranium has been very depressed because back in the 1970's a lot of power companies signed very long term contracts or stockpiled large quantities of natural Uranium and stored it when prices were lower. Then on top of those factors the demand side was less than projected because TMI sensationalism badly hurt the nuclear industry in the USA and new plant construction ground to a halt. This left a few large companies with stockpiles for plants that were no longer being built and many companies with stockpiles for several plants but only one plant completed. If you buy a contract to fuel a set of 3 plants for 10 years and then end up with only 1 plant you have a potential 30 year fuel stock for that plant. Then add in the fact that the USA and USSR/Russia arms reduction treaties added even more Uranium to the market, so much that the price crashed and the treaty had to be renegotiated to reassure producers that well under half of any years supply would be reclaimed weapon material. On top of all that France aggressivley pursued reprocessing and MOX use that even further reduces demand for fresh Uranium.

Of course once the price crashed a new balance on the enrichment side was reached, especially in the USA where everything was still done with cold war designed gasseous diffusion plants and SWU is expensive. To save money cheap natural Uranium was only depleted to .3% instead of .15%, which leaves a bit more depleted Uranium but saves big bucks on enrichment costs, it cost about twice as much to deplete the tails to .15%. Halving the tails assay changes about 4% of the total feed from Tails into Fuel, how much does the feed have to cost and how cheap does your SWU have to be to make doing that a good idea economically? It is my understanding that in France they even reduced tails depletion to .35% or for a time .4%, but I don't have any documentation to confirm or deny that this happenned.
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Re: Fission FAQ v 1.5

Unread postby cube » Wed 05 Apr 2006, 13:17:09

sch_peakoiler wrote:a layman's question on fast breeders.

Everywhere where I read about those breeders there are arguments being made that the reactors are unprofitable, unreliable and things.
I think (anybody who disagrees feel free to step in) MOST breeder reactors were not built with the primary intention of supplying energy.

Since breeders can "produce" Plutonium it has definite military applications. Modern nuclear bombs are fusion and not fission bombs. Therefore they do not recieve their power from plutonium per say. Instead the PU is used to create a fission reaction which produces massive amounts of heat. The heat "triggers" the fusion reaction. Basically every fusion bomb is actually 2 bombs in one. A fission reaction is necessary to create a fusion reaction.

Because of its significant military applications I would assume all breeder reactors must be government owned. That would explain their unprofitable history. :-D
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Re: Fission FAQ v 1.5

Unread postby Tanada » Wed 05 Apr 2006, 19:36:46

cube wrote:
sch_peakoiler wrote:a layman's question on fast breeders.

Everywhere where I read about those breeders there are arguments being made that the reactors are unprofitable, unreliable and things.
I think (anybody who disagrees feel free to step in) MOST breeder reactors were not built with the primary intention of supplying energy.

Since breeders can "produce" Plutonium it has definite military applications. Modern nuclear bombs are fusion and not fission bombs. Therefore they do not recieve their power from plutonium per say. Instead the PU is used to create a fission reaction which produces massive amounts of heat. The heat "triggers" the fusion reaction. Basically every fusion bomb is actually 2 bombs in one. A fission reaction is necessary to create a fusion reaction.

Because of its significant military applications I would assume all breeder reactors must be government owned. That would explain their unprofitable history. :-D


I guess it is better to be half right than all wrong ;)

The problem starts with the premesis, a breeder reactor has one primary design goal, which is to produce more fuel than it consumes. They were not designed primarily ofr energy production, but rather for fissile fuel production. So long as Uranium remains cheap and abundant they are rather like the F-T Coal-to-liquid process, yes they work and yes they produce a usable fuel, but the natural abundant fuel is cheaper and easier to handle so why go to the extra expense and difficulty?

Commercial breeder reactors are all designed to produce elecricity to help offset the expense of running them, do a web search on Fermi I at Lagoona Point in Monroe, MI. It was the first commercial breeder and provided a lot of lessons to the electric utilities on just what was needed to make a commercial breeder work. It was not an economic success, but it was a technical success, the reactor produced more fuel than it consumed along with some electric power which was sold over the grid. In order to be a commercial success a breeder has to either be so efficient that it overcomes the cost difference in fuel manufacturing and reprocessing. In the past with aqueous 'chop-leach' reprocessing that was a very high hurdle, the new pyrometalurgical process however is vastly cheaper and produces a lot less secondary waste and may provide an economic incentive to build commercial breeders.

Weapons material reactors are hugely different from commercial breeders, they require frequent or constant refueling to keep the bred material from degrading in quality, and frequent reprocessing to recover the bred material. For the military with their bottomless budgets for nuclear weapons this was not a handicap, but for the electric utilities it is the road to financial ruin.
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