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PostPosted: Feb 12, 2008 3:16 pm 
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it's been stated that the MSR and the thorium cycle in general are "proliferation resistant" but I don't think there are any illusions that it is "proliferation proof." This is confirmed by this report which confirms what I had suspected:
http://www.bnl.gov/est/files/pdf/FactSh ... actors.pdf

U-233 is perfectly suitable for use in a nuclear weapon, at least in theory. The thing which makes it difficult is that there would be some U-232 as well. This does not preclude the use in a weapon, but the short halflife of U-232 makes it much more radioactive and therefore difficult to handle. It could still be done with a decent hotbox system and minimal contact and a weapon with some shielding. (alternatively, make political prisoners assemble it and rotate them when their bleeding gums and nausia becomes too severe to continue).

But in any case, it's really the U-232 which is going to make the uranium recovered less suited for weapons use.

However looking at the aspects of protactinium separation, I'm wondering if this could be a hole in the process which would allow for much lower U-232. U-232 is the daughter product of Pa-232 just as U-233 is the daugher of Pa-233. Pa-233 has a half-life of 26.9 days but Pa-232 is only 1.3 days.

This seems as if it could cause a problem. Basically if you separate the protactinium and let it decay for about eleven days, for example, you've gone through eight half-lives of Pa-232 but less one half of a halflife cycle of Pa-233. Thus you still retain about three quarters of the Pa-233 you started out with but the Pa-232 has been diminished to less than half a percent of what you started with. You could do it for even longer before you start to loose a lot of the Pa-233.

Thus, at this point you could do the process over again, removing the uranium and retaining the protactinium and you would have a very high concentration of Pa-233 and very little Pa-232, which is where the U-232 would come from. This is not very difficult and could easily be done with what is available. The result is basically an easy source of weapons grade U-233.

Given this, I'm wondering if just leaving the protactinium in the mix might be an option worth considering. If Pa-233 happens to absorb a neutron before it decays it will result in Pa-234, which decays within hours to U-234. U-234 is not a fissile material but it is extremely fertile. It readily captures neutrons and the result is U-235, which is of course the fuel of most modern reactors.

The reaction does yield energy in the end, but it's certainly much less than optimal. You have to put three neutrons in to get one fission. (not exactly productive) and you also will generate some U-236 from the U-235 which does not fission. U-236 has a very low neutron cross-section but it is sufficient to generate a small amount of neptunium-237. Hence you will begin to end up with a tiny amount of minor actinides.

From what I have read, leaving in the Pa-233 is going to decrease the efficiency and neutron economy but it does not seem to be a deal breaker. Less than ideal, yes, but it might be worth taking a hit on this, both due to the proliferation concerns and the fact that it would be one less step in the process.


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PostPosted: Feb 12, 2008 3:19 pm 
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oops. I linked to the wrong PDF. Not I have to find the bookmark to the correct one. I know I have it somewhere.


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PostPosted: Feb 12, 2008 3:35 pm 
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Look, there's no doubt in my mind that if a nation-state decided to operate a thorium breeder in such a manner as to produce weapons-grade material, they could do so, using some variation of the processes you've outlined here.

The question to ask yourself is, why would they?

Assuming they have the nuclear technology to build a thorium breeder (quite a steep challenge, since it hasn't yet been done) why wouldn't they take the far easier route of doing what the United States,

and Russia,

and France,

and England did back in the 1940s and 50s and build a simple graphite-moderated uranium toaster.

The US called their uranium toaster Hanford and it made enough plutonium for Nagasaki and Trinity in about nine months.

The Russians called their uranium toaster RBMK and it led to Chernobyl. The English had Windscale and the French had another one too. Heck, some of these reactors even made power too!

If your goal is weapons, or if your goal is even weapons-disguised-as-power, there's a MUCH easier way to do it than to mess around with thorium.

That why (SHOCK) after 60 years no nation-state has ever built a production nuclear weapon around U233. The US (on a whim) tried one one time and it was a bit of a dud. They never tried it again.


On the other hand, if you did have a thorium breeder, it would have a breeding ratio near unity, and if you tried to "skim" some U233 out of the protactinium pot, you wouldn't be able to run the reactor very long.

There's this idea out in the world that unless nuclear power can be made PERFECT--something no other power source is expected to be--then we can't consider using it. Well it won't happen. The good news about nuclear power, and thorium nuclear power in particular, is that I think it can get closer to "perfect" than anything else we've got.


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PostPosted: Feb 12, 2008 3:39 pm 
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drbuzz0 wrote:
From what I have read, leaving in the Pa-233 is going to decrease the efficiency and neutron economy but it does not seem to be a deal breaker. Less than ideal, yes, but it might be worth taking a hit on this, both due to the proliferation concerns and the fact that it would be one less step in the process.


One way or another you are going to have to move bred fissile from the blanket to the core (in a two-fluid reactor). You either remove it from the blanket as Pa and let it decay outside the core (for optimal neutronic performance, efficiency, and for a clean blanket) or you let it decay to uranium in the blanket and remove it by fluorination and THEN put it in the core.

But you've got to get it over one way or another. By removing it as Pa you can keep the blanket really clean and avoid having to fluorinate the whole blanket on a regular basis, as well as keeping the blanket really free of fission products.


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PostPosted: Feb 12, 2008 3:48 pm 
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Here it was, the one time it was done:

Image

Quote:
MET stands either for "Military Effects Test" or "Military Effects Tower" (according to Frank Shelton). This was a LASL test of a composite U-233/plutonium bomb core (the first test by the U.S. to use U-233) in a Mk 7 HE assembly. The 30 inch diameter spherical implosion system weighed 800 lb.

The primary purpose was to evaluate the destructive effects of nuclear explosions for military purposes. For this reason, the DOD specified that a device must be used that had a yield calibrated to within +/- 10%, and the Buster Easy device design was selected (this test gave 31 kt and used a plutonium/U-235 core). LASL weapon designers however decided to conduct a weapon design experiment with this shot, and unbeknownst to the test effect personnel substituted the untried U-233 core. The predicted yield was 33 kt. The actual 22 kt was 33% below this, seriously compromising the data collected.


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 Post subject: Please, no borons
PostPosted: Feb 12, 2008 6:46 pm 
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Kirk Sorensen wrote:
build a simple graphite-moderated uranium toaster


As it happens, it's not so simple to build a graphite-moderated reactor, because it requires the graphite to be extremely pure. The normal graphite production process uses boron as a seed for the graphite crystals, which obviously won't work for a reactor. I remember reading about the German's attempt to get a graphite reactor going, but unfortunately they couldn't get it critical because of the boron contamination. The Americans did better, because their graphite's cross section was 10% less.

I'm not sure about this, but are there international controls on nuclear graphite production?

Also, didn't someone mention (Jonathan?) in another post that if one were to make the thorium blanket out of thorium chloride, it would be easy to extract the Pa as PaCl3, which one could presumably reduce, set aside, and wait for the U233 to build up. Maybe that's not feasible, but people, being clever, might nevertheless be able to think of some other way of extracting U233.

At any rate, while I would agree that U233 is out for making a bomb, I would agree with a post that Jaro made about six months ago saying that U233 can nevertheless be used for a nuclear device.

This possibility requires some thought about how MSRs should be licensed and inspected. We should also be prepared to identify molten salt technologies that should be restricted or regulated based on their ability to be exploited for weapons making purposes.


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 Post subject: Re: Please, no borons
PostPosted: Feb 12, 2008 7:00 pm 
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honzik wrote:
I'm not sure about this, but are there international controls on nuclear graphite production?


I highly doubt it, because I highly doubt that there aren't legitimate non-nuclear uses for ultrapure graphite. Can't you just pyrolise natural gas and get graphite that way?


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PostPosted: Feb 12, 2008 7:24 pm 
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I will have to go & check the original article, but the modified PACER concept, using FLiBe, probably relied on U233-based fission-fusion bombs....

.


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PostPosted: Feb 12, 2008 11:52 pm 
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Kirk Sorensen wrote:
Look, there's no doubt in my mind that if a nation-state decided to operate a thorium breeder in such a manner as to produce weapons-grade material, they could do so, using some variation of the processes you've outlined here.

The question to ask yourself is, why would they?



My god, I'm sorry I brought it up at all. There's a few reasons for mentioning it. First, if the US/Russia/Brittan/France is planning on exporting any reactor to a lot of countries, potentially some that are politically unstable then there should be no illusion about the proliferation aspects.

Secondly, it's a political issue. Everything you say makes sense and yet proliferation concerns are what killed everything from reprocessing in the US to the integral fast breeder. Therefore it's worth considering that if someone demonstrates extraction of weapons grade material from a molten salt reactor that's going to be an issue that will become a political liability.

I'm simply saying that if you are going to sell the proliferation resistance of a thorium MSR you MAY want to consider whether a design effecient enough to dispense with that one aspect might be adventagious to enhancing this capability.


I'm sorry I got everyone so angry on that one

:cry:


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PostPosted: Feb 13, 2008 3:37 am 
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drbuzz0 wrote:
My god, I'm sorry I brought it up at all. There's a few reasons for mentioning it. First, if the US/Russia/Brittan/France is planning on exporting any reactor to a lot of countries, potentially some that are politically unstable then there should be no illusion about the proliferation aspects.


For what it's worth, I have NEVER been an advocate of selling or exporting our best nuclear technology to other countries. One of the most compelling aspects of the LFTR to me is the idea we can package it up in a submarine, park it off the coast of another country and sell them kilowatt*hours.

If they want to develop the technology themselves, and are hell-bent on weapons, as I and others have pointed out, there's a lot easier paths to take to what they're after. Using thorium is like driving from Tampa to Miami by way of Detroit.

What's embarrassing to me is how nuclear vendors are already ready to sell off American technology to the highest bidder. Witness the terms of the latest agreement between Areva and the Chinese.

But the Chinese already have weapons...you guessed it--a uranium toaster.


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PostPosted: Feb 13, 2008 10:14 am 
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Quote:
Given this, I'm wondering if just leaving the protactinium in the mix might be an option worth considering. If Pa-233 happens to absorb a neutron before it decays it will result in Pa-234, which decays within hours to U-234. U-234 is not a fissile material but it is extremely fertile. It readily captures neutrons and the result is U-235, which is of course the fuel of most modern reactors.


drbuzz0,

Your concerns and questions are indeed very important and this debate has come up numerous times on this site. Proliferation resistance is something we need to maximize wherever possible and skipping the Pa removal step is something that I would guess a majority of experts would agree is wise. Your conjecture is correct that by having Pa removed and stored is a potential avenue for obtaining relatively clean U233 (i.e. little U232).

First some facts. Many of us on this site strongly favor the 2 Fluid design of having one salt with the U233 (and maybe a little thorium) and a separate salt for the thorium. The Single Fluid design however has been what most researchers have focused on since the late 1960s.

In a Single Fluid design it is much more difficult to try to skip Pa removal and still break even. The way to lower the neutron losses to Pa is to lower the average neutron flux it experiences (especially thermal neutrons). You can do this by simply having a much larger core or by having excess salt that you cycle in and out of the reactor loop. However, in a Single Fluid design, having more fuel salt means having much more fissile material to start. This is not a deal breaker but a serious impediment nonetheless.

In a 2 Fluid design we can lower losses to Pa down to almost nothing by simply increasing the volume of blanket salt. This means paying for more thorium and carrier salt but thorium is very inexpensive (the true potential cost of mass produced Flibe salt is unfortunately one of the big unknowns). For example, 1960s 2 Fluid designs had about 260 tonnes of thorium in the blanket salt versus about 70 tonnes in the later Single Fluid design.

Kirk has pointed out advantages to keeping a Pa removal system as part of the design for even the 2 Fluid system. It saves some money on up front thorium and carrier salt costs, improves the neutronics slightly and keeps the blanket salt cleaner of fission products (there will still be some though).

I would say that weighed against those advantages are overwhelming disadvantages even beyond the proliferation issue which itself is reason enough for most people. The Pa removal system means employing complex liquid bismuth reductive extraction and doing so on a very fast time scale to have any effect (3 to 10 days is typical to process the entire salt volume). Compare this to simpler vacuum distillation to remove fission products which we can do at a much slower rate (months to even years) and still break even.

A last point is that in designs without graphite and purposefully harder spectrums (higher fissile concentration) the losses to Pa are relatively much less of an issue and Pa removal schemes are rarely mentioned as being part of those designs.


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PostPosted: Feb 13, 2008 11:52 am 
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David wrote:
I would say that weighed against those advantages are overwhelming disadvantages even beyond the proliferation issue which itself is reason enough for most people. The Pa removal system means employing complex liquid bismuth reductive extraction and doing so on a very fast time scale to have any effect (3 to 10 days is typical to process the entire salt volume). Compare this to simpler vacuum distillation to remove fission products which we can do at a much slower rate (months to even years) and still break even.


I'm not sure where the "overwhelming disadvantages" for Pa removal show up. In your scheme you still have to fluorinate the entire blanket, which is much larger in volume than a Pa-removed blanket. The uranium you'll remove from the blanket still has the same purity that the removed Pa would decay to. What's the difference there?

And another question, how are you going to vacuum-distill a thorium-bearing blanket?

David wrote:
A last point is that in designs without graphite and purposefully harder spectrums (higher fissile concentration) the losses to Pa are relatively much less of an issue and Pa removal schemes are rarely mentioned as being part of those designs.


That introduces other concerns, such as maintaining the reactor in its maximally reactive configuration.


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PostPosted: Feb 13, 2008 12:49 pm 
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My original point was that Pa removal offers a simple way to get a higher concentration of U-233 by allowing the Pa-232 to mostly decay and then reseperate. Unless I am missing something I was of the impression that the outer blanket would be expected to contain a high enough concentration of U-232 to complicate handling of the uranium and fabrication into a weapon.


Here; the pdf I had been looking for: http://www.princeton.edu/~globsec/publi ... _1kang.pdf


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PostPosted: Feb 13, 2008 1:06 pm 
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The Pa-232 would have been formed from neutron absorption in Pa-231, which in turn would have been formed from neutron absorption in Th-230, which (unless you introduced it into the blanket) wouldn't be present in any significant amount in the thorium.


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PostPosted: Feb 13, 2008 1:40 pm 
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I'm not sure where the "overwhelming disadvantages" for Pa removal show up. In your scheme you still have to fluorinate the entire blanket, which is much larger in volume than a Pa-removed blanket. The uranium you'll remove from the blanket still has the same purity that the removed Pa would decay to. What's the difference there?


And another question, how are you going to vacuum-distill a thorium-bearing blanket?


Kirk,

As you should know it is much more difficult to remove Protactinium from a carrier salt than it is to remove uranium by fluoride volatility (just bubbling through HF and F2 gas). Also, even with a rapid Pa removal scheme, you will still need to occasionally fluorinate out produced U233 as the Pa always has a residency time (see data below). Besides the relative complexity of the two processes, there is also no real rush to remove the produced U233, with perhaps a 6 month cycle being fine (as opposed to 0.5 to 10 days for Pa removal).

The blanket will be producing U232 at the same time so whenever you fluorinate out the U233 you will also get appreciable amounts of the desired U232 coming with it. It is true though that in a 2 Fluid design the blanket salt may produce a somewhat lower ratio of U232/U233 if the spectrum is softer in the blanket region (compared to a Single Fluid design).

In regards to vacuum distillation of the blanket salt I do not think this would be necessary over the lifetime of the plant as the fission rate in the blanket should be exceedingly low. I can not say that for absolute fact but I would assume we could have a cycle time of several years for fission product removal, thus we could have use liquid bismuth extraction on a tiny daily volume. Alternatively, just use vacuum distillation and throw away some thorium with the fission products (I know that sounds wasteful but it is dirt cheap and virtually inexhaustible). See below how the U233 content in the blanket can be as low as 0.0005% UF4 by simple fluorination so there will be very few fissions taking place.

The ORNL work of the mid 60s on 2 Fluid designs did end up proposing to separate Pa but again they were dictated to follow any method that could increase the breeding ratio (to lower the doubling time). As mentioned often, we only need to break even now. Here is data from ORNL 3996 where they looked at with or without Pa removal. Especially take note that they looked to process the entire blanket salt TWICE per day to remove Pa. I can`t see anyone thinking that is economically feasible.

FROM ORNL 3996 August 1966 Page xi

Graphite Moderated 2 Fluid designs (using graphite plumbing to keep the salts separate).
Both have similar fuel salts with 0.22% UF4 (fissile)

With Pa removal

101 tonnes of Thorium in the blanket
Fluoride Volatility on 55 day cycle
Pa removal on 0.55 day cycle! Means processing 2350 cu ft per day
UF4 in blanket (not given, assumed very low)
Specific Inventory 681 kg (1000 MWe)
Breeding Ratio 1.071

Without Pa removal

260 tonnes of thorium in the blanket (thus 2.6 times the blanket salt)
Fluoride Volatility on 23 day cycle
0.0005% UF4 in the blanket salt (fissile)
Specific Inventory 769 kg
Breeding Ratio 1.049

At the time ORNL had no proven viable method of Pa removal so they just took a blind guess and said it would cost roughly the same as fluoride volatility (bismuth extraction was a couple years later).

So yes with a impossibly fast Pa removal rate there is a marginal improvement in the breeding ratio but I stand by my "overwhelming disadvantages" statement.

N.B. Later in the document (page 113) they show that the design with Pa removal has the same 0.0005% UF4 in the blanket salt. That doesn`t seem to make sense so I am guessing there is an error somewhere (either the other case should be higher or this one lower).


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