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PostPosted: Aug 18, 2008 10:43 am 
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Lars wrote:
Hello David,
Thanks for posting your talk. Regarding the need to remove Pa233. Is the following correct?

The concern here is preserving neutrons rather than the downstream products from a capture in Pa233. The ratio of Pa233 to Th232 captures should be equal to the ratio of (Pa233 inventory times Pa233 capture area) / (Th232 inventory times Th232 capture area). In your design the blanket volume is around 20m^3 for 400MWe {6.6m x (1.1m x 1.1 m -0.5m x 0.5m ) x 3.14}. The blanket will contain 35 t Th232 and 35kg of Pa233. The thorium capture area is around 3x the Pa233 capture. Net we lose 1/3000 neutrons per fission to Pa233 capture. Your design has about 5x the blanket as the French design (10m^3 for 1GWe) so you design should both lose 5x less neutrons to Pa233 capture and have fewer neutrons escaping altogether. The price is buying 5x more blanket salt and Th initially and processing 5x blanket volume. A side benefit is that the outside wall containing the blanket is exposed to fewer neutrons.

Another possibility is to simply store blanket salt away from the reactor to reduce the quantity of Pa exposed to the neutron source and to allow some decay. This trade would reduce Pa capture but not leakage at the price of initial salt and Th but not extra processing.



Yes, I have not calculated losses specifically but I typically go by the previous work of ORNL on 2 fluid designs of the 1960s. They proposed about to have 260 tonnes of thorium in about 100 m3 of blanket salt to significantly lower the losses to Pa. For my design I can adjust the blanket thickness to make sure I have roughly the same above ratio or as you mention, simply store some of the blanket salt outside the reactor to have enough volume.

Don't forget, the French design has plenty of thorium and thus Pa233 in the central core so they have to worry about Pa losses in the core as well as the blanket. However, the most recent French designs have a very hard spectrum so they don't have to worry too much about losses to Pa since at the higher neutron energies the cross section for Pa drops quite low.


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PostPosted: Aug 18, 2008 2:38 pm 
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David,

On the issue of the molten salt selection: has there been any consideration of using a different salt composition for the fuel and blanket salts? Can considering different salts be part of the design space? For example, could a blanket salt be NaF and ZrF while the core salt be standard FLiBe? That might make the salt less expensive; the lower cost might make a larger blanket more feasible, which in turn, might obviate the need for protactinium processing. It might also make the spectrum in the blanket (probably only slightly) harder, which might make lower the cross section of protactinium absorption.

I'm really not sure if these suggestions make sense, but I haven't seen any explicit discussion of different salts compositions for the blanket and the core in your talk.


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PostPosted: Aug 18, 2008 7:35 pm 
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David,

I think it might be useful to get an estimate for the cost of the FLiBe in a 1 GW reactor. I did some estimates which suggested it was tens of millions of dollars. At that level, sizing up by a factor of ten might cost less than a Pa extraction system.

-Iain


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PostPosted: Aug 18, 2008 8:17 pm 
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David wrote:
I got to meet Gary (who contributes on this site) face to face and he had lots of information for me to digest and contacts to seek out.

It was great meeting you. And that was an excellent talk you gave!

David wrote:
He`s trying to make a HW-MSR lover out of me too and made some interesting comments that SiC/SiC tubes might do just fine even in contact with water (i.e. no zircalloy or other extra barrier).

Ratz! Now the word is out that I'm a closet MSR fan. 8^)

-Gary


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PostPosted: Aug 18, 2008 8:26 pm 
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David wrote:
Yes, all reactors will produce roughly the same amount of fission products. Liquid fluoride reactors will actually put out a little less since the thermal efficiency is better than LWRs, meaning there are less fissions per GW year.

That would be an interesting question to look at. If you're doing much on-line separation of FPs, and some of the nasties have appreciable thermal capture XSs, I wonder how well you end up doing in the end.

Actually, I'd bet someone has looked at it, no? The calculation would be easy enough to do, given a fairly simple model of the chemistry. The hard part would be deciding on a metric for "goodness" of the result.

-Gary


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PostPosted: Aug 18, 2008 9:09 pm 
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I haven't seen a lot on the FP - the focus has been on the actinides. The online gas extraction helps to make isolation of some of the nasties easier. There are so many different FP's that I don't imagine solid fuel reactors will destroy much of the FP at all. MSRs have the flexibility but with the possible exception of 4 or so FP's mostly we can just let them decay (about 300 years). I would expect economics will drive that the FP's end up going to a central processing area rather than building a chemical processing plant at each reactor site.


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PostPosted: Aug 19, 2008 9:36 am 
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honzik wrote:
David,

On the issue of the molten salt selection: has there been any consideration of using a different salt composition for the fuel and blanket salts? Can considering different salts be part of the design space? For example, could a blanket salt be NaF and ZrF while the core salt be standard FLiBe? That might make the salt less expensive; the lower cost might make a larger blanket more feasible, which in turn, might obviate the need for protactinium processing. It might also make the spectrum in the blanket (probably only slightly) harder, which might make lower the cross section of protactinium absorption.

I'm really not sure if these suggestions make sense, but I haven't seen any explicit discussion of different salts compositions for the blanket and the core in your talk.



Yes Honzik, your exactly right, it would be a good idea to try to use a less expensive carrier salt for the blanket since we'd like much more volume. 78%NaF-22%ThF4 has been my favorite backup blanket salt which would be incredibly cheap. NaF-ZrF4 also can have substantial amounts of thorium. However, a drawback is that both salts have melting points a little over 600 C with that much thorium. Not a big problem though as we can easily keep its max temp at 700 C since we don't need to draw much heat out of the blanket (say running between 650 and 700). We'd pay a small penalty on neutron losses since Na is inferior to Li7 but probably worth the penalty.


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PostPosted: Aug 19, 2008 9:51 am 
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iain wrote:
David,

I think it might be useful to get an estimate for the cost of the FLiBe in a 1 GW reactor. I did some estimates which suggested it was tens of millions of dollars. At that level, sizing up by a factor of ten might cost less than a Pa extraction system.

-Iain



Yes the unknown cost of enriched Li7 is a huge problem we face. Everyone assumes the price won't be too enormous because there is such a big relative mass difference and Li7 already starts out at 92%. ORNL in the 70s and 80s estimated 120$ per kg of Li7 (99.995 pure I think). A recent Russian paper estimated 800$ per kg. The MSBR needed 49 m3 of salt which would have 13.6 tonnes of Li7 which is around 10 million at the 800$/kg rate. Thus not much relative to other capital costs but still significant. I can't see how it could cost much more than a thousand or two per kg and even then it is not a huge factor.

For 2 Fluid designs we typically only have 20 m3 or even less for the core salt. The blanket salt can get extensive, probably pushing 100 m3. As just mentioned in a previous post, we could save money here by using NaF-ThF4 but even if we stick with LiF-ThF4 the price is not too huge. BeF2 is also a factor if we use it, but it is much cheaper than you might think, about a quarter or less the price of Be metal by Be content (BeF2 is an intermediate stage of Be metal production).


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PostPosted: Aug 19, 2008 10:01 am 
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Gary wrote:
David wrote:
Yes, all reactors will produce roughly the same amount of fission products. Liquid fluoride reactors will actually put out a little less since the thermal efficiency is better than LWRs, meaning there are less fissions per GW year.

That would be an interesting question to look at. If you're doing much on-line separation of FPs, and some of the nasties have appreciable thermal capture XSs, I wonder how well you end up doing in the end.

Actually, I'd bet someone has looked at it, no? The calculation would be easy enough to do, given a fairly simple model of the chemistry. The hard part would be deciding on a metric for "goodness" of the result.

-Gary


Gary,

Did you mean to imply that taking FPs out quicker than other designs might the fission products a bit more radiotoxic since they would not have as long to transmute to something else?

If so, I think the overall effect would be small since in any modern design we'd probably not be trying to take out the fission products nearly as quickly as ORNLs 20 day cycle. Thus if it is 6 months or longer we'd really not have a big difference with other reactors and again a MSR produces less total since the thermal efficiency is lower.

On the other hand, for simplified molten salt designs that look to be just converter reactors by not bothering to remove soluble fission products, we might have a real benefit of radiotoxicity reduction if we can leave FPs in the reactor for years or even decades (such as ORNL's 30 Year Once Through design).

Do you think that might be the case? If you just keep irradiating fission products, does across the board transmutation improve or worsen things. I'd think the long term fission products would be the most interesting to look at. We'd certainly get rid of some, but would we be converting other short term FPs into long term ones?


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PostPosted: Aug 19, 2008 1:46 pm 
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David wrote:
It would be a good idea to try to use a less expensive carrier salt for the blanket since we'd like much more volume. 78%NaF-22%ThF4 has been my favorite backup blanket salt which would be incredibly cheap. NaF-ZrF4 also can have substantial amounts of thorium. However, a drawback is that both salts have melting points a little over 600 C with that much thorium. Not a big problem though as we can easily keep its max temp at 700 C since we don't need to draw much heat out of the blanket (say running between 650 and 700). We'd pay a small penalty on neutron losses since Na is inferior to Li7 but probably worth the penalty.


Thinking about the neutron losses with Sodium: would it be possible to add some BeF2 to the mixture and hope to get some significant reversal on the neutron losses via the n->2n reaction? Could it also help lower the melting point of the salt mixture?


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PostPosted: Aug 19, 2008 2:52 pm 
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David wrote:
Did you mean to imply that taking FPs out quicker than other designs might the fission products a bit more radiotoxic since they would not have as long to transmute to something else?

Well, there are a lot of saturating fission products. Essentially, the ones with largish thermal capture cross sections. So, you'd certainly get a different mix of stuff by separating on-line, but I wouldn't know which would be worse.

Personally, I'm not a big fan of radiotoxicity as a metric. Radiotoxicity is what you have to measure if you don't have a disposal scenario. Once you know the eventual fate of the "waste", you can use dose delivered to the environment (or to man, or whatever) as a metric, and the two can be very different.

-Gary


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PostPosted: Aug 19, 2008 5:05 pm 
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Here's another question: What kind of flow would one expect in the core? Would it be laminar or turbulent? Could there be a transition between laminar and turbulent flow, creating a flow instability? Would the pipe designs include a means of stabilizing the flow transition, if there is one? What about mechanical resonances in the pipes?

Is it possible that the reactivity coefficient w.r.t. temperature be greater (less negative) in turbulent flow where there is more mixing than in laminar flow, where one would expect local changes in temperature to remain relatively local relative to the flow stream?

Finally, if one were to move forward with this design, would there be code available to simultaneously simulate the thermal hydraulic / pipe mechanics / neutronics of the problem?


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PostPosted: Aug 20, 2008 1:07 pm 
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honzik wrote:
David wrote:
It would be a good idea to try to use a less expensive carrier salt for the blanket since we'd like much more volume. 78%NaF-22%ThF4 has been my favorite backup blanket salt which would be incredibly cheap. NaF-ZrF4 also can have substantial amounts of thorium. However, a drawback is that both salts have melting points a little over 600 C with that much thorium. Not a big problem though as we can easily keep its max temp at 700 C since we don't need to draw much heat out of the blanket (say running between 650 and 700). We'd pay a small penalty on neutron losses since Na is inferior to Li7 but probably worth the penalty.


Thinking about the neutron losses with Sodium: would it be possible to add some BeF2 to the mixture and hope to get some significant reversal on the neutron losses via the n->2n reaction? Could it also help lower the melting point of the salt mixture?



Yes, NaF-BeF2 is a great salt too and can hold a lot of thorium without going too high in melting point. One disadvantage over NaF-ThF4 or NaF-ZrF4-ThF4 is that having beryllium will lead to tritium production. A minor point perhaps as we have ways to deal with tritium. The n,2n bonus will be very small in the blanket though as it is for really high energy neutrons and by the time they reach the blanket they will have been slowed a bit.


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PostPosted: Aug 24, 2008 3:38 pm 
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Hi David;

I have not red all the posts on your reactor design or understood fully the ones that I have read. So forgive me for any stupidity or repetition that follows. I am just too excited by your thinking to keep silent any longer and am hoping to contribute.

In the area between the core and the blanket, might it not be possible to fill that area with 60mm silicon carbide coated graphite pebbles to provide for moderation? These pebbles could circulate slowly from top to bottom in an automatic inspection cycle (30 per day) to allow for the safe removal of damaged pebbles without exposing maintenance personnel to radiation.

This automatic cycle would enable continuous reactor operation as the failed graphite pebbles are continuously replaced by fresh new ones.

The undamaged pebbles would be reinserted from the bottom to the top of the reactor to continue the cycle.

Furthermore, at reactor startup, this pebble bed could use standard TRISO 60MM fuel enriched to less then 20% with the amount of TRISO fuel pebbles just enough to begin the thorium fuel ignition.

Once the thorium fuel has ignited, the TRISO fuel is gradually replaced with 60 MM silicon carbide coated graphite pebbles that serve as a moderator of the thorium chain reaction.

Here again, the PBMR like automatic pebble inspection process replaces all deteriorated TRISO and/or graphite pebbles.

A mix of TRISO and graphite pebbles in any proportion is possible to startup or optimize the thorium fuel chain reaction.



This may mitigate the use of LEU 235U (reference slide 44) in the TRISO fuel since it can be simply removed at any time without any impact on the core salt.

This pebble space can be helium cooled to remove heat from the TRISO fuel and/or graphite pebbles to keep this space and the associated reactor walls cool to avoid thermal stress. This space would also provide a safety zone to carry away any salt that leaks from faults in the partition walls during reactor operation.



Thank you; Axil.

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PostPosted: Aug 25, 2008 11:06 am 
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Axil wrote:
Hi David;

I have not red all the posts on your reactor design or understood fully the ones that I have read. So forgive me for any stupidity or repetition that follows. I am just too excited by your thinking to keep silent any longer and am hoping to contribute.

In the area between the core and the blanket, might it not be possible to fill that area with 60mm silicon carbide coated graphite pebbles to provide for moderation? These pebbles could circulate slowly from top to bottom in an automatic inspection cycle (30 per day) to allow for the safe removal of damaged pebbles without exposing maintenance personnel to radiation.

This automatic cycle would enable continuous reactor operation as the failed graphite pebbles are continuously replaced by fresh new ones.

The undamaged pebbles would be reinserted from the bottom to the top of the reactor to continue the cycle.

Furthermore, at reactor startup, this pebble bed could use standard TRISO 60MM fuel enriched to less then 20% with the amount of TRISO fuel pebbles just enough to begin the thorium fuel ignition.

Once the thorium fuel has ignited, the TRISO fuel is gradually replaced with 60 MM silicon carbide coated graphite pebbles that serve as a moderator of the thorium chain reaction.

Here again, the PBMR like automatic pebble inspection process replaces all deteriorated TRISO and/or graphite pebbles.

A mix of TRISO and graphite pebbles in any proportion is possible to startup or optimize the thorium fuel chain reaction.



This may mitigate the use of LEU 235U (reference slide 44) in the TRISO fuel since it can be simply removed at any time without any impact on the core salt.

This pebble space can be helium cooled to remove heat from the TRISO fuel and/or graphite pebbles to keep this space and the associated reactor walls cool to avoid thermal stress. This space would also provide a safety zone to carry away any salt that leaks from faults in the partition walls during reactor operation.



Thank you; Axil.



Axil,

Thanks very much for the enthusiastic comments and ideas. In general yes I have given a fair bit of thought to having graphite moderator between the inner core and outer blanket. I hadn't really thought about using pebbles in this region, there could indeed be advantages but also some things to be careful about. First off, if the barrier has pebbles, that also means empty space and we'd have to be careful about core salt never leaking into the barrier region or that would increase reactivity. That is something that probably could be managed though.

Pebbles would give you an easier way to change the graphite as it became too irradiated but I tend to prefer very simple systems and having a pebble cycling system in this confined space between core and blanket sounds a little complicated in general to me. I'll certainly keep it in mind though. If we want to have graphite moderation, I do tend to prefer pebbles, but right in the central core fluid. We can still cycle them in and out that way too.

In term's of having fuel (and fertile) in these balls and the balls inside the barrier, I don't think that could work though. We'd need to have a completely different, high pressure helium cooling system to take all the heat away from the pebbles. The gas would have to be at high pressure and then that means everything is thick walled. Just taking away a little gamma ray heating of moderator pebbles is one thing, but trying to take away almost all the heat produced is a completely different issue.

Having pebbles that have fissile and fertile material (TRISO) in the central core fluid is a different subject. In this case the salt can take the heat away at low pressure. Any time you start to put fuel or even just fertile elements in the pebbles, you introduce the subject of how do you remove the decay heat from them if you shut down. This would probably not be all that hard but it is certainly a major issue to be sure everything would work well. For the salt, we can transfer that easily to dump tanks. For pebbles though, if they have fission products inside giving off decay heat we have to be sure we can take care of that as well.


Thanks again for the ideas. Even if something isn't directly applicable, it can often lead to new ideas elsewhere.

David L.


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