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PostPosted: Aug 08, 2008 9:02 am 
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Molten Salt Reactors: A New Beginning for an Old Idea

Dr. David LeBlanc, Carleton University

12:00 Noon, Friday August 15, 2008
Library Auditorium
AECL - Chalk River Laboratories


Quote:
This talk will be open to all AECL staff and other personel who have access to the Chalk River site.


....sounds like this may be a good opportunity to meet Gary and a few other interested parties who couldn't make it to your last CRL presentation.

Looking forward to your (& others') comments on how it went.
(I don't expect to be in CRL on that date, unfortunately)


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PostPosted: Aug 15, 2008 12:44 pm 
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So how'd it go this time, David ?

No power outage again, I hope ?

According to this photo, there was quite a bit of traffic on Chalk River plant road.... no doubt all heading to your seminar.


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PostPosted: Aug 15, 2008 8:29 pm 
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Jaro,

My talk went very well, a good turnout and very well received. I can`t think of a single person that had a negative point of view in either talk up at Chalk. I am always pleased by the reaction of some of the younger scientist and engineers, they are working on all the problems and engineered solutions needed for LWR and CANDU designs and they can`t believe what molten salt reactors can do inherently. One young man in particular today couldn`t get over how we don`t need control rods to control power levels, we just draw more heat out of the flowing salt and that increases power for us due to the negative temperature reactivity coefficient.

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. 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).

While almost all AECL scientists or engineers seem very receptive, I still am having difficulty getting heard by the upper level folks that really decide things. Perhaps I should follow the advice of someone at the CNS conference back in June and get a petition going!

Here is a copy of my power point presentation. As always some slides are difficult to follow with only the slides to go by. If anyone has questions, feel free to ask.


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PostPosted: Aug 15, 2008 10:12 pm 
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David wrote:

Here is a copy of my power point presentation. As always some slides are difficult to follow with only the slides to go by. If anyone has questions, feel free to ask.


Regarding slide titled "Radiotoxicity PWR vs FBR* vs MSR"

Table labels vs title

PWR = PWR?
U/PU = FBR?
Th/U (thermal) = MSR?

Is it correct to say Th/U waste is ~1000 times less toxic than PWR waste? (Seems to hold for all time frames).


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PostPosted: Aug 15, 2008 10:24 pm 
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Hey David, I just skimmed the presentation but it looks great! This is fantastic work that you're doing!


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PostPosted: Aug 15, 2008 11:25 pm 
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Slide #23:
Quote:
Graphite now proven to be compatibility with fluoride salts
should be either
Graphite now proven to be compatible with fluoride salts, or
Graphite now proven compatibility with fluoride salts.
Slide #43:
Quote:
Hasteloy
should be Hastelloy.
Slide #52:
Quote:
Year 20+ Build them by the thousands…
Woo-hoo!


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PostPosted: Aug 16, 2008 12:23 am 
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charlesH wrote:
David wrote:

Here is a copy of my power point presentation. As always some slides are difficult to follow with only the slides to go by. If anyone has questions, feel free to ask.


Regarding slide titled "Radiotoxicity PWR vs FBR* vs MSR"

Table labels vs title

PWR = PWR?
U/PU = FBR?
Th/U (thermal) = MSR?

Is it correct to say Th/U waste is ~1000 times less toxic than PWR waste? (Seems to hold for all time frames).


Charles,

Yes, PWR without Pu recycle, U/Pu represents a liquid sodium fast breeder reactor that loses 0.1% of actinides (Np,Pu,Am,Cm) each time it processes fuel and Th/U is a thermal spectrum molten salt (liquid fluoride) that also loses 0.1% each time it processes fuel.

The molten salt design is so much lower for actinide radiotoxicity (the really long lived nasty part) because it has so little in the salt (maybe 100 kg) versus the many tonnes that a FBR needs to reprocess. The data is from a paper and more specifically from a presentation by S. David from France. A very interest add on to that data is that the leading fast breeder work is the IFR out of Argonne and they don`t come near a loss of 0.1% during reprocessing. Their goal seems to be around 0.5% and have rarely done better than 1 to 2% loss.

However, be careful how you read things. The MSR (LFTR) is not a 1000 times lower in radiotoxicity overall, but 1000 times lower for the actinide wastes. In the first few hundred years it is the fission products that are the strongest. The big benefit of molten salt operation is that the long term waste (after 300 to 1000 years) is much, much lower than the waste of LWRs, FBRs or LWRs with Pu recycle (the last option not shown) for 10s of thousands of years. Add to that you don`t have a plutonium mine for the next 1000 generations and pretty much any molten salt design seems even more attractive.


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PostPosted: Aug 16, 2008 12:27 am 
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Bill wrote:
Slide #23:
Quote:
Graphite now proven to be compatibility with fluoride salts
should be either
Graphite now proven to be compatible with fluoride salts, or
Graphite now proven compatibility with fluoride salts.
Slide #43:
Quote:
Hasteloy
should be Hastelloy.
Slide #52:
Quote:
Year 20+ Build them by the thousands…
Woo-hoo!


Thanks Bill, I am not the best proof reader but I did notice and enjoy the Woo-Hoo, the most sought after of responses...


Last edited by David on Aug 16, 2008 11:09 am, edited 1 time in total.

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PostPosted: Aug 16, 2008 1:39 am 
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David wrote:
charlesH wrote:
David wrote:

Here is a copy of my power point presentation. As always some slides are difficult to follow with only the slides to go by. If anyone has questions, feel free to ask.


Regarding slide titled "Radiotoxicity PWR vs FBR* vs MSR"

Table labels vs title

PWR = PWR?
U/PU = FBR?
Th/U (thermal) = MSR?

Is it correct to say Th/U waste is ~1000 times less toxic than PWR waste? (Seems to hold for all time frames).


Charles,

Yes, PWR without Pu recycle, U/Pu represents a liquid sodium fast breeder reactor that loses 0.1% of actinides (Np,Pu,Am,Cm) each time it processes fuel and Th/U is a thermal spectrum molten salt (liquid fluoride) that also loses 0.1% each time it processes fuel.

The molten salt design is so much lower for actinide radiotoxicity (the really long lived nasty part) because it has so little in the salt (maybe 100 kg) versus the many tonnes that a FBR needs to reprocess. The data is from a paper and more specifically from a presentation by S. David from France. A very interest add on to that data is that the leading fast breeder work is the IFR out of Argonne and they don`t come near a loss of 0.1% during reprocessing. Their goal seems to be around 0.5% and have rarely done better than 1 to 2% loss.

However, be careful how you read things. The MSR (LFTR) is not a 1000 times lower in radiotoxicity overall, but 1000 times lower for the actinide wastes. In the first few hundred years it is the fission products that are the strongest. The big benefit of molten salt operation is that the long term waste (after 300 to 1000 years) is much, much lower than the waste of LWRs, FBRs or LWRs with Pu recycle (the last option not shown) for 10s of thousands of years. Add to that you don`t have a plutonium mine for the next 1000 generations and pretty much any molten salt design seems even more attractive.


OK, so the FP (fission products) is the same for all types. Correct?

"A very interest add on to that data is that the leading fast breeder work is the IFR out of Argonne and they don`t come near a loss of 0.1% during reprocessing. Their goal seems to be around 0.5% and have rarely done better than 1 to 2% loss."

By loss you mean long term waste getting outside the reactor into the waste stream? Thus 0.1% is better than 0.5%?

May I use the slide in my discussions with others (e.g. James Hansen)? If yes, should I link to you as the source?


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PostPosted: Aug 16, 2008 6:45 am 
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Excellent stuff!
Can this be shared?


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PostPosted: Aug 16, 2008 8:18 am 
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meiza wrote:
Excellent stuff!
Can this be shared?


Yes, could I upload it to the pdf document list?


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PostPosted: Aug 16, 2008 9:17 am 
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Kirk Sorensen wrote:
Yes, could I upload it to the pdf document list?

How about linking to it, on the CNS-Chalk River Branch web site:

http://www.cns-snc.ca/branches/ChalkRiver/past_speak/DLeblanc_MSR_ChalkAug15.pdf


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PostPosted: Aug 16, 2008 10:31 am 
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Quote:
OK, so the FP (fission products) is the same for all types. Correct?

"A very interest add on to that data is that the leading fast breeder work is the IFR out of Argonne and they don`t come near a loss of 0.1% during reprocessing. Their goal seems to be around 0.5% and have rarely done better than 1 to 2% loss."

By loss you mean long term waste getting outside the reactor into the waste stream? Thus 0.1% is better than 0.5%?

May I use the slide in my discussions with others (e.g. James Hansen)? If yes, should I link to you as the source?


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.

0.1% loss means if you are reprocessing fuel (solid or liquid fluoride) that has 1000 kg of transuranics (Np,Pu etc) then you will lose about 1 kg to a waste stream since chemical reactions or processes are hard to be 100%. One of the leading proposed reprocessing system for fast breeders is the pyroprocessing method of the Integral Fast Reactor of Argonne and it does worse than 0.1%. Their process has many proliferation advantages compared to the PUREX process that others propose. I am not sure what losses of transuranics that PUREX experiences but it is by no means a "clean" process.

I should have referenced where I took that graph as it comes from S. David of France. Here a link to one of his power point presentations. Slide 20 shows both a thermal and fast spectrum Th-U MSR. You have a bit more radiotoxicity with a faster spectrum MSR because you have more transuranics in the salt. I can`t seem to find the exact presentation he had the simpler graph that I used. He wrote a paper on this several years ago, I should try to get that as well.

http://ipnweb.in2p3.fr/~pacs/pacs/activites/scenarios-systemes/INPC2004-sdavid.pdf


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PostPosted: Aug 16, 2008 11:13 am 
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Kirk Sorensen wrote:
meiza wrote:
Excellent stuff!
Can this be shared?


Yes, could I upload it to the pdf document list?



Sure thing, readers should bear in mind the slides have many things that are a bit oversimplified that I attempt to explain better in the real talk.


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PostPosted: Aug 17, 2008 8:55 pm 
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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.


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