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PostPosted: Dec 12, 2011 1:54 pm 
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@Cyril In the paper they say the neutronics support up to 40% burnup (400 GWd/tHM) without reprocessing.

@Jaro Where did you get the 3-D model? Are you sure you're not looking at the demo plant, which has fast flux test ports and other access for maintenance?

By the way, by some accounts the thermal U-Th reactor at Ft St Vrain achieved 170 GWd/t burnup. Actually it's by WNA's account, but when I asked them they couldn't produce a reference. Wikipedia says 90 GWd/t.

Edit: I wonder if anyone has ever tried making an amorphous metal out of DU. Zr is a common alloying element in amorphous metals. At least, displacements per atom may not be so structurally damaging to them.

-Carl


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PostPosted: Dec 12, 2011 2:31 pm 
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Fort St. Vrain used TRISO type fuel which is known to facilitate high burnup without venting. This is done by leaving a lot of space full of porous carbon where the fission gas can go into and by having a very strong ceramic coating.

Travelling wave reactor uses a different fuel, metal fuel with metal cladding. This fuel is cheaper to make and easier to reprocess, but can it take 550 dpa? In particular can the cladding take it? It's okay if the fuel gets smashed into goo since it only has to sit there, lots of space for swelling by the metal bonding technology which is awesome. The cladding however must perform a containment function for most of the fission products, even if it is vented fuel. Embrittlement point creeps up to higher and higher temperatures with more lattice displacement, when the reactor operating temperature has been reached it gets ugly. Recently I read about a lead cooled fast reactor fuel design, they were counting on 4-8 dpa and then they had to swap out the fuel because the cladding couldn't take any more.

How many dpa do the state of the art ODS zirconium claddings achieve?


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PostPosted: Dec 12, 2011 3:08 pm 
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Ft St Vrain had prismatic graphite fuel embedded with both fissile and fertile (thorium) TRISO particles. In fact it contained over 20 times more thorium than uranium. The fuel is said to be much cheaper to produce than normal fuel rods. Another spec sheet says it was rated to 100 GWd/t (6 years in the core). The reactor containment was prestressed concrete and the plant achieved 40% thermal efficiency. To my eye, simply iterating on this design would be a better engineering target to what TerraPower are undertaking. General Atomics is currently pushing a design called EM^2, but it is a fast reactor.

-Carl


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PostPosted: Dec 12, 2011 4:43 pm 
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clumma wrote:
@Jaro Where did you get the 3-D model? Are you sure you're not looking at the demo plant, which has fast flux test ports and other access for maintenance?

Its a still frame from the presentation video.

Yes, it is the demo plant -- as you can see in the photo, "TP-1" ( 1200 MWth).

Yes, there is some extra instrumentation compared to the 3000 MWth commercial model ("TPRP").
But the large access ports in the top of the dome will always be needed - just as they were needed in Phenix & Superphenix - for removing damaged equipment like the fuelling machine and the main instrument column.
You don't just build another plant, whenever a component becomes defective !

PS. in this design, the primary sodium pumps and HX's are located INSIDE the pool, next to the core: Maybe you're confusing those with "extra instrumentation" ??


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PostPosted: Dec 13, 2011 12:20 am 
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To the best of my understanding, they are still very uncertain about the fuel cladding since materials aren't usually tested beyond 500 dpa (because most materials fail before they reach that point). Though the HT9 alloy shows some promise, they are open to the idea of fabricating a new material. They are still testing things. However, neutron irradiation testing is hard to do, so they make an approximation by hitting the material with Fe+, a process that is also referred to as ion implantation.

So much abuse, poor HT9. :-/

_________________
"To know that we know what we know, and to know that we do not know what we do not know, that is true knowledge." ~ Copernicus


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PostPosted: Dec 13, 2011 4:01 am 
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clumma wrote:
Ft St Vrain had prismatic graphite fuel embedded with both fissile and fertile (thorium) TRISO particles. In fact it contained over 20 times more thorium than uranium. The fuel is said to be much cheaper to produce than normal fuel rods. Another spec sheet says it was rated to 100 GWd/t (6 years in the core).


It is more expensive fuel to buy, but the high burnup makes up for that. With a sufficiently high burn the economics become actually better, despite the higher initial fuel investment, due to less downtime for refuelling. Also like you say the fuel operates hotter which makes the power conversion more efficient.

Quote:
The reactor containment was prestressed concrete and the plant achieved 40% thermal efficiency. To my eye, simply iterating on this design would be a better engineering target to what TerraPower are undertaking. General Atomics is currently pushing a design called EM^2, but it is a fast reactor.


This is the philosophy that led to the development of the advanced high temperature reactor at UCB. They swapped out high pressure gas coolant, which is less economical and less safe, for a low pressure fluoride coolant, which is safer and more economical. Dr. Per Peterson's group is working on a pebble bed version.

http://pb-ahtr.nuc.berkeley.edu/index.html


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PostPosted: Dec 13, 2011 4:21 am 
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Yeah, I like the FHR/AHTR concepts quite a lot, and I agree the atmospheric-pressure operation (and greater thermal capacity) of molten salt is an advantage. Still, having thousands of spherical fuel elements bobbing around in there concerns me for some reason. I assume the moderation provided by the FLiBe results in a negative void coefficient for the thing... but I guess I should read up on it more. -Carl


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PostPosted: Dec 13, 2011 7:25 am 
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So far the neutronic analysis of the AHTRs have shown that FLiBe has negative void. More importantly, the total coefficient is negative.

The advantage of pebble fuel is easy shuffeling of pebbles, reducing fissile requirements, and lack of complex core internals and fuel (they are all the same little balls, amenable to mass manufacture). Also the power density is higher for pebble fuel.

Regarding the dpa rates, what do CANDU pressure tubes take? They seem to last quite long (though not 60 years) right in the middle of high flux.


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PostPosted: Dec 13, 2011 1:15 pm 
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Cyril R wrote:
Regarding the dpa rates, what do CANDU pressure tubes take? They seem to last quite long (though not 60 years) right in the middle of high flux.
Can't quote a dpa number off hand.
But the PTs will last at most 25 years (hence the currently on-going refurb projects at several old CANDU plants: Korea's Wolsong-1 restarted a few months ago, with a core full of new PTs)
Also, the neutron flux is likely to be considerably less damaging than an FBR: the fuel is oxide, not metal, and there is light water coolant in the PTs.


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PostPosted: Dec 13, 2011 2:24 pm 
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Thanks Jaro. Do you know if CTs last longer than PTs? With the lower pressure and temperature, they should last much longer, right? Or do they just get replaced with the PTs?

Vented TWR fuel would be quite low pressure, and sodium is chemically much nicer on metals than water, but the TWR cladding would operate at much higher temperature than Candu PTs.

I'd say the radiation damage to the cladding is THE biggest technical issue for the TWR. Not the fuel actually, that can swell as much as it wants, just displace the sodium fuel-cladding bond up into the plenum. Metal fuel might be better than oxide fuel, with its better thermal conductivity keeping temperatures low, and metal self annealing if it gets too hot locally. Whereas oxide fuel, while it might suffer less radiation swelling damage, will run very hot and not anneal itself out, more likely the opposite happens, when it cracks it further overheats from void thermal insulation, exagerbating fuel damage.


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PostPosted: Dec 16, 2011 3:49 am 
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It would be interesting to know how the CANDU pressure tubes would behave at lower pressure but higher temperature (500C) with a coolant like a salt eutectic (SnF2-PbF2) or a liquid metal (Mg-Al). Everyone talks about different fuels in Calandria configuration but the only coolants considered have been water or heavy water.
I hope that Mr Gate's computations show that irradiation can convert enough fertile matter for the wave to continue and proceed in his TWR.


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PostPosted: Dec 16, 2011 1:36 pm 
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jagdish wrote:
It would be interesting to know how the CANDU pressure tubes would behave at lower pressure but higher temperature (500C) with a coolant like a salt eutectic (SnF2-PbF2) or a liquid metal (Mg-Al). Everyone talks about different fuels in Calandria configuration but the only coolants considered have been water or heavy water.


The pressure tubes are made of zirconium niobium alloy. Zirconium gets severely attacked from fluoride melts, expecially less stable ones such as SnF2-PbF2. Though it would be interesting to know the solubility of ZrO2 in a fluoride melt. Such passivation protection, even if ZrO2 is insoluble in fluoride melts, is not likely to be effective in a molten salt cooled or fuelled reactor, because it is not likely to be practical to maintain makeup oxygen to keep the passivation layer intact.

Niobium however is compatible with fluorides, similar resistance as Hastelloy N. It does cost you neutrons in a thermal reactor.

Mg-Al is a horribly corrosive bile, and highly flammable when molten.

Lead is likely to work as a coolant with zirconium-niobium alloys. And you have a practical oxygen supply, by putting some in the helium reactor cover gas, for passivation upkeep. Lead also doesn't burn in air or water.


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PostPosted: Dec 16, 2011 3:19 pm 
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Quote:
Everyone talks about different fuels in Calandria configuration but the only coolants considered have been water or heavy water.


Not true, AECL came pretty close to implementing a low pressure, high temperature (425 outlet) oil as the coolant (cool heavy water still the moderator). They even ran a reactor with it in the 60s. Great stuff overall, many still lament it being abandoned. The fact that the stuff burns always threw me off but I never really looked into it too deeply.

David LeBlanc


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PostPosted: Dec 17, 2011 12:22 am 
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It may have been abandoned due to instability of oils at high temperature. I read about it once and then proceeded to search for more of less volatile moderator/coolants to reduce the pressure in tubes/reactor vessels. A per-fluorocarbon lubricant, Krytox (poly hexa-fluoro propyl ether) came nearest. I have not been able to convince many about it. Mg and Al have higher ignition points than Sodium and I have had some encouraging feedback. Tin or lead di-fluorides would be non-moderating but no more corrosive than other fluorides. Tetra-fluorides could be oxidizing. You could always have Nickel or copper coating if found necessary.


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PostPosted: Dec 17, 2011 6:32 am 
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Tin and lead fluorides are very corrosive. In fact you'd be better of with pure lead coolant than lead fluoride.

Lead is a pretty good coolant for a fast reactor. Lead is corrosive to nickel alloys but not to zirconium, niobium, molybdenum with proper oxygen control. Would be nice for a CANDU as well, especially if using radiogenic lead, from thorium deposits, when thorium decays it eventually becomes Pb-208, the best lead isotope for reactor coolant. Lead doesn't moderate, so in a lead cooled heavy water moderated CANDU it means you have a bimodal spectrum and can breed much better with both U-Pu and Th-U. Lead is compatible with the zirconium alloys used in CANDUs so you can just make the entire primary loop out of that, and maybe a TZM steam generator tubing.


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