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PostPosted: Jan 19, 2011 4:58 pm 
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Welcome Erik,

As I plowed more deeply into the graphite issues surrounding the MSBR design that they consider in WASH-1222, one of the key "desire"-ments for the graphite is that it have a low surface porosity to as to keep the occlusion of fission product gases down and to keep the breeding ratio high.

Well, there's more than one way to skin that cat.

Two-fluid design for a LFTR have a very different approach to graphite geometry than the one-fluid design in WASH-1222. Furthermore, if we're not shooting for a really high breeding ratio (like 1.05) then we can afford to relax the surface porosity constraint a bit. That will lead to longer graphite lifetime. If we adopt a core design where fuel is confined to relatively thin graphite channels and bulk moderator graphite is in form of shapes or pebbles that don't come in contact with fuel salt or have a porosity constraint, it can get even better.

So I'm not nearly as worried about graphite lifetime as I was when I first started reading the documents.


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PostPosted: Jan 19, 2011 5:35 pm 
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Kirk Sorensen wrote:
If we adopt a core design where fuel is confined to relatively thin graphite channels....

You mean "thin wall graphite channels", right ? ....or at least thin relative to channel diameter, correct ?


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PostPosted: Jan 20, 2011 8:17 am 
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Yes, thin-wall channels, where the wall thickness is less than half of the diameter of the channel, as opposed to the graphite blocks used in the MSBR design evaluated in WASH-1222, where the salt channel was of a small diameter relative to the overall dimension of the block.


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PostPosted: Jan 20, 2011 9:19 am 
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Kirk Sorensen wrote:
Yes, thin-wall channels, where the wall thickness is less than half of the diameter of the channel, as opposed to the graphite blocks used in the MSBR design evaluated in WASH-1222, where the salt channel was of a small diameter relative to the overall dimension of the block.

Good !

That addresses the FP problem:
I figure 1/8" thick RCC tube wall should be OK for non-pressure application.
Any Xe coming out the other side can be collected from the gas annulus around each tube.

That leaves the remaining issue of neutron flux damage to the bulk moderator -- better avoid graphite !


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PostPosted: Jan 21, 2011 10:56 am 
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Alternatively one could use graphite dust to fill the space between channels. Channels would have to be sleeved somehow so one could swap a channel without spilling the graphite filling. The reactor would likely be little larger than if solid graphite blocks were used as a moderator, but the core life-time may be much larger.

I very much like the HW-MSR concept, but it seems to me longer down the line (Gen6? :) )


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PostPosted: Jan 21, 2011 11:04 am 
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BeO is another candidate for the bulk moderator in this concept. It is neutronically positive due to n,2n > n,g.


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PostPosted: Jan 21, 2011 11:09 am 
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The question then is what material to use to build the sleeve? It sees the same neutron flux and has the same strength requirements.


The lifetime of graphite is set by the change in volume and resulting strength reduction.
The absorption of Xe is a problem for the neutron economy if your goal is to breed the most fuel possible. If your target is an iso-breeder you have a bit of neutron budget (around 6%) to use in engineering cost reductions. One place you might choose to spend it is here (around 0.5%). ORNL was successful at impregnating the graphite against helium penetration in an environment with no neutron flux (and no fission product impacts). The sealing was more than sufficient for the Xe needs. The seal is a deep penetrating seal rather than a surface treatment. The next step would have been to see what happens when the sealed graphite gets exposed to a neutron flux and eventually to fission product impacts.


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PostPosted: Jan 21, 2011 11:13 am 
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Cyril R wrote:
BeO is another candidate for the bulk moderator in this concept. It is neutronically positive due to n,2n > n,g.

Oxygen is something we really do not want running loose inside the reactor. Where does it go when you hit the Be with a neutron? There are lots of things that want the oxygen - are you sure the Be is what wants that oxygen the very most of all?


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PostPosted: Jan 21, 2011 11:30 am 
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BeO is very stable. It really likes oxygen. Some Be will eat a neutron and spew out two others, making it basically two helium nuclei, which will migrate somewhere. The rate of generation is not very big. But you are right the result is somewhat oxidizing conditions. That oxygen has to find a resting place. I think you can use BeO infused with Be metal as integrated metallic oxgyen getter, or just use Be metal only. You have to be somewhat careful about peak temperatures but Be metal has higher thermal conductivity than BeO at elevated temperatures, and a rather high melting point. Be metal also wants to react a bit with the graphite. Probably it will form beryllium carbide. I don't think that's a problem, but who knows. Beryllium carbide might be an option in itself. It has a minor disadvantage of being incompatible with some substances (IIRC water; BeO is more stable). BeF2 is another option, perfectly compatible with graphite. But it will undergo plastic deformation and this will have to be contained. Lots of options, but I don't have a good feel how the pros and cons weigh out.

Regarding the xenon problem. From the ORNL documents it appears the graphite dilatates uniformly upon irradiation, so that little cracking is caused, just a lower density with correspondingly increased permeability. ORNL was confident it would still be acceptable.

If not, then perhaps there are purging options to be considered to increase useful channel lifetime (if xenon poisoning is indeed the limiting factor). The Xenon probably does not penetrate deeply. Perhaps the channels can be vacuumed to suck some Xenon out periodically. In Jaro's siphon design the vacuum equipment is present. Alternatively, the channels can be removed by overhead crane and one end blocked mechanically, then pull a vacuum on them to suck the xenon out.


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PostPosted: Jan 21, 2011 11:43 am 
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Cyril R wrote:
From the ORNL documents it appears the graphite dilatates uniformly upon irradiation, so that little cracking is caused, just a lower density with correspondingly increased permeability. ORNL was confident it would still be acceptable.

The graphites ORNL was working with shrank first then expanded. They defined the lifetime of the graphite as when the graphite expanded back to its original size. That is the definition that limits the graphite lifetime in a reactor. Xe has nothing to do with it. Under the definition of removing the graphite when it returns to its original size yes ORNL was confident that the graphite would be fine.

But if you want to increase the graphite lifetime beyond this you are in new territory where ORNL did not go.

The graphite increases volume because in one dimension fresh graphite has tightly packed plates. When a neutron bumps into a carbon atom it knocks it around and frequently displaces it to reside between plates - increasing the volume. If you had loosely packed carbon to start with I could imagine it would not increase the volume so the concept of graphite dust etc. likely would be fine. But you need a container to hold the dust and so we have simply moved the problem. The container wall needs to meet the same requirements that the original graphite needed to meet.


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PostPosted: Jan 21, 2011 12:44 pm 
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RCC products are made by a CVD process (Chemical Vapour Deposition) into a carbon fiber "pre-form".
The process takes a long time -- particularly when the highest-density finished product is desired.
This makes it expensive.

If we just want tubes for channeling liquid flow, then the ultimate in density is NOT necessary -- only adequate mechanical properties (strength, stiffness) are required.

Somewhat porous, thin tube walls may actually be desirable -- both for avoiding Xe accumulation, and for reducing production costs. Perhaps helium could help speed the Xe through the wall ?

For the case of the HW-MSR, operating with vacuum siphons, this is probably a non-issue, since any significant gas seepage is likely to be inward from the outside of the fuel tubes -- ie. from the surrounding gas annulus, which is at ambient pressure.....

Or to put it in the form of a question: how much Xe can be absorbed in a porous wall with an adverse gas pressure gradient inside ?


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PostPosted: Jan 21, 2011 5:01 pm 
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Actually xenon retention could have something to do with effective graphite lifetime. If the retention is severe isobreeding will be inhibited, so you'd have to swap the graphite before it gets too clogged up with xenon. This is probably not that big a problem though, so the limiting factor factor will likely be graphite swelling.

Jaro's design does take care of the moderator retention issue by using a calandria, physically decoupled from the reactor. Could use BeO or D2O.


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PostPosted: Jan 23, 2011 11:38 am 
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Cyril R wrote:
....take care of the moderator retention issue by using a calandria, physically decoupled from the reactor. Could use BeO or D2O.

BeO sounds appealing at first thought (appart from other issues such as cost and toxicity - shared to some extent with D2O).
But as with other fuel-channel-type reactors with solid moderator (typically graphite), I see problems with cooling the bulk material.
Convective and radiative heat transfer from the fuel channels to the moderator can be addressed adequately.
But neutron and gamma heating will happen no matter what.
If there is NO cooling, then extreme temperatures are likely -- reversing the flow of radiative heat transfer, such that the fuel will become the coolant.
There may not be any calandria material for containing the BeO that can withstand such high temperatures (ditto for the graphite powder suggested by others).
Independent gas cooling of solid moderator is undesirable as it introduces complexity and reduces average moderator density, which increases neutron leakage (even worse for powdered graphite).
Moreover, unless the cooling gas is pressurized, the cooling is not very effective -- particularly when compared to cooling a liquid moderator, like D2O, which is simplicity itself.


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PostPosted: Jan 23, 2011 1:23 pm 
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That's very perceptive of you Jaro. The BeO itself may be able to take the high temps - thermal conductivity and melting point are excellent - but the zirconium alloy isn't going to like high temperatures. It is similar to fuel rods and calandria tubes: the latter last a lot longer due to higher strength at <100 C. Using BeO might be similar to asking fuel rods to last decades; a major challenge.


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