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PostPosted: Aug 02, 2012 10:41 am 
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Owen T wrote:
djw1 wrote:
Current theory is that moderator life is a function only of fluence and temperature.
Assuming fixed temperature, you can only get so much energy out of a kg of graphite.
From an economic point of view, everything else being equal, high power density
and short moderator life is cheaper than low power density and long moderator life.
You want to turn over your expensive inventory of graphite rapidly,

Reactor grade graphite costs up to $20/kg. It seems to me that the way forward
is high power density, short moderator life, and after a decay period recycle. Re-sintering
moderately radioactive graphite can't be that big a problem.


It's not just sintering. I believe it will need to be cooked for a while in vacuum to get any adsorbed xenon and krypton out and then re-densified with chemical vapor deposition to get the outer surface to be smooth and nonporous again.

Cooking radioactive stuff in high pressure flammable gas? Sounds like fun. Should be feasible, but it's not going to be too cheap.

I've never considered the cost of the graphite as signicant. djw1, how much would you expect to spend per GWe-yr replacing graphite? I think the bigger issue is to ensure that the reactor down time is minimized.

Talking with ORNL they suggest that in essence we would have to grind the graphite back to dust, add a binder and then reform it (pressure, temperature, and vacuum). I think there is no chance that reformed graphite is cheaper than new stuff. But PERHAPS if the machine can be sealed and automated and built in a low throughput form we could put such a machine at each site and over four years reform the graphite from the previous cycle. That would be a way to reduce the graphite waste and sounds politically attractive. Note that this will not reduce the total number of 14C atoms created but it will reduce the total volume of waste graphite. It is my presumption that the waste graphite will be treated as high level waste but very low radiation and heat generation so it should be quite suitable to geological disposal.


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PostPosted: Aug 02, 2012 11:51 am 
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Agree on cost. Based on a moderator life of 4 years,
we are estimating 0.015 cents/kWh for the graphite.
In the noise. My poorly stated point was that long moderator life
gains you nothing in levelized cost, and in fact
costs you a little due to the discounting process.

If we did not have the TRU waste problem,
there would be no point in attempting moderator recycling.
But since we do its worth looking into.

Cyril's idea of shaving off the surface raised in another thread
is not supported by the MSRE.
It is true that with one exception almost all the FP
in the MSRE graphite samples were found in the outer 0.25 mm.
See ORNL-TM-7207 Figure 14.
Unfortunately, that one exception is our old friend 137-Cs.
Most of FP in the sample appears to be 137-Cs
and it was spread more or less evenly through the whole 6 mm.
Cesium appears to have the ability to diffuse thru graphite
pretty well at operating temperatures. Perhaps
the solid state guys can tell us why,
and ideally come with a counter. If we could
somehow avoid the cesium, then recycling becomes
a lot easier, in part because almost all the other half-lives
are of the order of 30 days.

Jack


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PostPosted: Aug 02, 2012 2:00 pm 
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djw1 wrote:
Cyril's idea of shaving off the surface raised in another thread
is not supported by the MSRE.


Waste wasn't an issue in during the MSRE construction and operation. Burying was considered acceptable. And it is acceptable of course, since graphite is inert under conditions found in rocks and soil. Since then waste has become a perceived issue. It still isn't an issue. Just a silly perception, somehow unique to nuclear power (fossil plants can throw most of their waste in the atmosphere and store the rest openly in the biosphere and no one really cares enough about it to stop it).

Quote:
It is true that with one exception almost all the FP
in the MSRE graphite samples were found in the outer 0.25 mm.
See ORNL-TM-7207 Figure 14.
Unfortunately, that one exception is our old friend 137-Cs.
Most of FP in the sample appears to be 137-Cs
and it was spread more or less evenly through the whole 6 mm.
Cesium appears to have the ability to diffuse thru graphite
pretty well at operating temperatures. Perhaps
the solid state guys can tell us why,
and ideally come with a counter. If we could
somehow avoid the cesium, then recycling becomes
a lot easier, in part because almost all the other half-lives
are of the order of 30 days.


Not a solid state guy, but I know a few tidbits of information that could help explain things. Cs-137 has a gaseous parent so that makes sense. Also I know that - unlike molten fluoride salts - liquid alkali metals wet graphite, which would explain the diffusion behaviour (along with the fact that cesium would be liquid under all operating conditions of a MSR).

Cs-137 is however very easy to remove. If you have to recycle bulk graphite, that means you need to re-sinter it, which means handling powder. If you're doing that, it's not a major hassle to add a (vacuum) distillation step to remove the cesium. Graphite after all is about as nonvolatile as it gets.


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PostPosted: Aug 02, 2012 3:23 pm 
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Believe it or not the 14C is the isotope they worry about!! But burying it deep should be pretty easy since there is no heat load or radiation load. Recycling the graphite would be mostly PR. But we need to remember that PR has real value in today's world.

In any event, even if we scrape away the surface we probably remove 99% but not 100% of the TRUs. So, while this is a reasonable approach in a sane world I suspect it won't make any difference in the real world.


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PostPosted: Aug 02, 2012 5:55 pm 
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Lars wrote:
In any event, even if we scrape away the surface we probably remove 99% but not 100% of the TRUs. So, while this is a reasonable approach in a sane world I suspect it won't make any difference in the real world.


Cs-137 isn't a TRU. Why would TRU's go into the graphite to any depth? These will not be slammed into the graphite like the fission products do. Nor have they mobile radionuclide parents. Plus, they're all present as stable fluorides, which stay in the salt (and don't wet graphite). If there are any TRUs in the graphite at all, you'd expect them in a very shallow layer. It's not really a % game, more like ppm, possibly even ppb.


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PostPosted: Aug 02, 2012 8:35 pm 
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Good point.

So we wait a couple of years, we scape off the outside, grind up what's left, heat it up to get rid of the cesium,
what's left? A bit of 144Ce, a weak beta emitter? How hard can it be?

Somebody needs to do the numbers.

Jack


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PostPosted: Aug 02, 2012 9:20 pm 
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Cyril R wrote:
Lars wrote:
In any event, even if we scrape away the surface we probably remove 99% but not 100% of the TRUs. So, while this is a reasonable approach in a sane world I suspect it won't make any difference in the real world.


Cs-137 isn't a TRU. Why would TRU's go into the graphite to any depth? These will not be slammed into the graphite like the fission products do. Nor have they mobile radionuclide parents. Plus, they're all present as stable fluorides, which stay in the salt (and don't wet graphite). If there are any TRUs in the graphite at all, you'd expect them in a very shallow layer. It's not really a % game, more like ppm, possibly even ppb.


So if it is ppb what do you expect the regulators will say on how you may dispose of the "plutonium contaminated" graphite?

Like I said, in a sane world we could put it in a cave but in our political world we likely will have to argue about potential dangers even if we put it in a salt dome or in deep boreholes. A good hint is France's declaration that graphite was high level waste (and I presume that is just from the 14C levels since they haven't run a molten salt reactor I'd guess the graphite had no contact with fission products or plutonium ).


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PostPosted: Aug 03, 2012 5:49 am 
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Lars wrote:
Cyril R wrote:
Lars wrote:
In any event, even if we scrape away the surface we probably remove 99% but not 100% of the TRUs. So, while this is a reasonable approach in a sane world I suspect it won't make any difference in the real world.


Cs-137 isn't a TRU. Why would TRU's go into the graphite to any depth? These will not be slammed into the graphite like the fission products do. Nor have they mobile radionuclide parents. Plus, they're all present as stable fluorides, which stay in the salt (and don't wet graphite). If there are any TRUs in the graphite at all, you'd expect them in a very shallow layer. It's not really a % game, more like ppm, possibly even ppb.


So if it is ppb what do you expect the regulators will say on how you may dispose of the "plutonium contaminated" graphite?

Like I said, in a sane world we could put it in a cave but in our political world we likely will have to argue about potential dangers even if we put it in a salt dome or in deep boreholes. A good hint is France's declaration that graphite was high level waste (and I presume that is just from the 14C levels since they haven't run a molten salt reactor I'd guess the graphite had no contact with fission products or plutonium ).


A few ppb plutonium in graphite is roughly similar in radioactivity to granite (ranging 4000 - 60000 ppb U in granite, not to mention its daughters like radon). Granite has about 1000 Bq/kg. Coffee also has that.

What is the 14C activity per ton graphite? I don't know if this matters since it is the carbon that we're recycling in the first place. The "hot" surface area that gets grinded off is definately high level waste, but it isn't a big volume (less than 1% of the total graphite). If some C14 is in that it doesn't add to the difficulty of storage (there are fission products in this waste).


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PostPosted: Aug 03, 2012 7:56 am 
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Lars wrote:
A good hint is France's declaration that graphite was high level waste (and I presume that is just from the 14C levels since they haven't run a molten salt reactor I'd guess the graphite had no contact with fission products or plutonium ).
That declaration was based on chlorine-36, not C-14.
Chlorine is used to purify graphite during manufacture.
In a reactor the residual chlorine-35 yields an activation product, Cl-36, which is a long-lived (301,000 years) energetic (709 keV) beta emitter that is highly soluble in water.
Graphite wastes have an activity of roughly 1 MBq/kg, from Cl-36.
Even though C-14 is 500 times more radioactive, France's CNE said that it is Cl-36 that poses the problem in a repository because of its extreme mobility in the biosphere.


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PostPosted: Aug 03, 2012 9:11 am 
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jaro wrote:
Lars wrote:
A good hint is France's declaration that graphite was high level waste (and I presume that is just from the 14C levels since they haven't run a molten salt reactor I'd guess the graphite had no contact with fission products or plutonium ).
That declaration was based on chlorine-36, not C-14.
Chlorine is used to purify graphite during manufacture.
In a reactor the residual chlorine-35 yields an activation product, Cl-36, which is a long-lived (301,000 years) energetic (709 keV) beta emitter that is highly soluble in water.
Graphite wastes have an activity of roughly 1 MBq/kg, from Cl-36.
Even though C-14 is 500 times more radioactive, France's CNE said that it is Cl-36 that poses the problem in a repository because of its extreme mobility in the biosphere.


Long lived means low activity, beta is a less dangerous type of radiation, and chlorine doesn't bioaccumulate (in fact cycles very rapidly out of the body). This sounds like another nonsense radiophobia scaremongering thing. Lets look up the activity of natural uranium, hmm its more than 10 MBq/kg. High level waste, yeah right.

Of course, chlorine only occurs in volatile forms in graphite, so it will fly out along with the cesium removal step (and will be washed out along with the cesium in a closed loop scrubber). Plus, Kirk mentioned that there are alternatives to using chlorine in the production process, possibly F to replace Cl would work even better.

In other words we can design it out of the equasion.


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PostPosted: Aug 03, 2012 5:12 pm 
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The Cesium and chlorine will combine into CsCl which is medically valuable.
And then somebody will bring up Goiana. You cant win.

More to the point, the US limit for TRU is 3.7MBq of alpha emitting TRU
(less than Cyril''s number for uranium). If we can get below this number,
at least for the non-surface material, then we dont ahve TRU waste even in the US.

If we cant get below this number, we should be able to safely re-cycle.
Even the most obtuse regulator cant object to putting the graphite
back in the reactor.


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PostPosted: Aug 04, 2012 11:07 pm 
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If only someone (Russia, India or China) would use molten salt instead of sodium as coolant in their fast reactors, the rest of world would be less scared of fast spectrum reactors. NaF-ZrF4 would do quite well as the secondary coolant to begin with. Later they could dissolve the fuel in the same (or improved) salt eutectic and put it in the place of primary coolant inside the reactor vessel.
That could be the ultimate solution to the graphite waste problem. Also to the problem of 99.995% Li-7 availability.


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PostPosted: Jan 05, 2013 3:28 pm 
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jaro wrote:
Lars wrote:
A good hint is France's declaration that graphite was high level waste (and I presume that is just from the 14C levels since they haven't run a molten salt reactor I'd guess the graphite had no contact with fission products or plutonium ).
That declaration was based on chlorine-36, not C-14.
Chlorine is used to purify graphite during manufacture.
In a reactor the residual chlorine-35 yields an activation product, Cl-36, which is a long-lived (301,000 years) energetic (709 keV) beta emitter that is highly soluble in water.
Graphite wastes have an activity of roughly 1 MBq/kg, from Cl-36.
Even though C-14 is 500 times more radioactive, France's CNE said that it is Cl-36 that poses the problem in a repository because of its extreme mobility in the biosphere.


This company developed a novel method to produce solar grade silicon:
http://www.engineeringtv.com/video/RSis ... %20silicon
http://www.rsi-silicon.com/images/RSi_P ... ration.jpg

The method involves using highly purified carbon derived from sugar in an arc furnace. Could this be used to make moderator graphite?


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PostPosted: Sep 13, 2014 9:15 am 
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One option is using boron carbide (B11) as inserts in the graphite core. The inserts can be replaced instead of the whole core and the boron carbide molecular structure seems to whistand more than graphite.


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PostPosted: Sep 14, 2014 11:35 pm 
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I have often thought of Beryllium Carbide as moderator. However the best would be to do without one. It will reduce the core and reactor size and you could invest the savings in the initial fissile fuel.


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