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PostPosted: Sep 11, 2008 4:19 pm 
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Hi David,

I was thinking about the coefficient of reactivity with respect to changes in the flow rate. Given the geometry you have it seems that there might be a negative coefficient of reactivity w.r.t the flow rate. That means, the fuel flow rate goes higher, the reaction rate will drop. I'm hypothesizing because for a higher flow rate, more and more of the delayed neutrons will end up in the part of the reactor whose geometry does not support criticality. That means that relatively more of those delayed neutrons will end up being absorbed by the blanket.

This may or may not be a big deal, but depending on the control laws used to ramp up and down the reaction rate, this could cause a dynamic instability. Here's the idea. Suppose the control laws are written to increase the fuel salt flow rate in response to a higher load in order to transfer more heat to the heat exchanger. Changing that rate will lower the equilibrium temperature of the core (all things being equal), which might cause the feedback loop to increase the rate even more (i.e., a positive feedback loop). I suppose a bigger danger might result from shutting off the flow altogether, raising the temperature of the core too high.

Of course, all of this depends on the values of the various reactivity coefficients (thermal, flow-rate, etc) and the control mechanisms that are being suggested. Until you've the dynamic models in place, and the control laws written, it's hard to know exactly what might happen. I was just thinking about the problem though and thought I might pass along these concerns.


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PostPosted: Sep 11, 2008 10:02 pm 
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David [or anyone],

I'm newly exposed to this technology, so please bear with my naivete. I'm trying to develop a basic MSR presentation for the educated general public, but I need to understand a bit more, myself. Can you or anyone else answer some questions I have about your presentation.

General Attributes

Why is the vehicle salt BeF2 and LiF? Light to be moderator?

Does the 20 day cycle for fission products mean it takes 20 days to circulate the entire fuel load through the chemical separation system? And 10 days for the Pa233 means the thorium circuit is separate?

Does "specific inventory" of 1500 kg mean 1500 kg of U233 in the active core? Or the whole system?

Does breeding ratio of 1.06 and 20 year doubling time work out to about 20 months to convert 1 kg of U233, via thorium, to 1.06 kg of U233?

General Benefits
Transuranic waste production extremely low -- how low? How much plutonium? How does is depend on whether the MSR is started-up with U233, vs HEU, vs Pu? This can be a critical selling point to the general public.

Radiotoxicity graph
The "dose" is for a gigawatt-year produced by a thorium MSR?
I can't relate the graph line labels to "FBR" and "MSR".
Any web URLs I can quote?
Is this the proof that the waste need only be stored for 1000 years?

Problems
Is the Pa233 separation a proliferation concern because you want to leave it mixed with the U233 to make it difficult to work with because of x-rays? Don't you have to separate the Pa233 to allow it to decay to U233 without absorbing a neutron?

Does the U232 come from Pa233 + neutron - alphaparticle?

I thought U233 was a poor material for making weapons in any event. Weapons grade plutonium is plentiful and works well, no?

Liquid Bismuth Reduction
I don't understand the point, at all; sorry.

Aircraft Reactor Experiment
The temperature was 860C. Can an MSR reach 900-950C for efficient thermo-electrolysis production of hydrogen from water?

Graphite Free Version
No moderation? Fast spectrum neutrons?

System Advantages
For MSR moderated by graphite pebbles or tubes with channels?
Fissile inventory of 400 kg/GW is dramatically less than 1500kg/GW. Does this mean we would burn up 400 kg of U233 in about 5 months?

..................

If you can answer these, maybe I'll be smart enough to understand the rest...

Bob Hargraves


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PostPosted: Sep 12, 2008 12:07 am 
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robert.hargraves wrote:
David [or anyone],

I'm newly exposed to this technology, so please bear with my naivete. I'm trying to develop a basic MSR presentation for the educated general public, but I need to understand a bit more, myself. Can you or anyone else answer some questions I have about your presentation.

General Attributes

Why is the vehicle salt BeF2 and LiF? Light to be moderator?

Does the 20 day cycle for fission products mean it takes 20 days to circulate the entire fuel load through the chemical separation system? And 10 days for the Pa233 means the thorium circuit is separate?

Does "specific inventory" of 1500 kg mean 1500 kg of U233 in the active core? Or the whole system?

Does breeding ratio of 1.06 and 20 year doubling time work out to about 20 months to convert 1 kg of U233, via thorium, to 1.06 kg of U233?

General Benefits
Transuranic waste production extremely low -- how low? How much plutonium? How does is depend on whether the MSR is started-up with U233, vs HEU, vs Pu? This can be a critical selling point to the general public.

Radiotoxicity graph
The "dose" is for a gigawatt-year produced by a thorium MSR?
I can't relate the graph line labels to "FBR" and "MSR".
Any web URLs I can quote?
Is this the proof that the waste need only be stored for 1000 years?

Problems
Is the Pa233 separation a proliferation concern because you want to leave it mixed with the U233 to make it difficult to work with because of x-rays? Don't you have to separate the Pa233 to allow it to decay to U233 without absorbing a neutron?

Does the U232 come from Pa233 + neutron - alphaparticle?

I thought U233 was a poor material for making weapons in any event. Weapons grade plutonium is plentiful and works well, no?

Liquid Bismuth Reduction
I don't understand the point, at all; sorry.

Aircraft Reactor Experiment
The temperature was 860C. Can an MSR reach 900-950C for efficient thermo-electrolysis production of hydrogen from water?

Graphite Free Version
No moderation? Fast spectrum neutrons?

System Advantages
For MSR moderated by graphite pebbles or tubes with channels?
Fissile inventory of 400 kg/GW is dramatically less than 1500kg/GW. Does this mean we would burn up 400 kg of U233 in about 5 months?

..................

If you can answer these, maybe I'll be smart enough to understand the rest...

Bob Hargraves


I'll give it a whirl, to the best of my meager ability.

IIRC, FliBe is proposed because it has already been proven out, in the MSRE.
Specific inventory is, I think, the fissile inventory per unit something, in this case per (net?) GWe output.
I think you nailed breeding ratio, and a 20 year doubling time would mean, according to the rule of 72, approx 3.6% surplus fissile production pa, compounded (3.52% pa, to be precise). As to time to convert specific block X of Th to U233 or vice versa, I'm not too sure.

Fission products have to be removed, iirc, to allow breeding - the shorter cycle time for their removal, the lower average burden they impose on reactivity. A 20 day cycle would mean a mean cycle time of 20 days to scrub fission products, yes. Pa removal is, as you have seen, somewhat contentious - it may be possible to scrub Pa through a similar system, as it acts chemically in its own damn way, but instead of removing 5% of the fission product load over a day, scrubbing 10%.

TRU production - per unit energy, you have ~ 100x less heavy metal being irradiated, so since conventional fuel cycles convert >1% of the heavy metal loading to TRU (IIRC), you simply cannot produce as much TRU as a conventional fuel cycle. Further reducing TRU production is the lower mass numbers, and U233's greater propensity to fission upon neutron capture (compared to U235) and the continuous recycling of the fuel salt, burning up some of the little TRU that is produced. So conservatively, at least 100x less TRU production. Pretty much sod-all Pu239 is produced, and what does get produced gets burned up in situ or converted into higher isotopes of Pu.

Since Th is element 90, Pa is element 91 and U is element 92, I think beta decay is involved to vault the atom in question up the atomic scale. Complete Pa seperation is the ideal so none of it becomes U234, but with a big enough blanket, that loss can be made as small as desired. I believe U232 is ultimately produced from Th230 via Pa231.

U233 as a device core? I'm not sure if any such devices have been tested, but yes, weapons grade Pu (and weapons grade stupidity) is quite easy to get, eg via Hanford-style graphite piles and lightly toasting U235/238.

I think using liquid bismuth reduction came about due to ORNL's choice of a single fluid design, allowing Th to be extracted in presence of fission products. I think it's major benefit for 2-fluid operation is an easy way to swap bred U233 for fresh Th232.

Temp limits for an LFTR are primarily dependent on the heat exchangers, so if you used (frinstance) tungsten, then four digit temperatures should be no problem.

Not sure about the merits of having/not having graphite, so I'll leave it to those more knowledgeable than me.

I think those fissile inventories are quoted per GWe-yr - a design burning (say) 700 kg/GWe-yr would consume 700 kg fissile over the course of a year in a 1GW plant, over 6 months in a 2 GW plant, over 4 years in a 250 MW plant, etc.

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PostPosted: Sep 12, 2008 10:53 am 
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honzik wrote:
Hi David,

I was thinking about the coefficient of reactivity with respect to changes in the flow rate. Given the geometry you have it seems that there might be a negative coefficient of reactivity w.r.t the flow rate. That means, the fuel flow rate goes higher, the reaction rate will drop. I'm hypothesizing because for a higher flow rate, more and more of the delayed neutrons will end up in the part of the reactor whose geometry does not support criticality. That means that relatively more of those delayed neutrons will end up being absorbed by the blanket.

This may or may not be a big deal, but depending on the control laws used to ramp up and down the reaction rate, this could cause a dynamic instability. Here's the idea. Suppose the control laws are written to increase the fuel salt flow rate in response to a higher load in order to transfer more heat to the heat exchanger. Changing that rate will lower the equilibrium temperature of the core (all things being equal), which might cause the feedback loop to increase the rate even more (i.e., a positive feedback loop). I suppose a bigger danger might result from shutting off the flow altogether, raising the temperature of the core too high.

Of course, all of this depends on the values of the various reactivity coefficients (thermal, flow-rate, etc) and the control mechanisms that are being suggested. Until you've the dynamic models in place, and the control laws written, it's hard to know exactly what might happen. I was just thinking about the problem though and thought I might pass along these concerns.


Your completely right that stopping flow does lead to a small reactivity increase because some of the delayed neutrons that would have been emitted outside the core are now inside. It is always part of the planning and not too big of a concern. If flow suddenly stops, the reactivity momentarily goes up but since this raises the salt temp the reactivity then starts to drop. If flow is stopped it also means the salt isn't being cooled so quite quickly the temp rise of the salt shuts down the nuclear chain reaction. Decay heat will continue to heat the salt though so if you don't restart the flow soon the salt would need to be dumped to storage tanks (done automatically by the melting of freeze plugs).

I guess your main point though was about instabilities caused by flow variations and if some feedback problems could arise. Thats a question I don't have an answer for but something to keep in mind. One thing I should mention is that we typically want to size the power of our pumps such that you can't really have any sort of huge surge of flow. A problem that Oak Ridge always wanted to be careful about was that if you stopped flow for a few seconds, then suddenly surged overcooled salt from the heat exchangers back to the core this could really push the reactivity up. That is the main reason to keep the pumps at a certain minimum power. Since we can't really overcool the salt much without freezing it, even this scenario isn't really a valid concern.


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PostPosted: Sep 12, 2008 11:35 am 
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Robert,

I think fnord has answered many of your questions but I might as well start from scratch.

-LiF and BeF2 just happen to have the best combination of many properties, the most important being they have very low cross sections for the absorption of neutrons. There are other carrier salts proposed but LiF-BeF2 is the best (and most expensive unfortunately).

-20 day cycle means on average a soluble fission product will be removed within 20 days. The Oak Ridge work actually passed all the salt through a system every 10 days so the average efficiency of removal was about 50%. We'd go with much longer removal times these days as we no longer want to produce fissile material, just break even on breeding. In some designs this time could be pushing 10 or 20 years and still break even. 6 months is now a more typical value proposed.

-1500 kg is in the whole system (core, heat exchangers, piping etc)

-I am not sure on your math there, but the way to think about a Breeding Ratio of 1.06 is for every kg we burn, we'd then have an extra 0.06 kg. We burn about 800 kg per year so that means we produce about 50 kg extra per year. If you need 1500 kg to start the next reactor, that means about 30 years if you just use one reactor. Again we probably only want a breeding ratio of 1.00 so we don't ship out or ship in fissile material. Just to start the reactor and thats it. Startup can be with U233 produced elsewhere, waste plutonium or low enriched uranium (using some special tricks for LEU startup)

-The pure Th-U233 cycle will only have in the neighborhood of a few kilograms of plutonium or other transuranics in the salt. When we'd process for fission products we'd want to just put it back in the core to keep burning. We have to assume though that we might lose about 0.1% of this to waste each time. Thus we might be talking about a waste stream of 10s of grams of transuranics per year (versus 250 kg per GWe for Once Through PWR and maybe a few kg per year for a Sodium Cooled Fast Breeder since it needs to process tonnes per year).
For a MSR if we start it on Plutonium we'd be losing a little more to wastes but just for the first couple years (maybe a few hundred grams per year). Exact answers are not possible because we have so many different options on how we'll run things.

-In the graph on my slides, the red line labeled U/Pu is for a FBR (sodium cooled Fast Breeder Reactor). A FBR sends a significant amount to waste even though they try to recycle everything back into the core (when you process tonnes of Pu every year, a significant amount goes to waste). The results for a MSR is the bottom green line Th/U.

-Pa separation is an added proliferation concern because it decays into pure U233 that would be free of any U232 (which leads to lots of very hard gamma rays). If we don't isolate the Pa we do lose a few more neutrons to Pa absorption but in a 2 Fluid system you can minimize this by just having a bigger volume of thorium blanket salt.

-U232 comes from a couple different reaction, one from U233 but more from Th232. The reactions are rare but we only need a little U232 to really contaminate the uranium.

-U233 can be used to make a bomb so we still need to be mindful of this but yes, U to Pu239 is a much easier route.

-Liquid Bismuth Reduction is a method of removing fission products that involves contact the salt with liquid bismuth and a reducing agent (lithium). However, it is a quite complex method and for 2 Fluid designs that have no thorium in the fuel salt, we can use a simpler method called vacuum distillation. That is one of the big advantages of 2 Fluid designs over Single fluid (that mix thorium and U233 in a single salt)

- The proven material for use with MSRs is a nickel alloy called Hastelloy N. It is only rated up to 725 C. It is not a big stretch of the imagination to assume we could use higher temp metals like molybdenum or carbon based materials and get up to the 1000 C level.

-Graphite Free still has some moderation from the light salt atoms themselves. Thus we can still chose to have a fairly soft spectrum by just having a low concentration of fissile material. Often though when we look to remove graphite we also try to get a harder (fast) neutron spectrum to get the advantages of fast spectrums (and some disadvantages as well).

- A fissile inventory of 400 kg per GWe means how much is needed to start the reactor. Any reactor will burn about 800 kg per year per GWe (a bit more for PWRs since they have lower thermal efficiencies).


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PostPosted: Sep 12, 2008 12:13 pm 
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David wrote:
-LiF and BeF2 just happen to have the best combination of many properties, the most important being they have very low cross sections for the absorption of neutrons. There are other carrier salts proposed but LiF-BeF2 is the best (and most expensive unfortunately).

.............

- The proven material for use with MSRs is a nickel alloy called Hastelloy N. It is only rated up to 725 C. It is not a big stretch of the imagination to assume we could use higher temp metals like molybdenum or carbon based materials and get up to the 1000 C level.

I would just like to make a point here that the question of which salt is "best," addressed in the first part of the above quote, depends also on the materials question addressed in the second part of the quote.

And its not just a question of the highest possible operating temperature.

At least in the early part of MSR development, the LOWEST possible operating temperature (i.e. melting point) may also be an important consideration (in fact it always is).

If so, then the use of "carbon based materials" allows a much wider choice of salt compositions, because metallic vessel wall corrosion by fluorides is avoided.

As an example, mixtures of SnF2 and/or BiF3 with UF4 may be considered, potentially giving a melting point of about 200°C (depending on UF4 concentration, and with limited knowledge of phase diagrams in the SnF2-BiF3-UF4 system....)

This is far lower than LiF-BeF2, so I would hesitate to blithely claim that its the "best."

At the other extreme, non-metallic materials do indeed allow much higher operating temperatures -- likely allowing the use of straight UF4, without any carrier salt.

So I guess that "best" is really in the eye of the beholder, as they say....


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PostPosted: Sep 12, 2008 2:29 pm 
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Sure, best is in eye of the beholder and just as there are numerous possible designs we should be investigating there can be more than one best choice. Be careful though about wishing for very low melting point salts. A combination of good negative temperature reactivity and a melting point much lower than operation can be a big problem from the "cold slug" phenomenon if you pump in much colder and denser salts into the core (just like a PWR has to worry about an influx of cold water into the core). If you only operate 50 to 100 degrees above the melting point you don't really have to be too concerned about this.


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PostPosted: Sep 12, 2008 2:30 pm 
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David and/or Yaro

There are a number of emergency conditions that molten salt viscosity will bear upon as follows:

One, if the molten salt circulation stops, the molten salt temperature gradient will stratify and the hot salt will rise to the top of the reactor core. The melt plug at the bottom of the core might not melt since the coolest salt will have settled there.

There seems to be a relationship between the melting temperature of the melt plug, the viscosity of the molten salt, the height and width of the core, and the operating temperature of the core material at the top of the core (Hastelloy N.).

Another complicating factor is the use or non use of graphite pebbles in the design since there presents will reduce convective molten salt flow to some degree which will reduce but not eliminate molten salt temperature stratification.

Does the use of light water reactor fuel at startup introduce and complications here?


Has a model been developed to validate a 100% reliability solution to this condition? For me, this situation is too complicated to be solely resoluble through common sense.


Two, when the turboelectric generator drops load, the circulator will continue to run but no heat will be removed by the cooling loop. How long will heat build up in this situation, and will the melt plug melt in this situation? Does the amount of molten salt or the power density of the core affect this condition? Can this condition be modeled to validate a 100% reliability solution to this condition?


Three, can the design of the MSR afford to exclude passive emergency cooling?

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PostPosted: Oct 08, 2008 3:42 pm 
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Greetings,

I am currently on a brief speaking tour around the Toronto region. I gave talks at U of Ontario Institute of Technology (a new be rapidly growing university) and today up at Bruce Power (they own and run 8 CANDUs in ONtario). Things were very well recieved and I have some great new leads on help in modelling these systems. I speak tomorrow at McMaster University in Hamilton, hopefully I can get more converts there too.
Here is my latest talk version, I've tried to include a bit more explanation in the form of a few footnotes for those only seeing the slides. I'll give more details of my trip when I get back to Ottawa.

David L.


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PostPosted: Oct 08, 2008 4:53 pm 
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David,

You state on slide 16
"Only small amounts Plutonium are present, it is of poor quality (mostly 238Pu) and very hard to extract"

I have no concern at all that the Pu coming from a Th/U cycle is a proliferation concern since it would be so much harder to separate 239Pu from 238Pu than 235U from 238U. (I think this is a much tougher problem than chemically separating the Pu and so far as I understand 238Pu isn't a proliferation concern).

However, I am curious if you consider the falling drops of salt in an upflowing stream of F used by ORNL to be very hard?
If so, what are the concerns?


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PostPosted: Oct 10, 2008 9:45 am 
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Lars wrote:
David,

You state on slide 16
"Only small amounts Plutonium are present, it is of poor quality (mostly 238Pu) and very hard to extract"

I have no concern at all that the Pu coming from a Th/U cycle is a proliferation concern since it would be so much harder to separate 239Pu from 238Pu than 235U from 238U. (I think this is a much tougher problem than chemically separating the Pu and so far as I understand 238Pu isn't a proliferation concern).

However, I am curious if you consider the falling drops of salt in an upflowing stream of F used by ORNL to be very hard?
If so, what are the concerns?



I do have hope for the "falling drop" method of Pu fluorination to be a way we can take out any Pu to put back to burn off (instead of going to waste). The original ORNL report on it makes it sound like a good idea but I would still stick to the label of ""very hard" though based on the fact that in the following decade at ORNL they consistently termed Pu removal to be "uneconomical". Sure, they were not as worried about it going to waste as we now are but it certainly did seem like they that removing Pu to be hard. It is certainly one of the high priorities to examine Pu removal. We can certainly do it by Liquid Bismuth Reductive Extraction as well. So being "hard" to do is good from a proliferation sense but we don't want it to be impossible from a waste reduction point of view.


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PostPosted: Oct 10, 2008 11:13 am 
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David wrote:
Here is my latest talk version, I've tried to include a bit more explanation in the form of a few footnotes for those only seeing the slides.

From slide #60, "Possible AECL Interests"
Quote:
At a minimum, the use of clean molten salt as a low pressure coolant in CANDU design would appear worth investigating

David,

Wonderful though it would be, in theory, to have a low pressure coolant salt in CANDU, I fear that this is physically very challenging, from a fluid dynamics point of view.

Even with 300°C water (at 10.5 MPa - 1500 psi), it takes a lot of pumping power to circulate it through 380 fuel channels, associated feeder tubes, and the four steam generators -- 27 MWe to be exact.

Depending on the salt used, the required pumping power could be several times that (what salt mixture do you propose, when people in the audience ask ?).

And that doesn't even begin to address all the problems of salt compatability with the Zr-clad fuel bundles and channels (i.e. none).

Presumably completely different fuel bundle design would be required ? ....some sort of TRISO contraption ?

And you would still be stuck with the two complex fueling machines -- only now they would be full of salt, instead of high-pressure water.

Seems to me that fluid fuel is so much better -- especially because the far better heat transfer would let you cut the number of fuel channels to a fraction of 380, making the whole reactor much smaller at the same time (less D2O moderator).


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PostPosted: Oct 10, 2008 9:55 pm 
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Jaro,

I was just throwing the idea out there as a baby step for ingrained CANDU people. Yes, I did mean to imply TRISO type fuels to be cooled by a clean molten salts. The salts are very impressive in general at heat removal and basic 2LiF-BeF2 would probably be the salt of choice. I am just going by the work of Forsberg here, as they make a pretty strong case for these salts. I don`t really think the concept has all that much merit, but as I said, just throwing it out there. Yes, your right that any sort of online refueling would be very difficult with hot salt compared to hot heavy water.

David L.


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PostPosted: Oct 12, 2008 9:59 am 
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David wrote:
Yes, I did mean to imply TRISO type fuels to be cooled by a clean molten salts. The salts are very impressive in general at heat removal and basic 2LiF-BeF2 would probably be the salt of choice. I am just going by the work of Forsberg here,

With the high cost of fabrication of TRISO type fuels, it would make no sense whatsoever to use them with natural uranium in CANDUs, to get just ~7.5 MW*d/kg of burnup.
All the HTR designs using TRISO fuel go for burnups at least twenty times higher -- which of course requires fuel with U enrichment on the order of 8% to 10%.
With that kind of fuel, it probably makes little sence to use D2O moderator -- the salt-cooled, graphite moderated Gen-IV VHTR is the way to go, no on-line fueling machines required.

But, getting back to liquid-fueled CANDUs, the D2O would certainly allow you to operate with NU -- and to a much higher burnup than 7.5 MW*d/kg, if you keep removing the neutron poisons....

And if you think VHTR TRISO fuel is a pain to reprocess, then why not up the enrichment level *slightly* in the CANDU-MSR, to get near-break-even performance -- way more burnup than the VHTR, at much lower enrichment level....


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PostPosted: Oct 13, 2008 11:55 pm 
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David, I noticed that in your recent presentations, you have snuck in a second potential solution to "plumbing problem". At least as I interpret the nature of the problem, the problem is not so much that the two-fluid reactor plumbing is un-buildable, but rather that it has poor service life (particularly designs which use graphite). It's basically a "first wall" problem of neutron damage (hence the clever single fluid design with no wall between core and blanket).

That suggests that graphite moderation (and/or complicated plumbing) is fine as long as we provide a solution that makes replacement fast and affordable. Your recent slides state that lowering the temperature (to 600C or so) would extend the life of graphite and allow use of stainless steel, which to me sounds a lot cheaper to replace than carbon-carbon (the most expensive part of the Space Shuttle's skin) or nickel-based hastelloy (asssuming the whole reactor is discarded with the graphite moderator). Sounds like a winner to me! I would think graphite blocks in a steel tube would last longer (because it could be allowed to degrade more) than a graphite tube that had a structural function as well as moderating.

Can you talk a little more about why you don't like graphite anymore? Isn't it the case that the graphite's lifetime is only somewhat shorter than other barrier materials (like hastelloy)? Isn't it more dangerous (for spills etc) to have the fissile load so high as to not require moderation, particularly in a long-skinny core?

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