Energy From Thorium Discussion Forum

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PostPosted: Jan 10, 2010 11:47 am 
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In case people haven't noticed, the HW-MSR offers a useful kind of flexibility that is not readily matched by other concepts:

In the current highly restrictive political climate, it can be run on NU or SEU as a near breeder, while we await a more permissive future.

Towards the middle or end of the century, when (we hope) the political climate changes, the same reactor can be fueled with gradually increasing grades of fuel -- from LEU to HEU -- with increasing loads of thorium, making it into an excellent breeder.....

Other systems typically start at the other (wrong) end of the spectrum -- like jumping into the deep end of the pool, when the lifeguard on duty has put it off limits.....


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PostPosted: Jan 11, 2010 12:00 pm 
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It does seem reasonable that the economics of reprocessing would turn out to have a big scaling factor, but modularity (full factory production, high learning rate) is of course a major advantage as well.

Perhaps a large number of smaller modular reactors will use one or two shared on-site reprocessing facilities. The economics of reprocessing as well as modular reactors, plus there is little transport of radwaste, and a high degree of reliability can be achieved (spread out the maintenance and downtime).

Imagine a park of 10 x 300 MWe units. Or 30 x 100 MWe. Any number of individual reactors can be put together in power parks.


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PostPosted: Jan 12, 2010 7:40 am 
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Th-LEU MOX has already been designed for AHWR.
http://www.dae.gov.in/gc/ahwr-leu-broc.pdf
PHWR shall take a smaller proportion of 19.75%LEU. With approx. 90% thorium, it shall have a higher conversion ratio than the AHWR. It could be denatured as in the AHWR design or just have 4 of 37 pins of LEU. Reactor grade plutonium could also be used in place of LEU.
Others are in the process of designing thorium fuel for LWRs.


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PostPosted: Feb 07, 2010 6:03 am 
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AHWR300-leU MOX fuel is a denatured thorium fuel for a solid fueled reactor which itself has been designed but not built.
http://www.dae.gov.in/gc/ahwr-leu-broc.pdf
The basic concept of Th-20%leU MOX MOX can be extended to LWR with qualitatively similar but quantitatively slightly different characteristics. The proportion of leU will be greater than 22% of AHWR to 30-35%. A part of the power will be produced by fission of thorium. (39% in AHWR). The burn up of fuel will be comparable with 64,000(MWd/te) of the AHWR fuel.The denatured Uranium recovered by reprocessing is likely to be fit for LWR fuel. Due to a high burn up, the Pu shall be unfit for weapon use and may encourage the US to resume reprocessing of used fuel. It could, however, be fit for fast reactors or thorium fueled reactors.
Th-leU MOX can be a better way to introduce thorium in existing LWRs to reduce consumption of uranium as the supply position of uranium fuel is going to be tight after exhaustion of surplus weapons fuel. To make the LWR used fuel less proliferation attractive, it could be put through one DUPIC cycle. Reprocessing-shy countries like the US must introduce cheaper versions of CANDU for this purpose. Construction can be 'outsourced' to Indian firms for reduced costs.


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PostPosted: Feb 10, 2010 8:57 am 
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I'm thinking of a few changes that will be very beneficial to a solid fueled CANDU thermal breeder, perhaps someone here can comment.

The first thing to do is to increase the power density of the reactor to improve neutron economy. This will require a higher thermal conductivity fuel, such as uranium diboride (UB2), with a zirconium or metal boride clad. The boron will of course have to be nearly pure depleted boron (B-11).

Also, the high pressure has to go. Some kind of stable non-corrosive non-toxic, non-flammable coolant has to be used that doesn't absorb too many neutrons. Fluorocarbons have been suggested, like Krytox (Jagdish?). This allows safer operation at even a bit higher temperatures (400C?), easier refuelling, and fewer neutron absorbing structural materials.

Finally, the most practical approach is to use the calandria as breeder blanket. For example, a thorium boride suspension. This can then easily be filtered out and sintered into new fuel elements.

This kind of arrangement could be useful when reprocessing is a sensitive issue or the technology is unavailable. Uranium boride fuel can be stored even safer than uranium oxide, in dry storage casks. It fact it is being investigated for this very purpose.


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PostPosted: Aug 20, 2010 7:15 pm 
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Cyril R wrote:
... a higher thermal conductivity fuel, such as uranium diboride (UB2) ...


How much heat does it conduct?

(How fire can be domesticated)


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PostPosted: Aug 21, 2010 2:57 am 
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GRLCowan wrote:
Cyril R wrote:
... a higher thermal conductivity fuel, such as uranium diboride (UB2) ...


How much heat does it conduct?

(How fire can be domesticated)


I looked it up in Google books when I did a check for ceramic fuel properties but UB2 was really hard to find back then. Its in the Handbook of Condensed Matter and Materials Data, Volume 1:

http://books.google.nl/books?id=TnHJX79 ... de&f=false

51.9 W/m/K. It does not give temperature dependant thermal conductivities, though borides tend to hold their ground better than most ceramics in this respect...


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PostPosted: Aug 21, 2010 5:02 am 
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Al-Mg is a high thermal conductivity, low vapor pressure low fire hazard coolant melting at 449C. I do not yet know if it is as benign to container/cladding materials as sodium. It should be as easy to pump by electro magnetic pumping as sodium. It could be a candidate for heat transfer from a low pressure high temperature reactor. Heavy water is the best moderator known but has the limitation of volatility comparable to water. Perhaps it should be circulated through a cooler located outside the reactor core.
That said, AHWR design is not a breeder or even claimed as a near breeder. For fissile isotope self sufficiency, you have to have fast spectrum reactors.


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PostPosted: Mar 20, 2013 3:26 am 
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If you are running in a bi-modal fast/thermal modes, you may as well go the whole hog and have uranium in the fast zone and thorium as a thermal blanket.
You could have thorium as the first and last bundle in a tube as an axial blanket. You could also have the outermost two layers of tubes as a radial reflector/blanket.
In the individual core bundles, nearly 60% of total bundles, have Th-20%LEU pins like the AHWR fuel as drivers and some thorium pins as blanket. The radial blanket bundles could be changed frequently for maximum conversion. The core and axial blanket could be burnt to the maximum taking advantage of fast fission of U238 as well as U-235 initially and U-233 later. The blanket bundles could be metallic thorium which can be reprocessed in a single stage electro-refining to recover Thorium and U-233. It may be a breeder arrangement or utilised anyway for maximum conversion to U-233, for use in the LFTR when ready.


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PostPosted: Nov 14, 2016 9:26 pm 
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Sorry to dig up a very old thread but I cam across interesting information about this.

It appears that if you were to obtain 99% 90Zr and use it in the calandria and pressure tubes of a CANDU you could obtain ~10-11GWd/t with natural uranium fuel.
And more than that, apparently the practical burnup for a self sustaining thorium cycle jumps from ~5GWd/t to ~15GWd/t.
Which is high enough that it might actually be an economic cycle in a high uranium price environment.

I am also investigating if it can be used to further improve the neutron economy of an SEU fuel one and results are promising, with a ~15% improvement in burnup as a first estimate. It reduces non-fuel parasitic absorbances by more than a third.

99% 90Zr could be obtained from a plasma centrifuge at something resembling a reasonable cost, although it was marginally uneconomic at the time it was last researched in a serious way in the mid 80s.
But still, very interesting.


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PostPosted: Nov 16, 2016 8:10 am 
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I feel that a breeder or near breeder can be achieved by using a blanket as was the case with the shippingport experiment, even with light water. Two bundles at the ends of a tube could be a thorium axial blanket, amounting to nearly 17% of total fuel. Two outermost rings could be put in as a radial blanket.
While the axial blanket could be burnt in situ till the tube needs refuelling, the radial blanket could be irradiated for maximum conversion, 400 to 500 days. With a 19.8% LEU-thorium core fuel, there could be five blanket changes to one complete fuel change, resulting in a breeder. The fuel in radial blanket tubes could be metallic enabling a single step electro refining for separation of fissile U233.
The fissile U233 could be used in the same reactor or a fast reactor or even an LFTR.


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