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PostPosted: Jun 20, 2015 1:48 pm 
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Well - natural Zirconium has a rather low cross section for neutron absorption at most energies - which is why it is used in things like fuel cladding and CANDU pressure tubes.
But of the common Zr isotopes 90Zr has an even lower cross section, something on order of 0.08 barns.

Enrichment of Zirconium has been attempted by chromatography but does not appear to be going anywhere fast - however it occurs to me that we have a source of pure isotopic 90Zr.

We have many tonnes of 90Sr, much of it separated and thus uncontaminated by the heavier Zirconium isotopes.
So how about we just leave the separated strontium supply and occasionally extract the decay zirconium from it.

A study on isotopic enrichment with AVLIS put the value of 90% 90Zr for pressure tube manufacture at $1000/kg or more.
Completely pure 90Zr would likely be worth more - and high performance regimes like MSRs (Like salt and Transatomic's ZrH moderator) could probably increase that even further.


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PostPosted: Jun 21, 2015 7:09 am 
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Nice idea. I believe a typical LWR produces about 20kg/GWyr of Sr90.

I suspect most of it is vitrified away somewhere.

Do most MSR designs rely on Zr?


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PostPosted: Jun 21, 2015 8:34 am 
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Not particularly, but there are various designs using things like zirconium fluoride in their fuel salts and Transatomic uses ZrH as a moderator and has serious absorption problems.

One problem is that production will take a long time to approach equilibrium but not much can be done about that.


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PostPosted: Jun 24, 2015 4:12 pm 
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Too small a supply to be worth fussing over.


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PostPosted: Jun 24, 2015 5:13 pm 
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Certainly for fuel use, but if you made use of HEC or similar designs for CANDU pressure tubes, you only need a few thousand kilogrammes to make a substantial difference to the performance of the entire reactor.

If we piled up all the strontium we have we could get that kind of amount of material out every few years.
And once you have it you can recycle it after the plant is dismantled.


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PostPosted: Jun 25, 2015 7:59 am 
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Alex,
20kg/GW? The cumulative yield for 90-Sr in the Lawrence-Livermore data shows 5.78% for 235-U thermal.

http://ie.lbl.gov/fission/235ut.txt

Somewhere on this forum I recall reading a rule of thumb that a GW/year takes a ton of fuel. Is that ton metric?. If so, won't the cumulative yield be 57.8 kg, assuming no decay of 90-Sr in that year?


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PostPosted: Jun 25, 2015 10:26 am 
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I have found this patent from 1986 (filed 1982 by Westinghouse) "Separation of isotopes of zirconium
US 4568436 A"

https://www.google.com/patents/US4568436

It claims a 3 time lower barns value in thermal spectrum...
Quote:
Table I shows that the effective thermal neutron-absorption cross section of the zirconium could be reduced from its natural value of 0.09732 barns to 0.03281 barns, a factor of about three, if only the isotope 90 Zr were present. Correspondingly the fast neutron absorption cross section would be reduced from its natural value of 0.04121 barns to 0.00806 barns if only 90 Zr were present, a reduction by a factor of five.


They claim a 7% fuel cost saving for a pressurized water reactor.
Quote:
Computer calculations based on existing codes for specific thermal reactors have shown that if 90 Zr were separated and used in a pressurized water reactor having 4 loops, fuel cost savings of 7% would result. Similar calculations relative to removing 91 Zr and using the remaining isotopes in the reactor have shown that a 5% fuel cost saving would result. In either case the fuel cost saving is a substantial amount of money per reactor core loading, such that if all of the saving were assigned to the cost of performing the necessary isotopic separations, it would amount to $115 per pound of Zircaloy-4 alloy for the 90 Zr separation case, and $80 per pound of Zircaloy-4 alloy for the 91 Zr removal case. These figures are based on fuel costs of 50ยข/MBTU.


In a MSR like a LFTR the fuel cost is not a real problem.
So where are the advantages to use 90 Zr?
Where can it be used and could this mean smaller design, less salt....?


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PostPosted: Jun 25, 2015 12:17 pm 
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SteveMoniz wrote:
Alex,
20kg/GW? The cumulative yield for 90-Sr in the Lawrence-Livermore data shows 5.78% for 235-U thermal.
http://ie.lbl.gov/fission/235ut.txt
Somewhere on this forum I recall reading a rule of thumb that a GW/year takes a ton of fuel. Is that ton metric?. If so, won't the cumulative yield be 57.8 kg, assuming no decay of 90-Sr in that year?

Fission yields have to be adjusted for molecular weight - and add up to 200%.
Comes out at 22kg 90Sr per tonne of fission products.
Quote:
In a MSR like a LFTR the fuel cost is not a real problem.
So where are the advantages to use 90 Zr?
Where can it be used and could this mean smaller design, less salt....?

If you can reduce absorption by zirconium by a factor of five you could combine it with Depleted Molybdenum to build a CANDU SCWR "High Efficiency Channel" that has a similar total neutron cross section to the CANDU-BLW tube.
A supercritical water reactor with natural uranium or SEU. That would certainly be worth it.


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PostPosted: Jun 26, 2015 10:23 pm 
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I get it. 235-U becomes 90-Sr 5.78% of the time. The result masses 90/235 of the original.

(5.78)* (90/235) = 22.1


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PostPosted: Jul 03, 2015 7:46 pm 
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I found an interesting piece of work about using a plasma centrifuge to enrich Zirconium, in this case to improve neutron economy in CANDUs, probably by replacing the pressure tubes and potentially even the cladding.

Apparently it is not quite considered worth it - but this might prove useful in higher flux reactors, and also for reducing the amount of radioactive waste produced in reactor operations by improving a reactor's breeding ratio. Enriching stable isotopes to reduce waste production seems a good tradeoff. Especially since widespread use of zirconium in chemical catalysts is generating a market for 91Zr rich tails that have no problems with large neutron cross sections.
Also very interesting for the Transatomic people.


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PostPosted: Jul 04, 2015 2:49 am 
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Sr 90 is a beta emitter before it could get converted to zirconium. If you can go to the trouble of separating it, best use it as decay fuel in RTG before you think of using the end product.
It may be best to use existing metals or refractory oxides rather than finding exotic methods for isotope separation for structural materials. Such treatment should be reserved for materials used in neutronic reactions.


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