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PostPosted: Feb 17, 2015 6:58 pm 
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Yesterday I downloaded a paper (thanks Bruce) presented by P.R. Kasten at the 3rd International Conference on the Peaceful Uses of Atomic Energy (1964) which discusses a MSR concept (molten salt epithermal reactor or MOSEL) which effectively addresses most of the “issues, arguments, quibbles, & excuses” raised about them. (http://moltensalt.org.s3-website-us-eas ... pt_OCR.pdf)
(An easier to read pdf is ATTACHED)

Dr. Kasten (a German who had apparently collaborated with ORNL’s researchers) modeled a 1.3 GWt, 2 fluid breeder reactor whose core consists of a metallic heat exchanger containing roughly 42 volume percent of 233U-in-NaF/BeF fuel salt (no Th):16 volume percent metal: & 42 volume percent of a 232Th-NaF/BeF blanket salt. There is no moderator & its neutron spectrum should therefore be quite similar to that of the MSFR. Heat exchange to the secondary salt from which the reactor’s turbines extract power is solely via circulation of the blanket salt through an external HX – the fuel-side salt is circulated only to enhance heat exchange to the blanket salt and facilitate inert gas & noble metal scum removal (it also provides fuel dump capability).

According to Dr. Kasten, MOSEL could achieve CR’s > 1.05 with no reprocessing of its fuel-side salt other than occasional uranium recovery/recycle.

While his paper didn’t specify sizes & volumes, I’ve derived the following characteristics from some of the numbers that he did reveal:
• In-core fuel salt volume (probably over 90% of the total) is about 2.8 m3
• Total core volume (fuel, metal, & blanket) is about 6.7 m3
• Total metal (Hastelloy) in the core HX is about 6.4 tonnes
• Total 233U within the core is about 2 tonnes

He assumed that the core HX would be replaced at 3-5 year intervals (that’s similar to the core wall replacement interval assumed for the MSFR).

This concept has the following advantages relative to anything else I’ve see so far including the MSFR.

1. the fact that power is extracted solely from its blanket salt means that that HX would be subjected to a much lower neutron flux and less apt to be fouled with FP detritus than those of other MSR concepts (last longer)
2. eliminating the fuel salt HX, piping, etc. means that a much higher percentage of delayed neutrons would be kept within the core (e.g., >90% vs 40-50%) which should enhance safety
3. the same feature would reduce its fissile inventory to the same degree
4. Its high “breeding” capability means that it could achieve genuine sustainability (i.e., generate 100% of its own fissile from infinitely “renewable” thorium) with very little, very simple, and very cheap “reprocessing” (far less/cheaper than that required by any other sort of breeder I’m aware of).
5. This also translates to very little, very simple, and very cheap waste management (discard/vitrify everything in the “reprocessed” fuel salt slipstream except uranium)
6. Its core HX is small enough (“modular”) both in size & mass to render it relatively simple and cheap to replace.
7. it can achieve CR>1 with sodium-based solvent salts – concepts such as either of ORNL’s MSBRs (both one & two fluid) or the EU’s MSFR would require several (maybe tens of) million dollars worth of 7Li per reactor..
8. no 7Li requirement also means that its blanket salt would be cheap to produce which, in turn, would render a pool-type configuration reactor practical (immerse the core within a large pool of the blanket salt). This would enhance its breeding capability (lesser parasitic neutron absorption by 233Pa) and provide a large heat sink if the reactor were to be scrammed.

Please look the paper over & get back to me with your comments/analyses.


Attachments:
MOSELConcept_OCR (epithermal 2fluid).pdf [1.38 MiB]
Downloaded 378 times

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Last edited by darryl siemer on Feb 19, 2015 1:42 pm, edited 1 time in total.
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PostPosted: Feb 17, 2015 8:35 pm 
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Its certainly a very interesting concept - especially since the reactor power can be expanded almost without limit by simply extending the 'width' of the core.
Although I would worry about how well the Hastelloy/Inconel/another alloy I have never heard of would stand up to the constant neutron bombardment from virtually every neutron in the core. Building a reactor wall out of it is one thing - exposing it to that is quite another.

Its like PWR fuel cladding, but with enormously greater fluences.


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PostPosted: Feb 18, 2015 1:06 am 
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Looks familiar.....

I think this type of design is suitable for SMRs, but not for high power density.

I discussed this elsewhere in more detail.
viewtopic.php?f=29&t=4524


Attachments:
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PostPosted: Feb 18, 2015 6:18 am 
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Quote:
the fact that power is extracted solely from its blanket salt means that that HX would be subjected to a much lower neutron flux and less apt to be fouled with FP detritus than those of other MSR concepts (last longer)


Darryl,

The power first has to be extracted from the primary HX - which is in the middle of a MSR core!!!

This does not make the HX last longer.

We need a ballpark of 2 m2/MWt or 2600 m2 of HX surface area in the core. This is a heck of a lot more than the misleading couple of pipe snake bends in the diagram! It is more like a PWR, loads of tiny tubes, this is what you need.

As well, 1000C operating temperature is quite mad. ORNL was worried that Hastelloy N would not work well at 700C and were considering a dial down to 650C. MSRE operated at 650C for good reasons.

1000C, HX-in-core means expected lifetime in the order of hours to days. Any longer, if it hasn't failed from other causes, it will be so badly embrittled and radiation voided, it would not be a safe engineering material anymore.

This is a paper reactor.


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PostPosted: Feb 18, 2015 9:40 am 
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Cyril R wrote:
Quote:
the fact that power is extracted solely from its blanket salt means that that HX would be subjected to a much lower neutron flux and less apt to be fouled with FP detritus than those of other MSR concepts (last longer)


Darryl,

The power first has to be extracted from the primary HX - which is in the middle of a MSR core!!!

This does not make the HX last longer.

We need a ballpark of 2 m2/MWt or 2600 m2 of HX surface area in the core. This is a heck of a lot more than the misleading couple of pipe snake bends in the diagram! It is more like a PWR, loads of tiny tubes, this is what you need.

As well, 1000C operating temperature is quite mad. ORNL was worried that Hastelloy N would not work well at 700C and were considering a dial down to 650C. MSRE operated at 650C for good reasons.

1000C, HX-in-core means expected lifetime in the order of hours to days. Any longer, if it hasn't failed from other causes, it will be so badly embrittled and radiation voided, it would not be a safe engineering material anymore.

This is a paper reactor.


You are absolutely right, it is just a paper reactor & I've assumed that Dr. Kasten didn't screw up somewhere. I believe that the 1000 C figure mentioned in his article refers to the temp of the fuel salt at the middle of each fuel tube/channel within the core HX, not that of the entire core or the metal that it's made of. He assumed an approximately 700 C bulk temp & his calculations indicated that the temperature differential across the metal walls separating the fuel and blanket salt streams would only be about 20 degrees which in turn suggests a max metal temp of under 750 C. However, this means that all heck would quickly break loose upon pump failure if thermal expansion doesn't immediately kill the chain reaction. I don't know how to evaluate such things. Do you?

The scariest thing about this thing to me is the extremely high concentration of fissile (233U) in its fuel.- about 7 mole percent or 35 x that in most of the LFTR designs I've heard about and 2.5x that assumed for the MSFR. This opens up a host of inadvertent criticality scenarios in accident scenarios.

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Last edited by darryl siemer on Feb 18, 2015 10:30 pm, edited 1 time in total.

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PostPosted: Feb 18, 2015 6:26 pm 
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One miracle for me is how the DFR got a patent.


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PostPosted: Feb 18, 2015 6:44 pm 
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HolgerNarrog wrote:
One miracle for me is how the DFR got a patent.

Lots of things get patented......


Attachments:
LeBlancs_US_Patent_Application_20120183112.JPG
LeBlancs_US_Patent_Application_20120183112.JPG [ 29.84 KiB | Viewed 6296 times ]
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PostPosted: Feb 21, 2015 2:41 pm 
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jaro wrote:
Looks familiar.....

I think this type of design is suitable for SMRs, but not for high power density.

I discussed this elsewhere in more detail.
http://www.energyfromthorium.com/forum/ ... =29&t=4524


I went to your link- there is no such discussion there.

Do you have a reason for saying that this system would be unsuitable for high power density?

Re its power density: the figure mentioned in Kasten's paper (470 kW/liter) is only about one half of that in EBR II's core (60 MW generated within about 65 liters) which means that the metal comprising its HX core structure would be "seeing" about one half of the fast neutron flux that EBR II's stainless steel fuel cladding survived during burnups up to 170 GWd/t. A MOSEL breeder core's durability is an open question which, I suspect, can only be answered via realistic experimentation.

Again, Dr. Kasten's nominal fuel salt temperature (1000C) is not the temperature that the HX core metal would experience (how hot does the aluminum of a saucepan containing water being boiled over a 1500C air-methane flame get?)

The picture in the patent you ATTACHED isn't very helpful. I suspect that a plate-type, cross flow HX core configuration would be more practical for a blanket salt-cooled breeder (most discussions of MOSEL-type reactors assume a molten metal coolant).

Do you (anybody?) have a way of estimating what what would happen if the blanket salt pump fails? Ditto if a big gas bubble is pumped into the blanket salt side of the HX?

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PostPosted: Feb 21, 2015 4:27 pm 
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Solubility of thorium salts is much less than that of uranium or plutonium in molten salts. I wonder why thorium is still considered for dissolving in fuel or blanket salts.
I feel that high melting thorium metal should be used as a metallic blanket and the fuel should have uranium or plutonium only. The uranium could be U233 if and when available and desired.
The reactor under consideration could have a metallic thorium liner as blanket and the coolant envelope could be a clean salt. Even a boiler jacket full of water could be used. It will provide some thermalised neutron on the outer surface as a steady source of neutrons which could create additional Neutrons moving inwards to be absorbed by any thorium or uranium-238 which may be in the salt. If the water is reduced, the thermal fission will cease as in boiling water reactors.
The irradiated thorium could be electrolytically refined after 270 days cooling to extract the U-233.


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PostPosted: Feb 21, 2015 6:59 pm 
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jagdish wrote:
Solubility of thorium salts is much less than that of uranium or plutonium in molten salts. I wonder why thorium is still considered for dissolving in fuel or blanket salts.
I feel that high melting thorium metal should be used as a metallic blanket and the fuel should have uranium or plutonium only. The uranium could be U233 if and when available and desired.
The reactor under consideration could have a metallic thorium liner as blanket and the coolant envelope could be a clean salt. Even a boiler jacket full of water could be used. It will provide some thermalised neutron on the outer surface as a steady source of neutrons which could create additional Neutrons moving inwards to be absorbed by any thorium or uranium-238 which may be in the salt. If the water is reduced, the thermal fission will cease as in boiling water reactors.
The irradiated thorium could be electrolytically refined after 270 days cooling to extract the U-233.


Metallic thorium is extremely reactive (both air & water rapidly oxidize it) & any sort of solid blanket containing Th would subject the 233Pa to a much higher neutron flux (transmute it before it can decay to 233U) than would a Th-bearing fluid blanket.

ThF4 readily dissolves in LiF - There's a broad eutectic containing 22-29mole % ThF4 which melts at roughly 560C & would be fine blanket salt. According to NIST, the NaF/ThF4 eutectic contains about 24mole% Th & melts at about 619C.

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PostPosted: Feb 27, 2015 7:40 pm 
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I've put together a paper for an ASME journal (draft ATTACHED) which encourages its readers to work out some of the "engineering details" that must be to dealt with before a "real" MOSEL or MSFR could be built. I encourage you to get started on it now.

If anyone can answer (definitively) any of the questions it poses, please post your answer(s) here. Questions are welcome too.


Attachments:
2015 ASME doc.doc [1.15 MiB]
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PostPosted: Mar 04, 2015 5:24 pm 
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darryl siemer wrote:
I've put together a paper for ....


Here's the as-submitted version (today, ATTACHED)

I've added a few more footnotes, changed citation style, & reworded its conclusions. I'm hoping to get some "outsiders" interested in volunteering ideas & labor.


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2015 ASME entire doc.doc [1.16 MiB]
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PostPosted: Mar 04, 2015 7:06 pm 
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darryl siemer wrote:
darryl siemer wrote:
I've put together a paper for ....


Here's the as-submitted version (today, ATTACHED)

I've added a few more footnotes, changed citation style, & reworded its conclusions. I'm hoping to get some "outsiders" interested in volunteering ideas & labor.



Page 3:

Quote:
My next ballpark calculation demonstrates why such a nuclear renaissance could not be implemented with either today’s Gen II or the nuclear industry’s even-safer (and even more expensive) “advanced” Gen III light-water reactors (LWRs), all of which can consume only readily fissionable 235U and 238Pu, not the far more abundant 238U and/or 232Th.


Should read 239Pu.

As far as power conversion, why is supercritical CO2 Brayton necessary? Could an air breathing turbine, like the NACC proposed for the FHR, be used with an MSFR?

For the MSFR, you mention that only U and F would be recycled. 7Li wouldn't be recycled either? That seems like a costly material to throw away.


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PostPosted: Mar 04, 2015 9:53 pm 
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Cthorm wrote:
darryl siemer wrote:
darryl siemer wrote:
I've put together a paper for ....


Should read 239Pu.

As far as power conversion, why is supercritical CO2 Brayton necessary? Could an air breathing turbine, like the NACC proposed for the FHR, be used with an MSFR?

For the MSFR, you mention that only U and F would be recycled. 7Li wouldn't be recycled either? That seems like a costly... away.


Good catch on p 3's 239Pu. thanks

Due to CO2's superior "compressibility", CO2 Brayton is apt to be both more efficient & cheaper to build (smaller) than an open-air system which, in turn, should translate to lower cost electricity. It also provides another barrier to tritium loss. I'm not too excited about generating another customer for "fracked" natural gas.

As far as recycling the 7Li is concerned I just feel that it wouldn't be worth doing as far as the utility's owners are concerned. In any case, recovering it would probably be easier/cheaper after the glass logs containing it along with that scenario's discarded thorium have cooled off in the repository (it ain't gonna go anywhere).

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PostPosted: Mar 05, 2015 1:17 am 
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http://www.scientificamerican.com/artic ... ate-power/
The advantages of super-critical CO2 have been discussed It is claimed to be more compact and efficient as it can work at higher temperatures.
For compactness of the reactor core, it should be free of graphite moderator. Sodium in place of lithium can also cut initial costs. It may be possible at higher energy spectrum.
Corrosion due to salt will still have to be managed. SiC coatings may help.


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