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PostPosted: Mar 10, 2011 6:35 am 
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Lars wrote:
Just finished reading the report and learned a lot. There are several ideas that would be common with a LFTR. Question 2 is applicable in the context of a LFTR and would love to discover that it isn't a problem at all.

I have a few questions.
1) Was thorium considered as a burnable poison - would it work?
2) It seems like Smahtr will have virtually the same tritium production as a LFTR. It seems to me that the main tritium transfer path will be through the heat exchangers especially if a steam turbine is used. What is the forecast tritium release rate?
3) Your fissile consumption of 1600 kg LEU20 every 3 years for 125 MWth => 2.1 tonnes fissile / GWe-yr so at the end of life roughly half the fissile must still be in the fuel. Is it correct that the spent fuel contains roughly 10% enriched uranium? Is the plan that this is treated as once-through or is the plan to recycle the fuel in some form?
4) From the drawings it looks like an aircraft could take out all three passive cooling towers. Is this a plausible scenario? If so, how does the system handle it?



Thanks for your feedback Lars. It's great to get these comments and questions as they will help us in the next iteration of the design. Let me try to provide some answers to your questions and Sherrell can also chime in:

1) We discussed thorium as an option for burnable absorbers, but frankly did not have time to look at it and for the report section on reactivity control had just considered more traditional poisons. All of the analysis for cycle length was performed without reactivity control. One of our next steps is to design the control system including the burnable absorber and control rods.

2) On tritium, you are correct. As you can read in the the report we plan to have a tritium clean up system to control the tritium. We are not planning on using a steam power conversions system, our preferred approach is SCO2 and in addition to the intermediate loop, one option is the use a "salt vault" head storage system which will represent another barrier to release. We have not gone far enough with this to calculate a tritium release rate.

3) This concept is purposefully focused on just energy production, much as the NGNP, and therefore at this conceptual stage it is best to keep the fuel cycle as simple as possible, which means a one-batch, once-through core design. In order to maximize the cycle length we are using 19.75% enriched uranium. I don't have the number handy regarding the end of life enrichment, but it may very well be in the 10% range. The one-batch core certainly hurts our fuel utilization, but we selected it for simplicity to allow a cartridge core design that is removed in once piece. The fuel could certainly be reprocessed and re-enriched or used in other reactors, and there is some R&D going on in this area. We originally had a section in the report on fuel cycle, but decided to leave it out because we had not performed sufficient analysis. Certainly thorium is an option for these systems as is the deep-burn approach that is being proposed for HTGRs. For now using enriched uranium is a simpler starting point for the concept.

4) There are a couple of different drawings showing the DRACS towers. They would be positioned such that they could not be taken out by one aircraft. Note also that even if they are taken out by an aircraft that does not mean the primary system is not still functioning. This system has a six heat exchangers, 3 primary and 3 DRACS, only 2 (either 2 primary or 2 DRACS) are needed In fact, we didn't analyze it, but I suspect that for decay heat removal, only 1 primary heat exchanger would be sufficient.

Good questions/comments. Keep them coming!


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PostPosted: Mar 10, 2011 11:59 am 
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A relatively simple recycle option would be to make the fuel two cartridges each 2m high rather than a single 4m cartridge. In remote single reactor sites it would operate as you currently foresee but in quad sites you could have a machine that could remove a cartridge, turn it upside down, then reinsert it. This should give you a much more even burn vertically and increase your fuel utilization. Since it also means longer fuel lifetime it means a better chance that thorium in the fuel will be a win.

Seems like one could extend the idea more. Imagine a large AHTR built using the same modules. Seems like the "spent" fuel modules from the small units could serve as fuel modules for the large AHTR to achieve better fuel utilization without creating more waste or any fancy reprocessing.

For tritium, if I understood the old ORNL reports correctly, the vast majority of the tritium is extracted by the He sparge and only residuals left for trapping and removing in subsequent salt loops. You may need to put the He sparge back in. It is simpler in this case since there are no fission products involved so the extracted gas goes through the tritium trap (the old reports used hot CuO tubes but I think people also consider a metal like paladium?).


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PostPosted: Mar 10, 2011 12:38 pm 
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I also just finished reading the report and found it very interesting, an impressive piece of work. I also have a few questions. My background is less of a design and more of a operational viewpoint. But I do still have questions on some design decisions. I apologize in advance if the answers are in the document and I missed them.

1. Hastelloy-N is the stated alloy for the Reactor Vessel, which Hastelloy-N is this? The original or the modified listed in ORNL-4541 (Page 26) where ~1 w/% Hafnium and 1-2 w/% Niobium was added? I assume the original since the modified was done in support of fission product corrosion control. Or something developed later?

2. If any particular passive cooling tower is damaged, or worse all three, has your team calculated how much decay heat removal cooling can be provided to that reactor module by the salt vault (if installed) cooled by the other modules? I understand this would not be passive heat removal but would be a backup system to the loss of any particular tower. What about forced draft fans in the passive cooling towers? For operational flexibility that would allow you to remove 1 or 2 towers or DHX's from service for maintenance. Would not to be relied upon for licensing of course, simply to provide operational flexibility.

3. Your 2 of 3 PHX design is interesting. If you have 1 PHX Main pump out of service:
a) Is there equal flow through the shutdown PHX and the 2 operating PHX's?
b) Are there axial temperature differences across the core with 1 of 3 PHX Main pumps out of service?
b) If there are temperature differences how do you account for neutron flux tilting across the core?

4. Have you given consideration to the shape of your burnable poisons to try and use some self shielding such that they will burn out at the same rate as the fuel is consumed? I.e., the total reactivity the movable rods have to compensate for is maintained in a relatively narrow band?


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PostPosted: Mar 10, 2011 8:34 pm 
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Lars wrote:
Just finished reading the report and learned a lot. There are several ideas that would be common with a LFTR. Question 2 is applicable in the context of a LFTR and would love to discover that it isn't a problem at all.

I have a few questions.
1) Was thorium considered as a burnable poison - would it work?
2) It seems like Smahtr will have virtually the same tritium production as a LFTR. It seems to me that the main tritium transfer path will be through the heat exchangers especially if a steam turbine is used. What is the forecast tritium release rate?
3) Your fissile consumption of 1600 kg LEU20 every 3 years for 125 MWth => 2.1 tonnes fissile / GWe-yr so at the end of life roughly half the fissile must still be in the fuel. Is it correct that the spent fuel contains roughly 10% enriched uranium? Is the plan that this is treated as once-through or is the plan to recycle the fuel in some form?
4) From the drawings it looks like an aircraft could take out all three passive cooling towers. Is this a plausible scenario? If so, how does the system handle it?


Lars,

Some comments - not necessarily answers - to your questions. We really have hardly scratched the surface of the design at this point, so many, many details are TBD. At this point, SmATHR is really a feasibility demonstration rather than an optimized concept.

1. We talked about, but did not analyze Th fuels and breeder core configurations (a la homogeneous or heterogeneous). Our focus has been on use of homogeneous GCR UCO TRISO fuel formulations with 40-50% TRISO particle packing. The idea, of course, is to lower the technical barriers to the extent possible.

2. I would expect the tritium generation and transport to be similar for similar salts. Perhaps 1 Ci / MW-d if it is indeed similar to the MSBR. This is one of the areas where the FLiBe isotopics (MSBR assumed 99.995% Li-7) will be important. The topic is briefly discussed on page 4-2 of the report. And yes, transport across the heat exchangers is a key dynamic. But, we've not had an opportunity to look at the production and transport in detail. As noted in the report, we expect continuous tritium removal from both the primary and secondary loop salts will be required. There are a number of chemical and physical approaches to this.

3. I'm sorry (been a long day) but somewhere you lost me... Assuming solid cylindrical fuel option, 1556 kg * .1975 / (125 *3.52) MWt-yr = 0.698 kg U5 / MWt-yr = 698 kg/GWt-yr. ? A 40% power conversion efficiency increases this to 1.745 MT U5 / GWe-yr ??? In any event, I don't recall the burnup for this core - but I know we have it archived in our calc results. It is something we paid attention to, but the results didn't make it into the report.

4. Fig. 5.6 in the report is misleading. We are aware of this threat. The details of the "cooling tower" (I prefer "chimneys") for the DRACS are all TBD. Relative placement of the chimneys, grade-level placement, and hardening approach all need to be examined. The 3 loops are obviously no advantage if a single credible common-mode failure or external event can take them all out.

HTH,
Sherrell


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PostPosted: Mar 10, 2011 9:08 pm 
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USPWR_SRO wrote:
I also just finished reading the report and found it very interesting, an impressive piece of work. I also have a few questions. My background is less of a design and more of a operational viewpoint. But I do still have questions on some design decisions. I apologize in advance if the answers are in the document and I missed them.

1. Hastelloy-N is the stated alloy for the Reactor Vessel, which Hastelloy-N is this? The original or the modified listed in ORNL-4541 (Page 26) where ~1 w/% Hafnium and 1-2 w/% Niobium was added? I assume the original since the modified was done in support of fission product corrosion control. Or something developed later?

2. If any particular passive cooling tower is damaged, or worse all three, has your team calculated how much decay heat removal cooling can be provided to that reactor module by the salt vault (if installed) cooled by the other modules? I understand this would not be passive heat removal but would be a backup system to the loss of any particular tower. What about forced draft fans in the passive cooling towers? For operational flexibility that would allow you to remove 1 or 2 towers or DHX's from service for maintenance. Would not to be relied upon for licensing of course, simply to provide operational flexibility.

3. Your 2 of 3 PHX design is interesting. If you have 1 PHX Main pump out of service:
a) Is there equal flow through the shutdown PHX and the 2 operating PHX's?
b) Are there axial temperature differences across the core with 1 of 3 PHX Main pumps out of service?
b) If there are temperature differences how do you account for neutron flux tilting across the core?

4. Have you given consideration to the shape of your burnable poisons to try and use some self shielding such that they will burn out at the same rate as the fuel is consumed? I.e., the total reactivity the movable rods have to compensate for is maintained in a relatively narrow band?


Darn, you guys are good !

Well let's see.

1. I don't recall that we actually discussed this nuance. The structural guys used the Hast-N properties from the 2010 ASME B&P Vessel Code, and I frankly don't know which Hast-N alloy is in there. The trace elements may not affect the yield strength ratings that much anyway. Maybe Jess remembers. We'll check it out.

2. You are touching on an interesting aspect of the SmATHR thermal energy system. The 2-of-3 always-on strategy for the PHX loops and DRACS loops results in a system in which all three 50% PHX loops (or three 50% DRACS loops) are running in parallel in a "throttled-back" mode. A failure of one simply results in the other two picking up the load. Yes, we could put a fan in the chimneys to enable higher flow in the event of a failure in one, but we don't need it for a single loop failure and it "corrupts" the passive philosophy. (smile) Now the salt vault is a really exciting aspect of the system (at least to me), because it accomplishes so many desirable functions in terms of storing energy and buffering the load from the reactors, the reactors from the load, and the reactors from each other. You'll notice from Figs. 7.1-7.2 that we have not plumbed the salt vault to allow it to be cooled by anything other than the user's heat load. This could be done, at the expense of greater system complexity. I would love to have a summer student this summer to put together a dynamic model of a multi-reactor salt vault system with all the relevant phenomena (including heat loss from the vault to its surroundings.

3. At this point, we've only had a chance to do a RELAP analysis of a complete loss of flow with 20 second coast-down, 10-second scram delay, and a normal transition to DRACS cooling. The flow through the PHX with the failed pump would not be very substantial. I think it more likely any resulting spatial temperature variations would be azimuthal in nature due to lack of desired mixing in the downcomer, but I don't know. Once again, I'm afraid I have to say we haven't yet had a chance to get into those details.

4. This is really an question for Jess, but we've really not looked very closely at the integrated reactivity control system (poisons and rods). But your idea would be something we would normally consider in the course of the layout.


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PostPosted: Mar 10, 2011 11:41 pm 
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Sherrell Greene wrote:
2. I would expect the tritium generation and transport to be similar for similar salts. Perhaps 1 Ci / MW-d if it is indeed similar to the MSBR. This is one of the areas where the FLiBe isotopics (MSBR assumed 99.995% Li-7) will be important. The topic is briefly discussed on page 4-2 of the report. And yes, transport across the heat exchangers is a key dynamic. But, we've not had an opportunity to look at the production and transport in detail. As noted in the report, we expect continuous tritium removal from both the primary and secondary loop salts will be required. There are a number of chemical and physical approaches to this.

I'll be interested to see your conclusions on this as we have virtually the same problem (made a tad more challenging by the presence of Xe and Kr).
Quote:
3. I'm sorry (been a long day) but somewhere you lost me... Assuming solid cylindrical fuel option, 1556 kg * .1975 / (125 *3.52) MWt-yr = 0.698 kg U5 / MWt-yr = 698 kg/GWt-yr. ? A 40% power conversion efficiency increases this to 1.745 MT U5 / GWe-yr ???

Not too different than my quick calculation (2.1 versus 1.745). What I noticed is that we actually use around 0.9 MT/GWe-yr so roughly half the original 235U is left in the fuel when you change it out. This reactor looks very attractive from its simplicity of design and deployment but it isn't fuel efficient. The fuel should still have plenty of life left in it though so it could be reused in a number of ways even without any processing.

Quote:
HTH,
Sherrell

HTH = hope this helps?

Thank you and yes it does. Do you expect you will do a follow-on to this?


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PostPosted: Mar 11, 2011 6:44 am 
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A couple of quick responses:

On the burnable absorber question - we have not looked at reactivity control for this system yet and have not done calculations with burnable absorbers. We did do burnable absorber calculations for the larger AHTR system back in 2005 and spent considerable time on their location in the fuel assembly. There are many options available to us. USPWR_SO is exactly right, we need to manage the reactivity as much as we can with burnable absorbers to minimize the use of control rods and number of control rods.

On the fuel utilization comments/questions - certainly with high enrichment and TRISO fuel, the fuel cost for this reactor is going to be more expensive than, say, LWR fuel. Even with that, we do not believe that the fuel costs will be an issue given the unique aspects of this reactor (high temperature, high thermal efficiency). Regarding fuel management, there are several options that can increase the fuel utilization at the cost of taking an outage and if we went to a three batch refueling we would be able to increase the fuel utilization by 50%. In the end it may make sense to do this for economic reasons. And, given the simple nature of this reactor, perhaps the outage can be relatively short. But, our goal for this concept was a continuous period of operations is four years, which can only be achieved with a single batch core.

Any yes, we hope to continue to work on this concept.


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PostPosted: Mar 11, 2011 7:00 am 
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Jess Gehin wrote:

Any yes, we hope to continue to work on this concept.


How was the original work funded (DOE/NE, LDRD, other internal funds?) and what are the prospects for future funding?


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PostPosted: Mar 11, 2011 7:14 am 
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Sherrell Greene wrote:
... I think it more likely any resulting spatial temperature variations would be azimuthal in nature due to lack of desired mixing in the downcomer, but I don't know. Once again, I'm afraid I have to say we haven't yet had a chance to get into those details...



Thanks. I wrote axial and I meant radial temperature shifts. Resulting in an azimuthal neutron flux tilt. Such is the sign of true genius, answering the question that was meant, not the one that was communicated. :lol: :lol:

Thanks again.


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PostPosted: Mar 11, 2011 8:42 am 
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arcs_n_sparks wrote:
Jess Gehin wrote:

Any yes, we hope to continue to work on this concept.


How was the original work funded (DOE/NE, LDRD, other internal funds?) and what are the prospects for future funding?


This was an internally funded study to address what we feel is a gap in available nuclear reactor concepts. Specifically, a small, modular reactor for high temperature applications. Most of this work was performed last summer and it took some time to get the report finalized, reviewed and published, but we are glad to have it out and see folks like you all reading the report and commenting on it. We believed that there will be future opportunities for funding and several of the things that we learned in this study are being carried over to the work that we are doing on the larger AHTR system.


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PostPosted: Mar 11, 2011 5:15 pm 
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Here's the back story..

The AHTR was developed in the 2002 timeframe as the first or archetype "salt-cooled" reactor. Our concept work at that time was focused on the large, central-station electricity generation applications. AHTR attracted a lot of attention and educated many about the potential for this new type of very high temperature reactor (VHTR).

Subsequently, some limited additional AHTR concept development work was done on AHTR, re-labeled as the "liquid salt VHTR".

In 2008, we did a study of the potential issues associated with co-location of nuclear plants and bio-refineries (report ORNL/TM-2008/102). One of the major conclusions of that study was that something like 125 MWt was an ideal size for the reactor for such applications.

As I looked more and more into potential process heat markets, and thought about realistic development/deployment issues, technology evolution, etc, it became clear (at least to me) that "small is especially beautiful" if one is seeking to (finally) open the process heat market for nuclear power. And delivering affordable high-temperature heat could be transformational. So why not a small AHTR? We originally investigated a natural circulation SmAHTR concept, but a it became clear a Nat-Circ SmATHR would have difficulty achieveing the power level and physical size desired. Thus SmATHR as you see it today.

Then there's the salt vault. Though it's just a crude concept at present, I hope you can see from the report that something like the salt vault has the potential to enable multi-reactor deployments and applications that would be difficult to achieve without the high-temperature, compact, thermal energy storage potential of salt.

So, our goals with SmAHTR were to (1) demonstrate the feasibility and promise of small FHRs, (2) Identify the central design trades, (3) layout a system architecture and technology evolution path that can take us to much higher temperatures through time, and (4) address some of the process heat customer integration issues identified in our 2008 Nuc/Biofuels study – but in a generic manner that isn't tied solely to bio-refineries. In short, demonstrate such a significant potential value proposition for small AHTRs that those seriously interested in VHTRs will be compelled to take notice.

AHTRs are "the other VHTR". One size does not fit all. Only time will tell if they are really a better VHTR...

Sherrell

arcs_n_sparks wrote:
Jess Gehin wrote:

Any yes, we hope to continue to work on this concept.


How was the original work funded (DOE/NE, LDRD, other internal funds?) and what are the prospects for future funding?


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PostPosted: Mar 12, 2011 9:33 pm 
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Kirk,

Provided ondrejch agrees, can you move this discussion thread to the salt-cooled rectors area of the forum?
That's really where it belongs now that we have that area established.

Thanks!
Sherrell


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PostPosted: Mar 13, 2011 12:58 am 
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Sure, I think the whole thread can be moved over to a different category by a moderator.


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PostPosted: Mar 13, 2011 12:50 pm 
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The SmAHTR should be easy to prototype and decommission. One might want to build a 7 megawatt thermal prototype (similar size to the MSRE), and just throw away the heat. The fuel is a stable form suited for long term dry storage after a few years or geological disposal. Certainly more stable than spent LWR fuel. The reactor might be small enough to just put on a truck and take to a low level waste repository. This prototype should demonstrate all the passive cooling features and chemistry. Then everyone would get excited about the concept and would inspire the next round of funding.

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PostPosted: Mar 25, 2011 9:52 pm 
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I would much rather have a salt cooled TRISO fuel reactor underwater than a salt/fuel/fission product reactor.

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