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PostPosted: Sep 05, 2008 1:37 pm 
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Axil wrote:
The core and blanket only contain U233/U232, and only inextricable amounts of PU(***).

So back to the question of "how much U232, and where to get it," right ?


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PostPosted: Sep 05, 2008 1:56 pm 
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Axil wrote:
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Decay heat removal is integral to reactor safety since a reactor continues to produce heat at 7% of full power at shutdown and continues to be over or near 1% for several days (this is from all the radioactivity, not any fissions).



Wow, lots of posts in the last 24 hours. I tried to just "quote" the post you had the graph of how decay heat goes through the reactor but I guess it doesn't repeat attachments. Getting back to dealing with decay heat. Yes, that graph showed how a pebble bed (or perhaps a graphite matrix design?) deals with decay heat. You can't just assume that since we also want to use pebbles that it will also be OK. Pebble bed (or graphite matrix) designs are extremely limited on how they are built if they want to get rid of decay heat. They end up with bizarre annular cores etc so that all fuel elements are close to the outer walls. If we are looking to have a thorium blanket salt surrounding the core, we have to plan on that blanket salt draining (on purpose or by accident). In that case with a large insulation space there is no way that pebbles in the central core would not surpass the temperature limit from decay heat. You need a nicely conductive pathway for heat to get to the outside world. Yes, if we keep the salt circulating and externally cool the salt then we can take decay heat away from the pebbles, but then we getting back to engineered safe guards and triple redundancy instead of inherent safety.

Bottom line is if you want TRISO fueled pebbles you are adding an entirely new level of difficulty in how to handle all possible events or accidents.

I'll remind you that the system I have proposed for starting up on LEU (Low Enriched Uranium) is to use it in the central fuel salt as UF4 for the first year (or first few years) until enough U233 has built up in the blanket (kept in the blanket salt, or otherwise removed and saved). Once enough U233 is in saved up, the reactor can be restarted on the pure Th-U233 cycle. Similar to your idea of using TRISO fuel pebbles, the 1st cycle "dirty" salt should be processed to return Pu and other transuranics back into the pure Th-U233 cycle. However, the processing methods are the same as needed for the pure Th-U233 cycle (just a lot more transuranics to remove). In the TRISO mode of fuel in the pebbles, you either have to live with throwing out a good deal of transuranics or spend a great deal of money to process them out.

There are perhaps some small advantages over shipping LEU inside TRISO pebbles as opposed to shipping LEU as UF4 but I don't think the differences are that big.


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PostPosted: Sep 05, 2008 1:57 pm 
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Quote:
So back to the question of "how much U232, and where to get it," right ?


I will answer your question with a question.

Don’t U232 and other nuclear denaturants as well as U233 come as a natural consequence of thermal neutron radiation of thorium salt; which sometimes comes primarily from TRISO fuel (at startup) and sometimes comes primarily from U233 decay (after startup)?

As an armature, I would keep the reaction right at criticality and hope for the best. A professional would run a huge number of reaction simulations under every posible condition to find out what to adjust in each situation to get the results that he wants; which just might work.

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PostPosted: Sep 05, 2008 2:12 pm 
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Axil wrote:
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One could add the minor actinides from the LWR waste aswell but it would lower fuel utilisation. No reprocessing would be worthwhile after such a high burnup, unless to lower the long term radiotoxicity.



Note,TRISO radiation damage limit is 8 10exp21.



Axil,

Again your graph doesn't show up in my response but it was the one showing how much fast neutron damage TRISO fuel pebbles can handle. Do you know the exact units and conditions this quotes? Typically when we say a fast neutron flux we quote above some level. Oak Ridge work usually talks about fluences above 50 keV whereas I've seen other work quoting above 180 keV. Oak Ridge graphite in the form of hexagonal logs was thought to be good for a fast neutron fluence of 3 x 10exp22 n/cm2 (>50 keV). So if TRISO is only good for 8 x 10exp21 it is actually a much lower possible fluence before needing replacement. I was hoping the pebbles would be better than that but it probably makes sense as pebble bed systems are usually very low power density (to help with heat removal, including decay heat). Most of the molten salt systems we'd want to use them in would be pretty high power density (higher neutron flux) so they might only have a lifetime of a year or two (unless there is indeed a problem with the units or conditions of the quote above).

David L.


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PostPosted: Sep 05, 2008 2:15 pm 
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Axil wrote:
Reactor Cavity Cooling

The reactor cavity cooling system (RCCS; see Figure) is a constant flow, water-based cooling system that removes heat from the reactor cavity to protect the concrete walls of the cavity during both normal shutdown and accident conditions. It is comprised of standpipes that
line the inside of the cavity, and is a low-temperature, low pressure system with water temperatures below 30°C during normal active operation and reaching the boiling point only during emergency passive operation. The RCCS can operate in both an active mode by pumping
water through the standpipes, or a passive mode by boiling the water for approximately 72 hours. The passive mode time constraint is yet to be defined.



Im not sure I see the connection betwen your reply and what I wrote :?: Anyway the passive cooling system I mentioned a msr should have is the one you depicted in the attached picture.

Axil wrote:
In the diagram, all reflectors are graphite, only the outer reflector is permanent. I also suggest that the MSR design team shop around for a proven modular pebble bed reactor design (there are several about) as a development template.
.


Depends on the design, some designs also has a permanent center reflector like the on in your picture. In other designs dummy pebbles play the same role.

But back to a main point I was trying to make that probably got lost. You cant assume that TRISO fuel is inherently safe in any reactor configuration. The current HTR designs are small, thin and long to enable passive cooling to be enough to ensure safety during all situations. But if you make the reactor larger for instance passive cooling is no longer enough to keep the fuel below temperature limits. The center reflector is a example of a way to squeese more power out of the design without making it larger, flux flattening from the reflector limits peak temperatures in a loss of coolant accident.

A molten salt cooled HTR could probably be larger due to better heat transfer properties of the salt compared to helium, but then one has to ensure that the salt can never drain from the core or otherwise the pebbles will become to hot.

Edit, i just saw David beat me to it!


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PostPosted: Sep 05, 2008 2:18 pm 
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Axil wrote,


Quote:
The design of a reactor where the TRISO fuel never exceeds 1600C degrees in case of loss of coolant accident has been tested in china on their prototype test bed.

This is a worse case failure condition that covers the situation in which all molten salt coolant is lost in both the core and the blanket.



As I just mentioned. If you have a design that has the needed thorium blanket before the outside containment wall, there is no way that the pebbles will not overheat if both the core and blanket salt are lost (or drained).

David L.

Edit, Johan has beat me to this second comment!


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PostPosted: Sep 05, 2008 2:43 pm 
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David wrote:

Axil,

Again your graph doesn't show up in my response but it was the one showing how much fast neutron damage TRISO fuel pebbles can handle. Do you know the exact units and conditions this quotes? Typically when we say a fast neutron flux we quote above some level. Oak Ridge work usually talks about fluences above 50 keV whereas I've seen other work quoting above 180 keV. Oak Ridge graphite in the form of hexagonal logs was thought to be good for a fast neutron fluence of 3 x 10exp22 n/cm2 (>50 keV). So if TRISO is only good for 8 x 10exp21 it is actually a much lower possible fluence before needing replacement. I was hoping the pebbles would be better than that but it probably makes sense as pebble bed systems are usually very low power density (to help with heat removal, including decay heat). Most of the molten salt systems we'd want to use them in would be pretty high power density (higher neutron flux) so they might only have a lifetime of a year or two (unless there is indeed a problem with the units or conditions of the quote above).

David L.


There seems to be some uncertainties about how high fluence they can take. Computer codes predict failure rates at around 1% when fluence > 0.1 MeV approach 10^22 n/cm^2 but experiments has shown TRISO's seems to be surviving fine even over 10^25 n/cm2. This thesis(I have been slowely working my way through this behemot during the last 2 months since I arrived here in Delft) discuss it abit at page 24 http://www.tnw.tudelft.nl/live/pagina.j ... /bende.pdf

TRISO's where the silicon carbide layer is replaced with zirconium carbide is also beeing investigated and it seems like they can take higher fluences and temperatures but I havent read much about it yet.


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PostPosted: Sep 05, 2008 3:08 pm 
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Quote:
Do you know the exact units and conditions this quotes?



I did some leg work. All the numbers is saw reflect the maximum fluence below.

• Mainly based from the past German, American and Russian experience, it has been proven that current HTR UO2 fuel could achieve at an industrial scale the following
performances:

a burn-up ~ 10% FIMA (max 15 %FIMA achieved)

a maximum fluence ~ 2 to 3 x 10exp25 n.m-2, E>0.1 MeV

a temperature limit under accidental conditions of 1600°C

a failure rate ~ 10exp-5

My assumption is that for the 80% beep burn TRISO fuel the maximum fluence will to be somewhat greater.

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PostPosted: Sep 05, 2008 11:55 pm 
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Quote:
Getting back to dealing with decay heat.


David L, you are requiring that I solve the heat of decay at shutdown/loss of coolant condition. I hope I don’t inadvertently uncover your patent concepts dealing with this subject.

See my following post for my design assumptions.

Design Suggestion

The following describes a fail safe passive reactor cooling mechanism.

The outside double walls of both the core and the blanket tanks are hollow, forming a thin planar heat pipe (heat spreader) through hermetically sealing a 1mm space formed by the double walls thus forming a very thin hollow vessel in which a working heat transfer fluid operates in a closed-loop capillary recirculation system.

At the time of reactor construction, this heat pipe is formed by using a vacuum pump to exclude all gases and liquids from the empty heat pipe, and then the pipe is filled with a fraction of a percent by volume of a working heat transfer fluid, (sodium fluoride as coolant), which has be chosen to match the reactor operating temperature.

Due to the partial vacuum that is near or below the vapor pressure of the fluid, some of the coolant will be in the liquid phase and some will be in the gas phase.

Inside the pressure walls, an optional wick structure exerts a capillary pressure on the liquid phase of the sodium fluoride working fluid. This may be either finely powered graphite, fullerenes, sintered metal powder or some other material capable of exerting capillary pressure on the condensed liquid sodium fluoride to wick it back to the heated end. The heat pipe may not need a wick structure if gravity or some other source of acceleration is sufficient to overcome surface tension and cause the condensed liquid to flow back to the heated end. The design of this coolant flow system is yet to be defined

Heat pipes contain no mechanical moving parts and typically require no maintenance,.

The coolant, sodium fluoride, has been tentatively chosen based on the temperature conditions that the heat pipe must operate, (993–1700 C)

The advantage of this heat pipe is its great efficiency in transferring heat. It is a much better heat conductor than an equivalent cross-section of solid copper. In some applications, a heat flux of more than 230 MW/m² has been recorded

If detailed design warrants in normal operation, the heat pipe transfers reactor wall heat to the top of the reactor where a heat exchanger preheats cooled molten reactor salt which has just exited the turboelectric generator. When reheated by this sodium fluoride heated reactor salt heat exchanger procedure, it then reenters the reactor in its normal salt circulation cycle pattern.

The key enabling design factor is the selection of the appropriate coolant salt whose heat range provides the appropriate reactor wall temperature required for low cost steel reactor vessel construction. This design is yet to be determined. The Ideal salt temperature range for this requirement is 400 – 1700 C.

In a shutdown or coolant loss condition, if molten reactor salt is drained from the reactor core and/or blanket into the associated molten salt storage tank(s) under the reactor vessel by the melt of the reactor core plug(s), the heat exchanger is air cooled by fans to dissipate heat from the reactor pipe heat.

If heat pipe over temperature results from failure of the turboelectric generator, if required, air fans will automatically kick in to cool the reactor wall heat exchanger.

One additional advantage of this plan may be cool reactor wall temperatures.

This may enable the use of SA508/SA533 steels in the construction of the reactor vessel due to low reactor wall operating temperatures enabled by the heat pipe. This temperature will be close to the cool reactor salt temperature exiting from the turboelectric generator.

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Last edited by Axil on Sep 06, 2008 2:32 pm, edited 4 times in total.

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PostPosted: Sep 06, 2008 1:47 pm 
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johan wrote:
But back to a main point I was trying to make that probably got lost. You cant assume that TRISO fuel is inherently safe in any reactor configuration. The current HTR designs are small, thin and long to enable passive cooling to be enough to ensure safety during all situations. But if you make the reactor larger for instance passive cooling is no longer enough to keep the fuel below temperature limits. The center reflector is a example of a way to squeese more power out of the design without making it larger, flux flattening from the reflector limits peak temperatures in a loss of coolant accident.


I am a proponent of the “small reactor” school of design for the following reasons:

A low power density is a prerequisite for a safe reactor design. Passive safety measures, the prime attribute of a safe reactor becomes increasing ineffective as the power density of a reactor increases.

The economic disadvantage of a small vis-a-vis a large reactor can be compensated for through the mass production manufacture of the small reactor on an automated factory production line where consistent high level quality control can be applied to the robot enabled modular construction of the small reactor.

A small reactor core (about two or three meters) can be manufactured by many pressure vessel production companies. This competition through low bid selection insures the lowest reactor core production cost.

The complete fail safe and secure automation of the small reactor besides eliminating human operator error and human radiation danger also greatly increases reactor security and proliferation resistance.

This automation also provides small reactor clustering that can support the same levels of economies of scale that the large reactor can provide.

A small reactor cluster can grow in small increments to meet increasing power demands while keeping the need for construction capital low at any given time.

Small reactor construction and commissioning times are kept to two years or under.

The zone of danger for any emergency response for a reactor element in a cluster is consistently small. This restricts emergency response to the nuclear complex and not throughout the surrounding country side. In my opinion, this is a requirement for good public relations with the neighbors of nuclear power.

A reactor cluster is never completely unavailable because of the n – 1 availability characteristic of cluster topology.

Reactor maintenance and refueling of a cluster reactor element does not result in a cluster shutdown.

A centralize cluster control room provides a single point of control throughout the cluster.

This line of thinking provides the basic assumptions that form the foundation of the design suggestions that I offer on this thread.

With all those aforementioned design priories in mind, an important MSN reactor design goal is to maximize power output at the minimum reactor construction cost. My guess is that the ideal reactor power level is anywhere between 100 to 300 Megawatts.

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PostPosted: Sep 06, 2008 10:55 pm 
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Axil wrote:
Quote:
Getting back to dealing with decay heat.


David L, you are requiring that I solve the heat of decay at shutdown/loss of coolant condition. I hope I don’t inadvertently uncover your patent concepts dealing with this subject.

See my following post for my design assumptions.

Design Suggestion

The following describes a fail safe passive reactor cooling mechanism.

The outside double walls of both the core and the blanket tanks are hollow, forming a thin planar heat pipe (heat spreader) through hermetically sealing a 1mm space formed by the double walls thus forming a very thin hollow vessel in which a working heat transfer fluid operates in a closed-loop capillary recirculation system.

At the time of reactor construction, this heat pipe is formed by using a vacuum pump to exclude all gases and liquids from the empty heat pipe, and then the pipe is filled with a fraction of a percent by volume of a working heat transfer fluid, (sodium fluoride as coolant), which has be chosen to match the reactor operating temperature.

Due to the partial vacuum that is near or below the vapor pressure of the fluid, some of the coolant will be in the liquid phase and some will be in the gas phase.

Inside the pressure walls, an optional wick structure exerts a capillary pressure on the liquid phase of the sodium fluoride working fluid. This may be either finely powered graphite, fullerenes, sintered metal powder or some other material capable of exerting capillary pressure on the condensed liquid sodium fluoride to wick it back to the heated end. The heat pipe may not need a wick structure if gravity or some other source of acceleration is sufficient to overcome surface tension and cause the condensed liquid to flow back to the heated end. The design of this coolant flow system is yet to be defined

Heat pipes contain no mechanical moving parts and typically require no maintenance,.

The coolant, sodium fluoride, has been tentatively chosen based on the temperature conditions that the heat pipe must operate, (993–1700 C)

The advantage of this heat pipe is its great efficiency in transferring heat. It is a much better heat conductor than an equivalent cross-section of solid copper. In some applications, a heat flux of more than 230 MW/m² has been recorded

If detailed design warrants in normal operation, the heat pipe transfers reactor wall heat to the top of the reactor where a heat exchanger preheats cooled molten reactor salt which has just exited the turboelectric generator. When reheated by this sodium fluoride heated reactor salt heat exchanger procedure, it then reenters the reactor in its normal salt circulation cycle pattern.

The key enabling design factor is the selection of the appropriate coolant salt whose heat range provides the appropriate reactor wall temperature required for low cost steel reactor vessel construction. This design is yet to be determined. The Ideal salt temperature range for this requirement is 400 – 1700 C.

In a shutdown or coolant loss condition, if molten reactor salt is drained from the reactor core and/or blanket into the associated molten salt storage tank(s) under the reactor vessel by the melt of the reactor core plug(s), the heat exchanger is air cooled by fans to dissipate heat from the reactor pipe heat.

If heat pipe over temperature results from failure of the turboelectric generator, if required, air fans will automatically kick in to cool the reactor wall heat exchanger.

One additional advantage of this plan may be cool reactor wall temperatures.

This may enable the use of SA508/SA533 steels in the construction of the reactor vessel due to low reactor wall operating temperatures enabled by the heat pipe. This temperature will be close to the cool reactor salt temperature exiting from the turboelectric generator.



Axil,

My preference is typically to have the simplest solutions possible as elaborate solutions often introduce new problems. Heat piping is possible but would introduce a whole range of issues. For example, if you are planning for accidents than might drain the core and/or blanket salts, it is hard to be assured a liquid would not also be lost between walls. As well, the melting point of NaF is much higher than the temperature we might be looking to operate. Yes, many people assume that we`ll just run everything using carbon composites or SiC/SiC and have it as hot as we`d like but I prefer to plan for the materials and temperatures we have already proven (Hastelloy N, max 705 C), anything beyond that can be considered a bonus.

Again, I am not saying the concept of TRISO pebbles to start up a cycle is out of the question but it seems to me there are much simpler avenues to reach the same goals. Now, if we are talking about just using pebbles as moderator and take advantage of all the knowledge gained from Germany and elsewhere then I am quite interested. I am happy to hear that the pebbles themselves might be good for a fast neutron flux of upwards of 10^25n/cm, if that can be verified they might lure me away from my current preference of having no graphite at all in the cores (at least for some designs).

David L.


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PostPosted: Sep 07, 2008 2:22 am 
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David wrote:
Axil wrote:
Quote:
Getting back to dealing with decay heat.


Quote:
My preference is typically to have the simplest solutions possible as elaborate solutions often introduce new problems.



True, simplicity is the golden rule of the engineering art, but on the other hand, you can’t have too many passive safety systems in a reactor design. Overdesign is not a vice when it comes to reactors.

Quote:
As well, the melting point of NaF is much higher than the temperature we might be looking to operate.


Selection of the heat pipe salt would have been determined by the operational/emergency temperature range of the reactor. Remember, the heat pipe would need to operate at very high emergency temperatures.

Quote:
Heat piping is possible but would introduce a whole range of issues. For example, if you are planning for accidents than might drain the core and/or blanket salts, it is hard to be assured a liquid would not also be lost between walls.



Loss of liquid between the walls is not a critical melt down event since its occurrence does not interfere with the normal molten salt cooling cycle. Failure of one face of a double wall still leaves the wall fully functional, only the heat pipe has failed.

Furthermore, wall failure is detected by a pressure rise in the low pressure environment between the double walls. Pressure sensors send an immediate operator alarm which alerts operational personnel that a failure in this passive safety system has occurred. Shut down will proceed with the circulation pump operating.

Also, a double wall with failure detection is far safer than a single wall without failure detection.

Next, the heat pipe suggestion and the TRISO fuel suggestion are not necessarily connected.

Its been a reactor design assumption that the meltable molten salt core drain plug is 100% reliable. That is not good.

A failure of the meltable molten salt core drain plug is more probable than reactor wall failure. If light water reactor fuel or even U233 is used in reactor startup and both the core molten salt circulator pump and the drain plug fails, then you have a problem. A heat pipe will protect against this situation.

Come to think about it, you won’t know you have a faulty drain plug until the molten salt circulator pump fails and the reactor temperature goes hot. That’s when the drain plug might not work.

The heat pipe is an alternate heat removal system to the circulator pump. It might be less of a cleanup effort if the drain plug can be kept in tact if the heat pipe can drain enough heat through a circulator pump failure; and it might.

Please allow me to try to preempt some criticism of my scrupulous and perfectionist nature.

Remember, designers, builders and users of the Titanic (as well as the Olympic) claimed that those ships were unsinkable. This and other engineering disasters teach that designers and engineers must leave no stone unturned; many times over and then humbly hope for the best.

But if all I have done is to convince you to keep the power density low in your reactor design, for safety sake then I have done something to be proud of.

I know when to stop beating a dead horse. All good things must come to an end, but this conversation with you; an eminent and good natured professional, has been a great thrill and an inspiration for this amateur. I am very grateful, thank you.

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PostPosted: Sep 07, 2008 7:59 am 
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David wrote:
I am happy to hear that the pebbles themselves might be good for a fast neutron flux of upwards of 10^25n/cm, if that can be verified they might lure me away from my current preference of having no graphite at all in the cores (at least for some designs).

I think Axil was using different units:
Axil wrote:
a maximum fluence ~ 2 to 3 x 10exp25 n.m-2, E>0.1 MeV


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PostPosted: Sep 07, 2008 10:29 am 
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jaro wrote:
David wrote:
I am happy to hear that the pebbles themselves might be good for a fast neutron flux of upwards of 10^25n/cm, if that can be verified they might lure me away from my current preference of having no graphite at all in the cores (at least for some designs).

I think Axil was using different units:
Axil wrote:
a maximum fluence ~ 2 to 3 x 10exp25 n.m-2, E>0.1 MeV


Thanks Jaro, I did miss that it was m^2 instead of cm^2. However, Johan made a comment with roughly the same 10^25 values but in n/cm^2, I guess his comment was only that test showed they "seemed" to be fine up to 10^25. I know your a heavy water fan Jaro but have you heard any lifetimes for pebbles? The complication is that I am mainly interested in just plain pebbles but with a pyrolytic coating for fire safety. Most pebble bed work worry about fission gas release which we don`t care about at all, in fact, we`d prefer the fission gasses to be released into the surrounding salt. Actually that raises another problem with using TRISO fuels, we`d have all the problems of Xenon including the hit to the neutron economy.

My guess would be that pebbles are probably pretty similar lifetime to fast neutron fluence
compared to graphite logs or blocks. Graphite first shrinks and then expands and once you`ve surpassed the original volume I`d be worried about pebbles cracking and if they start falling apart you`ve got a serious issue for pumps and heat exchangers.


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PostPosted: Sep 09, 2008 1:56 pm 
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Quote:
I’d be worried about pebbles cracking and if they start falling apart you’ve got a serious issue for pumps and heat exchangers.

David, I need one more whack at that dead horse.

Design Suggestion.

I am much impressed with the uranium hydride reactor but I believe it can be improved upon through the use of molten thorium salt.

For example, it does not support generation of electricity and/or hydrogen from its heat output. The customer or third party OEM is required to provide this function.

Nevertheless, some of its design features and methods can be drawn to advantage for our thorium salt reactor. Like both uranium hydride and thorium hydride, molten thorium salt is self regulating in terms of temperature. I would like to put that molten thorium salt nuclear reaction characteristic to work for us.

Let’s eliminate the molten thorium salt circulators and heat exchangers in our current thinking and replace them with a redundant system of heat pipes and associated heat exchangers to carry reaction heat to the turboelectric generators and like equipment. Further, the heat pipes can be designed to provide suitable redundancy and overcapacity in addition to fail-safe cooling to dissipate energy generated by the decay of radioactive fission products.

I see a number of advantages in doing this as follows:

The molten salt circulators will have a low mean time between failure expectations. They will be operating in a very corrosive environment of high radiation, very hot corrosive and viscous molten salt, and if graphite pebble moderation is used, possible broken graphite and/or silicon carbide fragments from graphite pebble moderation, internal core graphite reflectors and/or TRISO startup fuel.

Removing them from the design will substantially improve reactor availability and safety. The core and blanket reactor operation will have no moving parts and will be completely passive.

Simplicity in the design will be increased to maximum levels. Such a nuclear power reactor would be inherently fail-safe from over-temperature excursions and may be mass-produced in some variant designs as turnkey modules due to this inherent design simplicity and compactness.

In addition, this will greatly reduce the complexity of the automation needed for automatic control.

Heat pipes operate effectively up to lengths that are limited by the aspect ratio (ratio of length to diameter) of the devices of about 100 to 1. Both the core and blanket outside wall can be used for planar heat piping containing a metal screen made of the same metal as the core and positioned to act as a wick for the working fluid. In addition, an axially located heat pipe in the core can collocate with the centrally located graphite reflector and pebble grappler.

The planar heat piping of the core wall is redundant to eliminate the possibility of total core heat pipe failure. The possibility of core wall failure is extremely small; a double core wall will bring that probability to zero.

Continuous removal of nuclear gaseous poisons and tritium can be done from the top of the reactor by a negative pressure extraction system. This is an improvement over the uranium hydride reactor design.

The molten thorium salt will reach thermal equilibrium inside the core with more nuclear activity near the heat pipes and less in isolated volumes away from the heat pipes.

The best core design is consistent with a cylindrical “LeBlanc” geometry: ‘Tube-Within-Shell’ design and oriented to the vertical.

The diameter will be a bit over one meter with no graphite moderator or the diameter will be a bit over two meters with graphite pebble moderation.

Graphite moderation is optional but heat pipes only heat transfer removes the problem of graphite deterioration. Graphite pebble rubble will harmlessly settle to the bottom of the core.

Fuel enrichment of U233 from the blanket to the core can be done in a reserved space at the bottom of the reactor where an automated processing system and separate neutron shielded storage tanks are not human accessible eliminating the need for denaturants.

The redundant heat exchangers one per heat pipe, transfer heat from the heat pipe working fluid to 70 bars helium. This will enable the use of all the GEN IV helium enabled peripheral equipment including electric generation and hydrogen production.

In the event of a helium gas failure of any heat pipe circuit, high volume redundant air fans transfer core heat from the surface of the failed heat exchanger to the reactor enclosure space where it is removed by the emergency stand pipe water cooling system during shutdown(see previous post).

The heat pipes will handle shut down latent radiation heating. Such radioactive decay energy starts at 7% of the operating power of the reactor but rapidly decays from that value.

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