Energy From Thorium Discussion Forum

Lars, my thoughts exactly.

I am no expert either. But isn't one of the large advantages of the MSR the strongly negative coefficient of reactivity WITHOUT any of the time delay effects of solid fuel reactors?
As I understand, this negative coefficient comes from the liquid fuel expanding out of the core. No time-lag from the delta-T between solid fuel and moderator. So large criticality excursions should be impossible because density changes in the fuel salt are communicated at sonic speed in the salt to the control mechanism (fuel expanding out of moderated core).

So, at first glance, it seems to me the delayed neutron effect would not play a role in the LCTR or lack thereof would not make an LCTR more difficult to control than an LFTR.

What have I overlooked?

That works in a LFTR where the delayed neutron component makes the reactivity swings occur over minutes, but from what I understand in a fast neutron reactor the reactivity swings can occur over rather small fractions of a second. Thermal expansion of the fuel in a fast reactor stops the reaction eventually but from what I remember, you're dependant on doppler broadening rather than the negative temperature coefficient to actually prevent the criticality excursion in the first place. If you have no place to dump the excess gigawatts from the criticality excursion, you could be facing a rather nasty accident.

Its also important to remember that different salt makeups can cause the fuel to have a positive temperature coefficient, which is why many of us here support two fluid reactors for safety reasons.

And a LCTR wouldnt have a moderated core.

Klaus writes "criticality excursions should be impossible"...

Hmmmmm.... That sounds really bad.... Three Mile Island was a criticality excursion??

By the way, I think I should post the reason why I started the post so we can stay on topic. My interest is in bringing the subject of Thorium and Molten Salt Reactors to a general audience. I must admit that I have been struggling with acronyms. I finally figured out that ORNL refers to Oak Ridge National Laboratory, not a place in France.

Anyways....

There are different groups of people who have different interest. Some people are interested in cheap, safe, abundant power. When speaking with these people it is clear that conversation should center on the virtues of the Liquid Fluoride reactors. It meets all their needs.

Other people are concerned about nuclear waste. These would include certain environmental groups. It would also notably include politicians from Nevada. These people may or may not be interested in a Thorium reactor. They certainly *would* be interested in a way to stop developing Yucca mountain as a waste dump. One of the things I have been thinking of doing is writing letters to the leading politicians in Nevada. The idea is that a conversation could be started about using the next generation of nuclear reactors as opposed to creating a waste dump...

Also, in case you think that I am just all BS.... I contacted the national headquarters of the Sierra Club earlier this week. I had a polite conversation with the person who answered the phone. I explained my interest and he gave me the phone number of a person on the energy committee in Vermont. I had a brief but interesting conversation with this person. I doubt that my actions will have any impact on the Sierra Club's policy. That said, I am proud of myself for reaching out to someone in the environmental community with the intention of making a real difference in the world. I will continue to use my creative powers to reach people and get them excited about Thorium power.

PS. I have shared with the people at my office and *they* are excitied about Thorium. Several people have Thorium posters on the walls of their cubicles.

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Three Mile Island (TMI) was a loss of heat removal from the primary system. I won't go into all the boring details but all fission reactors generate what is called decay heat from the daughter isotopes of the fission reaction. This heat must be removed for a long time or you will damage the core. Near worst case for this event was TMI where about 40% of the core was completely destroyed and the rest damaged beyond repair.

A reactivity excursion causing an uncontrolled criticality or if already at power an uncontrolled fission power excursion (this is the correct terminology for what I think you were trying to say) is an entirely different accident. This is what happened at Chernobyl. Obviously this type of accident has the potential for much greater consequences. A typical decay heat load at a large power station after core shutdown is in the tens of megawatts. A reactivity excursion on the other hand can release hundreds or thousands of megawatts of energy (even if only for a few seconds) causing catastrophic damage.



This is a good idea. Anyone using an acronym here should probably follow the convention and spell it out the first time with the acronym in paranthesis just after it, e.g., Pressurized Water Reactor (PWR). We should also have an acronym dictionary on the forums sticked somewhere. If we don't have one we should make one. Example:

LWR = Light Water Reactor
HWR = Heavy Water Reactor
PWR = Pressurized Water Reactor
BWR = Boiling Water Reactor
ORNL = Oak Ridge National Laboratory
ANL = Argonne National Laboratory
LFTR = Liquid Fluoride Thorium Reactor
MSBR = Molten Salt Breeder Reactor

A good overview of fluoride versus chloride over the last several posts. A couple things I think are a little off target though.

A faster spectrum isn't more difficult to control because of differences in the delayed neutron fraction but because the prompt neutron lifetime. The delayed neutron fraction is quite different between different fissile isotopes (large for U235, small for Pu239 and U233) but I don't think there is much difference between different neutron spectrums.

What is different is the lifetime of prompt neutrons. It is much shorter for fast reactors. Thus if you ever do have an event that pushes reactivity past the buffer delayed neutrons provide, the power rises much faster for a fast reactor than a thermal one. As long as you have a good negative temperature coefficient though things still should be fine.

A fluoride reactor can indeed run on only actinides (Pu,Am,Np etc) and this is the proposal of current Russian work called MOSART (MOlten Salt Actined Recycle and Transmutation). It is a design without graphite (except as reflectors to protect the outer vessel) but not a really hard spectrum. The limited solubility of PuF3 is an important issue but you can still run on only TRUs if you pick your salt right.

A chloride reactor or fluoride (thermal or fast) will burn about the same amount of actinides per year if that is all you add to the reactor. While a thermal spectrum means you'll be producing more higher actinides (Am, Cm) those will eventually burn off as well.

What does TRU stand for?

-5 points.

D'oooh, caught me in the act...

TRUs = TRansUranics

Everything above uranium such as Np, Pu, Am and Cm with plutonium being the main component. These are the real nasty bits in spent fuel that we'd like to burn off instead of sending to long term storage.

Thanks for the clarification. I thought however that negative temperature coefficients weren't 'fast enough' to mitigate very fast reactivity changes in fast neutron reactors however, and so they rely on doppler broadening as well. Whats the story behind this?



Then you're in new engineering questions as you would be with a liquid chloride reactor though, aren't you, since the salts we have the most experience with aren't the ones PuF3 is the most soluble in?



What are some other uses of the huge neutron surplus from a liquid chloride reactor besides breeding U233, making weapons material, and stabilizing long lived fission products anyways?

Huge surpluses of neutrons don't exist.

One could use some extra to reduce the fission product (FP) processing load. That is one could take twice as long process the entire volume of fuel salt if the neutronic balance can tolerate the neutrons captured by the FP's. (Still nice to remove the FP for safety reasons - it reduces the decay heat load and reduces the radioactivity of the core if an emergency shutdown ever happens).

Likewise, it can eliminate the need for Pa extraction.

So, in general we can use the extra neutrons to reduce the cost of the power plant.

OR to generate new fissile to start more plants faster.

When liquid chloride reactors are estimated to have breeding ratios of 1.7 or higher, I find that a bit hard to believe.

As another nuclear layman I would understand that:-
a. Liquid chloride reactors shall be fast ones due to lower moderating power of heavier nuclei of chlorine.
b. Chlorine shall have to be enriched in Cl-37 due to higher neutron absorption by Cl-35.
so you shall need to have the courage of conviction to use a fast-spectrum liquid reactor.

Chlorine-37 enrichment is important but I think it's a pretty tractable problem. Chlorine-37 is about a quarter of natural chlorine, so it's not exactly rare, and chlorine's pretty abundant. What's nice about the fuel reprocessing ideas I've had for spent fuel is that only the TRU's will need to be treated with Cl-37.

Just a couple extra points to add...

There is enough excess neutrons in the fast chloride reactors to open up some interesting options. First is that it is possible to use natural chlorine and still have a good breeding ratio. In fact, the study called REBUS 3700 managed to just break even on breeding with natural chlorine AND no blanket surrounding the core. They have a huge amount of leakage without a blanket but could still break even. Thus if they added a blanket, they could have a high breeding ratio. The paper I have on this is at home though, it is hard to dig up any direct links online for this work.

Another point and I am not sure if it applies for both natural uranium and Cl-37 is that you get a particularly nasty induced radioactive element, CL 36 I believe.

Finally, keep in mind that any fast spectrum reactor (fluoride or chloride) will take a very large starting fissile load, perhaps 10 to 20 tonnes of Pu (and Np,Am if we use spent fuel). Yes, we have plenty in spent fuel but that is still an awful lot. A more thermal fluoride (with graphite, heavy water, or just letting the salt moderate) can have really low starting loads of around a tonne or even much lower.

With isotopically-enriched Cl-37 we can make that Cl-36 problem go away, along with the generation of 35S from an (n,p) reaction in Cl-35. That sulfur can cause corrosion.