ORNL Offers a Brief Look at MSR/LFTR Economies

A recent ORNL report, Fast Spectrum Molten Salt Reactor Options, offers some insight into the cost lowering potential of MSR nuclear technology. Since Nuclear Green has always had an interest in the cost lowering potential of MSR technology, I intend to review the cost related information included in this report, while in some cases offering a context for that information.

The "Fast Spectrum" report does not offer and cost evaluation in terms of dollar costs, indeed this would not be possible. The report offers an overview of technical options, and no dollar cost evaluation is possible outside the context of a specific design project. The report acknowledges,

A confident assessment of the economic performance of an FS-MSR is not yet possible. Technology, regulatory requirements, and market conditions have changed significantly over the 40 years since the economic assessments accompanying the MSBR; therefore, the cost inferences drawn from the earlier work have such large error bands that they provide little guidance. Additionally, the neutron spectrum of the present evaluation alters the fuel cycle both in and outside the power plant site sufficiently that direct analogies to other reactor concepts are challenging. The most challenging aspect of reporting a cost for an FS-MSR, however, arises from the concept flexibility. A no-heavy-metal reprocessing design variant has a plant layout much different from that of a full-recycle plant intending to directly accept used LWR fuel as its fuel source. Similarly, a plant intending to produce gasoline as its primary product has an entirely different power cycle compared with an electricity generator.

Overall economic tendencies, however, can be estimated by comparing FS-MSR attributes with those of other nuclear power systems. A summary of FS-MSR attributes and their cost implications is provided in Table 2. A primary cost metric for any power plant is its thermal efficiency. FS-MSRs, as high- temperature power plants, are anticipated to have 45–48% thermal efficiencies, a 12–15% efficiency advantage over LWRs. As refueling for an FS-MSR would be performed on-line, the plant availability would be expected to eventually, once maintenance techniques were developed and matured, surpass that for an LWR.

Thus costs are evaluated in relationship to the cost of Light Water Reactor costs. For example, the absence of a fuel fabrication requirement would lower FMSR costs and indeed all MSR costs relative to LWRs

Other FMSR characteristics that would tend to lower capital costs noted by the Fast Spectrum report would include,

* No fuel handling equipment or pool storage facilities
* No irradiated cladding or matrix material in ultimate waste stream
* Large temperature reactivity coefficient
* No cladding- or matrix-based temperature limits in accident scenarios
* Safe shutdown possible through geometry control in accident scenarios
* Higher primary coolant volumetric heat capacity
* Visually transparent, low-pressure, chemically stable coolant

In addition to these cost lowering characteristics, the capacity of all MSRs to operate at a one atmosphere pressure offers a further and important cost lowering potential.

The ORNL Fast Spectrum report also noted characteristics of FMSRs that would lower electrical costs to customers, or increase utility revenue per unit of electricity generated. These include,

* No cladding-based burn up limits
* Higher operating temperature
* Flexible input fuel chemical form
* Flexible input fuel isotopic content

A number of FMSR characteristics that would raise capital costs include,

* No cladding as fission product barrier with a substitute fission product barrier
* Higher operating temperature
* Highly radioactive, fissile-bearing primary coolant
* Potential for safeguards concerns with separated material
* Material corrosion problems

In the case of materials corrosion, it should be noted that there is a low cost work around, if the designer is willing to accept a somewhat lower but still high by LWR operating temperature.

One MSR characteristics offer a mixed cost picture. This was:

* Continuous separation of fission products (and reduction of source term in accident scenarios)

Not only did continuous separation have potential to lower safety related manufacturing and construction costs, but it offered a potential for dramatically lowering regulatory costs. If gaseous and volatile fission products are removed from a reactor as they are produced by nuclear fission, then the motive for most aspects of nuclear regulation is disappears. Thus the cost of nuclear regulation can be lowered. In addition, if separated from the coolant salts, many fission products become salable, either immediately or after a laps of some time. Thus fission product separation can lead to a new revenue stream. Finally although fission product removal devices add to capital costs, their cost can be lowered if Molten Salt Reactors are mass-produced.

It should be noted that the capital cost raising and lowering picture is similar other forms of MSR, with factors such as coolant salt choice, core design including graphite use, and relative neutron speed (thermal, epithermal, and fast), effecting capital costs.

The Fast MSR offers a tool for managing the actinide content of nuclear waste. . In 1991, Uri Gat, and J. R. Engel of ORNL, and C. H. Dodds, of the University of Tennessee, proposed burning fissile fuel from dismantled nuclear weapons in LFTRs, as a means of nuclear deproliferation. That is the process of destroying the raw materials of nuclear weapons.

V. V. Ignatiev, S. A. Konakov, S. A. Subbotine, and R. Y. Zakirov of the Kurchatov Institute in Moscow, and K. Grebenkine proposed the use of Molten Salt Reactors as a means of disposing of nuclear waste. They noted that LFTRs had advantages over Liquid Metal reactors for nuclear waste disposal. The Russian research has lead to the development of the MOSART reactor design. The MOSART is a liquid salt fuel reactor concept intended to burn nuclear waste.
A similar proposal has come from Charles W. Forsberg of ORNL.

The integral Fast Reactor can perform a similar function. and while I have come to appreciate the IFR design, it still has safety problems that would not trouble a fast MSR design. In addition, fast reactors require very large start up charges, in comparison to MSR thermal thorium breeders. The large size of fast start up charges limits their scalability. Thus ten times as many LFTR can be started with the plutonium from nuclear waste, as IFRs or FMSRs. Fast reactors thus are an option for disposing of plutonium from nuclear waste. LFTRs can be started with Reactor grade plutonium (RGP), U-235 or U-233 if it is available. They can be started with a mixture of reactor fuels, or they can be started with all three. The current American stockpile of RGP is gig enough to start enough LFTRs to supply the entire American electrical demand and then some.

Some IFR advocates argue that high breeding ratio IFRs are rapidly scalable because they can produce a very large amount of nuclear fuel. But IFR design research and development has to date largely focused on IFRs capable of burning RGP with breeding ratios similar to those of thermal LFTR breeders. Higher IFR breeding ratios are undoubtedly possible, but they would require much R&D and would never be as safe as FMSRs. Thus RGP can be disposed of by fast reactors, but Lars Jorgensen has established that a fleet of thermal LFTRs can dispose of our entire stock of RGP in under 300 years. Thus if it is viewed as desirable to use the RGP found in "nuclear waste" to start thermal breeder LFTRs, it can be used to start hundreds of LFTRs. Since actinide disposal is the largest single problem associated with the so called nuclear waste issue, the thermal LFTR-RGP start option may well represent the best option for nuclear waste management.

At any rate ORNL FMSR report, offers further support for the contention that MSRs have the potential for lowering nuclear costs. The cost lowering features of the FMSR are all available in high scalable thermal spectrum LFTRs, as well as uranium fueled MSR designs. In addition small MSRs can be built in large numbers in factories. Factory produced small MSR/LFTR modules can be shipped by truck or by train to final assembly sites, or completely assembled in factories and shipped by barge. Not only can they be used to provide electrical power, but they can produce industrial heat, serve as the basis of combined heat and power systems, and even include bottom cycle desalinization. Thus the small MSR may prove to have not only a lower cost than conventional nuclear power plants, but superior versatility.

It is clear then that if breeder scalability and rapid manufacture is desirable, the thermal MSR/LFTR path holds significant advantages over the FMSR or IFR fast breeder approach.

What are the problems with LFTR technology?

What are the problems with MSR/LFTR technology? This turns out to be a hard question to answer. Since there are a large number of LFTR design options, however, it is difficult to identify a set of problems that shared all of the options. Rather we should talk about elective choices, and the problems that a MSR/LFTR designer would face if a certain option were chosen.

Protactinium would seemingly pose a problem for thorium breeding. The Protactinium nucleus is a very big target for neutrons in a LFTR core. Kirk Sorensen discussed the problems posed by Pa-233 and U-233 in MSR blanket salts:

From these cross-sections, you can see that thorium-232 has a moderate cross-section for absorption, but there’s so much of it in the blanket that it does almost all the neutron-absorbing (as we would want).

After absorbing a neutron, the Th-232 becomes Th-233, which has a monster absorption cross-section (almost 200x that of Th-232) but its half-life is so short (22 min) that it isn’t around very long to absorb a neutron.

Once it turns into Pa-233, the absorption cross-section is still over 5 times greater than the Th-232. That is one of the basic reasons why it’s so important to isolate the Pa-233 from the blanket–in order to prevent another neutron absorption. This is a key step that you just can’t do in a solid-core reactor that’s trying to “burn” thorium (and achieve a conversion ratio of > 1.0).

Finally, the Pa-233 decays to U-233 in 27 days. The U-233 has a huge cross-section, mostly for fission (531 barns) but with a lot of absorption (45 barns). Thus, uranium-233 left in the blanket will really want to gobble up blanket neutrons and cause fission. That leads to even more trouble, because that will deposit fission products in the blanket, complicating reprocessing and making the blanket “hot” with radiation from fission products.

All of these factors argue for getting protactinium of out the blanket and letting it decay to U-233 outside of the neutron flux. The U-233 can then be removed by fluorination to UF6 and adding it back to the core salt by reduction to UF4. Continuous refueling of the core means that excess reactivity in the core can be held to almost nothing, an extremely important consideration for safe operation that is very difficult to achieve in a solid-core reactor.

This problem would seem to be compounded in a single fluid LFTR, in which thorium breeding takes place in the same fluid that carries the fissionable nuclear fuel. protactinium is not easy to remove from molten salts. It turns out that it is a lot easier to wait until the Pa-233 is transformed by a gamma particle emission into U-233. There is, however, a proliferation related disadvantage to Protactinium separation in addition to the problem posed by the need to separate Protactinium out of its carrier salts. Dr Buzzo points out,

U-233 is perfectly suitable for use in a nuclear weapon, at least in theory. The thing which makes it difficult is that there would be some U-232 as well. This does not preclude the use in a weapon, but the short halflife of U-232 makes it much more radioactive and therefore difficult to handle. . . . .

it's really the U-232 which is going to make the uranium recovered less suited for weapons use.

However looking at the aspects of protactinium separation, I'm wondering if this could be a hole in the process which would allow for much lower U-232. U-232 is the daughter product of Pa-232 just as U-233 is the daugher of Pa-233. Pa-233 has a half-life of 26.9 days but Pa-232 is only 1.3 days.

This seems as if it could cause a problem. Basically if you separate the protactinium and let it decay for about eleven days, for example, you've gone through eight half-lives of Pa-232 but less one half of a halflife cycle of Pa-233. Thus you still retain about three quarters of the Pa-233 you started out with but the Pa-232 has been diminished to less than half a percent of what you started with. You could do it for even longer before you start to loose a lot of the Pa-233.

Thus, at this point you could do the process over again, removing the uranium and retaining the protactinium and you would have a very high concentration of Pa-233 and very little Pa-232, which is where the U-232 would come from. This is not very difficult and could easily be done with what is available. The result is basically an easy source of weapons grade U-233.

Therefore, according to Dr. Buzzo, it is undesirable to to separate Protactinium from its carrier salts, if you are worried about proliferation. In response to Dr. Buzzo's proliferation related concern, David LeBlanc commented,

Many of us on this site strongly favor the 2 Fluid design of having one salt with the U233 (and maybe a little thorium) and a separate salt for the thorium. The Single Fluid design however has been what most researchers have focused on since the late 1960s.

In a Single Fluid design it is much more difficult to try to skip Pa removal and still break even. The way to lower the neutron losses to Pa is to lower the average neutron flux it experiences (especially thermal neutrons). You can do this by simply having a much larger core or by having excess salt that you cycle in and out of the reactor loop. However, in a Single Fluid design, having more fuel salt means having much more fissile material to start. This is not a deal breaker but a serious impediment nonetheless.

In a 2 Fluid design we can lower losses to Pa down to almost nothing by simply increasing the volume of blanket salt. This means paying for more thorium and carrier salt but thorium is very inexpensive (the true potential cost of mass produced Flibe salt is unfortunately one of the big unknowns). For example, 1960s 2 Fluid designs had about 260 tonnes of thorium in the blanket salt versus about 70 tonnes in the later Single Fluid design.

In another thread, Dr. LeBlanc commented,

in a 2 fluid reactor you can have more blanket salt cycled in and out of your reactor to really lower the loses to Pa. Here are some numbers to give you an idea of losses (remember you can almost double the values since you often lose a second neutron to U234):

Single Fluid design with a 3 day Pa removal time, Pa losses are 0.0017 out of 2.23 neutrons

Single Fluid design WITHOUT Pa removal, Pa losses are 0.05 out of 2.23

2 Fluid design with lots of blanket salt, NO Pa removal (ORNL 1467) 0.0079 out of 2.22 (0.36%)

2 Fluid design with less blanket salt but Pa removal (ORNL 4528) 0.0002 out of 2.22.

The neutron losses for all designs noted by Dr. LeBlanc are acceptable for breeding purposes. Thus a LFTR designer has a number of potions to chose from, in creating a design that best meets breeding goals. As it turns out, according to Dr. LeBlanc, none of the options pose serious barriers to breeding goals, although one option - Single fluid design with no Pa removal - offers the most disadvantages.

There are a number of number of problems associated with core graphite in thermal MSRs. Graphite cores breeders offer huge scalability advantages, because they can be started with a small charge of fissionable material. It takes about 10 times more fissionable material to start a Fast Breeder reactor, than a graphite moderated thermal breeder requires. This makes an enormous difference in the number of reactors that can be started quickly. Well over a year ago I posted a comment on Brave New Climate,

By 2050 if not sooner, we will begin to need breeder technology in order to keep up with world energy demand. The question is which Generation IV technology will have the advantage. I have argued that LFTRs will, because they are far more scaleable, can be manufactured more rapidly, are more flexible, and will be perceived as safer, and less of a proliferation danger. (I am not arguing the last two are the case, I am now satisfied of IFR safety, and proliferation is an anti-nuclear canard,.) Claims of high IFR breeding ratios are not confirmed from IFR design plans. The only IFR designs I was able to locate on the Information Bridge, had a maximum breeding ration of 1 to 1.07, the same as 1970′s ORNL MSBR designs. Statements by IFR advocates indicate no higher breeding ration can be expected in the near term. Since as many as 12 LFTRs of equivalent power output can be started for every IFR, if the IFR has no breeding ratio improvement, it cannot be seen as the most likely LWR replacement. Never-the-less world wide we will see LMFBRs. I believe that the Indians are considering plans to build as many as 300 500 MW LMFBRs, and I fully expect Russia, China and Japan to enter the LMFBR race.

However, it turns out that the reactor grade plutonium from spent light water reactor fuel becomes a chocking point for Generation IV reactors. The world supply of unused reactor fuel now is sufficient to now start a very large number of LFTRs, but not of IFRs. Given that current IFRs designs have no breeding advantage over the LFTR, allotment of RGP to start LFTRs offers some enormous scale advantages. Given that LFTRs are likely to cost less to build, can be built more rapidly, and are likely to have less political and public opposition, it seems to me that IFR advocates are backing the wrong horse in the Generation IV breeder race.

Needless to say, IFR advocates were not pleased by this comment, but they have not been able to show that I was wrong. However, in order to wrap the thermal scalability LFTR advantage, we have to find ways to solve the graphite problems. The two major graphite problems, are

* graphite deterioration in a high neutron flux environment

* And a positive coefficient of reactivity associated with core graphite.

If we want to build a large number of LFTRs quickly then we have to find a workaround. On solution to the graphite deterioration problem is to replace the core graphite every few years. One way to accomplish this is by using graphite pebbles rather than a graphite core structure. The pebbles can be replaced as they reach a point where their deterioration become unacceptable. "Cyril R" points out,

Graphite pebbles have a lot of potential advantages. In one two fluid design, the pebbles are filled with blanket salt. This means every pebble is a barrier, and so barrier maintenance is potentially easier. However, circulating pebbles turns out to be a bit tricky, and with a lot of pebbles all containing liquid blanket salt, broken pebbles seem like a big risk in a true two fluid design. . . .

However, if solid graphite pebbles were to be used in a two fluid design like David's tube in shell, things would be easier. The pebbles wouldn't circulate, but act as fairly static moderator. The simple graphite pebbles would last longer and be easy to replace. Because pebbles have a high void fraction, the traditional MSR graphite density could not be achieved (probably at least half the graphite density).

Lars commented,

Graphite in the blanket would serve to slow the neutrons down. The slower spectrum will make all cross-sections larger so less blanket salt is required to absorb the neutrons. However, the cross-section of the fissile will grow dramatically faster than anything else - which is not good in the blanket. It means that we have to keep the u233/th232 ratio much lower to keep fission in the blanket rare. So one result is that we have to process the blanket faster to keep the u233 concentration down. I'm not sure how big an issue this is - the original 2-fluid ORNL designs were thermal and so they faced this issue and did not identify it as the reason to stop work on the 2-fluid design. But they also generally assumed they could process things at a pretty high rate.

Another effect of graphite in the blanket would be that any neutrons that hit the exterior wall would be slow so they are much easier to absorb and stop and are easier on that wall.

Another effect is that it becomes more cost effective to absorb a higher percentage of the neutrons in the blanket since the thorium in the salt is more effective at absorbing slower neutrons.

One BIG concern is that if there is a big break in the plumbing for the blanket salt you will drain the neutron absorber from the blanket. With the graphite present it will slow down and reflect many neutrons back toward the core. In normal operations the blanket salt will absorb most of them. With the blanket salt drained they will go back to the core. In other words, if you get a dramatic break in the blanket plumbing and drain the blanket salt the reactivity of the core will go up. This can not be allowed. The design would need to somehow guarantee that no matter what the reactivity of the core does not go up in any accident scenario.

David LeBlanc described his position.

your don't want graphite or other good neutron reflecting material in the blanket zone or you end up with reactivity problems if the blanket salt drains or even just gets hotter and less dense.

For the core, using graphite is always a serious option and pebbles certainly have some big advantages but they don`t really help with the core to blanket barrier issue. Even a core with graphite moderator you still need some sort of physical barrier to the separate the fuel and blanket salts (any Two Fluid or 1 and 1/2 Fluid needs barrier material). A graphite core of logs might make things a little easier because we could have a simple metal cladding wrapped around it that would need no structural strength of its own). My google tech talk also shows a method to use individual graphite logs bunched together as the barrier but too complicated to describe here.

So in general, pebbles versus logs is always going to be an interesting trade off of pros and cons. In terms of radiation damage, I still don`t know if anyone has a good idea of how long a pebble would be last. The expansion beyond original size is no longer of structural concern for the core (like it would be for logs) but it the pebble starts to crack due to expansion then we do have a big problem. ORNL seemed to be on the fence regarding this in their early studies with pebbles (which seemed always to be a Plan B that never got too deep a look).

In most cases, comments are or can be referenced back to ORNL MSR research. I could quote more of the graphite pebbles discussion which illuminates a number of problems, but this is enough to suggest that MSRs problems exist, but that solutions and work arounds are available. Each solution or work around may have its cost, so any MSR/LFTR design is going to offer a compromise. The question facing the LFTR designer is, which set of compromises works best given design goals. Because the graphite moderated LFTR is highly scalable even without a high breeding ratio, designing the LFTR to produce just one U-233 atom for every fissionable atom burned. Not only does this decrease proliferation risks, but it allows for more breeding ratio lowering compromises in the LFTR design.

There would be a set of problems for every MSR/LFTR design, but there appear to be an acceptable set of compromises for the problems we have looked at. At least some of the compromises I have reviewed, seem to have secondary benefits that are consistent with probable design goals. In the nearly 40,000 comments of the Energy from Thorium discussion section, no one single killer problem has yet popped up. This most likely means that development of various MSR designs including LFTRs will not involve serious development challenges, and we can be reasonably but not entirely certain that serious problems will not impede MSR/LFTR developmental progress.

Thus it can be asserted with reasonable certainty that the LFTR offers a potential long term solution to human energy needs, that is consistent with a high energy lifestyle, and which will not create the sort of safety, waste, proliferation and capitol cost problems associated with LWR power technology.

The D A Ryan MSR/LFTR critique: Not ready for Prime Time

A blogging engineer named D A Ryan has recently written a critique of Molten Salt Reactor/LFTR/thorium nuclear technology that is decidedly a mixed bag. Most of the comments land off the mark, some seem to wonder away from reality. Many comments in the Energy from Thorium Comment Forum can be taken as critiques of MSR/LFTR technology. Any serious attempt to criticize MSR technology should review the technical discussions on EfT, before making an assessment of molten salt technology. This is quite a chore, but familiarity with the subject is a requirement of serious criticism. While Ryan is familiar with EfT, he has not gone to the trouble of checking out his criticisms with the EfT discussion forum. Ryan then is not a serious critic of nuclear power, but that is not his point. Ryan's conclusion is actually his starting point,

clearly, as regards the current discussion, we cannot run the world on nuclear energy; indeed we’d struggle to meet a tiny portion of global energy needs, for any prolonged period (and I mean a lot less that we currently manage!) with nuclear power, neither generation IV reactors, nor Thorium, nor even Fusion power will help much on this point. Even the most optimistic nuclear energy program we can realistically conceive of still has a substantial energy gap that something else will have to fill. And given our limited fossil fuel supplies (long term at least) that inevitably means alot more renewable energy, which has to take priority over nuclear.

Does Ryan reach this conclusion by sound reasoning, or dies he misrepresent facts, and engage in fallacious arguments? We need go no further than this comment to realize how dubious Ryan's enterprise is,

Another issue is that graphite core. I detailed previously with regard to the HTGR (part 6.4.3) it’s a fire hazard, i.e. Graphite is basically ultra high grade coal! Thus we would need to put the MSR within a containment dome of sorts. Again, as with the HTGR, this dome need not be built to the same exacting standards of a LWR dome as we are merely trying to contain a graphite fire, not an out of control reactor. We would need an effective on plant fire control team and some form of fire detection and suppression system, within the containment dome and all the necessary gear that this entails. I refer you the relevant section of the HTGR anaylsis, but needless to say such an arrangement would involve certain costs.

But is there a fire danger from Graphite? My regular readers might recall thata few months ago I looked at some questions related to the topic of graphite flammability and looked at the question, "Did the Graphite in the Windscale Reactor Burn?" My finding was,

When The UK Nuclear Safety Advisory Committee (NuSAC) meet in 2009 to examine evidence from the Windscale reactor, it found,

• Inspections have shown that there was NOT a graphite fire: damage to graphite, caused by severely overheated fuel assemblies, was localised.

In a third post, titled "Did Chernobyl Graphite Burn," I reviewed sections of a document prepaired for the Nuclear Regulatory Commission by Brookhaven National Laboratory, NUREG/CR-4981 "A Safety Assessment of the Use of Graphite in Nuclear Reactors Licensed by the U.S. NRC." NUREG/CR-4981 was a document intended to answer the question for the NRC. The Brookhaven researchers determined,

Experimental studies on graphite burning have shown that for all the geometries tested which Involved the conditions of small radiation and conduction heat losses, it was not possible to develop self-sustained rapid oxidation for graphite temperatures below about 650 C when the air temperatures were below the graphite temperature. At both high and low flow rates, the graphite was cooled by heat losses to the gas stream even under conditions where other heat loss mechanisms such as radiation and conduction were negligible.

At temperatures above about 650°C, in realistic geometries where radiation is a major heat loss mechanism, graphite will burn only in a limited range of flow rates of air and only when the air temperatures are high. At low flow rates, inadequate ingress of air restricts burning. At high flow rates, the rate of cooling by the flowing gas can exceed the rate of heat produced by oxidation.

Studies have shown that burning will not occur when there is no mechanism to raise the graphite temperature to about 650°C [Schweitzer, 1962a-f]. If the temperature is raised above 650°C, burning will not occur unless a flow pattern is maintained that provides enough air to sustain combustion but not enough to cause cooling. Since the experiments were designed to minimize all heat losses other than those associated with the air flow, 650°C can be considered a lower bound for burning. . . . . in order to have self-sustained rapid graphite oxidation in any of these reactors certain necessary conditions of geometry, temperature, oxygen supply, reaction product removal and favorable heat balance must exist.

The Soviets claimed and American nuclear safety experts like H.J.C Kouts accepted the notion that graphite could burn like charcoal.

The emission of radionuclide continued for about nine days, aided by burning of the graphite. It is estimated that upwards of ten percent of the graphite in the core burned, in a manner similar to the rapid oxidation of charcoal.

We know that Kouts view cannot be correct, nuclear graphite does not burn like charcoal, and the assertion that only 10% of the Chernobyl core graphite burned does not suggest graphite was the major source of the Chernobyl fire. There were, of course other materials in the Chernobyl reactor core that burned hot enough to oxidize nuclear graphite.

Both the NRC and a separate study commissioned by the United States Department of Energy determined that graphite reactor could be operated safely.

The NRC's answer to the original question which I asked at the beginning of this series is ";yes, graphite does burn" but only under a very limited set of conditions. Given this information it is quite possible to design a reactor in which those conditions will never occur.

Ryan references the wikipedia for his claim that "Graphite is basically ultra high grade coal!" What the Wikipedia actually says is,

Graphite may be considered the highest grade of coal, just above anthracite and alternatively called meta-anthracite, although it is not normally used as fuel because it is difficult to ignite.

Nuclear graphite is even more difficult to ignite than natural graphite. The wikipedia article on graphite also states,

during a fire, the graphite expands and chars to resist fire penetration and spread . . . .

What can I say? Here we have clear evidence of extremely shallow research or something worse. Ryan claims that

That big graphite core is a serious worry. As one nuclear physics put it to me “graphite is basically just high grade coal”. Obviously enough, building a nuclear reactor core out of coal doesn’t sound like a sensible idea! Its worth remembering that part of what made Chernobyl the disaster that it was, and why Fukushima is likely to have a much smaller level of fallout (despite 4 reactors involved, one fuelled with MOX against a single reactor at Chernobyl fuelled with only lightly enriched uranium) is because the graphite moderated core at Chernobyl caught fire. It was this fire and the smoke it generated that allowed the radioactive material from the core to spread over such a large area.

So clearly any FMEA process would zero in on this as a major issue that needs tackling. We need to take care in our design to make sure that any potential fire can be safely contained. Obviously this means that any ideas we have about building HTGR’s without containment domes, as some supporters of these reactors suggest we can (and indeed the UK’s AGR’s and Magnox reactors were also built without containment domes), wouldn’t be a good idea. I should note that the containment dome over a HTGR wouldn’t need to be build to the same exacting standards as one over a LWR as our goal is to contain a fire, not a melting down reactor core. This is important as it’s largely been the delays and difficulties in pouring concrete for these cores that is responsible for the messy cost overruns on the various new LWR reactor projects that are ongoing.

Our HTGR’s containment dome would need to be fitted out with some form of automatic fire detection and suppression system, specifically one that can cope with a high temperature graphite fire. As the Windscale power plant fire showed proper planning and equipment would be essential. At Windscale the initially attempt to put out the fire using CO2 failed, as the high temperatures of the fire simply stripped the oxygen from the CO2. The operators finally gambled and poured in water, knowing that this risked setting off an explosion, which fortunately didn’t happen. So clearly we’d need to be better prepared, an inert gas (Nitrogen, Argon or Xenon) or Halon gas should do the trick, if we have enough of it on site. I would note that a number of Halon’s have some potentially nasty environmental issues, such as being known carcinogens and mucking up the ozone layer, so inevitably storing a large quantity of them on site (never mind using them!) would have some environmental implications. Also, we don’t want to be relying, as at Chernobyl, on the local fire crew showing up and doing a Matrosov. Having a dedicated on-site fire crew covering the plant at all times (or nearby covering several plants in a geographical area), as is standard practice for airports, would be sensible. This fire crew, would be specifically trained in dealing with a high temperature graphite fire and be properly equipped to tackle such an event (i.e have working radiation suits! Unlike the situation at Chernobyl). These measures would close off this safety loop hole, but it will come with a cost.

But as we have seen Nuclear graphite is extremely difficult to ignite, and only burns when subject to heat of over 650 C, and then only if supplied with oxygen, and is not in contact with materials that are below 650 C.

Lets list some of Ryan's errors in this account:

1. He failed to note that the difference between the easy ignition of coal, and the ignition difficulties of graphite.

2. He ignores the mention in the wikipedia of graphite fire qualities.

3. His claim that a graphite fire was the primary cause of the Chernobyl radioisotope release is inconsistent with the evidence that 90% of the Chernobyl graphite did not burn, and the research that demonstrates that graphite does not burn in the absence of other heat sources and very hot air.

4. His claim that the Windscale accident included a major graphite fire is based on nothing more than old speculation that is now known to be untrue.

5. The word Chernobyl is used 6 times, Yet the basic premise of the Chernobyl discussion, "the graphite moderated core at Chernobyl caught fire" has been shown to be false. The relatively small amount of core graphite burning that toke place, had to be ignited by an external source and had to be continuously fed by a stream of very hot air.

6. The suggestion that HRGRs pose a fire danger is not supported by evidence.

7. The claim that HTGRs need containment domes is not on the invalid HTGR fire danger argument, thus is not supported by valid reasoning.

8. The argument regarding the supposed problems and costs related to containment dome concrete, is supported by an invalid chain of reasoning, and therefore is invalid itself.

In his account of the alleged graphite fire problems of MSRs, Ryan simply points to his invalid arguments related to the alleged HTGR graphite fire problem and then claims that MSRs will need specially trained fire fighting teams, special equipment, and a containment dome. All of this is hot air, because Ryan has not established reasonable grounds for believing that a MSR graphite fire problem exists.

Ryan criticizes both EfT and Nuclear Green,

Another misconception is that LFTR’s can be air-cooled (here and here) rather than being dependant on the water cooling process we utilise in most other power stations.

The Here and Here refer to one of my posts and a post by Kirk Sorensen.

How does Ryan justify this?

He claims,

Firstly, fire safety, air is an oxidising substance. Fires start all the time at power stations (fossil fuel fired and nuclear ones), especially in the turbine halls and the last thing we want in an emergency is a load of big cooling fans blasting in air and literally fanning the flames! In this scenario we’d face the dilemma between stopping the fans and cutting of the source of cooling (forcing us to SCRAM the reactor to prevent a LOCA scenario) or risk the fire spreading out of control, possibly to the point where it compromises the reactor’s safety. This was of course very similar to the dilemma faced during the Windscale fire, which was air cooled (although in this case directly, rather than indirectly as we currently discussing). And on the subject of Windscale, you will recall what I said earlier about fires and that Graphite core, so we’d be opening a very serious potential safety loophole.

But in nuclear plants, the turbine room is always separated from the reactor by a fire wall. The reference to the Windscle fire is simply confused. In the Windscale reactor air passed directly through the reactor core. There is no proposal to allow cooling air into the reactor core. Thus the Windscale analogy collapses and with it Ryan's first fire argument.

Ryan's second objection has to do with cooling fans

Cooling fans also aren’t terribly reliable, which is why the MSRE was down for several months due to a cooling fan failure. Air based cooling is also very weather dependant, indeed I note that the fans at the MSRE seems to have failed in the summer, when they would have likely been struggling to cope with higher daytime temperatures.

If cooling fans are unreliable, we can use a stack effect to accomplish passive air cooling. The reference to the MSRE cooling fan failure refers to a single incident early in the multiyear MSR test. Once the problems with the cooling fan were identified and corrected, the fan ran for something like 20,000 hours without a problem. The observations about summer related problems is bogus.

Finally we have a third argument

Thirdly, it’s the matter of thermal efficiency. Air based cooling is not very efficient, largely because air has such a low heat capacity compared to water (1.15 against 4.2 J/kg K). A typical COP (Co-efficient of Performance) for fans would be of the order of 2 – 3.7, while you can get 5 – 7.5 with water based cooling. Assuming a COP of 3 (it would be more like 2.5 at the temperatures in question, but bear with me!) and assuming a 1,000 MWth LFTR with a thermal efficiency of 50% (to keep my numbers easy!) = 500 MWe. Our cooling fans, in order to dispose of that 500 MW’s of excess heat, would be consuming 166.67 MW of electricity, dropping our effective plant efficiency down to 33%, barely Rankine cycle levels! This is why we use water in most power stations for cooling.

This argument is preposterous. The B&W mPower reactor is designed to be either air or water cooled. The air cooled mPower is rated at 125 MWe, while the water cooled mPower is reportedly rated at 136 MW, but officially described as a 125 MW reactor.. Thus the active air cooling system at worst consumes 11 MWs more electricity than the water cooling system does. We could probably expect even better performance is we use a stack effect based cooling system.

Ryan adds a further argument:

Also this air based cooling argument strikes me as a bit of a red herring, LFTR fans essentially inventing reasons why their “precious” is better than anything else. With the exception of a few geothermal power stations in arid areas (or hydroelectric plants!), I’m unaware of any major power project that was derailed for lack of cooling water. Either you can use cooling towers (forced draught or natural convection types) and minimise water losses to an acceptable level or simply move the plant next to a ready water source and transmit the power to where it is needed. Many desert countries operate large thermal power stations from around the coasts and several such as Iran, UAE and Libya are even planning to build nuclear stations too. So I fail to see how “air based” cooling offers any real benefits.

In fact water based cooling is usually considered by engineers to offer more thermal efficiency but the fact that Ryan is unaware does not mean consideration is not being given. In fact, as I have indicated, plans for an air cooled mPower reactor are moving forward, and plans for large air cooled solar thermal power plants are being considered in the Southwestern United States.

In addition this passage suggests an attack on Kirk Sorensen and myself for our enthusiasm about Molten Salt nuclear technology. The use of the word “precious” is highly inappropriate, and the use of the quotation marks would untruthfully attribute the word to us. Neither Kirk nor I have used the term “precious” to refer to the LFTR or MSR technology.

Furthermore, Ryan makes other attacks on LFTR/MSR supporters. Ryan describe us

the LFTR fanatics need to come off the Kool-Aid. I’ve gone to great lengths to debunk many of their crazy ideas because such cargo cult science as they are promoting does a great disservice to science, and gets in the way of more realistic and practical proposals. They also serve to confuse the public, and I mean even Wired News appears to have been taken in by this con, which makes the whole job of real scientists pursing real projects, all that harder. As I’ve shown many of the supposed advantages of the LFTR are simply figments of certain bloggers overactive imaginations. The fact that many of the LFTR supporters are Libertarians, individuals not entirely known for their grasp of basic physics, economics or social norms doesn’t help matters

This amounts to little more than a string of attacks on the character of LFTR supporters. The mention of Libertarians is especially silly. LFTR supporters come from the extreme right the extreme left and everywhere in between. The chinese communists who recently decided to invest in LFTR R&D are certainly not Libertarians. Nor am I. We regard MSR technology as promising, and even very promising, but that does not make us cargo cult scientists. The foundational ideas for the LFTR came from Eugene Wigner and Alvin Weinberg, who both were major figures in history of reactor science. LFTR supporters continue to look critically at alternative MSR technologies. The discussion pages of EfT are filled with debate on numerous MSR technology related topics.

In addition to my critique, the blogger uvdiv offers us a

Summary of some of the biggest howlers

* Claims MSRs have "Isotope Separation Plants" which separate 233U and 232U (the trace contaminant)
* Warns of hazardous fission products, such as thorium isotope "T-232" [sic], which supposedly is a disadvantage of thorium-fuelled reactors because of its 14 billion year half-life
* Warns that electrolyzing nuclear fuel salts is energy-intensive
* Warns that heat inputs in fluoride reprocessing are energy-intensive
* Asserts that thorium MSRs are constrained to a lower temperature limit of 1,110 °C, the melting point of pure ThF4. Concludes MSRs must be built entirely from ceramics
* "Obviously, once we exhaust the world’s U-235 stockpiles, LFTR’s and any other Thorium fuelled reactors will cease to function."
* Argues against using molten fuel salt as a working fluid in a gas turbine

"Uvdiv suggest that Ryan holds a misconception that

He has the misconception that 233U/232U isotopic enrichment is necessary for MSR operation. Spends many paragraphs speculating on this imaginary thing, finally concluding it will consume up to 25% of an MSR's electric output:

Uvdid find absurdities in Ryans accounts of fission products. Ryan quite literally describes Th-232 (Ryan calls it T-232) a fission product. Then uvdiv notes a strange error in Ryan's text,

The dendrite problem above demonstrates that the LFTR/LFUR has a relatively narrow thermal window. Its filtering plant will not work if the temperature of the fluid drops much below a certain threshold and the danger of fuel solidification raises the risk of the reactor being damaged. With UF4 the solidification temperature is 1,036 °C and its vapourisation temperature is 1,417 °C. [...] With TF4 our “window” is 1110 – 1,680 °C, but again we can potentially move this by lowering the pressure (or raising it if we want to go the other way…not that we do!). A low vapour pressure also creates a few potential problems in terms of keeping the reactor sealed (air is more likely to leak in if the pressure inside is less than atmospheric…possibly starting a fire!) and maintaining a good flow rate from our pumps.

Uvdiv then comments,

I don't know what confusion of his provoked this nonsense. Maybe he hasn't done his basic research, that all MSR proposals involve solvating actinide fluorides in other fluoride salts -- mixtures of LiF, NaF, BeF2, ZrF4, and/or others -- with the mixture having far lower melting points than actinide fluorides. Or maybe he's under the illusion that individual components of a chemical solution precipitate out at their pure melting points. At any rate, his chain of reasoning starts from this major error and leads to others:

Uvdiv then lists another chain of errors,

With the LFTR however, I doubt you could operate one made out of any Nickel alloy, contrary to everything said on the internet. Bare in mind I’m thinking in terms of a good lengthy service life with a sensible factor of safety, not a flimsy test reactor in a lab (with a 100 mile exclusion zone!).

[...]

Thus the pressure vessel of any LFTR would likely have to formed out of Ceramics (very expensive and difficult to form, especially given how critical getting an air tight seal is given the graphite core) and key internal components out of Refractory metals, as would be the case for certain high temperature parts of any ISP (in both the LFTR and LFUR cases) given talk of operating temperatures in the range of 1600 °C.

[...]

So my instinct from a materials science point of view would be to drop the LFTR idea altogether and focus instead on a LFUR. While this isn’t able to use the Thorium cycle, the point was raised earlier about how the Thorium cycle isn’t all its cracked up to be. Its going to be a lot easier to build a LFUR than a LFTR, cheaper (relatively speaking) and likely safer too. Of course it does come at the disadvantage of a slightly awkward acronym! but overall that would be my focus of attention.

We see once more Ryan's tendency to compound errors.

Uvdiv points to one Ryan error that did make me laugh out loud,

Another misconception is that a LFTR or LFUR can operate on an open cycle with a gas turbine. While true, it could be run this way, there are a host of practical reasons not to do it. Not least of them the fact that our turbine would have to be designed to withstand having a mixture of molten salt and fluorided fuel passed through it at very high temperatures. This would be tricky to say the least, likely requiring the use of those super expensive refractory metals, and while using such materials to make the odd turbine blade is one thing, an entire turbine casing is an entirely different matter. It would likely cost much more than the reactor itself!

Ryan confuses open cycle hear powered gas turbines, with fluoride salt turbines which indeed would be a difficult matter.

At this point I think we have enough. Further discussion and debate can be found here, and here.

I originally intended to write Good, Bad and Ugly critique of Ryan's essay on MSR/LFTR technology, but even the good is so tainted with the bad and the ugly that I find this impossible to do.

Third Thorium Energy Alliance marks rapid progress toward fulfillment of dreams

My health, although still not as good as I would like has been improving since my hospitalization last December. I am, however, not in good enough health to travel. This is unfortunate because I would very much have liked to attend the Third Thorium Energy AllianceConference in Washington, D.C., earlier this month. Energy from Thorium has a brief account of the Conference in Energy from Thorium. In addition the Thorium Energy Alliance has posted Power Point Presentations from the Conference on its web page. Judging from the presentations it is probably safe to say that Molten Salt Reactor technology has entered the age of entrepreneurs.

Presentations by Kirk Sorensen, DavidLe Blanc and Charles S Holden indicated that they were either currently involved in entrepreneurial activities or were seeking entrepreneurial opportunities. Kirk has left Teledyne Brown to found a company, Flibe Energy, the purpose of Flibe is to product LFTRs and perhaps uranium fueled MSRs. The relationship between Teledyne Brown and Flibe is not clear, but the money to pay Kirk's salary has to come from somewhere. The Flibe prospectus indicates that the company founders envisage going after such markets as isolated communities, and medical isotopes, as well as stable fission product sales.

Charles "Rusty" Holden probably wants to go after some of the Same Markets Kirk is targeting. Holden's company, Thorenco LLC, is planning to build a 40 MW MSR. The reactor is designed to produce about 15 MWe at maximum. Holden intends to come out of the starting gate with a full LFTR. The reactor is a pool type reactor which involves a large pool of molten coolant salts acting as a thermal sink. I am not a big fan of pool type reactors, although they are safe. This is safety at a cost. The coolant pool will contain 93,200 Liters of coolant salt which will weigh 450 tons. The function of the pool is far from clear since the reactor design includes a dump tank.

The reactor core contains no graphite, Moderation will be by Beryllium in the form of BeF2 in the salts. This is a two fluid reactor with blanket salts doubling as coolants. The core structure uses metal (no doubt Hastelloy N) tubing. With the tubes containing fuel salts surrounded by outer tubes containing coolant salts. Since there is no core core graphite, neutron speed will be relatively bast, likely falling in the Epithermal range. The core is surrounded by a thorium reflector, neutron absorption in the reflector converts some thorium into U-233. The reactor will be a converter, and will require 1600 kilograms of U-233 fissile load, which is an enormous amount given the modest amount of U-233 14 kgs per year, which the reactor will burn,

At this point I will stop, and pronounce Holden's reactor DOA. Too much material goes into it, and too little electricity will come out. Fundamental questions are left unanswered, for example startup. 1600 kgs of U-233 is probably more U-233 than exists in the whole world right now. Where is the U-233 going to come from? There are quite a few more problems and questions. As David LeBlanc noted in his Conference talk, "Softer Spectrum" means "much smaller fissile start up." David is still on the outside looking for an opportunity. In his Conference talk, David noted,

␣ Corporate interest will always be difficult to attract
␣ No lucrative fuel fabrication contracts
␣ Min 15 year return on investment a tough sell to shareholders (no matter how big the return may be)
␣ Existing nuclear players have their choices in place

Money is still the hard part, at least for now. Despite this enormous progress is being made by LFTR/MSR advocates. A month ago, Kirk Sorensen marked the fifth anniversary of Energy from Thorium. At that time a handful of people knew what thorium was. Even fewer knew about Molten Salt Reactors. Kirk set out to educate people using social media tools, and others followed his lead. What Kirk has managed to do is to start a bottom up social movement.

Where are we headed? Japanese Researcher Takashi Kamei of Ritsumeikan University, Kyoto, Japan offered some answers in the wake of the great Japanese earthquake-tsunami of 2011, not to mention the crisis related to Fukushima reactors. Takashi pictures MSRs beginning a rapid increase around 2025and with the number of LFTRs growing more slowly before 2050. He also suggests a growing number of LFTRs after 2035. Takashi estimates a total MSR output of 258 GWe to 317 GWe by 2050. My view is that much more is possible. The future will belong to the dreamers.

The Molten Salt Reactor Family: Converters and Breeders

All reactors are capable of not only using nuclear fuel, but also of producing new nuclear fuel. This characteristic sets nuclear power apart from other energy sources. Nuclear fuel is fissionable, that is its atoms break apart after absorbing neutrons. Fissionable atoms are described as fissile. When a fissile atom fissions, it divides into two smaller atoms, and two or three neutrons. The neurons keep the nuclear process going. Atomic weapons are designed so that the neutrons will encounter more fissile atoms, which in turn split, releasing more neutrons. As long as there is only more fissile material for the neutrons to encounter, the cjain reaction will grow, until heat and energy produced by nuclear fission blows the device apart, ending the chain reaction.

In a reactor the chain reaction is much more control ed. The presence of atoms that can be converted into nuclear fuel will greatly alter the nuclear process found in atomic bombs. The atoms that can be converted into nuclear fuel are described firtile. When they absorb neutrons, firtile atoms undergo a process of nuclear transformation that turns them into fissionable atoms. These processes take time, and thus the conversion process is much slower than the chain reactions in atomic weapons.

Natural uranium contains two isotopes or types of atoms. U-235 and U-238. U-235 is fissile, but it is only 0.7% of all natural uranium. U-238, the other common uranium isotope makes up 99.3% of most uranium ores. Inder most circumstances the percentage of U-235 in uranium is too small to produce a chain reaction. There are, however a few exceptional circumstances that will allow natural uranium to produce a chain reaction. They involve the presence of one of two moderators. Moderators, at materials that slow neutrons down. Slow neutrons are more likely to be absorbed by U-235 than fast neureons, so much so in fact, that natural uranium, if moderated by pure graphite or heavy water, can be brought to a chain reaction.

U-235 has a love affair with slow neutrons, so the presence of a moderator heats things up, so to speak, as more and more neutrons get absorbed by U-235. But even with moderators some neutrons get absorbed by U-238, which then undergoes a conversion to Plutonium-239. Pu-239 is also fissile, so it becomes new nuclear fuel. But in Light Water Reactors plutonium is not the best nuclear fuel. Pu-239 fissions less often at thermal neutron speeds, and on average produces less than two neutrons per neutron absorption. In a thermal reactor Plutonium will only breed at a less than one to one ratio. U-235 produces just over two neutrons per neutron absorption, not enough to sustain breeding. in epithermal Light Water Reactors the neutron performance of both U-235 and Pu-239 are worse. Thus uranium fuel cycle, Light Water Reactors cannot be breeders.

There are two nuclear fuel cycle conversion processes involving thorium-232 and U-238:

U-238 (non-fissile) + n -> U-239 -> Np-239 -> Pu-239 (fissile)
Th-232 (non-fissile) + n -> Th-233 -> Pa-233 -> U-233 (fissile)

TU of Delft, the Netherlands, offers an explanation,

It is important to know how many new fissile nuclei can be formed by conversion for every fissile nucleus consumed. The ratio of (# of new fissile nuclei / # of consumed fissile nuclei) is known as the conversion ratio. This ratio can be estimated as follows:

For every fissile nucleus consumed, X new neutrons are released
For a stable chain reaction, one neutron is needed to sustain the reaction: X must be larger than 1
To have 1 new converted nucleus for every fissioned nucleus, one neutron is needed: X must be larger than 2
Neutrons will leak from the reactor, so X must be appreciably larger than 2 to make a practical reactor with a conversion ratio > 1.

The value of X is highly dependent on the energy of the incident neutrons, as shown in the figure, and X grows rapidly for high-energy interactions.

The following figure traces the effect of Neutron speed on neutron propagation for U-233, U-235 and Pu-239. Because some Neutrons are invariably lost in a reactor core, the breeding ratio will always be lower than eta.

Explanation: "In this figure the Greek letter eta is the symbol for X. Pu-239 has the highest value of X in the high energy region and thus Pu-239 is the best breeding fuel available. At thermal energies X=2.44 for U-235, which is enough to get a conversion ratio of around 0.5: for every U-235 nucleus consumed, 0.5 U-238 nucleus is converted to Pu-239. At higher energies X=3 or higher, providing the possibility of getting a conversion ratio of >1.

It should be noted however that when thorium is the fertile isotope thermal neutrons can produce a conversion ratio of > 1. There are very significant implications for this. As neutron speed declines the amount of fissionable material required to maintain a chain reaction declines as well. A thermal reactor can maintain a chain reaction with a little as 1/10 as much U-233 as a fast reactor. That means with a given amount of U-233, you can start 10 thermal breeders for every fast breeder you can start. The implications of this are enormous if you want to start a lot of breeder reactors quickly. Thorium Cycle Molten Salt Reactors (LFTRs) are capable of true breeding with a conversion ration of greater than one. Because it is possible to start so many more thermal reactors with a given amount of fissionable material, it is quite clear that fueling a very large number of LFTRs is possible. Over 40 years ago, Alvin Weinberg explained:

WHY DEVELOP MOLTEN-SALT BREEDERS?

Nuclear power, based on light-water-moderated converter reactors, seems to be an assured commercial success. This circumstance has placed upon the Atomic Energy Commission the burden of forestalling any serious rise in the cost of nuclear power once our country has been fully committed to this source of energy. It is for this reason that the development of an economical breeder, at one time viewed as a long-range goal, has emerged as the central task of the atomic energy enterprise. Moreover, as our country commits itself more and more heavily to nuclear power, the stake in developing the breeder rises—breeder development simply must not fail. All plausible paths to a successful breeder must therefore be examined carefully.

To be successful a breeder must meet three requirements. First, the breeder must be technically feasible. Second, the cost of power from the breeder must be low; and third, the breeder should utilize fuel so efficiently that a full-fledged-energy economy based on the breeder could be established without using high-cost ores. The molten-salt breeder appears to meet these criteria as well as, and in some respects better than, any other reactor system. Moreover, since the technology of molten-salt breeders hardly overlaps the technology of the solid-fueled fast reactor, its development provides the world with an alternate path to long-term cheap nuclear energy that is not affected by any obstacles that may crop up in the development of the fast breeder.

The molten-salt breeder, though seeming to be a by-way in reactor development, in fact represents the culmination of more than 17 years of research and development. The incentive to develop a reactor based on fluid fuels has been strong ever since the early days of the Metallurgical Laboratory. In 1958 the most prominent fluid-fuel projects were the liquid bismuth reactor, the aqueous homogeneous reactor, and the molten-salt reactor. In 1959 the AEC assembled a task force to evaluate the three concepts. The principal conclusion of their report was that the "molten-salt reactor has the highest probability of achieving technical feasibility."

This verdict of the 1959 task force appears to be confirmed by the operation of the Molten-Salt Reactor Experiment. To those who have followed the molten-salt project closely, this success is hardly surprising. The essential technical feasibility of the molten-salt system is based on certain thermodynamic realities first pointed out by the late R.C. Briant, who directed the ANP project at ORNL. Briant pointed out that molten fluorides are thermodynamically stable against reduction by nickel-based structural materials; that, being ionic, they should suffer no radiation damage in the liquid state; and that, having low vapor pressure and being relatively inert in contact with air, reactors based on them should be safe. The experience at ORNL with molten salts during the intervening years has confirmed Briant's chemical intuition. Though some technical uncertainties remain, particularly those connected with the graphite moderator, the path to a successful molten-salt breeder appears to be well defined.

We estimate that a 1000 MWe molten-salt breeder should cost $115 per kilowatt (electric) and that the fuel cycle cost ought to be in the range of 0.3 to 0.4 mill/kWh. The overall cost of power from a privately owned, 1000-MWe Molten-Salt Breeder Reactor should come to around 2.6 mills/kWh. In contrast to the fast-breeder, the extremely low cost of the MSBR fuel cycle hardly depends upon sale of byproduct fissile material. Rather, it depends upon certain advances in the chemical processing of molten fluoride salts that have been demonstrated either in pilot plants or laboratories: fluoride volatility to recover uranium, vacuum distillation to rid the salt of fission products, and for highest performance, but with somewhat less assurance, removal of protactinium by liquid-liquid extraction or absorption.

The molten-salt breeder, operating in the thermal Th-233U cycle, is characterized by a low breeding ratio: the maximum breeding ratio consistent with low fuel-cycle costs is estimated to be about 1.07. This low breeding ratio is compensated by the low specific inventory* of the MSBR. Whereas the specific inventory of the fast reactor ranges between 2.5 to 5 kg/MWe the specific inventory of the molten-salt breeder ranges between 0.4 to 1.0 kg/MWe. The estimated fuel doubling time for the MSBR therefore falls in the range of 8 to 50 years. This is comparable to estimates of doubling times of 7 to 30 years given in fast-breeder reactor design studies.

From the point of view of long-term conservation of resources, low specific inventory in itself confers an advantage upon the thermal breeder. If the amount of nuclear power grows linearly, the doubling time and the specific inventory enter symmetrically in determining the maximum amount of raw material that must be mined in order to inventory the whole nuclear system. Thus, low specific inventory is an essential criterion of merit for a breeder, and the detailed comparisons in the next section show that a good thermal breeder with low specific inventory could, in spite of its low breeding gain, make better use of our nuclear resources than a good fast breeder with high specific inventory and high breeding gain.

The molten salt approach to a breeder promises to satisfy the three criteria of technical feasibility, very low power cost, and good fuel utilization. Its development as a uniquely promising competitor to the fast breeder is, we believe, in the national interest.

Needless to say, Weinberg's cost estimates are not current, still there are plausible grounds for believing that both Molten Salt Converters and LFTRs will be less expensive than conventional reactors. In addition Molten Salt Reactors offer greater safety, a very significant reduction of the problems associated with disposal of used nuclear fuel, a reduced possibility of nuclear proliferation, and a probably lower energy costs. Molten Salt reactors offer a route to supplying peak demand electricity from post-carbon energy sources. They offer a potential for producing heat for industrial processes, and backing up renewable generated electricity.

Kirk Sorensen delivers a TED Lecture on the LFTR

The "Nuclear Energy Experts" and the Canadian Media

Norm Rubin is an expert on nuclear energy. I know this because the Mainstream Canadian media has made this assertion several times, the latest by Antonia Zerbisias, in the Canadian National Newspaper, the Toronto Stat. Zerbisias reports that Norm Rubin is

Energy Probe’s director of nuclear research and senior policy analyst.

But what does that mean? First Energy Probe is just a web page is a Canadian anti-nuclear franchise. Secondly the Energy Probe Research Foundation states that,

Norm Rubin is a highly sought-after public speaker and has more than 500 public speaking engagements under his belt. He has been with EPRF for more than 20 years, working as a researcher in the nuclear field, and is now one of Canada's leading critics of the nuclear industry. He has appeared at many hearings and court proceedings as an expert witness and has written extensively on the nuclear industry. Mr. Rubin has also made a number of appearances on radio and television shows.

Rubin is quoted by the The Star as saying,

“Thorium doesn’t eliminate the problems,” . . . “If the nuclear industry’s problem was affording uranium, then switching to thorium might solve their problem. But that’s not their problem. The fuel cost in today’s reactors is a tiny fraction of the total cost. That’s not what is giving the Ontario government sticker shock about the next two reactors at Darlington. They’re solving a non-problem by substituting a cheaper fuel for uranium. Unless they solve the big problems, they’ve got a curiosity there instead of a practical solution to anybody’s problems.”

Note that Rubin does not say what "the problem" is, or what he means by "the big problems." Rubin talks about practicle solutions, but how do we know that he has the slightest idea what he is talking about?

When I tried Googling Norm Rubin I found that he is listed as a staffer by the Energy Prob, but has only two posts to the Energy Probe pages to his credit. Two posts is not very much for someone who has carried the fancy title of Research Director for many years. Neither of Rubin's posts relate to thorium. His staff duties for the Energy Probe are not specified. He has no papers or essays on thorium or the LFTR listed on Google. The Energy Probe has not published any essays or research papers on thorium or the LFTR. He is not listed among the 950+ participants on Thorium/LFTR related discussions on Energy From Thorium. So who is this guy? A Google search has uncovered the fact that he once testified before the Ontario Energy Board's about the Integrated Power System Plan (IPSP), and reportedly he is so afraid of radiation that he has never been inside a nuclear power plant. Where is the work product that would serve as evidence of his expertise? Yet Canadian Media accepts him as an expert on Nuclear power. So much the worse for the quality of the Canadian Media.

Ok so Rubin is a talker, not a researcher. He is an anti-nuclear brand spokesperson. His title makes what he has to say seem more credible, but does not establish that there is any research at all behind Rubin's public pronouncements. The Canadian media accords Rubin expert status, because they wish to accord some balance to their stories which suggest that there is a valid scientific controversy about the safety and usefulness of nuclear power. But are we talking here about good journalism, or are we talking about journalists collaboration in a hoax that is contrary to the public interest?

How the LFTR would have survived the Japanese Earthquake/Tsunami

Future nuclear safety tests should include capacity to survive the events which lead to the Fukushima Dai-ichi nuclear plant crisis. The real survival test would require the same flawed backup generator system that was destroyed by the tsunami that struck the plant. Of course, we are not going to talk about every conceivable tsunami. For example, it is highly likely that the island of Oahu will rupture someday, dropping a large part of it into the Pacific Ocean, creating a huge tsunami. That would be a megatsunami. Since megatsunami can be up to 1000 meters high, we need not worry about a coastal reactor surviving all tsunami. No one will be around in the vicinity of a 1000 meter megatsunami to worry about a subsequent nuclear accident.

This leads us to the question of acceptable nuclear risks, a fit topic for another post. At any rate we are looking, right now, at how well the Liquid Fluoride Thorium Reactor (LFTR) or any otherMolten Salt Reactor (MSR) would have survive the natural disaster that overwhelmed the Fukushima Dai-ichi nuclear plant.

First it should be noted that Molten Salt Reactors do not require water cooling at all. Hence the loss of the emergency cooling generator system would have not been a serious problem by itself. Lets explore an accident scenarios that is at worst remote possibility, a key pipe ruptures triggering a loss of coolant accident. Say the entire content of the core coolant system - a liquid salt mixture - drains on to the floor of the reactor chamber. and forms a puddle. Since the nuclear fuel is dissolved in the coolant salts it will be deposited into the puddle. Now the interesting thing about a puddle is that its geometry is not at all conducive to a chain reaction, so the loss of coolant in turn triggers a withdrawal of nuclear fuel from the core, which in turn triggers a termination of the chain reaction, so the reactor automatically stops functioning.

Now the puddle, even though we can expect it to be short lived, might be a problem because radioactive gaseous fission products, dissolved in the fuel salt, are likely to bubble out along with volatile fission products.

Our puddle will not stay on the floor - it quickly drains into a pipe leading into a set of emergency coolant tanks, which are intended to hold the fuel mixture until the reactor can be repaired and restarted. The geometry of the tanks would be intended to prevent a chain reaction from occurring, but the fuel salt would include some radioactive fission products capable of generating the sort of post reactor shutdown heat that created so many problems in the Japanese reactors. Is there any way to insure that the liquid salt in the emergency coolant tanks does not start boiling and releasing a lot of nasty stuff? There turnout to be several passive solutions to this problem.

One: Draw air over a simple heat exchange system designed to dissipate some of the heat in the emergency tanks. Not too much heat, since we want the coolant liquid to remain hot and well, liquid. The heated air can be directed to a chimney, through which it flows into the atmosphere. The system thus requires no power, no controls and no operators. It works automatically, relying on the laws of nature to function.

Two: Rely on a thermal sink, most likely a molten or solid salt, such as the salt that carries the fuel. The inner emergency coolant tanks could be surrounded by an outer thermal sink tank. Or the coolant tank could be shaped like a donut, with inner and outer thermal sink tanks removing heat. A large enough thermal sink would probably be sufficient to dissipate heat without requiring any further heat transfer system. There is a cost for a large thermal sink salt tank system. The Integral Fast Reactor (IFR), for example, relies on this thermal inertia to prevent the reactor from overheating during an emergency shutdown, but this approach is likely to add considerably to IFR costs. However, a thermal salt vault approach combining energy storage with the safety function of decay hear dispassion, would probably would add minimally to reactor cost, while offering a source of reserve electricity.

Three: One proactive LFTR safety approach would be to remove some or even most of the fission products from the reactor. Removal of radioactive gases would be very desirable for a number of reasons. For example. as Uri Gat, and H.L. Dodds pointed out,

The source term, which is the inventory of radioisotopes in the reactor available for dispersion to the environment, contributes two-fold to an accident. The source term is the measure of the radiation which needs to be contained from reaching any sensitive location or target. The energy contained in the source term also provides the driving force for the dispersion of the source term as it is also a measure of the after heat, or the energy, to damage a reactor in the event of heat-removal failure or loss-of-coolant accident (LOCA). For an MSR, as for any fluid fuel reactor, on-line fuel processing can be applied. The on-line processing, at the least, removes the gaseous and volatile part of the source term. This part is the most likely to be dispersed when there is a breach of containment. Fuel processing also reduces the inventory of longer and long-lived isotopes as their accumulation is time dependent. The MSRs processing can be adjusted to have a small source term. The safety advantages of this small source term are many fold: The driving force for dispersion is reduced; the gaseous and volatile components, which are the most likely to disperse, are essentially all but eliminated; the long half-life isotopes (elements) are reduced such that the long-term effect of even the most unlikely accident is not severe; and, the short-lived isotopes require a proportionately short-term protection time till they decay. Thus, even a hypothetical severe accident is ameliorated a priori.

A properly designed processing facility quickly removes the separated radioisotopes from the purview of the reactor. This makes them totally unavailable to the reactor source term even under the most extreme hypothesized circumstances.

In addition to removing fission product gases, and volatile fission products, removal of nobel metals would be highly desirable from an operational point of view. Gat and Dodds state,

The source term, which is the inventory of radioisotopes in the reactor available for dispersion to the environment, contributes two-fold to an accident. The source term is the measure of the radiation which needs to be contained from reaching any sensitive location or target. The energy contained in the source term also provides the driving force for the dispersion of the source term as it is also a measure of the after heat, or the energy, to damage a reactor in the event of heat-removal failure or loss-of-coolant accident (LOCA). For an MSR, as for any fluid fuel reactor, on-line fuel processing can be applied. The on-line processing, at the least, removes the gaseous and volatile part of the source term. This part is the most likely to be dispersed when there is a breach of containment. Fuel processing also reduces the inventory of longer and long-lived isotopes as their accumulation is time dependent. The MSRs processing can be adjusted to have a small source term. The safety advantages of this small source term are many fold: The driving force for dispersion is reduced; the gaseous and volatile components, which are the most likely to disperse, are essentially all but eliminated; the long half-life isotopes (elements) are reduced such that the long-term effect of even the most unlikely accident is not severe; and, the short-lived isotopes require a proportionately short-term protection time till they decay. Thus, even a hypothetical severe accident is ameliorated a priori.

A properly designed processing facility quickly removes the separated radioisotopes from the purview of the reactor. This makes them totally unavailable to the reactor source term even under the most extreme hypothesized circumstances.

Removing all radioisotopes from a Molten Salt Reactor removes the protection that those isotopes afford. As long as the salt contains radioactive fission products, it will be far too dangerous to handle for nefarious purposes, such as the eternal bogeyman of nuclear proliferation. Salt processing can be conducted by automatic equipment inside the reactor core hot cell. The heat and radiation inside the hot cell would prevent anyone having near real-time access to a MSR.

One way of managing a reactor situation that is likely to lead to an accident, is to design a built in failure point, analogous to an electrical fuse or other weak link, which will fail before anything else. One such deliberate failure point in the MSR is the freeze valve; if a LFTR or other MSR begins to overheat, the freeze valve is designed to melt as Gat and Dodds explained,

The MSR can utilize freeze valves in critical locations or where desired. Freeze valves can be ordinary sections of pipe which are exposed to a cooling stream of environmental gas to the extent that it creates a frozen plug that blocks the flow and acts as a valve. Where such a valve has a safety function, as in draining the fuel to the storage tanks, it is prudent to design it such that the required flow is
gravity-driven. The frozen valve itself can be designed such that when the salt rises above a certain predetermined temperature the heat overrides the cooling, melts the frozen plug and opens the valve. Such an arrangement is passive, inherent and non-tamperable (PINT-safe).

Furthermore, the properly sized external cooling of the freeze valve cooling drive, such as an electric driven fan, will cease with any failure of the power and release the valve to melt and perform its safety function. This mode of operation is again PINT-safe.

Once the freeze melts, a MSR will simulate a total loss of coolant accident, with fuel/coolant salts dumped into a tank or tanks that are designed with a criticality inhibiting geometry. In his paper 2006 paper, Molten-Salt-Reactor Technology Gaps (Proceedings of ICAPP ‘06, Reno, NV USA, June 4–8, 2006, Paper 6295), MIT nuclear scientist Charles Forsberg stated,

Under emergency conditions, the liquid fuel is drained to passively cooled critically safe dump tanks. By the use of freeze valves (cooled sections of piping) and other techniques, this safety system can be passively initiated upon overheating of the coolant salt. MSRs operate at steady-state conditions, with no change in the nuclear reactivity of the fuel as a function of time. Last, the option exists to remove fission products online and then solidify those radionuclides into a stable waste form. This minimizes the radioactive inventory (accident source term) in the reactor core and potential accident consequences.

In addition to the very useful freeze plug, the capacity of molten salt to freeze at a still relatively high temperature is directly responsible for another MSR, its automatic leak control. As hot salt leaks from reactor piping, it begins to cool on contact with hot cooler air, and as it cools, it freezes, blocking further escape of coolant salt. Reportedly this mechanism is very effective in stopping leaks if they occur.

At this point I have established the case which I have sought to prove, the tsunami that destroyed the back up generators of the Fukushima Dai-ichi nuclear plant, would have left the safety systems of the LFTR in tact.

This is not the only MSR advantage. According to a group pf French nuclear scientist from the University of Grenoble, the MSR does not simply offer a high probability of safety, it offers an

excellent level of deterministic safety,

That is, safety depending only on the laws of nature and thus safety that is beyond doubt. The MSR is uniquely stable. It can be designed to safely operate without any human intervention, until such time as repairs or parts replacement is required. Thus, MSRs do not require on site operators, and indeed the stability and load following ability of the MSR are such that operators would have quite literally nothing to do. The absence of human operators would probably add to MSR safety, rather than inhibit it. that is safety that depends on the laws of nature and thus safety that is beyond doubt.

It is probably true that the safety systems of the AP-1000 and the ESBWR would have survived the Dai-ichi tsunami. But compared to the simple safety features of MSRs, the safety systems of even advanced LWRs are complex and expensive. Molten salt nuclear technology offers many potential cost saving advantages, and if all of them are employed, MSR costs could be substantially lower than the costs of LWRs. Part of the MSR advantage is higher safety at lower cost.

The Molten-Salt Reactor Demonstration Reactor

Ed Bettis, who was one of the inventors of the Molten Salt Reactor concept, and was a pioneering champion, went on to become lead an ORNL design group during the 1960's and 1970's. Bettis designed both one and two fluid reactors, including a pioneering ORNL two fluid 250 kWe Modular MSR. Bettis also realized that a practical MSR, based on MSRE tested technology and in ORNL-TM-3832 (DESIGN STUDIES OF A MOLTEN-SALT REACTOR DEMONSTRATION PLANT) offered a practical design for such a reactor. The purpose of the Demonstration Reactor project was to

represents a molten-salt reactor plant which is feasible to build, will produce a significant amount of electrical power, and will be a major step toward a useful family of breeder reactors.

The abstract of ORNL-TM-3832 reads

The MSDR, a 350-MW(e) Molten-Salt Reactor Demonstration Reactor, is based on technology much of which was demonstrated by the MSRE. The cylindrical vessel (26 ft diam by 26 ft high) houses a matrix of graphite slabs forming salt passages having 8 volume fraction in the core of 10%. . . In the secondary exchanger, heat is transferred to a stream of Hitec salt (in at 800 F, out a t 1000F). The Hitec oxidizes tritium to tritiated water which is removed and disposed of. The Hitec generates steam at 9OO F, 2400 psi in a boiler, super- heater, and reheater. Electricity is produced a t an overall efficiency of 36.6%. Soluble fission products are removed by discarding the carrier salt every 8 years after recovery of the, uranium by fluorination. Volatile fission products are removed by sparging the fuel salt with helium bubbles in the reactor primary system. The fuel cycle cost was estimated t o 0.7 mill/kWhr for inventory, 0.3 mill/kWhr for replacement, and 0.1 mill/kWhr for processing, giving a total of 1.1 mills/kWhr.

Although ORNL's primary reactor development focus during the early 1970's was on the development of a Molten Salt Breeder Reactor, Bettis suggested:

An alternative approach to the development of a commercial MSBR has also evoked interest. This approach emphasizes more rapid attain- ment of commercial size but more gradual attainment of high performance. The step beyond the MSRE is construction of a 300-MW(e) Molten-Salt Demonstration Reactor(MSDR). The purpose of the MSDR would be to demonstrate the molten-salt reactor concept on a semi-commercial scale while requiring little development of basic technology beyond that demonstrated in the MSRE.

We see from the abstract that many improvements would be possible with the 1972 design. Compared to the MSBR, Bettis and his associates proposed,

First, the MSDR has only such chemical processing as was demonstrated in the
MSRE and has no provision for removing fission product poisons on a short time cycle. Thisresults in a much less complicated chemical processing plant, although it means that the reactor has a breeding ratio less than one and i s therefore a converter. The second major simplification i s that the power density was made low enough for the graphite core to last the 30-year design lifetime of the plant, thus simplifying the reactor vessel and eliminating the equipment for replacing the core.

At the time ORNL MSBR plans called for the periodic removal and replacement of the MSBR Graphite core, as a solution to the problem of core swelling caused by neutron bombardment. This involved design complexities, and so Bettis proposed an alternate scheme to deal with graphite swelling, a scheme that involved core enlargement.

The ORNL-TM-3832 design although interesting is flawed. The MSDR designers, in an effort to solve core a graphite problem increased the amount of core graphite, this in turn increased the size of the core. But a large core increases reactor construction costs. From a cost viewpoint, it is probably better to replace a small core every few years, than to build a very large core, that will last for 30 years.

While ORNL-TM-3832 represents a serious attempt to simplify MSR design, it hardly represents the last word in MSR simplification. While we may appreciate the ingenuity of Bettis' 1972 design, a revolutionary innovation in core design by Dr. David LeBlanc has greatly simplified MSR core concepts. If the MSDR was altered by substituting Dr. LeBlanc's two tube core for the original core design the entire reactor design would require great alteration. The LeBlanc tube core would almost certainly lower MSR costs, compared to all ORNL core designs of the 1960's and 1970's.

Other problematic features of the Demonstration Reactor involved the use of LiF-BeF2 salts. Since the purpose is not producing nuclear fuel in the breeding range, other salts might carry significant advantages including lower costs, and the elimination of the tritium problem. By switching to another salt combination the tritium problem associated with LiF-BeF2 salts. The MSDR included a third heat transfer loop as part of its tritium control system, and that loop decreased thermal efficiency, increased reactor complexity and costs. If the goal of MSR design is breeding, LiF-BeF2 are the preferred carrier salts, but when breeding ceases to be the objective, then the possibility of using other salts comes into play.

The MSDR was designed to generate power through the medium of superheated steam turbines. Gas turbines, and particularly CO2 turbines would be preferable, if available, but they are not an option yet. As it is the use of superheated steam would make the MSDR more efficient than Light Water Reactors.

Thus it would appear that development of the original MSDR design is not warranted, but that development of the design concept could be. Compared to a MSBR (a LFTR), an advanced MSDR would be a design slam dunk, because of the reliance on tested technology and because of the simplicity of the design. Not only would the MSDR cost less to manufacture than LFTRs, it would probably cost significantly less than LWRs and IFR SMRs such as the ARC-100.

In our current energy situation, a MSDR type reactor would be highly desirable. Not only would it serve as a route to a LFTR type molten salt thorium breeders, but it would offer a potential low cost alternative to the Light Water Reactor, that would be both safe, and would reduce the nuclear waste problem.