Lightbridge metallic alloy fuel provides upgrade path for LWRs 1

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  1. Will this new fuel assembly allow a higher percentage of the actinides in the fuel to be burned?

    It’s about time someone developed an upgrade for that 50+ year old technology. Is there anything anywhere we are using that hasn’t been improved in 50 years?

        1. Well, there’s more water, but the enrichment is higher. It’s too difficult for me to guestimate what the total effect on the H/U ratio is without more information.

          1. Higher enrichment isn’t going to increase U-235’s cross-section for fast neutrons.  The neutrons are mostly going to zip through until they get to water and slow down.

            There’s a much greater volume of water in a Lightbridge fuel assembly than one made of conventional fuel pins.  I recall that one of the design goals of the LWBE core was to squeeze water out to harden the neutron spectrum; this is the opposite.

            If we’re lucky, someone from Lightbridge has a Google alert on this stuff and will drop by to give us the straight dope.

          2. Higher enrichment isn’t going to increase U-235’s cross-section for fast neutrons.

            It will increase the macroscopic cross-section, which is the relevant parameter here and which is directly proportional to the atom density. Increasing the enrichment both increases the macroscopic fission cross-section (over all energy ranges) of U-235 and decreases the macroscopic capture cross-section of U-238, all other things being equal.

            The macroscopic cross-section also depends on the (mass) density of the fuel, however, which is something that I don’t know, so this is why I say that I cannot determine how this fuel design is going to affect the neutron spectrum relative to conventional LWR fuel. All I can say is that it will still be in the LWR range.

    1. Progress at last. Waste reduced by 60%. Best news I’ve heard for the industry in years.

      Thanks for the link.

  2. Question – Will this fuel use more or less natural uranium?

    If fuel is 5% enriched now and will be 20% enriched using Lightbridge, then four times as much natural uranium is needed to make the fuel. But burn-up is not four times more so Lightbridge might use more uranium and leave only 79% of the U235 remaining.

    I get a little confused when trying to combine enrichment and burn-up to think about the original uranium. What is the clear way to think about this?

    1. There’s not only the amount of natural uranium to think about. There’s also the energy required to enrich it.

      But the real volume of waste will be the number of forests that will be felled to produce the documents to license the thing.

      1. @Brian Mays

        With modern centrifuge enrichment, energy per SWU is about 5-10% of what it was with gaseous diffusion enrichment. Much smaller deal than before. Besides, URENCO is smart enough to be working on U-Battery power plants to supply their enrichment facilities.

        Your new (or near future) employer did that long ago with 4 Framatome 900 MW plants adjacent to George Besse I. The conversion to centrifuges has freed up about 2600 MWe for sale to other customers.

        1. The figures I see for centrifuge enrichment are 40-60 kWh/SWU, compared to ~2500 for GD.  That’s about 2% as much, and puts the energy cost down into the noise.

    2. There is the option to downblend HEU from warheads, although many consider such action as a sin/shame.

      However, I do not agree that the Lightbridge design using 20% LEU burns up a lesser amount of U235 – I believe it burns up more. The LTB17-1024™ not only puts out 10% more thermal power, it runs for 6 months (30%) longer than standard pin fuel per fuel shuffle.

  3. Question – Will this fuel use more or less natural uranium?

    By natural uranium I mean the uranium found in nature? Will more mining or wells be needed to obtain uranium?

    For this question do not think about HEW and do not think about energy to enrich which are interesting but not for this question.

    1. @Martin Burkle

      I should disclose that I own stock in several uranium mining companies. IMO, it will be many years before any increased uranium consumption requiring a little more mining becomes a problem worth worrying about.

  4. Higher enrichment suggests more uranium (barring a step change in enrichment efficiency) but there is less uranium in the fuel,
    “Though still using low enriched uranium, longer operating cycles, higher thermal power and lower uranium content associated with the high alloy fuel require closer to 20% enrichment than the current 5% standard.”

    So the answer is maybe, you would need more technical details from the company. They do have some technical articles on their website for viewing.

    My question is, since many plants already run at 24 months, what will their power uprate be? Also what are the barriers (legal or regulatory) to higher enrichment, do laws need to be changed, or just regulations, or nothing?

  5. So, my default guess is that the Lightbridge fuel will use twice as much natural uranium as existing fuels.

    1. Not quite.  Producing 4.5% LEU from 0.7% NU with 0.2% tails assay:

      0.7 = 4.5 x + 0.2 ( 1 – x) -> 0.7 = 4.3 x + 0.2
      -> 4.3 x = 0.5 x = 0.116, 88.4% tails.

      18% LEU, all else the same:

      0.7 = 18 x + 0.2 (1 – x) -> 0.7 = 17.8 x – 0.2
      -> 17.8 x = 0.5 -> x = 0.0281, 97.9% tails.

      Unless I have my arithmetic really wrong, per unit U-235 in the product you’re only consuming about 3% more uranium.

      1. Lightbridge fuel will be getting less of a plutonium boost towards the end of each fuel cycle. From memory, standard UO2 fuel gets about half its heat from the fission of plutonium, rather than U235, during the last couple of months of burnup. The metal fuel will have less U238 to convert – ~40% of the metal atoms, instead of 95% – and a softer neutron spectrum.

      2. Unless I have my arithmetic really wrong, per unit U-235 in the product you’re only consuming about 3% more uranium.

        That’s per U-235 load, but the relevant quantity is how much natural uranium is required per GWh of energy produced.

        Based on the enrichments you chose, it takes a little over four times the amount of natural uranium to produce one tonne of higher-enriched fuel than to produce one tonne of (conventional) lower-enriched fuel. In another comment here, someone has posted a link to the WNA that says that “the target burn-up is … about three times that of oxide fuels” or around 200 GWd/MTU.

        So if it takes about four times the amount of mined uranium to produce a ton of fuel and you get about three times the amount of energy out of that ton, then this design will require about one-third (~33%) more mined uranium than current Generation II designs operating in the US today.

        1. Thank you Brian. I accept 1/3 more natural uranium.

          This is not a problem. There’s lots of cheap uranium and someday we may finance the engineering development (research is done) of reactors which have a burn-up of 90+%.

          John, your comment about burning less U238 forced me to read about burn-up. Light bridge is calculating a burn-up of 21% of the original uranium atoms both U238 and U235. The current burn-up rate is 6.5% of the original uranium in the fuel.

  6. Do you (collectively or individually) have an opinion on the Lightbridge fuel’s commercial appeal and/or viability? Is this design attractive to large strategic entity in the industry?

    1. Not very far off topic. “Without the Hot Air” has been an invaluable resource for most things environment and energy, when numbers are needed and hand-waving just won’t do. I view Prof MacKay as largely responsible for Britain’s continued rational energy policy. This is indeed a loss.

  7. The article mentions that this new fuel would require close to 20% fuel enrichment. I think I’ve heard it said before that you could get benefits even with conventional (Uranium-oxide ceramic) fuel pellets if you could enrich up to about 20%. Yet, that hasn’t historically been done.

    What are the reasons that hasn’t been done much for commercial power plants (at least in the US – I think plants in some other countries maybe do use higher enrichment?), and do those reasons apply to this new fuel as well (for example, I think 20% enrichment has been considered a proliferation concern; wouldn’t it still be a proliferation concern that would engender political opposition against this new fuel)?

    1. What are the reasons that hasn’t been done much for commercial power plants …

      Several reasons. Historically, higher enrichment has involved higher costs for fuel manufacturing. Therefore, a substantial advantage of a higher-enrichment design would have need to be demonstrated before a switch had been warranted.

      In addition, there has been a lot of work done — both engineering and regulatory — to optimize the 4-5 wt% fuel that the plants currently use. Much of this would need to be redone if the enrichment were substantially increased.

      I think 20% enrichment has been considered a proliferation concern; wouldn’t it still be a proliferation concern that would engender political opposition against this new fuel

      No. The 20% limit is generally considered low enough to not be a proliferation concern. That said, if a country is stockpiling a large amount of 20 wt% LEU for dubious reasons (e.g., Iran) then the proliferation concerns begin to resurface.

    2. Cladding failure is the concern about the higher enrichment – the cladding will not last long enough to consume the higher enrichment. The pellet swells due to the fission products created, especially the ones that are gaseous at those temperatures. The metal fuel rods conduct heat somewhat better, thus those FPs are at a significantly lower pressure. Also, the metal alloy crystal structure is more ductile than the ceramic oxide fuel and can ‘bend not break’ somewhat better.

  8. Executives in unregulated electricity markets seem more prone to want to shut down nuclear plants than to change the regulations that fail to correctly value nuclear energy, and that are preferential to unreliable “renewables” i.e. natural gas. I wonder whether the prospect of 10-17% uprates to the older reactors might change the calculations of these reptiles?

    1. If what I read about the ability of utility companies to purchase NG through captive subsidiaries (forbidden by PUHCA, since repealed) is correct, the prospects of markups on self-dealt natural gas are likely much greater than improved revenues from power uprates and fewer fueling outages.

  9. I suspect that Westinghouse, Areva and other reactor manufacturers are redesigning to accommodate a fuel that promises about 30% more bang for the buck. But, we will not hear about this, because this would threaten to cannibalize sales of the existing design. A 30% cost saving would obviously stimulate the nuclear renaissance. It might even put a cork in a few of the anti-nuclear zealots, who have seized on costs, and FOAK problems, as their main attack.

  10. Interesting piece, Rod – it seems “low hanging fruit” both for new build and existing plant.

    I suspect before broader take up there’ll have to be more work on (a) decay heat removal impacts, and (b) confirmation that it at very least doesn’t make worse, and at best eliminates hydrogen production issues. On the former, I’m conscious that AP1000 being uprated to form CAP1400 seems not to have hit limits for passive cooling, which in its turn suggests that with the 17% uprate option, and an equivalent rate in decay heat production, this could go into an AP1000 without any great compromise to passive cooling.

    One thought does occur, possibly more for this side of the pond than yours – what’s the impact on things like de-canning during reprocessing? The “metallurgical bonding” and the fuel itself being alloyed with zirconium makes it sound like something radically different from existing PUREX derived processes would be needed – but I’m no chemist. Could this be a better candidate than current fuel designs for electrochemical methods?

    1. @ Andy
      This is the thing that kept me awake last night. That IFR/PRISM is the poster-child for electrochemical processing followed by remote metallic fuel-pin re-fabrication, and that IFR’s U-Pu metallic fuel is 10% Zr, suggests a distinct possibility.

      It depends on what you want the reprocessed fuel for. IFR electrochemical processing separates the used fuel into three components: pure Uranium, (U + Pu + Minor Actinides), and fission products. The fission products “waste” is removed from the cycle, leaving the Uranium and transuranic fractions to be blended with a bit more natural or depleted Uranium, and cast into new fuel slugs.

      The Minor Actinides are intentionally kept with the Plutonium for proliferation resistance, and here is where fast reactors differ markedly from thermal. The minor actinides are a fast reactor fuel, but as they are formed by thermal neutron capture cascade from Pu-239, a fast neutron spectrum generates relatively little of them and transmutes the rest to fission products leaving only a small amount of minor actinides in equilibrium.

      If you apply this same electrochemical process to used LWR fuel — as has been done in the laboratory after reduction of oxide fuel to metal — then in principle you can use the recovered U+Pu+MA fractions much like you currently do with PUREX. Except after electrochemical processing these fractions are metallic mixtures, not aqueous. Which may have some advantage if you want to re-fabricate metallic fuel.

      But if the metallic fuel you want to re-fabricate is Lightbridge LWR fuel, you’re still stuck with the same twice-through limit on the Pu+MA fraction as you have with oxide fuels and PUREX: the minor actinides and heavier Plutonium isotopes build up to where they are no longer suitable for LWR fuel. But at least they’re separate from the shorter-lived and more radiotoxic fission products.

      You can continue the LWR fuel re-processing past the first pass, of course. But if you want to re-use it again as LWR fuel, you’ll need to dilute the Pu+MA fraction considerably more with fresh Uranium (plus first-pass Pu) as is done with REMIX fuel.

      Otherwise you will need a fast-spectrum reactor to burn it. In which case the 2nd-pass still-lightly-used LWR fuel is just so much feed-stock for fast reactors. I’m sure Lightbridge has considered all such possibilities. But they need to get their initial fresh fuel qualified and approved by NRC first. By the time that is done and the first pins burned it will be early to mid 2020’s, by which time we’ll have better idea of Gen IV fast reactor deployment time frames.

      See WNA Mixed Oxide (MOX) Fuel.

      1. That IS fascinating, isn’t it?

        Something I’d like to know is what pyroprocessing does with thorium.  Turning every LWR into a near-unity converter is a fascinating possibility, but if the reprocessing just shunts thorium into the waste stream it would make for expensive reuse and disposal.

        I could see segmented fuel bars which are physically cut into pieces for separate processing depending on whether they are Th-based breeding/poison/neutron shield or U-based driver fuel.

        1. Not a fuels engineer, but its probably simpler to keep the thorium breeder blanket pins separate from the U-Pu drivers. This is the traditional approach; then you can use whatever chemistry best suits which pin, and via shuffling adjust the relative neutron fluxes as breeding and burn-up progress. This scheme is implemented in Russia’s Radkowsky Thorium Reactor, which uses a metallic U-Pu fuel in the seed region, and (apparently) Thorium oxides in the blanket.

          The blanket pins don’t breed transuranics, I imagine their Th-Pa-U chemistry is similar to TMSR designs and don’t require electrolytic reduction. There may be eventual application of metallic blanket pins rather than oxide, but oxides and carbides have been where all the action is.

          What you do with the separated uranium is a different story. Don’t see why thorium-bred U-232,3,4 couldn’t be slotted into Lightbridge LWR metallic fuel, but its radiotoxicity complicates fabrication so that’s for them to determine, qualify, and license.

          WNA’s Thorium suggest many other possibilities though again, all previous thorium fuels research has been on oxides, carbides, or molten salt. And that experience counts for a lot.

          1. The issue I can see about separate pins is that you have wide variations in thermal output between the two types, and you might lose part or all of your power uprate.  If you have segmented pins heat will conduct between zones of different specific power output (assuming they’re small) and the fuel assembly as a whole can have the same output power.

          2. Oh I see. You want both the power up-rate and thorium breeding — simultaneously on an existing production reactor and without changing the core pin carrier configuration. What could go wrong? Not even NRC would have a problem with that!

            🙂

            Seriously. For pin fabrication and reprocessing reasons alone I’d think you’d have at least as much luck (that is to say, none whatsoever) with homogeneous pins. But if you aren’t going to add more pins to the core, then something’s gotta give. And if it ain’t power, it ain’t pin geometry or uranium content.

            Lightbridge’s uranium is already at the 20% LEU limit. Their 50% Zirconium fraction was chosen for high melting point; their sales rep might not want to lower.

            Any existing art on the metallurgical properties of Thorium/Protactinium in Zirconium? There’s over 50 years with Uranium/Plutonium.

            Breed-and-burn Th-Pu oxide pins have been tested for BWRs. But Plutonium is dear, Lightbridge does metal, and their primary target is existing commercial PWR’s. As you noted, commercial operators are unlikely to sacrifice power generation or lengthened re-fueling schedule for an innovative bit of in-plant uranium breeding. You want to breed U-233 from thorium, use a reactor spec’d for the purpose. That’s what all the cool kids do.

          3. Oh I see. You want both the power up-rate and thorium breeding — simultaneously on an existing production reactor and without changing the core pin carrier configuration. What could go wrong? Not even NRC would have a problem with that!

            Who wouldn’t want to have it all?

            Lightbridge’s uranium is already at the 20% LEU limit. Their 50% Zirconium fraction was chosen for high melting point; their sales rep might not want to lower.

            Thorium’s melting point is higher than zirconium.  This doesn’t necessarily mean U-Th alloys will have a higher MP, but it is suggestive.  However, that would require co-processing U and Th with the possibility of losing lots of Th along with the FPs as Th is chemically similar to rare earths.  OTOH, as it’s essentially a REE waste product today….

            Thorium has 80% higher density than zirconium.  50 wt% Th is going to take up substantially less volume:  more fuel in the same space.

            But if you aren’t going to add more pins to the core, then something’s gotta give.

            You have a point there.  Separate thorium elements increase the total volume required.  However, if the fuel design allows a 30% power uprate but the reactor cannot handle the extra output, you’re not losing anything if you sacrifice some water flow to get additional breeding material (perhaps a 5- or 6-pointed star configuration with different alloys in different points?).  If you can achieve a 30- or 36-month fuel cycle you’ll have something attractive to the customer, particularly if you are licensed to take the used fuel off their hands afterward.  If you can mechanically separate the different alloy sections for different handling (such as a Zr bonding layer thick enough for electrical discharge machining without getting into the fuel alloys) you may be able to get a value stream big enough to justify the extra work.  Or just treat it as a long-term investment; SNF isn’t going anywhere.

            I admit to being curious about the method Lightbridge proposes to clad the fuel elements.  Plasma-spraying would be able to coat almost anything, though explosive bonding has a certain dramatic attraction.

            Breed-and-burn Th-Pu oxide pins have been tested for BWRs. But Plutonium is dear, Lightbridge does metal, and their primary target is existing commercial PWR’s. As you noted, commercial operators are unlikely to sacrifice power generation or lengthened re-fueling schedule for an innovative bit of in-plant uranium breeding.

            The point of having a substantial amount of breeding going on in the core is to flatten the reactivity over time and extend fuel cycles further.  Being able to reprocess metal fuel without using wet chemistry, capturing uranium as metal to be cast back into fuel alloy with minimal handling (IFR-fashion) would be a major advance.  If handling can be automated, gamma emissions from U-232 daughters would not be as much of a factor.

          4. Well, IIRC Thorcon uses/intends to use a U-Th liquid fuel, and all the other TMSR somehow managed to separate the FP’s. I imagine a really good fuels engineer could figure it out for LIghtbridge.

            They’ve got an open req. Wanna do a joint apply?

            One of the beauties of LIghtbridge’s design is its simplicity. One-shot extrusion vs. fairly complex though still automated packing, sintering, machining, QA-ing, and cladding individual fuel pellets, then loading them in a Zircalloy tube.

            I kind of like that simplicity. Adding additional layers and maybe mechanically machining one of them off isn’t the direction I’d take. Yeah, I thought of using breeding to flatten core reactivity as well. The question is where to do it. Lightbridge already has a displacer at their pin core, though that might not be the best place to find thermal neutrons. Maybe put the thorium blanket just beneath the cladding and let uranium replace the displacer. Still have only a three-layer pin.

            That, or just put the thorium in separate blanket pins where the chemistry — metal, ceramic, pixie dust, whatever — residence time, neutron flux, and coolant flow is optimized for thorium. And the bred U-233 isn’t diluted by all that useless 238.

            Its all so far down the road, if ever. The problem with actinides is there’s so many of them. Truly an embarrassment of riches. Lightbridge has a New! Improved! Simplified! Uranium fuel design that Areva and half the reactor operators in this country are chomping at the bit to fabricate and use. They’ve got irradiation tests scheduled in Norway and Sweden. NRC has been notified and is making room. Its great to fantasize about what other cool things might be done with this tech, but the important thing is for them to get it Out There and making steam, which might not be as easy as it looks.

            Thanks for the discussion!

  11. Is there any film footage of any test where the fuel elements of a fully powered reactor are literally yanked/sprung clear out the “runaway” reactor to suspend in (confined) mid-air to see how or if they melt like white hot icicles? Is the main concern in meltdowns the massive reactor vessel melting along with the fuel which could be circumvented it the rods were no longer there? Were there ever any “wild” proposals like automatically physically ejecting reactor fuel elements out of a reactor and just dealing with the naked fuel cores lying on the floor or suspended without any water to boil and pressure up or metal to melt? Hope I’m posing this right!

    James Greenidge
    Queens NY

    1. Well… the reactor pressure vessel head is bolted on fairly tight, and the fuel assemblies held together rather firmly by their plenums as well. To break the whole thing apart and eject the core would require something on the order of an atom bomb.

      Okay. Maybe a bit less. But only a bit. You get the idea.

      Pause to consider what you mean by a “runaway” reactor. Either the core is covered with sufficient moderator/coolant to sustain the chain reaction, or it isn’t. If it is, then the core is cooled and not running away. If it isn’t, the core is exposed, but with no moderator there is no reaction.

      But there is a lot of heat still being generated by short-lived fission products trapped in the fuel rods. (Solid-fuel rods — MSR fans please don’t all pile on at once.) The FP decay heat is a known and finite quantity, generated over a known time profile that (eventually) goes to zero.

      In the meantime, during a loss-of-cooling accident (LOCA) the question is how hot does the fuel’s Zirconium cladding get before the emergency core cooling system (ECCS) flood it with water. Please see page 13 of the company’s current presentation.

      There are several things going here. Maybe more. Oxide fuel – UO2 — is ceramic and doesn’t conduct heat really well. Not compared to metals, whose heat conductivity pretty much matches their electrical. So
      “The temperature down the center of Lightbridge fuel is over 1000 °C cooler than the center of standard nuclear fuel; and with ~35% more fuel surface area there is significantly more margin to fuel failure. ” (CCP page 11)

      PWR’s operate at about 305 C, Lightbridge models their fuel’s center temperature to run 350 C. As I understand it 1000 C is the range where Zr+H20 reactions start to initiate, though they don’t really get underway until about 1300 – 1580 C. (See Hydrogen Generating Reactions in LWR Severe Accidents Table I. Reaction rate constant is hundred-fold higher at 1580 than 1000 C.

      But the Zr cladding is on the outside of an oxide fuel pellet, not it’s center, and Zr temperature is limited if the ECCS runs properly. Which for various reasons it didn’t at Fukushima-Daiichi but that’s another story.

      Loss-of-cooling accidents are modeled as part of a reactor’s design basis. The reactor must survive them. LBLOCA Analysis in a Westinghouse PWR 3-Loop Design gives an example of how such modeling was done in the ’90s. Google “LBLOCA”, and hope Rod or Brian or Rich or someone more knowledgeable than myself drops by for better explanation.

      Beneath the water reaction temperature, Zirconium has attractive properties for use in reactor fuels and fuel assemblies. Corrosion resistance, high melting temperature, compatibility with uranium. Replacing it isn’t easy, although silicon carbide is beginning to show some promise. This Lightbridge “everything is a metal” thing is a thing to wrap one’s head around. Quite apart from power and safety, it also opens intriguing fabrication and reprocessing possibilities. 😎

    2. @ James Greenidge
      Apologies for multiplicity of posts, but it’s an intriguing “what if.” Core-catching is not the only ultimate option, for instance Westinghouse AP1000 design floods the containment external to the reactor pressure vessel with sufficient emergency cooling water to prevent molten-core melt-through in the first place. See NRC Probabilistic Risk Assessment 19.39 In-Vessel Retention of Molten Core Debris.

  12. Thanks for the info tips Ed!
    It would be a nice engineering exercise to design such a “pop-top” reactor vessel that would facilitate forceful ejection of core elements in imminent meltdown via super springs or compressed gas or literally being shot out. I was told in SpaceWorld that the thousand ton concrete lids of Minuteman missiles silos are designed so that if hydraulics and springs fail to slide the lid away on launch that high explosives will literally blow it off without scratching the missile underneath (supposedly Polaris had the same thing in case the missile hatch was warped shut from stress during combat), so I just wonder whether such an concept ever surfaced way far back or made the drawing boards — and more intriguing, would such a “last ditch” meltdown preventer work via today’s technology? I wonder if the public would feel “safer” and accept nuclear better if they thought such a dramatic way to avoid a meltdown was available besides the “mere simple undramatic” solution of constantly dousing reactors with endless water. Again, just a non-techie asking!

    James Greenidge
    Queens NY

    .

    1. How many times you heard on Star Trek to jettison the core! Jettison the core!! before the antimatter reactors explode??

    2. @James Greenidge.
      That fission-product heat source term is not to be trivialized. Nor are the radioactive fission products themselves. Absent cooling water, freshly abused LWR fuel pins are going to take on a white-hot glow and melt. They’re going to. You’re inquiring into the feasibility of packing a half ton of shape charges around the reactor head and core, blowing the reactor head off, then spattering white-hot molten corium all over the containment building and halfway beyond.

      Sure, you could do it. But there’s no obvious advantage over just letting the stuff collect in a single well-contained shielded puddle at the bottom of the reactor pressure vessel or core-catcher. Given enough water and/or sacrificial concrete — and these are given — and a month or so of time, it will eventually cool down and freeze. In the same small, well-contained, and shielded place.

  13. It’s been known for quite some time that it should be possible to implement a closed 233U/Th based nuclear fuel cycle with CANDU type reactors (e.g., https://www.researchgate.net/publication/232388182_Investigation_of_CANDU_reactors_as_a_thorium_burner). By “closed” I mean that after the reactors are started with reactor-grade Pu, no “new” uranium would have to mined, enriched, etc. That paper also suggests that actinide burn per cycle could be as high as about 17% which translates to far less “reprocessing” per GWe than much of France’s currently unsustainable nuclear fuel cycle’s spent fuel has already undergone. It seems to me that Lightbridge-type fuel rods would be especially attractive for this application if it were possible to make their fuel “meat” in a way (e.g., sinter metallic BBs together) that preserves sufficient porosity to accommodate FP buildup. The reasons for these/my conclusions include: 1) metallic thorium is “easy” to dissolve (facilitates recycle) – ThO2 is a real bitch; 2) metallic fuels shouldn’t get nearly as hot as do the traditional oxide ceramic-based fuels; 3) at steady-state, far less plutonium – the world’s most politically charged element – would be generated/GWe than would be the case with either IFRs or Pu/U-based molten salt reactors (MCFRs); 4) it’s easier (cheaper) to separate rare earth FP from U than from Pu/MA; and finally, 5) since the nuclear industry’s leadership is already familiar with CANDU reactors, proposing to implement a genuinely sustainable Th-based nuclear fuel cycle with them isn’t apt to generate as much resistance as does attempting to do the same thing with any sort of molten salt reactor.

  14. Since posting my comment I’ve read several other reports that GOOGLING “CANDU thorium fuel cycle” dug up It seems that the conclusions of the RESEARCHGATE paper that I’d built my conceptual house of cards upon are inconsistent with expert consensus; i.e., while it’s probably possible to implement a self sustaining Th-based fuel cycle with CANDU-type reactors, doing so is apt to require too much reprocessing to be practical.

    That paper appears to be another example of how peer review isn’t infallible. Of course, I can’t tell for sure whether or not it’s “wrong” because neither its scenario nor the modification I’d suggested (substituting metallic for ceramic fuels) is exactly the same as any discussed by the other papers/reports/theses.

    Anyway, this means that the scenario I’d suggested is a lot more speculative than either I or any other pro nuke would like it to be.

    Can anyone following this thread this identify why the paper I’d referenced apparently reached the “wrong” conclusion?