Nuclear scientists raise problems with proposed Liquid Fluoride Thorium Reactors (LFTRs)

Thorium Cycle questions and problems http://daryanenergyblog.wordpress.com/ca/part-8-msr-lftr/8-3-thorium-lftr/Questions have also been raised by some nuclear scientists about the Thorium cycle, in particular the proposed one that the LFTR would use. I’m not a nuclear physicist so I’ll merely forward you on to the relevant paper here, and a rebuttal here. The crux of the argument seems to be the proliferation risk (I’ll come back to that one later), the fact that a number of its spend fuel outputs (such as Technetium-99) are “nasty stuff” with a long half life and the fact we’ll still need supplies of Uranium to get Thorium reactors going again whenever we have to turn it off (which will happen at least once a year or so during its annual maintenance shutdown). They also highlight a number of technical issues, which I discussed in the chapter on HTGR’s.

Certainly the fission products from a Thorium reactor are a worry, Technetium-99 has a half life of 220,000 years, uranium-232 produces thallium-208 (a nasty wee gamma emitter), Selenium-79 (another gamma emitter with a 327,000 year half-life), evenThorium-232 is a problem with its half life of 14 Billion years (and while the T-232 isn’t a major worry, all the time during this 14 Billion years it will be decaying and producing stuff that is!).

The UK based NNL (National Nuclear Laboratories) also pour cold water on the idea of Thorium fuelled reactors (see here). While the report is low on detail (they seem to be saying “trust us we’re scientists who work with nuke stuff… and we smoke pipes!”) they do highlight the major time delays it would take to establish and get working a Thorium fuel cycle (10-15 years with existing reactors, 30 with more advanced options), point out that under present market conditions its unlikely to be economically viable and will (as the points above raise) offer only a modest reduction in nuclear wastes.

MIT recently undertook a study of future nuclear fuel supplies. The Thorium cycle barely gets a mention, and even then its usually in relation to Fast Reactor programs (of which the US currently has none) and modifed LWR systems, rather than the MSR.

Obviously, once we exhaust the world’s U-235 stockpiles, LFTR’s and any other Thorium fuelled reactors will cease to function. Indeed long before then the spike in Uranium prices will have rendered MSR’s (and all other nuclear plants) uneconomically viable (of course there’s plenty who’d say that’s already the case!). The LFTR fans usually groan at this point and state that “all we need is a little plutonium”. Now while I’m quite sure that in the fantasy world which the LFTR fans inhabit Plutonium is available in any good hardware store but back in the real world, it’s a little harder to come by! As with the HTGR’s using Thorium (if its possible) would certainly help stretch things out….a bit! But not by nearly as much as the supporters of Thorium reactors would have you believe.

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One thought on “Nuclear scientists raise problems with proposed Liquid Fluoride Thorium Reactors (LFTRs)

  1. Thorium: Not ‘green’, not ‘viable’, and not likely

    1. Introduction
    “With uranium-based nuclear power continuing its decades-long economic
    collapse, it’s awfully late to be thinking of developing a whole new fuel cycle
    whose problems differ only in detail from current versions.”
    Amory Lovins, Rocky Mountain Institute, March 2009.

    A number of commentators have argued that most of the problems associated with
    nuclear power could be avoided by both:
     using thorium fuel in place of uranium or plutonium fuels
     using ‘molten salt reactors’ (MSRs) in place of conventional solid fuel reactor
    designs.

    The combination of these two technologies is known as the Liquid Fluoride Thorium
    Reactor or LFTR, because the fuel is in form of a molten fluoride salt of thorium and
    other elements.

    In this Briefing, we examine the validity of the optimistic claims made for thorium
    fuel, MSRs and the LFTR in particular. We find that the claims do not stand up to
    critical scrutiny, and that these technologies have significant drawbacks including:
     the very high costs of technology development, construction and operation.
     marginal benefits for a thorium fuel cycle over the currently utilised uranium /
    plutonium fuel cycles
     serious nuclear weapons proliferation hazards
     the danger of both routine and accidental releases of radiation, mainly from
    continuous ‘live’ fuel reprocessing in MSRs
     the very long lead time for significant deployment of LFTRs of the order of half
    a century – rendering it irrelevant in terms of addressing current or medium
    term energy supply needs

    1. Background

    1.1 What is thorium?

    Thorium is a heavy metal named after Thor, the Nordic God of thunder. The naturally
    occurring isotope, 232Th, is mildly radioactive with a very long half life of 14 billion
    years. Thorium presents a health hazard mainly from inhalation of dust, and from
    emissions of the powerfully radioactive gas radon (220Rn).
    It occurs mainly in deposits of rare earth metals. . As it has few uses requiring more
    than minimal volumes of material it is considered as radioactive waste – and requires
    careful and expensive handling to prevent contamination. It is three to four times
    more abundant in the Earth’s crust than uranium, and is especially plentiful in
    Australia, Norway, India, the USA and China.
    Although thorium can be used to make nuclear fuel, it is not fissile. But it is ‘fertile’ –
    that is, it can transformed into fissile material. Under neutron irradiation, typically
    provided by the fission of uranium or plutonium, it breeds the fissile uranium isotope
    233U. Thus any thorium fuel cycle needs to be initiated by an supply of existing
    fissile material.

    The thorium-uranium fuel cycle has some advantages over the dominant uraniumplutonium
    cycle, in terms for example, of the reduced production of long-lived actinides and somewhat diminished radio-toxicity overall. However, it also creates new hazards of its own. As far as radioactive fission products are concerned, there is little to choose between the two.
    (see Appendix 1 for further details)

    1.2 What are molten salt reactors?
    Unlike conventional nuclear reactors which use solid fuel in the form of rods or
    pellets, molten salt reactors (MSRs) use fuel in the form of a complex mixture of
    fluoride salts in a molten state. The salt mixture includes the fissile material (fissile
    isotopes of plutonium and/or uranium), together with any fertile material (such as
    thorium or 238U) together with other elements. The molten fluoride salt serves as the primary coolant, carrying heat away from the reactor, and delivering it to a secondary cooling circuit and ultimately to the steam turbines that generate electricity. In principle, MSR’s offer several potential advantages over conventional reactor designs:

     the reactor and its cooling circuits operate at near atmospheric pressure,
    reducing the chance of any explosion
     In the event of a reactor overheating, the fuel can simply drain out into a
    secondary container and the fission chain reaction will halt, reducing the risk of
    reactor meltdowns such as those experienced at Chernobyl and Fukushima
    But before ‘production’ MSRs can be built, there are significant technical problems to
    be overcome, among them:
     the development of corrosion-resistant materials capable of surviving for
    decades in a uniquely hostile environment – highly corrosive and subject to
    intense radiation including neutron bombardment
     and, the continuous fuel reprocessing that MSRs demand, requiring the
    development of hazardous, complex and currently experimental pyro-processing
    and electro-refining technologies on a production scale.
    If these technologies are successfully developed – and it cannot be taken for granted
    that they ever will – they are likely to be very expensive. Moreover, reprocessing will
    always represent a weak link from a safety and proliferation perspective.
    (See Appendix 2 for further details)

    2. Current State of Play

    2.1 Actual thorium reactors

    Thorium fuel has so far been used in about 30 operational reactors in conjunction with
    fissile uranium (235U / 233U) or plutonium (239Pu) to initiate the fuel cycle. Most of
    these were located in the USA, Germany, Netherlands and India. A single example
    operated in the UK, from 1965 to 1976: the Dragon Reactor at Winfrith, a heliumcooled
    test reactor evaluating fuel and materials for the European high-temperature
    reactor programme. It is currently partially decommissioned. Most thorium reactors have been of conventional designs originally intended for uranium fuel, such as pressurised water reactors, boiling water reactors and pressurised heavy water reactors. But thorium has also been included in more exotic designs, notably the molten salt breeder experiment (MRSE) reactor (see 2.3, below),
    the thermal breeder reactor (USA), and the liquid metal fast reactor (India). The only operational thorium reactors today are in India, which possesses abundant thorium reserves but little uranium. These are all solid fuel reactors. As of 2010, India had used only a small amount of thorium – approximately one tonne – in its reactors.

    2.2 Planned thorium reactors

    In December 2011, India announced its plans for a new generation of Advanced
    Heavy Water Reactors using a plutonium / uranium / thorium MOX (mixed oxide)
    fuel. The programme would begin with an initial test reactor whose construction could
    commence in 2013. Again, this would not be a molten salt reactor but would use
    conventional solid fuel.

    Norway’s Thor Energy is also intending to develop a thorium-plutonium MOX
    nuclear fuel, aimed at replacing conventional fuels in light water reactors (LWRs). It
    is currently seeking investment to irradiate thorium-plutonium oxide fuel pins in
    simulated LWR conditions in the Halden fuel-testing reactor. A separate project is to
    optimise thorium-plutonium fuels for boiling water reactors, while maximising the
    breeding of 233U. Thor Energy anticipates that 25-30% of power output could arise
    from the thorium.

    Proposals have been made to construct LFTR reactors in China, Japan and the US (see
    2.3). Initially these would be research reactors and the first ‘production’ LFTR would
    appear to be 20-30 years away (see 2.4).

    2.3 Actual molten salt reactors

    The molten salt reactor was originated in the 1950s as a potential power source for the
    USAF’s fleet of high altitude nuclear bomber aircraft. A working reactor was
    produced (under the Airborne Reactor Experiment) programme, but never
    commissioned.

    The technology was further developed at Oak Ridge National Laboratory in the 1960s
    under its MSRE (Molten Salt Reactor Experiment). The 7MW reactor employed
    fluoride salts of uranium and plutonium as fuel. In the 1970s, Oak Ridge built its
    Molten Salt Breeder Reactor (MSBR), which used as fuel fluoride salts of uranium,
    thorium and plutonium as its fuel.

    2.4 Planned molten salt reactors

    There are a number of proposals to build MSRs:
     In January 2011, the Chinese Academy of Sciences announced plans to develop
    the LFTR technology into commercial reactors. But it warned that 20 to 30
    years of research and development would probably be needed before an LFTR
    was operational.
     Flibe Energy was set up in 2011 to develop LFTR technology in the USA and
    worldwide. Its initial intention is to build a small test reactor. Ultimately, it aims
    to bring about a world with many thousands of LFTRs.
     The FUJI LFTR project in Japan is attempting to raise £300 million to build a
    10 MW ‘MiniFUJI’ research reactor. Following the 2011 Fukushima
    catastrophe, the project has a low chance of attracting the necessary finance.
     the UK’s Weinberg Foundation was established in September 2011 to act as a
    communications, debate and lobbying hub to promote thorium energy and the
    LFTR in particular. There are currently no plans in the UK to build an actual
    LFTR.

    Despite the resurgence of interest in the MSR / LFTR technology, there are no
    concrete plans to build even a single such reactor. China currently appears most likely
    to provide the funding necessary to develop LFTR technology due to that country’s
    relatively large nuclear programme and the government’s willingness to invest in new
    energy generation technologies. But even there any production-scale LFTR is unlikely
    to materialise for 20-30 years.

    2.5 New-found interests – why?

    Several factors underlie the current vogue of interest in thorium reactors. Perhaps the
    most important is the desire for energy and nuclear independence in countries with
    large thorium reserves and little uranium, or which have concerns about long-term
    price of uranium and its availability. This would appear to apply to India, China, the
    USA and Norway.

    Noting the large volumes of surplus thorium produced as waste in the mining of
    valuable rare earth metals, there is also a clear commercial interest among the mining
    companies concerned to give value to this waste. However, we have no evidence of
    any efforts by mining companies to drive forward the thorium project.
    A more significant factor is perhaps a deeply-rooted techno-optimism in human
    psychology – the desire to believe that one or other technology provides ‘the answer’
    to deep-rooted problems. Faced with the prospect of ‘peak oil’ and accelerating
    climate change from the burning of fossil fuels, those who are sceptical about the
    potential of renewable energy sources will naturally incline towards some other
    answer. For some, it would seem that thorium fills that particular ‘desire gap’.

    The established nuclear industry in the UK has little interest in thorium as such, since
    any use of thorium would create far more cost than it ever saved. However, the mere
    idea that there exists a notionally ‘green’ version of nuclear power could be seen by
    the nuclear industry as positive in public relations terms, and useful in promoting the
    persistence of nuclear power in the UK’s electricity mix.

    3. Thorium claims – and the reality

    Numerous advantages for thorium as a nuclear fuel and for the LFTR design over
    conventional solid fuel reactors have been claimed. In this section we consider each of
    these claims in turn.

    3.1 Abundance of thorium relative to uranium

    Claim: Thorium is several times more abundant in the Earth’s crust than uranium.
    Response: Thorium (232Th) is indeed more abundant than uranium, by a factor of
    three to four. But whereas 0.7% of uranium occurs as fissile 235U, none of the
    thorium is fissile. The world already possesses an estimated 1.2 million tonnes of
    depleted uranium (mainly 238U), like thorium a fertile but non-fissile material. So the
    greater abundance of thorium than uranium confers no advantage, other than a very
    marginal advantage in energy security to those countries in which it is abundant.

    3.2 Relative utility of thorium and uranium as fuel
    Claim: 100% of the thorium is usable as fuel, in contrast to the low (~0.7%)
    proportion of fissile 235U in natural uranium.

    Response: Thorium must be subjected to neutron irradiation to be transformed into a
    fissile material suitable for nuclear fuel (uranium, 233U). The same applies to the
    238U that makes up depleted uranium, which as already observed, is plentiful. In
    theory, 100% of either metal could be bred into nuclear fuel. However, uranium has a
    strong head start, as 0.7% of it is fissile (235U) in its naturally-occurring form.

    3.3 Nuclear weapons proliferation
    Claim: thorium reactors do not produce plutonium, and so create little or no
    proliferation hazard.

    Response: thorium reactors do not produce plutonium. But an LFTR could (by
    including 238U in the fuel) be adapted to produce plutonium of a high purity well
    above normal weapons-grade, presenting a major proliferation hazard. Beyond that,
    the main proliferation hazards arise from:
     the need for fissile material (plutonium or uranium) to initiate the thorium fuel
    cycle, which could be diverted, and
     the production of fissile uranium 233U.
    Claim: the fissile uranium (233U) produced by thorium reactors is not
    “weaponisable” owing to the presence of highly radiotoxic 232U as a contaminant.
    Response: 233U was successfully used in a 1955 bomb test in the Nevada Desert
    under the USA’s Operation Teapot and so is clearly weaponisable notwithstanding
    any 232U present. Moreover, the continuous pyro-processing / electro-refining
    technologies intrinsic to MSRs / LFTRs could generate streams of 233U very low in
    232U at a purity well above weapons grade as currently defined.

    3.4 Safety
    Claim: LFTRs are intrinsically safe, because the reactor operates at low pressure and
    is and incapable of melting down.

    Response: the design of molten salt reactors does indeed mitigate against reactor
    meltdown and explosion. However, in an LFTR the main danger has been shifted
    from the reactor to the on-site continuous fuel reprocessing operation – a high
    temperature process involving highly hazardous, explosive and intensely radioactive
    materials. A further serious hazard lies in the potential failure of the materials used for
    reactor and fuel containment in a highly corrosive chemical environment, under
    intense neutron and other radiation.

    3.5 State of technology
    Claim: the technology is already proven.

    Response: important elements of the LFTR technology were proven during the 1970s
    Molten Salt Breeder Reactor (MSBR) at Oak Ridge National Laboratory. However,
    this was a small research reactor rated at just 7MW and there are huge technical and
    engineering challenges in scaling up this experimental design to make a ‘production’
    reactor. Specific challenges include:
     developing materials that can both resist corrosion by liquid fluoride salts
    including diverse fission products, and withstand decades of intense neutron
    radiation;
     scaling up fuel reprocessing techniques to deal safely and reliably with large
    volumes of highly radioactive material at very high temperature;
     keeping radioactive releases from the reprocessing operation to an acceptably
    low level;
     achieving a full understanding of the thorium fuel cycle.

    3.6 Nuclear waste
    Claim: LFTRs produce far less nuclear waste than conventional solid fuel reactors.
    Response: LFTRs are theoretically capable of a high fuel burn-up rate, but while this
    may indeed reduce the volume of waste, the waste is more radioactive due to the
    higher volume of radioactive fission products. The continuous fuel reprocessing that
    is characteristic of LFTRs will also produce hazardous chemical and radioactive waste
    streams, and releases to the environment will be unavoidable.

    Claim: Liquid fluoride thorium reactors generate no high-level waste material.
    Response: This claim, although made in the report from the House of Lords, has no
    basis in fact. High-level waste is an unavoidable product of nuclear fission. Spent fuel
    from any LFTR will be intensely radioactive and constitute high level waste. The
    reactor itself, at the end of its lifetime, will constitute high level waste.
    Claim: the waste from LFTRs contains very few long-lived isotopes, in particular
    transuranic actinides such as plutonium.

    Response: the thorium fuel cycle does indeed produce very low volumes of
    plutonium and other long-lived actinides so long as only thorium and 233U are used
    as fuel. However, the waste contains many radioactive fission products and will
    remain dangerous for many hundreds of years. A particular hazard is the production
    of 232U, with its highly radio-toxic decay chain.

    Claim: LFTRs can ‘burn up’ high level waste from conventional nuclear reactors, and
    stockpiles of plutonium.

    Response: if LFTRs are used to ‘burn up’ waste from conventional reactors, their fuel
    now comprises 238U, 235U, 239Pu, 240Pu and other actinides. Operated in this way,
    what is now a mixed-fuel molten salt reactor will breed plutonium (from 238U) and
    other long lived actinides, perpetuating the plutonium cycle.

    3.7 Cost of electricity
    Claim: the design of LFTRs tends towards low construction cost and very cheap
    electricity.

    Response: while some elements of LFTR design may cut costs compared to
    conventional reactors, other elements will add cost, notably the continuous fuel
    reprocessing using high-temperature ‘pyro-processing’ technologies. Moreover, a
    costly experimental phase of ~20-40 years duration will be required before any
    ‘production’ LFTR reactors can be built.

    It is very hard to predict the cost of the technology that finally emerges, but the
    economics of nuclear fuel reprocessing to date suggests that the nuclear fuel produced
    from breeder reactors is about 50 times more expensive than ‘virgin’ fuel. It therefore
    appears probable that any electricity produced from LFTRs will be expensive.
    We must also consider the prospect that relatively novel or immature energy sources,
    such as photovoltaic electricity and photo-evolved hydrogen, will have become well
    established as low-cost technologies long before LFTRs are in the market.

    3.8 Timescale
    Claim: Thorium and the LFTR offer a solution to current and medium-term energy
    supply deficits.

    Response: The thorium fuel cycle is immature. Estimates from the UK’s National
    Nuclear Laboratory and the Chinese Academy of Sciences (see 4.2 below) suggest
    that 10-15 years of research will be needed before thorium fuels are ready to be
    deployed in existing reactor designs. Production LFTRs will not be deployable on any
    significant scale for 40-70 years.

    4. Thorium / LFTR prospects

    4.1 Timescales for thorium fuel

    The thorium fuel cycle is immature and unready for production-scale deployment.
    Although thorium fuels have been used in approximately 30 reactors, their nuclear
    dynamics and operational performance remain poorly characterised.
    India is already deploying thorium in its reactors as a component of mixed oxide
    (MOX) fuels comprising plutonium / uranium, and plans more of the same in its
    forthcoming Advanced Heavy Water Reactors. However, Norway’s Thor Energy and
    the UK’s National Nuclear Laboratory (NNL) both believe that considerable research,
    development and testing lies ahead before thorium fuels will be ready for operational
    use.

    As the NNL states, “Thorium reprocessing and waste management are poorly
    understood. The thorium fuel cycle cannot be considered to be mature in any area.” It
    estimates that 10-15 years work is required before thorium fuels will be ready for use
    in current reactor designs, and that their use in new types of reactor is at least 40 years
    away. [The Thorium Fuel Cycle – An independent assessment, NNL, August 2010]

    4.2 LFTR lead time: half a century
    The assessment of the Chinese Academy of Sciences as it embarks on its LFTR
    programme is that a production LFTR is 20-30 years in the future – rather shorter than
    the NNL’s estimate of 40 years (see 4.1).

    Given the hazards, such as the potential failure of reactor materials under intense
    neutron irradiation and chemical corrosion, risk-averse utilities and investors would
    want to observe the performance of any such full-scale LFTR for at least a decade and
    probably more, before embarking on any substantial LFTR programme. The lead time
    for nuclear construction is of the order of a decade, so this could add a further 20-30
    years before production LFTRs were deployed at full scale.

    The total lead time for LFTRs would therefore be a minimum of 40 years on the
    shortest estimates, or 70 years based on more conservative figures.
    4.3 Thorium and LFTRs – investment outlook

    The development of thorium / LFTR technologies represents a poor investment for
    national governments, utilities and private investors given:
     the marginal benefits to be derived from using thorium fuels in existing reactor
    designs;
     the very long-term nature of any benefit that may be realised from LFTRs, of
    the order of half a century;
     the uncertainty as to whether the very significant technical challenges of the
    LFTR will ever be overcome;
     the possibility that the materials used for reactor construction may degrade more
    rapidly than anticipated, causing early shut-down;
     the likely very high cost of LFTR electricity – especially when compared
    against the anticipated low future cost of electricity from renewable sources,
    solar in particular, over the applicable time frame.

    As NNL states: “thorium is competing with the uranium/plutonium fuel cycle which is
    already very mature. To progress to commercial deployment would demand major
    investments from fuel vendors and utilities … LWR and PHWR utilities would be
    unlikely to invest in thorium fuels to the levels required under current market
    conditions. The potential savings that thorium fuels offer and other claimed benefits
    are insufficiently demonstrated and too marginal to justify the technical risk that the
    utility would be exposed to.”We therefore see little prospect that LFTRs will present an economic solution if and when they are ever ready for large scale deployment. Any money invested in LFTRs, whether by governments, utilities or other investors, is likely to be wasted.
    Far better to invest in the renewable technologies that are already shaping our national
    and global future, and whose cost is rapidly falling – in the process developing
    valuable UK-based expertise and technologies, and accelerating the renewables
    revolution.
    Oliver Tickell, April / May 2012.

    Appendix 1 – The thorium fuel cycle
    Thorium is not itself fissile, however it is ‘fertile’. That is to say that, under neutron
    irradiation, it can be used to breed fissile material. In any thorium reactor, the
    naturally occurring 232Th is irradiated with neutrons from fissile material (for
    example, 235U, 233U or 239Pu). Some of the thorium nuclei capture a neutron and
    become 233Th. This isotope then undergoes beta decay to 233Pa (protactinium 233)
    which in turn beta decays to 233U, a fissile isotope of uranium.

    So in a thorium reactor, the fissile material is in fact uranium. The 233U behaves like
    the more familiar naturally occurring 235U. It has a fairly long half life of 160,000
    years, and like 235U, 233U is fissionable and can create and sustain a nuclear fission
    chain reaction, in which the neutrons emitted by one fission event trigger further
    fission events in other 233U nuclei. When 233U undergoes fission, it produces similar
    fission products as 235U, but in different proportions.

    Thorium fuel does possess some advantages over conventional uranium / plutonium
    fuels:
     the 232Th is more likely than the 238U to capture thermal neutrons;
     the resulting 233U is more likely to fission following neutron capture than is
    239Pu;
     fissioning 233U produces more neutrons to sustain the nuclear chain reaction.
    These factors combine to create a more efficient ‘neutron economy’ for thorium than
    for conventional nuclear fuels, making smaller reactors more viable. They also
    mitigate against the formation of long-lived transuranic isotopes such as plutonium.
    There is also one important disadvantage: the breeding of 232U, a non-fissile but
    strongly radioactive uranium isotope. This arises when the 233Pa absorbs a neutron
    before it decays to 233U. The resulting 234Pa may then expel a pair of neutrons to
    make 232Pa, which then undergoes beta decay to 232U.

    This isotope is typically present in small quantities with a 232U:233U ratio of well
    under 1%. But it presents a considerable hazard due to its short half life of under 70
    years and the rapid decay chain which follows, culminating in an ultra-hard 2.6 MeV
    gamma ray – capable of passing through a metre of lead. This powerful gamma
    irradiation creates a hazard to personnel and to unshielded electronic control systems.
    Consequently, thorium fuel requires far more shielding, and more stringent remote
    handling techniques than conventional nuclear fuels.

    But the greatest problem with the thorium fuel cycle is our relative inexperience of it,
    compared to the conventional uranium / plutonium fuel cycle. According to the UK’s
    National Nuclear Laboratory, “Thorium reprocessing and waste management are
    poorly understood. The thorium fuel cycle cannot be considered to be mature in any
    area.” The NNL estimates that 10-15 years of research and development will be
    required before thorium fuels are ready for production deployment in conventional
    reactors:
    “Starting from fabrication of a commercially-relevant mass of ThO2 fuel,
    which might take 1 or 2 years, the subsequent irradiation to full burnup would
    likely take 4 to 5 years. Subsequent post-irradiation examination might take another 1 to 2 years, so the overall timescale will be of the order of ~10 years. In practice, a gradual ramp-up to commercial scale loading might be necessary, leading to a more realistic timescale of about 15 years for commercial demonstration. This is comparable to the timescale that was
    required to commercialise MOX in LWRs.”

    Appendix 2 – Molten salt reactors
    A2.1 History of the molten salt reactor

    MSRs were first developed in the early 1950s as the US Airforce sought a novel
    power source for its fleet of high altitude nuclear bombers. Although a working
    reactor was developed under the ARE (Airborne Reactor Experiment) programme, it
    was never deployed.

    In the 1960s, the ARE technology was taken up by the Oak Ridge National
    Laboratory which conducted its own MSRE (molten salt reactor experiment) from
    1965 to 1969. This was based on a graphite-moderated reactor using fluoride salts of
    uranium and plutonium as fuel. Subsequently Oak Ridge built a Molten Salt Breeder
    Reactor (MSBR), which operated from 1970 to 1976. In its initial phase the MSBR
    used as fuel fluoride salts of 235U and thorium, later followed by using the 233U it
    had bred in the first phase, also with thorium. It was also tested using plutonium fuel
    (mostly 239Pu).

    The experiment demonstrated that the reactor design was viable. Particular successes
    included the breeding of 233U from the initial thorium; the subsequent use of the
    233U to re-initiate the a thorium fuel cycle; and the de-gasification of the molten salt
    fuel, extracting unwanted gases such as xenon (135Xe), an important neutron sink that
    would otherwise slow down or indeed halt the fission chain reaction.

    The MSBR also highlighted some unexpected hazards. For example, the nickelmolybdenum
    alloy used to build the reactor became brittle under thermal neutron irradiation, and suffered extensive surface cracking due to the presence of the fission product tellurium. This highlights the importance of developing materials capable of surviving the highly corrosive environment of an LFTR environment, and to withstand the intense neutron bombardment, over a multi-decadal timescale. It also raises the prospect that any material used may degrade well before its anticipated end of life and cause premature reactor closedown.

    It should also be pointed out that the power output of the MSBR was limited to just 7
    MW. Any production MSR built for power generation would be expected to have a
    thermal output closer to 500 MW, two orders of magnitude greater, with correspondingly greater fluxes of neutrons. This would create of host of challenges in engineering design, materials science and fuel reprocessing.

    A2.2 Molten salt processing
    A key benefit of MSRs is that they provide the ability to clean the fuel of unwanted
    fission products on a continuous basis. In conventional solid fuel reactors, fission
    products build up in the fuel rods or pellets, and some of these are powerful neutron
    absorbers, like the Xenon isotope 135Xe. It is the accumulation of these neutronabsorbing
    fission products that ultimately limits the lifetime of solid fuels by reducing the efficiency of the reactor’s ‘neutron economy’ until the nuclear fission chain reaction slows down or halts.

    In the course of molten salt reactor operation, other undesirable fission products also
    build up in the fuel. These include oxygen, which gives rise to particulate deposits of
    solid metal oxides, and highly corrosive sulphur and metals. These also require
    removal.

    Techniques for extracting these various contaminants from the molten salt fuel were
    originally developed at Oak Ridge. Typical processing temperatures are in the region
    of 400C to 600C and involve the use of highly reactive chemicals such as hydrogen,
    hydrogen fluoride and hydrofluoric acid. This creates a highly hazardous
    environment.

    Further ‘pyro-processing’ techniques that are highly applicable to MSRs were
    developed at the Argonne National Laboratory in the context of its Integral Fast
    Reactor (IFR) programme. These involve high temperature ‘pyrometallurgy’ and
    electro-refining.

    Note that these technologies could be used to produce very high purity streams of
    fissile uranium and plutonium well above weapons grade as currently defined (see
    A3.2 and A3.3 below).

    The continuous purification of the molten salt inevitably creates a waste stream of
    fission products in various combinations and in mixtures of reagents and waste
    chemicals arising from the process. The safe handling and disposal of these wastes,
    while minimising radioactive releases to the environment, presents further serious
    challenges in radio-chemical engineering.

    A2.3 Safety in operation
    Proponents of the LFTR claim important safety advantages for the technology:

     LFTRs are unable to suffer reactor core meltdown, as occurred at Chernobyl
    and Fukushima. If the core temperature rises too high, the liquid fuel expands
    and the fission chain reaction slows down. Also the removal of the ‘neutron sink’
    isotope 135Xe can be halted so as to slow down the fission chain reaction. As a
    fail-safe, a salt plug is included in the bottom of the reactor which will melt at a
    set temperature and allow the fuel to drain into a holding tank where fission will
    halt.
     the LFTR operates at near-atmospheric temperature is therefore less susceptible
    to explosive pressure release venting fuel and fission products to secondary
    containment or the atmosphere.
    These claims are broadly accurate. But while LFTRs do indeed reduce certain risks,
    other new risks appear. There is the risk of materials failure in the reactor / liquid fuel
    containment as the alloys become brittle under neutron irradiation, or suffer cracking
    and surface damage in the high-temperature, intensely corrosive reactor environment.
    Any accidental release of the hot fluoride salt fuel could be highly damaging owing to
    the fuel’s highly corrosive chemistry, and cause radioactive releases to the
    environment. Similar accidents could also take place in the continuous fuel
    reprocessing system, which will use highly reactive and potentially explosive
    chemicals such as hydrogen, fluorine and hydrofluoric acid, all at very high
    temperatures.

    Due to the intensely radioactive nature of some the isotopes that need to be handled
    during reprocessing, there is no scope for direct human intervention in the fuel
    reprocessing system or the reactor itself in the case of failure. All personnel will need
    to be shielded and will only be able to intervene via remote handling systems or
    robots, themselves subject to potential failure. In the event of an accident, there might
    be little alternative but to abandon the reactor, possibly for an extended period of
    time, until radioactivity declined to sub-lethal levels permitting human access.
    It is therefore hard to sustain with any certainty the idea that LFTRs are intrinsically
    safe. Indeed considerable dangers appear to be attached to LFTRs and their routine
    operation.

    Appendix 3 – Nuclear weapons proliferation

    A3.1 General considerations
    As already noted, thorium reactors work by breeding 233U, a fissile isotope of
    uranium. It has been stated that thorium reactors present no nuclear weapons
    proliferation hazard because they do not breed plutonium like conventional uranium
    reactors. However, there are a number of stages of the thorium fuel cycle in which
    fissile material for weapons could be diverted.

    First, the thorium fuel cycle needs to be initiated by externally supplied fissile
    material, whether uranium (235U or 233U) or plutonium (239Pu). Accordingly there
    is the risk that some of this externally-supplied fissile material could be diverted into
    weapons.

    Second, the 233U that is bred in thorium reactors is highly weaponisable. Such a
    bomb was exploded in the Nevada desert in 1955 as part of Operation Teapot, a series
    of 14 nuclear bomb tests conducted by the US government. In one of these tests, the
    Military Equipment Test or ‘MET shot’, engineers replaced the U235 core of the
    uranium / plutonium bomb with 233U. The bomb successfully detonated and the
    principle that 233U can be used to make nuclear bombs, with fearsome destructive
    potential, was firmly established.

    It has been suggested that the inevitable presence of 232U as a contaminant of 233U –
    as we have noted, a powerful gamma emitter though its decay products – renders
    U233 unusable as a bomb-making material, due to health damage to handlers and
    machinists, and disruption to electronics. Given that a 233U bomb has already been
    assembled and detonated using 1950’s technology, this is clearly not an insuperable
    problem. Pyroprocessing technologies could also be used to produce highly
    concentrated 233U untainted by 232U or any other uranium isotope from thorium fuel
    (see A3.2 below).

    One means that has been proposed to prevent the generation of sufficiently pure 233U
    to build a nuclear bomb is to add to the thorium fuel a significant percentage of
    natural or depleted uranium, rich in non-fissile 238U. It is impossible to chemically
    separate 233U from 238U since they are both the same element.
    However as the NNL notes, “Attempts to lower the fissile content of uranium by
    adding U-238 are considered to offer only weak protection, as the U-233 could be
    separated in a centrifuge cascade in the same way that U-235 is separated from U-
    238 in the standard uranium fuel cycle.” Indeed, owing to the greater difference in
    atomic mass the centrifuge separation would operate more efficiently.
    The presence of 238U in the fuel would also create another hazard. One of the
    advantages of using thorium fuel is its low level of conversion to long-lived
    transuranics. But if the fuel contains 238U as an anti-proliferation measure, then it
    will absorb neutrons (in the process degrading the ‘neutron economy’) and form
    plutonium. Next, ‘pyroprocessing’ technologies could be used to extract any plutonium
    from the molten salt fuel, potentially producing 239Pu at very high concentrations
    well above ordinary weapons-grade (see A3.2, below).

    A3.2 Weapons grade uranium (233U)

    The process whereby fissile 233U is bred in the thorium fuel cycle involves an
    intermediary stage, with the protactinium isotope 233Pa. The 233Pa undergoes beta
    decay to 233U with a half life of 27 days. But the 233Pa may first absorb a second
    thermal neutron to make 234Pa. This isotope may either decay to the undesirable
    uranium isotope 232U, or the non-fissile uranium isotope 234U.

    This makes it beneficial to remove the protactinium from the molten salt fuel before it
    can intercept a second neutron. Left to itself, away from the neutron flux, the 233Pa
    decays over a period of months to produce very pure 233U. Furthermore the removal
    of the protactinium leaves more neutrons to maintain the reactor’s neutron economy,
    maintaining both the thorium fuel cycle and the fission chain reaction.

    Oak Ridge demonstrated several effective methods of protactinium removal, for
    example, precipitation by addition of thorium oxide to the molten salt [Removal of
    protactinium from molten fluoride breeder blanket mixtures, C. J. Barton and H. H.
    Stone, 1966]. This technique precipitated 233Pa when present at just 0.1 parts per
    billion.

    But the authors note that the protactinium is highly radioactive and hazardous to
    handle even on the milligram scale, due to its short half life of 27 days, and its high
    combined beta and gamma energy of 570 KeV. For this reason, Oak Ridge used a
    mixture of 233Pa and much less radioactive 231Pa for its experiments, in which the
    231Pa was ~100,000 times more abundant than the 233Pa. This indicates that the
    handling of the highly radioactive 233Pa extracted from molten salt fuel would need
    to be done entirely by remote handling with any personnel shielded from radiation.
    This method of producing very pure fissile 233U, while highly desirable as regards
    reactor operation, represents a significant weapons proliferation hazard. Purities of
    233U well above accepted weapons grade (85% for 235U) would be achievable.

    A3.3 Weapons grade plutonium (239Pu)

    A similar approach could also be used to produce weapons-grade plutonium from
    molten salt fuel rich in 238U. In the normal operation of a uranium solid fuel reactor
    the 238U captures a neutron and then undergoes two beta decays to form 239Pu.
    However, the 239Pu itself has a high probability of capturing further neutrons to make
    240Pu. Hence the plutonium in spent fuel typically comprises under 80% 239Pu and
    most of the remainder is 240Pu with some 241Pu.

    240Pu is highly undesirable in a plutonium weapon since it can trigger premature
    fission giving rise to a low-yield explosion known as a ‘fizzle’. As a result weapons
    grade plutonium contains a maximum of 7% 240Pu. So in order to make weapons
    grade plutonium the fuel is only left for a short time in the reactor before
    reprocessing.

    The same result could be served in a molten salt reactor by including a high level of
    238U in the fuel, and extracting the 239Pu as it is formed. It would make sense to use
    the pyro-processing technology developed by the Argonne National Laboratory for its
    Integrated Fast Reactor – a high temperature (500C) electrolysis process using molten
    salts. This could be employed to concentrate the plutonium, together with some
    residual uranium and any minor actinides from the molten salt fuel [ProliferationProof
    Uranium/ Plutonium Fuel Cycles, By G. Kessler]. Standard aqueous methods
    such as PUREX could then be employed to purify the plutonium. This approach
    should be capable of producing plutonium with a very high 239Pu content well above
    normal weapons-grade.
    END

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